ML20207E826
| ML20207E826 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 05/21/1999 |
| From: | Peterson S NRC (Affiliation Not Assigned) |
| To: | |
| Shared Package | |
| ML20207E830 | List: |
| References | |
| NUDOCS 9906070170 | |
| Download: ML20207E826 (23) | |
Text
- pwouq'o, UNITED STATES 8
NUCLEAR REGULATORY COMMISSION n
5 j
WASHINGTON, D. C. 20555
\\...../
FLORIDA POWER CORPORATION CITY OF ALACHUA CITY OF BUSHNELL
]
CITY OF GAINESVILLE CITY OF KISSIMMEE CITY OF LEESBURG
. CITY OF NEW SMYRNA BEACH AND UTILITIES COMMISSION.
CITY OF NEW SMYRNA BEACH CITY OF OCALA ORLANDO UTILITIES COMMISSION AND CITY OF OF.LANDO SEMINOLE ELECTRIC COOPERATIVE. INC, CITY OF TALLAHASSEE DOCKET NO. 50-302 CRYSTAL RIVER UNIT 3 NUCLEAR GENERATING PLANT j
AMENDMENT TO FACILITY OPERATING LICENSE Amendment No.178 License No. DPR-72 1.
The Nuclear Regulatory Commission (the Commission) has found that:
A.
The application for amendment by Florida Power Corporation, et al. (the licensees), dated November 23,1998, as supplemented on January 29 and May 7,1999, complies with the standards and requirements of the Atomic Finergy Act of 1954, as amended (the Act), and the Commission's rules and regulations set forth in 10 CFR Chapter I; B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Commission; C.
There is reasonable assurance (i) that the activities authorized by this amendment can be conducted without endangering the health and safety of the public, and (ii) that such activities will be conducted in compliance with the Commission's regulations; D. -
The issuance of this amendment will not be inimical to the common defense and security or to the health and safety of the public; and 9906070170 990521 PDR ADOCK 05000302 P
PDR I
E - E.
The issuance of this amendment is in accordance with 10 CFR Part 51 of the Commission's regulations and all applicable requirements have been satisfied.
' 2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C(2) of Facility Operating License No. DPR-72 is hereby amended to read as follows:
Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.178, are hereby incorporated in the license. Florida Power Corporation shall operate the facility in accordance with the Technical Specifications.
3.
This license amendment is effective as of its date of issuance and shall be implemented prior to commencing Cycle 12 operation.
FOR THE NUCLEAR REGULATORY COMMISSION
-s
(-)\\% Ad \\' N,
Sheri R. Peterson, Chief, Section 2 Project Directorate 11 Division of Project Licensing Management Office of Nuclear Reactor Regulation Date of issuance: M 21, 1999 T'
s
ATTACHMENT TO LICENSE AMENDMENT NO.
TO FACILITY OPERATING LICENSE NO. DPR-72 DOCKET NO. 50-302 Replace the following pages of the Appendix A Technical Specifications with the attached pages. The revised pages are identified by amendment number and contains marginallines indicating the areas of change.
- Indicates spillover page.
Remove Paae Insert Pace 3.3-5 3.3-5 j
3.3-13 3.3-13 3.3-15 3.3-15 3.5-5 3.5-5 B 3.3-8 8 3.3-8 B 3.3-9 B 3.3-9 B 3.3-20 B 3.3-20 B 3.3-21 B 3.3-21 B 3.3-46 B 3.3-46 B 3.3-47 B 3.3-47 8 3.3-51 B 3.3-51 B 3.3-52 B 3.3-52 is 3.3-54 B 3.3-54 B 3.3-125A B 3.3-125A B 3.3-129 B 3.3-129 B 3.3-130 B 3.3-130 B 3.5-12 B 3.5-12 B 3.5-13 B 3.5-13 B 3.5-18 8 3.5-18 B 3.5-19 B 3.5-19*
F
+
RPS Instrumentation 3.3.1 i
Table 3.3.1-1 (page 1 of 1)
Reactor Protection System Instrumentation APPLICABLE CONDIT!DNS MODES OR REFERENCED OTHER FROM j
SPECIFIED REQUIRED SURVEILLANCE ALLOWABLE
' FUNCTION-CONDITIONS ACTION D.1 REQUIREMENTS VALUE 1.
Nuclear Overpower -
s.
High Setpoint 1,2(*)
F 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(D) 3(b) g gg 3,3,3,3
, $x gyp 4(b) $(b) 2.
RCS High Outlet Temperature 1,2 F
SR 3.3.1.1 s 618'F SR 3.3.1.4 SR 3.3.1.6 3.
RCS High Pressure 1,2 F
SR 3.3.1.1 s 2355 psis SR 3.3.1.4 SR 3.3.1.6 SR 3.3.1.7 4 RCS Low Pressure 1,2(*)
F SR 3.3.1.1 a 1900 psig l
SR 3.3.1.4 SR 3.3.1.6 i
SR 3.3.1.7 5.
RCS variable Low Pressure 1,2(*)
F SR 3.3.1.1 m (11.59
- ftwt
- SR 3.3.1.4 5037.8) psig i
6.
Reactor Building High 1,2,3(C)
F SR 3.3.1.1 s 4 psig l
Pressure SR 3.3.1.4 l
7.
Reactor Coolant Pump Power l'2(*)~
F SR 3.3.1.1 More than one punp Monitor (RCPPM)
SR 3.3.1.4 drawing 5 1152 or SR 3.3.1.6 1
1 4,400 kW SR 3.3.1.7 8..
Nuclear Overpower RCS Flow 1,2(*)
F SR 3.3.1.1 Nuclear Overpower RCS j
and Measured AXIAL 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 SR 3.3.1.7 1
9.
Main Turbine Trip (Control a 45% RTP H
SR 3.3.1.1 a 45 psig Olt Pressure)
SR 3.3.1.4 SR 3.3.1.6
-10.
Loss of Both Main Feedwater a 20% RTP 1
SR 3.3.1.1 a 55 psig
' Puups (Controt Dit SR 3.3.1.4 Pressure)
.11.
Shutdown Bypass RCS High 2(b) 3(b)
G SR 3.3.1.1 s 1820 psig l
Pressure SR 3.3.1.4 i
4(b) $(b)
SR 3.3.1.6 (a) - When not in shutdown bypass operation.
'(b) During shutdom bypass operation with any CRD trip breakers in the closed position and the LRD Control System (CRDCS) capable of rod withdrawal.
(c) With any CRD trip breaker in the closed position and the CRDCS capable of rod withdrawal.
l Crystal River Unit 3 3.3-5 Amendment No.178 e
n
-ESAS Instrumentation 3.3.5 ACTIONS CONDITION REQUIRED ACTION COMPLETION TIME C. -(continued)
C.2
NOTE---------
Only required for RCS Pressure--Low Parameter.
Reduce RCS pressure 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />
< 1800 psig.
l bNQ C.3
NOTE---------
Only required for RCS Pressure--Low Low Parameter.
Reduce RCS pressure 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />
< 900 psig.
6N.Q C.4
NOTE---------
Only required for Reactor Building Pressure High 1
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 /> 1
i 1
Crystal River Unit 3 3.3-13 Amendment No.178 i
L.
ESAS Instrumentation 3.3.5 Table 3.3.5-1 (page 1 of 1)
Engineered Safeguards Actuation System Instrumentation APPLICABLE MODES OR OTHER SPECIFIED ALLOWABLE PARAMETER CONDITIONS VALUE 1.
Reactor Coolant System Pressure - Low a 1800 psig r 1625 psig l
2.
Reactor Coolant System Pressure - Low Low 2 900 psig 2 500 psig 3.
Reactor Building Pressure-High 1,2,3 s 4 psig 4.
Reactor Building Pressure - High High 1,2,3 s 30 psig Crystal River Unit 3 3.3-15 Amendment No. 178 i
ECCS -Operating 3.5.2 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 with the 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 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 following valves in the HPI flow 24 months path are locked, sealed or otherwise secured in the correct position:
a.
MUV-2; b.
MUV-6; c.
MUV-10; d.
MOV-590; e.
MUV-591; f.
MUV-592; and g.
MUV-593.
(continued)
Crystal River Unit 3 3.5-5 Amendment No.178
F
]
RPS Instrumentation B 3.3.1 BASES BACKGROUND Channel Bypass (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 j
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.
I Shutdown Bvoass During plant cooldown, it is desirable 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 Iow Pressure trip will occur at 1900 psig and the rods will re-insert into the core. To I
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. 178
K
~
RPS Instrumentation B 3.3.1 BASES Shut'own Bvoass (continued)'
BACKGROUND d
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.
Inl addition to the Shutdown. Bypass RCS High Pressure trip,
,the nuclear overpower high flux trip setpoint is administrative 1y 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 Relay 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 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.
. APPLICABLE-Each of the analyzed accident's 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 (continued)
Crystal River _ Unit.3 8 3.3-9 Amendment No.178 1
i
RPS Instrumentation B 3.3.1 f
l BASES APPLICABLE 11.
Shutdown Bvoass RCS Hiah Pressure (continued) l SAFETY ANALYSES, LCO, and During. shutdown bypass operation with the Shutdown APPLICABILITY Bypass RCS High Pressure trip active with a setpoint 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 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 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.
i In MODES 1 and 2, the following trips shall be OPERABLE. These trips are designed to rapidly make 1
the reactor subtritical in order to protect the SLs during A00s and to function along with the ESAS to provide acceptable consequences during accidents.
1.a Nuclear Overpower-High Setpoint; 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. 178
F' RPS Instrumentation B 3.3.1 BASES-
- APPLICABLE
- 11. Shutdown Bvoass RCS Hiah Pressure (continued)
SAFETY ANALYSES, 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 1820 psig, provided the Shutdown Bypass RCS High Pressure l
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 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 (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 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 i
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 Naclear Overpower-Low Setpoint trip Functions are required to be OPERABLE if the CRD trip breakers are
= losed and the CRDCS-is ' capable of rod withdrawal. Under c
these conditions, the Shutdown Bypass RCS High Pressure and i
Nuclear Overpower-Low setpoint trips are sufficient to prevent-an approach to conditions that could challenge SLs.
(continued)
Crystal River Unit 3 B 3.3-21 Amendment No. 178
ESAS Instrumentation-B 3.3.5
- . BASES BACKGROUND related safeguards equipment will not inhibit-the overall ES (continued) '
valve is driven by either of two matrices, one is from Functions. Where a motor operated or a solenoid operated 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 Byoasset No provisions are made for maintenance bypass of ESAS instrumentation channels.
Operational bypasses are provided, as discussed below, to allow accident recovery actions to continue and, to allow plant cooldown without 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 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.
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 four bistables to provide two trip signals, at 1625 psig and 500 psig, and two bypass permissive signals, at 1800 psig and 900 psig.
(continued)
' rystal River Unit 3 B 3.3-46 Amendment No. 178 C
i
i ESAS Instrumentation B 3.3.5 BASES.
BACKGROUND Reactor Coolant System Pressure (continued)
The outputs of the three channels trip bistables, associated j
with the low RCS pressure (1625 psig) actuate bistable trip l 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 o
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
.I the High RB Pressure Parameter, and six pressure switches are u:ed for the High-High RB Pressure Parameter.
The output contacts of six High RB Pressure switches are i
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 are used in two sets of identical and independent actuation trains. The outputs of the liigh High RB Pressure switches are used in two-out-of-three coincident networks for RB l
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.
(continued)
Crystal River Unit 3 B 3.3-47 Amendment No. 178
7 ESAS Instrumentation B 3.3.5 BASES LC0 1.
Reactor Buildina Pressure-Hiah Setooint (continued)
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, thus ensuring that the RB high pressure actuation of j
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
'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 a timely actuation during a large break LOCA.
APPLICABILITY The ESAS instrumentation for each Parameter is required to be OPERABLE during the' following MODES and specified conditions.
1.
Reactor Coolant System Pressure--Low Setooint i
The RCS. Pressure-Low Setpoint actuation Parameter
)
shall be OPERABLE during operation above 1800 psig.
l This ensures the capability to automatically actuate safety systems and components during conditions indicative of a LOCA or SLB.
Below 1800 psig, the low l RCS Pressure actuation Parameter can 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 i
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 j
the RCS energy content is low.
l (continued)
Crystal River Unit 3 8 3.3-51 Amendment No. 178
p l'
ESAS Instrumentation B 3.3.5 i
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 removal bistable must be set with a setpoint of s 1800 psig.
The bypass removal l
does not need to function 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 5 900 psig. The bypass removal does not need to function for accidents initiated by RCS Pressure below the bypass removal setpoint.
l (continued)
Crystal River Unit.3 B 3.3-52 Amendment No.178
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ESAS Instrumentation B 3.3.5 BASES ACTIONS 161 (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.
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.
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 and is sufficient time to perform the Required Action.
-j 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 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 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.
l (continued)
Crystal River Unit 3 B 3.3-54 Amendment No.178 i
F t
PAM Instrumentation B 3.3.17 BASES a
The following table identifies the specific instrument tag ntsnbers for PAM instrunentation identified in Table 3.3.171.
FUNCTION CHANNEL A CHANNEL B 1.
Wide Range Neutron Flux N1-15 N!-1 or N1 15 NIR WI-14 N!-1 2.
RCS Hot Leg Temperature RC 4A TI4-1 RC-4B-TIR1 3.
RCS Pressure (Wide Range)
RC-158-P!2 or RC-158 PIR RC-159-P12 4.
Reactor Coolant RC-163A-LR1 (Hot leg level) and RC-163B-LR1 (Hot leg level) and Inventory RC-164A LR1 (Vessel Head level)
RC-164B-LR1 (Vessel Head level) 5.
Borsted Water Storage DM-7-LI or DH-7-LIR1 DH-37-LI Tank Level 6.
High Pressure injection A1: MU 23 F18 1 A1: MU-23 F112 Flow A2 MU-23-F110 A2: MU 23 FI6-1 B1: MU-23-F19 B1: MU 23-F15-1 B2: MU-23 F17-1 B2: MU 23 F111 7.
Containment Sump Water WD-303-LI or WD-303-LR WD 304-LI or WD-304-LR Level (Flood Level) 8.
Containment Pressure BS-16-PI BS-17 PI (Expected Post-Accident 1
i Range) 9.
Containment 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-1B/1C; ES Light Matrix "B": AHV-1A/1D; 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; l
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; e
MSV 130/148; SWV-47 thru 50/79 thru 86/109/110 1
- 12. Containment Hydrogen WS-11 CR WS-10-CR Concent ration
- 13. Pressuriter Level RC 1-LIR 1 RC-1-t!R-3
- 14. Steam Generator Water DTSG A: SP 25-L11 or OTSG A: SP-26 LI1 Level (Startup Range)
SP-25-LIR OTSG 8: SP 30-L11 OTSG B: SP-29-LII or SP-29 LIR (continued)
Crystal River Unit 3 B 3.3-125A Amendment No. 178
h PAM Instrumentation B 3.3.17 i
l BASES l
LC0 5.
Borated Water Storaae Tank (BWST) Level (continued)
BWST inventory is monitored by level instrumentation 1
with a span of 0 to 50 feet. Redundant monitoring I
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 LC0 deals specifically with this portion of the instrument string.
1 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 1
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 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 action is necessary to satisfy the long-term core cooling. requirements specified in 10 CFR 50.46.
6.
HPI Flow (Low Ranae)
HPI flow instrumentation is provided for verification and long term monitoring of HPI flow.
HPI flow is i
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 prnvides an input to a control room indicator.
Since th5 operator relies on the control room indication following an accident, the LC0 deals with this portion of the instrument string.
(continued) i Crystal River Unit 3 B 3.3-129 Amendment No. 178
PAM Instrumentation B 3.3.17 BASES l
LC0 -
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 distributior, through the HPI lines for LOCA core
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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 assumed in design calculations associated with Emergency Diesel Generator loading, ECCS pump j
available NPSH, and makeup tank (MUT-1) allowable overpressure versus level).
1 7.
Containment Sumo Water Level (Flood level)
Containment sump water level (Flood) is monitored by two channels of level indication, both of which are I
displayed in the control room on edgewise level
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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 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 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 are included within the LCO based upon this consideration.
(continued)-
Crystal River Unit' 3 B.3.3-130 Amendment No. 178
g' ECCS-Operating B 3.5.2 l
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 E
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 the core subcritical following the small break LOCA or an
.SLB.
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 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. With one ECCS train inoperable, the system is still l
l 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.
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(continued)
Crystal River Unit 3 B 3.5-12 Amendment No. 178 L
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ECCS-Operating B 3.5.2 1
l.
l-BASES LC0 Not all portions of the HPI flow path satisfy the 1
(continued) independence criteria discussed above.
Specifically, the j
HPI flow path downstream of the HPI/ Makeup pumps is not separable into two distinct trains, and is therefore, not
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independent. This conclusion is based upon analysis which J
shows, that in the event of a postulated break in the HPI injection piping, injection flow is required through a minimum of three (3) injection legs, assuming one pump j
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operation, or through a minimum of two (2) injection legs, f
assuming two HPI pump operation. When considering the impact of inoperabilities in this portion of the system, the same concept of maintaining single active failure protection must be applied. When components become inoperable, an assessment of the HPI systems ability to 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 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> 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 analysis.and must enter LC0 3.0.3 immediately.
In MODES 1, 2, and 3, an ECCS train consists of an HPI 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 i
provided to ensure an abundant supply of water from the BWST i
to the RCS via the HPI and LPI pumps and their respective discharge flow paths to each of the four cold leg injection i
nozzles and the reactor vessel.
In the long term, this flow j
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 failtre can disable both ECCS trains.
(continued)
Crystal River Unit 3 B 3.5-13 Amendment No.178 l
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ECCS-Operating F.-
B 3.5.2 p
BASES 7
(:
~ SURVEILLANCE
.SR 3.5.2.5' j
REQUIREMENTS L
(continued)-
Verification of the positions of the listed valves -in the HPI flowpath' ensures adequate flow-resistance in the overall system and-the individual HPI' lines.
Maintenance of adequate flow resistance Jand pressure' drop in the piping system,for each injection point is necessary in order to:
. (1) provide the proper flow split between injection points in accordance_with the assumptions 'used in the ECCS LOCA
~ analyses;' (2) provide an. acceptable level of total. ECCS flow to all injection points equal to or above values assumed in the ECCS LOCA analyses; (3) ensure adequate cooling flow to the' HPI pump mechanical seals; and (4) prevent HPI pump flow from exceeding 600 gpm when the system is in its minimum resistance configuration (600 gpm is the maximum HPI pump flow rate assumed in design calculations associated with Emergency Diesel Generator loading, ECCS pump available NPSH, and makeup tank (MUT-1) allowable overpressure versus level).
This 24 month Frequency is acceptable based on consideration of the design 1 reliability of valves that are-locked, sealed, or
- otherwise secured in position.
Verification of; correct valve position will be accomplished by assuring the mechanism that ' locks, seals or' secures the valves ~is intact. 'If the stop check valves or throttle valves'are repositioned, the valves must be returned to their. correct position and then secured. This "as-left" positioniverification ensures the HPI flow assumptions in the accident : analysis are maintained.
5R 3.5.2.6 1
- This Surveillance ensures that the flow controllers for the LPI throttle valves will automatically control the LPI train flow rate in the desired: range and prevent LPI pump runout as RCS pressure decreases after a LOCA. 'The 24 month Frequency is acceptable' based' on consideration of the design l
reliability (and confirming ~ operating experience) of the equipment.
1 (continued)
- Crystal River Unit.3 B 3.5-18 Amendment No. 178 a
b b
ECCS -Ope rati ng l-B 3.5.2 BASES a
SURVEILLANCE-SR 3.5.2.7 3
REQUIREMENTS j
(continued);
Periodic inspections of the reactor building emergency sump suction inlet ensure that it is unrestricted and stays in proper operating condition.
The 24 month Frequency is based on the need to perform this Surveillance under the conditions that apply during a plant outage and to preserve access to the location.
This Frequency has been found to be
-sufficient to detect abnormal degradation and has been confirmed by operating experience.
REFERENCES 1.
FSAR, Section 6.1.
3.
NRC Memorandum to V. Stello, Jr., from R.L. Baer,
" Recommended Interim Revisions to LCOs for ECCS Components," December 1, 1975.
4.
American Society of Mechanical Engineers, Boiler _ and Pressure Vessel Code,Section XI, Inservice Inspection, Article IWP-3000.
5.
.FTI 51-1266138-01, Safety Analysis Input to Startup
-Team Safety Assessment.
6.
FSAR,' Section 4.3.10.1.
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i Crystal' River. Unit 3' B.3.5-19
' Amendment No.178 l