ML20215E405
| ML20215E405 | |
| Person / Time | |
|---|---|
| Site: | Grand Gulf  (NPF-29-A-020, NPF-29-A-20) |
| Issue date: | 10/06/1986 |
| From: | Butler W Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML20215E407 | List: |
| References | |
| TAC-59440, TAC-61038, TAC-61039, TAC-61931, NUDOCS 8610150281 | |
| Download: ML20215E405 (29) | |
Text
.
p&**aq[C UNITED STATES 4
g
{ ) 3 s. f' ?
NUCLEAR REGULATORY COMMISSION E
IVASHING TON, D. C. 70555 LW
- 8 g %.-[f t
....+
MISSISSIPPI POWER & LIGHT COMPANY MIDDLE SOUTH ENERGY, INC.
SOUTH MISSISSIPPI ELECTRIC POWER ASSOCIATION DOCKET NO. 50-416 GRAND GULF NUCLEAR STATION, UNIT 1 AMENDMENT TO FACILITY OPERATING LICEriSE Amendment Nc. 20 License No. NPF-29 1.
The Nuclear Regulatory Commission (the Commission) has found that A.
The applications for amendment by Mississippi Power & Light Company, Middle South Energy, Inc., and South Mississippi Electric Power Association, (the licensees) dated August 12, 1985 (as amended September 25, 1985 and supplemented October 5 and October 22, 1985 and May 30, 1986), March 21, 1986 (as supplemented May 30,1986),
and July 15, 1986, complies with the standards and requirements of the Atomic Energy Act of 1954, as amended (the Act), and the Com-mission'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 I
in compliance with the Commission's regulations; D.
The issuance of this amendment will not be inimical to the common l
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 l
been satisfied.
2.
Accordingly, the license is amended by changes to the Technical Specifications as indicated in the attachment to this license amendment, and paragraph 2.C.(2) of Facility Operating License No. NPF-29 is hereby amended to read as follows:
l Technical Specifications The Technical Specifications contained in Appendix A and the l
Environmental Protection Plan contained in Appendix B, as revised l
through Amendment No. 20
, are hereby incorporated into this license.
l-Mississippi Power & Light Company shall operate the facility in accordance with the Technical Specifications and the Environmental Protection Plan.
8610150281 861006 l
PDR ADOCK 05000416 P
., 3.
The Technical Specification pages in this amendment are effective when the equipment modifications necessitating the changes on these pages are completed and the affected systems are made operable but not later than startup following the first refueling outage.
FOR THE NUCLEAR REGULATORY COMMISSION Odgimisigned W Walter R. Butler, Director BWR Project Directorate No. 4 Division of BWR Licensing
Attachment:
Changes to the Technical Specifications 4
Date of Issuance: October 6,1986 Previously concurred
- O LK for
[
\\/ ]
PD#4/LA*
PD#4 PM*
OGC*
PD#4/D M0'Brien LKintner:lb fiYoung WButler 09/19/86 09/19/86 09/25/86 g; /[./86 WG/tb
o 3.
The Technical Specification pages in this arrendment are effective when the equipment modifications necessitating the changes on these pages are completed and the affected systems are made operable but not later than startup following the first refueling outage.
FOR THE NUCLEAR REGULATORY COMMISSION b'$
u v
i Walter R. Butler, Director BWR Project Directorate No. 4 Division of BWR Licensing
Attachment:
Changes to the Technical Specifications Date of Issuance: October 6,1986 l
i
ATTACHMENT TO LICENSE AMENDMENT NO. 20 FACILITY OPERATING LICENSE N0. NPF-29 DOCKET N0. 50-416 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. Asterisk page(s) provided to maintain document completeness.*
Remove Insert 3/4 3-27 3/4 3-27*
3/4 3-26 3/4 3-28 3/4 3-28a 3/4 3-29 3/4 3-29*
3/4 3-30 3/4 3-30 3/4 3-31 3/4 3-31 3/4 3-31a 3/4 3-32 3/4 3-32*
3/4 3-32a 3/4 3-32a*
3/4 3-33 3/4 3-33 3/4 3-34 3/4 3-34 3/4 3-34a 3/4 3-35 3/4 3-35 3/4 3-35a 3/4 3-36 3/4 3-36*
3/4 3-63 3/4 3-63*
3/4 3-64 3/4 3-64 3/4 3-65 3/4 3-65 3/4 3-66 3/4 3-66*
3/4 4-9 3/4 4-9*
3/4 4-10 3/4 4-10 3/4 4-11 3/4 4-11 3/4 4-12 3/4 4-12 3/5 5-5 3/4 5-5 3/4 5-6 3/4 5-6*
INSTRUMENTATION 3/4.3.3 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION LIMITING CONDITION FOR OPERATION
~-
3.3.3 The emergency core cooling system (ECCS) actuation instrumentation channels shown in Table 3.3.3-1 shall be OPERABLE with their trip setpoints set consistent with the values shown in the Trip Setpoint column of Table 3.3.3-2 and with EMERGENCY CORE COOLING SYSTEM RESPONSE TIME as shown in Table 3.3.3-3.
APFLICABILITi.
As shown in TaDie 3.3.3-1.
ACTION:
a.
With an ECCS actuation instrumentation channel tric setpoint less conservative than the value shown in the Allowable.alues column of Table 3.3.3-2, declare the channel inoperable until the channel is restored to OPERABLE status with its trip setpoint adjusted consistent with the Trip Setpoint value.
b.
With one or more ECCS actuation instr'umentation channels inoperable, take the ACTION required by Table 3.3.3-1.
c.
With either ADS trip system "A" or "B" inoperable, restore the inoperable trip system to OPERABLE status within:
1.
7 days, provided that the HPCS and RCIC systems are OPERABLE.
2.
72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.
Otherwise, be in at least HOT SHUTDOWN within the.next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and reduce reactor steam dome pressure to less than or equal to 135 psig within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
SURVEILLANCE RE0VIREMENTS 4.3.3.1 Each ECCS actuation instrumentation channel shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL FUNCTIONAL TEST and CHANNEL CAllBRATION operations for the OPERATIONAL CONDITIONS and at the frequenc'es shown in Table 4.3.3.1-1.
4.3.3.2 LOGIC SYSTEM FUNCTIONAL TESTS and simulated automatic operation of all channels shall be performed at least once per 18 months.
4.3.3.3 The ECCS RESPONSE TIME of each ECCS trip functie; shown in Table 3.3.3-3 shall be demonstrated to be within the limit at least once per 18 months.
Each test shall include at least one channel per trip system such that all channels are tested at least once every N times 18 months where N is the total number of_ redundant channels in a specific ECCS trip system.
!i I
GRAND GULF-UNIT 1 3/4 3-27 j
l L
Q TABLE.3.3.3-1
- >5 EME'RGENCY CORE COOLING daTEM ACTilATION INSTPUMENTATION c)
E MINIMUM OPERABLE APPLICABLE CHANNELS PER OPERATIONAL f
[
TRIP FUNCTIL&
TRIP FUNCTION ")
CONDITIONS ACTION z
U A.
DIVISION I TRIP SYSTEM 1.
RHR-A (LPCI MODE) & LPCS SYSTEM a.
Reactor Vessel L ter Level - Low Low low,' Level 1 2(b) 1, 2, 3, 4*, 5*
30 b.
Drywell Pressure - High 2(b) 1,2,3 30 c.
LPCI Pump A Start Time Delay Relay 1
1, 2, 3, 4*, 5*
31 g) d.
Manual Initiation 1/ system 1, 2, 3, 4*, 5*
32 e.
Reactor Vessel Pressure - Ld'w"(Injection Permissive) 3 1,2,3 31 r
4* 5*
35 2.
AUTOMATIC DEPRESSURIZATION SYSTEM TRIP SYSTEM "A"#
a.
Reactor Vessel Water Level - Low Low Low, Level 1 2
1,2,3 30 b.
Drywell Pressure - High 2
1,2,3 30 c.
ADS Initiation Timer 1
1,2,3 31 l d.
Paactor Vessel Water Level - Low, Level 3 (Rermissive) 1 1,2,3 31 e.
LPCS Pump Discharge Pressure-High (Permissive) 2 1,2,3 31 f.
LPCI Pump A Discharge Pressure-High (Permissive) 2 1,2,3 31 g.
Manual Initiation 2/ system 1, 2, 3 32 h.
ADS Bypass Timer (High Drywell Pressure)'
1 1,2,3 32 2
1,2,3 32 i.
Manual Inhibit B.
DIVISION 2 TRIP SYSTEM i
m 1.
IhI ME a.
Reactor Vessel Water Level - Low,. Low Low, Level 1 2
1, 2, 3, 4*, 5*
30 0'I b.
Drywell Pressure'- High y$
2 1,2,3 30 9
c.
LPCI Pump B Start Time Delay Relay i
1 1, 2, 3, 4*, 5*
31 1/ system (b) 1, 2, 3, 4*, 5*
32 m6 d.
Manual Initiation g"
e.
Reactor Vessel Pressure - Low (Injection Per, missive) 3 1,2,3 31
- f I 4* 5*
35 O
I e
TABLE 3.3.3-1 (Continued) c35g EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION h
MINIMUM OPERABLE APPLICABLE m
CHANNELS PER OPERATIONAL TRIP FUNCTION TRIP FUNCTION (a)
CONDITIONS ACTION E
t; B.
DIVISION 2 TRIP SYSTEM (Continued) 2.
AUTOMATIC DEPRESSURIZATION SYSTEM TRIP SYSTEM "B"#
2f a.
Reactor Vessel Water Level - Low Low Low, Level 1 1,2,3 30 b.
Drywell Pressure - High 2
1,2,3 30 i
c.
ADS Initiation Timer 1
1,2,3 31 I
d.
Reactor Vessel Water Level - Low, Level 3 (Permissive) 1 1,2,3 31 e.
LPCI Pump B and C Discharge Pressure - High (Permissive) 2/ pump 1,2,3 31 f.
Manual Initiation 2/ system 1, 2, 3 '
32 g.
ADS Bypass Timer (High Drywell Pressure) 2 1,2,3 32 R>
h.
Manual Inhibit 1
1,2,3 32
,a Es f
q 3l" 2.
O n.
TABLE 3.3.3-1 (Continued)
EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION oC HINIMUM OPERABLE APPLICABLE CllANNELS PER OPERAll0NAL
' TRIP FUNCTION TRIP IUNCil0N(a)
CONUlil0NS ACTION n
r"-?
C.
DIVISION 3 TRIP SYSTEM C5 1.
IlPCS SYSTEM
)
1, 2, 3, 4*, 5*
33 a.
Reactor Vessel Water level - Low, Low, level 2 4
II 1,2,3 33 b.
Drywell Pressure - Highff 4
IC) c.
Reactor Vessel Water Level-High, level B 2
1, 2, 3, 4*, 5*
31 d.
Condensate Storage Tank level-Low 2
1,2,3,4*,5*
34 e.
Suppression Pool Water Level-liigh 2
1,2,3,4*,5' 34 f.
Manual Initiation ##
1 1, 2, 's, 4 *, 5*
32 D.
LOSS OF POWER 1.
Division 1 and 2 a.
4.16 kV Dus Undervoltage 4
1, 2, 3, 4**, 5**
30 (Loss of Voltage)
A b.
4.16 kV Bus Undervoltage 4
1, 2, 3, 4**, 5**
30 (BOP Load Shed) w,',
c.
4.16 kV Bus Undervoltage 4
1, 2, 3, 4**, 5**
30 (Degraded Voltage) 2.
Division 3 a.
4.16 kV Bus Undervoltage 4
1, 2, 3, 4**, 5**
30 (Loss of V61tage) b.
4.16 kV Bus Undervoltage 4
1, 2, 3, 4**, 5**
30 (Degraded Voltage)
~b A channel may be placed in an inoperable status for up to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> during periods of required surveillance (a) without placirq the trip system in the tripped condition provided at least one other OPERABLE channel in the same trip system is monitoring that parameter.
(b) Also actuates the associated division diesel generator.
(c) Provides signal to close llPCS pump discharge valve only, g
Provides signal to llPCS pump suction valves only.
b (d)
Applicable when the system is required to be OPERABLE per Specification 3.5.2 or 3.5.3.
E&
Required when applicable ESF equipment is required to be OPERABLE.
"1 5 P.
Not required to be OPERABLE when reactor steam dome pressure is less than or equal to 135 psig.
Prior to STARTUP following the first refueling outage, the injection function of Drywell Pressure -
gz High and Manual Initiation are not required to be OPERABLE with indicated reactor vessel water level P"g on the wide range instrument greater than Level 8 setpoint coincident with the reactor pressure less than 600 psig.
INSTRUMENTATION TABLE 3.3.3-1 (Continued)
EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION ACTION ACTION 30 -
With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip Function requirement:
a.
With one channel inoperable, place the inoperable channel in the tripped condition within one hour
- or declare the associated system (s) inoperable.
b.
With more than one channel inoperable, ce: are the associated system (s) inoperable.
ACTION 31 -
With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per ' rip Function requirement, declare the associated ADS trip system or ECCS inoperable.
ACTION 32 -
With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip Function requirement, restore the inoperable channel to OPERABLE status within 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> or. declare the associated ADS trjp system er ECCS inoperable.
ACTION 33 -
With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip Function requirement, place the inoperable channel (s) in the tripped condition within one hour
- or declare the HPCS system inoperable.
ACTION 34 -
With the number of OPERABLE channels less than required by the Minimum OPERABLE Channels per Trip Function requirement, place at least one iroperable channel in the tripped condition within one hour
- or declare the HPCS system inocerabic.
ACTION 35 -
With the number of OPERABLE channels les: inan required by the Minimum OPERABLE Channels per Trip Eunction-requirement, place the inoperable channel (s)'in the tripped' condition within one -
hour
- or declare the associated system {s).. inoperable. ;-
"The provisions of Specification 3.0.4 are no applicable.
GRAND GULF-UNIT 1 3/4 3-30 m endment No. 20 l E::ective Date:
---,,v-.
TABLE 3.3.3-2 n
5g EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION SETPOINTS o
ALLOWABLE F
TRIP FUNCTION TRIP SETPOINT VALUE A.
DIVISION l' TRIP SYSTEM E
1.
RHR-A (LPCI MODE) AND LPCS SYSTEM U
a.
Reactor Vessel Water Level - Low Low Low, Level 1
> -150.3 inches *
> -152.5 inches b.
Drywell Pressure - High 51.39psig 51.44psig c.
LPCI Pump A Start Time Delay Relay
< 5 seconds
< 5.25 seconds d
Manual Initiation NA NA e.
Reactor Vessel Pressure--Low (Injection Permissive) 516 psig, decreasing 452-534 psig, decreasing l 2.
AUTOMATIC DEPRESSURIZATION SYSTEM TRIP SYSTEM "A" a.
Reactor Vessel Water Level - Low Low Low, level 1
> -150.3 inches *
> -152.5 inches b.
Drywell Pressure - High 7 1.39 psig i 1.44 psig w) c.
ADS Initiation Timer 7 105 seconds 7 117 seconds l
d.
Reactor Vessel Water Level-Low, Level 3 I 11.4 inches
- I 10.8 inches w
J, e.
LPCS Pump Discharge Pressure-High 145 psig, increasing I25-165 psig, increasing f.
LPCI Pump A Discharge Pressure-High 125 psig, increasing 115-135 psig, increasing g.
Manual Initiation NA NA h.
ADS Bypass Timer (High Drywell Pressure)
< 9.2 minutes
< 9.4 minutes i
Manual Inhibit NA NA B.
DIVISION 2 TRIP SYSTEM 1.
Reactor Vessel Water Level - Low Low Low, level 1
> -150.3 inches *
> -152.5 inches b.
Drywell Pressure - High 7 1.39 psig 7 1.44 psig c.
LPCI Pump B Start Time Delay Relay 7 5 seconds 7 5.25 seconds Q
d.
Manual Initiation NA NA e.
Reactor Vessel Pressure--Low (Injection Permissive) 516 psig, decreasing 452-534 psig,' decreasing l NE 7g 2.
AUTOMATIC DEPRESSURIZATION SYSTEM TRIP SYSTEM "B" m+c Eig a.
Reactor Vessel Water Level - Low Low Low, Level 1
> -150.3 inches *
> -I52.5 inches
~'
b.
Drywell Pressure - High 51.39psig 51.44psig c.
ADS Initiation Timer i 105 seconds 5 117 seconds l
i
~
\\
TABLE 3.3.3-2 (Continued)
EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION SETPOINTS ALLOWABLE c)
. F, TRIP FUNCTION TRIP SETPOINT VALUE l
B.
DIVISION 3' TRIP SYSTEM (Continued)
E Z
2.
AUTOMATIC DEPRESSURIZATION SYSTEM TRIP SYSTEM "B" (continued) i w
d.
Reactor Vessel Water Level-Low, level 3
> 11.4 inches *
> 10.8 inches e.
LPCI Pump B and C Discharge Pressure-High 125 psig, increasing 115-135 psig, increasing f.
Phnual Initiation NA NA g.
ADS Bypass Timer (Iligh Drywell Pressure) 5 9.2 minutes 5 9.4 minutes h.
Manual Inhibit NA NA
. C.
DIVISION 3 TRIP SYSTEM w) 1.
HPCS SYSTEM wa a.
Reactor Vessel Water Level - Low Low, level 2
>-41.6 inches *
>-43.8 inches g
b.
Drywell Pressure ' High 5 1.39 psig 5 1.44 psig c.
Reactor Vessel Water Level - High, Level 8 5 53.5 inches
- 5 55.7 inches d.
Condensate Storage Tank Level - Low
> 0 inches
> -3 inches e.
Suppression Pool Water Level - High 7 5.9 inches 7 7.0 inches h
f.
Manual Initiation NA NA
?
t 1
k
=%*r l
Ekg-
\\
.. g i
1
~
c
l TABLE 3.3.3-2 (Continued)
EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION SETPOINTS i
El ALLOWABLE l
5 TRIP FUNCTION TRIP SETPOINT VALUE i
C i
25 D.
LOSS OF POWER i
1.
Division 1 and 2 I
a.
4.16 kV Bus Undervoltage 1.
4.16 kV Basis 2912 +0, -291 volts (Loss of Voltage) 2912 volts 2.
120 volt Basis 83.2 +0, -8.3 volts 83.2 volts 3.
Time Delay 0.5 +0.5, -0.1 seconds 0.5 seconds b.
4.16 kV Bus Undervoltage.
1.
4.16 ky Basis 3328 +0, -167 volts (BOP Load Shed) 3328 volts 2.
120 volt Basis 95.1 +0, -4.8 volts w
95.1 volts 30 3.
Time delay 0.5 +0.5, -0.1 seconds 0.5 seconds 3,
N' c.
4.16 kV Bus Undervoltage 1.
4.16 kV Basis 374'4 +93.6, -0 volts (Degraded Voltage) 3744 volts 2.
120 volt Basis 107 +2.7, -0 volts 107 volts 3.
Time Delay 9.0 1 0.5 seconds 9.0 seconds 2.
Division 3 i
a.
4.16 kV Bus Undervoltage 1.
4.16 kV Basis 3045 i 61 volts i
(Loss of Voltage) 3045 volts 2.
120 volt Basis 87 1 1.7 volts 87 volts 3.
Time Delay 2.3 + 0.2, -0.3' seconds 2.3 seconds 4
1
- See Bases Figure B 3/4 3-1.
s h
7 TABLE 3.3.3'-2 (Continued) i EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION SETPOINTS c3 o
i 55 ALLOWABLE l'
TRIP FUNCTION TRIP SETPOINT VALUE lE t1 D.
~ LOSS OF POWER (Continued) 2.
Division 3 (Continued) p.
t 1.
4.16 kV Basis 3661 1 102.5 volts b.
4.16 kV Bus Undervoltage (Degraded Voltage) 3661 volts 2.
120 volt Basis
.104.6 1 2.93 volts 104.6 volts 3.
Tim ( Delay 5 minutes /No LOCA
'S minutes i 30 seconds 4 seconds /LOCA (4.0 1 0.4 seconds)
~
Y
.k:
9 Sk
- ?.
El ??
=8 co j
e
TABLE 3.3.3-3 EMERGENCY CORE COOLING SYSTEM RESPONSE TIMES (SECONDS) 1.
LOW PRESSURE CORE SPRAY SYSTEM NA l
2.
LOW PRESSURE COOLANT INJECTION MODE OF RHR SYSTEM PUMPS A, 8 AND C NA l
3.
AUTOMATIC DEPRESSURIZATION SYSTEM NA 4.
HIGH PRESSURE CORE SPRAY SYSTEM
$ 27 5.
LCSS OF POWER NA GRAND GULF-UNIT 1 3/4 3-33 Amendment ho. 20 l
Effective Date:
~
TABLE 4.3.3.1-1 EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS c)
CHANNEL OPERATIONAL E
CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH
]
TRIP FUNCTION CHECK TEST CALIBRATION SURVEILLANCE REQUIRED g
A.
DIVISION I TRIP SYSTEM Q
1.
RHR-A (LPCI MODE) AND LPCS SYSTEM a.
l Low l f.ow, Level 1 S
M R(a) 1, 2, 3, 4*, 5*
b.
Drywell Pressure - High S
M R(a) 1, 2, 3 c.
LPCI Pump A Start Time l
Delay Relay NA M(b)
Q 1, 2, 3, 4*, 5*
d.
Manual Initiation NA R
Q 1, 2, 3, 4*, 5*
e.
Reactor Vessel Pressure -
j Low (Injection Permissive) S M
R(a) 1, 2, 3, 4*, 5*
R 2.
AUTOMATIC DEPRESSURIZATION SYSTEM TRIP SYSTEM "A"#
J, a.
Low Low Low, Level 1 S
M R
1,2,3 b.
Drywell Pressure-High S
M R
1, 2, 3 c.
ADS Initiation Timer NA M
Q 1, 2, 3 l
d.
Low, Level 3 S
M R(a) 1, 2, 3 e.
LPCS Pump Discharge Pressure-High S
M R(,)
1,2,3 f.
LPCI Pump A Discharge Pressure-Hioh S
M(b)
R(,)
1,2,3 g.
Manual Initi " lon NA R
NA 1,2,3 h.
ADS Bypass Timer (High Drywell Pressure)
NA M
Q 1, 2, 3 1.
Manual Inhibit NA R
NA 1, 2, 3 35 UI w
r+*
i
o TABLE 4.3.3.1-1 (Continued) 5 EMERGENCY CORE CdOLING SYSTEM ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS 5
o CHANNEL OPERATIONAL E
CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH 7
TRIP FUNCTION CHECK TEST CALIBRATION SURVEILLANCE REQUIRED g
B.
DIVISION 2 TRIP SYSTEM Z
1.
P{a) 1, 2, 3, 4*, 5*
Low Low Low, Level 1 S
M b.
Drywell Pressure - High S
M R)
1, 2, 3 c.
LPCI Pump B Start Iime Delay Relay NA M
Q 1, 2, 3, 4*, 5*
g) d.
Manual Initiation NA R
Q 1, 2, 3, 4*, 5*
e.
Reactor Vessel Pressure -
Low (Injection Permissival S M
R(a) 1, 2, 3, 4*, 5*
R l
1 Y
.T, l
n$
28 ag C~
o e
r 1
TABLE 4.3.3.1-1 (Continued) c)
[
EMERGENCY CORE COOLING A STEM ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL OPERATIONAL E
CH?NNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH TRIP FUNCTION CHECK TEST _
CALIBRATION SURVEILLANCE REQUIRED B.
DIVISION 2 TRIP SYSTEM (Continued)
[
2.
AUTOMATIC DEPRESSURIZATION SYSTEM TRIP SYSTEM "B"#
a.
Low Low Iow, Level 1 S
M R(a) 1, 2, 3 I) b.
Drywell Pr mure-High S
M R
1, 2, 3 c.
ADS Initiation Timer NA M
Q 1,2,3 l
d.
Low, Level 3 S
M R(a) 1, 2, 3 w
i e.
LPCI Pump B and C Discharge Pressure-High S
M(b)
R(3) 1,2,3 w
de f.
Manual Initiation NA R
NA 1, 2, 3 g.
ADS Bypass Timer (High Drywell Pressure)
NA M
Q 1,2,3 h.
Manual Inhibit NA R
NA 1, 2, 3 C.
DIVISION 3 TRIP SYSTEM 1.
HPCS SYSTEM a.
Reactor Vessel Water ievel -
1((a) 1, 2, 3, 4*, 5*
Low Low, Level 2 S
M R(a) b.
Drywell Pressure-High##
S M
1, 2, 3 a) c.
Reactor Vessel Water S
M 1, 2, 3, 4*, 5*
R Level-High, Level 8 QF d.
Condensate Storage Tank 23 Level - Low S
M R(a) 1, 2, 3, 4*, 5*
3@
e.
Suppression Pool Water
- TS Level - High S
M(b)
R(a) 1, 2, 3, 4*, 5*
[
f.
Manual Initiation ##
NA R
NA 1, 2, 3, 4*, 5*
3
" ' '.s
.i o
TABLE 4.3.3.1-1 (Continued)
[
EMERGENCY CORE COOLING SYSTEM ACTUATION' INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL OPERATIONAL E
CHANNEL FUNCTIONAL CHANNEL CONDITIONS FOR WHICH TRIP FUNCTION CHECK TEST C_AI m0ATION.
SURVEILLANCE REQUIRED g
D.
LOSS OF POWER
[
1.
Division 1 and 2 IM ')
R 1, 2, 3, 4**, 5**
4.16kVBusbndervoltage NA a.
(Loss of Voltage) b.
4.16 kV Bus Undervoltage NA M(,)
R 1, 2, 3, 4**, 5**
(B0P Load Shed)
M ')
R 1, 2, 3, 4**, 5**
I c.
4.16 kV Bus Undervoltage NA (Degraded Voltage) g 2.
Division 3 w
a.
4.16 kV Bus'lindervoltage NA NA R
1, 2, 3, 4**, 5**
O (Loss of Vo age)
T b.
4.16 kV Bus Undervoltage NA NA R
1, 2, 3, 4**, 5**
(Degraded Voltage)-
9 f
ga at 2
ni,+
r 5'E N
i I
I t
't I
TABLE 4.3.3.1-1 (Continued)
EMERGENCY CORE COOLING SYSTEM ACTUATION INSTRUMENTATION SURVEILLANCE REQUIREMENTS NOTATION Not required to be OPERABLE when reactor steam dome pressure is less than or equal to 135 psig.
Prior to STARTUP following the first refueling outage, the injection func-tion of Drywell Pressure - High and Manual Initiation are not required to be OPERABLE with indicated reactor vessel water level on the wide range instrument greater than Level 8 setpoint coincident with the reactor pres-sure less than 600 psig.
Applicable when the system is required to be OPERABLE per Specification 3.5.2 or 3.5.3.
Required when ESF equipment is required to be OPERABLE.
(a) Calibrate trip unit at least once per 31 days.
(b) Manual initiation switches shall be tested at least once per 18 months during shutdown.
All other circuitry associated with manual initiation shall receive a CHANNEL FUNCTIONAL TEST at least once per 31 days as a
~
part of circuitry required to be tested for automatic system actuation.
(c) DELETED (c) DELETED (e) Functional Testing of Time Delay Not Required 4
J GRAND GULF-UNIT 1 3/4 3-36
- - - _ -. _ _ - = _. _ _.
. -..... ~. - -. -
INSTRUMENTATION SEISMIC MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.7.2 The seismic monitoring instrumentation shown in Table 3.3.7.2-1 shall be OPERABLE.
APPLICABILITY:
At all times.
ACTION:
With one or more of the above required seismic monitoring instruments a.
inoperable for more than 30 days, prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within the next 10 days outlining the cause of the malfunction and the plans for restoring the instrument (s) to OPERABLE status.
b.
The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
SURVEILLANCE REQUIREMENTS 4.3.7.2.1 Each of the above required seismic monitoring instruments shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK, CHANNEL FUNC-TIONAL TEST and CHANNEL CALIBRATION operations at the frequencies shown in t
Table 4.3.7.2-1.
4.3.7.2.2 Each of the above required seismic monitoring instruments actuated during a seismic event greater than or equal tot 0.01 g shall be restored to OPERABLE status within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and a CHANNEL CALIBRATION performed within 5 days following the seismic event.
Data shall be retrieved from actuated instruments and analyzed to determine the magnitude of the vibratory ground motion.
A Special Report shall be prepared and submitted to the Commission pursuant to Specification 6.9.2 within 10 days describing the magnitude, freauency spectrum and resultant effect upon unit features important to safety.
GRAND GULF-UNIT 1 3/4 3-63
TABLE 3.3.7.2-1 SEISMIC MONITORING INSTRUMENTATION MINIMUM MEASUREMENT INSTRUMENTS INSTRUMENTS AND SENSOR LOCATIONS RANGE OPERABLE 1.
Triaxial Strong Motion Accelerometer a..
Containment foundation 0.001 to 1.0g 1
b.
Drywell O.001 to 1.0g 1
c.
SGTS Filter Train 0.001 to 1.0g 1
d.
SSW Pume House A 0.001 to 1.0g i
e.
Free Field 0.001 to 1.0g 1
f.
Reactor Piping Support 0.001 to 1.0g 1
l 2.
Triaxial Peak Recording Accelerograph a.
Containment Dome 0.01 to 2.0g 1
b.
Auxiliary Building Foundation 0.01 to 2.0g 1
c.
Diesel Generator 11 0.01 to 2.0g 1
d.
Control Building Foundation 0.01 to 2.0g 1
e.
Control Room 0.01 to 2.0g 1
f.
Reac: - Vessel Support 0.01 to 2.0g 1
g.
Deleted h.
Deleted i.
Deleted j.
Deleted k.
SSW Pump House B 0.01 to 2.0g 1
~
3.
Triaxial Seismic Switches a.
Containment Foundation (SSE) 0.025 to 0.25g 1*
b.
Containment Foundation (OBE) 0.025 to 0.25g 1*
d.
Drywell (OBE) 0.025 to 0.25g 1*
c.
Drywell (SSE) 0.025 to 0.25g 1*
4.
Vertical Seismic Trigger c.
Containment Foundation 0.005 to 0.05g 1*
5.
Horizontal Seismic Trigger a.
Drywell 0.005 to 0.05g 1*
"With control room annunciation.
GRAND GULF-UNIT 1 3/4 3-64 Amendment No. 20l Effective Date:
._._m.,
. m.. _. _..
1 TABLE 4.3.7.2-1 SEISMIC MONITORING INSTRUMENTATION SURVEILLANCE REQUIREMENTS CHANNEL
~
CHANNEL FUNCTIONAL CHANNEL INSTRUMENTS AND SENSOR LOCATIONS CHECK TEST CALIBRATION 1.
Triaxial Strong Motion Accelerometer a.
Containment Foundation M
SA R
b.
Drywell M
SA R
c.
SGTS Filter Train M
SA R
d.
SSW Pump House A M
SA R
e.
Free Field M
SA R
f.
Reactor Piping Support M
SA R
l 2.
Triaxial Peak Recording Accelerograph a.
Containment Dome NA NA R
b.
Auxiliary Building Foundation NA NA R
c.
Diesel Generator 11 NA NA R
d.
Control Building Foundation NA NA R
e.
Control Room NA NA R
f.
Reactor Vessel Support NA NA R
g.
Deleted h.
Deleted i.
Deleted j.
Deleted k.
SSW Pump House B NA NA R
3.
Triaxial Seismic Switches a.
Containment Foundation (SSE)
M SA R
b.
Containment Foundation (OBE)
M SA R
c.
Drywell (SSE)
M SA R
d.
Dcywell (OBE)
M SA R
4.
Vertical Seismic Trigger a.
Containment Foundation
- M SA R
5.
Horizonte1 Seismic Trigger a.
Drywell M
SA R
GRAND GULF-UNIT 1 3/4 3-65 Amendment No. 20 l
Effective Date:
INSTRUMENTATION METEOROLOGICAL MONITORING INSTRUMENTATION LIMITING CONDITION FOR OPERATION 3.3.7.3 The meteorological monitoring instrumentation channels shown in Table 3.3.7.3-1 shall be OPERABLE.
APPLICABILITY:
At all times.
ACTION:
a.
With one or more required meteorological monitoring instrumentation channels inoperable for more than 7 days, prepare and submit a Special Report to the Commission pursuant to Specification 6.9.2 within the next 10 days outlining the cause of the malfunction and the plans for restoring the instrumentation to OPERABLE status.
b.
The provisions of Specifications 3.0.3 and 3.0.4 are not applicable.
l
=
~
SURVEILLANCE REQUIREMENTS 4.3.7.3 Each of the above required meteorological monitoring instrumentation channels shall be demonstrated OPERABLE by the performance of the CHANNEL CHECK and CHANT'EL CALIBRATION operations at the frequencies shown in Table 4.3.7.3-1.
GRAND GULF-UNIT 1 3,4 3-66
REACTOR COOLANT SYSTEM OPERATIONAL LEAKAGE LIMITING CONDITION FOR OPERATION 3.4.3.2 Reactor coolant system leakage shall be limited to:
No PRESSURE B0UNDARY LEAKAGE.
a.
b.
5 gpm UNIDENTIFIED LEAKAGE.
c.
30 gpm total leakage.
1050 1 10 psig I gpm leakage at a reactor coolant system pressure of d.
from any reactor coolant system pressure isolation valve specified in Table 3.4.3.2-1.
2 gpm increase in UNIDENTIFIED LEAKAGE within any 4-hour pdriod.
e.
APPLICABILITY: OPERATIONAL CONDITION 5 1, 2 and 3.
ACTION:
With any PRESSURE BOUNDARY LEAKAGE, be in at least HOT SHUTDOWN within a.
12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in 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 />.
With any reactor coolant system leakage greater than the limits in b b.
and/or c, above, reduce the leakage rate to within the limits within i
4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
With any reactor coolant system pressure isolation valve leakage greater than tne above limit, isolata the high pressure portion of c.
the affected system from the luw pressure portion within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> by use of at least two closec manual or deactivated automatic valves, or be in at least HOT SHUTDOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
With one or more high/ low pressure interface valve pressure moni-d.
tors and/or interlocks inoperable, restore the inoperable monitor (s) and/or interlock (s) to OPERABLE ' status within 30 days or be in at least HOT SHUTDOWN within the next-12 hours and in COLD SHUTDOWN Qit foliewing 24 nours.
With any reactor coolant system UNIDENTIFIED LEAKAGE increase greater f
than 2 gpm within any 4-hour period, identify the source of leakage e.
increase as not service sensitive Type 304 or 316 austenitic stainless steel within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or be in at least HOT SHUTOOWN within the next 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and in COLD SHUTDOWN within the following 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
GRAND GULF-UNIT 1 3/4 4-9
REACTOR COOLANT SYSTEM SURVEILLANCE REQUIREMENTS 4.4.3.2.1 The reactor coolant system leakage shall be demonstrated to be within each of the above limits by:
~
a.
Monitoring the drywell atmospheric particulate and gaseous radioactivity at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, b.
Monitoring the drywell floor and equipment drain sump level and flow rate at least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, c.
Monitoring the drywell air coolers condensate flow rate at'least once per 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />, and d.
Monitoring :ne reactor vessel head flange leak detection system at least once per 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.
4.4.3.2.2 Each reactor coolant system pressure isolation valve specified in Table 3.4.3.2-1 shall be demonstrated OPERABLE by leak testing pursuant to Specification 4.0.5 and verifying the leakage of each valve to be within the specified limit:
a.
At least_once per 18 mbnths, and b.
- Prior to return'ing the'-valve to service following maintenance',
repair or replacement work on, the valve which could affect its
' lea. age rate.
The provisions of Specification 4.0.4 are not applicable for entry into OPERATIONAL CONDITION 3.
4.4.3.2.3 The high/ low pressure interface valves leakage pressure monitors shall be demonstrated OPERABLE with alarm and interlock setpoints per Table 3.4.3.2-2 and Table 3.4.3.2-3 by performance of a:
a.
CHANNEL FUNCTIONAL TE*~ at least once per 31 days, and b.
CHANNEL CALIBRATION at least once per 18 months.
GRAND GULF-L" !T 1 3/4 4-10 Amendment No. 2d Effective Date:
\\
TABLE 3.4.3.2-1 REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES VALVE NUMBER SYSTEM E21-F005 LPCS E21-F006 E22-F004 HPCS E22-F005 E12-F008 RHR E12-F009 E12-F023 E12-F041 A, B, C E12-F042 A, B, C E12-F050 A, B E12-F053 A, B E12-F308 E12-F394 E51-F063 RCIC E51-F064 E51-F065 E52-F076 E51-F013 TABLE 3.4.3.2-2 REACTOR COOLANT SYSTEM INTERFACE VALVES PRESSURE MONITORS - ALARM ALARM SETPOINT VALVE NUMBER
- SYSTEM, (psig)
E21-F005 to E21-F006 LPCS
$575 l
E12-F008 to E12-F006A RHR 5183 E12-F008 to E12-F006B RHR 1183 E12-F041A to E12-F042A RHR 1475 l
'E12-F041B to E12-F042B RHR 1475 l
E12-F041C to E12-F042C RHR 1475 l
GRAND GULF-UNIT 1 3/4 4-11 Amendment No. 20 l Effective Date:
TABLE 3.4.3.2-3 REACTOR COOLANT SYSTEM INTERFACE VALVES
~
PRESSURE INTERLOCK 5 INTERLOCK SETPOINT VALVE NUMBER SYSTEM (psic)
E12-F052 to E51-F064 RCIC
< 465 E12-F041A to E12-F042A RHR 1 475 l
E12-F0418 to E12-F0428 RHR 1 475 l
E12-F041C to E12-F042C RHR 1 475 l
E21-F005 to'E21-F006 LPCS 1 575 l
t i
GRAND GULF-UNIT 1 3/4 4-12 AmendmentNo.20(
Effective Date:
IMERGENCY CORE COOLING SYSTEMS SURVEILLANCE REQUIREMENTS (Continued) 2)
Low pressure setpoint of the:
(a) LPCI A and 8 subsystem loop to be 1 38 psig.
(b) LPCI C subsystem loop and LPCS system to be 1 22 psig.
(c) HPCS system to be 1 18 psig.
b)
Header delta P instrumentation and verifying the setpoint of the HPCS system and LPCS system and LPCI subsystems to be 1.2 0.1 psid change from the normal indicated AP.
3.
Verifying that the suction for the HPCS system is automatically transferred from the condensate storage tani to the suppression pool on a condensate storage tank low water level signal and on a suppression pool high water level signal.
4.
Verifying that the time required for each LPCI and LPCS injec-tion valve to travel from fully closed to fully open is 5, 29 seconds when tested pursuant to Specification 4.0.5.
d.
For the ADS at least once per 18 months by:
1.
Performing a system functional test which includes simulated aut: 'atic actuation of the system throughout its emergency operating sequence, but excluding actual valve actuation.
2.
Manually opening each ADS valve when the reactor steam dome pressure is greater than or equal to 100 psig* and observing that either:
a)
The control valve or bypass valve position responds accordir:1y, or b)
There is a corresponding change in the measured steem flow.
"The provisions of Specification 4.0.4 are not applicable provided the surveillance is performed within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after reactor steam pressure is adequate 1: perform the test.
GRAND GULF-UNIT 1 3/4 5-5 Amendment No. 20 l
Effective Date:
V EMERGENCY CORE COOLING SYSTEMS 3/4 5.2 ECCS - SHUTDOWN LIMITING CONDITION FOR OPERATION 3.5.2 At least two of the following shall be OPERABLE:
The low pressure core spray (LPCS) system with a flow path capable a.
of taking suction from the suppression pool and transferring the water through the spray sparger to the reactor vessel.
b.
Low pressure coolant injection (LPCI) subsystem "A" of the RHR system '
with a flow path capceie of taking suction from the suppression pool upon being manually re. aligned and transferring the water to the reactor vessel.
Low pressure coolant injection (LPCI) subsystem "B" of the RHR system c.
with a flow path capable of taking suction from the suppression pool upon being manually realigned and transferring the water to the
~
reactor vessel.
d.
Low pressure coolant injection (LPCI) subsystem "C" of the RHR system with a flow path capable of taking su'ction from the suppression pool upon being manually realigned and transferring the water to the reactor vessel.
The ogh pressure core 's, pray '(HPCS) system with a fl~ow path capable e.
~
of taxing s'uction from 06e of the following water sources and trans-
~
~
ferring the water through the spray sparger to the reactor vessel:
1.
From'the suppression pool, or
~
2.
When the suppression pool level is less than the limit or is drained, from the condenrate storage tank containing at least i
170,000 available gallons of water, equivalent to a level of 18 feet.
APPLICABILITY: OPERATIONAL CONDITION 4 and 5*.
ACTION:
With one of the above requirec subsystems / systems inoperable, restore
~
a.
at least two subsystems / systems to OPERABLE status within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or suspend all.aoerations that have a potential for draining the reactor vessel.
~
b.
With both of the above required subsystems / systems inoperable, scspend CORE ALTERATIONS and all operations that have a potential for draining the reactor vessel.
Restore at least one subsystem / system to OPERABLE status within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> or establish SECONDARY CONTAINMENT INTEGRITY within the next 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
n
~
The ECCS is not required to be OPERABLE provided that the reactor vessel head is removed, the cavity is flooded, the upper containment fuel pool gates are removed, the spent fuel pool gates are removed, and water level is maintained 4
within the ihmits of Specifications 3.9.8 and 3.9.9.
p GRAND GULF-UNIT 1 3/4 5-6
-