ML20082V379
| ML20082V379 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 05/04/1995 |
| From: | GEORGIA POWER CO. |
| To: | |
| Shared Package | |
| ML20082V378 | List: |
| References | |
| NUDOCS 9505090223 | |
| Download: ML20082V379 (36) | |
Text
,
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Edwin I. Hatch Nuclear Plant Thermal Stratification Iechnical Specifications Revision Request in ITS Format Paaa CharqJaftructions j
.Illst l 1
Insert Reolace l
3.3-30 3.3-30 3.3.32 3.3.32 I
i Jnsert Reolace 3.3-31 3.3-31 3.3-33 3.3-33 r
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9505090223 950504 PDR ADOCK 05000321 P
PDR IIL-4816 El-1
. ~.
l ATWS-RPT~ Instrumentation:
o-3.3.4.2 j
3.3 INSTRUMENTATION 3.3.4.2 Anticipated Transient Without Scram Recirculation Pump Trip (ATWS-RPT) Instrumentation i
LC0 3.3.4.2 Two channels per trip system for each ATWS-RPT instrumentation function listed below shall be OPERABLE:
a.
Reactor Vessel Water Level -- ATWS-RPT Level; and l-b.
Reactor Steam Dome Pressure -- High.
APPLICABILITY:
MODE 1.
f ACTIONS
NOTE--------------------------------------
Separate Condition entry is allowed for each channel.
CONDITION REQUIRED ACTION COMPLETION TIME' A.
One or more channels A.1 Restore channel to 14 days inoperable.
OPERABLE status.
QB A.2
NOTE---------
l Not applicable if inoperable channel is the result of an inoperable breaker.
j Place channel in 14 days trip.
i i
(continued)-
i HATCH UNIT 1 3.3-30 L:\\wpihateh\\its\\ unit l\\spees\\proposedi3-3-30495-195
'ATWS-RPT Instrumentation 1.--
3.3.4.2 i
SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE FREQUENCY SR 3.3.4.2.2 Perform CHANNEL FUNCTIONAL TEST.
92 days SR SJ3.4.2.3 Perform CHANNEL CALIBRATION. The 18 months Allowable' Values shall be:
a.
ATWS-RPT Level: 2: -73 inches; and b.
Reactor Steam Dome Pressure -- High:
s 1095 psig, v
-Perform LOGIC SYSTEM FUNCTIONAL TEST 18 months
. including breaker actuation.
r i
i i
2 HATCH UNIl 1 3.3-32 L:s.vwithsii.suniiis pec.sp 13m edt3-3-32495-195 i
ATWS-RPT Instrumentation 3.3.4.2 3.3 INSTRUMENTATION 3.3.4.2 Anticipated Transient Without Scram Recirculation Pump Trip l
(ATWS-RPT) Instrumentation-LCO. 3.3.4.2 Two channels per trip system for each ATWS-RPT instrumentation Function listed below shall be OPERABLE:
a.
Reactor Vessel Water Level -- ATWS-RPT Level and' l.
b.
Reactor Steam Dome ' Pressure -- High.
APPLICABILITY:
MODE 1.
t ACTIONS
NOTE-------------------------------------
i Separate Condition entry is allowed for each channel.
r CONDITION REQUIRED ACTION COMPLETION TIME A.
One or more channels A.1 Restore channel ~to 14 days inoperable.
OPERABLE status.
08 l
A.2
NOTE---------
r L
Not applicable if inoperable channel is the result of an i
inoperable breaker.
l Place channel in 14 days trip.
9 (continued)
I f
i i
HATCH UNIT 2 3.3-31 L:\\wr\\hateh\\its\\ unit 2\\ spec s\\ proposed \\3.3-31495-13 5
RPS Instrumentation 3.3.1.1 SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE FREQUENCY SR 3.3.4.2.2 Perform CHANNEL FUNCTIONAL TEST.
92 days SR 3.3.4.2.3 Perform CHANNEL CALIBRATION. The 18 months Allowable Values shall be:
a.
ATWS-RPT Level: 2: -73 inches; and b.
Reactor Steam Dome Pressure -- High:
s; 1095 psig.
SR 3.3.4.2.4 Perform LOGIC SYSTEM FUNCTIONAL TEST 18 months including breaker actuation.
i 1
1 l
i HATCH UNIT 2 3.3-33 k :hphtc h\\iin\\ unit 2\\npce >\\ proposed \\3-3-33495 - 135
ATWS-RPT Instrumentat,n 3.3.4.2
+
3.3 INSTRUMENTATION i
3.3.4.2 Anticipated Transient Without Scram Recirculation Pump Trip (ATWS-RPT) Instrumentation
)
1 l
LC0 3.3.4.2 Two channels per trip system for each ATWS-RPT instrumentation Function listed below sh 1 be OPERABLE:
a.
Jhkg) ev if, d
b.
Reactor Steam Dome Pressure -- High.
- ATWs-RPT APPLICABILITY:
MODE 1.
ACTIONS
NOTE-------------------------------------
Separate Condition entry is allowed for each channel.
CONDITION REQUIRED ACTION COMPLETION TIME A.
One or more channels A.1 Restore channel to 14 days inoperable.
OPERABLE' status.
QB A.2
NOTE---------
Not applicable if inoperable channel is the result of an inoperable breaker.
i Place channel in 14 days i
trip.
(continued)-
HATCH UNIT 1 3.3-30 Amendment No. 195 t
f
ATWS-RPT' Instrumentation 3.3.4.2 SURVEILLANCE REQUIREMENTS (continued)
SURVEILLANCE FREQUENCY SR 3.3.4.2.2 Perform CHANNEL FUNCTIONAL TEST.
92 days SR 3.3.4.2.3 Perform CHANNEL CALIBRATION. The 18 months Allowable Values shall be:
+
a.
es ter Level -
4dfff )
gg RPT tevReac # # in es; and
~b.
Reactor Stea me Pressure - High:
s 1095 psig.
SR 3.3.4.2.4 Perform LOGIC SYSTEM FUNCTIONAL TEST 18 months including breaker actuation.
Il HATCH UNIT 1 3.3-32 Amendment No. 195
[
ATWS-RPT Instrumentation l
3.3.4.2 i
3.3 INSTRUMENTATION 3.3.4.2 Anticipated Transient Without Scram Recirculation Pump Trip I
(ATWS-RPT) Instrumentation LCO 3.3.4.2 Two channels per trip system for each ATWS-RPT l
instrumentation Function listed below shall be OPERABLE:
a.
Reactor Vessel Water Level Lev g;
nd b.
Reactor Steam Dome Pressure - High.
.- A%5-WT APPLICABILITY:
MODE 1.
ACTIONS
~-----------------------------NOTE-------------------------------------
l Separate Condition entry is allowed for each channel.
CONDITION REQUIRED ACTION COMPLETION TIME A.
One or more channels A.1 Restore channel to 14 days inoperable.
OPERABLE status.
93 A.2
NOTE---------
Not applicable if inoperable channel is the result of an inoperable breaker.
Place channel in 14 days trip.
(continued)
HATCH UNIT 2 3.3-31 Amendment No. 135
ATWS-RPT Instrumentation 3.3.4.2 SURVEILLANCE REQUIREMENTS sontinued)
SURVEILLANCE FREQUENCY SR 3.3.4.2.2 Perform CHANNEL FUNCTIONAL TEST.
92 days SR 3.3.4.2.3 Perform CHANNEL CALIBRATION. The 18 months Allowable Values shall be:
a.
Reac" e
ater Level -
@ )
AiTWS-MLev '(:
4 hes; and
(
~lb b.
Reactor Stea ome Pressure - High:
- s; 1095 psig.
i SR 3.3.4.2.4 Perform LOGIC SYSTEM FUNCTIONAL TEST 18 months l
including breaker actuation.
i I
i i
i i
i
+
i HATCH UNIT 2 3.3-33 Amendment No. 135 i
Eachre 2 Edwin I. Hatch Nuclear Plant Thermal Stratification Technical SHAe=**- Revision Reauest in ITS Format Bases Changes I
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i i
e C
i i
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8 G
l I
IIL-4816 E2-1 f
6 8
ATWS-RPT Instrumentation B 3.3.4.2 8 3.3 INSTRUMENTATION B 3.3.4.2 Anticipated Transient Without Scram Recirculation Pump Trip n
(ATWS-RPT) Instrumentation BASES BACKGROUND The ATWS-RPT System initiates an RPT, adding negative reactivity, following events in which a scram does not (but should)' occur, to lessen the effects of an ATWS event.
Tripping the recirculation pumps adds negative reactivity from the increase in steam voiding in the core area as core flow decreases. When Reactor Vessel Water Level - ATWS-RPT l
Level or Reactor Steam Dome Pressure - High setpoint is reached, the recirculation pump drive motor breakers trip.
The ATWS-RPT System (Ref.1) includes sensors, relays, bypass capability, circuit breakers, and switches that are necessary to cause initiation of an RPT.
The channels include electronic equipment (e.g., trip units) that compares measured input signals with pre-established setpoints. When the setpoint is exceeded, the channel output relay actuates, which then outputs an ATWS-RPT signal to the trip logic.
The ATWS-RPT consists of two independent trip systems, with two channels of Reactor Steam Dome Pressure - High and two channels of Reactor Vessel Water Level - ATWS-RPT Level-in '
l each trip system.
Each ATWS-RPT trip system is a two-out-of-two logic for each Function.
Thus, either'two Reactor Water Level - ATWS-RPT Level or two Reactor l
Pressure - High signals are needed to trip a trip system.
The outputs of the channels in a trip system are combined in a logic so that either trip system will trip both recirculation pumps (by tripping the respective drive motor i
breakers).
I l
There is one drive motor breaker provided for each of the two recirculation pumps for a total of two breakers. The i
output of each trip system is provided to both recirculation l
pump breakers.
t I
I (continued)
HATCH UNIT 1 B 3.3-89 Uwp%akh\\its\\unklwanes\\ proposed \\3 3-89495-0
'l
i ATWS-RPT Instrumentation B 3.3.4.2 i
BASES I
APPLICABLE The individual functions are required to be OPERABLE in SAFETY ANALYSES, MODE 1 to protect against common mode failures of the i
LCO, and Reactor Protection System by providing a diverse trip to l
APPLICABILITY mitigate the consequences of a postulated ATWS event.
The l
(continued)
Reactor Steam Dome Pressure - High and Reactor Vessel Water Level - ATWS-RPT Level functions are required to be OPERABLE l
in MODE 1, since the reactor is producing significant power j
and the recirculation system could be at high flow. During i
this MODE, the potential exists for pressure increases or
[
low water level, assuming an ATWS event.
In MODE 2, the l
reactor is at low power and the recirculation system is at low flow; thus, the potential is low for a pressure increase or low water level, assuming an ATWS event.
Therefore, the ATWS-RPT is not necessary.
In MODES 3 and 4, the reactor is shut down with all control rods inserted; thes, an ATWS event is not significant and the possibility of a significant pressure increase or low water level is negligible.
In MODE 5, the one rod out interlock ensures that the reactor remains subcritical; thus, an ATWS event is not significant.
In addition, the reactor pressure vessel (RPV) head is not fully tensioned an( no pressure transient threat to the reactor coolant pre, we boundary (RCPB) exists.
l The specific Applicable Safety Analyses and LC0 discussions i
are listed below on a function by function basis.
l a.
Reactor Vessel Water level - ATWS-RPT Level l
Low RPV water level indicates the capability to cool the fuel may be threatened.
Should RPV water level l
decrease too far, fuel damage could result.
l Therefore, the ATWS-RPT System is initiated at a low i
level to aid in maintaining level above the top of the l
active fuel.
The reduction of core flow reduces the neutron flux and THERMAL POWER and, therefore, the rate of coolant boiloff.
l l
Reactor vessel water level signals are initiated from four level transmitters that sense the difference between the pressure due to-a constant column of water
)
(reference leg) and the pressure due to the actual
~
water level (variable leg) in the vessel.
l 2
(continued)
~
4 HATCH UNIT 1 8 3.3-91 L h p\\hatc h\\its\\ unit l \\ ham \\pnyumeJ\\3-3-914 95-0
ATWS-RPT Instrumentation
~
B 3.3.4.2 BASES APPLICABLE a.
Reactor Vessel Water Level - ATWS-RPT Level l
SAFETY ANALYSES, (continued)
LCO, and APPLICABILITY Four channels of Reactor Vessel Water Level - ATWS-RPT Level, with two channels in each trip system, are available and required to be OPERABLE to ensure that no single instrument failure can preclude an ATWS-RPT from this function on a valid signal. The Reactor Vessel Water Level - ATWS-RPT Level Allowable Value is l
chosen so that the system will not be initiated after a Level 3 scram with feedwater still available, and for convenience with the reactor core isolation cooling initiation.
b.
Reactor Steam Dome Pressure - Hiah F.xcessively high RPV pressure may rupture the RCPB.
An increase in the RPV pressure during reactor operation compresses the steam voids and results in a positive reactivity insertion. This increases neutron flux and THERMAL POWER, which could potentially result in fuel failure and overpressurization. The Reactor Steam Dome Pressure - High Function initiates an RPT for transients that result in a pressure increase, counteracting the pressure increase by rapidly reducing core power generation.
For the overpressurization event, the RPT aids in the termination of the ATWS event and, along with the safety / relief valves, limits the peak RPV pressure to less than the ASME Section III Code limits.
The Reactor Steam Dome Pressure - High signals are initiated from four pressure transmitters that monitor reactor steam dome pressure.
Four channels of Reactor Steam Dome Pressure - High, with two channels in each trip system, are available and are required to be OPERABLE to ensure that no single instrument failure can preclude an ATWS-RPT from this function on a valid signal.
The Reactor Steam Dome Pressure - High Allowable Value is chosen to provide an adequate margin to the ASME Section III Code limits.
(continued)
L:swe haithsit suniinsb.w.sproco as3 3-92495-0 HATCH UNIT 1 B 3.3-92 s
7
RCS P/T Limits B 3.4.9 BASES SURVEILLANCE SR 3.4.9.1 (continued)
REQUIREMENTS cooldown operations and RCS-inservice leakage and hydrostatic testing.
SR 3.4.9.2 A separate limit is used when the reactor is approaching criticality.
Consequently, the RCS pressure and temperature must be verified within the appropriate limits before withdrawing control rods that will make the reactor critical.
Performing the Surveillance within 15 minutes before control rod withdrawal for the purpose of achieving criticality provides adequate assurance that the limits will not be exceeded between the time of the Surveillance and the time of the control rod withdrawal.
i SR 3.4.9.3 and SR 3.4.9.4 Differential temperatures within the applicable limits ensure that thermal stresses resulting from the startup of an idle recirculation pump will not exceed design allowances.
In addition, compliance with these limits ensures that the assumptions of the analysis for the startup of an idle recirculation loop (Ref. 7) are satisfied.
)
Performing the Surveillance within 15 minutes before i
starting the idle recirculation pump provides adequate assurance that the limits will not be exceeded between the time of the Surveillance and the time of the idle pump start.
If the 145*F temperature differential specified in SR 3.4.9.3 cannot be determined by direct indication, an alternate method may be used as described below:
The bottom head coolant temperature and the RPV coolant can be assumed to be s 145 F if the following can be confirmed:
a.
One or more loop drive flows were > 40 percent of rated flow prior to the RPT,
(
(continued)
HATCH UNIT 1 B 3.4-50 k:\\wp\\ hatch \\its\\unitl\\ bases pmposed\\3-4-50495-0 s
i RCS P/T Lisits B'3.4.9 BASES i
SURVEILLANCE SR 3.4.9.3 and SR - 3.4.9.4 (continued)
REQUIREMENTS b.
High Pressure Coolant Injection (HPCI) and Reactor Core Isolation Cooling (RCIC) Systems have not injected since the RPT, c.
Feedwater temperature has remained > 300*F since the RPT, and d.
The time between the RPT and restart is < 30 minutes.
General Electric test data from BWR plants shows that stratification up to the 145 F differential does not occur any sooner than I hour following the RPT (Refs. 10 and 11).
Adding HPCI and RCIC injection, and feedwater temperature constraints provides assurance that the temperature differential will not be exceeded within 30 minutes of the RPT.
An acceptable means of demonstrating compliance with the temperature differential requirement in SR 3.4.9.4 is to compare the temperatures of the operating recirculation loop and the idle loop.
SR 3.4.9.3 and SR 3.4.9.4 have been modified by a Note that requires the Surveillance to be parformed only in MODES 1, 2, 3, and 4.
In MODE 5, the overall stress on limiting components is lower. Therefore, AT limits are not required.
i (continued)
HATCH UNIT 1 B 3.4-50a L%,wausnah.c.spmpo.auso495-0 l
V RCS P/T Limits B 3.4.9
\\
This page intentionally left blank.
P i
HATCH UNIT i B 3.4-50b k:\\wp\\ hatch \\its\\unitl\\ bases \\proposea\\3-4-so495-0 l
RCS P/T Limits B 3.4.9 BASES REFERENCES 3.
ASTM E 185-82, " Standard Practice for Conducting
-(continued)
Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels," July 1982.
4.
5.
Regulatory Guide 1.99, Revision 2, May 1988.
6.
ASME, Boiler and Pressure Vessel Code,Section XI, Appendix E.
7.
FSAR, Section 14.3.6.2.
8.
George W. Rivenbark (NRC) letter to J. T. Beckham, Jr.
(GPC), Amendment 126 to the Operating License, dated June 20, 1986.
9.
NRC No.93-102, " Final Policy Statement on Technical Specification Improvements," July 23, 1993.
0393, " Recirculation Pump Restart Without
- 10. _GE-NE-F~
Vessel eature Indication for E.I. Hatch Nuclear Plant," o.gril 9,1993.
1 11.
DRF A00-05834/6, " Safety & 10 CFR 50.92 Significant Hazards Consideration Assessment for RPV Stratification Prevention Improvements at Edwin I.
4 Hatch Nuclear Plant Units 1 and 2," April 1994.
J J
b I
l HATCH UNIT 1 B 3.4-52 L:1.psh.ichtit.suniitts..spropo os3 4-52495-0
1 ATWS-RPT Instrumentation B 3.3.4.2-l
)
B 3.3 INSTRUMENTATION B 3.3.4.2 Anticipated Transient Without Scram Recirculation Pump Trip (ATWS-RPT) Instrumentation BASES BACKGROUND The ATWS-RPT System initiates an RPT, adding negative reactivity, following events in which a scram does not (but i
should) occur, to lessen the effects of an ATWS event.
Tripping the recirculation pumps adds negative reactivity from the increase in steam voiding in the core area as core flow decreases. When Reactor Vessel Water Level - ATWS-RPT l
Level or Reactor Steam Dome Pressure - High setpoint is reached, the recirculation pump drive motor breakers trip.
The ATWS-RPT System (Ref.1) includes sensors, relays, I
bypass capability, circuit breakers, and switches that-are necessary to cause initiation of an RPT.
The channels include electronic equipment (e.g., trip units) that compares measured input signals with pre-establisheu l
setpoints. When the setpoint is exceeded, the channel output relay actuates, which-then outputs an ATWS-RPT signal to the trip logic.
The ATWS-RPT consists of two independent trip systems, with two channels of Reactor Steam Dome Pressure - High and two i
channels of Reactor Vessel Water Level - ATWS-RPT Level in l'
each trip system.
Each ATWS-RPT trip system is a two-out-of-two logic for each Function.
Thus, either two i
Reactor Water Level - ATWS-RPT Level or two Reactor l
I Pressure - High signals are needed to trip a trip system.
The outputs of the channels in a trip system are combined in a logic so that either trip system will trip both recirculation pumps (by tripping the respective drive motor i
breakers).
There is one drive motor breaker provided for each of the two recirculation pumps for a total of two breakers. The s
output of each trip system is provided to both recirculation pump breakers.
i (continued)
HATCH UNIT 2 B 3.3-89
(*paaithsii, simii 2sbanesspneo.eds3-3-89495...
ATWS-RPT Instrumentation B 3.3.4.2 BASES i
APPLICABLE The individual Functions are required to be OPERABLE in i
~ SAFETY ANALYSES, MODE 1 to protect against common mode failures of the LCO, and Reactor Protection System by providing a diverse trip to APPLICABILITY mitigate the consequences of a postulated ATWS event. The i
(continued)
Reactor Steam Dome Pressure - High and Reactor Vessel Water Level - ATWS-RPT Level Functions are required to be OPERABLE l in MODE 1, since the reactor is ' producing significant power l
and the recirculation system could be at high flow. During this MODE, the potential exists for pressure increases or low water level, assuming an ATWS event.
In MODE 2, the reactor is at low power and the recirculation system is at low flow; thus, the potential is low for a pressure increase or low water level, assuming an ATWS event. Therefore, the ATWS-RPT is not necessary.
In MODES 3-and 4, the reactor is shut down with all control rods inserted; thus, an ATWS t
event is not significant and the possibility of a
[
significant pressure increase or low water level is negligible.
In MODE 5, the one rod out interlock ensures that the reactor remains subcritical; thus, an ATWS event is t
not significant.
In addition, the reactor pressure vessel (RPV) head is not fully tensioned and no pressure transient threat to the reactor coclant pressure boundary (RCPB) exists.
The specific Applicable Safety Analyses and LC0 discussions are listed below on a function by Function basis.
a.
Reactor Vessel Water Level - ATWS-RPT Level l
Low RPV water level indicates the capability to cool the fuel may be threatened.
Should RPV water level decrease too far, fuel damage could result.
l Therefore, the ATWS-RPT System is initiated at a low level to aid in maintaining level above the top of.the i
active fuel.
The reduction of core flow reduces the neutron flux and THERMAL POWER and, therefore, the rate of coolant boiloff.
t Reactor vessel water level signals are initiated from four level transmitters that sense the difference l
between the pressure due to a constant column of water (reference leg) and the pressure due to the actual water level (variable leg) in the vessel.
l
[
t (continued) i HATCH UNIT ?
B 3.3-91 L:kplatch\\its\\ unit 2\\ bases \\pnposed\\3-3 91495-0 l
m v.
e e
-e.,,
I ATWS-RPT Instrumentation B 3.3.4.2 l
BASES 7
APPLICABLE a.
Reactor Vessel Water level - ATWS-RPT Level l
SAFETY ANALYSES, (continued)
LCO, and APPLICABILITY Four channels of Reactor Vesse' Nater Level - ATWS-RPT Level with two channels in each orip system, are available and required to be OPERABLE to ensure that no single instrument failure can preclude an ATWS-RPT from this Function on a valid signal. The Reactor Vessel Water Level - ATWS-RPT Level Allowable Value is l
chosen so that the system will not be initiated after a Level 3 scram with feedwater still available, and for convenience with the reactor core isolation cooling initiation.
b.
Reactor Steam Dome Pressure - Hiah Excessively high RPV pressure may rupture the RCPB.
An increase in the RPV pressure during reactor operation compresses the steam voids and results in a positive reactivity insertion. This increases neutron flux and THERMAL POWER, which could potentially result in fuel failure and overpressurization. The Reactor Steam Dome Pressure - High Function initiates an RPT for transients that result in a pressure increase, counteracting the pressure increase by rapidly reducing core power generation.
For the overpressurization event, the RPT aids in the termination of the ATWS event and, along with the safety / relief valves, limits the peak RPV pressure to less than the ASME Section III Code limits.
The Reactor Steam Dome Pressure - High signals are initiated from four pressure transmitters that monitor reactor steam dome pressure.
Four channels of Reactor Steam Dome Pressure - High, with two channels in each trip system, are available and are required to be OPERABLE to ensure that no single instrument failure can preclude an ATh!S-RPT from this function on a valid signal. The Reactor Steam Dome Pressure - High Allowable Value is chosen to provide an adequate margin to the ASME Section III Code limits.
(continued)
HATCH UNIT 2 B 3.3-92 W phk h\\its\\ unit 2hne s\\ptr yoned \\3-3-92495 - 0
i RCS P/T Limits B 3.4.9 BASES
. SURVEILLANCE SR 3.4.9.1 (continued)
REQUIREMENTS cooldown operations and RCS inservice leakage and hydrostatic testing.
l-SR 3.4.9.2 A separate limit is used when the reactor is approaching criticality. Consequently, the RCS pressure and temperature must be verified within the appropriate limits before withdrawing control rods that will make the reactor critical.
I Performing the Surveillance within 15 minutes before control rod withdrawal for the purpose of achieving criticality provides adequate. assurance that the limits will not be exceeded between the time of the Surveillance and the time of the control rod withdrawal.
SR 3.4.9.3 and SR 3.4.9.4 Differential temperatures within the applicable limits ensure that thermal stresses resulting from the startup of an idle recirculation pump will not exceed design allowances.
In addition, compliance with these limits ensures that the assumptions of the analysis for the startup of an idle recirculation loop (Ref. 7) are satisfied.
Performing the Surveillance within 15 minutes before starting the idle' recirculation pump provides adequate assurance that the limits will not be exceeded between the time of the Surveillance and the time of the idle pump start.
If the 145 F temperature differential specified in SR 3.4.9.3 cannot be determined by direct indication, on alternate method may be used as described below:
The bottom head coolant temperature and the RPV coolant can be assumed to be s 145 F if the following can be confirmed:
a.
One or more loop drive flows were > 40 percent of rated flow prior to the RPT, (continued)
HATCH UNIl 2 B 3.4-50 k:kp\\ hatch \\its\\ unit 2\\ bases \\ proposed \\3-4-50495- 0
RCS P/T Lizits B 3.4.9 BASES SURVEILLANCE SR 3.4.9.3 and SR 3.4.9.4 (continued)
REQUIREMENTS b.
High Pressure Coolant Injection (HPCI) and Reactor Core Isolation Cooling (RCIC) Systems have not injected since the RPT, c.
Feedwater temperature has remained > 300 F since the RPT, and d.
The time between the RPT and restart is < 30 minutes.
General Electric test data from BWR plants shows that stratification up to the 145"F differential does not occur any sooner than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> following the RPT (Refs. 10 and 11).
Adding HPCI and RCIC injection, and feedwater temperature constraints provides assurance that the temperature differential will not be exceeded within 30 minutes of the RPT.
An acceptable means of demonstrating compliance with the temperature differential requirement in SR 3.4.9.4 is to compare the temperatures of the operating recirculation loop and the idle loop.
SR 3.4.9.3 and SR 3.4.9.4 have been modified by a Note that requires the Surveillance to be performed only in MODES 1, 2, 3, and 4.
In MODE 5, the overall stress on limiting components is lower. Therefore, AT limits are not required.
(continued)
HATCH UNIT 2 B 3.4-50a Umphtch%nidhwaympondM50495-0 l
RCS P/T Limits B 3.4.9 This page intentionally left blank.
I HATCH UNIT 2 B 3.4-50b t*ps.tensii.surui2so..spropo.cos3-4 50495-0 l
l 1
RCS P/T Limits B 3.4.9 BASES i
REFERENCES 3.
ASTM E 185-82, " Standard Practice for Conducting j
(continued)
Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels," July 1982.
4.
3 5.
Regulatory Guide 1.99, Revision 2, May 1988.
6.
ASME, Boiler and Pressure Vessel Code,Section XI',
j Appendix E.
i 7.
FSAR, Section 15.1.26.
l 8.
Kahtan N. Jabbour (NRC) letter to W. G. Hairston, III (GPC), Amendment 118 to the Operating License, dated January 10, 1992.
9.
NRC No.93-102, " Final Policy Statement on Technical Specification Improvements," July 23, 1993.
i 10.
GE-NE-668-13-0393, " Recirculation Pump Restart Without l
Vessel Temperature Indication for E.I. Hatch Nuclear Plant " April 9, 1993.
11.
DRF A00-05834/6, " Safety & 10 CFR 50.92 Significant Hazards Consideration Assessment for RPV i
Stratification Pravention Improvements at Edwin I.
Hatch Nuclear Plar. L' nits 1 and 2," April 1994.
i i
3 l
HATCH UNIT 2 8 3.4-52 L:w si.hsii,suoii2xb.,,,'eropo,e413-4-52495-0 r
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ATWS-RPT Instrumentation B 3.3.4.2 B 3.3 INSTRUMENTATION B 3.3.4.2 Anticipated Transient Without Scram Recirculation Pump Trip (ATWS-RPT) Instru.nentation BASES BACKGP.0VND The ATWS-RPT System initiates an RPT, adding negative reactivity, following events in which a scram does not (but should) occur, to lessen the effects of an ATWS event.
Tripping the recirculation pumps adds negative reactivity from the increase in steam voiding in the core area as flo ses. When Reactor Vessel Water Level -
(f(p ATkh-WT Lev Reactor Steam Dome Pressure - High satpoin he recirculation pump drive motor breakers trip.
reac The ATWS-RPT System (Ref.1) includes sensors, relays, bypass capability, circuit breakers, and switches that are necessary to cause initiation of an RPT. The channels include electronic equipment (e.g., trip units) that t
compares measured input signals with pre-established setpoints. When the setpoint is exceeded, the channel output relay actuates, which then outputs an ATWS-RPT signal to the trip logic.
- ATW5-RPr The ATWS-RPT consists of two independent trip systems, with two channels of Reactor Steam Dome Pressure nd channels of Reactor Vessel Water Level -
MFf ve i
each trip system.
Each ATWS-RPT trip syst two-out-of-two logic for net-Thus, either two 4
two Reactor Reactor Water Level -
eve Pressure - High signals a
ed 1 ip a trip system.
The outputs of the chan 1els in a trip system are combined in a logic so that either rip system will trip both recirculation pumps (by tripping the respective drive motor breakers).
There is one drive motor breaker provided for each of the two recirculation pumps for a total of two breakers. The output of each trip system is provided to both recirculation 1
pump breakers.
(continued) 1 HATCH UNIT 1 B 3.3-89 REVISION 0
ATWS-RPT Instrumentation B 3.3.4.2 BASES 7 g gpp APPLICABLE The individua l functions are required to be OPERABLE in SAFETY ANALYSES, MODE 1 to prctect against common mode failures of the LCO, and Reactor Protection System by providing a diverse trip to APPLICABILITY mitigate the nsequences of a postulated ATWS event.
The (continued)
Reactor St n om Pr
- High and Reactor Vessel Water Level -
i@ L v nctions are required to be OPERABLE i MOD
,s the reactor is producing significant er and the recirculation system could be at high flow. During this MODE, the potential exists for pressure increases or low water level, assuming an ATWS event.
In MODE 2, the reactor is at low power and the recirculation system is at low flow; thus, the potential is low for a pressure increase or low water level, assuming an ATWS event. Therefore, the ATWS-RPT is not necessary.
In MODES 3 and 4, the reactor is shut down with all control rods inserted; thus, an ATWS event is not significant and the possibility of a significant pressure increase or low water level is negligible.
In MODE 5, the one rod out interlock ensures that the reactor remains subcritical; thus, an ATWS event is not significant.
In addition, the reactor pressure vessel (RPV) head is not fully tensioned and no pressure transient threat to the reactor coolant pressure boundary (RCP8) exists.
The specific Applicable Safety Analyses and LC0 discussions are luted below on a Function by Function basis.
hTD6-a.
eve d gs v
Low RPV water levei indicates the capability to co the fuel may be threatened.
Should RPV water 1
$ od bl decrease too far, fuel damage could result.
l Therefore, the ATWS-RPT System is initiated to aid in maintaining level above the top of the active fuel. The reduction of core flow reduces the neutron flux and THERMAL POWER and, therefore, the rate of coolant boiloff.
Reactor vessel water level signals are initiated from four level transmitters that sense the difference between the pressure due to a constant column of water (reference leg) and the pressure due to the actual water level (variable leg) in the vessel.
(continued)
HATCH UNIT 1 B 3.3-91 REVISION O
ATWS-RPT Instrumentation B 3.3.4.2 BASES
~
APPLICABLE a.
& Leve SAFETY ANALYSES, (continued)
'Q LCO, and APPLICABILITY Four
.els of Reactor Vessel Water Level -
fgtOS-RT~Leve with two channels in each trip system, a avai and required to be OPERABLE to ensure that no single instrument failure can preclude an ATWS-RPT from this Function on a ign he Reacter Wssel Water Level -#
eve lowable Value is chosen so that the uy ill e initiated after a Level 3 scram uith feedwater still available, and for convenience wi :h the reactor core isolation cocling initiation.
- ATOsMr b.
Reactor Steam Dome Pressure - Hioh Excessively high RPV pressure may rupture the RCPB.
An i, crease in the RPV pressure during reactor operation compresses the steam voids and results in a positive reactivity insertion. This increases p.utron flux and THERMAL POWER, which could potentially result in fuel failure and overpressurization. The Reactor Steam Dome Pressure - High Function initiates an RPT for transients that result in a pressure increase, counteracting the pressure increase by rapidly reducing core power generation.
For the overpressurization event, the RPT aids in the termination of the ATWS event and, along with the safety / relief valves, limits the peak RPV pressure to less than the ASME Section III Code limits.
The Reactor Steam Dome Pressure - High signals are initiated from four pressure transmitters that monitor reactor steam dome pressure.
Four channels of Reactor Steam Dome Pressure - High, with two channels in each trip system, are available and are required to be OPERABLE to ensure that no single instrument failure can preclude an ATWS-RPT from this function on a valid signal. The Reactor Steam Dome Pressure - High Allowable Value is chosen to provide an adequate margin to the ASME Section III Code limits.
(continued)
HATCH UNIT 1 B 3.3-92 REVISION O
RCS P/T Limits B 3.4.9 l
BASES i
SURVEILLAf4CE SR 3.4.9.1 (continued)
REQUIREMENTS cooldown operations and RCS inservice leakage and hydrostatic testing.
SR 3.4.92 A separate limit is used when the reactor is approaching criticality. Consequen;ly, the RCS pressure and temperature must be verified withir the appropriate limits before withdrawing control rods that will make the reactor critical.
Performing the Surveillance within 15 minutes before control rod withdrawal for the purpose of achieving criticality provides adequate assurance that the limits will not be exceeded between the time of the Surveillance and the time of the control rod withdrawal.
SR 3.4.9.3 and SR 3.4.9.4 Differential temperatures within the applicable limits ensure that thermal stresses resulting from the startup of an idle recirculation pump will not exceed design allowances.
In addition, compliance with these limits ensures that the assumptions of the analysis for the startup of an idle recirculation loop (Ref. 7) are satisfied.
Performing the Surveillance within 15 minutes before starting the idle recirculation pump provides adequate assurance that the limits will not be exceeded between the i
time of the Surveillance and the time of the idle pump neec+ 1.
c=> An acceptable means of demonstrating compliai (g}
temperature differential requirement in SR 3.4.9.4 is to compare tha temperatures of the operating recirculation loop i
and the idle loop.
SR 3.4.9.3 and SR 3.4.9.4 have been modified by a Note that requires the Surveillance to be performed only in MODES 1, 2, 3, and 4.
In MODE 5, the overall stress on limiting i
components is lower. Therefore, AT limits are not required.
(continued)
HATCH UNIT 1 B 3.4-50 REVISION O i
n g
INSERT 1
(
63.4-50)
If the 145 F temperature differential specified in SR 3.4.9.3 cannot be determined by direct indication, an alternate method may be used as described below:
The bottom head coolant temperature and the RPV cralant can be assumed to be 5145oF if the following can be confirmed:
a.
One or more loop drive flows were > 40 percent of rated flow prior to the RPT.
b.
High Pressure Ccoiant Injection (HPCI) and Reactor Core Isolation Cooling (RCIC) Systems havc not injected since the RPT.
c.
Feedwater temperature has remained > 300 F since the RPT. and d.
The time between the RPT and restart is < 30 minutes.
General Electric test data from BWR plants shows that stratification up to the 145oF differential does not occur any sooner than i hour following the RPT (Refs.
10 and 11).
Adding HPCI and RCIC injectio.'. and feedwater temaerature constraints provides assurance that the temperature differential viill not be exceeded within 30 minutes of the RPT.
RCS P/T Limits B 3.4.9 BASES REFERENCES 3.
ASTM E 185-82, " Standard Practice for Conduc. ting (continued)
Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels," July.1982.
4.
5.
Regulatory Guide 1.99, Revision 2, May 1988.
6.
ASME, Boiler and Pressure Vessel Code,Section XI, Appendix E.
7.
FSAR, Section 14.3.6.2.
8.
George W. Rivenbark (NRC) letter to J. T. Beckham, Jr.
(GPC), Amendment 126 to the Operating License, dated June 20, 1986.
9.
NRC No.93-102, " Final Policy Statement on Technical Specification Improvements," July 23, 1993.
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" A B I'l4 4-F HATCH UNIT 1 B 3.4-52 REVISION O
ATWS-RPT Instrumentation B 3.3.4.2 B 3.3 INSTRUMENTATION B 3.3.4.2 Anticipated Transient Without Scram Recirculation Pump Trip (ATWS-RPT) Instrumentation BASES BACKGROUND The ATWS-RPT System initiates an RPT, adding negative reactivity, 'llowing events in which a scram does not (but should) occur, to lessen the effects of an ATWS event.
Tripping the recirculation pumps adds negative reactivity from the increase in steam voiding in the core area as flo ses. When Reactor Vessel Water Level -
4f!)ge g g -@(>flev Reactor Steam Dome Pressure - High setpoint rea he recirculation pump drive motor breakers trip.
The ATWS-RPT System (Ref.1) includes sensors, relays, bypass capability, circuit breakers, and switches that are necessary to cause initiation of an RPT.
The channels include electronic equipment (e.g., trip units) that compares measured input signals with pre-established setpoints. When the setpoint is exceeded, the channel output relay actuates, which then outputs an ATWS-RPT signal to the trip logic.
Au-W The ATWS-RPT consists of tw in nt trip s tems, with two channels of Reactor Ste Dome _ Pres ig and channels of Reactor Vessel ater Level -
eve #?i each trip system.
Each AT -RPT trip system two-out-of-two logic for uncti Thus, either two Reactor Water Level -
eve r two Reactor Pressure - High signals ar ded rip a trip system.
The outputs of the channels in a trip system are combined in a logic so that either trip system will trip both recirculation pumps (by tripping the respective drive motor breakers).
There is one drive motor breaker provided for each of the two recirculation pumps for a total of two breakers. The output of each trip system is provided to both recirculation pump breakers.
(continued)
HATCH UNIT 2 B 3.3-89 REVISION 0
ATWS-RPT Instrumentation B 3.3.4.2 BASES kT(h5-N APPLICABLE The individual Functions are required to be OPERABLE in SAFETY ANALYSES, MODE 1 to protect against common mode failures of the LCO, and Reactor P otection System by providing a diverse trip to APPLICABILITY mitigate he consequences of a postulated ATWS event. The (continued)
Reactor Si e Pri
- High and Reactor Vessel Water Level - M(:V6 ev nctions are required to be OPERABLE in the reactor is producing significant power and the recirculation system could be at high flow.
During this MODE, the potential exists for pressure increases or low water level, assuming an ATWS event.
In MODE 2, the reactor is at low power and the
(
recirculation system is at low flow; thus, the potential is low for a pressure increase or low water level, assuming an ATWS event.
Therefore, the ATWS-RPT is not necessary.
In MODES 3 and 4, the reactor is shut down with all control rods inserted; thus, an ATWS event is not significant and the possibility of a significant pressure increase or low water level is negligible.
In BODE 5, the one rod out interlock ensures that the reactor remains subcritical; thus, an ATWS event is not significant.
In addition, the reactor pressure vessel (RPV) head is not fully tensioned and no pressure transient threat to the reactor coolant pressure boundary (RCPB) exists.
The specific Applicable Safety Analyses and LC0 discussions are listed below on a Function by Function basis.
ATOS-$9 a.
ev AP Low RPV water level indicates the capability to cool l
the fuel may be threatened.
Should RPV water level decrease too far, fuel damage could result.
G OVd \\etl Therefore, the ATWS-RPT System is initiated a to aid in maintaining level above the top of the active fuel.
The reduction of core flow reduces the neutron flux and THERMAL POWER and, therefore, the rate of coolant boiloff.
Reactor vessel water level signals are initiated from four level transmitters that sense the difference between the pressure due to a constant column of water (reference leg) and the pressure due to the actual water level (variable leg) in the vessel.
I (continued)
HATCH UNIT 2 B 3.3-91 REVISION 0
ATWS-RPT Instrumentation B 3.3.4.2 BASES Ams4Pr APPLICABLE a.
eve M SAFETY ANALYSES, (continued)
~~ Q LCO, and APPLICABILITY Four ls of Reactor Vessel Water Level -
ATQS-MTLev ith two channels in each trip system, a f
avai and required to be OPERABLE to ensure that no single instrument failure can preclude an ATWS-RPT from this Function on a sign he Reactor Vessel Water Level
,y Lev.fAlowableValue is chosen so that the ill e initiated after a Level 3 scram with feedwater still available,
)
and for convenience wi :h the reactor core isolation cooling initiation.
- k ilik5 M I b.
Reactor Steam Dome Pressure - Hioh Excessively high RPV pressure may rupture the RCPB.
An increase in the RPV pressure during reactor operation compresses the steam voids and results in a positive reactivity insertion. This increases neutron flux and THERMAL POWER, which could potentially result it. fuel f ailure and overpressurization.
The Reactor Steem Dome Pressure - High Function initiates an RPT for transients that result in a pressure increase, counteracting the pressure increase by rapidly reducing core power generation.
For the overpressurization event, the RPT aids in the termination of the ATWS. event and, along with the safety / relief valves, limits the peak RPV pressure to less than the ASME Section III Code limits.
The Reactor Steam Dome Pressure - High signals are initiated from four pressure transmitters that monitor reactor steam dome pressure.
Four channels of Reactor Steam Dame Pressure - High, with two channels in each trip system, are available and are required to be OPERABLE to ensure that no single instrument failure can preclude an ATWS-RPT from this Function on a valid signal. The Reactor Steam Dome Pressure - High Allowable Value is chosen to provide an adequate margin to the ASME Section III Code limits.
l (continued)
HATCH UNIT 2 B 3.3-92 REVISION 0
~
RCS P/T Limits B 3.4.9 BASES 4
SURVEILLANCE SR 3.4.9.1 (continued)
REQUIREMENTS cooldown operations and RCS inservice leakage and hydrostatic testing.
f SR 3.4.9.2 A separate limit is used when the reactor is approaching criticality. Consequently, the RCS pressure and temperature must be verified within the appropriate limits before withdrawing control rods that will make the reactor critical.
Performing the Surveillance within 15 minutes before control rod withdrawal for the purpose of achieving criticality provides adequate assurance that the limits will not be exceeded between the time of the Surveillance and the time of the control rod withdrawal.
SR 3.4.9.3 and SR 3.4.9.4 j
Differential temperatures within the applicable limits ensure that thermal stresses resulting from the startup of an idle recirculation pump will not exceed design allowances.
In addition, compliance with these limits ensures that the assumptions of the analysis for the startup of an idle recirculation loop (Ref. 7) are satisfied.
Performing the Surveillance within 15 minutes before starting the idle recirculation pump provides adequate assurance that the limits will not be exceeded between the Ay time of the Surveillance and the time of the idle pump nou start.
1.nseA L M
( Cd An acceptable means of demonstrating compliance with the temperature differential requirement in SR 3.4.9.4 is to compare the temperatures of the operating recirculation loop and the idle loop.
SR 3.4.9.3 and SR 3.4.9.4 have been modified by a Note that requires the Surveillance to be performed only in MODES 1, 2, 3, and 4.
In MODE 5, the overall stress on limiting components is lower. Therefore, AT limits are not required.
(continued)
HATCH UNIT 2 B 3.4-50 REVISION O
INSERT 1
{
@ g.4 -
If the 145oF temperature differential specified in SR 3.4.9.3 cannot be determined by direct irdication an alternate method may be used as described below:
The bottom head coolant temperature and the RPV coolant can be assumed to be 5145 F if the following can be confirmed:
a.
One or more loop drive flows were > 40 percent of rated flow prior to the RPT.
b.
High Pressure Coolant Injection (HPCI) and Reactor Core Isolation Cooling (RCIC) Systems have not injected since the RPT.
c.
Feedwater temperature has remained > 300oF since the RPT and d.
The time between the RPT and restart is < 30 minutes.
General Electric test data from BWR plants shows that stratification up to the 145of differential does not occur any sooner than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> following the RPT (Refs.
10 and 11).
Adding HPCI and RCIC injection, and feedwater temaerature constraints provides assurance that the temperature differential will not be exceeded within 30 minutes of the RPT.
RCS P/T Lioits B 3.4.9 BASES REFERENCES 3.
ASTM E 185-82, " Standard Practice for Conducting (continued)
Surveillance Tests for Light-Water Cooled Nuclear Power Reactor Vessels," July 1982.
4.
5.
Regulatory Guide 1.99, Revision 2, May 1988.
6.
ASME, Boiler and Pressure Vessel Code,Section XI, Appendix E.
7.
FSAR, Section 15.1.26.
8.
Kahtan N. Jabbour (NRC) letter to W. G. Hairston, III (GPC), Amendment 118 to the Operating License, dated January 10, 1992.
9.
NRC No.93-102, " Final Policy Statement on Technical Specification Improvements," July 23, 1993.
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HATCH CHIT 2 B 3.4-52 REVISION 0
-