ML20063F253
| ML20063F253 | |
| Person / Time | |
|---|---|
| Site: | 05200001 |
| Issue date: | 02/07/1994 |
| From: | Fox J GENERAL ELECTRIC CO. |
| To: | Poslusny C Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 9402140207 | |
| Download: ML20063F253 (33) | |
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ra A GE Nuclear Energy c vir c coww "r' MtX Amm e .%r > ,tw CA !)5 Un February 7, 1994 Docket No STN 52-001 Chet Posiusny, Senior Project Manager Standardization Project Directorate Associate Directorate for Advanced Reactors and License Renewal Office of Nuclear Reactor Regulation
Subject:
Submittal Supporting Accelerated ABWR Schedule - Response to Open Item F1,9 1
Dear Chet:
Enclosed are SSAR markups addressing the subject open item pertaining to venting procedures, testing of RCIC bypass, turbine trip reliability, materials selection, and increased capability for the ACIWA system. Each of these are separately addressed below:
Operating Procedu res for Venting Added Subsection l A.3.6, Procedure for Reactor Venting, in Amendment 33.
Iestinc of RCIC Bynan See modified Subsection I A.2.23 and new Subsection 1 A.3.8.
Enhanced Reliability for Turbine Trin Locie The turbine trip logic is within the scope of the ABWR Standard Plant and, therefore, is not a candidate for a COL action item. Altemately, we have expanded the treattaent of the information in Chapter 10 (enclosed) pertaining to turbine trip reliability.
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Increased - Canability for the ACIWA System See new Subsection 19.9.21 and revised Subsection 191.3.1.
8' lease provide a copy of this transmittal to Jerry Wilson.
Si n c e r 'ly, J x Advanced Reactor Programs ec: Alan Beard (GE) .
Norman Fletcher (DOE)
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injection valves will automatically close in order to prevent water from entering the main steamlines.
In the unlikely event that a subsequent low level recurs, the RCIC steam supply and injection valves uill automatically reopen to allow continued flooding of the vessel. The '
HPCF injection valves will also automatically reopen unless the operator previously closed them manually. Refer to Subsections 7.3.1-1-1.1 (HPCF) and 7.3.1.1.1.3 (RCIC) for additional details regarding system initiation and isolation logic.
1 A.2.23 Modify Break-Detection Logic to Prevent Spurious isolation of HPCI and RCIC Systems [lI.K.3(15)]
NRC Position The High-Pressure Coolant Injection (HPCI)and Reactor Core Isolation Cooling (RCIC) Systems use differential pressure sensors on elbow taps in the steamlines to their turbine drives to detect and isolate pipe breaks in the systems. The pipe-break-detection circuitry has resulted in spurious isolation of the HPCI and RCIC Systems due to the pressure spike which accompanies startup of the systems. The pipe-break-detection circuitry should be modified to that pressure spikes resulting from HPCI and RCIC System initiation will not cause inadvertent system isolation. l Response y The ABWR design utilizes the motor-driven h. OF System /rather than the turbine-driven HPCI System for high pressure inventory maintenance. Therefore, this position is only applicable to the turbine-driven RCIC System.
In the ABWR Standard Plant design, the high differential pressure signals which isolate the RCIC turbine are processed through the Leak Detection and Isolation System (LDS). Spurious trips are avoided because the RCIC has a bypass start system controlled by valves F037 and F045 (Figure 5.4-8, RCIC P&ID bO n receipt of RCIC start signals, bypass valve F045 opens to pressurize the line l downstream and accelerate the turbine. The bypass line via F045 is small (25A) and naturally limits the initial flow surge such that a differential pressure spike in the upstream pipe will not occu C
After a predetermined delay (approximately 5-10 seconds), steam supply valve F037 ,
opens to admit full steam flow to the turbine. At tMs stage, the line downstream is already pressurized. Thus, it is highly unlikely inat a differential pressure spike could '
occur dtuing any phase of the normal startup process. Su .S M e. ekt on i A. S . 8 -h c><
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1 A.3.2 noview and Modify Procedures for Removing Safety-Related Systems from Service
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Procedures shall be resiewed and modified (as required) for removing safety-related systems from senice (and restoring to senice) to assure operability status is known j (Subsections I A.2.18 and 1 A.2.19).
1A.3.3 In-Plant Radiation Monitoring Equipment and training procedures shall be provided for accurately determining the . !
airborne iodine concentration in areas within the facility where plant personnel may be .
present during the accident (Subsection I A.2.35). j ;
1A.3.4 Reporting Failures of Reactor System Relief Valves i Failures of reactorsystem reliefvalves shall be reported in the annual report to the NRC I (Subsection I A.2.21.1). j 1A.3.5 Report on ECCS Outages I Starting from the date of commercial operations, an annual report should be submitted which includes instance of ECCS unavailability because of component failure, - !
maintenance outage (both forced or planned), or testing, the following information j shall be collected: ,
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(1) Outage date T-e.3-t l
(2) Duration outage "Tkg cot, appb wM sk all p 4M p s .1 54ewh s y how 4- k R.c.T c-(3) - C. ..e of outage 4tst d4 Tea cl m - Sub se eMow t a . 2. 23 dw + a; 7 6% - b@ - ;
Emergency core cooling system or component involved (5) Corrective action taken ;
The above information shall be assembled into a report, which will also include a ;
discussion of any changes, proposed or implemented, deemed appropriate, to improve the availability of the emergency core cooling equipment (Subsection _1A.2.2.5). ,
1 A.3.6 Procedure for Reactor Venting ' l Procedurefshall be developed for the operators use of the reactor vents. (See Subse j IA.2.5) 1A.3.7 Testing of SRV and Discharge Piping The COL applicant will confirm that any'SRVs or discharge pipng installed that is not !
similar to those that have been tested will be tested in accordance with Subsection lA2.9. I
' Response to TMI Related Matters - Amendment 33 1A-35
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10.2 Turbine Generator 10.2.1 Design Bases 10.2.1.1 Safety Design Bases The turbine generator (T-G) does not serve nor support any safety function and has no safety design basis. The turbine generator is, however, a potential source of high ener gy missiles that could damage safety-related equipment or structures. The tu.bine is designed to minimize the possibility of failure of a '.urbine blade or rotor. Turbine integrity is discussed in Subsection 10.2.3. The effects of potential high energy missiles are discussed in Chapter 3. In addition, the main steam turbine stop valves are analyzed to demonstrate structural integrity under safe shutdown earthquake (SSE) loading conditions.
102.1.2 Power Generation Design Bases Power Generation Design Basis One-The T-G is intended for base load operadon. The gross generator outputs at reference guaranteed reactor rating and valves-wide-open (VWO) operation are given on the heat balances shown on Figures 10.1-2 and 10.1-3, respectively.
(" t Power Generation Design Basis Two-The T-G load change characteristics are compatible with the instrumentation and control system which coordinates T-G and rcactor operation.
Power Generation Design Basis Three-The T-G is designed to accept a sudden loss of fullload without exceeding design overspeed.
Power Generation Design Basis Four-The T-G is designed to permit periodic under load testing of steam valves important to overspeed protection, emergency overspeed trip circuits, and several other trip circuits.
i Power Generation Design Basis Five-The failure of any single component will not cause the rotor speed to exceed the design speed.
A I N 9E D-T A 4 10.2.1.3 Functional Limitations imposed by the Design or Operational Characteristics of l the Reactor Coolant System 10.2.1.3.1 Turbine Stop Valve During an event resulting in turbine stop valve fast closure, turbine inlet steam flow will not be reduced faster than that shown in Figure 10.2-1.
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5 INSERT A Power Generation Design Basis Six - The T-G is designed to support the plant availability goals by utilizing 2/3 or 2/4 coincident trip logic for all but the vibration trips (which are at least 2/2 per bearing). Similarly, all turbine control functions which are required. !
for power . generation will use at least dual redundant controllers and triply redundant control inputs.
Power Generation Design Basis Seven - The T-G auxiliary systems (stator cooling, lube . oil cooling, etc.) are designed either with enough redundancy to support full power operation with a single failure or to provide a signal to the reactor power control system to automatically reduce power to within the capability of the remaining on-line capacity.
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10.2.2.4 Turbine Overspeed Protection System In addition to the normal speed control function provided by the turbine control system, a separate turbine overspeed protection system is included. The turbine overspeed system is a highly reliable and redundant system which is classified as non-safety-related. dlaasi Protection against turbine overspeed is provided by the mechanica overspeed trip and ,
electrical backup overspeed trip. Redundancy is achieved by usin two independent channels from the signal source to the output device. The sensing device, line and output device are of a different nature for each individual channel in order to increase reliability.
The overspeed sensing devices are located in the front bearing standard and, therefore, are protected from the effects of missiles or pipe break. The hydraulic lines are fail-safe; that is, if one were to be broken, loss of hydraulic pressure would result in a turbine trip.
The electric trip signals are redundant. One circuit could be disabled by damage to the wiring, but the other system is failsafe (i.e., loss ofsignal results in a turbine trip). These features provide inherent protection against failure of the overspeed system caused by missiles or pipe whipping.
The electrical backup overspeed trip consists ofindependent circuits. Each circuit monitors a separate speed signal voltage and activates voltage comparators at various speed levels. The output of these circuits is used in tripping and monitoring of the turbine.
Two air relay dump valves are provided which actuate on turbine trip. The valves control air to the extraction non-return valves, which limit contributions to turbine overspeed from steam and water in the extraction lines and feedwater heaters. The closing time of the extraction non-return valves is less than 2 seconds.
Upon loss of genemtor load, the EHC System acts to prevent rotor speed from exceeding design overspeed. Refer to Table 10.2-1 for a description of the sequence of events following loss of turbine load. Failure of any single component will not result in Turbine Generator- Amendment 33 10.2-7
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rotor speed exceeding design overspeed (i.e.,120% of rated speed). The following component redundancies are employed to guard against overspeed: a
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(1) Main stop valves / Control valves !
s (2) Intennediate stop valves / Intercept valves (CIVs) .
(3) Primary speed control / Backup speed control (4) Fast acting solenoid valves / Emergency trip fluid system (ETS) ,
(5) Speed control /Overspeed trip / Backup overspeed trip :
The main stop valves and control valves provide full redundancy in that these valves are in series and have completely independent operating controls and operating . .
mechanisms. Closure of either all four stop valves or all four control valves shuts off all main steam flow to the HP turbine. The combined intennediate stop and intercept valves are also in series and have completely independent operating controls and :;
operating mechanisms. Closure of either valve or both valves in each of the six sets of combined intermediate stop and intercept valves shuts off all MSR outle,t steam flow to the three LP turbines. ol-leEsY An- -
The speed control unit utilizes wo speed signals. ncrease[f [. eed ends to close the confrol valves. In t... 3.. ! ..".! :m#-- -M in it s.i . e d agm.h Loss ofMspeed signals will initiate a turbine trip via the Emergency Trip System (ETS). ,
Fast acting solenoid valves initiate fast closure of control valves under load ' rejection condP ans that might lead to rapid rotor acceleration. The ETS initiates fast closure of ;
the valves whether the fast-acting solenoid valves work or not.
If speed control should fail, the overspeed trip devices must close~the steam admission valves to prevent turbine overspeed. The mechanical overspeed trip mechanism operates at 110% of rated speed. The electrical backup overspeed trip is set at 111% of - .
rated speed. Component redundancy and fail-safe design of the ETS hydraulic system and trip circuitry provide turbine overspeed protection. Three speed signals .
independent of the speed control unit provide input to the backup, ovenpeed trip. For reliability, two-out-of-three logic is employed in both mechanical and electrical overspeed trip circuitry. Single component failure does not compromise trip ..
protection. Two separate electrical buses supply power to the trip circuits. The primary power source is shaft moun ted. The station house power and battery systems are backup power supplies. Loss of power trips the turbine through fail-safe circuitry. j
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(3) Exhaust hood temperature and spmy pressure (4) Oil system pressures, levels, and temperatures (5) Bearing metal and oil drain temperatures ;
(6) Shell temperature (7) Valve positions (8) Shell and rotor differential expansion (9) Shaft speed, electrical load, and control valve inlet pressure indication (10) Hydogen temperature, pressure, and purity (11) Stator coolant temperature and conductivity (12) Stator-winding temperature (13) Exciter air temperatures (14) Turbine gland sealing pressure (15) Gland steam condenser vacuum (16) Steam chest pressure (17) Seal oil pressure 10.2.2.7 Testing The electrical and mechanical overspeed trip devices can be tested remotely at rated speed, under load, by means oflighted pushbuttons on the EHC test panel. Operation of the overspeed protection devices under controlled, overspeed condition is checked at startup and after each refueling or major maintenance outage.
Provisions for testing each of ^ - following devices while the unit is operating are included.:
(1) Main stop and control valves (2) Turbine bypass valves (3) Low pressure turbine combined intermediate valves (CIVs)
(4) Overspeed governor 10.2 10 Turbine Generator- Amendment 33
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Inspection of all valves of one type will be conducted if any unusual condition is discovered. !
10.2.4 Evaluation The turbine-generator is not nuclear safety-related and is not needed to effect or support a safe shutdown of the reactor.
The turbine is designed, constructed, and inspected to minimize the possibility of any -
major component failure.
Tlic turbine has a redundant, testable overspeed trip system to minimize the possibility of a turbine overspeed event.
Unrestrained stored energy in the extraction steam system has been reduced to an acceptable minimum by the addition of nonreturn valves in selected extraction lines.
The T-G equipment shielding requirements and the methods of access control for all areas of the Turbine Building ensure that the dose criteria specified in 10CFR20 for operating personnel are not exceeded.
All areas in proximity to T-G equipment are zoned according to expected occupancy times and radiation levels anticipated under normal operating conditions.
Specification of the various radiation zones in accordance with expected occupancy is listed in Chapter 12.
If deemed necessary during unusual occurrences, the occupancy times for certain areas will be reduced by administrative controls enacted by health physics personnel.
The design basis operating concentrations of N-16 in the turbin'e cycle are indicated in Section 12.2.
The connection between the low-pressure turbine exhaust hood and the condenser is made by mean of a stainless steel expansionjoint.
Since there is no nuclear safety-related mechanical equipment in the turbine area and since the condenser is at subatmospheric pressure during all modes of turbine operation, failure of thejoint will have no adverse effects on nuclear saf:ty related equipment.
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. ABWR. standantsatory Anstrsis Report liotwelllevel controls provide automatic makeu) or rejection of condensate to maintain a normal level in the condenser hotwells. On low hotwell water level, the makeup control valves open and admit condensate to the hotwell from the condensate storage tank. When the hotwell is brought to within normal operating range, the valves close. On high water level in the hotwell, the condensate reject control valve opens to divert condensate from the condensat: pump discharge (downstream of the polishers and auxiliaq condeniers) to the condensate storage tank; rejection is stopped when the hopell N d{egfgt,opibin *g *gk f[y cowAnwskpogogra,tjngraggjoh*
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During the initial cooling period after plant shutdown, the main condenser removes residual heat from the reactor coolant system via the turbine bypass system. However,if the condenser is not available to receive steam via the turbine bypass system, the reactor coolant system can still be safely cooled down using only Nuclear Island systems.
10.4.1.3 Evaluation During operation, radioactive steam, gases, and condensate are present in the shells of ,
the main condenser. The anticipated inventory of radioactive contaminants during operation and shutdown is discussed in Sectiors 11.1 and 11.3.
'.... i Necessag shielding and controlled access for the main condenser are provided ;
(Sections 12.1 and 12.3).
Hydrogen buildup during operation is not expected to occur due to provisions for continuous evacuation of the main condenser. During shutdown, significant hydrogen buildup in the main condenser will not occur, as the main condenser will then be isolated from potential sources of hydrogen. !
Main condenser tubeside circulating water is treated to limit algae growth and me, ni l long-term corrosion of the tubes and other components. Corrosion of the outsic e of the j condenser tubing is prevented by maintaining strict water quality using ee condensate l cleanup system described in Subsection 10.4.6.The construction materials used for the ' '
main condenser are selected such that the potential for corrosion by galvanic and other j effects is minimized. !
The potential flooding which would result from failure of the condenser is discussed in -l Section 3.4, which shows that failure of the condenser will not adversely affect any equipment required for safe shutdown of the reactor.
The loss of main condenser vacur m will cause a turbine trip and closure of the main steam isolation valves. The conseg 2ences of a turbine trip are discussed in Subsection ;
15.2.3. Should the turbine stop, control or bypass valves fail to close on loss of condenser vacuum, two rupture diaphragms on each turbine exhaust hood protect the condenser and turbine exhaust hoods against overpressure.
10.4-4 Other Features of Steam end Poner Conversion System - Amendment 31
23A6100 Rev. 9
. ABWR. StsndardSafatyAnslysisReport 10.4.1.4 Tests and Inspections Each condenser shell is to receive a field hydrostatic test before initial operation. This test will consist of filling the condenser shell with water and, at the resulting static head, inspecting all tubejoints, accessible welds, and surfaces for visible leakage and/or excessive deflection. Each condenser water box is to receive a field hydrostatic test with alljoints and external surfaces inspected for leakage.
10.4.1.5 Instrumentation Applications 10.4.1.5.1 Hotwell Water Level j gc , A 4 m The condenser hotwell water level is measured by[two level transmitters. These transmitters provide signals to an indicator, annunciator trip units, the plant computer, and the hotwell level control system. Level is controlled by two sets of modulating control valves. Each set consists of a normal and an emergency valve.
One set ofvalves allows water to flow from the condensate storage tank to the condenser hotwell as the level drops below the setpoint. If the levelincreases above another
,etpoint, the second set of valves located on the discharge of the condensate pumps opens to allow condensate to be pumped back to the storage tank. -
10.4.1.5.2 Pressure Condenser pressure is measured by gauges, pressure switches, and electronic pressure transducers. The pressure switches provide input signals to the turbine control s} stem ,
'i and the annunciator. Two pressure transducers provide input signals to the plant computer, a recorder, and a trip unit. The trip unit provides input signals to the Reactor Recirculation System and Steam Bypass and Pressure Control System. In addition, four independent and redundant safety-related pressure transmitters provide input signals tc the Nuclear Steam Supply System.
As condenser pressure increases above normal levels, an annunciator is activated. A further increase in pressure results in a turbine trip. As prc.sure increases toward a complete loss of vacuum, the main steam isolation valves and the turbine bypass nlves are closed to prevent overpressurization of the condenser shell.
The approximate serpoints for these functions are as follows:
P (1) High condenser pressure turbine alarms at 60.96 cm Hg vacuum.
(2) High condenser pressure turbine trips at 55.88 cm Hg vacuum. 1 (3) Bypass valve closes at 30.48 cm Hg vacuum.
(4) Main steam isolation valve closes at 17.8 to 25.4 cm Hg vacuum.
Other Feorures of Steam end Power Conversion System - Amendment 31 10.4-5
23A6100 Rsv. 3 ABWR. StandardSafetyAnalysis Report t
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g 10.4.6 Condensate Purification System
. The Condensate Purification System (CPS) pudfies and treats the condensate as required to maintain reactor feedwater purity, using filtration to remove suspended solids, including corrosion products, ion exchange to remove dissolved solids from condenser leakage and other impurities, and water treatment additions to minimize corrosion / erosion product releases in the power cycle.
10.4.6.1 Design Bases 10.4.6.1.1 Safety Design Bases The CPS does not serve or support any safety function and has no safety desigs bases.
10.4.6.1.2 Power Generation Design Bases Power Generation Design Basis One-The CPS continuously removes dissolved and-suspended solids from the condensate to maintain reactor feedwater quality. -
Power Generation Design Basis Two-The CPS removes corrosion products from the - ,.
condensate and from drains returned to the condenser hotwell so as to limit any accumulation of corrosion prodtscts in the cycle.
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Power Generation Design Basis Three-The CPS removes impurities. entering the power cycle due to condenser circulating water leaks as required to permit continued power operation within specified water qualitylimits as long as such condenser leaks are too small to be readily Iccated and repaired. ;
Power Generation Design Basis Four-The CPS limits the entry of dissolved solids into the feedwater system in the event oflarge condenser leaks, such as a tube break, to permit a reasonable amount of time for orderly plant shutdown.
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Power Generation Design Basis Hve-The CPS injects in the condensate such water treatm ent additives as oxygen and hydrogen as required to minimize corrosion / erosion -i product releaes in the power cycle. ;
Power Generation Design Basis Six-The CPS maintains the condensate storage tank water quality as required for condensate makeup and miscellaneous condensate supply senices.
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10.4.6.2.1 General Description s- l The Condensate Purification System (Figure 10.4-4a and 10.4-4b) consists of three 33%
capacity high efficiency filters arranged in parallel and operated in conjunction with a l normally closed filter bypass. The CPS also includes six bead resin, mixed bed ion exchange demineralizer vessels arranged in parallel with, normally five in operation and one in standby. A strainer is installed downstream of each demineralizer vessel to preclude gross resin leakage into the power cycle in case of vessel underdrain failure, j and to catch resin fine 'eakage as much as possible. The design basis for the CPS system will be to achieve the water quality efIluent conditions defined in the GE water quality l specification. The CFS components are located in the Turbine Building.
Provisions are included to permit air scrub cleaning and replacement of the ion exchange resin. Each of the demineralizer vessels has fail.open inlet and outlet isolation valves which are remotely controlled from the local CPS control panel.
A demineralizer system bypass valve is also provided which is manually gg orc aNh [control from the main control room. Pressure downstream of the deEninI'INiferys7tdIc#atch Y as E and !c p== b alarmed in the main control room to alert the operator.The bypass
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l is used only in emergency and for short periods of time until the CPS flow is returned to normal or the plant is brought to an orderly shutdmm. To prevent unpolished condensate from leaking through the bypass, double isolation valves are provided with
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10.4.6.2.2 Component Description l Codes and standards applicable to the CPS are listed in Section 3.2. The system is designed and constructed in accordance with quality group D requirements. Design data for major components of the CPS are listed in Table 10.4-4.
Condensate Filter-The CPS includes three 33% capacity backwashable high efficiency filters of such type as the hollow fiber or sintered metal filters.
Condensate Demineralizers-There are six 20% capacity demineralizer vessels (one on standby) each constructed of carbon steel and lined with stainless steel. Normal operation full load steady-state design flowrate is 2.521/s of bed. Maximum flowrates are 3.15 and 3.791/s for steady state and transient operation, respectively. The nominal bed depth is 102 cm.
t 10.4.6.2.3 System Operation l The CPS is continuously operated to maintain feedwater purity levels.
Other Features of Steam and Power Conversion System - Amendment 32 10.4-21
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Power Generation Design Basis Four-The CFS is designed to permit continuous long-term operation with one LI heater string out ofsenice at the maximum load permitted by the turbine manufacturer (approximately 85%). This value is set by steam flow g limitation on the affected LP turbine.
Power Generation Design Basis Hve-The CFS is designed to heat up the reactor feedwater to 215.55 C during fullload operation and to lower temperatures during part load operation.
Power Generation Design Basis Six-The CFS is designed to minimize the ingress or release ofimpurities to the reactor feedwater. l mes.T B y 10.4.7.2 Description 10.4.7.2.1 General Description ,
The Condensate and Feedwater System (Figures 10.4-5a,10.4-5b, and 10.4-6) consists of .
the piping, valves, pumps, heat exchangers, controls and instrumentation, and the -
associated equipment and subsystems which supply the reactor with heated feedwater in a closed steam cycle utilizing regenerative feedwater heating. The system described l in this subsection extends from the main condenser outlet to (but not including) the seismic interface restraint outside of containment. The remainder of the system, extending from the restraint to the reactor, is described in Chapter 5. Turbine cycle steam is utilized for a total of six stages of closed feedwater heating. The drains from each stage of the low-pressure feedwater heaters are cascaded through successively
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lower pressure feedwater heaters to the main condenser. The high-pressure heater drains are pumped backward to the reactor feedwater pumps suction. The cycle extraction steam, drains and vents systems are illustrated in Figures 10.4-7 and 10.4-8. ,
The CFS consists offour 33-50% capacity condensate pumps (three normally operating and one on automatic standby), three normally operated 33-65% capacity reactor feedwater pumps, four stages oflow-pressure feedwater heaten, and two stages of high-pressure feedwater heaters, piping, valves, and instrumentation. The con'densate pumps take suction from the condenser hotwell and discharge the deacrated condensate into one common header which feeds the condensate filter /demineralizers. Downstream of the condensate demineralizen, the condensate is taken by a single header and flows in parallel through five auxiliary condenser / coolers, (one gland steam exhauster condenser and two sets of SJAE condensers and offgas recornbiner condenser (coolers). i The condensate then branches into three parallel strings oflow pressure feedwater heaters. Each string contains four stages oflow-pressure feedwater heaters. The strings ,
join together at a common header which is routed to the suction of the reactor .r feedwater pumps. >
i' Another input to the feedwater flow consists of the drains which are pumped backward and injected into the feedwater stream at a point between the fourth stage low-pressure Other Features of Steam and Power Conversion System - Amendment 33 10.4-25 5
s , ,
l INSERT B i
Power Generation. Design Basis Seven - All CFS functions needed to support power. operation will use at least dual redundant controllers and triply redundant signals; a single control system failure will not cause an inadvertent pump trip or valve operation.
f f
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23A6100 Rev.2 ABWR. stadantSafetyAnalysisReport
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The drain tanks and tank drain lines are designed to maintain the drain pumps available suction head in excess of the pump required minimum under all anticipated operating conditions including, particularly, load reduction transients. This is achieved
- s. mainly by providing a large elevation difference between tanks and pumps (approximately 15.24m) and opthnizing the drain lines which would affect the drain system transient response, particularly the dmin pump suction line.
Heater Dmin Pmnps-Two motor-driven heater drain pumps operate in parallel, each .
taking suction from the heater drain tank and discharging into the suction side of the reactor feedwater pumps. The drain system design allows each heater drain pump to be indhidually removed from service for maintenance while the balance of the system remains in operation, while the affected string drains dump to the condenser.
Controlled drain recirculation is provided from the discharge side of the heater dmin pump to the associated heater drain tank. This ensures that the minimum safe flow through each heater drain pump is maintained during operation.
Reactor Feedwater Pumps-Three iden tical and independent 33-65% capacity reactor -
feedwater pumps (RFP) are provided. The three pumps manually operate in parallel and discharge to the high-pressure feedwater heaters. The pumps take suction ,
downstream of the last stage low-pressure feedwater heaters and discharge through the high-pressure feedwater heaters. Each pump is driven by an adjustable speed q gn n n,,, ~ +- d rw4 -
Isolation valves are provided which allow each reactor feed pump to be individually removed from se rvice for maintenance, while the plant continues operation at full power on the two remaining pumps.
Controlled feedwater recirculation is provided from the discharge side of each reactor ;
feed pump to the main condenser. This provision ensures that the minimum safe flow through each reactor feed pump is maintained during operation.
10.4.7.2.3 System Operation y g&dd Normal Operation-U er normal operating conditions, system operation is .
automatic. Automati evel control systems control the levels in all feedwater heaters, M S[RA cham bkthe heater drain tanks, and the condenser hotwells. Feedwater heater levels are controlled by modulating drain valves. Control valves in the discharge and recirculation lines of the heater drain pumps control the levelin the heater drain tanks. Valves in the makeup line to the condenser from the condensate storage tank u a m the return line to the condensate storage tank control the level in the condense; botwells.
During power operation, feedwater flow is automatically controlled by the reactor feedwater pump speed that is set by the feed pump speed control system. The control. ~ .
system utilizes measurements ofsteam flow, feedwater flow, and reactor level to regulate 10.4 28 Other Features of Steam and Power Conversica System - Amendment 32
r -. - - .- . . ~ . - .- -. -
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s.
314s100 Rev.a 7
', : ARWR. suahntserayAnsorsisneput 3 1 the feedwater pump speed. During startup, feedwater flow is automatically regulated by ;
the high-pressure heater bypass flow control valve. ,
Ten-percent step load and 5%/ min ramp changes can be accommodated without a ~
major effect on the CFS The system is capable of accepting a full generator load ,
~
rejection without reducing feedwater flow rate.
10.4.7.3 Evaluation The Condensate and Feedwater System does not serve or support any safety function.
Systems analyses show that failure of this system cannot compromis~e any safety-related j system or prevent safe shutdown.
t During operation, radioactive steam and condensate are present in the feedwater 1 heating portion of the system, which includes the extraction steam piping, feedwater: .i heater shells, heater drain piping, and heater vent piping. Shielding and access control are provided as necessary (Chapter 12). The CFS is designed to minimize leakage with'
. welded constniction utilized where practicable. Relief discharges and operating vents are channeled through closed systems.
t Ifit is necessary to remove a component from service such as a feedwater heater, pump, or control valve, continued operation of the system is possible by use of the multistring arrangement and the pro isions for isolating and bypassing equipment and sections of '!
the system. -
The majority of the condensate and feedwater pipmg considered in this section is -
located within the non-safety-related Turbine Building. The portion which connects to '
l- the seismic interface restraint outside the containment is located in the steam tunnel .
between the Turbine and Reactor Buildings. This portion of the piping is analyzed for-dynamic effects from postulated seismic events. The feedwater control system is designed to ensure that there will not be large sudden changes in feedwater flow'that --
g could induce water hammer.
10.4.7.4 Tests and inspections' !
10.4.7.4.1 Preservice Testing j
- Each feedwater heater and condensate pump receives a shop hydrostatic test which is ! q
< performed in accordance with applicable codes. All tubejoints of feedwater heaters are - l l
shop leak tested. Prior to initial operation, the completed CFS receives a field '
hydrostatic and performance test and inspection in accordance with the applicable' a code. Periodic tests and inspections of the system are performed in conjunction with scheduled maintenance outages.
'l d
1
-i Other Features of Steam and Power Conversion System - Amendment 32 10.4 29
".* i 22As1ocEav. 2 ABWR. saadantsatoryAnalysissoport
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10.4.7.4.2 Inservice inspections The performance status, leaktightness, and stmetural leaktight integrity of all system '
components are demonstrated by continuous operation.
s 10.4.7.5 Instrumentation Applications ,
Feedwater flow-control instrumentation measures the feedwater discharge flow rate -
from each reactor feed pump and the heater bypass startup flow control valve. These .
feedwater system flow measurements are used by the Feedwater Control System (Subsection 7.7.1 A) to regulate the feedwater flow to the reactor to meet system demands. y A bh l Pump flow is measured on the pump inlet line, and flow controis pro ide automatic pump recirculation flow for each reactor feedwater pump. Automaticjcontrols aln regulate the condensate flow through the auxiliary condensers (offgas recombiner !
condenser / coolers, gland steam condenser, and SJAE condensers) and maintains condensate pump minimum flow. Measurements of pump suction and discharge pressures are provided for all pumps in the system. Main feedpump suction pressure, discharge pressure and flow are indicated in the main control room. -
The high-pressure feedwater heater isolation valves are interlocked such that, if a string " '
of heaters were to be removed from service, the extraction non-return valves and isolation valves for those heaters would automatically close and the heater string bypass valve open. The low pressure feedwater heater isolation valves are interlocked such that, if a string of heaters were removed from senice, the extractions to the affected heaters which are equipped with nonreturn val es would automatically close.
&A Sampling means ar prmided for monitoring the quality of the condensate and final feedwater, as desc ed in Subsection 9.3.2. Temperature measurements are prmided for each stage off edwater heating. Steam pressure measurements are provided at each feedwater heater evel instrumentation and controls are provided for autcmatically regulating the he - ter drain flow rate to maintain the proper level in each feedwater ,
heater shell or heater drain tank. High-level control valves prmide automatic dump-to-condenser of heater drains on detection of high level in the heater shell. ,
seL M o d The total water volume in the CF is maintained through automatic makeup and rejection of condensate to the ondensate storage tank. The system makeup and -
rejection are controlled by the condenser hotwelllevel controllers.
10.4.8 Steam Generator Blowdown System (PWR)
Not applicable to the ABWR. .
L 10.4-30 Other Features of Steam and Power Conversion System - Amendment 32
4 . ,
l INSERT C .
The CSFS trip logic and control schemes will respectively use coin coincident logic and redundant control FRS and input signals to assure that plant availability goals are achieved and spurious trips are avoided. This specifically includes all FW. heater and drain tank level controllers, all CFS flow and minimum flow controllers, and pump suction pressure trips, FW heater string isolation /high level trips and CFS bypass system (s) operation.
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23A6100 Rev.1
. ABWR- St:nd:rdSattyAriysis R:pirt
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Stellite is used for hard facing of components which must be extremely wear resistant.
Use of high cobalt alloys such as Stellite is restricted to those applications where no 4
satisfactory alternative material is available. An alternative material (Colmonoy) has been used for some hard facings in the core area.
oncl Mo4mol Selecdton 12.3.1.1 Equipment Design [for Maintaining Exposure ALARA iz, L i, t. t Equ pme + D e r,se m Thi's sulisection describes sptcific components, as well as system design features, that aid in maintaining the exposure of plant personnel during system operation and ,
maintenance ALARA. Equipment layout to provide ALARA exposures of plant j personnel is discussed in Subsection 12.3.1.2. ~j (1) Pumps Pumps located in radiation areas are designed to minimize the time required for maintenance. Quick change cartridge-type seals on pumps, and pumps -
with back pullout features that permit removal of the pump impeller or mechanical seals without disassembly of attached piping, are employed to minimize exposure time during pump maintenance. The configuratio'n of -
piping about pumps is designed to provide sufIicient space for efiicient pump .
/
maintenance. Provisions are made for flushing and in certain cases chemically cleaning pumps prMr to maintenance. Pump casing drains pro ide a means I
for draining pumps to the sumps prior to disassembly, thus reducing the exposure of personnel and decreasing the potential for contamination.
Where two or more pumps conveying highly radioactive fluids are required for operational reasons to be located adjacent to each other, shielding is provided between the pumps to maintain exposure levels ALARA. An example of this situation is the CUW circulation pumps. Pumps adjacent to other highly radioactive equipment are also shielded to reduce the maintenance exposure, for example, in the Radwaste System Whenever possible, operation of the pumps and associated valving for radioactive systems is accomplished remotely. Pump control instrumentation is located outside high radiation areas, and motor or pneumatic-operated valves and valve extension stems are empicyed to allow operation from outside these areas.
(2) Instrumentation Instruments are located in low radiation areas such as shielded valve galleries, corridors, or control rooms, whenever possible. Shielded valve galleries provided for this purpose include those for the CUW, FPCC, and Radwaste (cleanup phase separator, spent resin tank, and waste evaporator) Systems.
Instruments required to be located in high radiation areas due to operational .
1232 Radiation Protection Design Features - Amendment 31
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23A6100 Rev.1
- ABWR staird
- rdsaferyA=1ysisa:p:rt 2
(8) SGTS Filters The SGTS filter is located in a separate shielded cubicle and is separated by a shield wall from the exhaust fans to reduce the radiation exposure of personnel durirt maintenance. The dampers located in the cubicles are remotelymperaten thus r quiring no access to the cubicle during operation.
A pneumatic transfer system is employed to remove the radioactive charcoal from the filter, requiring entry into the shielded cubicle only during the connection of the hoses to the SGTS filter unit.
IN GEPT A 12.3.1.2 Plant Design for Maintaining Exposure (ALARA)
This subsection describes features of equipment layout and design which are employed -
to maintain personnel exposures AIARA (1) Penetrations Penetrations through shield walls are avoided whenever possible to reduce the number of streaming paths provided by these penetrations. Whenever penetrations are required through shield walls, however, they are located to minimize the impact on surrounding areas. Penetrations are located so that the radiation source cannot "see" through the penetration. When this is not possible, or to provide an added order of reduction, penetrations are located to exit far above floor level in open corridors or in other relatively inaccessible areas. Penetrations which are offset through a shield wall are frequently employed for electrical penetrations to reduce the streaming of radiation ,
through these penetrations.
Where permitted, the annular region between pipe and penetration sleeves, as well as electrical penetrations, are filled with shielding material to reduce the streaming area presented by these penetrations. The shielding materials med in these applications include a lead-loaded silicone foam, with a density
. emparable to concrete, and a boron-loaded refractory-type material for -
applications requiring neutron as well as gamma shielding. There are certain penetrations where these two approaches are not feasible or are not . .
sufliciently effective. In those cases, a shielded enclosure around the penetration as it exits in the shield wall, with a 90 degree bend of the process pipe as it exits the penetration,is employed.
(2) Sample Stations Sample stations in the plant provide for the routine surveillance of reactor water quality. These sample stations are located in low radiation areas to reduce :he exposure to operating personnel. Flushing provisions are included ;
using dem:ncralized water, and pipe drains to plant sumps are provided to i bdiation Protection Design Features - Amendment 31 12.3-S '
.l
- 39w, f .
INSERT A l 12.3.1.1.2 Material Selection ;
In the ABWR design, radiation exposure potential has'been reduced through the removal or reduction of cobalt from many components exposed to reactor coolant as ,
compared to the current BWR fleet. Standards have been established that reduce ,
cobalt in pipe, core items, reactor internals, jet pumps, CRD pins, rollers and blades, feedwater heaters, and valves including MSIV's. Much of the cobalt shall be ' removed from contact with reactor coolant by eliminating Stellite and reducing cobalt _in the core stainless steel items. There are many components that still contain Stellite. These are mostly valves, including MSlV's, that have Stellite seats. At the present time there is no certified substitute for Stellite. EPRI is developing an iron base alloy callea NOREM, but it needs to be fully tested and accepted by the nuclear industry. 3 The present cobalt standards for the ABWR are: 0.02 wt percent for those items in the - !
core; 0.03 wt percent for those items in the internals; and 0.05 wt percent for all else.
These values will probably all be reduced over a period of time. Presently the cost to. ' i produce 0.02 wt percent cobalt stainless steel vs. 0.05 wt percent is about 8 times , so .
appreciable effort is involved to reduce cobalt content. It is clear that further work on Stellite substitutes and cobalt reduction in stainless steel must continue. In this regard, COL applicants should further reduce maintenance exposure throulgh material-selection by the following: ..I
- 1. Utilization of the progress of NOREM at EPRI and of the industry to produce a ,
workable material to be used in place of Stellite in valves and in 'other wear surfaces ' '!
exposed to reactor coolant.
- 2. Encouraging the metal industry to produce lower wt percentage cobalt stainless steel than now exists for the core components, the internals, the piping and the ,
feedwater heaters.
- 3. Utilization of the (a) water chemistry developments where xinc injection inhibits j cobalt plating on surfaces inside pipes and the RPV, and (b) hydrogen water !
chemistry developments as this system appears to flush out cobalt and other i radioactive products (sludge) from the cracks, crevices and traps that become.
radioactive sources that can expose maintenance personnel. ,
See Subsection 12.3.7.4 for COL license information requirements.
a I
b
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23A6100 R2v.1 ABWR sta d:rdSaintyA=lysisR: port 12.3.7.3 Requirements of 10CFR70.24 COL applicants will provide information showing that their plant meets the requirements of 10CFR70.24 or request an exemption from this 10CFR 70.24 requirement (Subsection 12.3.4.3).
12.3.8 References 12.Sl N. M. SchaefTer, ReactorShieldingforNuclearEngineers, TID-25951, U.S. Atomic Energy Commission (1973).
12.S2 J. I1. Hubbell, Photon Cross Sections, Attenuation Coefficients, andEnergy Absorption Coeficientsfrom 10 kev to 100 GeV, NSRDS-NBS20, U.S. Department of Commerce, August 1969.
12.13 RadiologicalIIralth IIandbook, U.S. Department of Health, Education, and ,
Welfare, Revised Edition, January 1970.
12.3-4 Reactor IIandbook, Volume III, Part B, E.P. Blizzard, U.S. Atomic Energy ;
Commission (1962).
12.3-5 Lederer, Hollander, and Perlman, Table ofIsotopes, Sixth Edition (1968).
12.% M.A. Capo, Polynomial Approximation of Gamma Ray Buildup Factorsfor a Point Isotropic Source, APEX-510, November 1958.
12.3-7 Reactor Physics Constants, Second Edition, ANIe5800, U.S. Atomic Energy Commission, July 1963.
12.S8 ENDF/&llI and ENDF/IkIV Cross Section Libraries, Brookhaven National Laboratory.
12.S9 PDS-31 Cross Section Library, Oak Ridge National Laboratory.
12.S10 - DLC-7, ENDF/B Photo Interaction Library.
- 12. '3,7 4 tA cdo v s a\ ~ S o.Ic b oa I
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Radiation Protection Design Features - Amendment 31 12.3-29
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23A6100 Rev. 3 l
'. ABWR StandardSafety Analysis Report 19.9.16 High Pressuro Core Flooder Discharge Valve As noted in Subsection 19D.7.7.5, the HPCF loop B pump discharge valve is in the drywell. Plant procedures should include independent verification that the valve is locked-open following maintenance.
19.9.17 Capability of Containment isolation Valves To insure that containment isolation valve capability does not reduce the containment capability, the COL applicant will demonstrate that the stresses of containment isolation l valves, when subjected to severe accident loadings of 0.77 MPa (97 psig) internal pressure and 260 C (500*F) temperature in combination with dead loads, do not exceed ASME Section III senice level C limits. In addition, the ultimate pressure l capability at 260 C (500 F) will be shown to be at least 1.03 MPa (134 psig).
19.9.18 Procedures to insure Sample Lines and Drywell Purge Lines Remain Closed During Operation As noted in Subsection 19.8.4.3, it is important that these lines be normally closed during plant operation. The COL applicant will develop procedures and administrative controls to ensure the valves are normally sealed closed and that the purge valves have motive power to the valve operators removed.
19.9.19 Procedures for Combustion Turbine Generator and Emergency Diesel Generators to Supply Power to Condensate Pumps The COL applicant will implement procedures for manual transfer of Combustion Turbine Generator (CTG) and Emergency Diesel Generator (EDG) power to the condensate pumps. As noted in Subsection 19.8.1.3, the ability of the EDG to power the condensate pumps provides a small reduction in the core damage frequency. The CTG is the preferred source if both the CTG and an EDG are available. Condensate pump support systems (lube oil, cooling water) are also needed if the pumps are to provide water to the RPV for substantial periods of time.
19.9.20 Actions to Assure Reliability of the Supporting RCW and Service Water.
Systems To assure the reliability of the RCW and Senice Water Systems, the COL applicant will ;
take the following action. At least each month, the standby pumps and heat exchangers are started and the previously running senice and sea water equipment is placed in a standby mode. i I E 9. 2 I Wou sm of ATCWA- Ep &cd sow ( AICWA) eJ w pw.,4 d is T-f o c cA c Phdtwk w eA-w cAcka w % A t L \a u gemph ,
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COL License information - Amendment 33 19.9-11/12
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ABWR St:rud:rd Skty Aulysis R:p:rt The success or failure of emergency DC power (station batteries) (node DP), and the emergency AC power and/or senice water (node APW) are taken into consideration in Figure 191-1 to accour.t for support system dependencies. Failure of all DC power results in a high-pressure core melt since all controlis lost, the high-pressure systems fail, and.
the reactor cannot be depressurized.The condidon ofsuccessful emergency DC and AC power and successful scram is indicated by the ET transfer and is described in detail in Figure 191-2. The condition of successful emergency DC and AC power, but with failure to scram is indicated by the ATWS transfer, and is described in Figure 191-3. ,
The condition of successful emergency DC and failure of emergency AC continues on Figure 191-1. The next question is whether or not there is a failure to scram (node C).
Failure to scram is considered as a Class IV core melt. With successful scram, RCIC (node UR) and firewater (node FA) are the only available means ofwater injection into the RPV since all AC power is lost. Since station batteries will eventually discharge resulting in loss of RCIC, or if RCIC fails, the reactor must then be depressurized (node X) to allow firewater injection. The loss of emergency DC power (station batteries)
[ results in a high-pressure core melt as shown in Figure 191-1.
The firewater system has diesel driven pumps**and Meb^Q*
See %b:e sh 19.9.Ei fo- cot Wu A-ab P all needed valv operated manually. No support systems are required for firewater operation. %e-
-firewater-pump i: housed in an-extemal-buikling4hed), rhc e collapse-would. net-l -prewrmh pumphem st2rdng-and+unning. The random failure probability of firewater is dominated by operator failure to initiate the system. For the upper branch, where RCIC is successful, the operator has 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> before the station batteries expire and RCIC trips. The human error probability (HEP) for this case is 1E-3. For the lower branch, where RCIC fails, the operator has only 30 minutes in which to depressurize the reactor and initiate firewater injection. For this case, the HEP is 0.1. In the event th'at the firewater diesel fails to start, the operator could make use of a fire truck, but this was not modeled.
If the RHR heat exchanger fails (node HX) due to the earthquake,it is presumed that the failure could include a pipe break that could partially drain the suppression pool into the RHR pump room. These core damage sequences are identified with a "P" (e.g.,
IB2-P). Fission product scrubbing would still be effective in preventing a large release.
The effects of possible flooding on equipment operation beyond the RHR room were ,
considered and found to be relatively insignificant because of the relatively high HCLPF of the heat exchangers (0.70), the ability of the operator to isolate the break, and the presence of the independent ACIWA (firewater) system.
191.3.2 LOSP with Emergency Power and Scram Event Tree In the event tree of Figure 191-2 (ET transfer), there are two similar dhisions depending on whether or not there is a stuck.open relief valve (node PC). If there is a stuck-open valve, the reactor will eventually depressurize causing loss of RCIC steam supply. The Seismic Margins Analysis - Amendment 33 19I-3 ,