ML20246E714
| ML20246E714 | |
| Person / Time | |
|---|---|
| Site: | Comanche Peak  |
| Issue date: | 08/23/1989 |
| From: | William Cahill, Walker R TEXAS UTILITIES ELECTRIC CO. (TU ELECTRIC) |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| TXX-89614, NUDOCS 8908290269 | |
| Download: ML20246E714 (70) | |
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File # 10010, 20014 C
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915.2 (clo)
Ref. # 10CFR50.34(b)
' William J CahiR,Jr.
Eaecutive Vice l' resident August 23, 1989 u
l 1
U. S. Nuclear' Regulatory Commission Attn: Document Control Desk i
Washington, D. C. -20555 l
SUBJECT:
COMANCHE PEAK STEAM ELECTRIC STATION (CPSES)
DOCKET N05. 50-445 AND 50-446 ADVANCE FSAR AND TECHNICAL SPECIFICATION SUBMITTAL IMPLEMENTATION OF P-9 PERMISSIVE:
REACTOR TRIP DN TURBINE TRIP BELOW 50% POWER 1
'l Gentlemen:
The enclosures' to this letter provide an advance submittal of FSAR and
)
-Technical Specification changes related to the Westinghouse P-9 permissive l
L modification along with related supporting documentation. These changes will be included in a future FSAR amendment and should also be reflected in the i
CPSES Unit 1 Technical Specifications when issued.
To facilitate NRC staff review of these changes, information related to the l
FSAR change (Enclosure 1) is organized as follows:
1.
Draft revised FSAR pages with changed portions indicated by a bar in the margin (denoted as " draft"), as they are to appear'in a future amendment (additional pages immediately preceding and/or following j
the revised pages are provided if necessery to understand the change).
]
2.
A line-by-line description / justification of each revised FSAR item.
3.
A copy of related SER/SSER sections.
4 The b'.,ld/ overstrike version of the revised FSAR pages referenced by the description / justification for each item identified above. The qq bold / overstrike version facilitates review of the revisions by highlighting each addition of new teYt in bold type font and overstriking with a slash (/) the portion of the text that is deleted.
In some cases, where the bold overstrike version is unavailable, a hand marked-up version is provided.
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ADOCK 05000445
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PDC 400 hah Olive Street LB Bl Dallas, Texas 75201 I
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TXX-89614-y August' 23. 1989 Page 2 of 2 I
l Information related to the Technical Specification change (Enclosure 2) is-l organized as-follows:-
1.
. Draft revised Technical Specification pages, with changed portions indicated by a bar in the margin (denoted as "dratt").
2-A description / justification of the revised Technical Specification.
l 3
A hand marked-up version of the Final Draft of the-CPSES Unit 1 Technical Specifications as certified on April Id. 1989.
I
- TV Electric requests the NRC to perform an. expedited review of the.above'FSAR
.)
and Technical Specification changes and inform us of acceptability.
j Sincerely, I
l Y
n'h' h
William J. Cahill, Jr.
I i
0 By:
M d
7 RogeVD. Walker Manager. Nuclear Licensing VPC/smp Enclosures c - Mr. R. D. Martin, Region IV l
Resident Inspectors. CPSES (3) 1 b
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]
I 1
Enclosure'1'to TXX-89614 August-23. 1989 Page 1 of 48 Advance.FSAR Change Related to Implementation of P 9 Permissive Reactor. Trip on Turbine Trip'Below 50% Power and Supporting Documentation-l.-
Item l' Draft Revised FSAR Pages pp. 2 - 17f Item 2.
One-by-line Description / Justification pp. 18 - 22 for.each.fSAR Change item Item 3 Related SER/SSER Pages pp. 23 - 32 Item 4
. Bold / overstrike version of the pp. 33 - 48 l-FSAR pages.
l 1
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' Enclosure 1 to TXX-89614' CPSES/FSAR'
.: Page 2 of 48:
relay rack. The protection relay rack is divided into four
- separate compartments, similar to the Solid State Protection System input cabinet described in Section 7.
2.3 to maintain I
channel' independence. Potential transformers from which the input signals for undervoltage and underfrequency relays are
' derived and.the protection relay rack are classified as Class 1E and are located in a Seismic Category I structure.
5.
Steam generator trip / Low-low steam generator water level trip l13 1
The specific trip function generated is the low-low steam 13 generator water level trip. This trip. protects the reactor from loss of heat sink in the event of a loss of feedwater to the steam generators. This trip is actuated on two out of four low-low water level signals occurring in any steam generator..
The logic is shown on Figure 7.2-1 Sheet 7.
l 13 6.
Reactor trip on a turbine trip (anticipatory) (BOP Scope) 13 The reactor trip on a turbine trip is actuated by two out of three logic from emergency trip fluid pressure signals or by all closed signals from the turbine steam stop valves. A turbine DRAFT trip causes a direct reactor trip above P-9.
The reactor trip on turbine trip provides additional protection and conservatism beyond that required for the health and safety of the public.
This trip is included as part of good engineering practice and prudent design. No cr?.d:t is taken in any of the safety analyses (Chapter 15) for this trip.
The turbine provides anticipatory trips to the Reactor Protection i
System from contacts which change position when the turbine stop 7.2-13 Draft Version t
h
c Enclesure'1,te TXX-89614 I4'
' Pag 2'3 of 48 CPSES/FSAR The intermediate range level trip and power range'(low setpoint)
~
' trip can only be blocked after satisfactory operation and permissive information are obtained from two out of four power range channels.
Four individual blocking switches are provided so that the low range power range trip and. intermediate range trip can be independently blocked (one switch for each train).
These trips are automatically reactivated when any three out of four power range channels are below the permissive (P-10) setpoint, thus ensuring automatic activation to more restrictive trip protection.
The development of permissives P-6 and P-10 is shown on Figure 7.2-1 Sheet 4.
All of the'permissives are digital: they are derived from analog signals in the nuclear power range and intermediate range channels.
Refer to Table 7.2-2 for the list of Prctection System interlocks.
2.
Blocks of reactor trips at low power Interlock P-7 blocks a reactor trip at low power (below DRAFT approximately 10 percent of full power) on a low Yector coolant flow in more than one loop, reactor coolant pump undervoltage, reactor coolant pump underfrequency, pressurizer low pressure or pressurizer high water level.
Refer to Figure 7.2-1. Sheets 5 and 6 for permissive appli,ations. The low power signal is derived from three out of four power range neutron flux signals below the setpoint in coincidence with two out of two turbine impulse chamber pressure signals below the setpoint (low plant load). See Figure 7.2-1, Sheets 4 and 16. for the derivation of P-7.
l 7.2-17 Draft Version i
?
Enclosure I to TXX-89614 CPSES/FSAR Page 4 of 48 uRAFT The P-8 interlock blocks a reacter trip when the plant is beloc approximately 48 percent of full power, on a low reactor coolant flow in any one loop. The block action (absence of the P-8 interlock signal) occurs when three out of four neutron flux power range signals are below the setpoint. Thus, below the P-8 setpoint,-the reactor will be allowed to operate with one inactive loop and trip will not occur until two loops are indicating low flow.
Refer to Figure 7.2-1, Sheet 4. for derivation of P-8, and Sheet 5 for applicable logic.
DRAFT The P-9 interlock blocks a reactor trip following a turbine trip when the plant is below approximately 50% of full power. The plant is designed for 50% load rejection capability (40% steam dump and 10% control rods insertion). The absence of the P-9 interlock (three out of four power range flux signals below the setpoint) blocks the immediate reactor trip following a turbine trip.
Refer to Figure 7.2-1, Sh. 4 for derivation of P-9 and Sh.16 for applicable logic.
Refer to Table 7.2-2 for the list of Protection System blocks.
5 7.2.1.1.4 Coolant Temperature, N-16 and Power Range Neutron 5
Detector Sensor Arranget ent 0032.56 5
The bypass line T ot and Tcold measurements are, as a result of h
incorporation of the Comanche Peak Protection and Surveillance Upgrade Package, being replaced with an N-16 power monitor and an in-line Tcold measurement.
In addition, the present two-section power range neutron detectors are being replaced with four-section detectors (used for both protection and control functions).
0 5
7.2.1.1.4.1 In-Line Fast Response Thermowell Installation of Cold 5
Leg Temperature Measurement The fast response in-line Tcold measurement is provided by an RTD in 1
a thin wall thermowell installed in the reactor coolant cold leg Draft Version 7.2-18
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lCPSES/FSAR Enclosure' I to TXX-89614 TABLE 7.2-2
.PagG 6 of 48 (Sheet 2)
PROTECTION SYSTEN INTERLOCKS Desig-nation Derivation Function Defeats the block of inter-mediate range reactor trip.
and intermediate range rod.
stops (C-1)
Input to P 11 Blocks of Reactor Trios P-7 Absence of P-7: 3/4 neutron Blocks reactor trip on:
flux (power range) below set-Low reactor coolant flow point (from P-10) in more than one loop, and undervoltage, under-2/2 turbine impulse chamber frequency, pressurizer DRAFT l
pressure below setpoint low pressure, and pressu-(from P-13) rirer high level P-8 Absence of P-8: 3/4 neutron Blocks reactor trip on low flux (power range) below reactor coolant flow in a setpoint single loop P-9 Absence of P-9: 3/4 neutron Blocks reactor trip on DRAFT flux (power range) below tu,'bine trip below DRAFT i
setpoint approximately 50% power DRAFT i
P-13 2/2 turbine impulse chamber Input to P-7 pressure below setpoint 1
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Enclosurp 1 to TXX-89614.
CPSES/FSAR Page 11 of 48>
The automatic Steam Dump System.is able to accommodate this abnormal 1
L load rejection and to reduce the effects of the transient-imposed upon the RCS.
By bypassing main steam directly'to the condenser an DRAFT artificial load is thereby maintained on the primary system. The Rod Control System can then reduce the reactor temperature to a new -
equilibrium value without causing overtemperature and/or overpressure conditions. - The steam dump steam flow capacity is 40 percent of full load steam flow at full load steam pressure.
If the difference between the reference Tavg (Tref). based on turbine impulse cMmber pressure and the lead-lag compensated auctioneered Tp. eSteeds a pre-determined amount, and the-interlock mentioned below 't satisfied, a demand signal will actuate the steam dump to maintain the RCS temperature within control range until a new equilibrium condition is reached.
.To prevent actuation of steam dump on small load perturbations, an independent load rejection sensing circuit is provided. This circuit senses the rate of decrease in the turbine' load as detected by the turbine impulse chamber pressure.
It is provided to unb b:.k the dump valves when the rate of load rejection exceeds a preset value corres-ponding to a 10 percent step load decrease or a sustained ramp load decrease of 5 percent / minute.
A block diagram of the Steam Dump Control System is shown on Figure 7.7-8.
7.7.1.8.1 Load Rejection Steam Dump Controller This circuit prevents large increase in reactor coolant temperature following a large, sudden load decrease. The error signal is a difference between the lead-lag compensated auctioneered Tavg and the reference Tavg based on turbine impulse chamber pressure.
7.7-21 Draft Version t
__-___._.m__-
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' Enclosuro I to TXX-89614 0
Paga 12)g, yavg signal is the, same as that used in the RCS. The leed- %
compensation for the Tavg signal is to compensate for lags in the
('
plant therma 1' response and in valve positioning.
Following a sudden load decrease, Tref is immediately decreased and Tavg tends to increase, thus generating an immediate demand signal for steam dump.
Since control rods are available, in this situation steam dump
. terminates as the error comes within the maneuvering capability of the control rods.
DRAFT 7.7.1.8.2 Plant. Trip Steam Dump Controller DRAFT' Following a reactor trip, as monitored by the reactor trip signal DRAFT (P-4), the load rejection steam dump controller is defeated and the turbine trip steam dump controller becomes active.
Since control rods are not available in this situation, the demand signal is the error signal between the lead-lag compensated auctioneered Tavg and the no-load reference Tavg. When the error signal exceeds a pre-determined setpoint the dump valves are tripped open in a prescribed sequence. As the error signal reduces in magnitude (indicating that the RCS Tavg is being reduced tow &rd the reference no-load value) the dump valves are modulated by the plant trip controller to regulate the rate of removal of decay heat and thus gradually establish the equilibrium hot shutdown condition.
7.7.1.8.3 Steam Header Pressure Controller Residual heat removal is maintained by the steam generator pressure controller (manually selected) which controls the amount of steam flow to the condensers. This controller operates a portion of the same steam dump valves to the condensers which are used during the initial transient following turbine reactor trip on load rejection.
I Draft Version 7.7-22
____--_-__--_____J
.- to.TXX-89614 CPSES/FSAR "i
Pag) 13 of 48
. TABLE 7.7-1 (Sheet 2)
PLANT CONTROL' SYSTEM INTERLOCKS Designation Derivation Function Defeats remote load dispatching (if load dispatching is used)
C-5 1/1 turbine-impulse-chamber Defeats remote load pressure below setpoint dispatching (if load dispatching is used)
Blocks automatic control rod 1
withdrawal C-7 1/1 time derivative (absolute Makes steam dump-1
.value) of turbine impulse -hamber valves available for pressure (decrease only) above either tripping or setpoint modulation t
C-8 Turbine trip, P/3 turbine Input signal to DRAFT emergency tr'p fluid pressure non NSSS turbine /
DRAFT below setpoint generator trip DRAFT logic DRAFT or 4/4 turbine valves closed l
- .4
- Enclosure - 1 to TXX-89614 7a
- Pag 2 14.of 48' CPSES/FSAR-TABLE 7.7-1
~
(Sheet 3)
PLANT CONTROL SYSTEM INTERLOCKS Designation Derivation Function ORAFT C-9 Any condenser pressure above Blocks steam dump to setpoint, or condenser 2/4 circulation water pump breakers open C-11 1/1 bank D control rod position Blocks automatic rod above setpoint withdrawal C-16 Reduced limit in coolant Stops automatic 5,0 '
temperature above normal turbine loading 50 setpoint until condition-50 clears 50 C-20 2/2 Turbine Impulse Chamber Arms AMSAC: below 70 Pressure setpoint blocks AMSAC 70 (See FSAR Section 7.8)
t.DCC 7.2-G/ 64-578) to'TXX-89614
[A p I5: ' lr lj5 Paga 15 of 48
+'
4026-u92 STE AM DUMP CONTROL la Manual Autil0NEERED
($ TEAM PRESSURE C0hTROL) 7 TUR8thE IMPULSE AVG REFEREhCE STAGE PRES $uRE no-LDAD T
p ayg RATE / LAG l'
C0wPENSAfl0N R cat tor LEAD / LAG W
k COMPEns&T On
~1Rar LOAD REJEttack
[,
IISTA BL E I
(~)
E
(* ) ('l I
(i DEFEAT LDAD REJECTION
$TE AM DUMP CONTROL:
l allow PLAuf TRIP $ TEAM DUMP CONTROL pi STABLES IISTABLES JtE ADE S PRESSURE
$ET PLANT TRIP PRES $uRE CON TROLLER l'
LDAD REJECTION Z
(*)
C0 h TROL L ER p
p I
L LOAD REJECTION CONTROL DR PLANT PI CONTROLLER TRIP CouTROL U
V V
LO AC REJECTION C0aTROL OR PL ANT U
TRIP CouTROL TRIP QPEN STEAM DwP v8Ld!5 NOTE: FOR ILOCRipG.UN-ILOCRING SIGNAL TO CONDENSER STEAM DUMP AUTO (igyg VALVES MANUAL COMTROL) ygf,Go 3, y p
($ TEAM PRESSUREi AIR SUPPLY TO CouTROL)
DUMP VALVES p
MODULATE ConDEuSER pump valves COMANCHE PEAK S.E.S.
FNAL SAFETY ANALYSIS REPORT UNITS 1 and 2 Block Diagram of Steam I
Dung Control System I
{
rGURE 7.7-8
{
E l
I
'y
. Enclosuro:1-to TXX-89614 CPSES/FSAR L*
Page'16 of,48 10.4.4.2.3.
System Operation
.j During operating. transients for.which the plant is. designed, the steam-I dump system is automatically regulated in the average reactor coolant temperature (tavg) control mode to maintain the programmed tavg.
A programmed reactor coolant reference temperature (tavg) corresponds to each turbine load.
During load variations when the reactor and turbine outputs are unbalanced, it deviates from the actual tavg of the primary coolant. The magnitude and the rate of this deviation, which depends on the transients, provides a signal that selects a new control rod pattern and activates the adequate number and mode of operation.for the dump valves.
l During a load reduction, the valves are modulated by temperature deviation through a load rejection controller. When the mode selection switch is in the steam pressure, the valves are modulated to maintain the steam header pressure setpoint.
DRAFT On large step load reductions (above 10 percent) or reactor trip, the steam dump valves open rapidly, in three seconds or less. During the three second period, while the turbine valves are closing and dump valves are opening, there is a temperature rise in the NSSS and an approximately 100-psi pressure rise in the steam generators.
In the initial part of a large step load transient, all dump valves go fully open. Then the valves are modulated closed in sequence to obtain a design load change in the reactor of five percent per minute. The valves are fully closed when the reactor power matches the turbine power.
DRAFT Reactor trip transfers steam dump control from the load rejection controller to the plant trip controller. The reactor trip signals are redundant.
1 Draft Version 10.4-28 l
L__
l p,
. to TXX-89614 CPSES/FSAR
- Page-17 of 48
'i 15.2.3 TURBINE TRIP 15.2.3.1 Identification of Causes and Accident Description i
~ DRAFT.
'For a turbine trip event, the reactor would be tripped directly (unless below approximately 50 percent power) from a signal derived from the turbine stop emergency trip fluid pressure and turbine stop valves. The turbine stop valves close rapidly (typically 0.1 seconds) on loss of trip fluid pressure actuated by one of a number of i
'possible turbine trip signals. Turbine trip initiation signals include:
1.
Generator trip.
2.
Low condenser vacuum.
3.
Loss of lubricating oil.
4.
Turbine thrust bearing failure.
5.
Turbine overspe.ed.
6.
Main steam reheat high level.
7.
Manual trip.
Upon initiation of stop valve closure, steam flow to che turbine stops DRAFT abruptly.
Sensors on the stop valves detect the turbine trip and initiate steam dump and, if above 50 percent power, a reactor trip.
The loss of steam flow results in an almost immediate rise in secondary system temperature and pressure with a resultant primary system transient as described in Section 15.2.2.1 for the loss of external load event. A slightly more severe transient occurs for the turbine trip event due to the more rapid loss of steam flow caused by the more rapid valve closure.
The automatic steam dump system would normally accomirodate the excess steam genaration.
Reactor coolant temperatures and pressure do not significantly increase if the steam dump system and pressurizer pressure control system are functioning properly.
If the turbine condenser were not available, the excess steam generation would be dumped to the Draft Version 15.2-6
Enc 10sure'1' to TXX-89614 CPSES FSAR AMENLMENT 77 h
DETAILED DESCRIPTION-Pap 3 1 y
Page 18 of 48
. FSAR Page (as amended)
Group Description f
7.2-13, 18-2 Instrumentation and Controls - Reactor Trip System -
Reactor Trip on a Turbine Trip (anticipatory) (BOP scope)
Addition:
I l
The current CPSES reactor trip system design includes a reactor trip (P-7 permissive interlock) following a turbine trip whenever the plant is.above 10% power.
i However, for plants with a 50% or greater load
-l l
rejection capability, a reactor trip is unnecessary below 50% power if the cause of the turbine trip is readily correctable. At CPSES, both Units are designed each with a 50% load rejection capability (i.e. 40%
steam dump capacity and 10% control rod insertion).
Taking advantage of this feature, CPSES.is implementing a design modification that will eliminate a reactor trip on turbine trip when the plant is at or below 50%
power. An P-9 permissive interlock is being installed to accomplish the above functions. In addition, to the above capabilities, it is anticipated that implementeaion of P-9 will reduce the down time required to restart the plant, thereby increasing plant availability. Also, the P-9 design modification is part of the CPSES trip reduction program.
FSAR Change Request Number: 89-579.3 Related SER Laction: 7.2.1 l
SER/SSER Impact: Yes SER Sec. 7.2.1 Reactor Trip System Description (p.7-7) states "[a] turbine trip causes a direct reactor trip above 10% power (P-7 interlock)."
7.2-17 2
Instrumentation and Controls - Reactor Trip System -
Reactor Trip on a Turbine Trip (anticipatory) (BOP scope)
Addition:
The addition of P-9 permissive design modification eliminates a reactor trip block on a turbine trip signal (i.e. P-7 permissive interlock) below 10% power.
FSAR Change Request Number: 89-579.9 Related SER Section: 7.2.1 SER/SSER Impact: No 7.2-18 3
Instrumentation and Controls - List of Reactor Trips -
Turbine (anticipatory) Trip Correction:
Correction of the P-8 permissive interlock setpoint.
The setpoint is incorrectly F.tated in the FSAR as 50%.
The correct value is 48%. This correction is also necessary for clarification that the P-8 and P-9 permissives are not duplicate permissives.
FSAR Change Request Number: 89-579.13 u---.--m.__. _ _ _ _ _ _. _ _ _. _ _.. _ _ _ -. _ _ _ _
Erisosure I to.TXX-89614 DETAILED DESCRIPTION Page 2 V*
.ia'e 19 of 48 g
FSAR Page-(as amended)
Group Description Related SER'Section: 7.2.1 l
SER/SSER lapact: No Table 7.2-1 2
See Sheet No(s):4 Instrumentation and Controls - List of Reactor Trips -
Turbine (anticipatory) Trip Revision:
Revised Table to reflect the addition of the P-9 permissive interlocks for a reactor trip signal on turbine trip.
FSAR Change Request Number: 89-579.2 Related SER Section: 7.2.1 SER/SSER Impact: No Table 7.2-2 2
See Sheet No(s):2 Instrumentation and Controls - Reactor Protection System Interlocks Revision:
The addition of the P-9 permissive design modification eliminates a " reactor trip" cn " turbine trip" in the absence of a P-7 permissive interlock (i.e. below 10%
power).
FSAR Change Request Number: 89-579.3 Related SER Section: 7.2.1 SER/SSER Impact: No Figure 7.2-1 2
See Sheet No(s):10 Instrumentation and Controls - Reactor Trip System-Reactor Trip on a Turbine Trip (anticipatory) B0P Revision:
Revised functional diagrams to reflect P-4 (reactor trip) instead of turbine trip as an input for the steam dump logic. Also, corrects wording regarding the " plant" trip controller. See also FSAR descriptions page 7.7-22.
FSAR Change Request Number: 89-579.11 Related SER Section: 7.2.1 SER/SSER Impact: No Figure 7.2-1 2
See Sheet No(s):2. 4 and 16 Instrumentation and Controls - Reactor Trip System -
Reactor Trip on a Turbine Trip (anticipatory) BOP Revision:
Revised functional diagrams to reflect the addition of the P-9 permissive interlock and associated logic.
FSAR Change Request Number: 89-579.10 Related SER Section: 7.2.1 SER/SSER Impact: No
l CPSES FSAR AMENDMENT 77 Enclesura'l to'TXX-89614 DETAILED DESCRIPTION Pag 3 3 Page 20 of 48 FSAR Pepe
-1st..quended)
Group Description 1
7.7 21 3
Instrumentation and Controls - Control Systems not Required for Safety - Steam Dump Control Correction:
The CPSES automatic Steam Dump System is designed to l
dump steam directly to the main condenser. The FSAR text incorrectly states that steam can be dumped to the atmosphere. This is not part of the system design.
Thus, the phrase "and/or atmosphere" is being deleted.
FSAR Change Request Number: 89-579.4 Related SER Section: 7.7.1 SER/SSER Impact: No 7.7-22 3
Instrumentation and Controls - Control Systems not Required for Safety - Plant Trip Steam Dump Controller Correction:
Revised the wording (i.e. changed " turbine" to " plant")
to better describe the plant design and to be consistent with related Sections of the FSAR.
FSAR Change Request Number: 89-579.12 Related SER Section: 7.7.1 SER/SSER Impact: No 7.7-22 2
Instrumentation and Controls - Control Systems not Required for Safety - Plant Trip Steam Dump Controller Revision:
The current plant design includes reactor trip on turbine trip when the plant is above 10% power (i.e.,
P-7). During this time, two groups of steam dumps are available to handle the steam load capacity upon de-creasing power following a reactor trip. However, with the increase in steam load capability associated with the P-9 modification, four groups of steam dumps are required to be available to handle the 40% steam dump load rejection capability when the plant is at or below 50% power, and the reactor has not tripped. Therefore, to retain the ability to reject 40% of the steam load a
to the steam dumps should a turbine trip be initiated from less than 50% power, the transition of the steam dump control from the " load rejection" mode to " plant trip" mode is now accomplished on a reactor trip instead of a turbine trip as previously done.
FSAR Change Request Number: 89-579.5 Related SER Section: 7.7.1 SER/SSER Impact: No Table 7.7-1 2
See Sheet No(s):2 and 3 Instrumentation and Controls - Control Systems not
EnclosurO I to TXX-89614 CPSES FSAR AMENDMENT 77 Page>21 cf 48 DETAILED DESCRIPTION Pag) 4
'FSAR Page (as amended)
Group Description Required for Safety - Plant Control System Interlocks Revision:
The P-9 permissive modification eliminated the C-8
. steam dump functions. C-8 now only provides an input-signal to the non-NSSS turbine / generator trip logic.
This is not a new function for C-8, but was not previously described in Table 7.7-1 of the FSAR.
FSAR Change Request Number: 89-579.6 Related SER Section: 7.7.1 SER/SSER Impact: No Table 7.7-8 2
Instrumentation and Controls - Control Systems not Required for Safety - Plant Control System Interlocks -
Revision:
The P-9 permissive modification eliminated the C-8 steam dump functions. Thus, the block diagram for the Steam Dump Control System is being revised to reflect the reactor trip input.
FSAR Change Request Number: 89-579.14 Related SER Section: 7.7.1 SER/SSER Impact: No 10.4-28 2
Steam and Power Conversion System - Other Features of Steam and Power Conversion System - Steam Dump System Operation Revision:
The current plant design includes a reactor trip on turbine trip when the plant is above 10% power (i.e.,
P-7). During this time, two groups of steam dumps are available to handle the steam load capacity upon de-creasing power following a reactor trip. However, with increase in steam load capacity associated with the P-9 modification, four groups ofsteam dumps are required to be available to handle the 40% steam dump load rejec-tion capacity when the plant is at or below 50% power and the reactor has not tripped. Therefore, to retain the ability to reject 40% of the load to the steam g
dumps should a turbine trip be initiated from less than 50% power, the transition of the steam dump control from the " load rejection" mode to " plant trip" mode is now accomplished on e reactor trip instead of a turbine trip as previously done.
FSAR Chenge Request Number: 89-579.7 Related SER Section: 10.4 SER/SSER Impact: No 15.2-6 2
Accident Analysis - Decrease in Heat Removal by the Secondary System-Turbine Trip Revision:
I
o CPSES FSAR AMENDMENT'77' Enclosura 1 to TXX-89614 DETAILED-DESCRIPTION Pege 5
- p,g,.22 cf 48 FSAR Page (as amended)
Group Description Revised the text to reflect the~ power range of 50%
associated with the P-9 design modification.
FSAR Change Request Number: 89 579.8 Related SER Section: 15.1.1 SER/SSER Impact: No 0
)
i-to TXX-89614 Page 23'cf 48 the design changes before the issuance of the Operating Liesnse.
The following items are included and are addressed in the referenced section of this report:-
(a) containment ventilation systes isolation signals (7.3.2.1)
(b) staam generator referenca leg taecerature ec=censation and icw-Icw ste.s ganarator level satsoint par II 3u11etin 79-21 (7.3.2.2)
(c) safety system set point metnocciogy (7.3.2.6)
(d) remote shutdown capability (7.4.2)
(e) confirmation of procedura review per IE Bulletin 79-27 (7.5.4)
(f) RCS pressure protection during lew-temperature operation (7.6.4)
(2) Common electrical power sources or sensor malfunctions which may cause multiple control,systes failures (7.7.2)
(3) Environmental qualification of control systans (7.7.2) 7.2 Reacter Trio Systas 7.2.1 Description The reactor trip systes (RTS) is designed to-automatically limit reactor operation within the limits established in the safety analysis.
This function is accomplished by tripping the reactor whenever predetermined safety limits are approached or reached.
The ATS monitors variables that are directly related to systes limitations or calculated from process variables. Vhenever a variable exceeds a setpoint, the reactor is tripped by insarting control i
rods.
The RTS initiates a turbine trip when a reactor trip occurs.
The RTS consists of sensors and analog and digital circuitry arranged in coincidence logic for monitoring plant St~nals from these channels are used in redundant logic trains. pareerters.Each of the two trains opens a sepa g
reactor trip breaker.
Dgring normal power operation, a direct current undar-voltage coil in each reactor trip brosker holds' the breaker closed.
For a reactor trip, the remeval of Opening either of two series power to the undervoltage coils opens the breakers.
connected breakers interrupts the power from the rod-dr"ve actor generator sets, and the control rods fall by gravity into the The rods cannot be withdrawn until the trip breakers are manually core.
reset, and the trip breakers cannot be manually uset until the abnormal condition that initated the trip is carrected.
Sypass breakers are provided to persit the tasting of the primary breakers.
The reactor trips Ifsted below are providad in the Comanche Peak design.
The numbers in parenthesis after each trip function indicate the coincident logic as, for exaeple, two out of four (2/4).
(1) nuclear power trips (a) power-range high-heutron-flu trip (2/4)
(b) intermediate-range high-neutron-flu trip source-range high-neutron-flu trip (1/2) (1/2)
(c (d
power-range high positive-neutron-flu-rate trip (2/4 power range high-negative-neutron-flu-rate trip (2/4,{
(e 7-4
~
(
1 Eaclosure I to TXX-89614 Page 24 of 48
~
(2) core thereal peuer trips (a
overtagerature FM trip (#4)
(b ove9 suer FM trip (u4)
(3) reacter coolant system pressurizer pressere end water level trips pressurizer low pressure trip (U4) pressurizer high pressure trip (&4) pressurizer high-uster level trip (U3)
~
(4) reacte'r coolant systes low flew tripe (a
law reacter coolant fisw (U3 per leep) U4)
(b reacter coolant pup underveltage trip ( 4 reacter coolant pup underfrequency trip v)4)
(c (5) steen generater lar law meter level trip (u4)
(6) tertine trip (anticipatory) ( F3 er 4/4)
(7) safety injection signal actuation trip.(#4, U3, or' VI)
(8) annual trip (VI)
The peuer-range high-neutron-fleet trip has tue histables for a high and a low trip settins. The hfMtfag trip is active durins all medes of operaties.
The latesetting trip 1s active sely dering reacter startap and shutdeun wherf the reacter is beler M pomer.
The intermediate-range tris provides protection defes reacter startg and shutdeun when the reacter Is helev M peser.
The.searce-rence trip prwides protectfen durfag reacter sg and shutdown when the neutrem fleet chesnel is below p 6 interlock ($ x 38 any).
A pomet range high-positive-neutron flat-rete trip occurs when a sudden abneraal increase in nuclear is detected. This trip ides departure free tracleste boiling (
)protectionassinstI redejectionaccidents from at$ suer and s active during all modes of.aparation.
A peuer-reage hip-nepet'ive-neutron-flat-rate trip ecturs when a sudden abnormal decrease in esclear power is detected.
This trip prwides protection against two er more dropped rods and is active during all medes of aparatten.
The core thereat-power tries are derived by detecting the amount of the FM present in the primary coolant systes. The FM concentration (an isetape of the nitrogen generated by neuteen activation of egygen contained in the water) which is present in the primary coolant is directly p tonal to the finsten rata in the core. Decay ** *h 16-15 feetape produces enery gamme rays which penetrate the well of the hi ressure piping.
FM concentration in the primary coolant is sent by asesuring the gemas radiation,eutside 7-5
.. to TXX-89614 Page 25,y gg pritory coolant piping.
The b l6 gEEBa radiation is detected by ion chambers located on the hot-og pipin sounted on opposite sides of the hot g of each coolant loop.
Detectors we og piping se that as try offacts can be compensated for the asasured F M.ymmetries due 1.o A total of four 16 gamme detectors and mounting assemblies are provided for each loop on the hot leg outside the biological shield.
An overtemperature M-16 trip protects the core against low departure from nucleate boiling ratio (DWR).
The setpoint for this trip is continuously calculated by analog circuits to compensate for the effects of temperature pressure, and axial puwer distribution effects on D ER limits.
IncreasesIn the axial flun difference, determined by the average of the upper two and the average of the lower two power range neutron detectors, beyond a predefined deadband result in 3 decrease in trip setpoint.
The trip setpoint is decreased when cold-leg tugwature increases and when primary coolant pressure decreases.
An overpower N-M hip protects against excessive power (fuel rod ratins protection).
The setpoint for each channel is continuously analog circuits to compensate for axial flux differences. y calculated 6y The theory, description, analysis, plant tests, and rience with F 16 systems was presented in Westinghouse Topical Aspert 9190, "> M Power Measurement System," which ses reviewed and approved by the staff. The appli-cant has documented the differences between the Comanche Peak N-M system and that described in WCAP-fl90.
The Comanche Peak rystem contains a temperature compensation tem vet contained in WCAP-9190.
The temperature compensation utilizes the cold-leg temperature esasurement for this compensation.
In addition, the F 16 systas in WCAP-9190 contains a background radiation compenst -
tion tem which is not in the Commenche Peak.FN systas.
This compensation is requined only if significant fuel failure exists, as' discussed in WCAP-9190.
If fuel failures should occur during operetten, the F M power seter will read high which anans the trip point will es reached eer11er then re The app 1Icant also clarified that the transit ties flow meter (TTFN) quired.
used via the N-16 systes to calibrate the reacter coolant flew is a portable electronic rock. Before the TTFM is connected, 411 of the bistables associated with that N-16 power monitor channel would be put in the tripped ande (as is done for test purpose) to preclude any control (TTFN) and protection ( F M power) interaction.
The pressurizer leepressure trip is used to protect assinst lou pressure that could lead to 05. The reactor is tripped when the pressurizer pressure (lead / lag compensated for rate of change) falls below a preset Ifeit.
This trip is blocked (P-7 laterlock) below approximately 15 ef full power to allow startup.
8 The pressurizer high pressure trip is used to protect the reactor coolant systas against s pressurizer low ystem overpressure.
The same trenesitters used for the pressure trip are used for the high pressure. trip. The reactor is tripped when pressurizer pressure exceeds a preset limit.
The pressurizer high-water-levat trip is provided as a bac to the pressurizer high-pressure trip and serves to prevent water relief th the pressurizer safety valves.
This trip is blocked (P-7 interlock) below approximately 10K of fall power to allow startup.
7-6
. to TXX-89614 Pege 26 of 48 Law reacter coolant flow is sensed by transmitters connected to elbow taps in each coolant leep.
The reacter is tripped when two out of three transmitters sense leu flew in a leep. This tH p protects the core fres low Om for a loss of primary coolant flow.
1 The reactor coolant pep undervoltage tHp is provided to protect against the low flow that can result fres loss of voltage to mere then one of the reacter coolant pump meters caused by a loss of power er a trip of the reacter coolant
{
pump breakers. One underveltage sensing relay.fs provided for each pop at the meter side of each reacter coolant pep breaker. These relays provide an output signal when the pep voltage goes below approximately 71K of rated voltage. Signals from these relays are time delayed to prevent spurious trips caused by short-tors weltage perturbations.
The reacter coolant pump underfrequency trip protects against low flow resulting freu underfrequency es a result of a major power pHei deter 6ance.
It trips the reacter when it senses an underfrequency condttien. One underfrequency sensing relay is providsd for each reacter coolant pep meter.. Signals for i
l any tuo of the four pump esters (ties delayed g to appremiestely 0.1 see to prevent spurious trips caused by short-tore frequency portuttations) will trip l
the reacter. The tHp is hypes'Jed if the power level is below approximately i
1 3 of full pewsr.
l The steam generator few-few unter level trip /fseheter flew misestch.'pretmets her,t (Ank in the evert of a sosteSed rkan A reacter trfp en a turhine tHp is actuated by tuo out of thres energe~ ncy-trip fluM prescura signals or b turbine staes step valves. y all (four out of four) closet signals free the A turbine tHp causes a direct reacter trip above 15 power (P-7 interlock).
j Asafetyinjectionsignalinitiatesareactertrip.
This tH p protects ths core against a loss of reacter coolant or overcooling.
The manual trip consists of tus switches. Operation of either switch de-one ses the undervel coils is seek logic train. At the same ties the b
r shunt cofis la brechers are enesvised, ukich provides a diverse means to ensure that the trip and bypass breakers are tripped.
The analog portion of the RTS consists of a partion of the pre:ess instruments-tion systes (p!$) and the nuclear instrumentattee system (NIS). The PIS includes these devices that osasure temperature, pressure, fluid flow, and level. The PI5 aise includes the power supplies si I conditioning, and bistables that provide initiation of protective ions. The NIS 'ncludes the neutree fluu moniteHag instruments, including poser supplies, signal conditioning, and histables that provide initiation of protmetive functions.
The dig (ital portion of the NTS consists of the solid state logic protection system 55LPS).
The55LP5takesbinaryinputs(vol ne voltage) from the PIS and NIS channels corresponding to normal / trip tions for plant parameters.
The 55LPS utilizes these 6tgnals "n the required logic combinations and generates trip signals (no voltage) to the undervoltage coils of the reactor tH p circuit breakers. The systes also provides annunciator, states light, and computa:-input signals that indicate the condition of the histable input signals, partial and 7-7
{
e J
e
. Enclosure I to TXX-89614 Page 27 of 48 4
full trip functions actuation functions., and the status of various blocking, pe missive, and In addition, the 55LPS includes the logic circuits for testing.
Analog signals derived from protection channels used for nonprotective functions such as control, remote process, and computer ienitoring are provided by the use of isolation amplifiers located in the protective system cabinets.
The isolation amplifiers are designed se that a short circuit, open circuit, or the application of credible fault voltages from within the cabim ts on the isolated output portions of the circuit (nonprotective side) will not affect the input signal.
The signals obtained from the isolation amplifiers are not returned to the protective system cabinets.
7.2.2 Requirements for Reacter Protection System Anticipatory Trip The staff issued a trench Technical Position (STP ICSB 26) that anticipatory trips should satisfy.all the requirements of IEEE Standard 279 oven if the safety analysis could demonstrate acceptable consequences withest the protection provided by such trips.
The basis of this positten uns that nonsafety grade equipment may degrade the capability of the protection systies for same events.
This sublect was an issue during the Construction Po mit review with *espect
{
to signs!r, derived from reactor coolant pump breakers. Pecausa the final l
desi r uses undervoltage and underfrequency detection and bncaesa the associated s m ers are located in prota ti m systm racks, this conc'>rn has been resolved.
I
. The anticipatory reacter tHp derived from turbine trip involves nonseismically qualified t.sraers locatei fe nenwienic structures.
Further, redundant channsla era routed in attal conduit escissively for those ciresits and separation in accordance witu redundant protection system channels is malntained.
The staff finds this design to be ucceptable.
7.2.3 Test and Calibration Features of the Safety Systems Regulatory Guide 1.22 describes acceptable methods for complying with GDC 21.
TheComanchePeakdesigncomplieswithReplatoryGuide1.22."PeriodicTesting of Protection System Actuation Functions.
The app 1tcant has determined that the following equipment cannet be tested at full power, so as not to damage egofpment or p oet plant aperation:
(1) manual actuation switches (2) reacter coolant pep breakers (3) turbine trip equipment from reactor trip Thestafffindsthattheapplicanthasproperlyjustifiedhisdesignstomeet the guidelines of Postion D.4 of Regulatory Guide 1.22.
The staff finds this to be acceptable.
7.2.4 Response Time Testing Regulatory Guide 1.118 (Position 6.3.4) requires that safety system response time measurements be made periodically to verify overall responsa time withiri 7-8
. to TXX-B9614 Page 28 of 48 7.7 Centrol Systems Not saadred for Safety 7.7.1 Description The plant control systems not required for safety described in this section include the following:
(1 reacter control system red control system monitoring and indicating plant control signals plant-control system interlocks (5
pressurizer pressure control systes (6
pressurizer unter level control system (7
steam rater meter level control system (4) steam control system The reacter control system (ACS) enables the nuclear plant to follow lead changes autametically, including the acceptance of step lead increases er decreenes of 13 and resp increases er decreases of E per sin within the.
(subjecttopossiblemonenlimitations). lead range of 15 to 105 er pressure relief The tan is a restering the coolas* averass temperature (T to within the temperature deseand followin aqr be perfereed at any time.g a change in 1 Manual centre red operation The full-length red control system receives red speed and direetten signals free the reacter coolant systes #
si Is.
a control bank in er out at a psMbebined speed.Nenusi control is provided to mov When the tuttine lead reaches ap
'any select tInd " automatic" mode, proximately 1 R ef rated lead, theretor and red action is then controlled the RCS.
A permissive interlock (C-5), derived free sensurements of turbine apulse chenber pressure, prevents automatic control when the turtine lead is below 1E.
In the "autaastic" mode, the rods are again withdrawn (or inserted) in a predetermined sequence by the automatic progressing with the control interlocks.
sur shutdeun banks are always in tas fully withdrawn position during normal operation, and they are seved to this position at a constant by annus1 control prior to criticality. A reacter trip signal causes s
to fall by gravity into the core.
that can be manipulated under autametic controlThe centrol banks are the only rods and each control bank is divided into too gregs to obtain ses11er increm, ental reactivity changes per stap. All red control cluster assemblies in a group atre electrically. paralleled to move simultaneously.
cluster centrol assembly. There is individual position indication for each red The monitoring and indicating plant signals described belev are available (1) Powr range channels are provided and are isportant because of their use in monitoring power distribution in the core within specified safe Itaits.
They are used to measure power level axial flux imbalance and radial flux tabalance. These channels are capable of recording overpow,er excurstens up to 203 of full power.
Suitable alares are der'ved from these signals.
7-29
1
. to TXX-89614 Page 29 of 48 Two separate systems are provided to sonen and dis (2)
.The digital rod position indication system (ORPIS) play contml rod positir measures the actual position of each full-length rod.
The control board display unit contains a column of light-emitting-diodes (LEDs) for each rod.
time At any given partIcularrod.the one LED illuminated in each column shows the position for Included in the system is a rod-at-bottom signal that operates a local alarm for each rod.
A control room annunciator is also actuated when any shutdown rod or control bank A rod is at bottom.
The demand position system controlsystemtoprovided(igitalreadestofthedemandedbankposition Operating procedures require the reactor operator to compare the demanded position from the DPS to the indicated (actual) readings from the DAPIS to verify operation of the rod control system.
When the reactor is critical, the normal indication of the status of l
reactivity in the core is the position of the control rod bank in reistion to reactor power (as indicated by the RCS loop AT) and the coolant average temperature.
These prameters are used to calculate inser$1en Ifmits for the control banks. Two alams are provided for each control bank.
The
" low" alarm alerts the operator to en approach to the rod insertfoe limits which will im boron addition by following normal procedures with the chemical volume control system. The " low-law" alam alerts the operator to take immediate action to add boron to the ACS by any one of several alternate methods.
(3) In a rod deviation function. performed as part of the DRP!$, an alam is generated if a preset limit is exceeded as a result of a comparison of any control rod with the other rods in a bank. The deviation alars of a shutd'am rod is actuated when a preset insertion limit is exceeded.
The demanded and measured red positten signals are aise monitored by the plant computer which provides a visual printest and an audible alam whentver an individus1 red position signal deviates from the et 'r rods '
in the bank by a preset limit. The alam can be set with appre fate allowance for instrument error and within sufficiently narrow limits to
, preclude exceeding core design het channel factors.
(4) A rod-bottos signal for tha full-length rods in the digital rod position system is used to operata a control miay, which generates the " rod bottaa rod drop" alare.
~
(5) A surveillance feature is provided to monitor core power distribution during nemel operation.
This surveillance feature monitors peak linear power density, and is designed for use as an advisor No credit on by the operator, Technical Specifications, ysis, y s is taken for this feature <n the accident anal and it is not relied or a ministrative control procedures as a basis for decisions concerning control of the plant.
The linear power shape is presented on a CRT disp ay in the control room.
TheplantcontrolsysteminGr5cksaredescribedinthefollowingparagraphs (1) Rod stops are provided to prevent abnormal power conditions which could result from excessive control rod withdrawal initiated by either a control system malfunction or operator violation of administrative procedures.
7-30
. to TXX-89614 Page 30 of 48.
Rod staps ars the C-1, C-2 C-3, C-4, and C-5 control interlocks.
The C-3 red stap derived from o,vertasperature F lf and the C-4 rod stop derived from over discussed below. power F 16, are aise used for, turbine lead runback,
(2) Autoestic tuttine load ru~nbeck is initiated iry' an approach to an overpower or overtemperature condition which, if reached, will be protected by reacter trip.
The runbecks are siew (< W of rated load within the capacity of avtamatic reactor load following. per sin) and are (3) A C-14 interlock is provided to limit tuttine loading during a rapid return to power transient when a reduction in reacter coolant. temperature is used to increase reacter power (through the negative moderator coef-ficient).
This interlock limits the drop in coolant temperature not to exceed cooldsun accident limits and preserves satisfactory steam generator operating conditions.
Subsequent avtamatic turbine loading can begin after the interlock has been cleared by an increase in coe ant tasperature which is accomplished by reducing the boren concentration in the coolant.
The reacter coolant system (RC5) pressure is controlled by using'aither the heaters (in the water region) er the sprey (in the stesa en) of the pres-surirer plus steen reifer for large transiests. The el cal innersion i
heaters are located near the bottaa of the pressurizer. A portion of the heater gree is proportievelly contes 11ed to correct small pressure variations.
These variations are de result of heat lossos, including heat lesses due to a small contir m s sprey Theremaining(backsp)heatersareturnedonwhenthe pressurizer pressure co.
I ntrolled sipal demand is approxiestely 105 proportional to heater power.
The orig nos41es are located on tap of the pressurizer.
. is initiated when the pressure centre 11er spesy demand signal is above a givenSpray
- setpoint.
The sprey rete increases proportionally with increasins spray demand sipiel until it reaches a maximum value.
Steam condensed by the spray reduces tie pressurizer pressure. A small continuous spray is normally maintained to reduce thermal stresses and theresi shock and to help maintain unifore water chasistry and tasperature in the pressurizer. power operated 1
relief valves limit systes pressure for large, positive pressure transients.
In the event of a 1 rejectioncapabilit lead reduction (not oncending the design plant load adverse condittens,y the pressurizer PORVs afght be actmeted for the most mest aspetive Doppler coefficient and the anximum i
incremental red worth. ThereliefcapacityofthePORVsIslargeenoughto limit the systas pressure to trip for the above conditten. prevent actuation of this high pressure reactor The water investery in the RC5 is maintained by ther chemical and volume control systan(CVCS). During normal slant operetten the charpng flow varies to produce the fler demanded by the pressurizer w,ater-leve controller.
The with the highest everage temperature (auctionsored) being used The pressurizer water level decreases as the lead is reduced free full lead.
The programmed level is designed to match as nearly as possible the level changes resulting from the coolant tasperature changes.
To control pressurizer water level during starty and sWtdown operations, the charging flow is annually regulated from the main control rosa.
e 7-31
Enclosure a to TXX-89614 Page 31 of 48 Each steam ' generator is equipped with a three-element feedseter flow controller, which maintains a programmed water level which is a function of turbine load.
The three-element feedwater controller regulates the feeduster valve b programmed level, and the pressure-compensated steam flo tinuously comparing the feedwater flow signal, the water-level si the In addition, the main turbine feedwater pump speed is varied to maintain a programeed Jressure differential between the steam header and the feed pump discharge weder.
The speed controller continuously compares the actual d with the p
raemed P which is a linear function of steam flow.
Continued delivery eedwater E$f,he steam generators is required as a sink for the heat stored t
o and generated in the reactor after a reactor trip and turbine trip. An override signal closes the feedwater valves when the average coolant temperature is below a given temperature and the reactor has tripped, Manual override of the feedwater control system is available at all times.
The steam dump system is designed to allow the turbine generator to accept a SGK loss of not load without, tripping the reactor.
The automatic steam dusp system is able to accommodate this abnormal load rejection and to reduce the effacts of the transient imposed upon the reactor coolant syst.am.
If the difference between the reference T T
pressuraandthelead-lagcompensal30a(ullfo)neeredTbased on turbine ispulse cha asceeds a predeter-minedamount,andtheinterlockmentionedbelowissIIIsfied,ademandsignal will actuate the ste diep to maintain the ACS temperature within control range until a new equilibrium condition is reached. To prwvent actuation of steam dump on small load rarturbations, an independent load-rejection sensing circuit is provided.
Tnis cittcit senses the rate of decrease in the turbine load as detected by the turbine impulse chamber pressure.
It Ys provided to unblock the dump valves when the rata of load rejection exceeds a preset value corresponding to a 10K step-lead decrease or a sustained ramp load decrease of SKhin.
7.7.2 Conclusions The control systems not required for safety control. include such systes partanters
.as reactivity, primary system pressure feedwater flow.-and turbine speed.
The scope of the review included descrI>tive inforestion,ical functional logic, instrumentation and electrical diagrams and Thereviewincludedtheapp1Icant'partialphyss design bases and ana yses.
drawings.
The basis for acceptance in the review has been conformance of the applicaat's designs, design criteria, and design bases for the control systems not required for safet Criteria,y to the Commission's regulations as set forth irr ths General Design and to applicable Regulatory Guides, Branch Technical Positions and i
industry standards as listed in Table 7-1, as well as the staff review and acceptance of similar systems.
It has been verified that control s Station (Dockets50-36g and 50-370)y the same as for the W required for safety are functionally McGuire previously reviewed and accepted by the staff.
The plant control systems are designed to ensure high reliability in any anticipated operational occurrences.
Equipment used in these systems is designed and constructed with a high level of reliability.
The plant control system will tend to prevent an undesirable condition in the operation of the plant that, if reached, will be protected by the reactor trip system.
Failure 7-32
.a.
9 to TXX-89614-Page 32 of 48' The staff has reviewed the applicant's system description and design criteria i
i for the components of the tumine gland sealing system and found twy conform to applicable regulations and industry standards, including Regulatory Guides 1.26 and 1.29.
Therefore, the staff finds the proposed turbine gland sealing system' acceptable.
10.4.4 Turbine Bypass System A
The turbine bypass system (referred to as the steam dump system by the applicant {
for each unit is designed to b turbine to the main condenser.ypass up.to 40% of main steam flow around tie This capacity, together with a 10% reactor auto-matic stepload reduction capability, is sufficient to withstand a 50% generator.
load loss.without tripping the reactor or turbine. The turbine bypass system is used to control reactor pressure as follows:
(1) during the reactor heatup to rated pressure; (2) while the turbine generator is being brought up to speed and synchronized; (3) during power operat<on when the reactor steam generator exceeds the transient turbine steam requirements; and (4) during accident
/
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^
The turbine bypass system consists of 12 air-operated modulating control valves mounted on a valve manifold. The manifold is connected to the main steam lines upstream of the turbine main stop valves, and the bypass valve outlets are piped to the main condenser. The turbine bypass valves are modulated by a signal nerated by the deviation from a set point of either the reactor coolant average reture or main steam header pressure, depending upon the control mode.
They may also be operated manually from the control room.
The turbine bypass valves are blocked from opening on loss of main condenser and they fail closed upon loss of air or control signal to the valves.
vacuum Periodleinservicetestsoneachvalvewillbeperformed.
If the turbine bypass system fails tt. operate (that is if all valves do not operateasdesired),excessivepressureriseinthemaInsteamgeneratorlifts the relief and/or safety valves. The safety valves are sized to accommodate more than the nomal full steam flowrate.
The adequacy of the turbine bypass system with respect to Branch Technical Positions ASB 3-1 and MEB 3-1 is evaluated in Section 3.6 of this report.
O The review of the turbine bypass syst.am included drawings, piping and instru-mentation diagrams, and descriptive infomation of the system in Section 10.4.4 of the FSAR. The design criteria and bases and design of 15e system met the acceptance criteria in Section II of SRP 10.4.4, and industry standards.
There-fore, the staff conc 1bdes that the system can perform its design function, and 1s, therefore, acceptable.
10.4.5 Circulating Water System The nonsafety-related (Quality Group D nonseismic Category I) circulating water systemassociatedwitheachreactorunitisdesignedtoremovetheheatrejected from the main condenser, auxiliary condensers turbine plant cocling water heat 3
exchanger, and condenser exhausting vacuum pum,p heat exchan rs.
The heat sink
]
and water supply for the circulating water system is Squaw reek Reserveir.
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Enclosure i to.TXX-89614 TABLE 7.2-2 l..
. Page 34 cf 48 l
(Sh et 2)
PROTECTION SYSTEM INTERLOCKS Desig-nation Derivation Function Defeats.the block of inter-mediate.*ange reactor trip and intermediate range rod stops (C-1)
Input to P-7 II Blocks of Reactor Trios P-7 Absence of P-7: 3/4 neutron Blocks reactor trip on:
flux (power range) below set-Low reactor coolant flow point (from P-10) in more than one loop, and undervoltage, under-2/2 turbine impulse chamber frequency, fd/5fdd f/fp/
pressurizer pressure below setpoint low pressure, and pressu-(from P-13) rizer high level P-8 Absence of P-8: 3/4 neutron Blocks reactor trip on low flux (power range) below reactor coolant flow in a setpoint single loop P-9 Absence of P-9: 3/4 neutron Blocks reactor trip on flux (power range) below turbine trip below setpoint approximately 50% power P 2/2 turbine impulse chamber Input to P-7 pressure below setpoint
- ~ to TXX-89614 CPSES/FSAR L.
'Page 35 ef 48 relay rack. The protection relay rack is divided into four.
separate compartments, similar to the Solid State Protection System input cabinet described.in Section 7.1.2.3 to maintain channel independence.
Potential transformers from which the input signals for undervoltage and underfrequency relays are derived and the protection relay rack are classified as. Class IE and are located in a Seismic Category I structure.
1 1
5.
Steam generator trip / Low-low steam generator water level trip 13 The specific trip function generated is the low-low steam 13 generator water level trip. This trip protects the reactor from loss of heat sink in the event of a loss of feedwater to the steam generators. This trip is actuated on two out of four low-low water level signals occurring in any steam generator.
The logic is shown on Figure 7.2-1. Sheet 7.
l13 6.
Reactor trip on a turbine trip (anticipatory) (BOP Scope)
' 13 The reactor trip on a turbine trip is actuated by two out of three logic from emergency trip fluid pressure signals or 'y all o
closed signals from the turbine steam stop valves. A turbine trip causes a direct reactor trip above P-9 7.
The reactor trip on turbine trip provides additional protection and conservatism beyond that required for the health and safety of the public.
This trip is included as part of good engineering practice and prudent design.
No credit is taken in any of the safety analyses (Chapter 15) for this trip.
The turbine provides anticipatory trips to the Reactor Protection l
System from contacts which change position when the turbine stop l
7.2-13 Bold /0verstrike Version
~
Enclosure li to TXX-89614
..e
' Page' 36 of 48 CPSES/FSAR j
The interme'iate range level trip and power range (low setpoint) d trip can only be blocked after satisfactory operation and permissive information are obtained from two out of four power range channels.
Four. individual blocking switches are provided so that the low range power range trip and intermediate range trip can be independently blocked (one switch for each train).
These trips are automatically reactivated when any three out of four power range channels are below the permissive (P-10)-
setpoint, thus ensuring automatic activation to more restrictive trip protection.
The development of permissives P-6 and P-10 is shown on Figure 7.2-1. Sheet 4.
All of the permissives are digital; they are derived from analog signals in the nuclear power range and intermediate range channels.
Refer to Table 7.2-2 for the list of Protection System interlocks.
2.
Blocks of reactor trips at low power Interlock P-7 blocks a reactor trip at low power (below approximately 10 percent of full power) on a low reactor coolant flow in more than one loop, reactor coolant pump undervoltage, reactor coolant pump underfrequency, pressurizer low pressure / or pressurizer high water level df fd/Bidd f(fp ifdddI.
Refer to Figure 7.2-1 Sheets 5/ and 6 ddd I6/ for permissive applications. The low power signal is derived from three out of four power range neutror: flux signals below the setpoint in coincidence with two out of two turbine impulse chamber pressure l
signals below the setpoint (low plant load).
See Figure 7.2-1, l
Sheets 4 and 16, for the derivation of P-7.
l l
7.2-17 Bold /0verstrike Version
Enclosura 1 to TXX-89614 CPSES/FSAR l,
Page 37 of 48 Th2 P-8 inttriock blocks a reactor trip wh n the plant is below approximately SE 58 percent of full power, on a low reactor coolant flow in any one loop. The block action (absence of the P-8 interlock signal) occurs when three out of four neutron flux l
power range signals are below the setpoint. Thus, below the P-8 setpoint, the reactor vill be allowed to operate with one inactive loop and trip will not occur until two loops are indicating low flow.
Refer to Figure 7.2-1 Sheet 4, for derivation of P-8, and Sheet 5 for applicable logic.
The P-9 interlock blocks a reactor trip following a turbine trip when the plant is below approximately SOE of full power. The plant is designed for 505 load rejection capability (405 steau dump and 107. control rods intertion). The absence of the P-9 interlock (three out of four power range flux signals below the setpoint) blocks the immediate reactor trip following a turbine trip. Refer to Figure 7.2-1, Sh. 4 for derivation of P-9 and Sh.
16 for applicable logic.
Refer to Table 7.2-2 for the list of Protection System blocks.
5 7.2.1.1.4 Coolant Temperature, N-16 and Power Range Neutron 5
Detector Sensor Arrangement 0032.56 hot and T old measurements are, as a result of 5
The bypass line T c
incorporation of the Comanche Peak Protection and Surveillance Upgrade Package, being replaced with an N-16 power monitor and an in-line Tcold measurement.
In addition, the present two-section power range neutron detectors are being rep 1 Geed with four-section detectors (used for both protection and control functions).
5 7.2.1.1.4.1 In-Line Fast Response Thermowell Installation of Cold 5
Leg Temperature Measurement 1
i The fast response in-line Teold measurement is provided by an RTD in a thin wall thermowell installed in the reactor coolant cold leg j
j Amendment 76 7.2-18 May 1, 1989
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~
The automatic Steam Dump System is able to accommodate this abnormal load rejection and to reduce the effects of the transient imposed upon the RCS.
By bypassing main steam directly to the condenser ind/d/
fMd ifddipHd/d/ an artificial load is thereby maintained on the primary system. The Rod Control System can then reduce the reactor temperature to a new equilibrium value without causing overtemperature and/or overpressure conditions. The steam dump steam flow capacity is 40 percent of full load steam flow at full load steam pressure.
If the difference between the reference Tavg (Tref) based on turbine impulse chamber pressure and the lead-lag compensated auctioneered Tavg exceeds a pre-determined amount, and the interlock mentioned below is satisfied, a demand signal will actuate the steam dump to maintain the RCS temperature within control range until a new equilibrium condition is reached.
To prevent actuation of steam dump on smal' load perturbations, an independent load rejection sensing circuit is provided. This circuit senses the rate of decrease in the turbine load as detected by the turbine impulse chamber pressure.
It is provided to unblock the dump valves when the rate of load rejection exceeds a preset value corres-ponding to a 10 percent steo load decrease or a sustained ramp load decrease r* 6 percent / minute.
A bloc,k diagrass of the Steau Dump Control System is shown on Figure 7.7-8.
7.7.1.8.1 Load Rejection Steam Dump Controller l
This circuit prevents large increase in reactor coolant temperature following a large, sudden load decrease. The error signal is a and difference between the lead-lag compensated auctioneered Tavg the reference T based on turbine impulse chamber pressure.
avg l
l 7.7-21
Enclesure 1"to TXX-89614 Page 43 of 48 CPSES/FSAR The Tavg signal =1s the same as that used in the RCS. The lead-lag compensation for the Tavg signal is to compensate for lags in the plant thermal response and in valve positioning.
Following a sudden load decrease. Tref is immediately decreased and T tends to avg increase, thus generating an immediate demand signal for steam dump.
Since control rods are available, in this situation steam dump terminates as the error comes within the maneuvering capability of the control rods.
7.7.1.8.2 Plant 7d/Widd Trip Steam Dump Controller Following a reactor fd/Midd trip. as monitored by the reactor fdf5fdd trip signal (P-4)..the load rejection steam dump controller is defeated and the turbine trip steam dump controller becomes active.
Since control rods are not'available in this situation. the demand signal is the error signal between the lead-lag compensated auctioneered Tavg and the no-load reference T vg. When the error signal exceeds a pre-a determined setpoint the dump valves are tripped open in a prescribed sequence. As the error signal reduces in magnitude (indicating that the RCS T is being reduced toward the reference no-load value) avg the dump valves are atoduiated by the plant trip controller to regulate the rate of removal of decay heat and thus gradually establish the equilibrium hot shutdown condition.
7.7.1.8.3 Steam Header Pressure Controller" l
i Residual heat removal is maintained by the steam ger,erator pressure controller (manur.11y selected) which controls the amount of steam flow to the condensers. This controller operates a portion of the same steam dump valves to the condensers which are used during the initial transient following turbine reactor trip on load rejection.
7.7-22
- to TXX-89614 CPSES/FSAR 7 --
_ Pag 2.44 of 48-TABLE 7.7-1 (Sheet 2)
PLANT CONTROL SYSTEM INTERLOCKS Designation Derivation Functiqn Defeats remote load dispatching (if load dispatching is used)
C-5 1/1 turbine impulse chamber Defeats remote load pressure below setpoint dispatching (if load dispatching is used)
Blocks automatic control rod
, withdrawal C-7 1/1 time derivative (absolute Makes steam dump value) of turbine impulse chamber valves available for pressure (decrease only) above either tripping or setpoint modulation C-8 Turnine trip, 2/3 turbine Input signal to emergency trip fluid pressure non-K$$$ turbine /
below setpoint generator trip legic or i
l 4/4 turbiac valves closed 516tts (tddd dddf tinitti til 1664 tilittl66 Tggg tintidtidt l
MdKdi ild66 866f instit ittistsid 16t ditMit ttifflit it i
66dd16ti46 l
1.
Enc 10 sura 1 to TXX-89614
.Page,45'of 48 CPSES/FSAR
+
TABLE 7.7-1 (Sheet 3)
PLANT CONTROL SYSTEM INTERLOCKS Designation Derivation Function Nd idfelds tfipt 2/3 tdf5fdd 516dKi itddd dddp 6difdddtf tti$ 11614 $fdiidfd tedif61 tid tiffidd 666t4 titfildt did 1/4 ffi$ Tggg tattidifinlit lidt ittp tilidi d6dtf611df
$6L lidtid C-9 Any condenser pressure above Blocks steam dump to setpoint, or condenser 2/4 circulation water pump breakers open C-11 1/1 bank D control rod positiow Blocks automatic rod above setpoint withdrawal C-16 Reduced limit in coolant Stops automatic 50 temperature above normal turbine loading 50 setpoint.
until csrAition SC c'itars 50
)
C-20 2/2 Turbine Impulse Chamber Arms AMSkC: below 70 Pr'ess are setpoint blocks AMSAC 70 1
(See FSAR Section 7.8)
_.-----_--_-_____-_--w
LDcC2 7.2-CP/ d4-579 Enc 10sure 1 to TXX-89614 Page 46 of 48 EA P IS 4 I3 u026-992 STEAN DuwP CoutdDL tu unhual AUCTtout[ RED
($TE Aw 'RE5$utt CONTROL)
TAVG 7 'O 0
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l allow PLauf Tt P STEAM DUMP CDsTROL Bl5 TABLES Cl$ TABLES MEADED PRE 33utt F
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l FNAL SAFETY ANALYSIS REPORT l
UNITS 1 and 2 Block Diagram of Steam Dump Control System FGURE ?.7-8
1:
' Enclesura 1 to TXX-89614 CPSES/FSAR
- Pag) 47 of 48 m
10.4.4.2.3 System Operation E
L During' operating transients for which the plant is designed, the steam dump system is automatically regulated in the average reactor coolant temperature (tavg) control mode to maintain the programmed tavg.
A programmed reactor coolant reference temperature (tavg) corresponds to each turbine load. During load variations when the reactor and turbine outputs are unbalanced, it deviates from the actual tavg of the primary coolant. The magnitude and the rate of this deviation, which depends on the transients, provides a signal that selects a new control rod pattern and activates the adequate number and mode of operation for the dump valves.
During a load reduction, the valves are modulated by temperature l
deviation through a load rejection controller. When the mode l
selection switch is in the steam pressure, the valves are modulated to maintain the steam header pressure setpoint.
On large step load reductions (above 10 percent) or reactor fdtefM trip, the steam dump valves open rapidly, in three seconds or less.
-l i
During the three second period, while the turbine valves are closing i
and dump valves are opening, there is a temperature rise in the NSSS and an approximately 100-psi pressure rise in the steem generators.
j L
In the initici part of a large step 1 cad transteat, all dump vt,1ves 90 fv11y open. Then the vahes are modulated closed in sequence to obtain a design load change in the reactor of five percent r>er winut2.
j 1
The valves are fully closed when the reactor power matches t4t; tvrbtur:
l powe r..
l Reacter 7d/Widd trip transfers stcom dump control f rno the lead l-rejection controller to the plant fd/5/U trip controller.
The i
1 L
reactor fd/Widd trip signals are redundant.
l l
l.
l l
Bold /0verstrike 10.4-28 Version
Enclosura 1 'to TXX-89614 CPSES/FSAR Paga 48 of 48 15.2.3 TURBINE TRIP 15.2.3.1 Identification 'of Causes and Accident Description For a turbine trip event, the reactor would be tripped directly (unless below approximately 5018 percent power) from a signal derived from the turbine stop emergency trip fluid pressure and turbine stop
.va M s.
The turbine stop valves close rapidly (typically 0.1 seconds) on loss of trip fluid pressure actuated by one of a number of possible turbine trip signals. Turbine trip initiation signals include:
1.
Generator trip.
2.
Low condenser vacuum.
3.
Loss of lubricating oil.
4.
Turbine thrust bearing failure.
5.
Turbine overspeed.
6.
Main steam reheat high level.
7.
Manual trip.
Upon initiation of stop valve closure, steam flow to the turbine stops abruptly.
Sensors on the stop valves detect the turbine trip and initiate steam rhap and, if above 5019 percent power, a reactor trip.
The loss of steam flow results in an almost immediate rise in secondary system temperature end pressure with a resultant primary system transient as described in Section 15.2.2.1 for the less of I
enternal lead event. A slightly more severe tr6nsie it ocevrs for the turbine trip event due to the vore rapid loss of steem flow caused by the mere rapid v41re clnure.
The automatic steam dump system would norsally accommodate the excess I
steam generation. Reactor coolant temper 6tures anc prusure da noc significantly ircrease if the steam dump system and pressurizer pressure control system are functioning properly.
If the turbine condenser were not available, the excess steam generation would be dumped to the Bold /0verstrike 15.2-6 Version
. to TXX-89614 August 23, 1989 Page 1 of 20 Advance Technical Specification Change Related to Implementation of P-9 Permissive Reactor Trip on Turbine Trip Below 50% Power end Supporting Documentation Item 1 Draft Revised Technical Specification pp. 2 - 9 Pages Item 2 Description / Justifications of pp. 10 - 11 Technical Specification Changes Item 3 Mark-up of Certified Technical pp. 12 - 20 Specifications i
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p L to TXX-89614 Pag 2 3 of 20 LIMITING SAFETY SYSTEM SETTINGS BASES Reactor Coolant Flow (continued)
On increasing power above P-7 (a power level of approximately 10%
of RATED THERMAL POWER or a turbine first stage chamber pressure at approximately 10% of full power equivalent), an automatic Reactor trip will occur if the flow in more than one loop drops below 90% of nominal full loop flow. Above P-8 (a power level of approximately 48%
of RATED THERMAL POWER) an automatic Reactor trip will occur if the flow in any-single loop drops below 90% of nominal full loop flow, l
Conversely, on decreasing power between P-8 and P-7 an automatic l
Reactor trip will occur on low reactor coolant flow in more than one -
loop and below P-7 the trip function is automatically blocked.
Steam Generator' Water level The Steam Generator Water Level Low-Low trip protects the reactor from loss of heat sink in the event of a sustained steam /feedwater i
flew mismatch resulting from locs of normal feedwater. The specified setpoint provides allowances fer starting delays of the Auxiliary feedwater System.
Undervoltaae and Underfreauency - Reactor Coolant Pumo Busses The Undervoltage and Underfrequency Reactor Coolant Pump Bus trips provide core protection against DNB as a result of complete loss of forced coolant flow. The specified setpoints assure a Reactor trip l
signal is generated before the Low Flow Trip Setpoint is reached.
Time delays are incorporated in the Underfrequency and Undervoltage l
trips to prevent spurious Reactor trips from momentary electrical L
power transients.
For undervoltage, the delay is set so that the time required for a signal to reach the Reactor trip bretkers folhwing the l
simultar, ecus trip of two or more reitetor coolant pucp bu circait breakers shall not exceed 1.2 seconds.
For underfrequer,cy, the de7ay is set so that the ttwa r$ quired for e signal to reach the Reactor trip breakers after the Underfrequency Trip Setpoint is rencoed shall L
L not exceed 0.3 second. On decreasing power tir) Urdervoltage and l
Underfrequency Reactor Coolant Pump Bus trips are automatically biccked by P-7 (a power level of approximately 10% ef RATED THERMAL POWER with a turbine first stage chamber presure at 4 proximately 10%
g of full power eat,valtnt!; and on increasing pwer, reinstatod j
automatically by P-7.
l-Turbine Trio A Turbine trip initiates a Reactor trip. On decreasing power the G
Reactor trip from the Turbine trip is automatically blocked by P-9 (a power le 01 of approximately 50% of RATED THERMAL POWER) and on increasi,J power, reinstated automatically by F-9.
COMANCHE PEAK - UNIT 1 B 2-7 DRAFT
", to TXX-89614 Paga 4 of 20 LIMITING SAFETY SYSTEM SETTINGS BASES' Safety iniection Inout from ESF If a Reactor trip has not already been generated by the Reactor Trip System instrumentation, the ESF-automatic actuation logic channels will initiate a Reactor trip upon any signal which initiates a Safety Injection. The ESF instrumentation channels which initiate a Safety Injection signal are shown in Table 3.3-2.
Reactor Trio System Interlocks The Reactor Trip System Interlocks perform the following functions:
P-6 On increasing power,. P-6 allows the manual block of the Source Range trip (i.e., prevents premature block of the Source Range trip), provides a backup block for Source Range Neutron Flux doubling, and de-energizes the high voltage to the detectors. On decreasing power, Source Range Level trips are automatically reactivated and high voltage restored.
P-7 On increasing power, P-7 automatically enables Reactor trips G
on low flow in more than one reactor coolant loop, reactor coolant pump bus undervoltage and underfrequency, pressurizer low pressure and pressurizer high level. On decreasing power, the above listed trips are automatically blocked.
i P-8 On increasing power, P-8 automatically enables the Reactor trip on low flow in one reactor coolant loop. On decreasing power, the P-8 automatically blocks the reactor trip on low flow in one recctor coelcnt loop.
P-9 Da increasing pcwer, P-g automatically enables Reactor trip on Turbine trip. On decreasing power, P 3 satomatically
!G l
blocks Reactor trip on Turbine trip.
I P-10 On increasing power. P-10 allcws the manual block of the l
IrJerradiate Rttpe trip and the Low Setpoint Power Range l
trip; e.ed automatically blocks the Source Range trip and de-1 energites the Source Range hign voltage power. On decreasing power, the latermediate Range trip and the Low Setpoint icuer Range trip are automatically reactivated.
Dro~ ides input to v
P-7.
P-13 Turbine first stage chamber pressure provides input to P-7.
COMANCHE PEAK - UNIT 1 B 2-8 DRAFT
f.
Enclosura 2-to TXX-89614 l'*'
Page 5'of 20 o
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. to TXX-89614 Page 6 of 20 TABLE 3.3-1 (Continued)-
TABLE NOTATIONS a
When the reactor trip breakers are in the closed position and C
the Control Rod Drive System is capable of rod withdrawal.
b Below the P-6 (Intermediate Range Neutron Flux Interlock)
O' Setpoint.
c Below the P-10 (Low Setpoint Power Range Neutron Flux D
Interlock) Setpoint d
Above the P-7 (At-Power) Setpoint.
D e
The applicable MODES and ACTION statements for these channels D
noted in Table 3.3-2 are more restrictive and therefore, applicable, f
Above the P-8 (3-loop flow permissive) Setpoint.
D 9
Above the P-7 and below the P-8 Setpoints.
D h
The boron dilution flux doubling signals may be blocked D
during reactor startup.
i Above the P-9 (Reactor trip on Turbine trip Interlock)
G Setpoint.
ACTION STATEMENTS ACTION 1 -
With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or be in HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
ACTION 2 -
With the number of OPERABLE channels one less than the Total Number of Channels, STARTUP and/or POWER OPERATION may proceed provided the following conditions are satisfied:
a.
The inoperabi~ channel is placed in the tripped condition within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, i
b.
The Minimum Channels OPERABLE requirement is met; however, the inoperable channel may be bypassed for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for surveillance testing of other channels per Specification 4.3.1.1, and l
COMANCHE PEAK - UNIT 1 3/4 3-5 DRAFT
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4 EnclosurG.2 to TXX-89614.-
i LPage 9'of'20 TABLE 4.3-1-(Continued)
TABLE NOTATIONS a Only if the reactor trip breakers happen to b'e in the closed position and the Control Rod Drive System is capable of rod withdrawal.
C b Below P-6 (Intermediate Range Neutron Flux Interlock) Setpoint.
c Below P-10 (Low Setpoint Power Range Neutron Flux Interlock)
- Setpoint, d Above the P-7 (At Power) Setpoint.
e Above the P-9.(Reactor trip on Turbine trip Interlock)
Setpoint.
G (1)
If not performed in previous 31 days.
(2) Comparison a calorimetric to excore power and N-16 power indication above 15% of RATED THERMAL POWER.
Adjust excore channel and/or N-16 channel gains consistent with calorimetric power if absolute difference of the respective channel is greater than 2L The provisions of Specification 4.0.4 are not applicable to entry into MODE 1 or 2.
(3) Single point comparison of incore to excore AXIAL FLUX DIFFERENCE above 15% of RATED THERMAL POWER.
Recalibrates if the absolute difference is greater than or equal to 3L For -
the purpose of these surveillance requirements, "M" is defined as at least once per 31 EFPD. The provisions of Specification 4.0.4 are not applicable for entry into MODE 1 or 2.
(4) Neutron and N-16 detectors may be excluded from CHANNEL CALIBRATION.
(5) Detector plateau curves shall be obtained and evaluated.
For the Intermediate Range Neutron Flux, Power Range Neutron Flux and N-16 channels the provisions of Specification 4.0.4 are not applicable for entry into MODE 1 or 2.
(6)
Incore-Excore Calibration above 75% of RATED THERMAL POWER.
D 3
For the purpose of these surveillance requirements "Q" is defined as at least once per 92 EFPD. The provisions of l
Specification 4.0.4 are not applicable for entry into MODE 1 i
or 2.
(7)
(NOT USED)
F I
COMANCHE PEAK - UNIT 1 3/4 3-12 DRAFT
,c T/S Page (as revised)
Group Description
-Table 2.2-1 2
See Sheet No(s):3 Technical Specification: Safety Limits and Limiting Safety System Settings - Reactor Trip System Instrumentation Trip Setpoints Addition:
The current CPSES reactor trip system design includes a reactor trip following a turbine trip whenever the plant is above 10% power (P 7 permissive interlock).
However, for plants with a 50% or greater load rejection capability, a reactor trip is unnecessary below 50% power if the cause of the turbine trip is readily correctable. At CPSES both Units are designed each with a 50% load rejection capability (i.e. 40%
steam dump capacity and 10% control rod insertion).
Taking advantage of this feature. CPSES is implementing a design modification that will eliminate a reactor trip on turbine trip when the plant is at or below 50%
power. A P-9 permissive interlock is being installed to accomplish the above functions. In addition to the above capabilities. it is anticipated that implementation of P-9 will reduce the down time required to restart the plant, thereby increasing-plant availability. Also, the P-9 modification is part of the CPSES ongoing Trip Reduction Program.
FSAR Change Request Number: TS89-090.1 1
Related SER Section: 7.2.1 f
SER/SSER Impact: Yes SER Section 7.2.1 Reactor Trip System Description (p.7-7) states "[a] turbine trip causes a direct reactor trip above 105 power (P-7 interlock)."
Table 3.3-1 2
See Sheet No(s):3 and page 3/4 3-5 Technical Specifications: Limiting Conditions for Operations - Reactor Trip System Instrumentation Revision:
Revised the LC0's to reflect the implementation of P-9 for the turbine trip and reactor trip systems interlock functions.
FSAR Change Request Number: TS89-090.3 i
Related SER Section: 7.2.1 SER/SSER Impact: Yes See SER/SSER Impact for TS 090.1.
Table 4.3-1 2
See Sheet No(s):3. 4 and page 3/4 3-12 Technical Specifications: Surveillance Requirements Reactor Trip System Instrumentation Revision:
Revised the surveillance requirements for turbine trip and reactor trip system interlocks to reflect implementation of P-9.
FSAR Change Request Number: TS89-090.4
E: closure 2 to TXX-89614. CPSES T/S REVISION Page 11 of 20 DETAILED DESCRIPTION Page 2 T ~
Je/S Page
'5 s revised)
Grous Description Related SER Section: 7.2.1 SER/SSER Impact: Yes See SER/SSER Impact for TS-89-090.1 8 2-7, 8 2
Technical Specificottons: Safety Limits and Limiting Safety Systek Settines - Bases Revision:
Revised the Limiting Safety System Settings Bases to reflect the taplementation of P 9. A Reactor trip will automatically be blocked at or below 50% power and reinstated above 50% power.
FSAR Change Request Number: TS89-090.2 Related SER Section: 7.2.1 SER/SSER Impact: Yes See SER/SSER Impact for TS89-090.1-I I
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Undervoltage and UEderfrecuency - Reactor Cnolant Pump Busses The Undervoltage and Underfrequency Reactor Coolant Pump Bus trips pro-vide core protection against DNS as a result of complete loss of forced coolant flow.
The specified setpoints assure a Reactor trip signal is generated before the Low Flow Trip Setpoint is reached. Time delays are incorporated in the Underfrequency and Undervoltage trips to prevent spurious Reactor trips from momentary electrical power transients.
For undervoltage, the delay is set so that the time required for a signal to reach the Reactor trip breakers following the simultaneous trip of two or more reacter coolant pump bus circuit breakers shall not exceed 1.2 seconds.
For underfrequency, the delay is set so that the time required for a signal to reach the Reactor trip breakers after the Underfrequency Trip Setpoint is reached shall not exceed 0.3 second.
On decreasing power the Undervoltage and Underfrequency Reactor Coolant Pump Bus trips are automatically blocked by P-7 (a power level of approximately 10%
of RATED THERMAL POWER with a turbine first stage chamber pressure at approximately 10% of full power equivalent); and on increasing power, reinstated automatically
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by P-7.
Turbine Trip A Turbine trip initiates a Reactor trip.
On decreasing ~ power the Reactor trip from the Turbine trip is automatically blocked by P-39(a power level of Ts.stc40 approximatelyJe% of RATED THERMAL POWER); and on increasing power, reinstated automatically by t't.
So F M
Safety Injection Input from ESF If a Reactor trip has not already been generated by the Reactor Trip System instrumentation, the ESF automatic actuation logic channels will initiate a Reactor trip upon any signal which initiates a Safety Injection.
The ESF instrumentation channels which initiate a Safety Injection signal are shown in Table 3.3-2.
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' Pa9tIMT6G2%AFETY SYSTEM SETTINGS BASES l#
6/ndervoltageandUn'derfrequency-ReactorCoolantPumpBusses(Continue
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Reactor Trip System Interlocks The ileactor Trip System interlocks perform the following functions:
P-6 On increasing power P-6 allows the manual block of the Source Range trip (i.e., prevents premature block of Source Range trip), provides a backup block for Source Range Neutron Flux doubling, and deener-gizes the high voltage to the detectors.
On decreasing power, Source Range Level trips are automatically reactivated and high voltage restored.
P-7 On increasing power P-7 automatically enables Reactor trips on low flow in more than one reactor coolant loop, reactor coolant pump g,gg.go -
bus undervoltage and underfrequency, tur;ir.; tri;;, pressurizer low pressure and pressurizer high level.
On decreasing power, the above listed trips are automatically blocked.
P-8 On increasing power, P-8 automatically enables the Reactor trip on low flow in one reactor coolant loop.
On decreasing power, the P-8 automatically blocks the reactor trip on low flow in one reactor coolant loop.
+
P-10 On increasing power, P-10 allows the manual block of the Intermediate Range trip and the Low Setooint Power Range trip; and automatically blocks the Source Range trip and de-energizes the Source Range high voltage power.
On decreasing power, the Intermediate Range trip and the Low Setpoint Power Range trip are automatically reactivated.
4 Provides input to P-7.
P-13 Turbine first stage chamber pressure provides input to P-7.
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Enclosuro'2 to'TXX-89614 Page 17 of 20 TABLE 3.3-1 (Continued)
TABLE NOTATIONS aOnly if the reacter trip breakers happen to be in the closed position and the Control Rod Drive System is capable of rod withdrawal.
bBelow the P-6 (Intermediate Range Neutron Flux Interlock) Setpoint.
CBelow the P-10 (Low Setpoint Power Range Neutron Flux Interlock) Setpoint.
dAbove the P-7 (At Power) Setpoint
'The applicable MODES and A h h tatements for these channels noted in W
Table,3.3-2 are more restrictive and therefore, applicable.
I 3
AbovetheP-8(3-loopflowpermissive)[etpoint.
9Above the P-7 and below the P-8 tpoints.
hThe boron dilution flux. doubling signals may be blocked during reactor startup.
1.%.g p.ct (?caetec +c ? on h ** +M M'~ low Se+Poin t Ts-89-o%
ACTION STATEMENTS ACTION 1 - With the number of OPERABLE channels one less than the Minimum Channels OPERABLE requirement, restore the inoperable channel to OPERABLE status within 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> or be in HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
ACTION 2 - With the number of OPERABLE channels one less than the Total Number of Channels, STARTUP and/or POWER CPERATION may proceed provided the following conditions are satisfied:
a.
The inoperable channel is placed in the tripped condition within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, b.
The Minimum Channels OPERABLE requirement is met; however, the inoperable channel may be bypassed for up to 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> for surveillance testing of other channels per Specification 4.3.1.1, and c.
Either, THERMAL POWER is restricted to less than or equal to 75% of RATED THERMAL POWER and the Power Range Neutron i
Flux Trip Setpoint is reduced to less than or equal to 85% of RATED THERMAL POWER within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />; or, the QUADRANT POWER TILT RATIO is monitored at least once per 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> per Specification 4.2.4.2.
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Enclosure Eto TXX-89614
~
- ' ~ v Paga D of 20 TABLE 4.3-1 (Continued)
TABLE NOTATIONS fo5r&
a y-Only if the reactor trip breakers happen to be in the closed and the Control Rod Drive System is capable of rod withdrawal.
A bBelow P-6 (Intermediate Range Neutron Flux Interlock) Setpoint.
C8elow P-10 (Low Setpoint Power Range Neutron Flux Interlock) Setpoint.
dAbove the P-7 (At Power) Setpoint.
e Above. 4 P-9 (Reac+er Mp u h b*me Mp I4er tocx ) 5+pnt.
Ts49-mo (1) If not performed in previous 31 days.
(2) Comparison of calorimetric to excore power and N-16 power indication above 15% of RATED. THERMAL POWER.
Adjust excore channel and/or N-16 channel gains consistent with calorimetric power if absolute difference of the respective channel is greater than 2%. The provisions of Specification 4.0.4 are not applicable to entry into MODE 1 or 2.
(3) Single point comparison of incore to excore AXIAL FLUX DIFFERENCE above 15% of RATED THERMAL POWER.
Recalibrates if the absolute difference is greater than or equal to 3%.
For the purpose of these surveillance requirements "M", is defined as at least once per 31 EFPD.
The provisions of Specification 4.0.4 are not applicable for entry into MODE 1 or 2.
(4) Neutron and N-16 detectors may be excluded from CHANNEL CALIBRATION.
I (5) Detector plateau curves shall be ootained and evaluated.
For the Intermediate Range Neutron Flux, Power Range Neutron Flux and N-16 channels the provisions of Specification 4.0.4 are not applicable for entry into MODE 1 or 2.
l (6) Incore - Excore Calibration, above 75% of RATED THERMAL POWER.
For the
{
purpose of these surveillance requirements "Q" is defined as at least a
once per 92 EFPD.
The provisions of Specification 4.0.4 are not applic-able for entry into MDDE 1 or 2.
(7) Each train shall be tested at least every 62 days on a STAGGERED TEST BASIS.
(8) The surveillance frequency and/or MODES specified for these channels in Table 4.3-2 are more restrictive and therefore applicable.
(9) Quarterly surveillance in MODES 3", 4, and S shall also include verifica-8 a
g' tion that permissives P-6 and P-10 are in their required state for exist-ing plant conditions by observation of the permissive annunciator window.
f Quarterly surveillance shall include verification of the Boron Dilution 1
/
/
Alarm Setpoint of less than or equal to an increase of twice the count r)._
/,
rate within a 10-minute ; e :.1 a.w,b M --* # f p :' u r j
nw g - % ". period. TH
="
gean s i;..,
9; a
~.
. k
\\ l-COMANCHE PEAK - UNIT 1 3/4 3-11
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