ML20080M230
| ML20080M230 | |
| Person / Time | |
|---|---|
| Site: | Limerick  |
| Issue date: | 02/16/1995 |
| From: | Stolz J Office of Nuclear Reactor Regulation |
| To: | Philadelphia Electric Co |
| Shared Package | |
| ML20080M235 | List: |
| References | |
| NPF-39-A-089, NPF-85-A-052 NUDOCS 9503030155 | |
| Download: ML20080M230 (40) | |
Text
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PHILADELPHIA ELECTRIC COMPANY DOCKET N0. 50-352 LIMERICK-GENERATING STATION. UNIT 1-
'l AMENDMENT TO FACILITY OPERATING LICENSE i
Amendment No. 89' License No. NPF-39
'1.
The Nuclear Regulatory Comission (the Commission) has found that:
A.
The application for amendment by Philadelphia Electric Company (the licensee) dated October 29, 1993, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act),
and the Comission'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 Comission; i
C. -There is reasonable assurance (i) that the activities authorized by i
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 Comission's regulations; D.
The issuance of this amendment will not be inimical to the comon 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 I
the Comission's regulations and all applicable requirements have been satisfied.
I I
9503030155 950216 PDR ADOCK 05000352 p
t 8
t 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. NPF-39 is hereby amended to read as follows:
Technical Soecifications The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No.
89, are hereby incorporated into this license.: Philadelphia Electric Company shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
3.
This license amendment is effective as of its date of issuance and shall be implemented within 30 days.
1 FOR THE NUCLEAR REGULATORY COMMISSION L
hn. Stolz, Direct'oJ-r,ect Directorate M D vision of Reactor Projects - I/II Office of Nuclear Reactor Regulation
Attachment:
Changes to the Technical Specifications i
Date of Issuance: February 16, 1995 1
l
.l
.i ATTACHMENT TO LICENSE AMENDMENT NO. 89 FACILITY OPERATING LICENSE NO. NPF-39
~ DOCKET NO. 50-352 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 the area of change.
Remove Insert 2-4 2-4 8 2-8 8 2-8 3/4 3-3 3/4 3-3 t
3/4 3-6 3/4 3-6 3/4 3-7 3/4 3-7 3/4 3-11 3/4 3-11 3/4 3-18 3/4 3-18 3/4 3-23 3/4 3-23 3/4 3-26 3/4 3-26 3/4 3-27 3/4 3-27 3/4 3-31 3/4 3-31 3/4 6-19 3/4 6-19 3/4 6-22 3/4 6-22 3/4 6-24 3/4 6-24 3/4 6-31 3/4 6-31 B 3/4 3-1 B 3/4 3-1 B 3/4 3-2 B 3/4 3-2 2-m
TABLE 2.2.1-1 REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS r-5 ALLOWABLE g
FUNCTIONAL UNIT TRIP SETPOINT VALUES n*
- 1. Intermediate Range Monitor, Neutron Flux-High s 120/125 divisions s 122/125 divisions of full scale of full scale E
- 2. Average Power Range Monitor:
- a. Neutron Flux-Upscale, Setdown s 15% of RATED THERMAL POWER s 20% of RATED THERMAL POWER
~
- b. Neutron Flux-Upscale
- 1) During two recirculation loop operation:
a) Flow Biased s 0.66 W+ 66% with s 0.66 W+ 68% with a maximum of amaximumof b) High Flow Clamped s 115% of RATED s 117% of RATED THERMAL POWER THERMAL POWER
- 2) During single recirculation loop operation:
a) Flow Biased s 0.66 W+ 61%
s 0.66 W+ 63%
b) High Flow Clamped Not Required,
Not Reauired,
OPEMBLE OPEMBLE
- c. Inoperative N.A.
N.A.
- d. Downscale 2 4% of RATED 23% of RATED m1 THERMAL POWER THERMAL POWER
- 3. Reactor Vessel Steam Dome Pressure - High s 1037 psig s 1057 psig
- 4. Reactor Vessel Water Level - Low, Level 3 2 12.5 inches above instrument 2 11.0 inches above zero*
instrument zero g
- 5. Main Steam Lina Isolation Valve - Closure s 8% closed s 12% closed
- 6. DELETED DELETED DELETED g
- 7. Drywell Pressure - High s 1.68 psig s 1.88 psig
- 8. Scram Discharge Volume Water Level - High
- a. Level Transmitter s 260' 9 5 8" elevation **
s 261' 5 5 8" elevation s
- b. Float Switch s 260' 9 5 8" elevation **
s 261' 5 5 8" elevation
- 9. Turbine Stop Valve - Closure s 5% clos s 7% clos g
- 10. Turbine Control Valve Fast Closure, Trip 011 Pressure - Low 2 500 psig 2 465 psig a
- 11. Reactor Mode Switch Shutdown Position N.A.
N.A.
- 12. Manual Scram N.A.
N.A.
See Bases Hgure B 3/4.3-1.
Equivalent. to 25.45 gallons / scram discharge volume.
______.----,--__.-v-
LIMITING SAFETY SYSTEM SETTINGS BASES REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS (Continued) i 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 enough.above the fuel to assure that there is adequate protection for the fuel and pressure limits.
i 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 MSIVs are closed automatically from measured parameters such as high steam flow, low reactor water level, high steam tunnel temperature and low steam line pressure.
The MSIVs closure scram anticipates the pressure an,d flux transients which could follow MSIV closure and thereby protects reactor vessel pressure and fuel thermal / hydraulic Safety Limits.
6.
DELETED 7.
Drywell Pressure-Hiah 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 to the primary containment. The trip setting was selected as low as possible without causing spurious trips.
l I
LIMERICK - UNIT 1 B 2-8 Amendment No. 89
TABLE 3.3.1-1 (Continued)
REACTOR PROTECTION SYSTEM INSTRUMENTATION 5
n 7
APPLICABLE MINIMUM OPERATIONAL OPERABLE CHANNELS c3 FUNCTIONAL UNIT CONDITIONS PER TRIP SYSTEM (a)
ACTION w
6.
DELETED DELETED DELETED
- DELETED l-7.
Drywell Pressure - High 1, 2(h) 2 1
8.
Scram Discharge Volume Water Level High I
a.
Level Transmitter 1, 2 2
1 5(i) 2 3.
b.
Float Switch 1, 2 2
1
{
5(i) 2 3
9.
Turbine Stop Valve - Closure 1(j) 4(k) 6 10.
Turbine Control Valve Fast Closure, Trip Oil Pressure - Low 1(j) 2(k) 6 k
11.
Reactor Mode Switch Shutdown 3
Position 1, 2 2
1 i
k 3, 4 2
7 E
5 2
3 12.
Manual Scram 1, 2 2
1.
cn 3, 4 2
8 5
2 9
f e
,--.--e-
,,-..--w a
,y c -
.c-
-s-a gr.
. wv n
a-w n
7
~
TABLE 3.3.1-2 E
REACTOR PROTECTION SYSTEN RESPONSE TINES m
RESPONSE TINE y
FUNCTIONAL UNIT (Seconds) 7 I.
Intermediate Range Monitore g
- a. Neutron Flux - High N.A.
-i
- b. Inoperative N.A.
2.
Average Power Range Monitor *:
- a. Neutron Flux - Upscale, Setdown N.A.
- b. Neutron Flux - Upscale 1)
Flow Biased 50.09 2)
High Flow Clamped 50.09
- c. Inoperative N.A.
w2
- d. Downscale N.A.
{
3.
Reactor Vessel Steam Dome Pressure - High 50.55 4.
Reactor Vessel Water Level - Low, level 3 sl.05 5.
Main Steam Line Isolation Valve - Closure
$0.06 6.
DELETED
. DELETED
[-
I 7.
Drywell Pressure - High
.N.A.
(
8.
Scram Discharge Volume Water Level - High E
- a. Level Transmitter N.A.
- b. Float Switch N.A.
g 9.
Turbine Stop Valve - Closure 50.06 cn 10.
Turbine Control Valve Fast Closure, Trip Oil Pressure - Low
$0.08**
11.
Reactor Mode Switch Shutdown Position N. A..
12.
Manual Scram N.A.
- Neutron detectors are exempt from response time testing.
Response time shall be measured from the detector output or from the input of the first electronic component in the channel.
- Measured from start of turbine control valve fast closure.
TABLE 4.3.1.1-1 u-REACTOR PROTECTION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS E
CHANNEL OPERATIONAL 7
CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH FUNCTIONAL UN; CHECK TEST CALIBRATION (a) SURVEIILANCE REQUIRED cz Z
l.
a.
Neutron Flux - High S/U.S(b)
S/U(c),W R
2 S
W(j)
R 3,4,5 b.
Inoperative N.A.
W(j)
N.A.
2,3,4,5 2.
Average Power Range Monitor (f):
a.
Neutron Flux -
S/U,S(b)
S/U(c),W SA 2
Upscale, Setdown S
W(j)
SA 3,5(k) b.
Neutron Flux - Upscale 1)
Flow Biased S,D(g)
S/U(c),Q W(d)(e),SA 1
w1
- 2) High Flow Clamped S
S/U(c),Q W(d)(e),SA 1
w c.
Inoperative N.A.
Q(j)
N.A.
1, 2, 3, 5(k) d.
Downscale S
Q SA 1
3.
Reactor Vessel Steam Dome Pressure - High S
Q R
1,2(h) 3.
4.
Reactor Vessel Water Level-
,{
Low, level 3 S
Q R
1, 2 a
5.
Main Steam Line Isolation f
Valve - Closure N.A.
Q R
1 6.
DELETED DELETED DELETED DELETED DELETED l
,0 7.
Drywell Pressure - High S
Q R
1, 2 8.
Scram Discharge Volume Water Level - High a.
Level Transmitter S
Q R
1, 2, 5(1) b.
Float Switch N.A.
Q R
1, 2, 5(i) m--
.m w.
. +
c
,e
,-v.....
_,s.., -,.-i
-.. f..
TABLE 3.3;2-1 Cg ISOLATION ACTUATION INSTRUMENTATION 5g MINIMUM
. APPLICABLE.
ISOLATION OPERABLE CHANNELS OPERATIONAL-TRIP FUNCTION SIGNAL (a)
PER TRIP SYSTEM (b)
CONDITION
_' ACTION.
c=
-]
1.
MAIN STEAM LINE ISOLATION a.
Low, low-Level 2 B
2 1, 2, 3 21 2)
Low, Low, Low-Level 1 C
2 1, 2, 3 21 b.
DELETED DELETED DELETED DELETED DELETED l
c.
Main Steam Line Pressure - Low P
2 1
22' d.
Main Steam Line Flow - High E
2/!ine 1, 2, 3 20 w
e.
Condenser Vacuus - Low Q
2 1,
2**, 3**
21 f.
Outboard MSIV Room Temperature - High-F(f) 2 1, 2, 3
'21 g
g.
Turbine Enclosure - Main Steam g
Line Tunnel Temperature - High F(f) 14 1, 2, 3.
21-f h.
Manual Initiation NA 2
1, 2,- 3 24 F
2.
-RHR SYSTEM SHUTDOWN COOLING MODE ISOLATION M
a.
Reactor Vessel Water Level Low - Level 3 A
2 1, 2, 3 23-b.
Reactor Vessel (RHR Cut-In.
23 Permissive) Pressure - High V
2 1, 2, 3-c.
Manual Initiation NA 1
1,2,3-24
.. ~
. ~.
TABLE 3.3.2-2 Cg ISOLATION ACTUATION INSTRUMENTATION SETPOINTS 5g ALLOWABLE TRIP FUNCTION TRIP SETPOINT VALUE 1.
MAIN STEAM LINE ISOLATION a.
- 1) Low, low - Level 2 2 - 38 inches
- 2 - 45 inches 2)
Low, Low, Low - Level 1 2 - 129 inches
- 2 - 136 inches b.
DELETED DELETED DELETED
- 1 c.
Main Steam Line Pressure - Low 2 756 psig 2 736 psig d.
Main Steam Line mi Flow - High 5 108.7 psid s 111.7 psid e.
Condenser Vacuum - Low 10.5 psia 210.1 psia /s 10.9 psia f.
Outboard MSIV Room Temperature - High 5 192*F s 200*F.
g.
Turbine Enclosure - Main Steam Line Tunnel Temperature - High s 165 F
- s 175 F F
h.
Manual Initiation N.A.
N.A.
2.
RHR SYSTEM SHUTDOWN COOLING MODE ISOLATION g
if a.
{
Low - Level 3 2 12.5 inches
- 2 11.0 inches im b.
Reactor Vessel (RHR Cut-in Permissive ) Pressure - High 5 75 psig s 95 psig c.
Manual Initiation N.A.
N.A.
,g,.
,,._m.
,_._,,m.
.-..-%,,,-,.m
-w w.
...m..,r.
,,, =. -,
,,...,,c%
v,
..,,4
,,,.,r...
,3w..m..e.,
...,,w,
I
]_APLE 3.3.2-3 ISOLATION SYSTEM INSTRUMENTATION RESPONSE TIME TRIP FUNCTION RESPONSE TIME (Seconds)#
1.
MAIN STEAM LINE ISOLATION a.
Low, low - Level 2 s 13(a)**
i 2)
Low, Low, Low - Level I s 1.0*/5 13(a)**
i b.
DELETED DELETED l
j c.
Main Steam Line Pressure - Low-s 1.0*/5 13(a)**
d.
Main Steam Line Flow - High s 0.5*/s 13(a)**
e.
Condenser Vacuum - Low N.A.
f.
Outboard MSIV Room Temperature - High N.A.
~
g.
Turbine Enclosure - Main Steam Line Tunnel Temperature - High N.A.
h.
Manual Initiation N.A.
^t 2.
RHR SYSTEM SHUTDOWN COOLING MODE ISOLATION a.
Reactor Vessel Water Level Low - Level 3 s 13(a) b.
Reactor Vessel (RHR Cut-In Permissive) Pressure - High N.A.
c.
Manual Initiation N.A.
3.
REACTOR WATER CLEANUP SYSTEM ISOLATION a.
RWCS a Flow - High 5 13##
b.
RWCS Area Temperature - High N.A.
c.
RWCS Area Ventilation 6 Temperature - High N.A.
i d.
SLCS Initiation N.A.
I e.
Low, low - Level 2 s 13(a) f.
Manual Initiation N.A.
LIMERICK - UNIT 1 3/4 3-23 Amendment No. 29, 89 f
.4 TABLE 3.3.2-3 (Continued)
{
ISOLATION SYSTEM INSTRUMENTATION RESPONSE TIME l
TRIP FUNCTION RESPONSE TIME (Seconds)#
f.
Outside Atmosphere To Refueling Area a Pressure - Low N.A.
j g.
Reactor Enclosure Manual Initiation-N.A.
h.
Refueling Area Manual Initiation N. A.-
TABLE NOTATIONS (a)
Isolation system instrumentation response tf.ne specified includes 10 seconds diesel generator starting and 3 seconds for sequence loading delays.
(b)
DELETED
- Isolation system instrumentation response time for MSIV only. No diesel 5
generator delays assumed for MSIVs.
- Isolation system instrumentation response time for associated valves except MSIVs.
- Isolation system instrumentation response time specified for the Trip Function actuating each valve group shall be added to isolation time shown in Tables 3.6.3-1, 3.6.5.2.1-1 and 3.6.5.2.2-1 for valves in each valve group to obtain ISOLATION SYSTEM RESPONSE TIME for each valve.
- With 45 second time delay.
i i
LIMERICK - UNIT 1 3/4 3-26 Amendment No. 6, 89
TABLE 4.3.2.1-1 ISOLATION ACTUATION-INSTRUMENTATION SURVEILLANCE REQUIREMENTS C
' #i CHANNEL
.0PERATIONAL l 5 CHANNEL FUNCTIONAL CHANNEL COM)ITIONS FOR Wi!CH R
TRIP FUNCTION CHECK TEST CALIBRATION SURVEILLANCEREDUIRED h
1.
MAIN STEAM LINE ISOLATION 5
a.
- 1) Low, Low, Level 2 S
Q R
1, 2, 3
~
- 2) Low, Low, Low - Level 1 S
Q R
1, 2, 3 b.
DELETED DELETED DELETED DELETED DELETED c.
Main Steam Line Pressure - Low S
Q R
1 d.
Flow - High S
Q R
1, 2, 3 C
e.
Condenser Vacuum - Low S
Q R
1, 2**, 3**
,i>
f.
Outboard MSIV Room Temperature - High S
Q R
1, 2, 3 g.
Turbine Enclosure - Main Steam Line Tunnel Temperature - High S
Q R
1, 2, 3 k1 h.
Manual Initiation N.A.
R N.A.
1, 2, 3 2.
RHR SYSTEM SHUTDOWN COOLING MODE ISOLATION if a.
Low - Level 3 S
Q R
1, 2, 3 ea b.
Reactor Vessel (RHR Cut-In S
Q R
1, 2, 3 i
y Permissive) Pressure - High y
c.
Manual Initiation
.N.A.
'R N.A.
1, 2, 3
,.__...~%
-.4 r.
...,4
.,w a
~*
TABLE 4.3.2.1-1 (C:ntinued)
ISOLATION ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS E
CHANNEL OPERATIONAL CHANNEL FUNCTIONAL CHANNEL C0f0lTIONS FOR WHICH g
j-TRIP FUNCTION CHECK TEST CALIBRATION SURVEILLANCEREQUIRED n
7.
SECONDARY CONTAINNENT ISOLATION i
a.
Reactor Vessel Water Leve'l e
'5 Low, Low - Level 2 S
Q R
1, 2, 3 -
w b.
Drywell Pressure ## - High S
Q R
1, 2, 3 c.l.
Refueling Area Unit i Ventilation Exhaust Duct Radiation - High S
Q R
2.
Refueling Area Unit 2 Ventilation Exhaust Duct Radiation - High S
Q R
d.
Reactor Enclosure Ventilation Exhaust Duct Radiation - High S
Q R
1, 2, 3 e.
Outside Atmosphere To Reactor Enclosure a Pressure - Low N.A.
M Q
1, 2, 3 w
h f.
Outside Atmosphere To Refueling Area a Pressure - Low N.A.
M Q
g.
Reactor Enclosure Manual Initiation N.A.
R N.A.
1, 2, 3 g
l b h.
Refuelin Area
(
Manual I!itiation N.A.
R N.A.
5 handling irradiated fuel in the refueling area secondary containment, or (2 during CORE
- Required when (1)(3) during operations with a potential for draining the reactor vessel with)the vessel ALTERATIONS, or head removed and fuel in the vessel.
p
- When not administratively bypassed and/or when any turbine stop valve is open.
g
- During operation of the associated Unit 1 or Unit 2 ventilation exhaust system.
ifThese trip functions (2a, 6b, and 7b) are common to the RPS actuation trip function.
l
?
m.
2 2
-+,-e v--
. 6--.
--<-m
- _se, m
s '.
TABLE 3.6.3-1 Cg PART A - PRIMARY CONTAINMENT ISOLATION VALVES 5
E INBOARD OUTBOARD ISOL.
c PENETRATION FUNCTION ISOLATION ISOLATION MAX.ISOL.
SIGNAL (S), NOTES
. P&ID -~
NUMBER BARRIER BARRIER TIME.IF APP.
IF APP.
[
(SEC)(26)
(20) 0038 CONTAINMENT INSTRUMENT 59-1005B (CK)
NA 59-GAS SUPPLY - HEADER 'B' HV59-129B 7
C,H,S 003D-2 CONTAINMENT INSTRUMENT 59-Ill2(CK)
NA GAS SUPPLY TO ADS VALVES HV59-1518 45 M
59~
E&K 007A(B,C,D)
MAIN STEAM LINE HV41-1F022A 5*
C,E,F,P,Q 6
41 1-
'A'(B,C,D)
(B,C,D)
HV41-lF028A 5*
C,E,F,P,Q 6
1 w
4 001B 45 ^
EA 6
V E
.01B NA 6,1 T B, THIS TABLE) k 008 MAIN STEAM LINE DRAIN HV41-IF016 30-C,E,F,P,Q 4-41 3
HV41-lF019 30 C,E,F,P,Q a
009A FEEDWATER 41-IF010A(CK) HV41-lF074A(CK)
NA 41 NA 41-1036A(CK)
NA HV41-130B 45
.M HV41-133A 45 HV41-109A NA 32 O
HV41-IF032A(CK)
NA HV55-1F105 30 7
HV44-IF039(CK)
NA (X-981 41-10L6(X-98, NA 31 X-44)
TABLE 3.6.3-1 (Continued).
PART A - PRIMARY CONTAINMENT ISOLATION VALVES 5
E INBOARD OUTBOARD ISOL.
PENETRATION FUNCTION ISOLATION ISOLATION MAX.ISOL.
SIGNAL (S).
NOTES P&ID c:
NUMBER BARRIER BARRIER TIME.IF APP.
IF APP.
}
(SEC)(26)
(20) 025 DRYWELL PURGE SUPPLY HV57-121(X-201A) 5**
B,H,S,U,W,R,T 3,11,14 57
~
HV57-123 5**
B,H,S,U,W,R,T 3,11,14 HV57-109 6**
B,H,S,U,W,R,T 11 (X-201A)
HV57-131 5**
B,H,S,U,W,R,T 11 (X-201A)
HV57-135 6**
B,H,S,U,W,R,T 11 HYDROGEN RECOMBINER "B" HV57-163 9
B,H,R,S 3,11,14 INLET FV-C-00-101B 90 8,H,R,S 11 026 DRYWELL PURGE EXHAUST HV57-114 5**
B,H,S,U,W,R,T 3,11,14,33 57
?
HV57-Ill 15**
B,H,S,U,R,T 11 SV57-139 5
10 HV57-115 6**
B,H,S,U,W,R,T 11,33 HV57-Il7 5**
B,H,S,U,R,T 11 SV57-145 5
B,H,R,S 11
(
HYDROGEN RFCOMBINER "A" HV57-161 9
B,H,R,5 3,11,14 g
INLET
{
FV-C-D0-101A 90 B,H,R,S 11 N
027A CONTAINMENT INSTRUMENT 59-1128(CK)
NA 59 g
GAS SUPPLY TO ADS VALVES HV59-151A 45 M
~
H,M,&S 028A-1 RECIRC LOOP SAMPLE HV43-lF019 10 B
43 HV43-IF020 10 B
~
028A-2 DRYWELL H2/02 SAMPLE SV57-132 5
B,H,R,S
' ll 11 57 SV57-142 5
B,H,R,5 E
028A-3 DRYWELL H2/02 SAMPLE SV57-134 5
B,H,R,S 11 57 SV57-144 5
B,H,R,S 11
..,,,._m
.~.m,,
-,,,.,---,c
-,s
+
4
=
w s.-
TABLE 3.6.3-1 (Continued)
PART A - PRIMARY CONTAllWENT ISOLATION VALVES i
m M
^
INBOARD OUTBOARD ISOL.
PENETRATION FUNCTION ISOLATION ISOLATION MAX.ISOL.
SIGNAL (S),
NOTES P&ID c
NUMBER BARRIER BARRIER-TIME.IF APP.
IF APP.
(SEC)(26)
(20)-
040G-1 ILRT DATA ACQUISITION 60-1057 NA 11 60-
~
60-1058 NA 11 040G-2 ILRT DATA ACQUISITION 60-1071 NA 11 60 60-1070 NA 11 040H-1 CONTAINMENT INSTRUMENT 59-1005A(CK)
NA 59 GAS SUPPLY - HEADER 'A' HV59-129A 7
C,H,5 042 STANOBY LIQUID CONTROL 48-IF007(CK)
NA 48 (X-Il6)
HV48-IF006A 60 29 043B MAIN STEAM SAMPLE HV41-lF084 10 B
41 T'
HV41-1F085 10 B
044 RWCU ALTERNATE 41-1017 NA 5,31 41 RETURN 41-1016(X-9A, NA X-98)
PSV41-112 NA f
045A(B,C,0) LPCI INJECTION 'A'(B,C,D) HV51-lF041A(B,C,NA 9,22 51
[
42A(B,C, 7
9,22 o
D)
HV51-lF017A 38
,P (B,C,D)
O w 050A-1 DRYWELL PRESSURE HV42-147B
-45 10 42 INSTRUMENTATION
~."
053 DRYWELL CHILLED WATER HV87-128 60 C,H 11,.
87 SUPPLY - LOOP 'A' HV87-120A-60 C,H 11 w*
HV87-125A 60 C,H.
11
...-..u...-..
4..
~.
,.cv..w-
.,. ww, a
--c,e
,.v.
..w.
,a
.,-m....
.....~
.y_
1 TABLE 3.6.3-I (Continued) bh FART B - PRIMARY CONTAINMENT ISOLATION EXCESS FLOW CHECK VALVES 9
E INBOARD OUTBOARD ISOL.
i PENETRATION FUNCTION ISOLATION ISOLATION MAX.ISOL.
SIGNAL (S), NOTES.
P&ID' c:
NUMBER BARRIER BARRIER TIME.IF APP.
IF APP.
25
. (SEC)(26)
(20)
-e 003A-1 INSTRUMENTATION
'D' XV41-lF0700 1
41 MAIN STEAM LINE FLOW XV41-IF0730 003A-2 INSTRUMENTATION
'A' XV43-IF003A 1
43 RECIRC PUMP SEAL PRESSURE 003C-1 INSTR. - HPCI STEAM' FLOW XV55-lF024A 1
55 4
003C-2 INSTR. - HPCI STEAM FLOW XV55-lF024C 1
55 0030-1 INSTR.
'A' MAIN STEAM XV41-lF070A
- 1 41-u, g
LINE FLOW XV41-lF073A i'
007A(B,C,D)
INSTR.
'A'(B,C,D) MAIN (HV41-IF022Af B, :
5*
C,E,F,P,Q 6
41 STEAM LINE PRESSURE C,D) SEE PARI A HV41-IF028A 5*
C,E,F,P,Q 6
THIS TABLE)
,C,D) SEE RT A THIS TABLE)
' EA HV40-IF001B 45 g
,K,P) SEE
. 6.
RT A THIS il TABLE)
J XV40-101B(F, 1,6 g
K,P) f 020A-1 INSTR - RPV LEVEL XV42-lF045B l-42' jd 020A-2 INSTR
'B' LPCI DELTA P XV51-1028 1
51 gg 020A-3 INSTR
'D' LPCI DELTA P XV51-103B 1
51 0208-1 INSTR - RPV LEVEL'
.XV42-lF045C' 1
42 0208-2 INSTR-
'C' LPCI DELTA P XV51-102C 1
' 51
.m m.m.. - - -.
.,-me__
..mw~
,+,.,~,..,,.,..s_.---_._.,w-..,,,.-,o,.,w
.u ec
.wm
....w v, %a
.a m.,
,n.,
- O
-3/4.3' INSTRUMENTATION BASES 3/4.3.1 REACTOR' PROTECTION SYSTEM INSTRUMENTATION The reactor protection system automatically initiates a reactor scram to:
a.
Preserve the integrity of the fuel cladding.
b.
Preserve the-integrity of the reactor coolant system.
c.
Minimize the energy which must be adsorbed following a loss-of-coolant accident, and d.
Prevent inadvertent criticality.
This specification provides the limiting conditions for operation necessary to preserve the ability of the system to perform its intended i
function even during periods when instrument channels may be out of service because of maintenance. When necessary, one channel may be made inoperable for brief intervals to conduct required surveillance.
The reactor protection system is made up of two independent trip systems.
There are usually four channels to monitor each parameter with two channels in each trip system. The outputs of the channels in a trip system are combined in a logic so that either channel will trip that trip system.
The tripping of both trip systems will produce a reactor scram.
The system meets the intent of IEEE-279 for nuclear power plant protection systems.
Specified surveillance intervals and surveillance and maintenance outage times have been determined in accordance with NEDC-30851P, " Technical Sgecification Improvement Analyses for BWR Reactor Protection System, as a proved by the NRC and documented in the NRC Safety Evaluation Re) ort (SER) letter to T. A.
Pickens from A. Thadani dated July 15, 1987. The )ases for t e trip settings of RPS are discussed in the bases for Specification 2.2.1.
Automatic reactor trip upon receipt of a high-high radiation signal from the Main Steam Line Radiation Monitoring System was removed as the result of an analysis performed by General Electric in NED0-31400A. The NRC approved the results of this analysis.as documented in the SER (letter to George J. Beck, BWR Owner's Group from A. C. Thadani, NRC, dated May 15,1991).
The measurement of response time at the specified frequencies provides assurance that the protective functions associated with each channel are completed within the time limit assumed in the safety analyses.
No credit was taken for those channels with response times indicated as not applicable.
Response time may be demonstrated by any series of sequential, overlapping or total channel test measurement, provided such tests demonstrate the total channel response time as defined.
Sensor response time verification may be demonstrated by either (t sensors with certified response times.(2) utilizing replacem 1
1 LIMERICK - UNIT 1 B 3/4 3-1 Amendment No. 53, 89
INSTRUMENTATION BASES 3/4.3.2 ISOLATION ACTUATION INSTRUMENTATION This specification ensures the effectiveness of the instrumentation used to mitigate the consequences of accidents by prescribing the OPERABILITY trip setpoints and response times for isolation of the reactor systems. When i
nece;sary, one channel may be inoperable for brief intervals to conduct required surveillance.
Specified surveillance intervals and maintenance outage times have been determined in accordance with NEDC-30851P Supplement 2 " Technical Specification Improvement Analysis for BWk Instrumentation Common to RPS and ECCS Instrumentation" as approved by the NRC and documented in the NRC Safety Evaluation Report (SER) (letter to D. N. Grace from C. E. Rossi dated January 6, 1989) and NEDC-31677P-A, " Technical Specification Improvement Analysis for BWR the NRC SER (letter to S. D. Floyd from C.p) roved by the NRC and documented in Isolation Actuation Instrumentation " as a E. Rossi dated June 18,1990).
Automatic closure of the MSIVs upon receipt of a high-high radiation signal from the Main Steam Lit Radiation Monitoring System was removed as the result of an analysis performet by General Electric in NED0-31400A. The NRC approved the results of this analysis as documented in the SER (letter to George J. Beck, BWR Owner's Group from A. C. Thadani, NRC, dated May 15,1991).
Some of the trip settings may have tolerances explicitly stated where both the high and low values are critical and may have a substantial effect on safety. The setpoints of other instrumentation, where only the high or low end of the setting have a direct bearing on safety, are established at a level away from the normal operating range to prevent inadvertent actuation of the systems involved.
Except for the MSIVs, the safety analysis does not address individual sensor response times or the response times of the logic systems to which the sensors are connected. For D.C. operated valves, a 3 second delay is assumed before the valve starts to move.
For A.C. operated valves, it is assumed that the A.C.
power supply is lost and is restored by startup of the emergency diesel generators.
In this event, a time of 13 seconds is assumed before the valve starts to move.
In addition to the pipe break, the failure of the D.C. operated is concurrent with the valve is assumed; thus the signal delay (sensor response) ding delay.
10-second diesel startup and the 3 second load center loa The safety analysis considers an allowable inventory loss in each case which in turn determines the valve speed in conjunction with the 13-second delay.
It follows that checking the valve speeds and the 13-second time for emergency power establishment will establish the response time for the isolation functions.
i Operation with a trip set less conservative than its Trip Setpoint but within its specified Allowable Value is acceptable on the basis that the
)
' difference between each Trip Setpoint and the Allowable Value is an allowance 1
for instrument drift specifically allocated for each trip in the safety analyses.
3/4.3.3 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION The emergency core cooling system actuation instrumentation is provided to initiate actions to mitigate the consequences of accidents that are beyond the ability of the operator to control. This specification provides the OPERABILITY requirements, trip setpoints and response times that will ensure effectiveness of the systems to provide the design protection. Although the instruments are listed by system, in some cases the same instrument may be used to send the actuation signal to more than one system at the same time.
LIMERICK - UNIT 1 B 3/4 3-2 Amendment No. 33, 53, 69, 89
,gmog:\\
y-G E
UNITED STATES i
NUCLEAR REGULATORY COMMISSION g... + j#
WASHINGTON, D.C. 20555-0001 4
PHILADELPHIA ELECTRIC COMPANY j
DOCKET NO. 50-353 LIMERICK GENERATING STATION. UNIT 2 AMEN 0 MENT TO FACILITY OPERATING LICENSE 4
Amendment No. 52 i
License No. NPF-85 1.
The Nuclear Regulatory Comission (the Comission) has found that:
A.
The application for amendment by Philadelphia Electric Company (the licensee) dated October 29, 1993, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act),
and the Comission's rules and regulations set forth in 10 CFR Chapter I; t
B.
The facility will operate in conformity with the application, the provisions of the Act, and the rules and regulations of the Comission; 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 Comission's regulations; D.
The issuance of this amendment will not be inimical to the comon 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 Comission's regulations and all applicable requirements have been satisfied.
t-l r
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. NPF-85 is hereby amended to read as follows:
Techni:a1 Snecifications The Technical Specifications contained in Appendix A and the Environmental Protection Plan contained in Appendix B, as revised through Amendment No.
52, are hereby incorporated into this license.
Philadelphia Electric Company shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
3.
This license amendment is effective as of its date of issuance and shall be implemented within 30 days.
FOR THE NUCLEAR REGULATORY COMNISSION A
ohn F. Stolz, Direct P
s:t Directorate I-2 Division of Reactor Projects - I/II Office of Nuclear Reactor Regulation Attachsent:
Changes to the Technical Specifications Date of Issuance: February 16, 1995 t
4 1
i I
ATTACHMENT TO LICENSE AMENDMENT NO. 52 FACILITY OPERATING LICENSE NO. NPF-85 DOCKET NO. 50-353 I
i 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 the area of change.
i Remove Insert j
2-4 2-4 B 2-8 B 2-8 3/4 3-3 3/4 3-3 i
3/4 3-6 3/4 3-6 3/4 3-7 3/4 3-7 3/4 3-11 3/4 3-11 3/4 3-18 3/4 3-18 3/4 3-23 3/4 3-23 3/4 3-26 3/4 3-26 3/4 3-27 3/4 3-27 3/4 3-31 3/4 3-31 3/4 6-19 3/4 6-19 3/4 6-22 3/4 6-22 3/4 6-24 3/4 6-24 3/4 6-31 3/4 6-31 B 3/4 3-1 B 3/4 3-1 B 3/4 3-2 B 3/4 3-2
TABLE 2.2.1-1 REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS r-52 ALLOWABLE FUNCTIONAL UNIT TRIP SETPOINT VALUES
- 1. Intermediate Range Monitor, Neutron Flux-High s 120/125 divisions s 122/125 divisions of full scale of full scale-E
- 2. Average Power Range Monitor:
t
- . Neutron Flux-Upscale, Setdown s 15% of RATED THERMAL POWER s 20% of RATED m
THERMAL POWER
- b. Neutron Flux-Upscale
- 1) During two recirculation loop operation:
a) Flow Biased s 0.66 W+ 62%, with s 0.66 W+ 64%, with a maximum of a maximum of b) High Flow Clamped s 115% of RATED s 117% of RATED THERMAL POWER THERMAL POWER
- 2) During single recirculation loop operation:
a) Flow Biased s 0.66 W+ 57%,
s 0.66 W+ 59%,
b) High Flow Clamped Not Required Not Required rp OPERABLE OPERABLE
- c. Inoperative N.A.
N.A.
- d. Downscale a 4% of RATED a 3% of RATED THERMAL POWER THERMAL POWER k
- 3. Reactor Vessel Steam Dome Pressure - High s 1096 psig s 1103 psig a
- 4. Reactor Vessel Water Level - Low, Level.3 a 12.5 inches above instrument a.11.0 inches above zero*
instrument zero
- 5. Main Steam Line Isolation Valve - Closure s 8% closed s 125 closed
- 6. DELETED DELETED DELETED l
t
- 7. Drywell Pressure - High s 1.68 psig s 1.88 psig w
- 8. Scram Discharge Volume Water Level - High
- a. Level Transmitter s 261' 1 1/4" elevation **
s 261' 9 1/4" elevation
- b. Float Switch s 261' 1 1/4" elevation **
s 261' 9 1/4" elevation
- See Bases Figure B 3/4.3-1.
- Equivalent to 25.58 gallons / scram discharge volume.
I
LIMITING SAFETY SYSTEM SETTINGS BASES REACTOR PROTECTION SYSTEM INSTRUMENTATION SETPOINTS (Continued) 1 4.
Reactor Vassel 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 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 MSIVs are closed automatically from measured parameters such as high steam flow, low reactor water level, high steam tunnel temperature, and low steam line pressure.
The MSIVs 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.
l 6.
DELETED 7.
Drywell Pressure-Hiah i
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 i
to minimize the possibility of fuel damage and reduce the amount of energy being added to the coolant and to the primary containment. The trip setting was selected as low as possible without causing spurious trips.
LIMERICK - UNIT 2 B 2-8 Amendment No. 52
TABLE 3.3.1-1 (Continued)
Cj(
REACTOR PROTECTION SYSTEM INSTRUMENTATION 5
R APPLICABLE MINIMUM OPERATIONAL OPERABLE CHANNELS c5 FUNCTIONAL UNIT CONDITIONS PER TRIP SYSTEM (a)
ACTION
--e N
6.
DELETED DELETED DELETED DELETED l
7.
Drywell Pressure - High 1,2(h) 2 1
8.
Scram Discharge Volume Water Level - High a.
Level Transmitter 1, 2 2
1 5(i) 2 3
b.
Float Switch 1, 2 2
1
{
5(i) 2 3
9.
Turbine Stop Valve - Closure 1(j) 4(k) 6 10.
Turbine Control Valve Fast Closure, Trip 011 Pressure - Low 1(j) 2(k) 6 11.
Reactor Mode Switch Shutdown 5
Position 1, 2 2
1 1
3, 4 2
7 3
5 2
3 E
g:
12.
Manual Scram 1, 2 2
1 3, 4 2
8 1
g 5
2 9
l i
i
.um
________.__._______.____.___._,_.__,____..___,________mm._____m a
_,__m
,,,._7,
..y,ym,
,p-.
3,,,...,,,_g.,
,,,,.,,..p
,9-
TABLE 3.3.1-2 f
REACTOR PROTECTION SYSTEM RESPONSE TIMES RESPONSE TIME 23 FUNCTIONAL UNIT (Seconds) n
[*
1.
gg a.
Neutron Flux - High N.A.
Il b.
Inoperative N.A.
~
2.
Average Power Range Monitor *:
a.
Neutron Flux - Upscale, Setdown N.A.
b.
Neutron Flux - Upscale 11 Flow Biased 50.09 2)
High Flow Clamped 50.09 c.
Inoperative N.A.
d.
Downscale N.A.
o, s
- ]
3.
Reactor Vessel Steam Dome Pressure - High 50.55 Es 4.
Reactor Vessel Water Level - Low, Level 3 s1.05 5.
Main Steam Line Isolation Valve - Closure 50.06-6.
DELETED
' DELETED l'
7.
Drywell Pressure - High N.A.
kI 8.
Scram Discharge Volume Water Level - High o
a.
Level Transmitter N.A.
t l'
b.
Float Switch N.A.
E 9.
Turbine Stop Valve - Closure 50.06 10.
Turbine Control Valve Fast Closure,
[,
Trip 011 Pressure - Low
$0.08**
11.
Reactor Mode Switch Shutdown Position N.A.
12.
Manual Scram N.A.
- Neutron detectors are exempt from response time testing. Response time shall be measured from the detector output or from the input of the first electronic component in the channel.
- Measured from start of turbine control valve fast closure.
-_,, _ =
,w,
%,,~
-.~.-.. -- -
v-
,e
+
- <-a v
- -. - ~
-n-w
.--w
wa TABLE 4.3.1.1-1
~.
C REACTOR PROTECTION SYSTEM INSTRUMENTATION SURVEILLANCE REQUIREMENTS E
CHANNEL OPERATIONAL CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH FUNCTIONAL UNIT CHECK TEST CALIBRATION (a) SURVEILLANCE REQUIRED' Q
l.
a.
Neutron Flux - High S/U.S(b)
S/U(c),W R
2 m
S W(j)
R 3,4,5 b.
Inoperative N.A.
W(j)
N.A.
2,3,4,5 2.
Average Power Range Monitor (f):
a.
Neutron Flux -
S/U,5(b)
S/U(c),W SA 2
Upscale, Setdown S
W(j)
SA 3, 5(k) b.
Neutron Flux - Upscale 1)
Flow Biased S,0(g)
S/U(c),Q W(d)(e),SA 1
- 2) High Flow Clamped S
S/U(c),Q W(d)(e),SA 1
O c.
Inoperative N.A.
Q(j)
N.A.
1, 2, 3, 5(k) d.
Downscale S
Q SA 1
3.
Reactor Vessel Steam Dome Pressure - High S
Q R
.1, 2(h)
N o
4.
Reactor Vessel Water Level-h Low, Level 3 S
Q R
1, 2 m
?,
5.
Main Steam Line Isolation g:
Valve - Closure N.A.
Q R
1 6.
DELETED DELETED DELETED DELETED DELETED l
g 7.
Drywell Pressure - Hign S
Q R
1, 2 8.
Scram Discharge Volume Water Level - High a.
Level Transmitter S
Q R.
1, 2, 5(i)'
b.
Float Switch N.A.
Q R
1, 2, 5(i) a
~
m m
w..
,e
,,._y c
+
O
- o; ;
TABLE 3.3.2-1 I
ISOLATION ACTUATION INSTRUMENTATION
~ [
M n
MINIMUM APPLICABLE ISOLATION OPERABLE CHANNELS OPERATIONAL c.
z TRIP FUNCTION SIGhAL fa) PER TRIP SYSTEM (b)
C0lWITION-ACTION i
I.
MAIN STEAM LINE ISOLATION N
a.
Low, Low-Level 2 8
2 1, 2, 3
-21!
2)
Low, Low, Low-Level 1 C
2 1, 2, 3.
- 21 b.
DELETED DELETED DELETED DELETED DELETED.
l:
c.
Main Steam Line Pressure - Low P
2
' l 22 w1 d.
Flow - High E
2/line 1, 2,'3 20' e.
Condenser Vacuum - Low Q.
2 1,
2**, 3**
21 f.
Outboard MSIV Room Temperature --High F(f) 2 1,. 2, 2 3 '.
211 k
. Turbine Enclosure - Main Steam g.
Line Tunnel Temperature - High F(f) 14 1,. 2, 3 ;
21-
?,
h.
Manual Initiation MA 2
. 1, 2, 3
. 24.-
2.
RHR SYSTEM SHUTDOWN COOLING MODE ISOLATION a.
Reactor Vessel Water Level-Low - Level 3 A
2 1, ? ~ 3 ~
23 i
b.
Reactor Vessel-(RHR Cut-In Permissive) Pressure - High
- V 2
- 1, 2, 3 23.-
c.
Manual Initiation NA 1
. 1, 2,~3 24-
_______,.,we.
.m.
w
-man,.-..4 a.m.
.es.y..,._
w...,-.%,
s
.m5
.--.-w,.g-m,_.,,.,4-%,
-.%w
,gue.m_.,
+.
w-
+<.
m
.- 3 TABLE'3.3.2-2 E
ISOLATION ACTUATION INSTRUMENTATION SETPOINTS 9
R ALLOWABLE
{
TRIP FUNCTION TRIP SETPOINT VALUE E
1.
MAIN STEAM LINE ISOLATION m
a.
- 1) Low, Low - Level 2 2 - 38 inches
- 2 - 45 inches
- 2) Low, Low, Low - Level 1 2
129 inches
- 2 - 136 inches b.
DELETED DELETED DELETED c.
Main Steam Line Pressure - Low a 756 psig a 736 psig d.
{
Flow - High s 122.1 psid s 123 psid Y
e.
Condenser Vacuum - Low 10.5 psia 210.1 psla/s 10'9 psia
- t E
f.
Outboard MSIV Room Temperature - High s 192*F s 200*F g.
Turbine Enclosure - Main Steam k
Line Tunnel Temperature - High s 165'F s 175'F i
3.
,g h.
Manual Initiation M.A.
N.A.
n" 2.
RHR SYSTEM SHUTDOWN COOLING MODE ISOLATION a.
Low - Level 3 2 12.5 inches
- 2 11.0 inches U
b.
Reactor Vessel (RHR Cut-in Permissive) Pressure - High s 75 psig s 95 psig c.
Manual. Initiation M.A.
' N.A.
b m-m
..m.
m m
mm m
-s
-e m
--rv.
-t-- -.
v.
=m-mm-*we
-*o.=
w a
-s'
=+<w.
d
- +-
TABLE 3.3.2 -
ISOLATION SYSTEM INSTRUMENTATION RESPONSE TIME TRIP FUNCTION RESPONSE TIME (Seconds)#
j; 1.
MAIN STEAM LINE ISOLATION a.
Low, Low - Level 2 s 13(a)**
l 2)
Low, Low, Low - Level 1 s 1.0*/s 13(a)**
t b.
DELETED DELETED l
c.
Main Steam Line Pressure - Low
$ 1.0*/s 13(a)**
d.
Flow - High 5 0.5*/s 13(a)**
e.
Condenser Vacuum - Low N.A.
f.
Outboard MSIV Room Temperature - High N.A.
g.
Turbine Enclosure - Main Steam Line Tunnel Temperature - High N.A.
h.
Manual Initiation N.A.
2.
RHR SYSTEM SHUTDOWN COOLING MODE ISOLATION a.
Reactor Vessel Water Level Low - Level 3 s 13(a) b.
Reactor Vessel (RHR Cut-In Permissive) Pressure - High N.A.
c.
Manual Initiation N.A.
3.
REACTOR WATER CLEANVP SYSTEM ISOLATI0ff a.
RWCS a Flow - High 5 13##
b.
RWCS Area Temperature - High N.A.
c.
RWCS Area Ventilation a Temperature - High N.A.
d.
SLCS Initiation N.A.
e.
Low, low - Level 2 s 13(a) f.
Manual Initiation N.A.
LIMERICK - UNIT 2 3/4 3-23 Amendment No. 52
e TABLE 3.3.2-3 (Continu:d)
ISOLATION SYSTEM INSTRUMENTATION RESPONSE TIME
- !!P FUNCTION RESPONSE TIME (Seconds)#
f.
'Outside Atmosphere To Refueling Area A Pressure - Low N.A.
g.
Reactor Enclosure Manual Initiation N.A.
h.
Refueling Area Manual Initiation N.A.
TABLE NOTATIONS (a)
Isolation system instrumentation response time specified includes 10 seconds diesel generator starting and 3 seconds for sequence loading delays.
(b)
DELETED
- Isolation system instrumentation response time for MSIV only.
No diesel generator delays assumed for MSIVs.
- Isolation system instrumentation response time for associated valves except MSIVs.
- Isolation system instrumentation response time specified for the Trip function actuating each valve group shall be added to isolation time shown in Tables 3.6.3-1, 3.6.5.2.1-1 and 3.6.5.2.2-1 for valves in each valve group to obtain ISOLATION SYSTEM RESPONSE TIME for each valve.
- With 45 second time delay.
t 1
i LIMERICK - UNIT 2 3/4 3-26 Amendment No. 52 9
TABLE 4.3.2.1-1 a
ISOLATION ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL OPERATIONAL CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH TRIP FUNCTION CHECK TEST CALIBRATION SURVEILLANCE REQUIRED cz U
l.
MAIN STEAM LINE ISOLATION
~
a.
Low, Low, level 2 S
Q R
1,. 2, 3 2)
Low, Low, low - Level 1 S
Q R
1, 2, 3 b.
DELETED DELETED DELETED DELETED DELETED-c.
Main Steam Line Pressure - Low S
Q R
1 d.
Main Steam Line Flow - High S
Q R
1, 2, 3 w
e.
Condenser Vacuum - Low S
Q R
1, 2**, 3**
f.
Outboard MSIV Room Temperature - High S
Q R
1, 2, 3 g.
Turbine Enclosure - Main Steam Line Tunnel Temperature - High S
Q R
1, 2, 3 h.
Manual Initiation N.A.
R N.A.
- 1. 2, 3 E
2.
RHR SYSTEM SHUTDOWN COOLING MODE ISOLATION a.
Low - Level 3 S
Q R
1, 2, 3 b.
Reactor Vessel (RHR Cut-In S
Q R
1, 2, 3 g
Permissive) Pressure - High c.
Manual Initiation N.A.
R N.A.
1, 2, 3 m
m u
w wwv*
~.
TABLE 4.3.2.:.-;. (Continued)
C ISOLATION ACTUATION..NSTRUMEN"A" ION SURVEILLANCE REQUIREMENTS z
CHANNEL OPERATIONAL p
CHANNEL FUNCTIONAL CHANNEL' CONDITIONS FOR WHICH:
TRIP FUNCTION CHECK TEST CALIBRATION SURVEILLANCE REQUIRE E
7.
SECONDARY CONTAllelENT ISOLATION Z
a.
Reactor Vessel Water Level Low, Low - Level 2 S
Q R
1, 2, 3 m
b.
Drywell Pressure ## - High S
Q R
1, 2, 3' c.1.
Refueling Area Unit 1 Ventilation Exhaust Duct Radiation - High S
Q R
2.
Refueling Area Unit 2 Ventilation Exhaust Duct Radiation - High S
Q R
- f.
w d.
Reactor Enclosure Ventilation
- 1 Exhaust Duct Radiation - High S
Q R
1,2,3 w
~
d, e.
Outside Atmosphere To Reactor Enclosure a Pressure - Low N.A.
M Q
1, 2, 3
,i ~
f.
Outside Atmosphere To Refueling Area a Pressure - Low N.A.
M Q
g.
Reactor Enclosure k
Manual Initiation N.A.
R N.A.
1, 2, 3 a
I h.
Refueling Area g
Manual Initiation N.A.
R N.A.
5
- Required when (1) handling irradiated fuel in the refueling area secondary contr.inment, or (2) during CORE O
ALTERATIONS, or-(3) during operations with a potential for draining the reactor vessel with the vessel head removed and fuel in the vessel.
,u
- When not administratively bypassed and/or when any turbine stop valve is open.
- During operation of the associated Unit 1.or Unit 2 ventilation exhaust system.
- These trip functions (2a, 6b, and 7b) are common to the'RPS-actuation trip function.-
l i
~.
TABLE 3.6.3-1 PART A - PRIMARY CONTAINMENT ISOLATION VALVES R
INBOARD OUTBOARD ISOL.
PENETRATION FUNCTION ISOLATION ISOLATION MAX.ISOL.
SIGNAL (S),. NOTES P&ID.
E NUMBER B.ARRIER BARRIER TIME.IF APP.IF APP.
Z (SEC)(26)
(20)
~
0038 CONTAINMENT INSTRUMENT 59-2005B (CK)
NA 59-GAS SUPPLY - HEADER
'B' HV59-2298 7
C,H,S 003D-2 CONTAINMENT INSTRUMENT 59-2112(CK)
NA GAS SUPPLY TO ADS VALVES HV59-2518 45 M
59 E&K 007A(B,C,D)
MAIN STEAM LINE HV41-2F022A 5*
C,E,F,P,Q 6
41 l.
'A'(B,C,D)
(B,C,0) 2 (B,C,D)
HV40-2F0018 45 EA 6
i (F,K,P)
G (XV40-201B NA 6,1 (F,K,P)
SEE PART B, THIS
- TABLE) 008 MAIN STEAM LINE DRAIN HV41-2F016 30 C,E,F,P,Q 4
41 HV41-2F019 30 C,E,F,P,Q 009A FEEDWATER 41-2F010A(CK)
NA 41 HV41-2F074A(CK)
NA j
41-2036A(CK)
NA re HV41-230B 45 g:
HV41-233A 45 HV41-209A NA 32 HV41-2F032A(CK)
NA m
3 HV55-2F105 30 7
HV44-2F039(CK)
NA (X-98)-
41-2016(X-98, NA 31 X-44)
+,
4..--
w--
e ww+
w a+w m
=------mu---
~
r i
TABLE 3.6.3-1 (Continued).
W
]
PART A - PRIMARY CONTAINMENT ISOLATION VALVES 5
R INBOARD OUTBOARD ISOL.
E PENETRATION FUNCTION ISOLATION ISOLATION MAX.ISOL.
SIGNAL (S),
NOTES P&ID Z
NUMBER BARRIER' BARRIER TIME.IF APP. IF APP.
(SEC)(26)
(20)'
m 025 DRYWELL PURGE SUDPLY HVS7-221(X-201A) 5**'
B,H,S,U,W,R,T 3,11,14.-
57 HV57-223 5**
8,H,S,U,W,R,T 3,11,14 HV57-209 6**
B,H,S,U,W,R,T'
-11 (X-201A)
HV57-231 5**
B,H,S,U,W,R,T 11 (X-201A)
HV57-235 6**
B,H,S,U,W,R,T 11 HYDROGEN RECOMBINER HV57-263 9
B,H,R,S 3,11,14 "B"
INLET.
FV57-00-201B 90 B,H,R,S 11,34 w1 026 DRYWELL PURGE EXHAUST HV57-214 5**
B,H,S,U,W,R,T 3,11,14,33 57 m
4, HV57-211 15**
B,H,S,U,R,T 11 no SV57-239 5
10
-HV57-215 6**
B,H,S,U,W,R,T 11,33_
i i
HV57-217 5**
B,H,S,U,R,T 11 SV57-245 5
B,H,R,S 11' HYDROGEN RECOMBINER HV57-261 9
B,H,R,S 3,11,14 "A"
INLET FV57-00-201A 90 8,H,R,S 11,34 027A CONTAINMENT INSTRUMENT 59-2128(CK)
NA 59
, [
GAS SUPPLY.TO ADS VALVES HV59-251A 45 M
g H,M,&S 028A-1 RECIRC LOOP SAMPLE HV43-2F0'19 10
-B 43 HV43-2F020 10 B-028A-2 DRYWELL'H2/02 SAMPLE SV57-232 l
SV57-242.
5 B,H,R,S 11 57 5
B,H,R,S-11 028A-3 DRYWELL H2/02 SAMPLE SV57-234 5
B,H,R,S 11
.57 SV57-244
-5 B,H,R,S 11 i
TABLE 3.6.3-1 (Continued)
PART A - PRIMARY CONTAINMENT ISOLATION VALVES e
9 INBOARD OUTBOARD ISOL.
E PENETRATION FUNCTION ISOLATION ISOLATION MAX.ISOL.
SIGNAL (S),
NOTES P&ID NUMBER BARRIER BARRIER TIME.IF APP. IF APP.
(SEC)(26)
(20) m 040G-1 ILRT DATA ACQUISITION 60-2057 NA 11 60 60-2058 NA 11 040G-2 ILRT DATA ACQUISITION 60-2071 NA 11 60 60-2070 NA 11 040H-1 CONTAINMENT INSTRUMENT 59-2005A(CK)
NA 59 GAS SUPPLY - HEADER
'A' HV59-229A 7
C,H,S 042 STANDBY LIQUID CONTROL 48-2F007(CK)
NA 48 (X-Il6)
HV48-2F006A 60 29 043B MAIN STEAM SAMPLE HV41-2F084 10 8
41 HV41-2F085 10 8
044 RWCU ALTERNATE 41-2017 NA 5,31 41 RETURN 41-2016(X-9A, NA X-98)
PSV41-212 NA Y
045A(8,C,D)
LPCI INJECTION 'A'(B,C,D) HV51-2F041A(B,C, NA 9,22 51 D)(CK)
(
HV51-242A(B,C, 7
9,22 D)
HV51-2F017A 38 (8,C,D)
M 050A-1 DRYWELL PRESSURE HV42-2478 45 10 42 INSTRUMENTATION 053 DRYWELL CHILLED WATER HV87-228 60 C,H 11 87 SUPPLY - LOOP
'A' HV87-220A 60 C,H.
11 HV87-225A 60 C,H
.11
u.a w.
.--- a----
.a +
<-w--
,--------r,-
-:--w-J
TABLE 3.6.3-1 (Continued)
PART B - PRIMARY CONTAINMENT ISOLATION EXCESS FLOW CHECK VALVES e
E INBOARD OUTBOARD ISOL.
E PENETRATION FUNCTION ISOLATION ISOLATION MAX.ISOL.
SIGNAL (S),
NOTES P&ID p
NUMBER BARRIER BARRIER TIME.IF APP. IF APP.
(SEC)(26)
(20) m 003A-1 INSTRUMENTATION
'D' XV41-2F0700 1
41 MAIN STEAM LINE FLOW XV41-2F0730 003A-2 INSTRUMENTATION
'A' XV43-2F003A 1
43 RECIRC PUMP SEAL PRESSURE 003C-1 INSTR. - HPCI STEAM FLOW XV55-2F024A 1
55 003C-2 INSTR. - HPCI STEAM FLOW XV55-2F024C 1
55 003D-1 INSTR.
'A' MAIN STEAM XV41-2F070A 1
41
[
LIhE FLOW XV41-2F073A 007A(B,C,D) INSTR.
'A'(B,C,D) MAIN (HV41-2F022A(B, 5*
C,E,F,P,Q 6
41 STEAM LINE PRESSURE C,D) SEE PART A (HV41-2F028A 5*
C,E,F,P,Q 6
THIS TABLE)
(B,C,D) SEE PART A THIS TABLE)
(HV40-2F0018 45 EA 6
[
(F,K,P) SEE o
PART A THIS
(
TABLE) a XV40-201B(F, 1,6 K,P) 020A-1 INSTR - RPV LEVEL XV42-2F045B 1
42 M
020A-2 INSTR
'B' LPCI DELTA P XV51-2028 1
51 020A-3 INSTR
'D' LPCI DELTA P XV51-203B 1
51 0208-1 INSTR - RPV LEVEL XV42-2F045C 1
42 0208-2 INSTR
'C' LPCI DELTA P XV51-202C
-1 51
_,-.,,,,..,s-.
,.. - - ~, -., - -
--.-,ev----
4..-
-e..-
.m m
---m e
.e 3/4.3 INSTRUMENTATION
.I BASES 1
3/4.3.1 REACTOR PROTECTION SYSTEM INSTRUMENTATION The reactor protection system automatically initiates a reactor scram to:
a.
Preserve the integrity of the' fuel cladding.
]
b.
Preserve LM integrity of the reactor coolant system.
1 i
c.
Minimize the energy which must be adsorbed following a i
loss-of-coolant accident, and d.
Prevent inadvertent criticality.
j This specification provides the limiting conditions for operation j
necessary to preserve the ability of the system to perform its intended function even during periods when instrument channels may be out of service because of maintenance. When necessary, one channel may be made inoperable for brief intervals to conduct required surveillance.
]
The reactor protection system is made up of two independent trip systems.
There are usually four channels to monitor each parameter with two channels in
.)
each trip system.
The outputs of the channels in a trip system are combined t
in a logic so that either channel will trip that trip system. The tripping of both trip systems will produce a reactor scram. The system meets the intent of IEEE-279 for nuclear power plant protection systems. Specified surveillance intervals and surveillance and maintenance outage times have been determined in accordance with NEDC-30851P, " Technical Specification Improvement Analyses for BWR Reactor Protection System," as approved by the NRC and documented in the NRC Safety Evaluation Report (SER) (letter to T. A.-
Pickens from A. Thadani dated July 15, 1987. The bases for the trip settings of RPS are discussed in the bases for Specification 2.2.1.
i Automatic reactor trip upon receipt of a high-high radiation signal from the Main Steam Line Radiation Monitoring System was removed as the result of an analysis performed by General Electric in NED0-31400A. The NRC approved the results of this analysis as documented in the SER (letter to George J. Beck, j
BWR Owner's Group from A. C. Thadani, NRC, dated May 15,1991).
The measurement of response time at the specified frequencies provides l
assurance that the protective functions associated with each channel are completed within the time limit assumed in the safety analyses. No credit was taken for those channels with response times indicated as not applicable.
Response time may be demonstrated by any series of sequential, overlapping or total channel test measurement, provided such tests demonstrate the total channel response time as defined.
Sensor response time verification may be demonstrated by either (1) inplace, onsite or offsite test measurements, or (2) utilizing replacement sensors with certified response times.
i LIMERICK - UNIT 2 B 3/4 3-1 Amendment No. 17, 52 j
O INSTRUMENTATION l
BASES 3/4.3.2 ISOLATION ACTUATION INSTRUMENTATION I
This specification ensures the effectiveness of the instrumentation used to mitigate the consequences of accidents by prescribing the OPERABILITY trip setpoints and response times for isolation of the reactor systems. When necessary, one channel may be inoperable for brief intervals to conduct required surveillance.
Specified surveillance intervals and maintenance outage times have been determined in accordance with NEDC-30851P Supplement 2 " Technical Specification Improvement Analysis for BWk Instrumentation Common to RPS and ECCS Instrumentation" as approved by the NRC and documented in the NRC Safety i
Evaluation Report (SER) (letter to D. N. Grace from C. E. Rossi dated January 6, 1989) tion Actuation Instrumentation "pecification Improvement Analysis for BWR and NEDC-31677P-A, " Technical S the NRC SER (letter to S. D. Floyd from C.p> roved by the NRC and documeented in j
Isola as a E. Rossi dated June 18,1990).
Automatic closure of the MSIVs upon receipt of a high-high radiation i
signal from the Main Steam Line Radiation Monitoring System was removed as the result of an analysis performed by General Electric in NED0-31400A. The NRC approved the results of this analysis as documented in the SER (letter to George J. Beck, BWR Owner's Group from A. C. Thadani, NRC, dated May 15,1991).
l Some of the trip settings may have tolerances explicitly stated where both the high and low values are critical and may have a substantial effect on safety. The setpoints of other instrumentation, where only the high or low end of the setting have a direct bearing on safety, are established at a level away i
from the normal operating range to prevent inadvertent actuation of the systems l
involved.
j Except for the MSIVs, the safety analysis does not address individual sensor
[
response times or the response times of the logic systems to which the sensors are connected.
For D.C. operated valves, a 3 second delay is assumed before the valve starts to move. For A.C. operated valves, it is assumed that the A.C.
power supply is lost and is restored by startup of the emergency diesel generators.
In this event, a time of 13 seconds is assumed before the valve i
starts to move.
In addition to the pipe break, the failure of the D.C. operated i
is concurrent with the l
valve is assumed; thus the signal delay (sensor response) ding delay. The safety i
10-second diesel startup and the 3 second load center loa analysis considers an allowable inventory loss in each case which in turn determines the valve speed in conjunction with the 13-second delay.
It follows that checking the valve speeds and the 13-second time for emergency power establishment will establish the response time for the isolation functions.
Operation with a trip set less conservative than its Trip Setpoint but l
within its specified Allowable Value is acceptable on the basis that the i
difference between each Trip Setpoint and the Allowable Value is an allowance for instrument drift specifically allocated for each trip in the safety j
analyses.
l 3/4.3.3 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION l
The emergency core cooling system actuation instrumentation is provided to initiate actions to mitigate the consequences of accidents that are beyond the j
ability of the operator to control. This specification provides the OPERABILITY requirements, trip setpoints and response times that will ensure effectiveness i
of the systems to provide the design protection. Although the instruments are listed by system, in some cases the same instrument may be used to send the actuation signal to more than one system at the same time.
i k
LIMERICK - UNIT 2 B 3/4 3-2 Amendment No. 77, 32, 52
?
-