ML20049H285
| ML20049H285 | |
| Person / Time | |
|---|---|
| Site: | 05000447 |
| Issue date: | 02/12/1982 |
| From: | GENERAL ELECTRIC CO. |
| To: | |
| References | |
| NUDOCS 8202230063 | |
| Download: ML20049H285 (15) | |
Text
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I i GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O ;
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i CilAPTER 10 l
! STEAM AND POWER CONVERSION SYSTEM 1
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- GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0
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SECTION 10.1 1
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i CONTENTS ,
i Section Title Page i
10.1
SUMMARY
DESCRIPTION 10.1-1 1
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GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O i
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i 10.1
SUMMARY
DESCRIPTION i
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Applicant will supply.
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GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O SECTION 10.2 CONTENTS Section Title Page
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10.2 TURBINE GENERATOR 10.2-1
- 10.2.1 Design Bases - Functional Limi-tations by Design or Operational Characteristics of the Reactor Coolant System 10.2-1 10.2.1.1 Turbine Stop Valve Fast Closure 10.2-1 10.2.1.2 Turbine Control Valve Closure 10.2-1
} 10.2.1.2.1 Turbine Control Valve Fast i Closure 10.2-1 10.2.1.2.2 Turbine Control Valve Normal Closure 10.2-1
) 10.2.1.3 Automatic Load Maneuvering j Capability 10.2-2 10.2.2 Description 10.2-2 10.2.3 Turbine Disk Integrity 10.2-2 10.2.4 Evaluation 10.2.2 i
ILLUSTRATIONS i
Figure Title Page
- 10.2-1 Turbine Stop Valve Closure Characteristic 10.2-3 10.2-2 Turbine Stop Valve Fast Closure Characteristic 10.2-4 r
l 10.2-3 Acceptable Range for Control Valve Normal I Closure Motion 10.2-5 1
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GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 10.2 TURBINE GENERATOR Applicant will supply.
10.2.1 Desigp Bases - Functional Limitations by Design or Operational Characteristics of the Reactor Coolant System The turbine control valves must be capable of full-stroke openings and closings at no greater than 7 seconds for adequate pressure control performance.
10.2.1.1 Turbine Stop Valve Fast Closure During any event resulting in turbine stop valve fast closure, turbine inlet steam flou must not be reduced faster than indicated in Figure 10.2-1.
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10.2.1.2.2 Turbine Control Valve Normal Closure l
I l The turbine control valve steamflow shutoff, upon step reduction I
to zero in pressure regulation demands (no resulting bypass steam flow demand), must not be reduced slower than indicated in Figure 10.2-3. Any single control system failure or turbine-generator event must not cause a faster steam flow reduction l than indicated in Figure 10.2-3 without initiating an immediate reactor trip.
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10.2-1
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 10.2.1.3 Automatic Load Maneuvering Capability Within the automatic load following region of the power / flow operating map, steamflow will automatically respond to a load demand step as follows:
(1) 0.9 x in 10 sec for lxl< 10% (except as stated in (3))
(2) 0.9 x in x see for lxl> 10% (except as stated in (3))
wnere x is in percent of power at rated core flow on the rod line in which the transient takes place.
(3) If the step demand is positive and greater tnan 10%
Nuclear Boiler Reactor (NBR) , then the response to the first 10% NBR will be as stated in (2) with tne balance of the demand response being accomplisned by a ramp at a rate of 15% NBh/ min.
10.2.2 Description Applicant will supply.
10.2.3 Turbine Disk Integrity Applicant will supply.
10.2.4 Evaluation Applicant will supply.
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GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 O\ 100 m - / -
ACCEPTABLE REGION FOR TUR0lNE STOP VALVE CLOSURE RESPONSE E
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0 0.1 0.2 l 0 TIME AFTER START OF STOP VALVE CLOSURE MOTION (sec)
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I \ _ ,1 Figure 10.2-1. Turbine Stop Valve Closure Cnaracteristic 10.2-3
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0
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VALVE FAST CLOSURE RESPONSE O
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AS THE INI I PE CEN LO MULT L ED BY O 8s c 0 T TIME AFTER START OF CONTROL VALVE F AST CLOSURE MOTION (sec)
Figure 10.2-2. Turbine Control Valve Fast Closure Characteristic 10.2-4
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 l I
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P = INITIAL STEAMF LOW, PERCENT S NUCLEAR BOILER RATED 9
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ACTUAL CONTROL VALVE FULL g 80 -
STROKE CLOSURE TIME (SLOWEST)
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.; FOR NORMAL CLOSURE T2 (Tv/100) P
$ OF TURBINE CONTROL T3 =
(Tv - 0.5)/100 P D VALVE E
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l 0 Ti T3 T2 l TIME AFTER START OF CONTROL VALVE NORMAL CLOSURE MOTION (sec) i l
m Figure 10.2-3. Acceptable Range for Control Valve Normal Closure Motion 10.2-5/10.2-6
t GESSAR II 22A7007 i 238 NUCLEAR ISLAND Rev. O I
l SECTION 10.3 i
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1 10.3 MAIN STEAM SUPPLY 10.3-1 4
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' GESSAR II 22A7007 l 238 NUCLEAR ISLA!;D Rev. 0 1
10.3 MAIN STEAM SUPPLY The main steam description, criteria, and design contained within the scope of the Nuclear Island are presented in Section 5.4. Tne Applicant shall provide the remaining material required by this section.
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l GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 SECTION 10.4 l
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CONTENTS I,
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- 10.4-1 10.4 OTHER FEATURES OF STEAM AND POWER CONVERSION SYSTEM
- 10.4.1 Main Condensers 10.4-1 !
10.4-1 l 10.4.2 Main Condenser Evacuation System 10.4.3 Turbino Gland Sealing System 10.4-1 f 10.4.4 Turbine Byr,3s System 10.4-1 !
i 10.4.4.1 Design Bis. 10.4-1 f 10.4.4.2 System De 7ription 10.4-1 10.4.4.2.1 Operationa. Function 10.4-1
- 10.4.4.2.2 Bypass Valves 10.4-2 j 4 10.4.4.2.3 Classification 10.4-2
!, 10.4.4.3 System Evaluation 10.4-2 10.4.4.4 Tests and Inspections 10.4-2
} Instrumentation Application 10.4-3 10.4.4.5 10.4.5 Circulating Water System 10.4-3 10.4.6 Condensate Cleanup System 10.4-3 l
10.4.7 Condensate and Feedwater Systems 10.4-3 ;
10.4.7.1 Design Basis 10.4-4 10.4.7.1.1 Power Generation Design Bases 10.4-4 10.4.7.2 System Description 10.4-5 10.4.7.2.1 Instrumentation 10.4-5 10.4.7.3 System Evaluation 10.4-6 10.4.8 Steam Generator Blowdown System (PWR) 10.4-7 10.4.9 Auxiliary Feedwater System (PWR) 10.4-7 10.4-i/10.4-ii
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 10.4 OTilER PEATURES OF STEAM AND POWER CONVERSION SYSTEM Applicant will supply.
10.4.1 Main Condensers Applicant will supply.
10.4.2 Main Condenser Evacuation System Applicant will supply.
10.4.3 Turbine Gland Sealing System Applicant will supply.
10.4.4 Turbine Bypass System p
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10.4.4.1 Design Basis Applicant will supply.
10.4.4.2 System Description 10.4.4.2.1 Operational Function l
The turbine bypass system controls primary steam pressure by sending excess steam flow directly to the main condenser. This permits independent control of reactor pressure and power during f
reactor vessel heatup to rated pressure as the turbine is brought up to speed and synchronized under turbine speed-load control.
Following main turbine / generator trips, the turoine bypass will 10.4-1
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 10.4.4.2.1 Operational Function (Continued) control reactor overpressure within its capacity and in accordance with the steam generation rate. The bypass valves are automatically closed whenever vacuum in the main condenser falls below a preset value.
10.4.4.2.2 Bypass Valves The turbine bypass system consists of several automatically-operated, regulating-type bypass valves connected by appropriate piping to the main steamlines upstream of the main turbine stop valves. The bypass valves are required to nave regulation capability and a fast-opening response approximately equivalent to the fast closure of the turbine stop and control valves. The bypass valve regulation is designed to prevent spurious or unnecessary opening of the bypass valves about their cracking point due to control signal noise or minor transients.
10.4.4.2.3 Classification The steam bypass system is classified as a primary power generation system; that is, it is not a safety system, and its operation is essential to the power production cycle.
10.4.4.3 System Evaluation The effects of a malfunction of the turbine-bypass system valves and the effects of such failures on other systems and components are evaluated in Chapter 15, Accident Analysis.
10.4.4.4 Tests and Inspections The opening and closing of the turbine bypass system valves will be checked during initial startup and snutdown for performance and timing. The bypass steam line upstream of the bypass valves 10.4-2
GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 l 10.4.4.4 Tests and Inspections
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will be hydrostatically tested to confirm leakage tightness.
Visual inspection of pipe weld joints will confirm the exterior condition of the weld.
10.4.4.5 Instrumentation Application Controls and valves are designed so that the bypass valves steam flow is shut off if the control system loses its electric power or hydraulic system pressure. For testing the bypass valves during operation, the strrke time of the individual valves is increased during testing to limit the rate of bypass flow increase and decrease to approximately 1%/sec of reactor rated flow.
Upon turbine trip or generator load rejection, the start of the bypass valve steam flow will not be delayed more than 0.1 sec I \ after the start of the stop valve or the control valve fast-U closure moti.on. A minimum of 80% of the rated bypass capacity will be established within 0.3 see after the start of the stop valve or the control valve closure motion. Refer to Subsection 7.1.1.7.E, Pressure Regulators and Turbine-Cenerator Controls.
10.4.5 Circulating Water System Applicant will supply.
10.4.6 Condensate Cleanuo System Applicant will supply.
10.4.7 Condansate and Feedwater Systems The feedwater lines description, criteria, and design contained
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v within the scope of the Nuclear Island are presented in Section 5.4.
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GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 10.4.7.1 Design Basis 10.4.7.1.1 Power Generation Design Bases The condensate and feedwater systems provide a dependable supply of high quality feedwater to the reactor. The cystem provides the required flow at the required pressure and temperature to the reactor allowing sufficient margin for continued flow under anticipated transient conditions.
(1) Performance Requirements During operation at rated temperature and pressure, the feedwater system supplies the reactor with feedwater from the reactor feed pumps. This system has sufficient capacity to provide 105% NBR steady-state and 115% NBR transient flow. The feedwater heaters provide the required temperature of feedwater to the reactor with six stages of closed feedwater heating. The final feedwater temperature is 420"F at rated power.
(2) Feedwater Quality Pumped-forward heater drains are sufficiently de-aerated in the shells of the pumped (third-stage) feedwater heaters to maintain a level of 200 ppb (or less) oxygen content in the final feedwater supplied to the reactor l
during normal full-load operation (measured upstream of the isolation valves).
i Other feedwater characteristics are maintained within the following limits during normal plant operation.
Conductivity 10.1 pmho/cm at 25 C (measured after the condensate treatment (system) 10.4-4
GESSAR TI 22A7007 238 NUCLEAR ISLAND Rev. 0
(g 10.4.7.1.1 Power Generation Design Bases (Continued)
V Metallic impurity <15 ppb of which no more than 2 ppb may be copper (measured before the outboard isolation valve)
Chloride (as Cl) Feedwater chloride concentrations shall be maintained so that reactor water chloride concenuration limits are maintained.
To minimize the corrosion-product input ta the reactor, a start-up recirculation line is provided from the reactor feedwater supply lines, downstream of the high-pressure feedwater heaters, to the main condenser.
(3) Design Codes O
All components of the condensate and feedwater system that contain the system pressure are designed and constructed in accordance with the applicable codes as referenced in Section 3.2.
10.4.7.2 System Description 10.4.7.2.1 Instrumentation Feedwater flow-control instrumentation measures the individual feedwater pump flow rates from the condensate and feedwater system. These measurements are used by the feedwater control system, which regulates the feedwater flow to the reactor to meet system demands. The feedwater control system is described in Subsections 7.1.1. 7. D ar i 7. 7,1. 4. Total feedwater flow is used in heat balance calculations.
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GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. 0 10.4.7.2.1 Instrumentation (Continued)
Short term isolation of the feedwater lines is accomplished by use of two check valves (one inboard and one outboard of the containment) in each of the two feedwater lines. Should a line break occur outside the containment, these check va]ves prevent significant loss of reactor coolant inventory and provide i, solation. A motor-operated gate valve which is capable of being remotely closed from the control room is provided for long term leakage protection upon operator judgment that continued makeup from the feedwater source is unnecessary.
Instrumentation and controls regulate pump recirculation flow rate for the condensate pumps, condensate booster pumps, and reactor feed pumps. Sampling means are provided for monitoring the quality of the final f eedwater as described in Subsection 9.3. 2.
Temperature measurements are provided for each stage of feedwater heating and thc3e include measurements at the inlet and outlet on both the steam and water sides of the heaters. Steam-pressure measurements are provided at each feedwater heater. Instrumenta-tion and controls are provided for regulating the heater drain flow rate to maintain the proper condensate level in each feed-water heater shell or heater drain tank. High-level alarm and automatic dump-to-condenser action on high level are provided.
10.4.7.3 System Evaluation During operation, radioactive steam and condensate are present in i the feedwater heating portion of the system which includes the l extraction steam piping, feedwater heater shells, heater drain piping, and heater vent piping. Shielding and controlled access are provided as necessary (Section 12. 3 ) . (Applicant will supply.) The condensate and feedwater system is designed to 1
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GESSAR II 22A7007 238 NUCLEAR ISLAND Rev. O I \ 10.4.7.3 System Evaluation (Continued) b minimize leakage with welded construction used where practical.
Relief discharges and operating vents are handled through closed-systems.
The condensate and feedwater system is not required to effect or support the safe shutdown of the reactor or to per 7orm in the operation of reactor safety features. If it is necessary to remove a component such as a feedwater heater, pump, or control valve from service, continued operation of the system is possible by use of ti;e multistream arrangement and the provisions for isolating and bypassing equipment and sections of the system.
The analysis of both the condensate and feedwater individual component failures is bounded by the feedwater component system failure analysis. This analysis is covered in Chapter 15 under
() the following subsections: 15.1.1, Loss of Feedwater Heater; 15.3.2, Feedwater Controller Failure; and 15.2.7, Loss of Feedwater Flow. The effects of equipment malfunction on the reactor coolant system are provided in Chapter 15, Nuclear Safety Operational Analysis. Also included in Subsection 15.6.6, Feedwater Line Break, are the isolation provisions that preclude release of excess radioactivity to the environment.
10.4.8 Steam Generator Blowdown System (PWR)
Not applicable to BWR, 10.4.9 Auxiliary Feedwater System (PWR)
Not applicable to BWR.
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