ML20010G024
| ML20010G024 | |
| Person / Time | |
|---|---|
| Site: | Rancho Seco |
| Issue date: | 02/23/1981 |
| From: | BABCOCK & WILCOX CO. |
| To: | |
| Shared Package | |
| ML20010G015 | List:
|
| References | |
| RTR-NUREG-0737, RTR-NUREG-737, TASK-2.E.1.2, TASK-TM BWNP-20004, TAC-42933, NUDOCS 8109150239 | |
| Download: ML20010G024 (39) | |
Text
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.a-BWNP-20004 (6-76) 8ABCOCK & WILCOX NUCLEAR PCWER GENERAflON DIVISION TECHNICAL DOCUMENT SYSTEM DESCRIPTION 15 1120580 01 Doc. ID - Serial No., Revision No.
for AUXILIARY FEEDWATER SYSTEM i
FOR SACRAMENTO MUNICIPAL UTILITY DISTRICT RANCHO SECO l
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PAGE 1 8109150239 810908 PDR ADOCK 05000312 P
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BWP-20005 (6-76)
BABCOCK & WILCOX NUCLEAR PCwit GENERATION DIYt31CN NUMBER RECORD OF REVISION is-1120ss0-<'t REY. NO.
CHANGE SECT / PARA.
DESCRIPTION / CHANGE AUTHORIZATION 01 This revision incorporates approved CI/A Number 88-618s-00.
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N. mc //
Title Z.3 f65 8/
Date DATE:
2-20-31 PAGE 2
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BABCOCK & WILCOX Numeen NUClf AR POWER GENERATCN OfV!$CN TABLE OF CONTENTS / EFFECTIVE PAGE llST 15-1120580-01 SECTION TITLE PAGE 000. NO.
1.0 SCOPE 4
15-1120580-01 2.0 SYSTEM REQUIREMENTS 4
15-1120580-01 5
15-1120580-01 6
15-1120580-01 7
15-1120580-01 8
15-1120580-01 9
15-1120580-01 10 15-112G580-01 11 15-1120580-01 12 15-1120580-01 13 15-1120580-01 TABLE 2-1 OTSG EMERGENCY FEEDWATER CHEMISTRY REQUIREMENTS 14 15-1120580-00 3.0 DESIGN DESCRIPTION 15 15-1120580-01 16 15-1120580-00 17 15-1120580-00 18 15-1120580-01 19 15-1120580-01 20 15-1120580-01 21 15-1120580-01 22 15-1120580-00 23 15-1120580-01 24 15-1120580-01 25 15-1120580-01 26 15-1120580-00 27 15-1120580-01 28 15-1120580-01 4.0 SYSTEM LIMITS, PRECAUTIONS AND SETPOINTS 28 15-1120580-01 29 15-1120580-01 30 15-1120580-01 FIGURE 3.3-1 AC POWER DISTRIBUTION TO COMPONENTS IN AFWS AND SVC WATER SYSTEM -
RANCHO SECO (LaTER) 31 15-1120580-01 FIGURE 3.3-2 125 VDC AND VITAL 120 V;E POWER DISTRIBUTION (LATER) 32 15-1120580-01 DATE:
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BABCOCK & WILCOX wumeen auct:A4 POWER GENERADON OMSCN 15-1120580-01 TABLE OF CONTENTS / EFFECTIVE PAGE LIST SECTION TITLE PAGE 000. NO.
FIGURE 4.2-1 FLUX TO FEEDWATER SETPOINT (LATER) 33 15-1120580-01 TABLE 4.2-1 EFW SYSTEM SETPOINTS 34 15-1120580-01 APPENDIX A TABULATION OF DRAWING NLMBERS VS.
FIGURE NUh' IRS FOR RANCHO SECO AFW SYSTEM A-1 15-1120580-01 APPENDIX B INSTRUMENTATION REQUIREMENTS B-1 15-1120580-01 B-2 15-1120580-01 APPENDIX C FLUX /FEEDWATER SETPOINT C-1 15-1120580-01 l
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BABCOCK & WILCOX Numen NUCLEAR POWER GlNERATioN olvlSloN TECHNICAL DOCUMENT 15-1120580-01 1.0 SC OPE This document contains the system description for auxiliary (01) feedwater (AFW).
The requirements for this system come from three sources - first, the functional requirements needed to properly interface the AFW system with the nuclear steam supply system (NSSS);
second, NUREG-0578, Short Term Lessons Learned Report; third, NUREG-0667, Transient Response of B&W Designed Reactors. This document contains the criteria necessary to upgrade the AFW system to comply with the Standard Review Plan Section 10.4.9, Branch Technical Position ASB10-1 and other standards generally applied to new de* gns.
In implementing these requirements, some exceptions may be taken where the improvement in system reliability is so small that the required modification is not justified for an operating plant.
Note that "feedwater", as used in this document, refers to AFW unless otherwise stated.
2.0 SYSTEM REOUIREMENTS The AFW system requirements are listed below.
(01) 2.1 NSS Interface Requirements 2.1.1 Maximum Feedwater Flow The maximum allowable FW fl ' is 1650 gpm per steam generator (SG).
This maximum FW flow limit s based on a tube vibration crossflow velocity limit of 5 ft/s.
his limit must not be exceeded at any steam pressure.
2.1.2 Minimum Available Feedvater Flow The AFW system must be sized so that a minimum of 760 gpm (total)
(01) can be delivered to either one or both SGs at a SG pressure of 1050 psig. This flow must be available for all accident conditions con-sidered in the design basis for the plant even with a single active t
failure in the system.
Note: BAW 1610, Analysis of B&W NSS Response to ATWS events, (01)
January 1980, used AFW flowrates of 1480 gpm at 15 sec for the loss of feedwater event and 740 gpm at 15 sec/1480 gpm at 40 see for the loss of offsite power case on 177 FA plants. These are nominal flows assuming no failuges. Any significant deviations from these values
=ust be justified.
DATE:
pAGE 2-20-31 4
BWNP-20007 (6-76)
BABCOCK & WILCOX yyy,,
NUCEAa power oENERATioN CIVISION TECHNICAL DOCUMENT 15-1120580-01 2.1.3 Maximum Automatic Initiation Time The system shall be designed so that the minimum AFW flow is established within 50 seconds after an initiation signal is (01) reached. This requirement is based on:
A.
Maintaining continuity in reactor coolant system (RCS) flow in the transition from ferced to natural circulation when the RC pumps (RCPs) are tripped.
B.
Reducing the probability of boil off of the entire inventory of water immediately following a loss of main FW occurrence.
C.
Providing margin to prevent overpressurization of the RCS following a loss of main FW event and reactor trip.
NOTE: The 50 secon_d delay includes instrumentation time delay, (01) diesel startup, diesel sequencing, pump acceleration time and valve stroke time.
2.1.4 Initiation and Control Requirements 2.1.4.1 General Requirements The requirements to which the AFW control system shall be designed are:
A.
The system shall provide automatic actuation of AFW, for the conditions specified in Section 2.1.4.2.
The capability for bypassing certain initiations shall be provided for unit startup or shutdown in accordance with the IEEE-279 provisions for shutdown bypssses.
B.
The system shall be designed to minimize overcooling following a loss of main FW event. This feature of the system is not required to meet the single failure criterion.
C.
The system, including control valve positioners, sensors, control and actuation signals and their auxiliary supporting systess, shall be designed as a safety grade (IE) system to the exte st possible. As such, i t shall be independent of the ICS, NNI, and other non-safety sysu D.
Redundancy and testability shall be provided to enhance the reliability demanded of a safety grade system.
l DATE:
2-20-31 PAGE 5
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BA8 COCK & Wil.COX NOCLEAR Powtt GENERATION olvisiCN TECHNICAL DOCUMENT 15-u20580-ol E.
A single failure shall neither prevent actuation of AFW when required nor spuriously actuate the system. This criterion shall apply to the AFW system and its auxiliary supporting features.
In addition to this single failure, all failures which can be predicted as a condition or a result of the initiating event requiring AFW shall be ectslaered.
F.
Indication of AFW status, flowrate and OTSG 1evel shall be available to the operator.
G.
The capability for a manual override of the automatic functioning of the system shall be provided. This condition shall be annunciated in the control room.
H.
The capability for manual initiation of AFW shall be provided.
1.
The capability for manual initiation and control shall be provided in the main control room. The capability for future installation of control from a remote shutdowu panel shall be provided.
J.
The system shall be designed to prevent or minimize cycling of the AFW control valves during normal plant operation when the AFW system is not in operation.
K.
The system shall provide the capability to control the atmospheric dump valves to a single, predetermined setpoint and in addition shall have manual override capability.
L.
Provisions shall be made to initiate AFW to mitigate the conse-(01) quences of a LOFW transient by a flux-feedwater ratio trip signal.
2.1.4.2 Actuation Requirements AFW shall be automatically initiated af ter the occurrence of any of the following conditions:
o Loss of all main FW as a minmum, as i..'tcated by the loss of both
=ain FW pumps, i.e., low pump discharge essure.
o Low level in either SG.
o Loss of all 4 RCPs.
o Low pressure in either SG if main FW is isolated on this parameter.
ESTAS ECCS actuation (high RB pressure or low RC pressure).
(01) o o
Power /MFW flow.
(01)
DATE:
2-20-81 PAGE 6
BWNP-20007 (6-76)
BA8 COCK & WILCOX NUCLEAR POWER GENERATloM DIVI $loN Numeen TECHNICAL DOCUMENT 15-1120580-01 2.1.4.3 Level
- Requirements Three adjustable level setpoints are required.
A.
Following AFW actuation, the level setpoint shall be automat-ically selected to approximately 3 feet if one or more RCPs are (01) running.
B.
Following AFW actuation, the level setpoint shall be automat-ically selected to approximately 20 feet if all 4 RCPs are tripped.
C.
Provision for manual selection of a high level setpoint of approximately 31.5 feet shall be provided. This setpoint will be (01) selected by the operator in accordance with operating guidelines.
- For the purpose of AFW design, " LEVEL" refers to the equivalent height of a saturated liquid column (900 psia) referenced. rom the top of the lower tube aheet.
2.1.4.4 Flowrata Requirement (01)
The objective of the flowrate control is to ninimize overcooling for low DH conditions. The EFW flow rate is controlled by the rate of level increase (see Section 2.1.4.3 for level definition). A level rate of 2 to 4 inches per ninute has been estimated to provide adequate RCS cooling. This fill rate is varied as a function of steam generator pressure in the range of 800 to 1050 psig for the transient conditions which require AFW.
Since the level rate control is a first of a ki'nd control scheme, the system must be tested in place to guarantee that the setpoint is sufficiently high to provide adequate cooling for the maximum heat load.
The level rate limit shall be adjustable under administrative control.
In operation, the EFW flowrate is modulated to hold the level rate at the setpoint.
2.1.5 Steamline Break /Feedwater Line Break (01) l A steaaline break or FW line break that depressurizes a SG shall cause the isolation of the main steamlines and main FW lines on the depressurized SG.
If isolation of the SG does not isolate the break, AFW shall be provided only to the intact SG.
No single active l
failure in the system shall prevent AFW from being suppplied to the intact SG nor allow AFW to be supplied to the broken SG.
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DATE:
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BWNP-20007 (6-76)
BABCOCK & WILCOX su een NUCLEAR Powlt otNERAfloM OlVI$loN TECHNICAL DOCUMENT 15-1120580-01 To meet these requirements the following design shall be implemented:
(01)
A.
Isolation - Low steam pressure (below approximately 600 psig) in either SG will isolate the main steamlines and main FW line to the affected SG.
B.
SG Selection -
o If both SGs are above 600 psig, supply AFW to both SGs.
o If one SG is below 600 psig, supply AFW to the other SG.
o If both SGs are below 600 psig but the presso._
afference between the two SGs exceeds a fixed setpoint (approximately 150 psig) supply AFW only to the SG with the higher pressure.
If both SGs are below 600 psig and the pressure difference is o
less than the fixed setpoint, supply AFW to both SGs.
(01) 2.1.6 Steam Generator Overfill Provisions must be made in the design to tecninate a main FW and AFW overfill condition. Provisions must also be made to manually bypass (01) the AFW overfill setpoint following a LOCA to permit establishing an OTSG level which will support steam condensation natural circulation in the RCS.
2.2 Fluid System Requirements 2.2.1 Branch Technical Position ASB10-1 BTP ASB10-1 places the following requirements on the AFW system:
l A.
The auxiliary FW system should consist of at least two full l
capacity, independent systems that include diverse power j
sources.
3.
Other powered components of the auxiliary FW systam should also use the concept of separate and multiple sources of motive energy.
An example of the required diversity would be two separate auxiliary FW trains, each capable of removing the afterheat load of the reactor system, having one separate train powered from either of two AC sources and the other train wholly powered by steam and DC electric power.
C.
The piping arrangement, both intake and discharge, for each train should be designed to permit the pumps to supply FW to any combination of SGs. This arrangement should take into account pipe failure, active component f ailure, power supply failure, or I
l DATE:
2-20-81 pAGE 3
L
BWNP-20007 (6-76) 9ABCOCK & WILCOX Numeen NUCLEAR Powlt GENERATION DivtSloN TECHNICAL DOCUMENT 15-112o580-01 control system failure that could prevent system function. One arrangement thr.t would be acceptable is crossover piping containing valves that can be operated by remote manual control from the control room, using the power diversity principle for the valve operators and actuation systems.
D.
The auxiliary FW system should be designed with suitable redundancy to offset the consequences of any single active component failure; however, each train need not contain redundant active components.
E.
When considering a high energy line break, the system should be so arranged as to assure the capability to supply necessary auxiliary FW to the SG despite the postulated rupture of any high energy section of the system, assuming a concurrent single active failure.
NOTE: If the AFW system is not used (and therefore not pressurized) during startup, hot standby and shutdown conditions, then a high energy line break in the AFW system only needs to be considered between the SG and the first check valve upstream of the SG.
2.2.2 Water Sources Seismic Category I water sources shall be provided of sufficient volume to remove decay heat for four hours and to subsequently cooldown the plant to the decay heat removal (DHR) system pressure.
2.2.3 AFW Pump Prctection The system design shall protect the AFW pump from runout and cavita-tion due to high energy line breaks or single failures in the system. Any automatic pump trip features must (a) not override automatic initiation of AFW, or (b) be designed as a Class LE system.
2.2.4 AFW Support Systems The requirements for diverse power sources and operation with a single failure also apply to the AEW support systems. These systems include:
o Electrical power to support systems.
o Compressed air.
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i BWNP-20007 (6-76)
BABCOCK & WII.COX NucuAn mwen cesena. ion Onn$loM Nu e.
TECHNICAL DOCUMENT 15-1120580-01 2.2.5 Cross Connects (ut)
AFW system shall be designed to allow either pump to feed either steam generator. Cross connects provided for this purpose shall include normally open remotely operated isolation valves.
2.2.6 Alarms As a minimut, the following alarm outputs are required:
o High SG level.
o Low SG level.
o Low source water level.
o Low AFW pump discharge pressure.
o Low AFW pump suction pressure.
Steam line valves HV-20569 and HV-20596 not open.
o o
AFW cross connect valves EV-31826 and EV-31827 not open.
2.2.7 Indication As a minimum, the following indication shall be provided to the operator.
o AFW flow to each SG*.
o Startup range SG 1evel*.
o Operate range SG 1evel*.
o Wide range SG level.
o Key valve positions.**
o Water source inventory.
o Control system status (level setpoint selected).
o Steam pressure to each SG.
o AFW pump status indication.
o Indications needed to check the status of AFW support systems.
o Additional primary system indication as required to monitor system functions and operations *.
o Status of the EFIC system (bypass, test, tripped, etc.)
- Depending on the extent of compliance to R.G. 1.97, these indications may be required to be safety grade.
DATE:
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BWNP-20007 (6-76) i BABCOCK & NVILCOX i
NUMSER NUCitA4 PoWit GtMERATioM DIVI $ ION TECHNICAL DOCUMENT 15-1120580-01
- Direct position indication (e.g., valve stem p;41 tion) shall be provided for all automatically operated valves and all remote manual power operated valves. Local manual valves in the flow path shall be locked open. Strict administrative control should be exercised over the use of these valves.
2.2.8 Physical Separation System components and piping shall have sufficient physical separation or shielding to protect the essential portions of the system from the effects of internally and externally generated missiles.
Functional capability of the system shall also be assured for fires and the maximum probable flood.
2.2.9 Fluid Flow instabilities The system design shall preclude the occurrence of fluid flow instabilities; e.g.,
water hammer, in system inlet piping during normal plant operation or during upset or accident conditions.
2.2.10 Operational Testing Provisions shall be made to allow periodic operational testing.
2.2.11 Water Chemistry The requirements of the B&W Water Chemistry Manual, BAW-135, shall be met.
The normal water source shall neet the requirements in Table 2-1.
4.3 Codes and Standards l
l l
The AFW system shall consider the requirements of the following codes and standards:
I A.
General Design Criterion 2*,
Design Bases for Protection Against Natural Phenomens, as related to structures housing the system and the system itself being capable of withstanding the ef fects of natural phenocena such as earthquakes, tornadoes, hurricanes, and floods.
3.
General Design Criterion 4*,
Environmental and Missile Design l
Bases, with respect to seructures housing the system itself being capable of withstanding the effects of external missiles and internally generated missiles, pipe whip, and jet impingement forces associated with pipe breaks.
t
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DATE:
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3WNP-20007 (6-76)
BABCOCK & WILCOX Numeen NUCLEAA POWER GENEeATlQN DIVI $loM TECHNICAL DOCUMENT 15-1120580-o1 C.
General Design Criterion 5*,
Sharing of Structures Systems and Components, as related to the capability of shared systems and components important to safety to perform required safety functions.
D.
General Design Criterion 19*, Control Room, as related to the design capability of system instrumentation and controls for prompt hot shutdown of the reacter and potential capability for subsequent cold shutdown.
E.
General Design Criterion 44*, Cooling Water, to assure:
(1) The capability to transfer heat loads from the reactor system to a heat sink under both normal operating and accident conditions.
(2) Redundancy of components so that under accident conditions the safety function can be performed assuming a single active component failure.
(This may be coincident with the loss of offsite power for certain events.)
(3) The capability to isolate components, subsystems, or piping if required so that the system safety function will be maintained.
F.
General Design Criterion 45*, Inspection of Cooling Water System, as related to design provisions made to permit periodic inservice inspection of system components and equipment.
G.
General Design Criterion 46*, Testing of Cooling Water System, as related to design provisions made to permit appropriate functional testing of the system and components to assure structural integrity and leak-tightness, operability and performance of active components, and capability of the integrated system to function as intended during normal, shutdown, and accident conditions.
H.
Regulatorv Guides 1.22, Feb 1972*
Periodic Testing of Protection System Actuation Functions 1.26, Rev 3, Sept 1978*
Quality Group Classifications and Standards for Water, Steam and Radioactive Waste Containing Components 1.29, Rev 3, Sept 1978*
Seismic Design Classification 1.47, May 1973 Bypassed and Inoperable Status Indication 1.53, June 1973 Application of the Single Failure Criterion DATE:
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BABCOCK & Wit.COX meu NUCLEAR PCwle GENERAflON DIVl$lON TECHNICAL DOCUMENT 15-112o580-01 1.62, Oct 1973 Manual Initiation of Protective Actions 1.75, Rev 2, Sept 1978 Physical Independence of Electrical Systems
[1.97,Rev1,Aug1977 Instrumentation to Assess Plant Conditions During and Following an Accident}
1.102, Rev 1, Sept 1976 Flood Protection for Nuclear Power Plants 1.
IEEE Standards 279-1971*
Criteria for Protection Systems for Nuclear Power Generating Stations (for initiation portions of AFW System) 323-1971*
General Guide for Qualifying Class I Electrical Equipment 338-1971 Trial Use C-iteria for Periodic Testing of Protection Systems 344-1971*
Seismic Qualification of Class 1E Electrical Equipment 379-1972 Trial Use Guide for the Application of the Single Failure Criterion 384-1974 Separation of Class 1E Equipment and Circuits
- As a minimum, B&W recommends that these standards be met.
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DATE:
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BA8 COCK & WILCOX NUcLEAa Powte GENERATION DIVISION TECHNICAL DOCUMENT 15-1120580-00 TABLE 2-1 OTSG Emergency Feedwater Chemistry Requirements pH at 77F Same as normal requirement (a)
Disssolved oxygen (0 )
2 OTSG at < 250F No reauirement (see hydrazine)
OTSG at > 250F No rmal 7 ppb max Upse t 100 ppb max for a period not to exceed I week Total iron 100 ppb max Hydrazine Catalyzed hydrazine OTSG at < 25CF Added to at least 300% of stoichiometric oxygen concen-tration OTSG at > 250F 20-100 ppb residual Cation Conductivity
- 1. 0 sho/cm, max for a period not to exceed 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> I4)o.5-9.3 at 77F - Austenitic stainless steel. feedwater heater tubes and stainless steel or copper-nickel reheater tubes.
9.3-9.5 at 77F - Carbon steel feedwater heater tebes or combinations of carbon ateel and stainless steel feedwater ard/or reheater tubes.
DATE:
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BABCOCK & WILCOX NUCLEAa PQwit otNEAAftoM OlVtSloN TECHNICAL DOCUMENT 15-1120580-01 3.0 DESIGN DESCRIPTION 3.1 Summary Description The AFW system consists of two interconnected trains, capable of supplying suziliary feedwater (AFW) to either or both SGs from either water source under automatic or manual initiation and control. A piping and instrumentation diagram is included as Figure 3.1-1 of this report.
The system pumps (AFW pumps) take suction from either the condensate storage tank or from the Folsom South Canal and discharge to the SGs.
In the flow path between the AFW pumps and the SGs there are isolation valves, check valves, control valves, flow instrumenta-tion, and pressure instrumentation to control the flow of AFW to the SGs. The fluid system design is described in Section 3.2.
The instrumentation system design is described in Section 3.4.
3.2 Fluid System Design The AFW system is designed to provide a minimum of 760 gpm of AFW (01) to the SGs at 1050 psig within 50 seconds of system initiation signal. The system is designed as two interconnected trains with redundant components to insure that the system will meet these requirements with a single failure. Figure 3.1-1, depicts the piping and instrumentation diagram.
3.2.1 Suction The primary water source for both AFW trains is the Seismic Category I condensate storage tank, T-358.
Although there are other connections to this tank, they draw through an internal str. u-pipe l
which assures that a minimum of 250,000 ysilons is held in reserve l
exclusively for the AFW system. Water is suppliad from this tank to the AFV pumps by separate 8-inch lines containing locked open manual valves MCM-057, MCM-058, FWS-045, FWS-046, and check valves MCM-059 and MCM-060.
l Alternative AFW system suction sources are available from the on-site l
reservoir and the Folsom South Canal. These alternate sources enter the cross connect in the suction piping between locked closed manual valves PWC-076 and PWC-079. Suction must be manually transferred from the condensate storage tank to the reservoir or the Folsom South l
Canal by opening the locked closed manual valves PWC-076 and PWC-079, l
closing the locked open manual valves MCM-057 and MCM-058, and either: (1) operating the Folsom South Canal transfer pumps and l
valves or (2) opening motor operated valve HV-43011 to obtain gravity flow from the on-site reservoir. The suction cross connect l
DATE:
2-20-81 PAGE
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BWNP-20007 (6-76)
SA8 COCK & WILCOX wu ses nuct:4a powee cemenAnow omsc" 15-1120580-00 TECHNICAL DOCUMENT also includes pressure relief valves PSV-31800 and PSV-31900. The operators are alerted to perform this succion transfar by redundant, safety grade low level alarm from the condensate storage tank.
In addition, tank level is redundantly indicated in the control room.
3.2.2 Pumps and Discharge Cross-Connect AFW Trsin A pump, P-318, is a combination turbine-driven motor-driven pump with both the turbine and electric motor on a common shaft. Either motive source can drive the pump at its rated capacity of 840 gpm at 1150 psig with a normal recirculation flow of 60 gpm. The turbine driver is used as the primary motive source for this pump. The motor driver can be manually initiated.
AFW Train B pump, P-319, is a motor-driven pump which has the same rated capacity and recirculation flow as the Train A pump.
The pumps discharge through check valves and locked open manual valves into 6-inch cross-connected discharge lines. The cross-connection line contains two normally-open motor-operated valves (HV-31826 and HV-31827). This cross-connect permits either pump to feed either or both steam generators.
3.2.3 Auxiliary Feedwater Flow Control Valves The flow of AFW to each steam generator is contrc. ed by normally closed pneumatically operated control valves (FV-2c327, FV-20528, FV-X1, and FV-X2) in parallel paths.
Initiation and control instrumentation for these valves is described in Section 3.4 of this report.
3.2.4 Auxiliary Feedwater Isolation Valves Each steam generator can be isolated from AFW flow by normally-open motor-ope rated valves (FV-20577, FV-20578, FV-X3, and FV-X4).
These valves are located in the parallel lines downstream of the AFW con-trol valves.
Laitiation and control instrumentation for these valves is described in Section 3.4 of this report.
- 3. 2. 5 Recirculation and Test Lines Recirculation and test lines are connented to the discharge piping of both pumps. Recirculation for pump protection la accomplished with normally open flow paths to the condensers consisting of small lines with check valves, restricting orifices, and lucked-open manual valves.
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o BWNP-20007 (6-76)
BASCOCK & WILCox MUCLEAA Powse GENERATON DIVISCN III TECHNICAL DOCUMENT 15-1120580- m Full flow test capability is provided through a 6-inch line which intersects the AW system cross-connect between the two normally open motor-operated valves HV-31826 and HV-31827. This full flow test path is isolated from the cross-connect during normal operation by a nonna11y closed pneumatically operated control valve (WS-X5).
Either AW train can be full-flow tested by opening valve WS-X5 and s tarting the appropriate AW pump. The full capability of both AW trains to supply AW on demand is maintained during the test since either a channel A or B AW initiation signal will result in automatic closure of valve WS-X5 through its fail closed on loss-of-air design. The AW systs is, therefore, automatically restored to its normal configuration.
3.2.6 Steam Supply for the AFWS Turbina Steam supply for the AW pumo P-318 turbine is obtained from both steam generators through six-inch lines containing check valves MSS-051 and MSS-052, locked-open manual valves MSS-049 and MSS-050, and normallympen motor operated valves HV-20569 and HV-20596. The check valve and motor operated valve provide redundant isolation capability to preclude blowing down the good steam generator in the event of steam line or feed line break.
Downstream of these valves the lines join to form a common supply to the pump turbine.
Upstream of the turbine is a normally closed DC motor operated valve FV-3 0801. A description of the controls for this valve is contained in Section 3.4 Turbine exhaust is vented to the atmosphere.
i i
3.2.7 Key Valve Positions Direct position indication (e.g., valve sten position) is required i
on all automatically operated and remote manual power operated valves. To comply with this requirement, the following valves require position indication:
FV-20527 WS-X5 FV-20528 FV-30801 FV-20577 HV-20569 FV-20578 HV-20596 FV-X1 HV-31826 FV-X2 HV-31827 FV-X3 FV-X4 DATE:
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7-24-80 17 t
BWNP-20007 (6-76)
BABCOCK & NVILCOX Munne NUCLEMI Powl0 otNGRAT!oM DIVI $loN TECHNICAL DOCUMENT 15-1120580 01 3.3 Supp<rting Systems The AFW pumps, pump motor and turbine are self-contained entities without dependencies on secondary support systems. The bearings on the turbine and both pumps are lubricated by slinging oil from reservoirs near the bearings. Lube oil cooling is accomplished by heat transfer to the pumped fluid.
3.3.1 Power The two AFW trains are powered from diverse power sources. AFW prmp P-318 is turbine driven with an AC power back-up motor, and AFW pump P-319 is AC power motor driven with back-up power from the diesel generator. The following valves required to operate the AFW system are also on AC power with back-up power from the diesel generator:
FV-20577, FV-20578, Fv-X3, FV-X4, KV-20569, EV-20596, HV-31826, HV-31827, and FWS-X5.
In the highly unlikely event of a loss of all AC power, the turbine driven pump AFW train derives its power from the steam generators for the pump and from a battery-backed DC buss for its steam supply valve. Valve FV-30801 requires battery backed DC power.
3.3.2 Service Air AFW flow control valves FV-20527, FV-20528, FV-X1 and FV-X2 are con-eccted to the qualified redundant air supply system with redundant valves in the same train being connected to a different air supply system.
3.4 Instrumentation Description It should be noted that all setpoints and values used in the follow-(01) l ing discussion are approximate and are given for purposes of illustration.
l The emergency feed initiation and control system (EFIC) is an instrumentation system designed to provide the following:
1.
Initiation of auxiliary feedwater ( AFW),
2.
Control of AFU at appropriate setpoints (approx. 3, 20 and 31.5 (01) feet),
l l
3.
Level rate control when required to inimize overcooling, 4
Isolation of the main steam and main feedwater lines of a de-pressurized steam generator, 5.
The selection of the appropriate steam generator (s) under i
conditions of steasline break or sain feedwater or emergency l
feedwater line break downstream of the last check valve, DAT :
2-20-81 pAGE 18
BWNP-20007 (6-76)
SABCOC3C & NVILCOX Nuct:4a rowen o!NERAT oN 000$ oN TECHNICAL DOCUMENT 15-1120580-01 6.
Termination of main feedwater to a steam generator on approach to overfill conditions, 7.
Termination of AFW to a steam generator on approach to overfill conditions, and 8.
Control of atmospheric dump valves to predetermined setpoint.
The emergency feed initiation and control system (EFIC) is illus-trated in Figures 3.4-1 thru 3.4-9.
Figure 3.4-1 illustrates the (01)
EFIC organization while the remaining figures illustrate the individual logics that comprise the system. The interface of the EFIC with the secondary plant is illustrated in Figure 3.1-1.
The EFIC - see Figure 3.4 consists of four channels (A,5,C, & D).
Each of the four channels are provided with input, initiate, and vector logics. Channels A and B also contain trip logics and control logics.
Each channel monitors inputs by means of the input logic, ascertains whether action should be initiated by means of the initiate logic and cetermines whien SGs should be fed by means of the vector logic.
Channels A and B monitor initiate signals from each of the four initiate logics by means of the trip logics to transmit trip signals when required. Channels A and B also exercise control of emergency feedwater flow to the SG by means of control logics to maintain SG level at prescribed values once AFW has been initiated.
In addition, Channels A and B also monitor SG A and B overfill signals originating in the Channel A, B, C and D initiate logics. By means of trip logics, Channel, A and B terminate main feedwater to a steam generator that is approaching overfill.
3.4.1 Input Logic The input logic, depicted in Figures 3.4-3a, 3.4-3b, and 3.4-3c is (01) located in each of the channels. The input logic:
1.
Receives analog input signals, (01) 2.
Provides input buffering as required, 3.
Compares analog signals to appropriate setpoints to develop digital signals based on analog values, 4
Provides for the injection of test stimuli, DATE:
2-20-81 pAGE 19
BWNP-20007 (6-76)
BABCOC/ & NVILCOX NUCLEA4 PoWim MINEAAflON DWI$loN TECHNICAL DOCUMENT 15-11205a0 01 5.
Provides buf fered Class 1E signals and isolated non-1E signals, and 6.
Provide signals to the remaining channel logic.
(01) 3.4.2 Initiate Logic The initiate logic, depicted in Figure 3.4-4 is lecated in each channel. The initiate logic derives its inputs from the input logic and provides signals which result in the issuance of trip signals via the trip logics in Channels A and B.
The initiate logic issues a call for AFW trip (to the trip logic) aen:
1.
All four RC pumps are tripped.
2.
Both main feedwater pumps are tripped (i.e. low discharge pressure).
3: The level of either steam generator is low.
4.
Either steam generator pressure is low.
5.
Deleted.
(01) 6.
Flux to MFW flow ratio trip is present.
(01)
Other functions of the initiate logic are:
3.
Signal approach to SG A overfill when SG A level exceeds a high level setpoint.
4.
Signal approach to SG B overfill when SG B level exceeds a high level setpoint.
5.
Provide for manually initiated individual shutdown bypassing of RC pumps, main feedwater pumps, and SG pressure initiation of AFW as a function of permissive conditions. The bypass (es) are automatically removed when the permissive condition terminates.
6.
Provida for saintenance bypassing of an EFIC initiate logic.
DATE:
2-20-81 pAGE 20
BWNP-20007 (6-76)
BABCOCK & WILCOX
(
wet:4a powee cesananon on, vow TECHNICAL DOCUMENT 15-1120580-01 3.4.3 Trip Logic The trip logic is illustrated in Figure 3.4-5.
The trip logic of the EFIC employs a 2(1-out-of-2) format. This format provides for easy one step testing from input logic test switches to the initiated con-trollers. Testing is facilitated by locating the AND portion of the 2(1-out-of-2) logic in the controller. A characteristic of coincidence logic systems is that a test stimulii inserted at the input propagates to the first AND element of the system and no further. Since the first AND element of the EFIC is in the controller, test stimuli inserted at the input logic will be propagated to each controller. EFIC testing philosophy is discussed in Section 3.4.6.
The trip logic is provided with five 2(1-out-of-2) trip networks.
These networks monitor the appropriate outputs of the initiate logics in each of the channels and output signals for tripping:
1.
2.
SG A main steamline isolation.
3.
SG B main steamline isolation.
4.
SG A main feedwater isolation.
5.
SG B main feedwater isolation.
It should be noted, for the later discussion of the vector logic, that the trip logic cutputs a signal when a 2(1-out-of-2) trip of AFW occurs. Also, note the presence of the vector enable switch.
It should also be noted that the EFW trip logics are input by the (01)
Emergency Safety Features Actuatien System (ESFAS) Emergency Core Cooling trip signals to assure that EFW is initiated coincident with Emergency Core Cooling actuation.
Refer to Figure 3.4 trip logics are contained in Channels A and 3 only per the two train AFW system.
For each trip function, the trip logic is provided with two manual trip switches. This affords the operator with a means of manually tripping a selected function by depressi'g both switches. The use of two trip switches allows for testing the trip switches and also reduces the possibility of accidental manual initiation.
Once a trip of the trip bus occurs, the trip is latched. A manual reset switch is provided for breakdown of the latch. Once a trip 1
DATE:
2-20-81 PAGE 21
BWNP-20007 (6-76) 8A8 COCK & WILCOX NUCLEAR Powet GENERAfloM DIVl$loN TECHNICAL DOCUMENT 15-1120580-o0 occurs, the trip can only be removed by manual reset action following return of the initiating parameter to an untrip value except as described in the next paragraph.
So that the operator may resume manual control of EFIC initiated devices following a trip, each trip logic is provided with a manual pushbut ton.
Operation of the manual pushbutton:
1.
Will have no ef fect on the trip logic so long as a trip condition does not exist.
2.
Will remove the trip from the trip bus only so long as the switch is depressed in the case of a one half trip (either bus but not both tripped).
This allows for testing the manual function.
3.
Will remove the trip from both busses so long as a full trip (both busses are tripped) exists. This is accomplished by means of latching logic.
Institution of the manual function also breaks the trip latches so that, if the initiating stimuli clears, the trip logic will revert to the automatic crip mode in preparation for tripping if a parameter returns to the trip reg io n.
Trip signals are transmitted out of the EFIC by activating a relay thereby gating power onto trip busses.
In this manner, the EFIC provides power to energize the control relays whose contacts form the AND gates in the controllers.
3.4.4 vector Logic The vector logic - Figure 3.4 appears in each of the EFIC channels - Figure 3.4-i.
The vector logic monitors:
1.
SG pressure signals, 2.
SG (A and 3) overfill signals, and 3.
AFW trip signals (vector enable) originating in Channel A and B trip logics.
The vector logic developes signals for open/close control of steam generator A and B auxiliary feedwater valves.
DATE:
PAGE y
~~
7-24-80
. ~
n' BWNP-20007 (6-76)
BA8 COOK & WILCOX Numen NUCLEA4 Powlt o&NERATioN DIVl$loN TECHNICAL DOCUMENT 15-1120580-01 The vector logic outputs are in a neutral state until enabled by trip signals (vector enaole) from the channel A or B trip logf es.
Once enabled, the vector logic will issue close commands to tre valves associated with any SG for which an overfill signal exists.
Note that the AFW overfill limit may be manually bypassed. Manual (01) bypass can only be initiated under permissive condi?. ions of AFW trips in Channel A and/or Channel B.
When enabled and with no overfill signals present, the valve open/close commands are determired by the relative values of steam generator pressures as follow.:
SG A Valve SG B Valve Pressure Status Command Command SG A & B > Setpoint Open Open SG A > Setpoint & SG B < Setpoint Open Close SG A < Setpoint & SG B > Setpoint Close Open SG A < Setpoint & SG B < Setpcint and SG A & B within 150 Open Open SG A 150 psi > SG B Open Close SG B 150 psi < SG A Close Open 3.4.5 Control Logic The control logic is depicted in Figure 3.4-2.
(01)
For each SG (A and B) there are two controls which are selected by (01) transfers T1 and T5 respectively. The three foot level setpoint control is automatically selected when an AFW trip occurs with one or more reactor coolant pumps operating. A level rate control with a twenty foot setpoint is selected when an AFW trip occurs with no reactor coolant pumps operating. The three foot level control requires no explanation. However, the rate control is more involved.
The characteristics of the rate limited follower are important in the following discussions. As the level signal changes, the rate output of the follower will follow it exactly so long as the rate of change does not exceed the predetermined rate limit values. The rate limit values given (4 inches per minute for increasing level DATE:
2-20-81 PAGE 23
BWNP-20007 (6-76) 8ABCOCK & WILCOX wycteAa powen GENERAfloM olvt$loN sucen TECHNICAL DOCUMENT 15-1120s80-01 rates and 200 inches per minute for decreasing level rates) are ap-proximate for purposes of illustration. (The rate limit for (01) increasing lavels is variable between 2 to 4 4.nches per minute as a function of steam generator pressure.) If level rate is increasing at greater than four inches per minute, the output of the rate limit-ed follower will increase at four inches per ainute. Once the rate of increase decreases to four inches per minute or less the output rate of increase will follow the input rate of increase. The func-tion is similar for decreasing level except that the rate limit is approximately 200 inches per minute. A side benefit of the rate limited follower is attentuation of noise whose ef fective rate is in excess of four inches per minute or 200 inches per minute respectively.
Reference Figure 3.4 with no RC pumps operating the twenty foot (01) setpoint will be selected and applied to one input of the low selec-tor.
As SG level falls, the output of the rate limited follower will lag actual level by twelve inches (twelve inch bias added to the level signal in the summer). When the rate limited signal (level plas twelve inches) becomes less than twenty feet, the rate lialter signal will appear at the subtractor (delta). The uutput of the subtractor will be approximately a negative one foot level error signal which will start opening the control valve ever wider thru the proportional plus integral. The increasing D.ow should halt the drop in level and ultimately start the level to increase toward the se t point.
If the level increase is more rapid than four inches per minute, the error signal out of the subtractor will decrease. This is due to the fact that the direct level input to the subtractor is not rate limited while the rate limited signal is.
This action will control the control valve so that the rate of approach to the setpoint does not exceed four inches per minute.
When level exceeds nineteen feet, the low selector will lock the twenty foot setpoint into the subtractor. During the last foot of level increase the error output of the subtractor will gradually reduce.
Transfer T4 is provided for future use by the user.
It allows for (01) l selection of hand control from either the main control room or tb e auxiliary shutdown panel.
See Figure 3.4 transfer logics T3 and T7 allow for selection of a (01) manually inserted setpoint (illustrated as a thirty foot setpoint).
The logic is arranged so that manual may be selected before and after an AFW trip. However, the twenty foot setpoint will automatically be selected on the occurrence of an AFW trip.
See Figure 3.4 transfer logics T2 and T6 allow for selection of (01) hand control of emergency feedwater control velves before and af ter an AFW trip. However, automatic operation will automatically be selected on the occurrence of an AFW trip.
1 DATE:
2-20-81 PAGE 24
BWNP-20007 (6-76)
BABCOCK & WILCOX un,,
NUCitAt Powie GENitATCN otVISCN TECHNICAL DOCUMENT 15-1120580-01 In addition, EFIC Channel A is provided with a pressure control loop for the steam generator A atmospheric dump valves. EFIC Channel B is provided with a pressure control loop for the steam generator 3 atmospheric dump valves. Transfer T8 describes provisions for future (01) transfer of ADV control to a location outside the main control room.
The steam generator atmospheric dump valve control logic requires no (01) explanation.
3.4.6 EFIC Trip Testing Figure 3.4-7 illustrates the trip philosophy of the EFIC in simpli-fled form for one EFIC trip function (e.g., AFW trip). For purposes of the following discussion, the test pushbuttons associated with each bistable is capable of forcing the bistable input into the trip region. The bistables employ a low dead band so the bistable will reset once the pushbutton is released.
Complete trip r.esting (input to controllers) may be initiated from the input logic in each of the channels. Depressing the pushbutton in Channel A will trip the Channel A bistable and:
1.
The Channel A initiate logic will transmit initiate signals to both the Channel A and B trip logics.
2.
The Channel A and 3 trip logics will half trip (trip one of the two trip busses).
3.
The Channel A and B trip logics will latch in the half trip.
The half trip will be retained after reset of the bistable. This tests the latching circuit.
4.
Each controller receiving the half trip will acknowledge the half trip by transmitting a test confirmation signal assuming all controllers are functioning properly.
5.
A f't11 complement of test confirm signals will satisfy the AND gate in both Channel A and B.
The result is that the confirm lamps will indicate test success.
6.
The trip logic reset switches can now be depresed to reset the half trip.
The confirm lamp should go out.
7.
If some but not all controllers were to respond due to a malfunction, the confirm lamp will flash.
(Off normal may be indicated by some means other than flashing in the final design.)
~
CATE:
2-20-81 PAGE 25
~
BWNP-20007 (6-76) 8ABCOCK & WILCOX NUCLEAL Powet oC'*eATCH olvt33CN TECHNICAL. 000UMENT ts-t12o580-00 8.
The foregoing tests may be conducted from each channel in turn to test the ability to transmit trips from all channels.
9.
The foregoing tests may be conducted for all trip functions from all channels for complete trip testing.
10.
Tests as described above may also be conducted by use of the local and remote manual trip and reset switches.
NOTE: The utilization one out-of-two taken twice logic allows for the foregoing test philosophy while minimizing the proba-bility of inadvertent initia tion.
3.4.7 EFIC Signal Application Figure 3.1-1 illustrates the application of EFIC signals to a simplified auxiliary feeedwater system. Salient features of the arrangement are:
1.
The channel A AW trip signal starts the electric emergency feedwater pump. Both the Channel A and B trip logics admit steam to the turbine powered auxiliary feedwater pump. With this arrangement, at least one pump will be started with a single failure of the A or 3 trip logics.
Also, given a failure of channel A, B, C, or D initiate logics, both pumps will be started due to the 2(1-out-of-2) character of the trip logic. The cross-connect between the discharges of the two auxiliary feedwater pumps allows either pump to supply feedwater to both SGs.
2.
If the cause of the AW trip is low SG pressure in SG A, AW will be tripped as in 1 above.
In addition, the trip logics in channels A and B will issue SG A main steamline and main feedwater isolation trip signals. The channel A and B trip logics will redundantly isolate SG A main feedwater. With the occurrence of low pressure in SG A, main feedwater to that generator will be tir.ninated in the presence of a single failure.
3.
Isolation of SG B main steam and main feedwater lines occurs in the same way as described in 2 above for SG A except that the channel A and B SG B main feedwater and main steanline trip logies are employed.
DATE:
pAGE 26 7-2 '.-80
BWNP-20007 (6-76)
BABCOCK & WILCOX
..t.
NUCLEAR Powet GtNIGADoM DIVl$lON TECHNICAL DOCUMENT 15-112o5a0-01 4.
Given the condition where both SG pressures are low, the events described in both 2 and 3 above will occur.
5.
7 se auxiliary feedwater path to each SG consists of parallel control valves and parallel isolation valves. This allows feeding when required in the presence of a single valve failure.
It also allows closure of the flow path when required in the presence of a single failure. Since each of the four valves receives vector close signals from different channe:.s the path will be closed when required by the vector logics in the presence of the failure of a single vector logic.
In the open direction, the isolation valves receive open vector commands, from channels C and D, when feeding of tbc SG is required. The control valves, under these conditions will open as dictated by the control logics in channels A and B.
In this way, a generator will be fed when required in the presence of a failure of channel A, B, C, or D.
3.4.8 OTSG Level Sensing Figure 3.4-8 contains the proposed arrangement for OTSG level sensing. The acceptability of this design will depend on the accuracy of the measurement. This accuracy will be determined in the detailed design.
To provide for low level control and initiation signals for the auxiliary feedwater, four differential pressure transmitters (dP transmitters) will be added. The sensing lines for these transmitters will be connected between the unused existing level sensing connetions located 251 inches above the datum 1.ue of the OSTG (277' above the face of the tube sheet) and the drain line connections located 7-1/2" below the ; ace of the tube sheet.
To provide high level control and overfill protection signals, four dP transmitters will be added. The upper sensing connections will be saa1 folded with the upper sensing line of the existing operating range level tra?.smitters. The lower sensor connections will be sanifolded with the lower sensing line of the existing operating (01) range level transmitters.
There are four drain line connections (located approx. 7-1/2" below the face of the tube sheet) which can be used for the lower sensing lines of the low level dP.
These will be manifolded as necessary to (01) best serve the redundancy requirements.
DATE:
2-20-31 PAGE 27
3WNP-20007 (6-76)
BABCOCK & WILCOX Nu-s ee pouctEAR Powlt GENERATION OlVI$loN TECHNICAL DOCUMENT 15-1120580-01 3.4.9 Interface with Valve and Pump Controllers All valve and pump controllers shall be designed such that signals from the EFIC system will override any other control signals. Also, when an EFIC signal is removed, the controller design shall be such that valves (other than the AFW control valve) will not change position and pumps will not change state without a specific manual command. When the vector logic close command to the AFW control valve is removed, the control valve shall be positioned as required by the AFW control system or the manual control as selected.
4.0 SYSTEM LIMITS, PRECAUTIONS AND SETPOINTS 4.1 Limits and Precautions 4.1.1 AFW Flow Limits Maximum allowable flow -
1650 gpm/SG Minimum allowable flow -
760 gpm (01) 4.1.2 Deleted.
(01) 4.1.3 AFW Pump Suction Pressure P-318 minimum NPSH -
18 feet at 840 gpm (01)
P-319 minimum NPSH -
18 feet at 840 gpm (01) 4.1.4 System Limits (Design)
Pressure -
1600 psig Temperature -
316 F 4.1.5 Minimum Pump Recirculation P-318 -
60 gpm
(
P-319 -
60 apa DATE:
2-20-a1 PAGE 28
)
e BWNP-20007 (6-76)
BABCOCK & WILCOX I
NUCLEAa Powlt GENetMioN OlVl$loN TECHNICAL DOCUMENT 15-1120580-01 4.2 Setpoints (01)
All setpoints given in this section and defined as " nominal" are instrument calibration points.
Instrument string errors as defined in Appendix B were used in the analyses to determine tne conservative maximum and sinimum setpoint values. The maximum and minimum setpoints represent the earliest and latest assumed actuation point for use in analysis.
For the purpose of this discussion, " Level" refers to the equivalent height of a saturated water column (1065 psia) referenced from the top of the lower tube sheet.
The flux to feedwater ratio setpoint is shown on Figure 3.3-2.
This setpoint was developed as an anticipatory trip for loss of feedwater events. The equation used for this setpoint and the errors and delay times are shown in Appendix C.
This trip function is located in the RPS and is used to trip the reactor. An output from the RPS will feed the EFIC to initiate EFW.
4.2.1 Low SG Level EFW Initiate Setpoint (01)
This is a protective setpoint designed to initiate EFW flow to a steam generator following loss of sain feedwater flow. The low range level instrumentation is used to monitor low level in the steam generators. For setpoints see Table 4.2-1.
4.2.2 EFW Control Level (3 Foot Level) Setpoint (01)
This is a level control setpoint designed to be automatically selected following initiation of EFW if one or more reactor coolant pumps are providing forced circulation. The low range level instru-mentation is used to monitor steam generator level at this point and to provide signals to the EFIC control system. For setpoints see Table 4.2-1.
4.2.3 Natural Circulation control Level (20 Foot Level) Setpoint (01)
This is a level control setpoint designed to be automatically selected following initiation of EFW if all four reactor coolant pumps are tripped. For 177 FA plants, 20 feet of steam generator level provides a thermal center in the steam generator at a higher elevation than that of the reactor. Controlling steam generator level at a ninimum level of 20 feet insures natural circulation of the reactor coolant system fluid. The full range level instrumer.tation is used to monitor steam generator level at this point and to provide signals to the EFIC control system. For setpoints see Table 4.2-1.
DATE:
2-20-31 PAGE 29
BWNP-20007 (6-76)
BA8 COCK & WILCOX NUCLEAA Powlt GENERAflopi olVl31oM TECHNICAL DOCUMENT 15-1120580-01 4.2.4 Steam Generator Overfill Setpoint (01)
This is a protective setpoint designed to automatically terminate AFW or main feedwater flow to a steam generator. This setpoint is required to prevent steam generator level from increasing to a level at which feedwater would flow into the main steam lines. This setpoint can be manually bypassed to allow the setpoint described in Section 4.2.5 to be reached. The high range level instrumentation is used to monitor steam generator level at this point. For setpoints see Table 4.2-1.
4.2.5 CCCS Fill Limit Setpoint (31.5 Feet Level)
(01)
This is a level control setpoint designed to be manually selected following a LOC A.
This setpoint will establish a steam generator feedwater level which will support steam condensation natural circulation. To preclude terminating feedwater flow before this setpoint is reached, the steam generator overfill setpoint described ie. Section 4.2.4 must be manually bypassed. The full range level instrusentation is used to monitor steam generator level in this region.
For setpoints see Table 4.2-1.
4.2.6 Low Steam Generator Pressure Setpoint (01)
This is a pressure setpoint designed to automatically isolate the main steam lines and main feedwater lines to the affected steam generator. This setpoint will isolate the steam generator only if one steam generator is affected. The other steam generator will not be isolated.
If both steam generators are below this setpoint the EFIC system will deteroine which steam generator to supply and which to isolate. Pressure irstrumentation string requirements are given in Appendix B.
For setpoints see Table 4.2-1.
4.2.7 Steam Generator Differential Pressure Setpoint (01)
This is a pressure setpoint designed to automatically determine, by comparing the difference in steam generator pressures, which steam generator is to be isolated and which steam generator is to be fed.
Pressure instrumentation string requirements are given in Appendix 3.
For setpoiots see Table 4.2-1.
4.2.8 Atmospheric Dump Valve Operating Setpoint (01)
This is a pressure setpoint designed to automatically open the atmospheric dump valves to relieve steam generator pressure. This i
setpoint is lower than the steam generator relief valve lift point l
and will therefore decrease the frequency of challenges to the relief valves. The control system provides the operator with the capability to manually override this setpoint.
Pressure instrumentation string l
requirements are given in Appendix B.
For setpoints see Table 4.2-1.
1 DATE:
2-20-31 pAGE 30 l
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DATE:
2-21-31 FWE 33 iER I A:. :
13. _ <' 5 3 1 -. ).
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't---
o' BWNP-20007 (6-76)
BABCOCK & WILCOX NUCLEAR PCwit GENtaATION OtYlSION TECHNICAL DOCUMENT 15-1120580-01 TABLE 4.2-1 (01)
EFW SYSTEM SETPOINTS NOMINAL MAXIMUM SETPOINT MINIMLH SETPOINT SETPOINT SETPlin NORMAL ACC IDENT NORMAL ACC IDENT Low SG Level Initiate 6"
12" N/A
> 0" N/A EFW Control Level **
36" 42" N/A 30" N/A Natural Circulation Control Level 240" 248" 255" 232" 204" Steam Generator overfill
- 374" 380" N/A 368" N/A ECCS Fill Limit 379" N/A 394" N/A 343" Low Steam Generator 600 625 S75 Prescure psig psig
- N/A psig
- N/A Stcas Generator 150 200 100 Differential Pressure psig psig
- N/A psig
- N/A 1020 1045 995 ADV Opening Setpoint psig psig
- N/A psig
- N/A
- It is reco= mended that this setpoint be established at 15" above the no?al operating range of the steam generator. This setpoint was used in the.tnalyses establishing setpoints for this document.
- Accident environments were not considered for this measurement since current procedures require the RC pumps to be tripped on a low pressure ESFAS.
If this procedure is changed, these setpoints should be reevaluated.
- It is assumed that the steam generator pressure measurements will be located outside the reactor building and therefore accident environment errors do not apply.
DATE:
2-20-81 PAGE 34
BWNP-20007 (6-76)
BANCOCK & WILCOX NUCLtAt Powlt GINitATION OlVf$lCN TECHNICAL DOCUMENT 15-1120580-01 TABULATION OF DRAWING NL3BERS VS. FIGURE NLHBERS FOR RA CHO SECO AFW SYSTEM FIGURE NUMBER B&W DRAWING NLHBER 3.1-1 1121327D 3.4-1 1122924F (01) 3.4-2 1122923D (01) 3.4-3a 1122930C (01) 3.4-3b 1122928C (01) 3.4-3c 1122927C (01) 3.4-4 1122922D (01) 3.4-5 1122926E (01) 3.4-6 1122925C (01) 3.4-7 1122921C (01) 3.4-8 1121322C 3.4-9 11229208 (01) l l
l l
i i
I DATE:
2-20-81 APPENDIX A, PAGE A-1
BWNP-20007 (6-76)
SABCOCK & WILCOX NUCLEAR Powlt GENEEAfloM olVf$loN Numeen TECHNICAL DOCUMENT 15-It20580-01 INSTRUMENTATION REQUIREMENTS (01)
J j
1.
Low Range Level Instrument String a.
Span 150" b.
Scale 0-150" c.
Pressure -
-- 1200 psig d.
Temperature 6000F e.
Instrument String Errors:
- e.1 Normal Operating Environment
- 6"
- e.2 Small LOCA Environment -
- 15 "
e.3 Reference Leg Heatup
-20" See Note 1.
2.
High Range Level Instrument Strg a.
Span
--- 3 0 0 "
b.
Scale
- - - 100" - 400" c.
Pressure --
- 1200 psig d.
Temperature
- 6000F e.
Instrument String Error:
- e.1 Normal Operating Environment -
--- *6"
- e.2 Small LOCA Environment ---
-- *15 "
e.3 Reference Leg Heatup
--- -21" See Note 1.
3.
Full Range Level Instrument String a.
Span ---
401.5" l
b.
Scale
400" c.
Pressure -
1200 psig d.
Temperature
6000 F i
e.
Instrument String Error:
- e.1 Normal Operating Environment t8"
- e.2 Small LOCA Environment ----
-- *15 "
e.3 Ref e rence Leg Heatup ----------
-21" See Note 1.
DATE:
2-20-81 APPENDIX 3, PAGE 3-1 1
l
BWNP-20007 (6-76)
BA8 COCK & WILCOX NUCLEAa Powla otNt9ATION OlVissON TECHNICAL DOCUMENT 15-1120580-01 4.
Pressure Instrument Strings a.
Span
-- 0-1200 psig b.
Response Time -
1 Second c.
Instrument String Error:
- e.1 Normal Operating Environment ----- *25 psi t
- Normal Operating Environment
--- 800F to 1400F / 100% RH
- Small LOC A Environment -
800F to 2500F / 100% RH NOTES:
I 1.
Level measurement to be density / pressure compensated over a pressure range of atmospheric to 1050 psig assuming a saturated volume of steam and water.
Since the level measurement is density compensated, the unit
" inches" refers to the actual level in the steam generator over the specified pressure range.
DATE:
2-20-81 APPENDIX B, PAGE B-2 l
t
g...., -
BWNP-20007 (6-76)
BABCOCK & WILCOX NVQtAR PCwit GENitAfiCN OlVl13ON TECHNICAL DOCUMENT 15-1120580-0t FLUX /FEEDWATER SETPOINT (01)
The following is the equation for the nominal setpoint used in Figure 4.2-1:
1.9 WFW - 21
=
Where: WFW = Feedwater flowrate in % secondary flow
$ = Neutron flux measured in % full power The errors and delay time used in developing this setpoint are:
Flow measurement error = 5.5%
Flux measurement error = 6%
Delay time = 2 seconds.
DATE:
2-20-81 APPENDIX C, PAGE C-1