ML20195K003
| ML20195K003 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 11/23/1998 |
| From: | FLORIDA POWER CORP. |
| To: | |
| Shared Package | |
| ML20195J992 | List: |
| References | |
| NUDOCS 9811250142 | |
| Download: ML20195K003 (43) | |
Text
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l FLORIDA POWER CORPORATION CRYSTAL RIVER UNIT 3 i DOCKET NO. 50-302/ LICENSE NO. DPR-72 ATTACIIMENT D LICENSE AMENDMENT REQUEST #241, REVISION 0 HIGH PRESSURE INJECTION SYSTEM MODIFICATIONS l
Proposed Revised ITS and ITS Bases Pages l
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9811250142 981123 P; PDR '
P ADOCK 05000302{.2 PDR
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i RPS Instrumentation 3.3.1 I l
Table 3.3.1 1 (page 1 of 1)
Reactor Protection System Instrunentation J
l APPLICABLE CON 0lfl0NS MODES OR REFERENCED l
OTHER FROM SPECIFIED REQUIRED j
SURVEILLANCE ALLOWA8LE FUNCTION CONDITIONS ACTIch D.1 REQUIREMENTS VALUE 1
- 1. Nuclear Overpower -
- a. High Setpoint 1,2(a) t SR 3.3.1.1 s 104.9% RTP SR 3.3.1.2 SR 3.3.1.5 SR 3.3.1.7
- b. Low Setpoint 2(b) 3(b)
, G SR 3.3.1.1 s 5% RTP 4(b) ,.
$a 3 s i 5 l
- 2. RCS High outlet Tenterature 1,2 F SR 3.3.1.1 s 618'F SR 3.3.1.4 i SR 3.3.1.6 l
- 3. RCS High Pressure 1,2 F SR 3.3.1.1 s 2355 psig l'
SR 3.3.1.4 SR 3.3.1.6 l SR 3.3.1.7
- 4. RCS Low Pressure 1,2(a) F SR 3.3.1.1 a 1900 psig SR 3.3.1.4 l '
ER 3.3.1.6 SR 3.3.1.7
- 5. RCS Variable Lou Pressure 1,2(*) F SR 3.3.1.1 m (11.59
- T t SR 3.3.1.4 5037.8) psYo * --
- 6. Reactor Building High 1,2,3(C)
Pressure F SR 3.3.1.1 s 4 psig SR 3.3.1.4 SR 3.3.1.6
- 7. Reactor Coolant Pump Power 1,2(8) F SR 3.3.1.1 More than one pump Monitor (RCPPM) SP 3.3.1.4 drawing a 1152 or j SR 3.3.1.6 y,14,400 kW i SR 3.3.1.7 l
- 8. Nuclear Overpower RCS Flow 1,2I *) F SR 3.3.1.1 Nuclear Overpower RCS and Measured AXIAL POWER SR 3.3.1.3 Flow and AX1AL POWER IMBALANCE SR 3.3.1.5 IMBALANCE setpoint SR 3.3.1.6 envelope in COLR SR 3.3.1.7
- 9. Main Turbine Trip (Control a 45% RTP H SR 3.3.1.1 2 45 psig Olt Pressure) SR 3.3.1.4 SR 3.3.1.6
- 10. Loss of Both Main Feedwater a 20% RTP I SR 3.3.1.1 a 55 psto Punps (Control Oil SR 3.3.1.4 Pressure) SR 3.3.1.6
'# 11. Shutdown Bypass RCS High 2(D) 3(D)
, G SR 3.3.1.1 s 1820 psig Pressure SR 3.3.1.4 l 4(b) 5(b)
, SR 3.3.1.4 (a) When not in shutdown bypass operation.
.(b) During shutdown bypass operation with any CRD trip breakers in the closed position and the CR0 Control System (CRDCS) capable of rod withdrawal.
(c) ' With any CRD trip breaker in the closed position and the CRDCS capable of rod withdrawal.
Crystal River Unit 3 3.3-5 Amenciment No.
l l- ESAS Instrumentation 3.3.5 ACTIONS l CONDITION REQUIRED ACTION COMPLETION TIME l
C. (continued)
C.2 --------NOTE---------
L Only required for RCS i Pressure--Low l l
Parameter.
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Reduce RCS pressure 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> 1
< 1800 psig. l j AND !
C.3 --------NOTE---------
Only required for RCS Pressure--Low Low l Parameter. 1 Reduce RCS pressura 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />
< 900 psig.
1 6HQ C.4 --------NOTE---------
Only required for Reactor Building Pressure High setpoint and High High Parameter.
Be in MODE 4. 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> l
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Crystal River Unit 3 3.3-13 Amendment No.
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l ESAS Instrumentation 3.3.5 Tabte 3.3.5-1 (page 1 of 1)
Engineered Safeguards Actuation System Instrumentation i
APPLICA8LE l PARAMETER MODES OR OTHER SPECIFIED CONDITIONS ALLOWABLE (
VALUE k 14. Reactor Coolant System Pressure -Low ' a 1800 psig a 1625 pois
!- l l.
[. 2. Reactor Coolant System Pressure-Low Low . 't 900 pels a 500 pois l
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- 3. Reactor' Building Pressure - High < '1,2,3 s 4 psig !
- . 4 Reactor Building Pressure- High High 1,2,3 s 30 pais l
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(. .. Crystal River Unit .3 3.3-15 Amendment No.
l ECCS-Operating 3.5.2 i i
SURVEILLANCE REQUIREMENTS l SURVEILLANCE FREQUENCY SR 3.5.2.1 Verify each ECCS manual, power operated, 31 days and automatic valve in the flow path, that '
l is not locked, sealed, or otherwise secured in position, is in the correct position.
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SR 3.5.2.2 Verify each ECCS pump's developed head at In accordance l
the test flow point is greater than or with the 1 equal to the required developed head. Inservice Testing Program SR 3.5.2.3 Verify each ECCS automatic valve in the 24 months flow path that is not locked, sealed, or otherwise secured in position, actuates to the correct position on an actual or :
l simulated actuation signal.
l SR 3.5.2.4 Verify each ECCS pump starts automatically 24 months !
on an actual or simulated actuation signal.
SR 3.5.2.5 . Verify correct position for valves 24 months throttled in the HPI flow path that are locked, sealed or otherwise secured in position.
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' (continued) l l
l l Crystal River Unit 3 3.5-5 Amendment No.
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RPS Instrumentation B 3.3.1-BASES-BACKGROUND Channei Bv'oass (continued) contacts from the other channels with the channel bypass relay. If any contact is open, the second channel cannot be bypassed. The second condition is the closing of the key switch. When the bypass relay is energized, the bypass contact closes, maintaining the channel trip relay .in an energized condition. All RPS trip logics are reduced to a two-out-of-three logic in channel bypass.
Shutdown Bvoass During plant cooldown, it is desir able to maintain the safety rods withdrawn to provide shutdown capabilities in the event of unusual-positive reactivity additions -
(moderator dilution, etc.). However, if the safety rods are withdrawn too soon following reactor shutdown as RCS pressure ~is decreased, an RCS Low Pressure trip will occur at 1900 psig and the rods will re-insert into the core. To l avoid this, the protection system allows the operator to bypass the low pressure trip and maintain shutdown capabilities.
During the cooldown and depressurization, the safety rods are inserted prior to the low pressure trip of 1900 psig.
The RCS pressure is decreased to less than 1820 psig, then each RPS channel is placed in shutdown bypass.
In shutdown bypass, a normally closed contact opens and the operator closes the shutdown bypass key switch in each RPS channel. This action bypasses the RCS Low Pressure trip, Nuclear Overpower RCS Flow and Measured AXIAL POWER IMBALANCE trip, Reactor Coolant Pump overpower /underpower trip, and the RCS Variable Low Pressure trip, and inserts a new RCS High Pressure, 1820 psig trip. The operator can now l withdraw the safety rods for additional available reactivity insertion.
The insertion of the new high pressure trip performs two functions. First, with a trip setpoint of 1820 psig, the l bistable prevents operation at normal system pressure, 2155 psig, with a portion of the.RPS bypassed. The second function is to ensure that the bypass is removed prior to normal operation. When the RCS pressure is increased during (continued)
Crystal River Unit 3 B 3.3-8 Amendment No.
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RPS Instrumentation B 3.3.1 BASES BACKGROUND Shutdown'Bvoass -(continued) g a plant heatup, the safety rods are inserted prior to reaching 1820 psig. The shutdown bypass is removed, which l returns the RPS to normal, and system pressure is increased to greater than 1900 psig. The safety rods can then be l withdrawn and remain at the full out condition for the rest of the heatup.
In addition to the Shutdown Bypass RCS High Pressure trip, the nuclear overpower high flux t.-ip setpoint is administratively reduced to 5% RTP while the RPS is in shutdown' bypass. This provides a backup to the Shutdown Bypass RCS High Pressure trip and allows low temperature physics testing while preventing the generation of any significant amount of power.
Module Interlock and Test / Interlock Trio Relav Each channel and each trip module is. capable of being I individually tested. When a module is placed into the test mode or is removed from the system, it causes the test /
interlock trip relay to de-energize and to indicate an RPS channel trip. Under normal conditions, the channel to be tested _is placed in bypass before a module is tested. This ensures the channel trip relay remains energized during testing and the channel does not trip.
1 APPLICABLE Each of the analyzed accidents and transients can be SAFETY ANALYSES, detected by one or more RPS Functions. The accident LCO, and- analysis contained in Chapter 14 of the FSAR takes credit APPLICABILITY for most RPS trip Functions. Functions not specifically credited in the accident analysis were qualitatively credited in the safety evaluation report (SER) written for the CR-3 operating license. Functions not specifically credited include high RB pressure, high RCS temperature, main turbine trip, shutdown bypass-RCS pressure high, and loss of both main feedwater pumps. ,
The LC0 requires all instrumentation performing an RPS Function to be OPERABLE. Failure of any instrument renders the affected channel (s) inoperable and reduces the 4
(continued)
Crystal River Unit 3 B 3.3-9 Amendment No.
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RPS Instrumentation B 3.3.1 BASES
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APPLICABLE 11. Shutdown Byoass RCS Hiah Pressure (continued)
- ' SAFETY ANALYSES, !
LCO, and During shutdown bypass operation with the Shutdown APPLICABILITY Bypass RCS High Pressure trip active with a setpoint 3 of s 1820 psig and the Nuclear Overpower-Low Setpoint l !
set at or below 5% RTP, the trips listed below can be !
bypassed. Under these conditions, the Shutdown Bypass i RCS High Pressure trip and the Nuclear Overpower-Low '
Setpoint trip prevent conditions from reaching a point where actuation of these Functions would be required.
1.a Nuclear Overpower-High Setpoint; l
- 4. RCS Low Pressure;
- 5. RCS Variable Low Pressure;
- 7. Reactor Coolant Pump Power Monitors; and
- 8. Nuclear Overpower RCS Flow and Measured AXIAL POWER IMBALANCE.
The Shutdown Bypass RCS High Pressure Function's Allowable Value is selected to ensure a trip occurs before producing THERMAL POWER.
The RPS satisfies Criterion 3'of the NRC Policy Statement.
In MODES 1 and 2, the following trips shall be OPERABLE. These trips are designed to rapidly make the reactor subcritical in cider to protect the SLs during A00s and to function along with the ESAS to -
provide acceptable consequences during accidents, l i
1.a Nuclear Overpower-High Setpoint; i
- 2. RCS High Outlet Temperature;
- 3. RCS High Pressure;
- 4. RCS Low Pressure; '
- 5. RCS Variable Low Pressure; (continued)
Crystal. River Unit' 3 B 3.3-20 Amendment No.
t o ~ w m -9 % e ,--m s -
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RPS Instrumentation B 3.3.1 BASES APPLICABLE
-SAFETY ANALYSES,
- 11. Shutdown Bvoass RCS Hiah Pressure (continued)
LCO, and 7. Reactor Coolant Pump Over/Under Power; and
~ APPLICABILITY
- 8. Nuclear Overpower RCS Flow and Measured AXIAL POWER IMBALANCE.
Functions 1, 4, 5, 7, and 8 may be bypassed in MODE 2 or below (higher numerical MODE) when RCS pressure is below l
1820 psig, provided the Shutdown Bypass RCS High Pressure j '
and the Nuclear Overpower-Low setpoint trip are placed in operation. Under these conditions, the Shutdown Bypass RCS High Prossure trip and the Nuclear Overpower-Low setpoint trip prevent conditions from reaching a point where actuation of these Functions is necessary.
Two other Functions are required to be OPERABLE during portions of. MODE.1. These are the Main Turbine Trip ,
(Control Oil Pressure) and the loss of Main Feedwater Pumps l (Control Oil Pressure) trip. These Functions are required ~
to be OPERABLE above 45% RTP and 20% RTP, respectively. !
Analyses presented in BAW-1893 (Ref. 5) showed that for 1 operation below these power levels, these trips are not necessary to minimize challenges to the PORVs as required by NUREG-0737 (Ref. 4).
Because the only safety function of the RPS is to interrupt power to the CONTROL RODS, the RPS is not required to be OPERABLE in MODE 3, 4, or 5 if the reactor trip breakers are open, or the CRDCS is incapable of rod withdrawal.
Similarly, the RPS is not required to be OPERABLE in MODE 6 when the CONTROL RODS are decoupled from the CRDs. However, in MODE 2, 3, 4, or 5, the Shutdown Bypass RCS High Pressure and Nuclear Overpower-Low Setpoint trip Functions are required to be OPERABLE if the CRD trip breakers are closed and the CRDCS is capable of rod withdrawal. Under these conditions, the Shutdown Bypass RCS High Pressure and Nuclear Overpower-Low setpoint trips are sufficient to prevent an approach to conditions that could challenge SLs.
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(continued)
~ Crystal River Unit 3 B 3.3-21 Amendment No.
! ESAS Instrumentation B 3.3.5 BASES BACKGROUND related safeguards equipment will not inhibit the overall ES (continued) Functions. Where a motor operated or a solenoid operated valve is driven by either of two matrices, one is from actuation train A and one from actuation train B. Redundant i ES pumps are controlled from separate and independent i
i actuation channels.
Enaineered Safety Feature Actuation System Bvoasses No provisions are made for maintenance bypass of ESAS instrumentation channels. Operational bypasses are i
provided, as discussed below, to allow accident recovery actions to continue and, to allow plant cooldown without l spurious ESAS actuation.
The ESAS RCS pressure instrumentation channels include ,
permissive bistables that allow manual bypass when reactor i pressure is below the point at which the low and low low pressure trips are required to be OPERABLE. Once permissive ;
conditions are sensed, the RCS pressure trips may be I manually bypassed. Bypasses are automatically removed when bypass permissive conditions are no longer applicable.
No more than two (of the three) High RB Pressure channels may be manually bypassed after an actuation. The manual bypass allows operators to take manual control of ES Functions after initiation to allow recovery actions.
B_eactor Coolant System Pressure i RCS pressure is monitored by three independent pressure transmitters located in the RB. These transmitters are separate from the transmitters that provide an input to the Reactor Protection System (RPS). Each of the pressure signals generated by these transmitters is monitored by four bistables to provide two trip signals, at 1625 psig and 500 psig, and two bypass permissive signals, at 1800 psig and 900 psig.
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Crystal Riyer Unit 3 B 3.3-46 Amendment No.
ESAS Instrumentation B 3.3.5 BASES BACKGROUND Reactor Coolant System Pressure (continued)
The outputs of the three channels trip bistables, associated with the low RCS pressure (1625 psig) actuate bistable trip I auxiliary relays in two sets (actuation trains A and B) of identical and independent trains. The two HPI trains each use three logic channels arranged in two-out-of-three coincidence networks. The outputs of the three bistables associated with the Low Low RCS Pressure (500 psig) actuate bistable trip auxiliary relays in two sets (actuation trains A and B) of identical and independent trains. The two LPI trains each use three logic channels arranged in two-out-of-three coincidence networks for LPI Actuation. The outputs of the three Low Low RCS Pressure bistables also trip the automatic actuation relays, via a LPI bistable trip auxiliary relay, in the corresponding HPI train as previously described.
Reactor Buildina Pressure ESAS RB pressure signal information is provided by 12 pressure switches. Six pressure switches are used for the High RB Pressure Parameter, and six pressure switches are used for the High-High RB Pressure Parameter.
The output contacts of six High RB Pressure switches are used in two sets of identical and independent actuation trains. These two trains each use three logic channels.
The outputs of these channels are used in two-out-of-three coincidence networks. The output contacts of the six RB pressure switches also trip, via a pressure switch trip auxiliary relay, the automatic actuation relays in the corresponding HPI and LPI trains as previously described.
The output contacts of six High High RB Pressure switches 1 are used in two sets of identical and independent actuation trains. The outputs of the High High RB Pressure switches are used in two-out-of-three coincident networks for RB Spray Actuation. The two-out-of-three logic associated with each RB Spray train actuates spray pump operation when the High-High RB signal and the HPI signal are coincident in that train.
l (continued)
Crystal River Unit 3 8 3.3-47 Amendment No.
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ESAS Instrumentation B 3.3.5 l
' BASES l
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l LC0 1. Reactor Buildina Pressure-Hiah Setooint I (continued) 1
.The RB Pressure-High Setpoint Allowable Value s 4 psig i was selected to be low enough to detect a rise in RB Pressure that would occur due to a small break LOCA,
- thus ensuring that the RB high pressure actuation of the safety systems will occur for a wide spectrum of break sizes. The trip setpoint also causes the RB coolers to shift to low speed (performed as part of the HPI logic) to prevent damage to the cooler fans l L
due to the increase in the density of the air steam ;
mixture present in the containment following a LOCA.
i l 2. Reactor Buildina Pressure-Hiah Hiah Setooint i
L The RB Pressure-High High Setpoint Allowable Value s 30 psig was chosen to be high enough to avoid {
j actuation during an SLB, but also low enough to ensure j a timely actuation during a large break LOCA.
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APPLICABILITY The ESAS instrumentation for each Parameter is required to ;
be OPERABLE during the following MODES- and specified conditions.
l l 1. ' Reactor Coolant System Pressure-Low Setooint L The RCS Pressure-Low Setpoint actuation Parameter i
shall be OPERABLE during operation above 1800 psig. l l .This ensures the capability to automatically actuate l
safety systems and components during conditions indicative of a LOCA or SLB. Below 1800 psig, the low l l RCS Pressure actuation Parameter can be bypassed to 1 i avoid actuation during normal cooldown when safety '
system actuations are not required.
l The allowance for the bypass is consistent with the plant transition to a lower energy state, providing
- greater margins to core and containment limits. The response to any event, given that the reactor is
-already shut down, will be less severe and allows sufficient time for operator action to provide manual safety system actuations. This is even more appropriate during plant heatup from an outage when
, the RCS energy content is low.
i-(continued)
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Crystal River Unit 3 B 3.3-51 Amendment No.
I
r-ESAS Instrumentation B 3.3.5 L BASES
-APPLICABILITY. 1. Reactor Coolant System Pressure-low Setooint (continued)
To ensure the RCS Pressure-Low trip is not bypassed j
'. when required to be OPERABLE by the safety analysis, l each channel's bypass removal bistable must be set i with a setpoint of s 1800 psig. The bypass removal l does not need to function for accidents initiated from l RCS Pressures below the bypass removal setpoint. I
- 2. Reactor Coolant System Pressure-low low SetDoint 1
- The RCS Pressure-tow Low Setpoint actuation Parameter shall be OPERABLE during operation above 900 psig.
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' This ensures the capability to automatically actuate i safety systems and components during conditions indicative of a LOCA. Below 900 psig, the low low RCS I t Pressure actuation Parameter can be bypassed to avoid i j actuation during normal plant cooldown. l The allowance for the bypass is consistent with plant transition to a lower energy state, providing greater !
l margins to core and containment limits. The response j
- to any event, given that the reactor is already tripped, will be less severe and allows sufficient ;
time for operator action to provide manual safety 1 l system actuations. This is even more appropriate i L
during heatup from an outage when the RCS energy i content is low. l l To ensure the RCS Pressure-Low Low trip is not
! bypassed when assumed OPERABLE by the safety analysis, each channel's bypass removal bistable must be set with a setpoint of s 900 psig. The bypass removal does not need to function for accidents initiated by l- RCS Pressure below the bypass removal setpoint.
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l (continued) 4 I Crystal River Unit 3 B 3.3-52 Amendment No. I i
ESAS Instrumentation B 3.3.5 L
BASES ACTIONS Bd (continued)
Condition B applies when one required instrumentation channel in one or more RB Pressure Parameters becomes inoperable. If one required channel is inoperable, placing it in a tripped Condition leaves the affected actuation train in-one-out-of-one condition for actuation and the other actuation channel in a two-out-of-two condition (making the worst case assumption the third channel in each
' actuation train is not OPERABLE). In this condition, if another'RB Pressure ESAS channel were to fail, the ESAS instrumentation could still perform its actuation function.
i For RB Pressure Parameters, all affected pressure switch L
trip auxiliary relays must be tripped to comply with this Required Action. This is norrr. ally accomplished by tripping the affected pressure switch test switch.
The 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> Completion Time is based on engineering judgment l and is sufficient time to perform the Required Action.
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C.1. C.2. C.3. and C.4 t
[
If Required Actions A.1 or B.1 cannot be met within the '
associated Completion Time, the plant must be placed in a 1 MODE in which the LC0 does not apply. To achieve this status, the plant must be placed in at least MODE 3 within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and, for the RCS Pressure-Low Parameter, to < 1800 l psig, for the RCS Pressure-Low Low Parameter, to
< 900 psig, and for the RB Pressure High Parameter and High High Parameter, to MODE 4 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full t
L power conditions in an orderly manner and without challenging plant systems.
SURVEILLANCE All ESAS Parameter instrumentation listed in Table 3.3.5-1 REQUIREMENTS are subject to CHANNEL CHECK, CHANNEL FUNCTIONAL TEST, CHANNEL CALIBRATION, and response time testing.
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1 (continued)
Crystal River Unit 3 B 3.3-54 Amendment No.
E
( ,_ _
r PAM Instrumentation B 3.3.17 BASES The following table identifies the specific instrunent tag nunbers for PAM instrunentation identified in Table 3.3.17 1.
FUNCTION CHANNEL A CHANNEL B
- 1. Wide Range Neutron Flux N1 15-NI 1 or NI 15 N!R NI 14 N!-1
- 2. RCS Hot leg Temperature RC-4A T14-1 RC-4B-flR1
- 4. Reactor Coolant RC-163A LR1 (Hot leg level) and RC 1638-LR1 (Hot leg level) and Inventory RC 164A LR1 (Vessel Head level) RC 1648 LR1 (Vessel Head level)
- 5. Borated Water Storage DH 7 L1 or DH-7 LIR1 DH 37 L1 Tank Level
- 6. High Pressure Injection A1: MU-23 F18-1 A1: MU 23 F112 Flow A2: MU 23 Fito A2: MU 23 FI6 1 B1: MU 23 FI9 B1 MU 23 FIS-1 82: MU-23 F17 1 82: MU 23-FI11
- 8. Containment Pressure BS 16 P! BS-17 PI (Expected Post Accident Range)
- 10. Containment Isolation ES Light Matrix "A": AHV 18/1C; ES Light MatriA "B": AHV 1A/10; j valve Position CAV-1/3/4/5/126/429/430/433/434; CAV 2/6/7/431/432/435/436; '
CFV-11/12/15/16; LRV 70/72; MUV- CFV-29/42;LRV-71/73;MUV-258 thru -261/567;WDV 3/60/94/ 18/27/49/253; WDV 4/61/62/405; 406; WSV 3/5/28 thru 31/34/35/ WSV 4/6/26/ l 42/43 27/32/33/38/39/40/41 ES Light Matrix "AB": CFv 25 thru 28; Civ 34/35/40/41; DW-160; i MSV 130/148; SWV-47 thru 50/79 thru 86/109/110 1
- 12. Containment Hydrogen WS 11-CR WS 10 CR Concentration
- 13. Pressurizer level RC 1 LIR 1 RC 1 LIR 3
- 14. Steam Generator Water OTSG A: SP 25 L11 or OTSG A: SP-26-L11 Level (Startup Range) SP-25-LIR OTSG B: SP-30 LII OTSG B: SP-29 L11 or SP T).LIR (continued)
Crystal River Unit 3 8 3.3-125A Amendment No.
PAM Instrumentation B 3.3.17 BASES l
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. LCO ' -
- 5. Borated Water Storaae Tank (BWST) level l (continued) .
i BWST inventory is monitored by level instrumentation with a span of 0 to 50 feet. Redundant monitoring :
capability is provided by three independent level measurements. Two level transmitters provide input to l control . room indicators, and one of these channels is recorded in the control room. The control room indications are the primary indications used by the operator. Therefore, the LC0 deals specifically with l this portion of the instrument string. '
i During a design basis LOCA, the Reactor Building Spray, low Pressure Injection (LPI) and High Pressure l Injection (HPI) Systems are automatically aligned to :
obtain suction from the BWST. As the BWST inventory '
is pumped into the RCS and containment, coolant will be lost through the break and will accumulate in the reactor building sump. The operator is required to I switch LPI and RB Spray suction to the reactor
' building emergency sump from the BWST when the BWST level reaches a s)ecified level setpoint. At this same time if the ICS pressure is greater than the LPI pump shutoff head, it will also be necessary to switch the suction of the HPI pumps to the discharge of the LPI pumps to ensure the capability to inject flow to the RCS since the HPI pumas do not have the capability of drawing coolant from tie sum). BWST level is a Type A variable because it is tie primary. indication used by the operator to determine when to initiate the switch-over to sump recirculation. This operator l action is necessary to satisfy the long-term core L
cooling requirements specified in 10 CFR 50.46.
l 6. HPI Flow (Low Ranoel HPI flow instrumentation is provided for verification and long term monitoring of HPI flow. HPI flow is determined from differential pressure transmitters.
Two channels in each of the four injection lines, for a total of eight low range indicators, provides this indication. One transmitter is calibrated to a range of 0-200 gpm. Each differential pressure measurement t
provides an input to a control room indicator. Since
- t. the operator. relies on the control room indication following an accident, the LCO deals with this portion of the instrument string.
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(continued)
[. Crystal River Unit 3 8 3.3-129 Amendment No.
e PAM Instrumentation ,
B 3.3.17 i BASES I
i LCO 6. HPI Flow (Low Ranae) (continued)
Although 4 high range flow indicators (0-500 gpm) readout on the main control board, they are not used to accomplish any safety functions. The HPI lines will have preset throttle valves, .stop check valves, and crosstie lines to: (1) create the desired flow l distribution through the HPI lines for LOCA core cooling; (2) ensure adequate cooling flow to the HPI pump mechanical seals; and (3) prevent HPI pump flow :
from exceeding 600 gpm (maximum HPI pump flow rate l assumed in design calculations associated with !
Emergency Diesel Generator loading, ECCS pump available NPSH, and makeup' tank-(MUT-1) allowable overpressure versus level).
- 7. Containment Sumo Water Level (Flood Level)
Containment sump water level (Flood) is monitored by two channels of level indication, both of which are displayed in the control room on edgewise level indicators. Channel A and B sump flood level indication are recorded in the associated 'A' and 'B' EFIC Rooms. Each instrument encompasses a range of ;
0-10 feet above the sump and provides information to l the operator related to gross leakage in the Reactor ;
Building. This leakage may be indication of degradation in the reactor coolant pressure boundary )
(RCPB) which would require further investigation and 1 action. These instruments are not assumed to provide ;
information required by the operator to take a ;
mitigation action specified in the accident analysis. !
As such, they are not Type A variables. However, the monitors are deemed risk significant (Category 1) and j are included within the LC0 based upon this i consideration.
(continued)
\
Crystal River Unit 3 B 3.3-130 Amendment No.
ECCS-Operating l
B 3.5.2
! BASES L APPLICABLE LOCA, the RCS depressurizes as primary coolant is ejected j SAFETY ANALYSIS ~through the' break into the containment. The nuclear (ccntinued) reaction is terminated either by moderator voiding during 1
.large breaks or CONTROL R00 assembly insertion for small breaks. Following depressurization, emergency cooling water {
1 is injected into the reactor vessel core flood nozzles, then !
flows into the downcomer, fills the lower plenum, and j refloods the core. .
l The LC0 ensures that an ECCS train will deliver sufficient l water to match decay heat boiloff rates soon enough to I minimize core uncovery for_ a large break LOCA. It also ;
ensures that the HPI pump will deliver sufficient water for j a small break LOCA and provide sufficient boron to maintain l L the core subcritical following the small break.LOCA or an SLB. l In the LOCA analyses, HPI and LPI are not credited until 35 seconds after actuation of the ESAS signal. This is based on a loss of offsite power and the associated time l delays in startup and loading of the emergency diesel
! generator (EDG). Further, LPI flow is not credited until l RCS pressure drops below the pump's shutoff head. For a-large break LOCA, HPI is not credited at all.
l_ The.ECCS trains. satisfy Criterion 3 of the NRC Policy l_ Statement.
l
!: j LC0 In MODES 1, 2, and 3, two independent (and redundant) ECCS
! trains are required to ensure that at least one is available, assuming a single active failure in the other l train. With one ECCS train inoperable, the system is still I capable of mitigating an event, providing a concurrent single failure does not occur. Hence, the 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> ACTION addressing a loss of redundancy is appropriate.
l L
L I
i-j (continued)
Crystal River Unit 3 B 3.5-12 Amendment No.
ECCS-Operating B 3.5.2 BASES
, \
LC0 l Not all portions of the HPI flow path satisfy the l
l (continued) independence criteria discussed above. Specifically, the l l HPI flow path downstream of the HPI/ Makeup pumps is not l- separable into two distinct trains, and is therefore, not i independent. This conclusion is based upon analysis which l I L shows, that in the event of a postulated break in the HPI l injection piping, ' injection flow is required through a minimum of three (3) injection legs, assuming one pump- i operation, or through a minimum of two (2) injection' legs, L
assuming two HPI pump operation. When considering the impact of:inoperabilities in this portion of the system, the; l same concept of maintaining single active failure protection. !
l' must be applied. When components become inoperable, an i
assessment of the HPI systems ability to perform'its safety i function must be performed. If the system can continue to l l
' perform its safety function,-without assuming a single active failure, then the 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> loss cf redundancy ACTION i is appropriate. -If the inoperability renders the system, as
! is, incapable of performing its safety function, without I postulating a single active failure, then the plant is in a '
condition outside the safety analysis and must enter LCO 3.0.3 immediately, i i In MODES 1, 2, and 3, an ECCS train consists of an HPI t
subsystem and an LPI subsystem. Each train includes the piping, instruments, and controls'to ensure an OPERABLE flow path capable of taking suction from the BWST upon an ESAS signal and manually transferring suction to the reactor building emergency sump.
During an event requiring ECCS actuation, a flow path is
! provided to ensure an abundant supply of water from the BWST to the RCS via the HPI and LPI pumps and'their respective discharge ficw paths to each of the four cold leg injection nozzles and the reactor vessel. In the long term, this flow
! path may be manually transferred to take its supply from the reactor building emergency sump and to supply its flow to the RCS via two paths, as described in the Background section.
L The flow path for each train must maintain its designed degree of independence to ensure that no single active 1 failure can disable both ECCS trains.
l I
i l (continued) ,
Crystal River Unit 3 B 3.5-13
(.
Amendment No.
i
ECCS-Operating B 3.5.2 BASES 1
gRVg SR 3.5.2.5 i (continued) grificgtion of thg positigs of vgives throttged in tgg HPI l sg!Y! an khE$n iO!Oua$ hgs 15!es * $nthnan a e[ !
hQ e hknhe i rd :
fkeprovide er hSoOoi!
$hego$tI s ! n ces!! SN gectSon po Ots a$ak e kbprovi ana!!gitbetwgenptS0$ekeve"ohkoa E f'ow kle 0 $$a$ " kS$ nk0r abhh!thc00$ih$$$ 0 hr!m xc!NbSn[b0 wb!ntheSs $sNIkt ib$m0m iso h$$e ahSO 0"in e$i aOulait $$ $$$t YY h E and akeb $0[fMUk-L$aS0wa e00e !!ur v!rsus koSked,rath de seakSd, ! $heor otherwlse!eS5abIl$
e!5*0" v !'$ a$"are y*0h* a secured n posit'on. SR 3.5.2.6 l f thot$ h"kOe!"w$$$au$ mat a [co0So belh$th$n CS esSuh dN! k! b
$h r a L8EA?'"th a41 rte"""t I e$!abSSt equipment.y (!nb m $g op!"akSSg !p h OSch o tNe 00$
SR 3.5.2.7 i Oct$ n inSeE" ens 0"e $ hat $tl! u0he Yr t "S $$[S$ay5 Y On E$e OS!d k "hehh [sSu0N1$nSPudN"$0e a00N Eo tk ko!SE15n " S$s* q" fkS!Onh0nYkobe
" " g" cS h hby o h$ king esp $ en!S REFERENCES 1. 10 CFR 50.46.
- 2. FSAR, Section 6.1.
hc $0nb0 nk!r in 0 ecemer$,NN' th[CO$h6EbCE' l Components,
- 4. American Society of Mechan' cal Eng< neers, Boiler and l kn! pes [$o A fi $ lW N00" ' "
l
- 5. FTI 51-L266138-01 Safet I Team Satety Assess, ment. y Analysis Input to Startup a
- 6. FSAR, Section 4.3.10.1.
l Crystal River Unit 3 B 3.5-18 Amendment No. NOTE - Valid Until Cycle 12 Only
i FLORIDA POWER CORPORATION CRYSTAL RIVER UNIT 3 LX)CKET NO. 50-302/ LICENSE NO. DPR-72 1 ATTACHMENT E l LICENSE AMENDMENT REQUEST #241, REVISION 0 HIGH PRESSURE INJECTION SYSTEM MODIFICATIONS l 1 Proposed Revised ITS and ITS Bases Pages - Strikeout Version j l 1 l l l l 1 4 1
' I
(: RPS Instrumentation 3.3.1 Table 3.3.1-1 (page 1 of 1) Reactor Protection System Instrunentation APPLICABLE CONDITIONS NODES OR REFERENCED OTHER FROM SPECIFIED REQUIRED SURVElLLANCE ALLOWABLE FUNCTION CONDITIONS ACTION D.1 REQUIREMENTS VALUE
- 1. Nuclear Overpower-
- e. High Setpoint 1,2(a) F SR 3.3.1.1 5 104.9% RTP SR 3.3.1.2 SR 3.3.1.5 SR 3.3.1.7
- b. Low setpoint 2(b) 3(b)
, G SR 3.3.1.1 s 5% RTP 4(b) $(b)
- 2. RCS High Outlet Tenperature 1,2 F SR 3.3.1.1 s 618'F SR 3.3.1.4 SR 3.3.1.6 3
- 3. RCS Nigh Pressure 1,2 F SR 3.3.1.1 s 2355 palg SR 3.3.1.4 SR 3.3.1.6 3 SR 3.3.1.7
- 4. RCS Low Pressure 1,2(a) F SR 3.3.1.1 a 4400 j M pois SR 3.3.1.4 SR 3.3.1.6 SR 3.3.1.7 3
- 5. RCS Variable Low Pressure 1,2(*) F SR 3.3.1.1 a (11.59
- TMt*
SR 3.3.1.4 5037.8) psie SR 3.3.1.6
- 6. Reactor Building High 1,2,3(C)
Pressure F SR 3.3.1.1 s 4 psis SR 3.3.1.4 SR 3.3.1.6
- 7. Reactor Coolant Pump Power 1,2(a) F SR 3.3.1.1 More than one pump Monitor (RCPPM) ' SR 3.3.1.4 drawing a 1152 or i
SR 3.3.1.6 1 14,400 kW SR 3.3.1.7
- 8. Nuclear Overpower RCS Flow 1,2(a) F SR 3.3.1.1 Nuclear overpower RCS and Measured AX1AL POWER SR 3.3.1.3 Flow and AXIAL POWER IMBALANCE SR 3.3.1.5 IMBALANCE setpoint SR 3.3.1.6 envelope in COLR l SR 3.3.1.7 l
- 9. Main Turbine Trip (Control a 45% RTP H SR 3.3.1.1 a 45 psig 011 Pressure) SR 3.3.1.a.
SR 3.3.1.6
- 10. Loss of Both Main Feedwater a 20% RTP I SR 3.3.1.1 a 55 psis Ptaps (Control ot t SR 3.3.1.4 Pressure) SR 3.3.1.6
-11. Shutdown Bypass RCS High 2(b) 3(b)
, G SR 3.3.1.1 s 4MO 1829pois Pressure SR 3.3.1.4 4(b) 5(b)
, SR 3.3.1.6 (a) When not in shutdown bypass operation.
(b) During shutdown bypass operatior,with any CRD trip breakers in the closed position and the CRD Control System (CRDCS) capable of rod withdrawal. (c) With any CRD trip breaker in the closed position and the CRDCS capable of rod withdrawal. Crystal River Unit 3 3.3-5_ Amendment No. 449
.-...__..~...__.. .. _ _ _ _ _ _ _ . _ _ _ _ . . _ _ . . . _ _ . . _ _ . _ _ . _ . _ . . _ _ _ _ . . . _ . _ _ _ _ _ l: - I' ESAS Instrumentation 3.3.5
'A_CTIONS l CONDITION REQUIRED ACTION COMPLETION TIME L
i :C. (continued) C.2 --------NOTE--------- L Only required for RCS Pressure--Low l Parameter. ! Reduce RCS pressure 12 hours
< 4700 180gpsig.
f M C.3 --------NOTE--------- Only required.for RCS ' Pressure--Low Low ' l Parameter. Reduce RCS pressure 12 hours '
< 900 psig.
E
--------NOTE---------
t C.4 Only required for Reactor Building Pressure High setpoint and High High Parameter. Be in MODE 4. 12 hours L l c ;, l-
- Crystal River Unit 3 3.3-13 Amendment No. 449 l
ESAS Instrumentation 3.3.5 Table 3.3.5-1 (page 1 of 1) Engineered Safeguards Actuation System Instrumentation l APPLICABLE MODES OR OTHER SPECIFIED ALLOWABLE PARAMETER CONDITIONS VALUE
- 1. Reactor Coolant System Pressure-Low a 43g0 @ psig a 4500 $ psig
- 2. Reactor Coolant System Pressure- Low Low a 900 psis a 500 pels 3-
-3. Reactor Building Pressure-High 1,2,3 s 4 pelg
- 4. Reactor Building Pressure-High High 1,2,3 s 30 pais Crystal River Unit 3' 3.3-15 Amendment No. M9
! 1 l ECCS-Operating i 3.5.2 l 1 l SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY ! SR 3.5.2.1 Verify each ECCS manual, power operated, 31 days ' and automatic valve in the flow path, that is not locked, sealed, or otherwise secured in position, is in the correct position. SR 3.5.2.2 Verify each ECCS pump's developed head at in accordance the test flow point is greater than or l with the equal to the required developed head. Inservice Testing Program l i SR 3.5.2.3 Verify each ECCS automatic valve in the 24 months flow path that is not locked, sealed, or otherwise secured in position, actuates to the correct position on an actual or simulated actuation signal. SR 3.5.2.4 Verify each ECCS pump starts automatically 24 months on an actual or simulated actuation signal. SR 3.5.2.5 Verify the ccrrect :ctting: cf step: for 24 months the fcllcWir.g MP1 stcp check vcivc;;
- c. MUV 2
- b. MUV 6
- c. MUV 10 Vi^flTyfd3FFEEtToiifi 6ElfdEI^VITVes thrsttisd finIthe:LHPlifisw; pathithaWsFs
]cikedissialndib@theywi,selsfjbredlin~
ppsjtJpnj (continued) l l Crystal River Unit 3 3.5-5 Amendment No. M9 t
RPS Instrumentation B 3.3.1 BASES BACKGROUND Channel Byoass (continued) contacts from the other channels with the channel bypass relay. If any contact is open, the second channel cannot be bypassed. The second condition is the closing of the key switch. When the bypass relay is energized, the bypass contact closes, maintaining the channel trip relay in an energized condition. All RPS trip logics are reduced to a two-out-of-three logic in channel bypass. Shutdown Bvoass During plant cooldown, it is desirable to maintain the ! safety rods withdrawn to provide shutdown capabilities in the event of unusual pocitive reactivity additions (moderator dilution, etc.). However, if the safety rods are withdrawn too soon following reactor shutdown as RCS pressure.is decreased, an RCS Low Pressure trip will occur at -18001900 psig and the rods will re-insert into the core. To avoid'IhTs, the protection system allows the operator to bypass the low pressure trip and maintain shutdown capabilities. 1 During the cooldown and depressurization, the safety rods are inserted prior to the low pressure trip of 4400 1900 psig. The RCS pressure is decreased to less than 4Me^' 20 psig, then each.RPS channel is placed in shutdown bypas}s}. , In shutdown bypass, a normally closed contact opens and the operator closes the shutdown bypass key switch in each RPS channel. Th h action bypasses the RCS Low Pressure trip, Nuclear Overpower RCS Flow and Measured AXIAL POWER IMBALANCE trip, Reactor Coolant Pump overpower /underpower trip, and the KCS Variable Low P.ressure trip, and inserts a i new RCS High Pressure, H2O 1820 psig trip. The operator I can now withdraw the safety F6ds for additional available I reactivity insertion. The insertion of the new high pressure trip performs two functions. First, with a trip setpoint of 4 M G 1820 psig
~^
the bistable prevents operation at normal system pFissure, , 2155 psig, with a portion of the RPS bypassed. The second function is to ensure that the bypass is removed prior to normal operation. When the RCS pressure is increased during (continued) Crystal River Unit 3 8 3.3-8 @Tdisjp=i i= No. 7
( RPS Instrumentation B 3.3.1 BASES
)
BACKGROUND Shutdown 8voass (continued) a plant heatup, the safety rods are inserted prior to reaching U 20 1820 psig. The shutdown bypass is removed, which returns'thT RPS to normal, and system pressure is increased to greater than M00 1900 psig. The safety rods can then be withdrawn and remaiii^it the full out condition for the rest of the heatup. In addition to the Shutdown Bypass RCS High Pressure trip, the nuclear overpower high flux trip setpoint is administratively reduced to 5% RTP while the RPS is in shutdown bypass. This provides a backup to the Shutdown Bypass RCS High Pressure trip and allows low temperature physics testing while preventing the generation of any significant amount of power. Module Interlock and Test / Interlock Trio Relav Each channel and each trip module is capable of being individually tested. When a module is placed into the test mode or is removed from the system, it causes the test / interlock trip relay to de-energize and to indicate an RPS channel trip. Under tarmal conditions, the channel to be tested is placed in bypass before a module is tested. This ensures the channel trip relay remains energized during testing and the channel does not trip. APPLICABLE Each of the analyzed accidents and transienti can be SAFETY ANALYSES, detected by one or more RPS Functions. Tne accident LCO, and analysis contained in Chapter 14 of the fSAR takes credit APPLICABILITY for most RPS trip Functions. Functions not specifically credited in the accident analysis were qualitatively credited in the safety evaluation report (SER) written for the CR-3 operating license. Functions not specifically credited include high R8 pressure, high RCS temperature, main turbine trip, shutdown bypass-RCS pressure high, and loss of both main feedwater pumps. The LC0 requires all instrumentation performing an RPS Function to be OPERABLE. Failure of any instrument renders the affected channel (s) inoperable and reduces the (continued) l Crystal River Unit 3 B 3.3-9 @@MRcthkr, No. 7
-.. . . - _ - . - - . . - - . - - - . - ~ . . - - - - . - -
RPS Instrumentation B 3.3.1 ,_ BASES l l l APPLICABLE 11. Shutdown Bvoass RCS Hiah Pressure l SAFETY ANALYSES, (continued) LCO, and During shutdown bypass operation with the Shutdown APPLICABILITY Bypass RCS...H.igh Pressure trip active with a setpoint of 5 W201820 psig and the Nuclear Overpower-Low Setpoint sit ~it or below 5% RTP, the trips listed below can be bypassed. Under these conditions, the l Shutdown Bypass RCS High Pressure trip and the Nuclear Overpower-Low Setpoint trip prevent conditions from reaching a point where actuation of.these Functions would be required. 1.a Nuclear Overpower-High Setpoint;
- 4. RCS Low Pressure;
- 5. RCS Variable Low Pressure;
- 7. Reactor Coolant Pump Power Monitors; and i
- 8. Nuclear Overpower RCS Flow and Measured AXIAL I POWER IMBALANCE. j l
l i The Shutdown Bypass RCS High Pressure Function's ' Allowable Value is selected to ensure a trip occurs before producing THERMAL POWER. The RPS satisfies Criterion 3 of the NRC Policy Statement. In MODES 1 and 2, the following trips shall be I 0PERABLE. These trips.are designed to rapidly make { the reactor subcritical in order to protect the SLs i during A00s and to function along with the ESAS to ; provide acceptable consequences during accidents, l.a Nuclear Overpower-High Setpoint;
- 2. RCS High Outlet Temperature;
- 3. .RCS High Pressure;
- 4. RCS Low Pressure;
- 5. RCS Variable. Low Pressure;
- (continued) l Crystal River Unit 3 B 3.3-20 @~$m]e {"=i
- icr. No. 7
. ~ . . . - . . _ - _. . _ _ _ ._, , ._ -
.. - .-.- . , .. .- .. - . - . - . .. - - - . - . - - - - . - - - .. - ~ .- -
p l L RPS Instrumentation l B 3.3.1 l BASES l APPLICABLE 11. Shutdown Bvoass RCS Hiah Pressure (continued)- SAFETY ANALYSES, i LCO, and 7. Reactor Coolant Pump Over/Under Power; and l
' APPLICABILITY
- 8. Nuclear Overpower RCS Flow and Measured AXIAL POWER IMBALANCE.
l Functions ~1, 4, 5, 7, and 8 may be bypassed in MODE 2 or j below (higher numerical MODE) when RCS pressure is below 1 1 MBO1820psig,providedtheShutdownBypassRCSHigh ! Pressure and the Nuclear Overpower-Low setpoint. trip are placed in operation. Under these conditions, the Shutdown Bypass RCS High Pressure trip and the Nuclear Overpower-Low
.setpoint trip prevent ccnditions from reaching a point where actuation of these Functions is necessary.
, Two other Functions are required to be OPERABLE during i portions of MODE 1. These are the Main Turbine Trip l (Control Oil Pressure) and the loss of Main Feedwater Pumps (Control Oil Pressure) trip. These Functions are required [ to be OPERABLE above 45% RTP.and 20% RTP, respectively. Analyses presented in BAW-1893 (Ref. 5) showed that for operation below these power levels, these trips are not l necessary to minimize challenges to the PORVs as required by NUREG-0737 (Ref. 4). I Because the only safety function of the RPS is to interrupt i power to the CONTROL RODS, the RPS is not required to be ' OPERABLE in MODE 3, 4, or 5 if the reactor trip breakers are open, or the CRDCS is incapable of rod withdrawal. l Similarly, the RPS is not required to be OPERABLE in MODE 6 when the CONTROL = RODS are decoupled from the CRDs. However, in MODE 2, 3, 4, or 5, the Shutdown Bypass RCS High Pressure and Nuclear Overpower-Low Setpoint trip Functions are required to be OPERABLE if the CRD trip breakers are L L closed and the CRDCS is capable of rod withdrawal. Under these conditions, the Shutdown Bypass RCS High Pressure and Nuclear Overpower-Low setpoint trips are sufficient to ' prevent an approach to conditions that could challenge SLs. I (continued) L Crystal River Unit 3 B 3.3-21 M@@{"= b h , No. 7
ESAS Instrumentation B 3.3.5 BASES BACKGROUND related safeguards equipment will not inhibit the overall ES (continued) Functions. Where a motor operated or a solenoid operated valve is driven by either of two matrices, c..e is from actuation train A and one from actuation train B. Redundant ES pumps are controlled from separate and independent actuation channels. ) Enaineered Safety Feature Actuation System Bypasses ! No provisions are made for maintenance bypass of FSAS 1 instrumentation channels. -Operational bypasses are i provided, as discussed below, to allow accident recovery F actions to continue and, to allow plant cooldown without i spurious ESAS actuation. + ) The ESAS RCS pressure instrumentation channels include permissive bistables that allow manual bypass when reactor pressure is below the point at which the -low and low low pressure trips are required to be OPERABLE. Once permissive conditions are sensed, the RCS pressure trips may be i manually bypassed, Bypasses are automatically removed when bypass permissive conditions are no longer applicable. No more than two (of the three) High RB Pressure chcnnels may be manually bypassed after an actuation. The manual bypass allows operators to take manual control of ES Functions.after initiation to allow recovery actions. Reactor Coolant System Pressure RCS pressure is monitored by three independent pressure transmitters located in the RB. These transmitters are separate from the transmitters that provide an input to the Reactor Protection System (RPS). Each of the pressure signals generated by these transmitters is monitored'by fcur 62$ psig..ard bistables totwo 500 psig, and provide bypasstwo trip signals, permissive at M001~ MOO 1800 signals, at psig and 900 psig. (continued)
. Crystal River Unit 3-- B 3.3-46 KMif,:vi:i= No. H
]
l:
- . _. . _ = - -
ESAS Instrumentation B 3.3.5 BASES l BACKGROUND Reactor Coolant System Pressure (continued) The outputs of the three channels trip bistables, associated with the low RCS pressure (M001625 psig) actuate bistable trip auxiliary relays in two seti'(ictuation trains A and B) i of identical and independent trains. The two HPI trains each use three logic channels arranged in two-out-of-three coincidence networks. The outputs of the three bistables associated with the Low Low RCS Pressure (500 psig) actuate bistable trip auxiliary relays in two sets (actuation trains A and B) of identical and independent trains. The two LPI trains each use three logic channels arranged in two-out-of-three coincidence networks for LPI Actuation. The outputs of the three Low Low RCS Pressure bistables also trip the automatic actuation relays, via a LPI bistable trip auxiliary relay, in the corresponding HPI train as ; previously described. 1 Reactor Buildina Pressure ESAS RB pressure signal information is provided by 12 pressure switches. Six pressure switches are used for the High RB Pressure Parameter, and six pressure switches ! are used for the High-High RB Pressure Parameter. 1 The output contacts of six High RB Pressure switches are used in two sets of identical and independent actuation trains. These two trains each use three logic channels. The outputs of these channels are used in two-out-of-three coincidence networks. The output contacts of the six RB pressure switches also trip, via a pressure switch trip auxiliary relay, the automatic actuation relays in the l corresponding HPI and LPI trains as previously described. The output contacts of six High High RB Pressure switches are used in two sets of identical and independent actuation trains. The outputs of the High High RB Pressure switches are used in two-out-of-three coincident networks for RB Spray Actuation. The two-out-of-three logic associated with each RB Spray train actuates spray pump operation when the High-High RB signal and the HPI signal are coincident in that train. l (continued) l Crystal River Unit 3 B 3.3-47 Q'dg@Revi:ica No M
ESAS Instrumentation B 3.3.5 I BASES l LC0 1. Reactor Buildina Pressure-Hiah Setooint (continued) l ' The RB Pressure-High Setpoint Allowable Value s 4 psig was selected to be low enough to detect a rise in RB Pressure that would occur due to a small break LOCA, i thus ensuring that the RB high pressure actuation of the safety systems will occur for a wide spectrum of l break sizes. The trip setpoint also causes the RB coolers to shift to low speed (performed as part of l the HPI logic) to prevent damage to the cooler fans i due to the increase in the density of the air steam mixture present in the containment following a LOCA.
- 2. Reactor Buildina Pressure-Hiah Hiah Setooint The RB Pressure-High High Setpoint Allowable Value s 30 psig was chosen to be high enough to avoid actuation during an SLB, but also low enough to ensure i l
a timely actuation during a large break LOCA. 1 APPLICABILITY The ESAS instrumentation for each Parameter is required to be OPERABLE during the following MODES and specified conditions.
- l. Reactor Coolant System Pressure-Low Setooint The RCS Pressure-Low Setpoint actuation Parameter '
shall be OPERABLE during operation above N GO 1806 psig. This ensures the capability to automatiEally actuate safety systems and components during conditions indicative of a LOCA or SLB. Below WOG
'1800 psig, the low RCS Pressure actuation Parameter Eaii~be bypassed to avoid actuation during normal cooldown when safety system actuations are not required.
The allowance for the bypass is consistent with the plant transition to a lower energy state, providing greater margins to core and containment limits. The response to any event, given that the reactor is already shut down, will be less severe and allows sufficient time for operator action to provide manual safety system actuations. This is even more appropriate during plant heatup from an outage when the RCS energy content is low. (continued) i j Crystal River Unit 3 B 3.3-51 @f@[tRevi:icn No. 44
ESAS Instrumentation , B 3.3.5 BASES APPLICABILITY 1. Reactor Coolant System Pressure-low Setooint (continued) To ensure the RCS Pressure-Low trip is not bypassed when required to be OPERABLE by the safety analysis, each channel's bypass remov_al_ bistable must be set _ with a setpoint of s M 00 1800 psig. _The bypass removal does not need to fd6Etion for accidents ' initiated from RCS Pressures below the bypass removal setpoint.
- 2. Reactor Coolant System Pressure-Low Low Setooint The RCS Pressure-Low Low Setpoint actuation Parameter ,
shall be OPERABLE during operation above 900 psig. This ensures the capability to automatically actuate safety systems and components during conditions indicative of a LOCA. Below 900 psig, the low low RCS Pressure actuation Parameter can be bypassed to avoid actuation during normal plant cooldown. The allowance for the bypass is consistent with plant transition to a lower energy state, providing greater margins to core and containment limits. The response to any event, given that the reactor is already tripped, will be less severe and allows sufficient- ' time for operator action to provide manual safety system actuations. This is even more appropriate during heatup from an outage when the RCS energy content is low. To ensure the RCS Pressure-Low Low trip is not bypassed when assumed OPERABLE by the safety analysis, each channel's bypass removal bistable must be set with a setpoint of s 900 psig. The bypass removal does not need to function for accidents initiated by RCS Pressure below the bypass removal setpoint, i I~ l-(continued) I Crystal River Unit 3 B 3.3-52 @ @ @Revi:ir. No. M i
f. ESAS Instrumentation i B 3.3.5 l l l BASES i l ACTIONS fL1 l (continued) Condition B applies when one required instrumentation , l channel in one or more RB Pressure Parameters becomes '
- inoperable. If one required channel is inoperable, placing I
it in a tripped Condition iraves the affected actuation train in one-out-of-one condition for actuation and the l other actuation channel in a two-aut-of-two condition l (making the worst case assumption the third channel in each , actuation train is not OPERABLE). In this condition, if l another RB Pressure ESAS channel were to fail, the ESAS ' instrumentation could still perform its actuation function. For RB Pressure Parameters, all affected pressure switch trip auxiliary relays must be tripped to comply with this Required Action. This is normally accomplished by tripping the affected pressure switch test switch. I The 72 hour Completion Time is based on engineering judgment ' and is sufficient time to perform the Required Action. C.l. C.2. C.3. and C.4 If Required Actions A.1 or B.1 cannot be met within the associated Completion Time, the plant must be placed in a MODE in which the LC0 does not apply. To achieve this status, the plant must be placed in at least MODE 3 within l 6 hours. and, for the RCS Pressure-Low Parameter, to 1
< WOO 1800 psig, for the RCS Pressure-Low Low Parameter, to < 900~~piig, and for the RB Pressure High Parameter and High High Parameter, to MODE 4 within 12 hours. The allowed Completion Times are reasonable, based on operating experience, to reach the required unit conditions from full power conditions in an orderly manner and without l challenging plant systems.
SURVEILLANCE All ESAS Parameter instrumentation listed in Table 3.3.5-1 REQUIREMENTS are subject to CHANNEL CHECK, CHANNEL FUNCTIONAL TEST, CHANNEL CALIBRATION, and response time testing. (continued) i Crystal River Unit 3 8 3.3-54 @[@MRevi:icr, No. -14 l
Y l PAM Instrumentation B 3.3.17 ; BASES 1 I 1 The following table identifies the specific instrunent tag runbers for PAM instrunentation identified in Table 3.3.17-1. FUNCTION CHANNEL A CHANNEL 8
- 1. Wide Range Neutron Flux N!-15 NI-1 or NI 15 NIR HI 14-N1-1
- 2. RCS Hot leq Tenperature RC 4A-TI4 1 RC-48 TIR1
- 3. RCS Pressure (Wide Range) RC-158-P!2 or RC 158 PIR RC-159 Pl2 4 Reactor Coolant RC-163A-LR1 (Hot leg level) and RC-1638-LR1 (Hot leg level) and inventory RC-164A-LR1 (Vessel Head level) RC-1648-LR1 (Vessel Head level)
- 5. Borated Water Storage DH 7 L1 or DH-7 LIR1 DH-37 LI Tank Level 1' ll
- 6. High Pressure Injection A1: MU 23 FIB-1 Flow A1: MU 23 F112 A2: MU 23-FI10 A2: MU 23 FI6-1 B1: MU-23-FI9 81: MU 23-F15-1 B2: MU 23 F17-1 B2: MU-23 Ft11
- 7. Containment surp Water WD 303 L1 or WD-303-LR WD-304 LI or WD 304-LR tevel (Flood Level)
- 8. Containment Pressure BS-16 PI BS-17-PI (Expected Post-Accident Range)
- 9. Contairenent Pressure BS 90 PI or BS 90-PR BS-91 PI or BS-91-PR (Wide Range)
- 10. Containment Isolation ES Light Matrix "A": AHV 18/1C; ES Light Matrix "B": AHV-1A/10; Valve Position CAV-1/3/4/5/126/429/430/433/434; CAV 2/6/7/431/432/435/436; CFV 11/12/15/16; LRV-70/72; MUV- CFV 29/42;LRV-71/73;MUV-258 thru -261/567;WDV 3/60/94/ 18/27/49/253; WDV 4/61/62/405; 406; WSV-3/5/28 thru -31/34/35/ WSV 4/6/26/
42/43 27/32/33/38/39/40/41 ES Light Matrix "AB": CFV-25 thru 28; CIV 34/35/40/41; DWV-160; MSV 130/148; "'" "; SW 47 thru 50/79 thru 86/109/110
- 11. Contairment Area RM-G29 RI or RM G29 RIR RM G30 R1 Radiation (High Range)
- 12. Containment Hydrogen W1 11 CR WS-10 CR Concentration
- 13. Pressurizer Level RC-1 LIR-1 RC 1 LIR-3
- 14. Steam Generator Water OTSG A: SP 25 L11 or OTSG A: SP 26 L11 Level (Startup Range) SP 25 LIR OTSG B: SP 30 L11 OTSG B: SP 29 L11 or SP 29 LIR (continued)
Crystal River Unit 3 8 3.3-125A giiisiidi@Revicica No. W l
I PAM Instrumentation B 3.3.17 BASES LC0 5. Borated Water Storaae Tank (BWST) level (continued) BWST inventory is monitored by level instrumentation with a span of 0 to 50 feet. Redundant monitoring capability is provided by three independent level measurements. Two level transmitters provide input to control room indicators, and one of these channels is recorded in the control room. The control room indications are the primary indications used by the operator. Therefore, the LCO deals specifically with this portion of the instrument string. During a design basis LOCA, the Reactor Building Spray, low Pressure Injection (LPI) and High Pressure Injection (HPI) Systems are automatically aligned to obtain suction from the BWST. As the BWST inventory is pumped into the RCS and containment, coolant will be lost through the break and will accumulate in the reactor building sump. The operator is required to switch LPI and RB Spray suction to the reactor building emergency sump from the BWST when the BWST level reaches a s)ecified level setpoint. At this same time if the RCS pressure is greater than the LPI pump shutoff head, it will also be necessary to switch the suction of the HPI pumps to the discharge of the LPI pumps to ensure the capability to inject flow to the RCS since the HPI pum)s do not have the capability of drawing coolant from tie sump. BWST level is a Type A variable because it is the primary indication used by the operator to determine when to initiate the switch-over to sump recirculation. This operator action is necessary to satisfy the long-term core cooling requirements specified in 10 CFR 50.46.
- 6. HPI Flow (Low Ranae) i flPI Tfl697115 t?udfiliH6n3sT W6Vi^difff6EVsFi fEstT65 !
andfjongltermimonitoringfofijP ! determined from differential ressure p.ljflo(~HPI^flarii^~^~ transmitters. l Two channels in each of the four injection lines, for a total of eight low range indicators, provides this indication. One transmitter is calibrated to a range of 0-200 gpm. Each differential pressure measurement provides an input to a control room indicator. Since the operator relies on the control room indication following an accident, the LC0 deals with this portion of the instrument string. ) 1
) I (continued) I Crystal River Unit 3 B 3.3-129 piindiiie@cdden No. 44 1
PAM Instrumentation B 3.3.17
- BASES LC0 6. HPI Flow flow Ranae) (continued)
I Although 4 high range flow indicators (0-500 gpm) readout on the main control board, they are not used to accomplish cperatica c;n notanychiev safety functions. runcut re- Twc HPI[uftEcir suction 50er c. H0xcver, single HP!:rdle:: while in : piggy b :k ecnfiguratien, c:n pump [cration,
- ..itvc runcut. 0 cre cr thrcttle HPI ficw te revent runcut using the %siHPIsl:inssisin 1 hase;iiresejew g:n-- instruments.tS thro ttisMal Vesil
' rosstiedlinesstog )?creathitheldesired;flowl~"
c distributionsthroug ithe1HPI ninesiforYLOCAicore j
'c001ingW(2)iensureE dequateic6ollingsf1owit6stheiltPI pumpj me ch anic alXsh al sM and ;(3 )ipprevent% HPI? pumpifl ow f romiexc eed ingT600f gpm M ma nimum ;HPli pumpFfl owirate-~
h s s umed sin %de s i gnW al cul at i 6ns es soci a teh"W)th"~ EmepgencyiDiese10 Generator %1oadi~o',nECCSipsep" n avai}able]NPSHeindim PVefpresgurgyersusg.akeup[tankg(ROJT , 1)iaggillj s ev_eljj In addition t =cnitoring for and preventin- HPI ump runcut, :: described beve, the low r:nge fiew r d e in trument; provide indication cf the required
" cver the flew rate: cf interest to determine
- curaiinei'01tioncriteriahavebeen::t.
if HPI Cert in HPI line break: -)
-inch break:) re-(dcuble ended uire i 01: tionrupture 2nd the HP.
of the faulted larger Iinete;;;ur ; 3cquate HPI flew i; injected inte the reacter 2001:nt ;y; tem.
- 7. Containment Sumo Water Level (Flood levell Containment two channelssump water of level level (Flood)th indication, bo of which areis monitored by displayed in the control room on edgewise level indicators. Channel A and B sump flood leiel indication are recorded in the associated 'A' and 'B' EFIC Rooms. Each instrument encompasses a range of 0-10 feet above the sump and provides informetion to the operator related to gross leakage in the Reactor Building. This leakage may be indication of degradation in the reactor coolant pressure boundary (RCPB) which would require further investigation and action. These mtruments are not assumed to provide information required by the operator to take a mitigation action specified in the accident analysis.
However monitors are deemed ris significant Assuch,theyarenotTkpeAvariables.are and inclu this consideration. (continued) Crystal River Unit 3 B 3.3-130 g M P,:vi:icn No. 44 r r --- - - + - - - - -
Y ECCS-Operating B 3.5.2 BASES APPLICABLE LOCA, the RCS depressurizes as primary coolant is ejected SAFETY ANALYSIS through the break into the containment. The nuclear (continued) reaction is terminated either by moderator voiding during large breaks or CONTROL R0D assembly insertion for small breaks. Following depressurization, emergency cooling water is injected into the reactor, vessel core flood nozzles, then flows into the' downcomer, fills the lower plenum, and refloods the core. The LC0 ensures that an ECCS train will deliver sufficient water.to match decay heat boiloff rates soon enough to minimize core uncovery for a large break LOCA. It also ' ensures that the HPI. pump will deliver sufficient water for , a small break LOCA and provide sufficient boron to maintain i the core subcritical following the small break LOCA or an ' SLB. In the LOCA analyses,.HPI and LPI are not credited until l 35 seconds after actuation of the ESAS signal. This is l based on a loss of offsite power and the associated time i delays in startup and loading of the emergency diesel generator (EDG). Further, LPI flow is not credited until RCS pressure drops below the pump's shutoff head. For a large break LOCA, HPI is not credited at all. The ECCS trains satisfy Criterion 3 of the NRC Policy Statement. LC0 In MODES 1, 2, and 3, two independent (and redundant) ECCS trains are required to ensure that at least one is available, assuming a single active failure in the other train. For en::ple, the d :ign of the HPI injection v:lv:: ("UV 23, "UV 24, "UV 25, :nd "UV 25) :ll;w; p:wcr t; th
- ter Operator: to bc ::100ted between : n =:1 cr b:ch p p uer supply. Powering en ::t of MPI v:1v:: frc: th
- ltern:t ::wcr :upply clirinate: the independen:: :d lign; b:t; tr;in: Of HP! valve: to re:Civ -
- ES b;;. In thi: condition, Wil6ToiEEds:w:r
?6Iin fr:m the MppifMljy the system is still cidib1'i WERff5iBig an event, providing a concurrent single failure does not occur.
Hence, the 72 hour ACTION addressing a loss of redundancy is appropriate. L l' l I (continued) r Crystal River Unit 3 B 3.5-12 @MP,:vi:ica No. 6 l [
I ECCS-Operating B 3.5.2 l { BASES l l LCO Ccaversely, nNot all portions of the HPI fl6Eli{tj System (continued) satisfy the ihdependence criteria discussid'~ibove. Specifically, the HPI ff6MTith System downstream of the < l HPI/ Makeup pumps is nof~ispaf~able into two distinct trains, ' j and i: therefore, not independent. This conclusion is based l upon analysis which shows, that in the event of a postulated j i break in the HPI injection piping, injection flow is l required through a minimum of three (3) injection legs, . assuming one pump operation, or through a minimum of two (2) injection legs, assuming two HPI pump operation. When ) considering the impact of inoperabilities in this portion of ! the system, the same concept of maintaining single active l
-failure protection must be applied. When components become !
inoperable, an assessment of the HPI systems ability to i perform its safety function must be performed. If the system can continue to perform its safety function, without assuming a single active failure, then the 72 hour loss of redundancy ACTION is appropriate. If the inoperability. renders the system, as is, incapable of performing its safety function, without postulating a single active failure, then the plant is in a condition outside the safety l analysis and must enter LC0 3.0.3 immediately. 1 In MODES I, 2, and 3, an ECCS train consists of an HPI l subsystem and an LPI subsystem. Each train includes the piping, instruments, and controls to ensure an OPERABLE flow path capable of taking suction from the BWST upon an ESAS signal and manually transferring suction to the reactor building emergency sump. During an event requiring ECCS actuation, a flow path is provided to ensure an abundant supply of water from the BWST l to the RCS via the HPI and LPI pumps and their respective discharge flow paths to each of the four cold leg injection nozzles and the reactor vessel. In the long term, this flow path may be manually transferred to take its supply from the reactor building emergency sump and to supply its flow to the RCS via two paths, as described in the Background section. The flow path for each train must maintain its designed degree of independence to ensure that no single active l failure can disable both ECCS trains. 4 f (continued) ! Crystal River Unit 3 B 3.5-13 Amendment No. M3
ECCS-Operating l B 3.5.2 BASES HM RM(conIinue[) SR 3.5.2.5 h u-Suryg 1]ggcp ensureg, thy 3;gsc,,yglyggj[gjg,}he tLm! yYg.
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- hen $hFjiA%!Mih G'49" fiE5?:8Md4Rn'A' N "8M"
- W SR 3.5.2.6 l
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gsNCb"*e'Sure'de"'5!e a7 er*a 18Eh*" S0!"hS$ah$makah a4 ,"38 re$la0S5t ($ NEE mngop0Eak$ng !pr c oh t equipment. SR 3.5.2.7
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SE-En S$s" gg g$S!l0nk0ng""kobe con lEm!O by oper$Ekng esp $ enc REFERENCES 1. 10 CFR 50.46.
- 2. FSAR, Section 6.1.
khc!EOeh"fnk0rinR Componenk, ecem er , O(6[E8?76kEbbE' o b Wm%!D2R8hm!R(** ^"* '"9"' t St'rt"P 5-au i
- 6. FSAR, Section 4.3.10.1. I Crystal River Unit 3 B 3.5-18 AMiiidisniRevicien No. W NOTE - Valid ~Unti'l Cycle 12 Only
FLORIDA POWER CORPORATION CRYSTAL RIVER UNIT 3 DOCKET NO. 50-302/ LICENSE NO. DPR-72 l l ATTACHMENT F l l LICENSE AMENDMENT REQUEST #241, REVISION 0 HIGH PRESSURE INJECTION SYSTEM MODIFICATIONS i l Simplified Drawing of the HPI System l 1 l 1 l 1 l 1 I l v - --
K. _ i i J f 3 i o I o I~ l I l FROM .. V ( DHP-18
~
1500-3 1500-7 l V A/B 0 OHv-12 I D k uuv-2 2M" A
-'e "uuv-26 wu-23-rE3 FROM s.
uuv-se
' 6, 3'W 4-
^/8 BWST uUv-58 uuv-59 uuv-i v C
QI 2M' n -25 uuP-1C yy.23-FE1 4" HuV-62 uuv-3 uuv-63[ uuv-4[ I B-ES 6* D uuv-16 yyy_ig ) uuv-s ) (uu-27-n 1500-3 15e0-7 3' 4' i FROM 4.uuv-65 MUT W6, uuv-7 wu -5 e) (uu-24-rE uuv-s4 uuv-se e-Es uup-is 2M, k uuv-31 uuv-27
-C8G- !
E uuv-se j (THROTTLED) uuv-sa7 uuv-68 uuv-8 uuv-69 uuv-9 r uuv-sa9 Mu-23-FE4 8' 2M- v - m9, ^/8 3, _ uuv-te n uul-24 e 4' FROM e. p uuv-72 3, N% uu-23-rE2 BWST - 4 N wuv-73 uuv-7e uuv-11 v C P
,4 A/B O uuv-23 o
ogy_ii] g uuP-1A FROM DHP-1A s H 1 l 2 l 3 l 4 l 5
,, n , , , e i v i is u, w -w _j INSIDE OUTSIDE I R.B.
\ 7~ -
R.B. I A l SECONDARY 2 ) I l SHIELD WALL l l l (D-RING) 1500-7 1500-4 1 l SI SI i _ LOCKED pa l N2 k N1 1500-4 1500-5 PRE-THROTTLED POSITION WAntTENANCE VALVES . (TYP, 4 PLCS) l [ (TYP. 4 PLCS.) "UV-37 y 24- 82 X X to.
+ >Nuuv-is4 PEN 33k )
N (RCP-1D) e uuv-57s t l uuv-593 l l LOOP B M PEN 336 j r y Q 5 M463 2M" l l Edh B1
~
WUV-592 l l l I
'$$f" I C l
' i
-366[S (TYr), cE"$iYsS0$0STION 2 PLCS.) l L I t I l TO RCP
'. pen 35dsH> CN -~"i l ) C 5 SEAL I
I uuv-is2 _1 7 l INJECTION D y r I l l l l l l I APERTutltB e I CARDIl r l Ako AP' # AM7 - f ~[uuv-595 l t L O. MVV-588 l l 2n- l i
*' I r uuV-59i PEN 435 ! WUV-43 I ugy_S74 X >< X C>q 2r [ q h/-594 to. T uuv-16i l
(RC 1A) - uuv-59e I pen 434 l LOOP A t 3, X $ uuV-573 N MUV-160 2M- I l N uuV-42 A2 (RCP-18) I 9
- C 15e0-7 1500-4 I l
I NUCLEAR ENGINEERING t i - LEGEND CRYSTAL RIVER I UNIT 3 I GRIDA HPI UPGRADE PROJECT RED - PROPOSED MODIFICATIONS BLACK - EXISTING CONFIGURATION M EJ gDJbTNOWh a 11-13-98 I
/ WAD mw we NTS f [ f on arm scar g 1
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