ML20107H340
| ML20107H340 | |
| Person / Time | |
|---|---|
| Site: | 05200001 |
| Issue date: | 04/16/1996 |
| From: | GENERAL ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML20107H327 | List: |
| References | |
| NUDOCS 9604240086 | |
| Download: ML20107H340 (19) | |
Text
_.
ABM Design contuviDocument PROPOSED CHANGES CHANGE PACKAGE NO.10 Changes to Technical Specifications 1
9604240086 960406 PDR ADOCK 05200001 PDR A
SLs 2.0
\\
2.0 SAFETY LIMITS (SLs) 2.1 SLs 1
2.1.1 Reactor Core SLs 2.1.1.1 With the reactor steam done pressure < 5.41 MP G or-core flow < 10% rated core flow:
THERMAL POWER shall be s 25% RTP.
2.1.1.
th the reactor steam done pressure 2 5.41 MPaG an core flow 210% rated core flow:
- F+
MC shall be 21.07.
2.1.1.3 Reactor vessel water level shall be greater than the top of active irradiated fuel.
e i
f 1
2.1.2 Reactor Coolant System Pressure SL j
i Reactor steam dome pressure shall be s 9.13 MPaG.
}
[
2.2 SL Violations With any SL violation, the following actions shall be completed:
l 2.2.1 Within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />, notify the NRC Operations Center, in accordance l
with 10 CFR 50.72.
I 2.2.2 Within 2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />s:
2.2.2.1 Restore compliance with all SLs; and i
2.2.2.2 Insert all insertable control rods.
2.2.3 Within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, notify the [ General Manager-Nuclear Plant and Vice President-Nuclear Operations] and the [offsite reviewers specified in Specification 5.5.2, "[0ffsite] Review and Audit").
(continued) i ABWR TS 2.0-1 Rev, D. Design Control Docisnent/ Tier 2
f SSLC Sensor Instrumentation 3.3.1.1 I
l Table 3.3.1.1 1 (Page 3 of 7)
$sLC sensor Instrumentation APPLICABLE CONDIT!0Ns MODES OR REFERENCED OTNER FROM SPECIFIED REQUIRED REQUIRED SURVEILLANCE ALLOWABLE FUNCTION CONDITIONS CNANNELS ACTIONS REQUIREMENTS VALUE 7.
Reactor Vessel Water Levet. Low, l
Level 2 7a. ESF Initiation 1,2,3 4
N SR 3.3.1.1.1 m ( ) em st 3.3.1.1.5 SR 3.3.1.1.9 st 3.3.1.1.10 st 3.3.1.1.13 7b. Isolation initiation.
1,2,3 4
K sa 3.3.1.1.1 a t 3 ce st 3.3.1.1.5 st 3.3.1.1.9 SR 3.3.1.1.10 st 3.3.1.1.14 (f) 4 L
st 1.3.1.1.1 SR 3.3.1.1.5 sr. 3.3.1.1.9
?At 3.3.1.1.10
/
SR 3.3.1.1.14
) d 7c. SLCS and FWRB initiation 1,2 4
G SR 3.3.1.1.1 h[ 1,6 M SR 3.3.1.1.6 SR 3.3.1.1.11 8.
Reacto.- Vessel Water Level -Low, Level 1.5 Sa. ESF Initiation.
1,2,3, 4
N st 3.3.1.1.1 2 [ 3 cm SR 3.3.1.1.5 4('},5(*)
SR 3.3.1.1.9 st 3.3.1.1.10 SR 3.3.1.1.13 ab. Isolation Initiation.
1,2,3 4
Q SR 3.3.1.1.1 I 3 cm SR 3.3.1.1.5 st 3.3.1.1.9 SR 3.3.1.1.10 SR 3.3.1.1.14 Sc. ATWs ADS Inhibit.
1, 2 4
H st 3.3.1.1.1 t I 3 cm SR 3.3.1.1.5 SR 3.3.1.1.9 sa 3.3.1.1.10 9.
Reactor Vessel Water Level-Low, Level 1 9a. ADS A, CAMS A, LPFL A &
1,2,3, 4
N SR 3.3.1.1.1 2 I ) em LPFL C Initiation SR 3.3.1.1.5 SR 3.3.1.1.9 4(e)* $(e)
SR 3.3.1.1.10 SR 3.3.1.1.13 9b. ads B, Diesel Generator, 1,2,3, 4
N SR 3.3.1.1.1 2 I ) cm RCW, CAMS B, & LPFL B SR 3.3.1.1.5 sR 3.3.1.1.9 Initiation 4(e)' $(e)
SR 3.3.1.1.10 f
SR 3.3.1.1.13 g
(Continued)
ABWR TS 3.3-12 Rev. O. Design Control Document / Tier 2 i
i
l SRNM Instrumentation 3.3.2.1 SURVEILLANCE REQUIREMENTS (Continued)
SURVEILLANCE FREQUENCY SR 3.3.2.1.3 NOTE Not required to be met with four or less fuel assemblies adjacent to the SRNM and no other fuel assemblies in the associated core quadrant.
l Verify count rate is 2 3.0 cp,
~~~~~~~~n'~~
- 12. hours i 4lce uf' L
during CORE u
b l
ALTERATIONS l
AND 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> SR 3.3.2.1.4 Perform CHANNEL FUNCTIONAL TEST.
[7]' days l-l l
SR 3.3.2.1.5 Perform CHANNEL FUNCTIONAL TEST.
[31] days i
i SR 3.3.2.1.6 NOTE
(
Neutron detectors are excluded.
1 I
Perform CHANNEL CALIBRATION.
18 months i
i l
4 ABWR TS 3.3-42 Rev.1. Design 0:ntrol Documen:/ Tier 2 1
- - - -._ = _ _ _ - _
S/RVs 3.4.2 SURVEILLANCE REQUIREMENTS i
SURVEILLANCE FREQUENCY SR 3.4.2.1 Verify the safety function lift setpoints e
of the required S/RVs are as follows:
In 3,c e,y.g3mcf W5 t
Number of Setpoint S/RVs (MPaG)
Tw se ce 2
7.92 1 0.0792
^#f Progt h 4
7.99 0.0799
- f'*:*i' u
i 4
8.19 1 0.0819 l
Following testing, lift settings shall be within i 1%.
NOTE--------------------
Not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> i
after reactor steam dome pressure is 2 6.55 MPaG.
Verify each required S/RV opens when 18 months o a manually actuated.
Sf eac alv solen ~d ABWR TS 3.4-3 Rev. O, Design Control Document / Tier 2
._-=.
RCW/RSW System and UHS-Operating-3.7.1 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.1.1 Verify the water level of each UH3 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />
[ spray pond] is 2 [ ] m.
SR 3.7.1.2 Verify the water level in each RSW pump 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> well of the intake structure is 2 [ ] m.
gsw at A tole b +o Verify the[m?% 4 M. R.cw/#.3wewatertemperature[+f 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> SR 3.7.1.3 head
^0
.2 h c. Wy
(= 3 3.3
- c.
NOTE--------------------
Isolation of flow to individual components does not render RCW/RSW System inoperable.
Verify each RCW/RSW division and associated 31 days UHS [ spray network] division manual, power operated, and automatic valve in the flow path servicing safety related systems or components, that is not locked, sealed, or otherwise secured in position, is in the correct position.
SR 3.7.1.5 Verify each RCW/RSW division and associated 18 months UHS [ spray network] division actuate on an actual or simulated initiation signal.
i ABWR TS 3.7-3 Rev. o. Design control Document / Tier 2
RCW/RSW System and UHS - Shutdown 3.7.2 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.2.1 Verify the water level of each UHS 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />
[ spray pond) is 2 [ ] m.
SR 3.7.2.2 Verify the water level in each RSW pump 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> well of the intake structure is 2 [ ] m.
at the inie+ eo Verify the =it.Swrm water temperatur 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> SR 3.7.2.3
'Jll0 '
Me RcW/R.sw he.o e.x cb ; Z'C.
6 3 3. S
- C.
ogavs tS SR 3.7.2.4
NOTE--------------------
Isolation of flow to individual components does not render RCW/RSW System inoperable.
Verify each RCW/RSW division and associated 31 days UHS [ spray network] division manual, power operated, and automatic valve in the flow path servicing safety related systems or components, that is not locked, sealed, or otherwise secured in position, is in the correct position.
SR 3.7.2.5 Verify each RCW/RSW division and associated 18 months UHS [ spray network] division actuate on an actual or simulated initiation signal.
1 ABWR TS 3.7-6 Rev. O, Design Control Document / Tier 2
l RCW/RSW System and UHS - Refueling 3.7.3 SURVEILLANCE REQUIREMENTS SURVEILLANCE FREQUENCY SR 3.7.3.1 Verify the water level of each UHS [ spray 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> pond] is 2: [ ] m.
L l
l SR 3.7.3.2 Verify the water level in each RSW pump 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> well of the intake structure is 2: [ ] m.
l t
at me inte+ %
l g3w 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Verify the eu..se wate/r temperatureAcf U:lS-l SR 3.7.3.3 9.cw R.s'W h e.d 4.; 3;"C. A t3
- f: 33.'3 #c.
e.x ek.gev r l
NOTE------------------------
Isolation of flow to individual components l
does not render RCW/RSW System inoperable.
Verify RCW/RSW division and associated UHS 31 days
[ spray network] division manual, power operated, and automatic valve in the flow path servicing safety related systems or components, that is not locked, sealed, or l
otherwise secured in position is in the correct position.
SR 3.7.3.5 Verify each RCW/RSW division and assoicated 18 months UHS [ spray network] division actuate on an actual or simulated initiation signal.
t J
ABWR TS 3.7-8 Rev. D. Design Control Document / Tier 2
Design Features 4.0 4.0 DESIGN FEATURES (continued) 4.3 Fuel Storage 4.3.1 Criticality
]
4.3.1.1 The spent fuel storage racks are designed and shall be maintained with:
a.
Fuel assemblies having a maximum k-infinity of 1.35 in the normal reactor core configuration at cold conditions; b.
k s 0.95 if. fully flooded with unborated water, which iE21udes an allowance for uncertainties as described in Section 9.1 of the DCD Tier 2.
l l
4.3.1.2 The new fuel storage racks are designed and shall be maintained with:
a.
Fuel assemblies having a maximum k-infinity of 1.35 in pfat &
the normal reactor core configuration at 20 *C;
$ 0.95 if fully flooded with unborated water, which b.
k 'c#i$ 1udes an allowance for uncertainties as describe 4, \\;we, Section 9.1 of the DCD Tier 2; c.
k,,, s 0.98 if moderated by aqueous foam, which includes an allowance for uncertainties as described in Section 9.1 of the DCD Tier 2; and d.
A nominal [approximately 16] cm center to center distance between fuel assemblies placed in storage racks.
4.3.2 Drainaae J
The spent fuel storage pool is designed and shall be maintained to prevent inadvertent draining of the pool below 3.1 m above the top of the active fuel.
]
4.3.3 Capacity 4.3.3.1 The spent fuel storage pool is designed and shall be maintained with a storage capacity limited to no less than 2354 fuel assemblies (270% of full core discharge).
ABWR TS 4.0-2 Rev. O, Design control Document / Tier 2
S/RVs B 3.4.2 BASES ACTIONS The 14 day Completion Time to restore the inoperable -
(continued) required S/RVs to OPERABLE status is based on the relief capability of the remaining S/RVs, the low probability of an i
event requiring S/RV actuation, and a reasonable time to complete the Required Action.
)
B.1 and B.2 i
With less than the minimum number of required S/RVs OPERABLE, a transient may result in the violation of the ASME Code limit on reactor pressure.
If the inoperable required S/RV cannot be restored to OPERABLE status within the associated Completion Time of Required Action A.1 or if two or more required S/RVs are inoperable, the plant must be brought to a MODE in which the LCO does not apply. To achieve this status, the plant must be brought to at least MODE 3 within 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> and to MODE 4 within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. The
{
allowed Completion Times are reasonable, based on operating experience, to reach the required plant conditions from full power conditions in an orderly manner and without challenging plant systems.
SURVEILLANCE SR 3.4.2.1 REQUIREMENTS This Surveillance demonstrates that the required S/RVs will open at the pressures assumed in the safety analysis of Reference 2.
The demonstration of the S/RV safety function lift settings must be performed during shutdown, since this is a bench test. The lift setting pressure shall correspond to ambient conditions of the valves at nominal operating temperatures and pressures.
The S/RV setpoint is i 3% for OPERABILITY; however, the valves are reset to 1% during the Surveillance to allow for drift.
Th: I" = nth Trcquency.;;.: ::1 :ted h::::: thi:
("-veill m.=
d M perfened during :hutds; ::ndition: -
M i: h::d n the tia ht= :: e "emy a r
Sjweif'eal in a cc o rda* c*--.
fQ fy i4 e ib w,% Ke Z wrv;<e b yre+l*~ fv=ya~-
(continued)
ABWR TS B 3.4-8 Rev. O. Design Control Document / Tier 2
S/RVs B 3.4.2
(
BASES SURVEILLANCE SR 3.4.2.2 REQUIREMENTS (continued)
A manual actuation of each required S/RV is performed to verify that, mechanically, the valve is functior.ing properly and no blockage exists in the valve discharge line. This can be demonstrated by the response of the turbine control valves or bypass valves, by a change in the measured steam flow, or any other method suitable to verify steam flow.
Adequate reactor steam dome pressure must be available to perform this test to avoid damaging the valve.
Sufficient time is therefore allowed after the required pressure is l
achieved to perform this test. Adequate pressure at which l
this test is to be performed is [6.55] MPaG (the pressure recommended by the valve manufacturer).
Plant startup is 4
allowed prior to performing this test because valve OPERABILITY and the setpoints for overpressure protection are verified, per ASME requirements, prior to valve installation. Therefore, this SR is modified by a Note that states the Surveillance is not required to be performed until 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> after reactor steam dome pressure is 2 ([6.55] MPaG). The 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> allowed for manual actuation after the required pressure is reached is sufficient to achieve stable conditions for testing and provides a i
reasonable time to complete the SR.
If the valve fails to actuate due only to the failure of the solenoid but is j
capable of opening on overpressure, the safety function of the S/RV is considered OPERABLE.
[The18monthonaSTAGGEREDTESTBASISFrequencyensures
( that each solenoid for each S/RV is alternately tested The
' Frequency is consistent with SR 3.4.2.1 to ensure that the S/RVs are manually actuated following removal for refurbishment or lift setpoint testing.
I i
j REFERENCES 1.
ASME, Boiler and Pressure Vessel Code,Section III.
2.
DCD Tier 2, Section 5.2.2.
l 3.
DCD Tier 2, Chapter 15.
P i
ABWR TS B 3.4-9 Rev. O, Design Control Document / Tier 2
R3, o ECCS -0perating 8 3.5.1 BASES 1.03 7
BACKGROUND The RCIC System is designed to pr vid core cooling for a
( Continued )
wide range of reactor pressures,
.04 MPaG to 8.12 MPaG. Upon receipt of an init ation signal, the RCIC turbine accelerates to a specified speed.
As the RCIC flow increases, the turbine control valve is automatically adjusted to maintain design flow.
Exhaust steam from the RCIC turbine is discharged to the suppression pool. A full flow test line is provided to route water from and to the suppression pool to allow testing of the RCIC System during normal operation without injecting water into the RPV.
For the station black out scenario, where all AC power from the offsite AC circuits and from the standby diesel generators are assumed to be lost, RCIC is designed to provide makeup water to the RPV.
Diverse alternatives to RCIC are provided by the Combustion Turbine Generator (CTG) and the AC-Independent Water Addition (ACIWA) mode of RHR(C)
(References 13 and 14).
If RCIC is inoperable, water can be injected into the RPV either by powering other ECCS subsystems from the CTG or by the Fire Protection System (FPS) using the ACIWA mode of RHR(C).
The ECCS pumps are provided with minimum flow bypass lines, g
which discharge to the suppression pool.
The valves in these lines automatically open to prevent pump damage due to overheating when other discharge line valves are closed or RPV pressure is greater than the LPFL pump discharge pressures following system initiation.
To ensure rapid delivery of water to the RPV and to minimize water hammer i
effects, the ECCS discharge line " keep fill" systems are designed to maintain all pump discharge lines filled with water.
i The ADS (Ref. 1) consists of 8 of the 18 S/RVs.
It is designed to provide depressurization of the primary system during a small break LOCA if RCIC and HPCF fail or are unable to maintain required water level in the RPV. ADS operation reduces the RPV pressure to within the operating pressure range of the low pressure ECCS subsystems (LPFL),
so that these subsystems can provide core cooling.
Each ADS valve is supplied with pneumatic power from either its own dedicated accumulator located in the drywell, or from the atmospheric control system (ACS) directly when pneumatic power from the accumulators is not needed. The ACS also supplies the nitrogen (at pressure) necessary to assure the ADS accumulators remain charged for use in emergency actuation.
)
(continued)
ABWR TS B 3.5-4 Rev. O, Design Control Document / Tier 2
.... -,.. -.... ~...
ECCS-Operating B 3.5.1 BASES LCO With less than the required number of ECCS subsystems j
(continued)
OPERABLE during a limiting design basis LOCA concurrent with the worst case single failure, the margins to the limits specified in 10 CFR 50.46 (Ref. 7) would be reduced.
Furthermore, all ECCS subsystems are assumed to be initially available in the comprehensive set of analyses performed to satisfy the single failure criterion required by 10 CFR 50.46 (Ref. 7). Thus all ECCS subsystems must be OPERABLE.
i l
The ECCS is supported by other systems that provide automatic ECCS initiation signals (LCO 3.3.1.1, "SSLC Sensor Instrumentation" and LCO 3.3.1.4, "ESF Actuation i
Instrumentation"), cooling and service water to cool rooms i
containing ECCS equipment (LCO 3.7.1, " Reactor Building i
Cooling Water (RCW) System, Reactor Service Water (RSW) i System and Ultimate Heat Sink (VHS)-Operating", LC0 3.7.2, "RCW/RSW and UHS-Shutdown" and LCO 3.7.3 "RCW/RSW and UHS-a i
Refueling"), electrical power (LCO 3.8.1, "AC Sources-Operating," and LCO 3.8.4, "DC Sources-Operating").
l A LPFL subsystem may be considered OPERABLE during alignment j
and operation for decay heat removal when below the actual j
RHR cut in permissive pressure in MODE 3, if capable of i
)
being manually realigned (remote or local) to the LPFL mode i
and not otherwise inoperable.
At-these low pressures and decay heat levels, a reduced complement of ECCS subsystems can provide the rew! red core cooling, thereby allowing j
operation of an RHR shutdown cooling loop when necessary.
I 1.03 j
i APPLICABILITY All ECCS subsystems are required to be OP BLE during MODES 1, 2, and 3 when there is consider le energy in the reactor core and core cooling would be quired to prevent fuel damage in the event of a break in he primery system piping.
In MODES 2 and 3, the RCI tem is not required to be OPERABLE when pressure is s
.0 MPaG since other ECCS subsystems can provide sufficient ow to the vessel.
In MODES 2 and 3, the ADS function is not required when pressure is s 0.343 MPaG because the low pressure ECCS subsystems (LPFL) are capable of providing flow into the RPV below this pressure.
ECCS requirements for MODES 4 and 5 are specified in LCO 3.5.2, "ECCS-Shutdown."
(continued)
ABWR TS B 3.5-7 Rev. O, Design Control Document / Tier 2
RCW/RSW System and UHS-Operating C 3.7.1 BASES
]
LCO (continued) d.
The associated piping, valves, instrumentation, and controls required to perform the safety related g g un g o ga g q Ey Lg g g
OdElfAB[LITY$fTh[Iihisbasedonamaximumfwater temperatureg f 25'S with OPERABILITY of each division uirinofminimum water level at or above elevation [mean sea ieve". (equivalent to an indicated level of 2: [ ] m) and six OPERABLE spray networks].
The isolation of the RCW/RSW System to components or systems i
may render those components or systems inoperable, but does not affect the OPERABILITY of the RCW/RSW System.
i APPLICABILITY In MODES 1, 2, and 3, the RCW/RSW System and UHS are required to be OPERABLE to support OPERABILITY of the equipment serviced by the RCW/RSW System and UHS, and are required to be OPERABLE in these MODES.
In MODES 4 and 5, the OPERABILITY requirements of the RCW/RSW System and UHS are specified in LCOs 3.7.2, "RCW/RSW and UHS-Shutdown" and 3.7.3, "RCW/RSW and UHS-Refueling".
ACTIONS Ad If one RCW pump and/or one RSW pump and/or one RCW/RSW heat exchanger and/or one [ spray network] in the UHS in the same division is inoperable, action must be taken to restore the inoperable component (s), and thus the division affected, to OPERABLE status within 14 days.
In this condition sufficient equipment is still available to provide cooling water to the required safety related components and sufficient heat removal capacity is still available to adequately cool safety related loads, even assuming the worst case single failure. Therefore, continued operation for a limited time is justified.
The 14-day Completion Time is reasonable, based on the low probability of an accident occurring during the 14 days that one or more components are inoperable in one division, the number of available redundant divisions, the substantial (continued)
ABWR TS B 3.7-4 Rev. O. Design Control Document / Tier 2
RCW/RSW System and UHS-Operating B 3.7.1 l
l l
BASES L
i SURVEILLANCE SR 3,7.1.1 REQUIREMENTS This SR ensures adequate long term (30 days) cooling can be maintained. With the UHS water source below the minimum level, the affected RCW/RSW division must be declared l
inoperable. The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Frequency is based on operating l
experience related to trending of the parameter variations during the applicable MODES.
SR 3.7.1.2 This SR verifies the water level in each RSW pump well of the intake structure to be sufficient for the proper operation of the RSW pumps (net positive suction head and l
pump vortexing are considered in determining this limit).
l The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Frequency is based on operating experience related to trending of the parameter variations during the applicable MODES.
a e m e. r Me t gg,4ev
-)-o 4b.e. g.cw /R.S W h
k e.wc ho. der SR 3.7.1.3 Verification of the UHS temperature ensures that the heat removal capability of the RCW/RSW System is within the assumptions of the DBA analysis.
The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Frequency is based on operating experience related to trending of the parameter variations during the applicable MODES.
SR 3.7.1.4 Verifying the correct alignment for each manual, power operated, and automatic valve in each RCW/RSW and associated UHS [ spray network) division flow path provides assurance that the proper flow paths will exist for RCW/RSW operation. This SR does not apply to valves that are locked, sealed, or otherwise secured in position, since these valves were verified to be in the correct position prior to locking, sealing, or securing.
A valve is also allowed to be in the nonaccident position and yet considered in the correct position, provided it can be automatically l
realigned to its accident position. This SR does not l
require any testing or valve manipulation; rather, it involves verification that those valves capable of potentially being mispositioned are in the correct position.
(continued)
ABWR TS B 3.7-7 Rev. o, oesign control occumentnier 2
(
RCW/RSW System and UHS-Shutdown B 3.7.2 l-BASES SURVEILLANCE SR 3.7.2.1 REQUIREMENTS This SR ensures adequate long term (30 days) cooling can be maintained.
With the UHS water source below the minimum level, the affected RCW/RSW division must be declared inoperable.
The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Frequency is based on operating experience related to trending of the parameter variations during the applicable MODES.
SR 3.7.2.2 This SR verifies the water level in each RSW pump well of the intake structure to be sufficient for the proper operation of the RSW pumps (net positive suction head and pump vortexing are considered in determining this limit).
The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Frequency is based on operating experience related to trending of the parameter variations during the applicable MODES.
a+ the inte +
to 4he. Rcw/RTW l
ad ex ebaMtv3 l
SR 3.7.2.3 p Qw Verification of the temperature ensures that the heat removal capability of the RCW/RSW ystem is within the assumptions of the DBA analysis. The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Frequency is based on operating experience related to trending of the j
parameter variations during the applicable MODES.
l SR 3.7.2.4 Verifying the correct alignment for each manual, power l
operated, and automatic valve in each RCW/RSW and associated i
UHS [ spray network] division flow path provides assurance that the proper flow paths will exist for RCW/RSW operation.
This SR does not apply to valves that are locked, sealed, or otherwise secured in position, since these valves were i
verified to be in the correct position prior to locking, l
sealing, or securing. A valve is also allowed to be in the nonaccident position and yet considered.in the correct l
position, provided it can be automatically realigned to its accident position.
This SR does not require any testing or valve manipulation; rather, it involves verification that those valves capable of potentially being mispositioned are j
in the correct position.
This SR does not apply to valves i
I (continued) i ABWR TS B 3.7-13 Rev. O. Design Control Document / Tier 2 1
1
RCW/RSW System and UHS-RGfueling B 3.7.3 f
BASES ACTIONS A.1 and A.2 (continued) diesel generator made inoperable and LCO 3.9.7, " Residual Heat Removal (RHR)-High Water Level" for RHR shutdown
- cooling made inoperable.
This is in accordance with LCO 3.0.6 and ensures the proper actions are taken for these components.
i SURVEILLANCE SR 3.7.3.1 i
REQUIREMENTS l
This SR ensures adequate long term (30 days) cooling can be j
maintained. With the UHS water source below the minimum level, the affected RCW/RSW division must be declared i
inoperable. The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Frequency is based on operating experience related to trending of the parameter variations i
during the applicable MODES.
9 i
l SR 3.7.3.2 This SR verifies the water level in each RSW pump well of 3
the intake structure to be sufficient for the proper operation of the RSW pumps (net positive suction head ~and pump vortexing are considered in determining this limit).
The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Frequency is based on operating experience related to trending of the parameter variat ns durinJ the applicable MODES.
6N IME 4 MMM 40 ft.S w w a h r SR 3.7.3.3 99b Verification of thej[4 hts temperature ensures that the heat removal capability of the RCW/RSW ystem is within the assumptions of the DBA analysis. The 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> Frequency is based on operating experience related to trending of the parameter variations during the applicable MODES.
SR 3.7.3.4 Verifying the correct alignment for each manual, power operated, and automatic valve in each RCW/RSW and associated UHS [ spray network] division flow path provides assurance (continued)
ABWR TS' B 3.7-17 Rev. O. Design Control Document / Tier 2
3 Rev.o-ABWR oesten cenererDocumentirier2 e-(AS%Iwk*f D % kN Rcw/tt3 maintain the unit in a safe shutdown condition. Theptf5 temperaturgis nob**
D' 7
to exceed 85'C average-cb v ca L o C A.
(2) In the event of an accident, the UHS is designed to provide sufficient cooling water to the RSW System to safely dissipate the heat for that accident. The amount of heat to be removed is provided in Tables 9.2-4a,9.2-4b and 9.2-4c.
(3) The UHS is sized so that makeup water is not required for at least 30 days following an accident and design basis temperature and chemistry limits for -
safety-related equipment are not exceeded.
i (4) The UHS is designed to perform its safety function during periods of adverse site conditions, resulting in maximum water consumption and minimum cooling capability.
(5) The UilS is designed to withstand the most severe natural phenomenon or site-related event (e. g., SSE, tornado, hurricane, flood, freezing, spraying, pipe whip, jet forces, missiles, fire, failure of non-Seismic Category I equipment, flooding as a result of pipe failures or transportation accident),
and reasonably probable combinations ofless severe phenomena and/or events, without impainng its safety function.
l (6) The safety-related portion of the UHS shall be designed to perform its f
required cooling function assuming a single active failure in any mechanical or electrical system.
(7) The UHS is designed to withstand any credible single failure of man-made structural features without impairing its safety function.
(8) All safety-related heat rejection systems shall be redundant so that the essential cooling function can be performed even with the complete loss of one division. Single failures of passive components in electrical systems may lead to the loss of the affected pump, valve or other components and the partial or complete loss of cooling capability of that division but not of other divisions.
(9) The UHS and any pumps, valves, structures or other components that remove heat from safety systems shall be designed to Seismic Category I and ASME Code,Section III, Class 3, Quality Assurance B, Quality Group C, IEEE-279 and IEEE-308 requirements.
(10) The safety-related portions of the UHS shall be mechanically and electrically separated.
(11) The UHS is designed to include the capability for full operational inspection and testing.
9.2-6 Watersystems
R:v. 0 s
ABWR oesion controloceanentmer2 The MUWP System is not safety-related. However, the systems incorporate features that assure reliable operation over the full range of normal plant operations.
9.2.10.4 Tests and inspections The MUWP System is proved operable by its use during normal plant operation.
Portions of the system normally closed to flow can be tested to ensure operability and integrity of the system.
i l
Flow to the various systems is balanced by means of manual valves at the indisidual takeoffpoints.
9.2.11 Reactor Building Cooling Water System 92.11.1 Design Bases 9.2.1.1.1.1 Safety Design Bases j
(1) The Reactor Building Cooling Water (RCW) System shall be designed to j
remove heat from plant aaxiliaries which are required for a safe reactor
{
shutdown, as well as those auxiliaries whose operation is desired following a i
LOCA, but not essential to safe shutdown._
Q,y +o me CRsw wa+ev ** er mb"v^s o f 3 5 +c.
uou aityYs'/c sNn eatremovla requirementduringa The heat removal c
base LOCA with the maximum U!IS t=p-"e (i.. 35 C msgc). As shown in Table 9.2-4a, the heat removal requirement is higher during other plant operation modes, such as shutdown at 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br />. However, the RGV System is designed to remove this larger amount of heat to meet the requirements in Subsection 5.4.7.1.1.7.
(2) The RCW System shall be designed to perform its required cooling functions following a LOCA, assuming a single active or passive failure.
(3) The safety-related portions and valves isolating the non-safety-related portions of the RCW System shall be designed to Seismic Category I and the ASME Code,Section III, Class 3, Quality Assurance B, Quality Group C, IEEE-279 and IEEE-308 requirements.
(4) The RCW System shall be designed to limit leakage to the environment of radioactive contamination that may enter the RCW System from the RHR System.
(5) Safety-related portions of the RCW System shall be protected from flooding, spraying, steam impingement, pipe whip, jet forces, missiles, fire, and the effect of failure of any non-Seismic Category I equipment, as required.
9.2 20 WaterSystems