ML20070D804
| ML20070D804 | |
| Person / Time | |
|---|---|
| Site: | Palisades  |
| Issue date: | 02/15/1991 |
| From: | Brian Holian Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20070D807 | List: |
| References | |
| NUDOCS 9103010206 | |
| Download: ML20070D804 (16) | |
Text
-
[kf kj UNITED STATES g
NUCLEAR REGULATORY COMMISSION s
wAssiwatoN, D. C. 20666 k.....,/
l t
CONSUMERS POWER COMPANY PALISADES PLANT DOCKET NO. 50-255 AMENDMENT TO PROVISIONAL OPERATING LICENSE Amendment No. 136 License No. DPR-20 1.
The Nuclear Regulatory Commission (the Comission) has inund that:
A.
The applications for amendment by Consumers Power Company (the licensee) dated June 13,1990(asrevisedNovember9and December 7, 1990) and November 2, 1990 complies with the standards and requirements of the Atomic Energy Act of 1954, as amended -
t (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 (1) 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 E.
The issuance of this amendment is in accordance with 10 FR Part 51 of the Comission's regulations _and all applicable requt ments 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 3.B. of Provisional Operating License No. DPR-20.is hereby amended to read as follows:
9103010206 910210 PDR ADOCK 05000255 P
.._.. _._.~.
2 Technical Specifications The Technical Specifications contained in Appendices A and B, as revised through Amendment No.136, are hereby incorporated in the license.
The licensee shall operat the facility in accordance with the Technical Specifications.
3.
This license amendment is effective as of its date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION hre L. B. Marsh, birector Project Directorate 111-1 Division of Reactor Projects Ill/IV/V Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications 3
Date of issuance: February 15, 1991 9
l l..
ATTACHMENT TO LICENSE AMENDMENT NO. 136 PROV1510NAL OPERATING LICENSE NO. OPR-20 DOCKET NO. 50-255 Revise Appendix A Technical Specifications by removing the pages identified below and inserting the attached pages. The revised pages are identified by the amendment number am! contain marginal lines indicating the area of change.
RfMOVE INS'RJ 3-29 3-29 3-31 3-31 3-33 3-33 3-77 3-77 3-76 3-78 3-81 3-81 3-135 3-135 3-136 3-136 4-3 4-3 4-R 4-8 4-10 4-10 4-88 4-88 5-3 5-3 2
R
/
f ID
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3.3 EMERGENCY CORE COOLING SYSTEM Aeolicability Applies to the operating status of the emergency core cooling system.
Ob.iective To assure operability of equipment required to remove decay heat from the core in either emergency or normal shutdown situations.
Soecificationi Safety In_iection and Shutdown Coolina System,1 3.3.1 The reactor shall not be made critical, except for low temperature physics tests, unless all of the following conditions are met:
a.
The SIRW tank contains not less than 250,000 gallons of water with a boron concentratica of at least 1720 ppm but not more than 2000 ppm at a temperature not less than 40'F.
b.
All four Safety injection tanks are operable and pressurized to at least 200 osig with a tank liquid level of at least 174 inches and a maximum Yevel of 200 inches with a boron concentration of at least 1720 ppm bat not more than 2000 ppm.
c.
One low prt ssure Safety injection pump is ope-ble on each bus.
d.
One high prassure Safety injection pump is operable on each bus.
J e.
Both shutdtwn heat exchangers and both component cooling heat exchangers are operable, f.
Piping and valves shall be operable to provide two flow paths from the SIRW tank to the primary cooling system, g.
All valves, piping and interlocks associated with the above components and required to function during accident conditions are operable.
h.
The 1.ow Pressure Safety injection Flow Control Vaive CV 3006 shall be opened and disabled (by isolating the air supply) to prevent spurious closure, i.
The Safety injection bottle motor-operated isolation valves : hall be opened with the electric power supply to the valve motor disconnected.
j.
The Safety Injection min 1 flow valves CV 3027 and 3056 shall be opened with HS-3027 and 3056 positions to maintain them open.
3 29 Amendment No.31,7b l01,136
1L.
3.3 EMERGENCY CORE COOLING SYSTEM (Continued)
If Specification a. and b. cannot be met, an orderly shutdown shall c.
te initiated and the reactor shall be in hot shutdown condition within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />, and cold shutdown within the next 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
BJULi.A The normal procedure for starting the reactor is, first, to heat the primary coolant to nesr operating temperature by running the primary coolant pumps. The reactor is then made critical rods and diluting boron in the primary coolant.'g withdrawing control With this mode of start up, the energy stored in the primary coolant during the approach to criticality is substantially equal to that during power operation and, therefore, all engineered safety features and auxiliary cooling systems are required to be fully operable. During low temperature physics tests, there is a negligible amount of stored energy in the primary coolant; therefore, an accident comparable in severity to the design basis accident is not possible and the engineered safeguards' systems are not required.
The SIRW tank contains a minimum of 250,000 gallons of water containing 1720 ppm boron.
This is sufficient boron concentration to provide a 5%
shutdown margin with all control rods withdrawn and a new core at a temperature of 60'F.
Heating steam is provided to maintain the tank above 40*F to prevent freezing. The 1% boron (1720 ppm) solution will not precipitate out above 3PF.
The source of steam durin steam line in the turbine cycle. g normal plant opec' ion is extraction The limits for the safety injection tank pressure and volume assure the required amount of water injection during an accident and are based on values used for the accident analyges.
The minimum 174 inch level corresponds to-a volume of 1040 ft and the maximum 200-inch level 3
corresponds to a volume of 1176 ft.
Prior to the time the reactor is brought critical, the valving of the safety injection system must be checked for correct alignment and appropriate valves locked. Since the system is used for shutdown cooling.
the valving.ill be changed and must be properly t.ligned prior to start-up of the reactor.
The operable status of the various.systcms and components is to be demonstrated by periodic tests. A large fraction of these tests will be pr* formed while the reactor is operating in the power range.
If a moonent is found to be inoperable, it will.be possible in most cases to
- d ect repairs and restore the system to full operability within a relatively short time.
For a single component to be inoperable does not negate the ability of the system to perform its. function, but it reduces the redundancy provided in the reactor design and thereby limits the 3-31 Change No. 1 Amendment No JJ7,136 i
I
3.3 EMERGENCY CORE COOLING SYSTEM hili (continued) demonstrate that the maximum fuel clad tismperatures that could occur over the break size spectrum are well below the melting temperature of zirconiur (3300'F).
Malfunction of the Low Pr6ssure Safety injection Flow control valve could defeat the Low Pressure Injection feature of the ECCS; therefore, it is disabled in the 'open' mode (by 19 t.ing the air supply) during plant operation. This action assures that it will not block flow during Safety Injection.
The inadvertent closing of any one of the Safety injection bottle isolation valves in conjunction with a LOCA has not been analyzed.
To provide assurance that this will not occur, these valves are electrically locked open by a key switch in the >.ontrol room.
In addition, prior to
}
critical the valves are checked open, and then the 480 volt breakers are 5
opened. Thus, a failure of a breaker and a switch are required for any of the valves to close, insuring both HPSI pumps are inoperable when the PCS temperature is
<260'F or the shutdown cooling isolation valves are open eliminates PCS mass additions due to inadvertent HPSI pump starts.
Both HPSI pumps starting in conjunction with a charging / letdown imb lance may cause 4
10CFR50 Appendix G limits to be exceeded when the PCS temperature is
<260'F.
When the PCS temperature is 2 430'F, the-pressurizer safety
-valves ensure that the PCS pressure will not exceed 10CFR50 Appendix G.
The requiopert to have both HPSI trains operable above 325'F provides added assurat.ce that the effects of a LOCA occuring under LTOP conditions would be mitigated.
If a LOCA occurs when the primary system temperature is less than or equal to 325'F, the pressure would drop to the level where low pressure safety injection can prevent core damage.
Therefore, when the PCS temperature is 2260'F and $325'F operation of the HPSI system would not cause the 10CFR50 Appendix G limits to be exceeded nor is HPSI system operation necessary for core cooling.
HPSI pump testing with the HPSI pump manual discharge valve closed is permitted since the closed valve eliminates the possibility of-pump testing being the cause of a mass addition to the PCS.
Referen'sgi (1)
FSAR, Section 9.10.3; (2)
FSAR, Section 6.1, (3)
FSAR, Section 14.17 (4)
Letter, H.G.Shaw (ANF) to R.J.Gerling (CPCo), " Standard Review Plan Chapter 15 Disposition of Events Review for Changes to Technical Specifications Limits for Palisades Safety injection Tank Liquid Levels", April 11, 1990, 3-33 Amendment No. JJ, JJ, JSJ, JJ7, JJJ,136 e
n 1
3.17 LNSTRUMENIATION AND CONTROL SY&TiMS (Cont'd) if the bypass is not effected, the out of service channel (Power Removed) assumes a tripped condition (except high rate of change power, variable high power and high pressurizer pressure), W which results in a one out-of three channel logic, if, in the 2 of 4 logic system of either the reactor protective system or the engineered safeguards system, one channel is bypassed and a second channel manually placed in a tripped condition, the resul%g l gic is 1 of 2.
At rated power, the minimum operable variable high power level channels is 3 in order to provide adequate flux tilt detection.
If only 2 channels are operable, the reactor power level is reduced to-70% rated power which protects the reactor from possibly exceeding design peaking factors due to undetected flux tilts.
The engineered safeguards system provides a 2 out of 4 logic on the signal used to actuate the equipment connected to each of the 2 emergency diesel generator units.
Two source-range channels are available any time reactivity changes l
are deliberately being introduced into the reactor and the neutron power is.not visible on the wide range nuclear instrumentation or above 10"% of rated power.
This ensures that redundant source range instrumentation is available to operators to monitor effects of reactivity changes when neutron power levels are only visiole on the source-range channels.
In the event only one source-range channel is l available and the neutron power level is sufficiently high that it is being monitored by both channels of wide-range instrumentation, a j
startup can be performed in accordance with footnote (d) of Table 3.17.4.
The Recirculation Actuation System (RAS) initiates on a 1 out of 2 taken twice logic scheme.
Any one channel declared inoperable shall be placed in a bypass condition to ensure protection from an inadvertent RAS Actuation.
Since the bypassing of a channel introduces the possibility for a failure to receive an automatic RAS actuation signal, the time period in the bypassed condition is limited.
The Zero Power Mode Bypass can be used to,) bypass the low flow, steam generator low pressure, and TM/LP trips (
for all four Reactor Protective system channels to perform control rod testina or to perform low power physics testing below normal operating temperatures. The requirement to maintain cold shutdown boron concentration when in the bypass condition provides additional assurance that an accidental criticality will not occur.
To allow low power physics testing at_ reduced temperature and pressure, the requirement for cold shutdown boron concentration is not required and the allowed power is increased to 10"%.
References (1) Updated FSAR, Section 7.2.7 (2) Updated FSAR, Section 7.2.5.2 3-77 4
Amendment No. JJE, 12/,135
. _ _ _ _ _. _.. _.... -_ ~..
Table 3.17.1 Instrumentation Ooeratino Reauirement for Reactor Protective System Minimum Minimum Permissible Operable Degree of Bypass A
Functional Channels Redundancy Conditions 1.
Manual (Trip 1
None None Buttons) (g) 2, Variable High 2(*'d) 1(d)
None Power Ltvel (g) 3.
Wide Range 2
1 Below 10%(*) or Above l
Channels (g) 15% Rated Power (a)
Except as Noted in (c).
4.
Thermal Margin /
2 " ' )
1 Below 10%(') of Rated low Pressurizer Power ( and greater than Pressure (g) cold shutdown boron concentration.
5.
High-Pressurizer 2(b) 1 None Pressure (g) 6.
Low Flow Loop (g) 2 1
Bel ow l*0%(') of Rated N
)
Power and greater than cold shutdown boron concentration.
7.
Loss of Load (h) 1 None None 8.
Low Steam 2/ Steam 1/ Steam None Generator Water Gen
- Generator Level (g) 9.
Low Steam 2/ Steam 1/ Steam Below l*0%(*) of Rated Generator Pressure Gen
- Generator Power and greater
)
(g) than cold shutdown boron concentration.
10.
High Containment 2"'
1 None Pressure (g)
(a)
Bypass automatically removed.
(b)
One of the inoperable channels must be in the tripped condition.
(c)
Two channels required if TM/LP, low steam generator or low-flow channels are bypassed.
(d)
If only two channels are operable, load shall be reduced to 70% or -
-less of rated power.
(e)
For low power physics testing,10% may be increased to 10% and cold shutdown baron concentration is not required.
(f)
_ Axial Offset' operability requirements-are given_in Specification 3.11.2.
(g)
Required operable if any clutch power supply is energized.
_(h)
_ Automatically bypassed below 15% power.
3-78 Amendment No. JJE, J29,136
Table 3.17.4 Instrumentation Operatina Reouirements for Other Safety Feature Functions 1
Minimum Minimum Permissible Operable Degree of Bypass 16 Functional Unit Channels Eqdyn(gngy Conditions 1
_SIRWT Low-Level 4
NA(b)
One channel may be Switches inoperable for a period of 7 days.
- 2 AT-Power 3*
1 None Comparator 3
(Deleted) 4 Air Cooler Service 1
None None Water Flow Instruments 5
Primary and 1
None N/A Secondary Rod Insertion and Out-of-Sequence Monitors 6
Fuel Pool Building 1
None As Requested Under Crane Interlocks Administrative Controls.*
7 Source-Range 2
1*
Not required Above 10"% l Channels of Rated Power.
(a)
Crane shall not-be used to move material past the fuel storage pool unless the interlocks are available.
(b)
If a channel is declared inoperable, it shall be placed in a bypass condition. Minimum degree of redundancy is not applicable to the SIRWT low-level switches.
(c)
If only two channels are operable, load shall be reduced to 70% or less of rated power.
(d)
Minimum operable channels shall be one (1)- and minimum degree of redundancy.is zero -(0) if shutdown neutron power levels indicated on the wide range channels are greater than three times the lowast decade l in which__ neutron visibility can be confirmed. Neutron visibility will be confirmed through observation of reactivity changes on neutron power level-(including a 1/M plot during reactor start-up) and -
comparing the observed changes to the changes noted on previous-similar start-aps.
Instrumentation operability will also-be verified by comparison among the three operable channels to ensure their individual responses are in agreement.
3-81 Hjfst/Mfj/3 Amendment No. JJJ, JJf q
136
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' Table 3.25.1 ALTERNATE SHUTDOWN MINIMUM EOUIPMENT 4
!!2 :
Instrumentation Minimum Eauioment Readout location 1
Pressurizer Pressure 1
C150
~(PI 0110) 2 Pressurizer Level 1
C150 (LI-0102E) 3 Reactor Coolant Hot leg 1/ Loop C150A Temperature (TI-Oll2HAA).
(TI-0122HAA)-
4 Reactor Coolant Cold 1/ Loop C150A Leg Temperature (T1-Oll2CAA)
(TI 0122CAA) 5 Steam Generator 1/S.G.
C150A Pressure (PI-0751E)
(PI-0752E)
.6 Steam Generator Level 1/S.G.
-C150
-(LI 0757C)
(LI 0758C) 7 Source Range 1
C150A Neutron Monitor (NI-1/3C)-
C150 Suction Pressure (PS 07410)
~9
-SIRW Tank Level 1
C150A (LT-03328) 10 Auxiliary Feedwater 1/S.G.-
C150 Flow Rate l
(FI-0727B)-
(FI 07498)-
c l
3-135 Amendment No.-J22,136
Table 3.25.1 (Continued)
ALTERNATE SHUTDOWN MINIMUM E0VIPMENT Transfer Minimum Switch da Switches Eauioment location Function 11 HS 0102A 1
C150 Contr61 Room Alarm.
12 HS 01028 1
C150 S/G Level Indications.
Pressurizer level Indications.
Aux. FW Flow Indication.
Aux. Fw Flow Control.
x 13-HS 0522C C150 Opens Aux. Fw Pump Steam Valve CV-05228.
14 HS-0102C 1-C150A Control Room Alarm.
S/G Pressure Indications..
Hot / Cold Leg Temperature Indications.
Neutron Monitor System Power.
Minimum Controls U2 Control Circuits Eauioment
- f. ram Function t
.15 Auxiliary FW 1
C150
. Controls B S/G Aux FW Flow' Control
- Flow Control Valve (HIC 07270)
(CV 0727).
16 Auxiliary FW 1
C150 Controls A S/G Aux. FW Flow Control Flow Control-Valve (HIC-0749C)
(CV-0749).-
3-136 Amendment No. -J22,136 l
.y l
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TABLE 4.1.1 Minious Frequencies for Checirs, Cetibratiorn and Testing of Reactor Protective Systee(S)
Surveitience Channet Description Function Frcouency Surweiitence Methad Comperison of four-power chamet readings.
1.
Power Rerge Safety Channets
. a. Check (7) e.
- b. Check (3)
D b.
Channet adjustment to agrea with heet botence cetcutation. Repeet whenever fIun-AT power comparators eterms.
I c.
Test M(2) c.
Internet test signet.
d.
Calibrate (6)
R d.
Channet stigrument through meesurement/edjusternt of internet test points.
i Cooperison of chemet Indications.
l l
2.
vide-Range e.
Check 5
e.
Neutron Monitors b.
Test P
ts.
Internet test signet, Chemet eligrunent through meesurement/edjustment of c.
Collbrate R
c.
internet test points.
Comperison of four seperate totet flow truficatione.
3.
Reactor Cootent Flow e.
Check e.
b.
Cetibrate R
b.
Known dif ferentist pressure opptled to sensors.
c.
Test M(2) c.
Sistable trip tester.(T)(4) 4.
Thernet Margin / Low e.
Check: (8)
S e.
Check:
Pressuriger Presstre (1) Tenperature (1) Comperison of four separate calculated trip pressure set point Indications.
Irput (2) Pressure (2) Comparison of four pressurfrer pressure Irput indications. Some se 5(e) below.
b.
Calibrate R
b.
Calibrate:
(1) Temperature (1) Knoisi resistence stest n uted for RTD coinct dent with knours pressure and power Irput.
Input (2) Pressure (2) Port of 5(b) below.
Irput c.
Test M(2) c.
Sistable trip tester.(1)
Coupertson of four seperate pressur, Indications.
5.
High-Pressurizer Pressure e.
Check (8)
S e.
b.
Calibrete R
b.
Knaun pressure s miled to sensors.
c.
Test M(2) c.
Sistable trip tester.(1) 4-3 muurrutment No.M, M, US, JM,135
=.
TABLE 4.1.2 Minimum Frequencies for Checkt Calibrations and Testing of Cngineered Safety Feature instriseertation Controls (Contd)
Survei11ance Channel Descrtotton Functton Frecuency Sgreet11ance Method
- 13. Safety injection Tank Level
- a. Check 5
- a. Verify that level and pressure indication is between tnd Pressure Instrunents
$44 A,t high high/ low alams for level and pressurs
- b. Calibrate R
- b. Known pressure and differenttal pressure app 1ted to pressure and level sensors.
- c. Check R
- c. Functlonal Check on high and low level alams.
l
- 14. Boric Acid Tank Level Switches
- a. Test E
- a. Pts, tank below low-level alam point to verify switch operatton.
- 15. Boric Acid Heat Tracing Systen
- a. Check D
- a. Observe temperature recorders for proper readings.
- a. Check 5
- a. Co m are four t. b a.t pressuee Indications.
Circuits
- b. Test
- R
- b. Signal to eeter relay adjusted with test device to verify MSIV circutt logic.
- 17. SIRW Tank Temperature
- a. Check M
- a. Comare independant te m erature readouts.
Indication and Alams
- b. Calibrate R
- b. Known resistance applied to indicating loop.
- 18. Low-Pressure Safety injection
- a. Check P
- a. Observe valve is open with air supply isolated.
Flow Control Valve CV-300F 1
l
- 19. Safety Injection Bottle
- a. Check P
- a. Ensure each valve open by observing valve positton indicatton Isolation Vaives and valve itself. Then lock open breakers and control power key switches.
- 20. Safety inject ton Wintflow a'. Check P
- a. Vertfy valves open and HS-3027 and 3056 po,sittened to Valves Cv-3027, 3056 maintain them open.
NOTES:
(1) Calibration of the sensors is performed during calibration of Itam 5(b). Table 4.1.1.
(2)All monthly tests will be done on only one channel at a time to prevent protection system actuation.
(3)Calibratton of the sensors is perfomed during calibration of Item 7(b). Table 4.1.1.
(4) Required when PCS is >1500 psta.
4-8 wt no. se. ss. Jer. ns.136
/""" ' "'
l
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- - - ~-
1 TAstC 4.1.3 Minlaus Frequencies for Checks, Calibrations and Testing of Miscellaneous instrtsamtation and Controts (Cont'd)
Surveiitonce Channel Description Function Frequency Surveillance method
' 1.
Source Range Neutron Monitors
- a. Check 5
Comerison of both charnet conot rate indications when in l
s.
service.
- b. Test P
b.,
Internet Test Signals.
- c. Calibrate R-c.
Channet alignment through measurcarnt/adjustmannt of internal test points.
2.
Prime y Rod Position
- a. Check a.
Comparison of output data with secondary RPIS.
IrC cation System
- b. Check M
b.
Check of power deperwient insertion limits ammitoring systeva.
- c. Calibrate R
c.
Physicalty measured rod drive psition used to verify system accuracy. Check rod position interlocks.
3.
Secondary Rod Position
- a. Check S
' a.
Comparison of output data with primary RPIS.
Indication System
- b. Check M
b.
Same as 2(b) above.
- c. Calibrate R
Sanne as 2tc} above, includifts out-of-sequence storm c.
function.
4 Area Monitors
- e. Check D
Normat readings observed and internal test signets Note: Process Monitor a.
used to verify instrument operation.
Surveillance Requirements
- b. Catibrate R
b.
Exposure to known enternal radiation source.
are tocated in Tables
- c. Test M
c.
'4.24-1 and 4.24-2 Detector exposed to remote operated radiation check source or integrat electronic check source.
5.
Emergency Plan Radiation
- a. Calibrate
.A a.
Empesure to known radiation source.
Inst rtsnents.
- b. Test M
b.
Battery heck.
,~
6.
Environmental Monitors
- a. Check M
a.
Operationat check.
- b. Calibrate A
b.
Verify airflow irwitcator.
7.
Pressurizer Levet
- a. Check S
Comparison of two wide and two narrow range Ynstrtsments a.
- b. Calibrate R
iih a t level readings.
[
b.
Known differentist pressure applled to sensor.
- c. Test M
Signal to meter relay adjusted with test device.
c.
4-10 D
a="*=t "o-30,37,38,83,I15,118,136 i
n febte 4.21.1 (Cont'd)
ALf(3NA7E $NUIDOWN se0NI}DRIMG !NSTRtsef WiAiION SURVEIttAact atouimE8EENT$
Surveitionce Channet Description isnetton FreoJency Survelliance swthod Internet test signal (performued under febte 4.1.3) l
- 7. Source Range meutron peanitor a.
Test Prior to a.
l stert m "
(u!-1/3C)
I
- 8. Auniilary feeduater tcw Suction j
l Pressure 5esitch (PS-0T&10) a.
Catibrate Refueting a.
Apply known preneure to pressure sensor.
cycle Compare trukpendent levet readines.
- 9. Stav Tank Lewt Indication a.
Check (1) cuarterly a.
-(11-03328) b.
Calibrate 3efurtine b.
Appty knoen dif ferentist pressure l
cycle to f ewet sensor.
- 10. Auxiliary feed ater flow tate (2)
Indication Apply knamn dif ferential pressure to sensor (s).
(f I-07273) a.
Catibrate Refueling a.
(fI-07498) cycie
- 11. Aunittery feeduster Ftou controt (3) a.
Check mefueting a.
Verify Control.
valves (CV-GT2T & CV-0T&9) cycle
- 12. Auxiliary feedseter Pump Intet a.
Check Refuelin0 a.
Verify Centrol.
stem. Vatwe (CV-05229) cycle NOTES:(1) cuarterly checks are not rewired den the plant is less than 325'F.
(2) satisfies Table 4.1.3-15 Sepirement.
(3) See specification 4.96.
(4) Pr!or to each start w, If not done previous weet.
(ment page is 4-90) 4-88 Amene-ne me.ss, 122,136
~.
1 l
t 5.3 NUCLEAR STEAM SUPPLY SYSTEM (NSSS) (Cont'd) 5.3.2 Reactor Core and Control a.
The reactor core shall approximate a right circular cylinder with an equivalent diameter of about 136 inches and an active height of about 132 inches, b.
The reactor core shall consist of approximately 43,000-Zircaloy 4 clad fuel rods containing slightly enriched uranium in the form of sintered UO2 pellets. The fuel rods shall be grouped into 204 assemblies.
A core plug or plugs may be used to replace one or more fuel assemblies subject to the analysis of the resulting power distribution.
c.
The fully-loaded core shall contain approximately 211,000 pounds UO and approximately 56,000 pounds of Zircaloy-4.
Poison may be 2
placed in the fuel bundles for long term reactivity control,
- d. -The core excess reactivity shall be controlled by a combination of boric acid chemical shim, cruciform control rods, and mechanically fixed absorber rods where required.
Forty-five l
control rods shall be distributed throughout the core as shown in Figure 3-5 of the FSAR.
Four of these control rods may consist of part-length absorbers, 5.3.3 Emeraency Core Coolina System An emergency core cooling system shall be installed consisting of various subsystems es:h with internal redundancy. These subsystems shall include four safety injection tanks, two high-pressure and two low pressure safety injection pumps, a safety injection and refueling water storage tank, and interconnecting piping as shown in Section 6 of the FSAR.
5-3 Amendment No. 27,fS,83,136