ML20114B364
| ML20114B364 | |
| Person / Time | |
|---|---|
| Site: | Hope Creek  |
| Issue date: | 08/17/1992 |
| From: | Chris Miller Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20114B366 | List: |
| References | |
| NUDOCS 9208260230 | |
| Download: ML20114B364 (18) | |
Text
.
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'n UNITED STATES 3
i NUCLEAR REGULATORY COMMISSION WASHINGTON. D.C. 20565 EUBLIC SERVICE lLECTRIC & GAS _ COMPANY ATLANTIC CITY ELECTRIC COMPR E DOCKET N0. 50-3M HOPE CREEK GENERATING SJATION AMENDMENT TO FACILITY OPERATING LICE!(S1 Amendment No. 53 License No. NPF-57 1.
The Nuclear Regulatory Commission (the Commission or the NRC) has found that:
A.
The application for amendment filed by the Public Service Electric &
Gas Company (PSE&G) dated February 3, 1992, as supplemented by letter dated June 16, 1992, complies with the standards and recuirements of the Atomic Energy Act of 1954, as amended (the Act),
anc 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 set forth in 10 CFR Chapter I; 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 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 t) this license amendment, and paragraph 2.C.(2) of Facility Operating Lic nse No. NPF-57 is hereby amended to read as follows:
9208260230 920817 PDR ADOCK 05000354 P
. (2)
Technical Specifications and Environmental Protection Plan The Technical Specifications contained in Appendix A, as revised through Amendment No. 53, and the Environmental Protection Plan contained in Appendix B, are hereby incorporated into the license.
PSE&G shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
3.
The license amendment is effective as of its date of issuance and shall be implemented within 180 days of the date of issuance.
FOR THE NUCLEAR REGULATORY COMMISSION b
C. M
[#CharlesL. Miller, Director Project Directorate 1-2 Division of Reactor Projects - 1/11 Office of-Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications Date of Issuance: August 17, 1992
-~-~w
AlTACHMENT T0 llCENSE AMEN 0 MENT N0. 53 FACllITY OPERATING LICENSE NO. NPF-51 DOCKET NO. SQ-111 Replace the following pages of the Appendix "A" Technical Specifications with the attached pages.
The revised pages are identified by Amendment number and contain vertical lines indicating 'he area of change. Overleaf pages provided tomaintain-documentcompleteness.j Remove Insert 2-5 2-5 B 2-7 8 2-7*
B 2-8 8 2-8 3/4 3-3 3/4 3-3 3/4 3-4 3/4 3-4 3/4 3-5 3/4 3-5*
3/4 3-6 3/4 3-6 3/4 3-7 3/4 3-7 3/4 3-8 3/4 3-8 3/4 3-11 3/4 3-11*
3/4 3-12 3/4 3-12 3/4 3-17 3/4 3-17*
3/4 3-18 3/4 3-18 3/4 3-25 3/4 3-25*
3/4 3-26 3/4 3-26
TABLE 2.2.1-1 REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS (continued)
E N
ALLOWABLE FUNCTIONAL UNIT TRIP SETPOINT VALUES 6.
This ites intentionally blank 7.
Drywell Pressure - High
$ 1.68 psig i 1.88 psig 8.
Scram Discharge Volume Water Level - High a.
Float Switch Elevation 110' 10.5" Elevation 111' O.5" b.
Level Transmitter / Trip Unit Elevation 110' 10.5"*
Elevation 111' 4.5*
9.
Turbine Stop Valve - Closure
$ 5% closed 5 7% closed 10.
Turbine Control Valve Fast Closure, n.
Trip Oil Pressure - Low 1 530 psig 1 465 psig 11.
Reactor Mode Switch Shutdown Position NA NA 12.
Manual Scram NA NA
^80.5" above instrument zero EL 104' 2" for Level Transmitter / Trip Unit A&B (South Header) 83.25" above instrument zero EL 103' 11.25" for Level Transmitter / Trip Unit C&D (North Header)
I t
a II a
E
4
-+
o LIMITING SAFETY SYSTEM SETTINGS,
_ BASES EQC10RPROTECTIONSYSTEMINSTRUMENTATIONSETPOINTS(Continutd)
C Average Power Range Monitor (Continued)
Because the flux distribution associated with uniform rod withdrawals does not involve high local peaks and because several rods must be moved to change power by a significant amount, the rate of power rise is very slow.
Generally the I
heat flux is in near equilibrium with the fission rate.
In an assumed uniform rod withdrawal approach to the trip level, the rate of power rise is not more than 5% of RATED THERMAL POWER per minute and the APRM system would be more than adequate to assure shutdown before the power could exceed the Safety Limit.
The 15% neutron flux trip remains active until the mode switch is placed in the Run position.
The APRM trip system is calibrated using heat balance data taken during i
steady state conditions.
Fission chambers provide the basic input to the system and therefore the monitors respond directly and quickly to changes due to transient operation for the case of the Fixed Neutron Flux-Upscale set-point; i.e. for a power increase, the THERMAL POWER of the fuel will be less than that indicated by the neutron flux due to the time constants of the heat transfer associated with the fuel.
For the Flow Biased Simulated Thermal
. Power-Upscale setpoint, a time constant of 6 t 0.6 seconds is introduced
-into the flow biased APRM in order to simulate the fcel thermal transient characteristics.
A more conservative maximum value is used for the flow biased setpoint as shown in Table 2.2.1-1.
The APRM setpoints were selected to provide adequate margin for the Safety-Limits and yet allow operating margin that reduces the possibility of unneces-sary shutdown.
The flow referenced trip setpoint must be adjusted by the specified formula in Specification-3.2.2 in order to maintain.these margins-when CMFLPD is greater than or equal to FRTP.
3.
Reactor Vessel Steam Dome Pressure-High High pressure in the nuclear system could cause a rupture to the nuclear system process barrier resulting in the release of fission products.
A pressure increa e while operating will also tend _to increase the power of the reactor by compressing voids thus adding reactivity.
The trip will quickly reduce the neutron flux,-counteracting'the pressure increase.
The trip setting is slightly higher than the operating pressure to permit normal operation without spurious
. trips. -The setting provides for a wide margin to the maximum.11owable design l
pressure and. takes into account the location of the pressure measurement compared l
to the highest pressure that occurs in the system during a tonsient.
This trip l -
setpoint is effective at low power / flow conditions when ths. :*aine control valve fast closure and' turbine stop valve closure trip are bypassed.
For a load rejection or turbine trip under these conditions, the transient analysis indicated an adequate margin to the thermal hydraulic limit.
HOPE CREEK B 2-7 1
- i i
LIMITING SAFETY SYSTCM SETTINGS BASES REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS (Continued) 4.
Reactor Vessel Water Level-Low The reactor vessel water level trip setpoint has been used in transient analyses dealing with coolant inventory decrease.
The scram setting was chosen far enough below the normal operating level to avoid spurious trips but high i
enough above the fuel to assure that there is adequate protection for the fuel i
and pressure limits.
5.
Main Steam Line Isolation Valve-Closure The main steam line isolation valve closure trip was provided to limit the amount of fission product release for certain postulated events.
The MSIV's are closed automatically from measured parameters such as high steam flow, low reactor water level, high steam tunnel temperature, and the low steam l
line pressure.
The MSIV's closure scram anticipates the pressure and flux transients which could follow MSIV closure and thereby protects reactor vessel pressure and fuel thermal / hydraulic Safety Limits.
6.
This item intentionally blank 7.
Drywell Pressure-High High pressure in the drywell could indicate a break in the primary pressure boundary systems or a loss of drywell cooling.
The reactor is tripped in order to minimize the possibility of fuel damage and reduce the amount of energy being added to the coolant and the primary containment.
The trip setting was selected as low as possible without causing spurious trips.
8.
Scram Discharge Volume Water Level-High The scram discharge volume receives the water displaced by the motion of the control rod drive pistons during a reactor scram.
Should this volume fill up to a point where there is insufficient volume to accept the displaced water at pressures below 65 psig, control rod insertion would be hindered.
The reac-tor is therefore tripped when the water level has reached a point high enough to indi' ate that it is indeed filling up, but the volume is still great enough c
to accommodate the water from the movement of the rods at pressures below 65 psig when they are tripped.
The trip setpoint for each scram discharge volume is equivalent to a contained volume of approximately 35 gallons of water.
HOPE CREEK S28 Amendment No.53
)
W g..
TABLE 3.3.1-1 (Continued) 3 m
' REACTOR PROTECTION SYSTEM INSTRUMENTATION
=
'l n
t.:
. m x
APPLICABLE MINIMUM
(
OPERATIONAL OPERABLE CHANNELS FUNCTIONAL UNIT
. CONDITIONS PER TRIP SYSTEN (a)
ACTION
\\
. This item intentionally blank:
l 6.
- 7... Drywell Pressure'- High 1, 2(h) 2 1
8.
Scras Discharge Volume Water Level -~High i
a.
Float Switch 1, 2(g) 2 1
l l
5 2
3
,s i
[
b.
Level Transmitter / Trip Unit
,1,2(g) 2 1
5 2
3 II) 4(k) 9.
Turbine Stop Valve - Closure I
6
- 10. Turbine Control Valve Fast Closure, II) 2(k) 6 L
Valve Trip Systee Oil. Pressure - Low I
11.
Reactor Mode Switch Shutdown I
Position 1, 2 2
1 I
3, 4 2
7 5
2 3
I k
i s
- 12. Manual Scram..
1, 2 2
1 i
g 3, 4 2
8
)
i 5
2 9
ae
+
O i
i r
i i
~
TABLE 3.3.1-1 (Continued)
REACTOR PROTECTION SYSTEM INSTRUMENTATION ACTION ACTION 1 Be in at least HOT SHUTDOWN within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />.
ACTION 2 Verify all insertable control rods to be inserted in the core and lock the reactor mode switch in the Shutdown position within one hour.
ACTION 3 Suspend all operations involving CORE ALTERATIONS
- and insert all insertable control rods within one hour.
ACTION 4 Be in at least STARTUP within 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />.
ACTION 5 This ACTION is deleted ACTION 6 Initiate a reduction in THERMAL POWER within 15 minutes and reduce turbine first stage pressure to less than the automatic bypass setpoint within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
ACTION 7 Verify all insertable control rods to be insert 6d within one hour.
ACTION 8 Lock the reactor mode switch in the Shutdown position within one hour.
ACTION 9 Suspend all operations involving CORE ALTERATIONS *, and insert all insertable control rods and lock the reactor mode switch in the SHUTOOWN position within one hour.
"Except replacement of LPRM strings provided SRM instrumentation is OPERABLE per Specification 3.9.2.
l HOPE CREEK 3/4 3-4 Amendment No. 53 l
TABLE S.3. M (Continued)
REACTOR PROTECTION SYSTEM INSTRUMENTATION TABLE NOTATIONS (a)
A channel may be placed in an inoperable status for up to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> for required surveillance without placing the trip system in the tripped condition provided at least one OPERABLE channel in the same trip system is monitoring that parepeter.
(b)
This function shall be automatically bypassed when the reactor mode switch is in the Run position.
(c) Unless adequate shutdown margin has been demonstrated per Specifica-tion 3.1.1, the " shorting links" shall be removed from the RPS circuitry prior to and during the time any control rod is withdrawn *,
'(d)
The non-coincident NMS reactor trip function logic is such that all channels go to both trip systems.
the Minimum OPERABLE Channels Per the Trip System are 4 APRM 2 SRMS.
(e)
An APRM channel is inoperable if there are less than 2 LPRM inputs per level or less than 14 LPRM inputs to an APRM channel.
(f)' This function is not required to be OPERABLE when the reactor pressure vessel head is removed per Specification 3.10.1.
(g) This function shall be automatically bypassed when the reactor moda switch is not in the Run position.
(h) This function is not required to be OPERABLE when PRIMARY CONTAINMENT INTEGRITY is not required.
(i) With any control rod withd vn Not cpplicable to control rods removed per Specification 3.9.10.'
9.10.2.
(j) This function shall be auto.ically bypassed when tambine first stage pressure is < 159.7 psig equivalent to THERMAL POWER lass than 30% of l
RATED THERMA [ POWER. To allow for instrument accuracy, calibration, and drift, a setpoint of 5, 135.7 psig is used.
l (k) Also actuates the E0C-RPT system, 4
l "Not required for control rods removed per Specification 3.9.10.1 or 3.9.10.2.
~
I HOPE CREEK 3/4 3-5 Amendment No. 35 DEC 181989 -
.g.
-s l
w
'l
' TABLE 3.3.1 ^
g o
REACTOR PROTECTION SYSTEM RESPONSE TIMES m
.E 3
RESPONSE TIME
" FUNCTIONAL UNIT (Seconds) lu
- 1. :
i a.
Neutron' Flux.- High~
NA
.b.
-InoperativeL
,NA 2.
Average Power Range Monitor *:..
a.
Neutron'. Flux - Upscale,'Setdown NA b.
Flow Biased Simulated Thermal Power - Ups:: ale
< 0.0s '-
1 c.
Fixed Neutron Flux - Upscale 3.09 d.
Inoperative NA 3.
Reactor: Vessel Steam "Jone' Pressure -' High
< 0.55 t
E 4.
Reactor. Vessel-Water Level - Lov.: Level 3 7 1.05 5.
Main Stean Line, Isolation Valve - Closure 7 0.06 i
Y
' 6.
This item intentionally blank' l
7.
Drywell Pressure - High NA I
8.
Scram Discharge Volume Water Level
- High NA J
l a;
Float Switch NA
.i b.
Level Transmitter / Trip Unit NA-9.
Turbine'Stop; Valve.
Closure
< 0.06
'10.
Turbine Control Valwe: Fast Closure,
. Trip Oil Pressure - Low.
< 0.080-
'11.
Reactor Mode Switch Shutdown Position HA
- 12. Manual Scram 1 NA
" Neutron detectors are exempt from response time-testing. ' Response time shall be measured s(
from'the detector output or from the. input of the first electronic component in the channel.
a
- Not including simulated thermal: power time constant, 6 1 0.6 seconds.
s
[
fMeasured from start of turbine control valve fast closure.
i
.O I
TABLE 4.3.1.1-1 5
A REACTOR PROTECTION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS a
CHANNEL OPERATIONAL x
CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH
}
FUNCTIONAL UNii CHECK TEST CALIBRATION (a) SURVEILLANCE WEQUIRED l
1.
a.
Neutron Flux - High S/U(b) 5 S/U(c),W R
2 5
W R
3, 4, 5
)
b.
Incperative NA W
NA 2,3,4,5 Average Power Range Monitor (I):
2.
a.
Neutron Flux -
5/U(b) 5 S/U(C),W SA 2
i I
Upscale, Setdown S
W SA 3,4,5 l
b.
Flow Biased Simulated I9)
S/U(c),q g(d)(e) SA,R(h) y y
Thermal Power - Upsca'..
5,D u
w c.
Fixed Neutron Flux -
0 Upscale 5
S/U(C),Q W(d), SA 1
l d.
Inoperative NA Q
NA 1,2,3,4,5 3.
Reactor Vessel Steam Dome Q(k)
R 1, 2 Pressure - High 5
4.
Low, Level 3 5
Q(k)
R 1, 2 5.
Main Steam Line Isolation
{
Valve - Closure NA Q
R 1
f6.
This item intentionally blank
="
7.
Drywell E
Pressure - High S
Q(k)
R 1, 2 6
i e
- ~
TABLE 4.3.1.1-1 (Contir-REACTOR PROTECTION SYSTEM INSTRUMENTATION A X tj JIREMENTS 2,
CHANNEL OPERATIONAL Q
CHANNEL FUNCTIONAL.
CONDITIONS TOR WHICH
+
^
FUNCTIONAL UNIT CHECK TEST s,.
SURVEILLANCE REQUIRED S
8.
Scram Discharge Volume Water Level - High a.
Float Switch NA Q
R 1,2,5(I) b.
Level Transmitter / Trip Q) k R
1, 5(3)
Unit S
9.
Turbine Stop Valve - Closure NA Q
R 1
10.
Turbine Control Valve Fast Closure Valve Trip System Oil Pressure - Low NA Q
R 1
(d) 11.
Reactor Mode Switch Shutdown Position NA R
NA 1,2,3,4,5 m
12.
Manual Scram NA W
NA 1,2,3,4,5 W Neutron detectors may be excluded from CHANNEL CALIBRATION.
(b) The IRM and SRM channels shall be determined to overlap for at least % decades during each startup after entering OPERATIONAL CONDITION 2 and the IRM and APRM channels shall be determined to overlap for at least % decades during each controlled shutdown, if not performed within the previous 7 days.
(c) Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> prior to startup, if not performed within the previous 7 days.
(d) This calibration shall consist of the adjustment of the APRM channel to conform to the power values calculated by a heat balance during OPERATIONAL CONDITION 1 when THERMAL POWER > 25% of RATED THERMAL POWER. Adjust the APRM channel if the absolute difference is greater than 2% of RATED THERMAL POWER.
Any APRM channel gain adjustment made in compliance with Specification 3.2.2 shall not be included in determining the absolute difference.
p (e) This calibration shall consist of the adjustment of the APRM flow biased channel to conform to a g
calibrated flow signal.
g-(f) The LPRMs shall be calibrated at least once per 1000 effective full power hours (EFPH) g using the TIP system.
(g) Verify measured core flow (total core flow) to be greater than or equal to established core flow at the r+
y existing recirculation loop flow (APRM % flow).
(h) This calibration shall consist of verifying the 6 1 0.6 second simulated thermal power time constant.
(i) This item intentionally blank (j) With any control rod withdrawn.
Not applicable to control rods removed per Specification 3.9.10.1 or 3.9.10.2.
(k) Verify the tripset point of the trip unit at least once per 92 days.
e
=
5
' TABLE 3.3.'2-1.
A ISOLATION ACTUATION INSTRUMENTATION E
m VALVE ACTUA-
. m TION GROUPS MININUM APPLICABLE.
x
, IRIP FUNCTION OPERATEQY' OPERA 8tE CHANNE OPERATIONAL PERTRIPSYSTEMg)-
SIGNAL CONDITION
').CTION 1.
PRIMARY CONTAllMENT ISOLATION a.
Reactor Vesse? Water Level 1)-
Low Low,' Level 2 1,2,8,9, 2
1,2 3 20 12, 13, 14, 15, 17, 18 2)
Low Low Low, level 1 10, 11, 15, 16 2
1, 2, 3 20 b.
Drywell Pressure - High 1, 8, 9, 10, 2
1, 2 '3 20 11, 12, 13, 14, 15, 16, i
17, 18 Y
c.
Reactor Building Exhaust 1, 8, 9,.12 O
Radiation - High 13, 14, 15, 3
1,2,3 28
-17, 18 d.
Manual Initiation 1,'8, 9, 10 1
1, 2.-3 24 11, 12, 13, 14, 15, 16, 17, 18 2.
SECONDARY CONTAINMENT. ISOLATION a.
r 19 ')
I t
low Low,. Level 2 2
1, 2, 3 and
- 26 I
19 ')
b.
~Drywell Pressure - High 2
1,2,3 26 c.
Refueling Floor Exhaust Radiation - Hig.,
19(c) 3 1, 2, 3 and
- 29 d.
Reactor Building Exhaust gc)
Radiation - High 19 3
1, 2, 3 and
- 28 I
19 ')
e.
Manual Initiation 1-1, 2, 3 and
- 26 l
ee W -
y
j' r
~,
i TABLE 3.3.2-1 (Continued)-
I ISOLATION ACTUATION INSTRUMENTATION 2
A VALVE ACTUA--
TION GROUPS--
MINIMUM
-APPLICABLE OPERATEyY.
OPERABLE CHANNE OPERATIONAL PERTRIPSYSTEM(g) q TRIP FUNCTION
' SIGNAL CONDITION ACTION 3.
MAIN STEAM LINE ISOLATION a.
I!eactor Vessel' Water Level -
1 2
1,2,3 21 Low Low Low, Level 1
' fD) b.
Main Steam Line Radiation -
'2 2
1,2,3M 28 High,lHigh c.
Main Steam Line Pressure'-
1 2
1 22 '
. 3 Low-d.
Main Steara Line. Flow - High 1
2/line 1,2,3 20 A
Condenser' Vacuum Low I'
2 1, 2**, 3**
21:
e.
w4 f.
Main Steam Line Tunnel 1
2/line 1, 2, 3 21-
' Temperature - High m
g.
Manual Initiation 1, 2, 17 2
1,2,3 25 4.
REACTOR WATER CLEANUP SYSTEM ISOLATION I
a.
RWCU A Flow - High 7
1/ Valve ')
'1, 2, 3 23 b.
RWCU A Flow - High, Timer 7
1/ Valve ')
1, 2, 3 23 I
c.
RWCU Area Temperature - High 7
6/ Valve *)
- 1, 2, 3 23 I
d.
RWCU Ar?a Ventilation a 7
6/ Valve ')
1,'2, 3 23 1
I g
Temperatare-High e.
SLCS Initiation 7(I) 1/ Valve *)
1, 2, 5#
23 I
f.
Reactor Vessel Water 7
2/ Valve ')
1, 2, 3 23 I
Level - Low Low, Level 2
<+
f g.
Manual Initiation 7
1/ Valve *)
1, 2, 3 25 I
Y N
g
TABLE 3.3.2-1 (Continued) 25 ISOLATION ACTUATION INSTRUMENTATION g;
h{
TABLE NOTATION i
i l
This table notatiod identifies which valves, in an actuation group, are closed by a particular trip signal.
If I
all valves in the group are closed by the trip signal, only the valve group number will be Ifsted. If only j
certain valves in the group are closed by the trip signal, the valve group number will be listed followed by, in parentheses, a listing of which valves are closed by the trip signal.
TRIP FUNCTION VALVES CLOSED BY SIG8AL 1.
PRIMARY CONTAIPMENT ISOLATION a.
1)
Low Low, Level 2 1 (HV-5834A, HV-5835A, HV-5836A, HV-5837A), 2, 8, 9, *.':, i 13,14,15 (HV-5154, HV-5155),17,18 2)
Low Low Low, Level 1 10, 11, 15(HV-5126 A&B, HV-5152 A&S, HV-5147 HV-5148 HV-5162), 16 b.
Drywell Pressure - High 1 (HV-5834A, HV-5835A, HV-5836A, HV-5837A), 8, 9,1",
d 11, 12, 13, 14, 15, 16, 17, 18 c.
Reactor Building Exhaust Radiation - High 1 (HV-5834A, HV-5835A, HV-5836A, HV-5837A), 8, 9, 12, 13, 14, 15, 17 (HV-5161), 18 d.
Manual Initiation 1 (HV-5834A, NY-5835A, HV-5836A, HV-5837A), 8, 9, 10 11, 12, 13, 14, 15, 16, 17 (HV-5161), IP 2.
SECONDARY CONTAIPMENT ISOLATION a.
19 Low Low, level 2 19 b.
Drywell Pressure - High Refueling Floor Exhaust Radiation - High 19 c.
d.
Reactor Building Exhaust Radiation - High 19 19 e.
Manual Initiation
Y W
TABLE 3. 3. 2 (Continued)..
m ISOLATION ACTUATION-INSTRUMENTATION g
TABLE ~ NOTATION g
n TRIP FUNCTION VALVES CLOSED BY SIGNAL' 3.
-MAIN STEAM LINE ISOLATION
.i a.
Reactor Vessel Water Level 11 (HV-F022A, B, C & D, HV-F028A, E, C & D, HV-F067A, B,
' 1 Low Low Low, Level 1 lC & D HV-F016, HV-F019);
b.
Main-Steam Line Radiation - High, High 2
c.
Main Steam Line Presure - Low 1 (as above)..
.]
d.
Main' Steam Line Flow - High 1 (as above) w e.
Condenser Vacuum - Low 1 (as above) f.
Main' Steam'Line Tunnel 1 (as above)
Temperature - High g.
Manual-Initiation 1 (as above), 2, 17 (SV-J004A-1, 2, 3, 4 & 5) 4.
REACTOR WATER CLEANUP SYSTEM ISOLATION a.
RWCU A Flow - High 7
b.
RWCU A Flow - High, Timer-7 g
a c.
RWCU Area Temperature - High 7
t E.
c r
rm
1 IABIE 3.3.2-2 (Continued)
IS01 All0N ACTUAI10N INSIRUMENIATION SLIPOINIS om ALLOWACLE IRIP FUNC110N TRIP SEIPOINI VALUE m
7 7.
RilR SYSTEM SliUIDOWN COOLING MODE ISOLATION a.
Low, level 3
> 12.5 inches *
> 11.0 inches b.
Reactor Vessel (RitR Cut-in Permissive) Pressure - liigh 5 82.0 psig 5 102.0 psig c.
Manual Initiation NA NA
{
^See Bases Figure B 3/4 3-1.
- lhese setpoints 'are as f ollows:
[
160*f - RWCU pipe chase room 4402 140 f - RWCU pump room and heat exchanger rooms m
135 f - RWCU pipe chase room 4505
- 30 minute time delay.
- 15 minute time delay.
- The hydrogen water chemistry (ifWC) system shall not be placed in service until reactor power reaches 20%
of RATED TilERMAL POWER.
Af ter reaching 20% of RATED TilERMAL POWER, and prior to operating the HWC system, the normal full power background radiation level and associated trip setpoints may be increased to levels previously measured during full power operation with hydrogen injection.
Prior to decreasing below 20% of RATED TitERMAL POWER and af ter the llWC system ha been shutoff, the background level and associated setpoint shall be returned to the normal full power values.
If a power reduction event occurs so that the reactor power is below 20% of RATED TilERMAL POWER without the required setpoint change, control rod motion shall be suspended (except for scram or other emergency actions) until the nec(ssary setpoint adjustment is made.
g M
c> =
a
~
'z uM o
l TABLE 3.3.2-3 l
ISOLATION SYSTEM INSTRUMENTATION RESPONSE TIME TRIP FUNCTION RESPONSE TIME (Seconds)#
1.
PRIMARY CONTAINMENT ISOLATION a.
Low Low, Level 2 NA 2)
Low Low Low, Level 1 NA b.
Drywell Pressure - High NA c.
Reactor Building Exhaust Radiation - High NA d.
Manual Initiation NA 2.
SECONDARY CONTAINMENT ISOLATION a.
Reactor Vessel Water Level-Low Low, level 2 NA b.
Drywell Pressure - High NA c.
Refueling Floor Exhaust Radiation -
1 4.0 High(b) d.
Reactor Building Exhaust 5 4.0 Radiation - High(b) e.
Manual Initiation NA a
3.
MAIN STEAM LINE ISOLATION a.
Reactor Vessel Water Level - Low Low Low, Level 1 1 1.0*/1 13(,),,
b.
Main Steam Line Radiation - High, High(a)(b)
< 13(a)**
7 1.0*/< 13((a)**
c.
Main Steam Line Pressure - Low 7 0.5*/7 13 a),,
d.
Main Steam Line Flow-High e.
Condenser Vacuum - Low RA f.
Main Steam Line Tunnel Temperature - High NA g.
Manual Initiation NA 4.
REACTOR WATER CLEANUP SYSTEM ISOLATION a.
RWCU 4 Flow - High NA b.
RWCU A Flow - High, Timer NA c.
RWCU Area Temperature - High NA d.
RWCU Area Ventilation a Temperature - High NA e.
SLCS Initiation NA f..
Reactor Vessel Water Level - Low Low, Level 2 NA g.
Manual Initiation NA 5.
REACTOR CORE ISOLATICN COOLING SYSTEM ISOLATION a.
RCIC-Steam Line a Pressure (Flow) - High NA b.
RCIC Steam Line a Pressure (Flow) - High, Timer NA c.
RCIC Steam Supply Pressure - Low NA d.
RCIC Turbine Exhaust Diaphragm Pressure - High NA HOPE CREEK 3/4 3-26 Amendment No. 53
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