ML20095J375
| ML20095J375 | |
| Person / Time | |
|---|---|
| Site: | 05200001 |
| Issue date: | 04/24/1992 |
| From: | Stirm R GENERAL ELECTRIC CO. |
| To: | Pierson R NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM), Office of Nuclear Reactor Regulation |
| References | |
| EEN-9261, MFN-102-92, NUDOCS 9205010292 | |
| Download: ML20095J375 (125) | |
Text
..
GE NuclearEnergy L usm man amu %
?,u -
..w sx :w a s m April 24,1992 MFN No.102-92 Docket No. STN 52-00t EEN-9261 Document Control Desk U.S. Nuclear Regulatory Commission Washington, D.C. 20555 Attention:
Robert C. Pierson, Director Standardization and Non Power Reactor Project Diiectorate
Subject:
GE Response to the Resclution of OutstanJing issues 140 and 144-of AHWR DSER SECY 91355 :
,!-[
Enclosed are thirty-four (34) copies of the GE response to Outstanding Issues 140 and 144 bothi
- ~
pertaining to the revision of Appendix 15E -lATW'S analysis. Since the changes are so extensive, a complete Appendix 15E (including the unchanged pages) is included in this transmittal.
.O L;
it is intended that GE will amerd the SSAR with this revised Appendix 15E.
f Sincerely, i
9
' R. C. Stirm, Acting Manager Regulatory and Analysis Services l
M/C 444, (408) 925-6948 l
cc: F. A. Ross (DOE)
L N._ D. Fletcher
.(DOE) l' C. Poslusny, Jr.
(NRC)
G. Thomas
- (NRC)
F
- R. C. Berglund (GE)
-1
- J. F. Quirk '
(GE) o 4
j.
i 4
9
]
L-canona j
\\D 9205010292.920424--
PDR. ADOCK- 05200001
..i.h
].
PDR-
Q
)
APPENDIX 15E ATWS PERFORMANCE EVALUATION O
O ABWR Standard Plant SECTION 15E l
CONTENTS Section
.Illig
.fagg 15E.1 INTRODUCTION 15E.1 1 15E.2 PERFORMANCE REOUIREMENTS 15E.2-1 15E3 A,NALYSIS CONDITIONS IJE3-1 15E.4 ATWS LOGIC AND SETPOINTS 15E.4-1 15E.5 SELECTION OF EVENTS 15E.5-1 15E.6 TRANSIENT RESPONSES -
15E.61 -
l 15E,7 CONCLUSION 15E.7-1 15E.8 REFERENCE
. 15E.8 O 4-e s
!O
O Anwa Standard Plant 15E.1 INTRODUCTION Typical ATWS events are analped for ABWR to confirm the design for ABWR.
The procedure and assumptions used in this analysis are consistent with those used in the analyses for the operating plants as docu. mied in Section 15E.8, Reference 1.
v d
1 ABWR s
Standard Plant 15E.2 PERFORMANCE REQUIREMENTS As identified in Section 15E.8, Reference 1, the design should meet the following requirements:
(1)
Fuel Inter <rity - The long term core cooling capacity shall be assured by meeting the cladding temperature and oxidation criteria of 10CFR50.46 (1.c.,
peak cladding temperature not exceeding -
1207 C or 220& F, and the local oxidation of the cladding not exceeding 17% of the total cladding thickness).
(2)
Containment Integrity - The long term containment capab!!ity shall be maintained. The maximum containment pressure shall not exceef the design pressure (3.16 kg/cm g) of the containment structure. The suppression O'
pool temperature shall be limited to values shown in Table 15E.2-1.
(3)
Primary SystelD - The system transient pressure shall be limited such that the maximum primary stress within the reactor coolant pressure boundary (RCPB) dm no exceed the emergency limits as dc.Sned in the ASME-code, Section 11_I _ If practical, the peak pressure should be limited to the upset limits in order to allow for more economical equipment design.
(4). Lonn-Term Shutdown Canha - Subsequent to an ATWS event, the reactor shall be brought a safe shutdown condition, and be cooled down and maintained in a cold shutdown condition.
These performance requirements are summarized in Table 15E.2-1.
4 Q
ABWR -
Standard Plant Table 15E.21 PERFORMANCE REQUIREMENTS Maximum RPV Peak Maximum Fuel Costalament P m sure FoolTemperature latarity P m sure ARI/RPT 105.5 kg/cm*g 97.2 C' Coolable 3.16 kg/cm g 2
Geometry FMCRD/RFT 105.5 kg/cm*g 97.2 C*
Coolable 3.16 kg/cm g _
2 Geometry 2
Boron /RPT 105.5 kg/cm 6 Containment Coolable
- 3.16 kg/cm*g Design Geometry Pressure O
l i
i
- $7.2 *Cpool temperature should not be reached befom the reactor reaches the hot shutdown condition.
ABWR Standard Plant 15E.3 ANALYSIS CONDITIONS Due to the extremely. low probability of the occurrence of an ATWS, nominal parameters and initial conditions have been used in this analysis and also in Section 15E.8, Reference
- 1. Tables 15E.31 and 15E.3 2 list the initial conditions and equipment performance characteristics, which are used in the
- analysis, r
J I
a i
'f
.i 1-4
+
\\
t t
i
)
i f
r e
e
+
-h
(
1
+
5 I
~......_u_-_.-_.,-.
....s.,-..~._..2..:_,._,
4 1
0
^8wa 1
Standard Plant Table 15E.3-1 INITIAL OPERATING CONDITIONS Parameter h
2 Dome Pressure (kg/cm g) 72.1 Core Flow (Mkg/hr)/(%NBR) 52.2/100 Vessel Diameter (m) 7.06 Number of Fuel Bundles 872 f
Power (MWt)/(%NBR) 3926/100 Steam / Feed Flow (kg/sec)/(%NBR) 2123/100-Feedwater Temperature ( C) 215.6 Void Reactivity Coefficient ( /%)
-9.7 Doppler Coefficient ( /C)
-0.504 i
ARI/FMCRD Reactivity Curve D Curve
' Suppression Pool Volume (mb/
3580/28,1 (Full NBR FW Flow-Min) -
l Initial Suppression Pool Temperature ( C) 37.7 Condensate Storage Temperature (C) -
48.9 i
. 4
?
r I
i
- e r-
.,.y.
-r,-
ABWR Standard Plant O
Table 15E.3 2
- EQUIPMENT PERFORMANCE CHARACTERISTICS Parameters Yalut 3.0 Nominal Closure 9me of MSIV (sec)
Relief Valm System Capacity (% NBR Steam Flow)/No. of Valves 913 at 1st setpoint/
18 Relief Valve Setpoint Range (kg/cm*g) 80.5/84.0 Relief Valve Opening Time (sec) 0.15 2
Pressure Drop Below Setpoint for Relief Valve Closure (kg/cm )
53 Relief Valve Closure Time Delay (sec) 0.9 Relief Valve Closure Time Constant (see) 0.2 RCIC Low Water LevelInitiation Setpoint Level 2 HPCF Low Water Level Initiation Setpoint
. Level 1.5
- HPCF Start Time (sec) 20 I
HPCF/RCIC H!,h Water Level Shutoff Setpoint Level 8 Number of HPCF Pumps 2
HPCF Flow Rate per Pump (kg/sec)/(% NBR Steam Flow) 50.4/237 RCIC Start Time (sec)
,2.29 RCIC Flow Rate (kg/sec)/(% NBR Steam Flow) 50.4/237 ATWS Dome Pressure Sensor Time Constant (sec) 0.5 ATWS Logic'%e Delay (sec) -
0.03 '
I HPCF and RCIC high levelshutofis imiependent of drywellpressure for ADYS mitigation.' Automatic reset is required so restart will autor satically occur iflent retums below the level setpoint. Manual action to control level in -
the nonnal range is perferred rather than automatic cycling between L8/L2 during the post hot shutdown phase of any A ni$ event.
The nominalflow wisus pressure head curve is used. Due value given for ABlVR is at S2.7 kgfem*g.
V
.. +.
-n.
ABWR 1*
Standard Plant Table 15E.3 2 EQUIPMENT PERFORMANCE CHARACTERISTICS (Continued)
Parameter h
Recirculation Pump System Inertia (Kg-m )
21.5 Delay before Start of Electro Hydraulic Rod Insertion 1.0/39.0 (with/without off site power)(sec)
Electro-Hydraulic Control Rod Insertion Time (sec).
'135 ARI Rod Insertion Time (sec) 25' RHR Pool Cooling Capacity (Kcal/sec/ C)/
265/1.57
(% NBR at 3& C AT).
Water Level Setpoint above Which RHR Pool Ca>1ing is Micwed Level 1 Serpoint for Iew Water Level Closure of MS' Level 1.5 Setpoint for 1.ow Steamline Pressure Closure of MSN (k;/cm g) -
52.7 Setpoint for Automatic Pool Ocding (C) 43.3 RHR Start Time (sec) 20 -
e F
.I e
g.
O t
-~,..._.,....A,.s.~
ABWR Standard Plant 15E.4 NIWS LOGIC AND SETPOINTS The mitigation of ATWS events is accomplished by multitude of equipment and procedures.
These include ARI, FMCRD run in, feedwater runback, RPT, recirculation runback, ADS inhibit, and SLCS. The logie of these ATWS mitigation is presented in Figures 15E.41.
The followings are the initiation signals and i
setpoint for the above response.
(1) ARI and FMCRD run-m 4
High pressure (1125 psig), or
-1
- level 2 i
(2) SLCS initiation
- High p' essure (1125 psig), and SRNM not downsca. : for 3 minutes, or -
- Level 2 and SRNM not downscale for 3 ~
minutes, or l
Manual ARl/FMCRD run-in signals and SRNM i
uot downscale for 3 minutes (3) RI'T (RIPS not connected to M/G set)'
- High pressure (1125 psig)
(4) RPr(RIPS connected to M/G set)
- Level 2 I
(5) Recirculation runback (10%/second)
- Any scram sipals, or j
- Any ARI/FMCRD run-in signals (6) Feedwater runback
- High pressure (1125 psig), and SRNM not downscale for 2 minutes (7) ADSinhibit f
- High pressure (1125 psig), and SRNM not downscale for 2 minutes, or.
Level 2 and SRNM not downscale for 25 seconds t
~
- Also tripped at Level 3,' which is not a part of ATWS mitigation
,( _
6, s
1.
\\
i
[
t
O
- 0*
CwANNE
i 7 ?-
.itoeacn 4P ASDi
_ IFg sr.A
- 7.* m a c< :
evpAss -
sem4u cou.ow C*~*"
/
A>p Asc.
'p p sr.s E
~1 t
=3-l*
2 __+,3
, >; y.b C c-,Au-
,F F sr.c h E f c scaAV W 3
>4
-b c..i.,,, p F sto
- a, g l[SF.B.>
2 SF. A -*
9ERCENCv
=2_
l p
g$ gg. C'e RFC CONTRCEER 1
r f
3,,e 4
m l
RC & IS I sko-*-
I " 0 s.c -
C ^" ***
2
~
SCRRi
< mr ap* -.
El A.e l iN EAC-N AC
- g,,,,,2f O, 7,,
.g C 7[ _ _I 205 1
'EA g
f fq 2
d
.NVEA*E4 C :-Pc ap' v 3 l
5 j,,A egg
- 1 mi I
b /+
LOGC BY =AACWa.:I lu e a ami c.uo. 4
_a
-1 a
m yog U@
n
~~
,- [, "7 SOUD-STATE lu nw ARi maaen e i$
2' "* M 2
- 1 CONTACTCA LOG.C.
to ASO A "Q
SSLC - ECCS
- T/
r
-* v7 0
( apy t v>, ? t }
C aov tvt 2 0 r
2 f
@)'/
= 1
- e"'ASD F 4
g, y, 3m>y$
~
1
= T*
CRPV Lvt 2 iv}
)
g m.RPr to Aso a E
2/+ pyrr l
SB & PC 5
3 m ao s
~
J $
2 c coue==s 3
/+ mPr
,k O1 to ASO E (DOME m a M)
$h e-RWT
=1 m ASD H g
h M 6 See Dy M
~ :::::
m ASD C J
g 2/ >I6 5ec 0tv M c em,. 3 C Na a L3 5
E
- OO
/
2 c mo,.,
-g
- is see oi, l>RWT
/
3 to Aso = -
Figure 15E.4-1. ATWS Mitigation Logic -
)
L
,.._..m
.-,.-,-.,,--.,,,,,,.e
,-,...y
, - - - = _. -
6 l
ABWR Standard Plant 15E.5 SELECTION OF EVENTS (3)1oss of Feedwater Based on conclusions from the evaluations for This imusient is less severe than the above two operating BWR plants as documented in Section eventt However, it is the only event which is miti ate. by ARI or FMCRD run.in initiated from 15E.8, Reference 1, the following limiting 6
eve ts were selected to demonstrate the the low level signals. Thus, this event is analyzed to performance of the ATWS capabilities. They are show that the low level trips are capable to mitigate grouped into three categories. The first the event, category includes events which demonstrate ATWS mitigation on the most severe and limiting (4) loss of Feedwater Heata cases. The second category has events which are generally less severe for ATWS analysis but This transient is very mild as the increase of neutron are analyzed to show the sensitivity of key flux never reaches the scram setpoint. The reactor ATWS parameters to these events. In each above shutdown is initiated by operator action. The main case, the recirculation pump trip, ARI, concer a that peak linear heat generation rate may electrical insertion of the control rod drives, exceed performance criteria when FMCRD run-in is boron injection and other ATWS mingation initiated. The analysis is to show that the actions are assumed to occur on the appropriate recirculation traback can mitigate this event.
signals. No operator action is assumed, unless specifically mentioned.- The third category
. Category 2. Moderate Impact Events covers the cases has only minor impact to the reactor vessel and containment. They are (5) Turbine Trip with Bypass Valves Open discussed briefly to support the assumption that they do not significantly influence the This transient usually produces higher neutron flux, design of ATWS mitigation. No analysis was
- heat flux, and vessel pressure than those from MSIV performed for event;in the third category, closure event due to the fast closure of the turbine stop valves. However, the availability of main condenser significantly reduces the amount of steam Category 1. Limitine Events discharged into the suppression pool.
(1) Main Steam Isolation Valve (MSIV) Closure (6) Loss of Condenser Vacuum Generic studies have shown that this transient The initial transient behavior of this event is similar produces high neutron flux, heat flux, vessel to that of turbine trip as the reduction of vacuum in pressure, and suppression pool temperature, the main condenser initiates turbine stop valves The maximum values from this event are, in most closure. When the isolation setpoint is reached, the cases, bounding of all events considered.
- MSIVs start to close. The event follows the pattern of MSIV closure in suppression pool temperature :.nd (2)1oss of Normal AC Power containment pressure.
This transient is less severe than the MSIV (7) Feedwater Controller Failure at Maximum closure in terms of vessel pressure, heat flux, Demud neutron flux, and suppression pool temperature. However, because the loss of This transient produces peak values of key power to the condensate and feedwater pumps parameters similar to those of turbine trip case. The causes the Icedwater flow to cease, very low availability of main condenser significantly reduces vessel water levels are expected. Thus,-the. the load of suppression pool frem steam discharge -
capability of the ECCS to recover the water from S/RVs.
level will be tested.
p ABWR Cl Standard Plant Category 3. Minimum Imnact Events (10) Inadvertent Opening of All Bypass Valves (8) Recirculation Flow Controller Failure at This event initiates a gradual decrease of the vessel Maximum Demand pressure and power. It is followed by a rapid rise of pressure and power after the closure of MSIV on low This transient is not severe enough to trip any steam line pressure. The characteristics of the ATWS logie nor initiate HPCF or RCIC flow. It remammg portion of this transient is very much the is considerably milder than the MSIV closure or same as the MSIV closure event except it starts at a turbine trip ATWS cases. This is a short term much lower initial power level. The steam discharged transient with a sudden power rise and into the containment is much less than that in the relatively small pressure increase. The entire MSIV closure event. The same conclusio i is also transient is over within 30 seconds by which true for other key parameters, time the reactor settles out to a new equilibrium condition of less than 100% rated (11) Shutdown Cooling (RHR) Malfunction -
power. Since the peak pressure stays below the Decreasing Temperature lowest S/RV setpoint, steam flow discharge to the suppression pool does not take place.
This event can only occur at very low pressure. The shutoff head of the shutdown cooling pumps is less Since the transient is not severe enough to than 300 psig. In this condition, the reactor has trip the ATWS logic or initiate HPCF or RCIC almost no voids in it and therefore only little if any flow as the feedwater and level control is positive reactivity is increased. Hence, this event is maintained. Manual ARI/FMCRD run-in has to noi considered further.
be initiated by operator in the case of manual
/]
scram fails. The success of ARI or FMCRD All transient analyses, unless otherwise specified,
(
/
run.in with recirculation runback can bring the were performed with the REDYA code. Other codes reaetor to hot sbutdown just like normal used in special analysis are ODYNA and PANACEA.
scram. If control rods fail to insertion after operator action, the boron injection would bring the reactor to below 1%.
(9) Startup of the Idle Recirculation Pump The abnormal startup of an idle recirculation pump requires the inverter to proside electric current much higher than the normal to counter the much higher reverse flow. This overcurrent requirement activates the overcurrent protection logic of the electric bus which supplies the power to the idle RIP. This electric bus is tripped by the protection logic. Consequently, the other RIPS powered by this electric bus are also tripped. Therefore, this event 's similar to the trip of three recirculation pump event. Since the scram is never initiated and there is.no steam discharged into the suppression pool, there is no impact to the ATWS mitigation design.
l Therefore, further transient specific enalyses have not been done.
f
[ f\\
b l
/
(
Standard Plant 15E.6 TRANSIENT RESPONSES the FMCRDs. The insertion of the control rods is successful in bringing the reactor to hot shutdon.
For every event selected for analysis, three Peak values of key parameters are shown in Table cases were analyzed. The first one shows the 15E.6.11 for the ARI case and Table 15E.6.1-2 for ATWS performance with ARI. This case is the FMCRD run-in case. In the case that control intended to show the effectiveness of the ARI rods fail to insert, the reactor will be brought to hot design. The second.ase, which uses FMCRD shutdown by automatic boron injection in about 19.4 run in, assuming a total failure of ARI, was aninutes from the beginning of the event. The performed to show the backup capability of transient behavior of this case is listed in Table FMCRD run in. The third case was analyzeo to 15E.6.13. The reactor system response is presented show the in-depth ATWS mitigation capability of by Figures 15E.6.1-1 to 15E.6.1-4 for ARI actinted, the ABWR, in this case, both ARI and FMCRD Figures 15E.6.1-5 to 15E.6.1-8 for FMCRD run-in run in are assumed to fail. Automatic boron case and Figures 15E.6.19 to 15E.6.1-12 as SLCS injection with a 180 seconds delay, are relied operating, respectively. The normalized axial powei upon to mitigate the transient event.
shape change during FMCRD run-in we presented in Figure 15E.6.113. The increase of the local power if the ARI and FMCRD run in fail at the same density does not violate the performlace criteria.
time, which has extremely low probability of occurrence, the peak reactor would still be 15E.6.2 Loss of AC Power controlled by the Recirculation runback and relief valves. However, the nuclear shutdown in this event, all scram signal paths, including valve wi1I thea re1y on tbc automatie SLCS position, high flux, high pressure, low level, and all injection. The boron would reach the core 60 manual attempts have been assemed to fil.
p seconds after the initiation. The operation of Q
both SLCS pumps generate a 100 gpm volumetric The loss of AC power has the fdlowing effects:
flow rate of sodium pentaborate. The nuclear shutdown would begin when boron reaches the (1) An immediate load rejection will occur. This will core.
cause the turbine control valves to close.
15E.6.1 Main Steam Isolation Valve Closure (2) As a result of the lead rejection, four of the ten recirculation pumps will trip.
This transient is considered an initiating event caused by either operator action or (3) Due to the loss of power to the condensate
- instrument failure. Scram signal paths that pumps, feedwater will be lost.
are assumed to fail include valve position, high neutron flux, high vessel pressure, and (4) The reactor will be isolated after loss of main all manual attempts. A short time after the, condenser v acuum.
MSIVs have closed completely, the ATWS high pressure setpoint is reached, which initiates Figures 15E.6.21 to 15E.6.2-4 show the transient four of the ten recirculation pumps to trip and behavior under ARI activation, Figures 15E.6.2-5 to the rest start to runback. The combined effect 15E.6.2-8 for FMCRD run-in and Figures 15E.6.2-9 of the trip and runback reduces the core flow to 15E.6.2-12 for automatic SLCS, respectively, and increases core voids, thereby reducing power generation which limits pressure increase The fast closure of the turbine control valves causes a and steam discharge to the supprpssion pool.
rapid increase of pressure, and the ATWS high The ATWS high pressure signal causes the pressure setpoint is reached shortly after the control actuation of ARI and the electric insertion of valves have closed. Because the four pumps have already tripped at this time on the load rejection F
-(
_ Standard Plant signal, only six remaining pump will start pumps. At about 22 seconds, low water (L2) is also runback. The ATWS high pressure signal reached. This trips remaining recirculation pumps, initiates the rod insertion. The rod activates ARI, FMCRD run-in, starts SLCS clock, insertions are successful in bringing the -
and initiated RCIC. Successfulinsertion of control reactor to hot shutdown. If both modes of rod rods brmgs the reactor to hot shutdown. Failure of I
insertion fail, the ATWS high pressure signal rod insertion will initiate SLCS upon the timer run up also initiates the timer for SLCS. After while SRN'4 signal is not downscale. AT about 16.9 confirming the rod insertion failure by minutes the reactor becomes hot shut down as the
+
monitoring the high pressure and SRNM boron concentration reaches sufficient value. Tables not downscale signal for 3 minutes, the SLCS 15E.63-1 to 15E.63-3 show the summary of peak will be initiated. At 16.9 minutes, the values of key parameters for the three cases.
reactor is brought to hot shutdown when enough boron concentration is built up in the reactor core.
15E.6.4 Loss of Feedwater Heater Tables 15E.6.2-1 to 15E.6.2-3 show the summary This transient does not trip any automatic ATWS of peak values of key parameters for the three logic. ARI,- FMCRD run-in, and SLCS timer are events.
assumed to be initiated by operator at about 10 minutes after the beginning of this event. At this 15E.63 Loss of Feedwater time, the reactor has settled in a new steady state at a higher power level. There is no steam discharge to This event does not have rapid excursions as in the suppression pool because of the relatively low some of the other events but is a'long-term vessel pressure. Figures 15Ei, &1 to 15E.6.4-4 show power reduction and depressurization. Since the transient behavior for ARI, Figures 15E.6.4-5 to
,_/
the pressure begins to fall at the onset of the 15E.6.4-8 for FMCRD run-in and Figures 15E.6.4-9 transient, the need for relief valves does not to 15E.6.4-12 for SLCS case, respectively. IJpon the arise until isolation occurs very late in the failure of rod insertion, SLCS can bring the reactor to event and only single valve cycling is expected hot shutdown at about 333 minutes, to handle decay heat. The containment limits are not approached.
The mild nature of this transient forestalls any i
significant peak values for the key parameters '
In this event all feedwater flow'is assumed to normally associated with ATWS study. However, the be lost in about five seconds.- Figures slow insertion rate of FMCRD run-in allows the l
15E.6.31 to 15E.6.3-4 show the transient reactor to re-establish quasi-steady axial power shape.
I' behavior for ARI activated. Figures 15E.63-5 The peak value of these new profiles, which were to 15E.63-S represent FMCRD run-in event. The calculated by the PANACEA code, are shown in-mitigation of this event by SLCS is illustrated Figure 15E.6.4-13. The peak cladding temperature in Figures 15E.63-9 to 15E.63-12.
does not exceed the coolable geometry criteria Figure 15E.6.4-14 presents the normalized axial After the loss of feedwater has taken place, power shape change.during the event. -Table the pressure,' water level and neutron flux 15E.6.41 shows the peak values of the key begin to fall. Around 6.5 seconds low water -
parameters for FMCRD run-in case. Ihe same (L3) is reached. This trips four recirculation values apply to ARI and SLCS cases as well.
i i
f ABWR Standard Plant 150.6.5 Turbine Trip with Bypass concerned and similar to the MSIV riosure case with respect to suppressic:: t> col temperature and The initial characteristics of this transient pressure. Fir D *.s 15E.6.61 to 15E.6.6-4 show the are much like the MSIV ciente described in transient beha,ar fot ARi event, Figures 15E.6.6 5 to Section 15E.6.1 with a rapid steam shutoff.
15E.6.6 8 for FMCRD run in case and Figures Pressure and power increases which are limited 15E.6.6 9 to 15E.6.C 12 for SLCS condition, by the action of the relief valves and respectively. The high pressure ATWS setpoint is RPT/ recirculation runback. As this event reached shortly after the closure of turbine stop progresses, however, the availability of the valves. The high pressure initiates trip for four of the main condenser makes it possible for the relief ten RIPS and runback of the other Jr. It starts AR1, valses to be closed after about 48 seconds.
FMCRD run in and iiLCS timer. A successful This terminates the steam discharge to the insertion of control rods brings the reactor to hot
{
suppression pool. Figures 15E.6.51 to 15E.6 4 shutdown. Otherwise, the injection of boron is show the transient behavior for ARI, Figure:
initiated t,pon SR?iM not downscale and high 15E.6.5 5 to 15E.6 8 for FMCRD run-in and pressure signals. As the poison reaches sufficient Figures 15E.6.5 9 to 15E.612 for SLCS cases, concentration in the core, the reactor achieves hot respectively.
shutdown in about 19.1 minutes. Tables 15E.6.6-1 to 15E.6.6 3 show the summary of peak values of key The closur he turbine stop valves causes a parameters for these events.
rapid inci of pressure, the ATWS high pressure sei, at is reached shortly after the 15E.6.7 Feedwater Contr aller Failure closure. The 1.igh presst re initiates four of the recirculation pumps to trip and the rest to The initial portion of this transient results in a O
start runback, initiates ARI, FMCRD run in and gradual p;wer increase, then a sharp pressure rise SLCS timer. Upon successful insertion of ti.:
and power peak as the turbine stop valves close at control rods, the reactor achieves hot high water level. The long term segment of this shutdown. If the rods fail to insert into the transient is similar to tha.'. of turbine trip with bypass core, the SLCS will be initiated by SRNM valves operating. The discharge of steam into the not-downscale and the high pressure signal when suppression pool is minimbed by the availability of the timer run up. In this case, the hot the main condenser and turbine bypass valves.
shutdown b " cached at about 19 minutes.
Figures 15E.6.71 to 15E.6.7-4 show the transient Tables 15E4 61 to 15E.6.5-3 show the summary behavier for AR1, Figures 15E.6 7 5 to 15E.6.7-8 in of peak values of key parameters for these FMCRD run-in and Figures 15E.6.7 9 to 15E.6.712 events.
for SLCS case, respectively.
15E.6.6 less of Condenser Vacuum The closure of the turbine stop valves starts a rapid increase of pressurt. The ATWS high pressure This transient s' arts with a turbine trip setpoint is reached shortly after the valve closure.
because of the low condenser vacuum, therefore, The high pressure trips four of the ten recirculation the bef nning is the same as the turbine trip pumps and starts runback of the othes six, initiates i
event (see section 15E5.4). Houver, the MSIVs AR1, FMCRD run in, and SLCS timer, The reactor and turbine bypass valves also close after the reaches hot sh".tdown once the control rods complete condenser vacuun has further dropped to their the insertion into the core. If the rod insertion fails, closure setpoints, relief valves cycling the initiatloc of SLCS is confirmed by SRNM increases considerably compared tg the original not. downscale and the hot shutdow? is achieved at turbine trip case. Hence, this event is about 20 minutes. Tables 15E.6.71 to 15E.6.7-3 similar to the turbine trip event as far as the show the summary of peak values of key parameters peak power med g.ressure characteristic are for tbesc events.
'O V
z
O ABWR Standard Plant s
Table 15E.6.1 1 MSIV CLOSURE
SUMMARY
(ARI) l't!n
.Ilsa Mazimum Neutron Mux (%)
451 1.7 see Madmum Vessel Bottom Preuvre (kg/cm g) 91.3 4.6 see Madmum Average Heat Hus (%)
131 3.0 see Mazimum Bulk Suppression Pool T.'mperature C C) :
59.9 303 min Assc. dated Co.tainment Pressure (kg/cm g) 0.24 303 min Peak Claddmg Temperature (C) 613 17.9 see s
e Q
ABWR Standard Plant Table 13E.6.12 MSIV CLOSURE
SUMMARY
(FMCRD RUN.IN) -
Talut.
.Bar.
Mazimum Neutron Flux (%)
451 1.7 see 2
Maximum Veuel Bottom Preuure (kgn2 3) 913 4.6 see Maximum Average Heat Flux (%)
131 3.0 see Maximum Bulk Suppreulon Pool Temperature C C) 65E 148 min Anociated Containment Preuure (kg/cm g) 032 148 min Peak Cladding Temperature (C) 536 8.5 see O
O
.g O
.______2__
i i
ABWR Standard Plant l
s t
Table 15E.6.13 MSIV CLOSURE
SUMMARY
(BORON INJECTION)
Xalat.
.Ilmt.
Maximum Neutron Flux (%)
451 1.7 sec Maximum Veuel Bottom Preuure (kg/cm g) 91 3 -
4.6 see Maximum Average Heat Flux (%)
131 3.0 see Maximum Bulk Suppreuion Pool Temperature (C) 81.6 33.4 min 2
Auociated Containment Preuure (kg/cm g) 0.63 33.4 min Peak Cladding Temperature (C) 697 140.0 sec i
i O
l k
F h
i I
L
?
e e
l i
I Y
t
_,.-n,--
-.,---v--r-
+-.#
.s-.-.-
m-%,
y
..w--
,%em.. ~,,
,w
- v. y g
-.me-y e.y',
m.-x.-
c.+,w..
ABWR Standard Plant Table 15E.6.2.;
LOSS OF AC POWER
SUMMARY
(ARI) i
.Yalue.
.' Dant.
f Maximum Neutron Flux (%)-
170 0.69 see 2
Maximum Vessel Bottom Pressure (kg/cm g) -
84.9 3.0 see l
Maximum Average Heat Flux (%)
102
~ 0.89 see 58.5 351 min Maximum Bulk Suppreuion Pool Temperature (C) _
Anoeinted Conta' ment Pressure (kg/cm g) 0.22 351 min m
i 1.
1 O
ABWR Standard Plant Table 15E.6.2 2 I
LOSS OF AC POWER
SUMMARY
(FMCRD RUN IN) falar.
11mt.
Mazimum Neutron Mux (%)
170 0.69 see Maximum VesselBottom Pressure (kg/cm g) 84.9 3.0 see Maximum Average Heat Rux (%)
102 0.89 sec Maximum Bulk Suppression Pool Temperature ( C) 59.2 --
325 min Associated Containment Pressure (kg/cm g) 0.23 325 min O
4 O
p
,a
O ABWR Standard Plant Table 15E.6.2 3 LOSS OF AC POWER
SUMMARY
(BORON INJECTION)
Value 3 hat.
Maximum Neutron Flux (%)
453 371 see Maximum Vessel Bottom Preuure (kg/cm g) 84.9 3.0see Mavimum Average Heat Flux (%)
102 0.89 see Mazimuru Bulk Suppression Pool Temperature ( C) 65 163 min 2
Anociated Containment Picssure (kg/cm g) 0.31 163 min I
O f
e
^.
'i
O ABWR Standard Plant Table iSE.6.31 LOSS OF FEEDWATER
SUMMARY
(ARI)
Yalne
.Tunt Maximum Neutron Flux (%)
116 424 see 2
Maximum Vessel Bottom Pressure (kg/cm 3) 75.4 430 sec Maximum Average Heat Flus (%)
116.
430see Maximum Bulk Suppression Pool Temperature (C) 58.0-384 min Associated Containment Pressure (kg/cm g) 0.22 384 a
O
1 t
ABWR Standard Plant i
1 Table 15E.6.3 2 f
LOSS OF FEEDWATER
SUMMARY
(FMCRD RUN.IN)
L XalR
.'DIK f
Maximum Neutron Flux (%)
116 424see 3
Maximum Vessel Bottom Pressure (kg/cm'g)
-75.4
.430 see Maximum Average Heat Flux (%)
116 430 sec j
t Maximum Bulk Suppression Pon! Temperature ( C) 57.9 383 min f
2 Associated Containment Pressure (kg/cm g) 0.22 383 min l
I
- r I
1 r
b I
i L
i i
i
~~
- i h
?
i f
5 L
f G
9 I
r i
1
' I s
1 r
.-r--
+.--~~,,..,n,,+-
,,n.,v,.w-,.,,,,-.~-a,,w-
.w,,
,.,..~,v,
+,,,,,,,,..,..
O ABWR Standard Plant Table 15E.63 3 LOSS OF FEEDWATER SUMhMRY (BORON INJECTION)
Value 21st.
Maximum Neutron Max (%)
116 424 see Maximum Vessel Bottom Pressure (kg/cm g) 75.4 430 see Maximum Average Heat Rux(%)
116 430 see Maximum Bulk Suppres.lon Pool Tempera 6are (C) 63.1 212 min 2
Associated Containment Pressure (kg/cm g) 0.28 212 min O
____m.
_ _. _. - = _ _ -
ABWR O
Standard Plant Table 15E.6,41 LOSS ONE FEEDWATER HEATER
SUMMARY
(FMCRD RUN.IN)
Xalut.
Jims.
Ma unum Neutron Flux (%)
116 424 see 2
Maximum Veuct Bottom Preuure (kg/cm g) 75.4 430 see Maximum Average Heat Flux (%)
116 430 see Maximum Bulk Suppreuion Pool Temperature' (C) 2 Anociated Containment Preuure (kg/cm g) o o
O
" taitial values 9
6
_.___a,__--___
-,a,-----w_
ABWR O
Sandard Plant Table 15E.6.51 TURBINE TRIP WITH BYPASS
SUMMARY
(ARI)
.Yala
. Dan.
Maximum Neutton Flux (%)
757 0.79 sec 2
Maximum Vesel Bottom Preuure (kg/cm g) 87.1-2.46 see Maximum Average Heat Flux (%)
126.5 1.10 see MaxLnum Bulk Suppreulon Pool Temperature ('C) 34.4 33 see Auociated Containment Pressure (kg/cm g) 0.01 33 see O
t O
_ _ _. _ _ _, 1 l
f ABWR Standard Plant Table 15E.6.5 2 TURBINE TRIP WITH BYPASS
SUMMARY
(FMCRD RUN.IN)
Yalu
.Ilm.
1 Maximum Neutton Flux (%)
757 0.79 sec 2
Maximum Veuel Bottom Preuure (kg/cm g) 87.1 2.46 see 1
f
- Maximum Average Heat Flux (%)
126.5 1.10 see Maximum Bulk Suppreulon Pool Temperature ( C) 34.7 90Sec 2
Associated Containment Preuure (kg/cm g) 0.02 90 sec O
f a
9 l
e l O 4
--. -.-----.. ~ -..,...,... -,.,.,,, -
--ry.-~.-
mv.---,
-+s.,
y----
y-
- ~, *
,m--
.~e.--
l h
ABWR Standard Plant l
Table 15E.6.5 3 i
l l
TURBINE TRIP WITH BYPASS
SUMMARY
(BORON INJECTION) i Value 3 hat Maximum Neutron Flux (%)
757 0.79 sec Maximum Vessel Bottom Pressure (kg/cm g) 87.1 2.46 see Maximum Average Heat Flux (%)
126.5 1.10 see Maximum Bulk Suppression Pool Temperature (C) 42.1 12 min Associated Containment Pressure (kg/cm g) 0.07 12 min O
W e
P u-
+3--eg M
nua
'-5dWW-w
.F g_
J-A y4 ig-4
-*Tr
'y>tt
=
p-Pra
-wur
+-e-..
waw-wwr vA-
+
n
ABWR Standard Plant Table 15E.6.61
!hSS OF CONDENSER VACUUM
SUMMARY
(ARI)
Jala
. time.
Maximum Neutron Flux (%)
757 0.79 see Maximum Veuei Bottom Preuure (kg/cm g) 87.1
' 2.46 see Maximum Average Heat Flux (%)
127 1.10 sec r
Maximum Bulk Suppreuion Pool Temperature (C) 59.4 316 min Auociated Containment PrM.ie (kg/cm g) 0.24 316 min O
O
l' ABWR Standard Plant 3
Table 15E.6.6 2 I
SUMMARY
(FMCRD RUN.IN) fala
. Bat.
l Mazimum Neutron Mux (%)
757 0.79 see 2
Maximum Vessel Bottom Pressure (kg/cm g) 87.1' 2.46 sec
?
Maximum Average Heat Mux (%)
127 1.10 sec l
Maximum Bulk Suppression Pool Temperature (C).
60.7 282 min Associated Containment Preswre (kg/cm g) e 25 282 min i
i f
l.
1 s
l i
i t
l.
1 l-r e
4 Y
O L
n
.-*__-.-*.-___.,,__m
,e f,w,-
.eg.a
._.v,,y,pp,,,.gp,p.
,,pg,_
O inwa jitandard Plant Table 15E.6.6 3 LOSS OF CONDENSER VACUUM
SUMMARY
(BORON INJECTION)
Yalue
- 31st, Maximum Neutron Flux (%)
757 0.79 see 2
Maximum Vessel Bottom Preuure (kg/cm g) 87.1 2.46 tec Maximum Average Heat Flux (%)
127 1.10see Maximum Bulk Suppreulon Pool Temperature C C) 80.1 49.$ min Anociated Containment Preuure (kg/cm g) 0.59 49.5 min O
t
ABWR Standard Plant Table 15E.6.71 FEEDWATER CONTROLLER FAILURE
SUMMARY
(ARI)
.Yalut 2 hat Maximum Neutron Flux (%)
647 19.9 see Maximum Vessel Bottom Pressure (kg/cm g).
87.0 21.6 see Maximum Average Heat Flux (%)
127.7 203 see Maximum Bulk Suppression PoolTemperatt e (C) 34.6 48 see Anociated Containment Pressure (kg/cm g) 0.01 48 see i
O e
4 0
.--__.--.-__-_-_-____m_.._m.__.
ABWR Standard Plant Table 15E.6.7 2 FEEDWATER CONTROLLER FAILURE
SUMMARY
(FMCRD RUN IN)
YAtR 31st.
Mazimum Neutron Flux (%)
647 19.9 sec 2
Maximum Vessel Bottom Pressure (kg/cm g) 87.0 21.6 sec Maximum Average Heat Flux (%)
127.7 203 see Maximum Bulk Suppression Pool Temperature (C) 34.6 60 see Associated Containment Pressure (kg/cm g) 0.01 60sec O
i O
48wn Standard Plant Table 15E.6.7 3 FEEDWATER CONTROLLER FAILURE
SUMMARY
(BORON INJECTION)
Islur.
Ilmt j
Maximum Neutron Flux (%)
647 19.9 see 2
Maximum Veuel Bottom Pressure (k /cm g) 87.0 21.6 see F
Maximum Average Heat Flux (%) -
127.7 203 see Maximum Bulk Suppression Pool Temperature ( C) 34.8 48 see Auociated Containment Pressure (kg/cm g) 0.02 48 see O
B l
h
't
_. _. -.. _.. ~..
i O~
~
O O
Figure 15E.6.1-1. ABWR MSIV Closure, ARI l
AT3 REDYA02V ABWR ' SSAR ATWS MSIVC SRI-RPT 13
- 1. 4 DOME PRES:iURE
= 10' I
m
- 1. 2 i
O_
i W
i Dw W
W i
Q' 1.
Q_
~
t
~.
I'
- 0. 8 O.
50.
100.
150.
200.
r,. or-rer o\\
n.
n.c..
i.-
i i
Figure 15E.6.1-2. ABWR MSIV Closure, ARI i
i.
AT3 REDYA02V ABWR SSAR ' ATWS MSIVC SRI-RPT 14 i
. 150.
- NEUTRON Ft UX (7.)
i 1
2 AVE SURFACE HEAT FLUX (7.)
i FLOW ((7.) /LB) 2 CORE INLET i
INLET SUBC00 LING BTU 4-2 Y
1 i
l I
r 100.
2 9
5
)
i i
2 4
y N.
x L
50.
\\
\\
/
l I
7 i
c c
s
\\_2 2
i-O.
~ ~ ~ - - '
50
.100.
.150.
200.
0.
.~
r~-m
?
,..-,,--,-.--4
, +, - ~.. -.,,. - - -
.-,c,-
~<
O-O O
Figure 15E.6.1-3. ABWR MSIV Cosure, ARI l
AT3 REDYA02V ABWR SSAR ATWS f'
MSIVC SRI-RPT 15 g
Nx 2
[ EVE
'.[Er SEP-SKIRT)
ITY (($)) (FT)
D LEVEL 2
ENSE l
NET REACTIV flTY $
ROD REACTP
~
q i
h.
O.
l 3
-10.
4 3
-20.
O.
50.
100.
150.
200.
r,.. r-ter or n....
O O
O I
i Figure 15E.6.1-4. ABWR MSIV Closure. ARI i
1 i
AT3 REDYA02V ABWR : SSAR ATWS MSIVC SRI-RPT 16 120.
-[
FEEDWATER FLOW t
2 REllEF VALVE FLOW 3 VESSEL STEAM FLOW.
I ECCS SCALED FLOW
,3 i
s 1
I 1
1 f
5' I
~.
~ 80.
j.
g i
5 L2 I
.il '
I i'
F-i \\f
<C Q'
ig
. k 5
O.
r' M.
g 5
r 40.- 'y 3
j r
O-
-_J
' L.L.
.l t
u,.
r j
2' 3
i
.' h h*3 s.
a a
n n s c
.I' O.
50 100..
'150.
200.
O.
.s y
,,y,
.,.,p._y._.
4
O O
O Figure 15E.6.1-5. ABWR MSIV Closure, FMCRD Run-in p
i AT3 REDYA02V l
i t
ABWR SSAR ATWS MSIVC FMCRD-RPT
- 1 l
- 1. 4
= 10' DOME PRESSURE j
s t
i.
A
[
<(
l
- g
- 1. e
}
O_
i
~
LJ i
i W.
.D
.1 v)
(A La.
M 1.
i Q_
i
~
N i
[
- 0. 8 O.
O. 5 1.
- 1. 5
- 2. = 10' n.
n,,.,
r.
e r (crn\\
(-
< - ~,
%)
%)
Figure 15E6.1-6. ABWR MSIV Closure, FMCRD Run-in AT3 REDYA02V ABWR SSAR ATWS MSIVC FMCRD-RPT #2 150.
NEUTRON Ft UX (%)
2 AVE SURFACE HEAT FLUX (7.)
FLOW ((7.)
3 CORE INLET 00 LING BTU /LB)
INLET SUBC 1
i 100.
4 1
I 3
3 3
3 1
L
0 O.
O. 5 1.
- 1. 5
- 2. = 10'
/nrn%
O O
O Figure 15E.6.1-7. ABWR MSIV Closure, FMCRD Run-in AT3 REL"A02V ABWR SSAS ATWS I
MSIVC
-RPT #3 f
b FT.REF SEP-SKIRL)
^
r lTY (($)) (FT)
) LEVEL a W ENSEI f
U 2
Nr 1(EAC \\
D R
14 flTY $
p
/
2 f
5 L
\\1 \\jn
\\
3
-10.
~
s
~.
1
-20.-
O.
O. 5 1.
- 1. 5 2.=10' T... r-rcr ns r.
O O
O
~
Figure 15E.6.1-8. ABWR MSIV Closure, FMCRD Run-in AT3 REDYA02V ABWR SSAR A.TWS MSIVC Fh/CRD-RPT #4 120.
FEEDWATER FLOW i
2 REllEF VALVF FLOW 2 VESSEL STEnM FLOW ECCS SCALED FLOW a
3, 1
l 80.
g L2J F-l l
<C (r
LL O
M 40.
3 O
1 Li_
3 3
2 3 2 3 2
3 2 3 2
' ' 'I ' ' '
I 0.
O.
O. 5 1.
- 1. 5
- 2. = 10'
U L-)
5-f~s v
Figure 15E.6.1-9. ABWR MSIV Closure, SLCS AT3 REDYA02V ABWR SSAR ATWS MSIVC SLCS-RPT f l
-f
- 1. 4 DOME PRESSURE 3
i x10 i
I
~
If,in m
(
G
- 1. 2 i
i g
i s
i i,.
v khh f
bl U
vL Z
D i
m i
m W
i 1
1.
y O_
l k
A O.8111'b -
O.
O. 5 1.
- 1. 5
- 2.
- 10' r,, a r-rc r c.\\
c.
K-O-
~
O O
Figure 15E.6.1-10. ABWR MSIV Closure, SLCS 1
AT3 REDYA02V l
ABWR SSAR ATWS MSIVC SLCS-RPT f 2 150.
NEUTRON Fl UX (7.)
i i
2 AVE SURFACE HEAT FLUX (7.)
s CORE INLET FLOW (7.)
i INLET SUBC00 LING (BTU /LB) 4 l
100.
i l
i 50.
I' 1
k l
kI s
0'. ' ' '
~
O O. 5 1.
- 1. S 2.=10'
,.,r
/croi
-..,m,,c ie
( )n
)-
~
Figure 15E.6.1-11. ABWR MSIV Closure, SLC AT3 REDYA02V l
MSIVC SLCS-RF,h 3 10.
3, D LEVEL (FT))
LEVEL (FT REF SEP-SKIRT i
2 W'R SENSE 1TY (($))
s NET REACIl\\
YlTY $
l ROD REACTI 4
3 a
4 3 4 8
4 4
4 g7
%j. '3
(
5 3
3 y I
%v'
-10.
'l'
-20.
O.
O. 5 1.
- 1. 5
- 2. > 10'
,. a r-tcr nr y gg.,
O-O-
O l
Figure 15E.6.1-12. ABWR MSIV Closure, SLCS i
MSIVC SLCS-RPT A 4 120.
FEEDWATER FLOW i
2 REllEF VALVE FLOW 2 VESSEL STEJ\\M FLOW ECCS SCALED FLOW 3,
l l
80.
O L_1 1-
<C T
1 i
O M
40.
3
's O
_J LL
^
W
,q I
O.
I'
0.
O. 5 1.
- 1. 5
- 2. = 10'
,.,._s
ea4-
- n]
5 E;:Q k
7.// T'\\\\..~
S~ z
/
4 e
s.\\
/
>/
\\_
E m
i u
p/q' C
i+
t 5
=, s 1
).-
y M
O h
6 2
rg
=
$ 3
[
Q o
n' p
2 E&
3., 8$-
[
5,
\\a
+-
r l
l I
g%
4>
O s
og N]%l W[
d5 1
z>
n m
i a$
~
N
~
8 sq w 7\\
m a.
e
<g b.
No O
p' 5
- b. -
N L
'a
+
1N '.N ci
<su I
l's \\!,
s x
e a
t~
e o
1 x
=
g o
y e
q f
Q o
-4 b.
t NV e
\\x
\\..
~
l
'o J L o..yt m.
m m
e c
4
A
}
Figure 15E.6.2-1. ABWR Loss of AC Power, ARI AT3 REDYA02V ABWR SSAR ATWS LOSS AC-ARI g1
- 1. 4
= 10'
- DOME PRESSURE O
e g
- 1. 2 R
i v
t
,4 g
y
]
3 m
M LJ 1
1.
'g
- a N
aud>
e
.dle
'I'
- 0. 8 O.
200.
400.
600.
800.
T... r-
/ c-r ei c...
n,,n
..... =.
~
o o
o
- Figure 15E.6.2-2. ABWR IAss of AC Power, ARI AT3 REDYA02V ABWR SSAR ATWS
~
LOSS AC-ARI #2 150.'
NEUTRON F1 UX (%)
i 2 AVE SURFACE HEAT FLUX (%)
a CORE INLET FLOW (
- INLET SUBC00 LING TU/LB)'
t I
.100.
2 1
' 50.
2 4
l S 4 3
4 I
E 4
c
A O
O O
Figurr i 'E.6.2-3. ABWR Loss of AC Power, ARI AT3 REDYA02V ABWR 'SSAR ATWS LOSS AC-ARI #3 l
10~
W R S(FT.REF SEP-SKIRT)
LEVEL i
2 ENSED LEVEL (FT)-
2 s NET' REACTIVITY ($)
ROD REACTIVITY ($)
2 2
2.
s 2 2
f y m2 g,
0.
\\___,
e
.)
~
3
-10.
e e
W 6-
' ' ' I ' ' '
-20.-
600.
800.
- 0.'.
200.
-400.
~...,,;,,,,
r...,-
rcr ns -
O
~
~
O O
Figure 15E.6.2-4. ABWR Loss of AC Power, ARI AT3 REDYA02V
~
LOSS AC-ARI g4
- 120.
i-FEEDWATER - FLOW 2 RELIEF VALVE FLOW 2 VESSEL STEAM FLOW ECCS SCALED FLOW' s
s P
O. 80; td '
s
- t l--
i
<C i
t cc.
.g-r,
.O 3
W t
40.
I-
.3 b
O 1r-
.g i.;
Lt '
q e
4 f.
--/
a a
s s
1'
O.
600.
O, 200.
400.
800.
w e
v
d
O
~
O O
Figure 15E.6.2-5, ABWR Loss of AC Power, FMCRD Run-in AT3 REDYA02V ABWR '. 3AR ' ATWS LOSS AC-FMCRD # 1
- 1. 4
=10 i DOME PRES:iURE 5
4 e
m I
. g
.1.' 2 l
Q_1 v
t t
t t
. y-
- g' i
D:
w l;
r i
g
. g.
- 3.
~
/
- a.
q, q.
e 6
m I
f 0. 8 O.
200..
400 600.
1800.
i,. - a
..s
O O
O
~
Figure 15E.6.2-6. ABWR Loss of AC Power, FMCRD Run-in AT3 REDYA02V ABWR SSAR ATWS LOSS AC-FMCRD #2 150.
NEUTRON F1 OX i
2 AVE ' SURFACEI FLUX (7.)
3 CORE INLET FLO INLET SUBC00 LING U/LB) 4 l-l I
e 100.
i i
'l
- e t
. O.
t 8
4 4
i l 5
4 4
4 3
4 I
1
- O.
200..
400.
63 800.
O-O O
~
Figure 15E.6.2-7. ABWR Loss of AC Power, FMCRD Run-in AT3 REDYA02V ABWR SSAR ATWS LOSSAC-hCRD# 3 10.
D LEVEL (FT))
i LEVEL (FT,REF SEP-SKIRT 2 W R SENSE QTY (($))
ITY $
s NET REACTiV ROD. REACil' jl 1
8 _M.
,3_,,-
v 0..
O i
/,
y
?
I
-10.
4 l
-20.
t' 600..
800.
O 200.
400.
/c r n \\ -
v.
-sr-
d Os O
O Figure 15E.6.2-8. ABWR Loss of AC Power, FMCRD Run-in AT3 REDYA02V ABWR SSAR ATWS LOSS AC-FMCRD #4 120.
FEEDWATER FLOW i
2 RELIEF VALVE-FLOW 3 VESSEL STEAM - FLOW ECCS SCALED FLOW l-80.
g W-3 F-4O' r
u-O
~
5 p
g 40.
. O -
l2-r.
E
- k..
I 4
?
.W
./
2 2
2 2
~
34{
f 3
3 4 3 i2 41 23
- 81 3
24 4 3
4 3 600.
'800
. 0.
200.
'400..
. < - r-e
%.,s
Q~
O-O
~
Figure 15E.6.2-9. ABWR Loss of AC Power, SLCS AT3 REDYA02V ABWR' SSAR ATWS LOSS AC-SLCS #1
- 1. 4 DOME PRESSURE x10' G
21.2 i
i g
'i Q.
if,,
f bf l
L e
a l
W i
y)
Ld CE 1.'
j 1
i
~
i r
g.
9 Ws 0.8
' ' '
- I ' '
O.
0.- 5 1.
- 1. 5 2.=10'
,.x me
O O
O Figure 15E.6.2-10. ABWR Loss of AC Power, SLCS r
AT3 REDYA02V ABWR SSAR ATWS LOSS AC-SLLS #2
~
i NEUTRON Fl.UX s AVE SURFACEHA FLUX (%)
FLOW ((%) /LB) 2 CORE INLET 00 LING BTU INLET SUBC 4
~
100.
l
~
i l
l i
b.
J l k
l f'
2
' ~
O. '
0.
O. 5 1.
- 1. 5
- 2. = 10'
)
Figure 15E.6.2-11. ABWR Loss of AC Power, SLCS AT3 RIDYA02V i
}
LOSS AC-,5 pS '#3 10 LEVEL (FT.REF:SEP L KIR 2
i 2
2 W R SENSED LEVEL (FT 2 NET REACTIVITY 2
ROD REACTMTY I
,4 ts (f 2At
-4 1g*/
N ( f"'
s>
/
~"
j l'
-10.
l
'I
I
-20.
0.
O. 5 -
1.
- 1. 5
. 2. = 10'
~
,i
f'%
(m%A O
f%
_O I
Figure 15E.6.2-12. ABWR Loss of AC Power, SLCS AT3 REDYA02V ABWR SSAR ATWS LOSS AC-SLCS #4 120.
FEEDWATER FLOW i
2 RELIEF VALVE FLOW 2 VESSEL STEAM FLOW ECCS SCALED FLOW 4
i i
6 80.
O LJ f-
<C Q'
6 h
'Lt..
O i
40.
2 223 3 3O i
f
_J 5
i 4
4 4
2
.-3 2
s a
s 2
f e
4,'
c 1 n n'n n'nl' n n'f\\'n'r\\"f i>,
o.
0.
O. 5 1.
- 1. 5
- 2. = 1 G' in, n\\
-r..,,-
o
-I
.[
Figure 15E.63-1. ABWR Imss of Feedwater Flow, ARI AT3 REDYA02V ABWR SSAR - ATWS LOSS FW-ARI #1
- 1. 4 DOME PRESSURE 5
i x10 I
i m
g _.1. 2 V
1 e
b D'
Dw i
\\'
,j i
- y j'
.1 e.
4
~
t i i l i I l l t~
g 600.
800.
- 0.
200.
'400.
.<~.--s.
O O
O Figure 15E.63-2. ABWR Loss of Feedwater Flow, ARI AT3 REDYA02V ABWR SSAR ATWS LOSS FW-ARI #2 150.
!i NEUTRON Ft UX (%)
i l
2 AVE' SURFACE HEAT FLUX (%)
FLOW ((%)TU/LB) s CORE INLET 00 LING B INLET SUBC 4
100.
E 3
f g 2 4
- h. Lq 50.
4 4
)
2 4
q 4
4 4
1
%/\\g>,d, M1, j~'
4
- 3. '.
4
- j-^'*.4 h
t
>B
~
~ ~ '
O.
0.
-200.
400.
600.
800.
g g
g s.
Figure 15E.63-3. ABWR Loss of Feedwater Flow, ARI AT3 f E YA02V F
- 3 10.
- LEVEL (FT.REF SEP-SK!RT) r 2 W R SENSED LEVEL (FT) i flIY ($))
ITY ($
i e
s. NET REACTIV l
-* ROD REACTI' l
j i' W
(
=
n 0.
- ~
y 7,,f L
'b l
-i o-
[
,\\/
l m ',
4 5,
i.
'd '
I''
-20.
~ 400.-
O.
-. 200.
600.
800.
y O
O O
j i
Figure 15E.63-4. ABWR Loss of Feedwater Flow, ARI AT3 REDYA02V ABWR SSAR ATWS -
LOSS FW-ARI g4 FEEDWATER FLOW
~
i
- REllEF VALVE FLOW 3 VESSEL STEAM ' FLOW ECCS SCALED FLOW 4
\\,
80.
O.
L4J.
l L I-( -
3
.(-
tr Lu-
'O-S 40.
i 5
^
O'
. _j -
.LL
_ 3
,4 4
3 3
3 3 --
'lg g
g g' 'g l
g 0.
200.-
400.
6001 800.
O O
O Figure 15E.6.3-5. ABWR Imss of Feedwater Flow, FMCRD Run-in AT3 REDYA02V ABWR SSAR ATWS LOSS FW-FMCRD. h 1
- 1. 4 DOME PRES:iURE x 10' i
4 e
- 1. 2 m
1 1
1 f
b
'I O'
1
)(n E
'~ : u,
(>
/
~
~
f l
~
'I'
I
- 0. 8 0.
200.
400.
600.
800.
t O
O O
Figure 15E.63-6. ABWR Loss of Feedwater Flow, FMCRD Run-in AT3 REDYA02V ABWR SSAR ATWS LOSS FW-FMCRD A 2
.150.
HEUTRON F1 UX (%)
i 2 AVE SURFACE HEAT FLUX (%)
FLOW ((%)TU/LB) 2 CORE INLET 0 LING B INLET SUBC0 e
' 100.- c h
3 I
[
,y i
50.
4 4
1
=
i A>
^ '"---
O.
O.
200.
400.
600,.
800.
O' O
O Figure 15E.63-7. ABWR loss of Feedwater Flow, FMCRD Run-in AT3 RED A02V ABWR SSAF TWS LOSS FW-l i DA 3 10.
?
LEVEL (FT REl~ SEP-SKIRT) i i
'ITY (($)) (FT)
D LEVEL 2 W R SENSE s-NET. REACTi\\
/ITY $
r 4 ROD. REACTI' q
h a
7 7
3
-10' e
4
~
4 4
I'
-20.
O.
'200..
400.
600.
800.
O O
O L
l.
l :-.
Figure 15E.63-8. ABWR less of Feedwater Flow, FMCRD Run-in AT3 REDYA02V ABWR ' SSAR ATWS LOSS FW-FMCRD A 4-10.
FEEDWATER FLOW i
. REllEF VALVE FLOW -
3' VESSEL STEAM FLOW ECCS SCALED FLOW (3
1 i
80.
g g.
i l-5
<C ih cc Lti_
O 5
M
. 5.
40...
i
- O
_J
- LL-'
}_
3 4
~
J J-3 3
3 I
0.
'i
- 0. '
200.
400.
600.
800.
m m.
6 Figure 15E.63-9. ABWR Loss of Feedwater Flow, SLCS AT3 REDYA02V ABWR SSAR ATWS LOSS FW-SLCS #1 1.'4 x10' DOME PRESSURE mI
- 7 W
O_
8 t
v 1
Ld.
d E
3 9
W W
i-Ld i.
Y
- 0. 8 O.
- 0. 5
-1.
- 1. 5
- 2. = 10'
/
-r
.- m\\
a--
(
(
Figure 15E.63-10. ABWR Loss of Feedwater Flow, SLCS I
AT3 RED"A02V ABWR SSAR ATWS LOSS FW-SLCS #2 150.
NEUTRON FL UX (%)
i 2 AVE SURFACE HEAT FLUX (%)
FLOW ((%)TU/LB) s CORE INLET 00 LING B INLET S8JBC 1C;0.
I t
S i
f 50.
4 I
4
~
3 1
3
- o. L k'l d@adALAW 0.
O. 5 1.
- 1. 5
- 2. > 10'
Figure 15E.6.3-11. ABWR bss of Feedwater Flow, SLCS 1
LOSS FW-SbCS#3 f
10.
k D LEVEL (FT))
LEVEL (FT.REF SEP-SKIRT i
2 W R SENSE ilY ((1))
h s NET REACTIV flTY D ROD REACTI' 4
- Q i.*
1 s
g (pg g-
[
g'/'
gr L
(
-10.
j
- (
m M
6 4
m
I'
-20.
1]
- 2. = 10' 0.
O. 5 1.
O O
O Figure 15E.63-12. ASWR Loss of Feedwater Flow, SLCS AT3 PEDYA07!
ABWR SSAR ATWS LOSS FW-SLCS #4 FEEDWATER FLOW i
> RELIEF VALVE FLOW 3 VESSEL STEAM FLOW ECCS SCALED FLOW 80.
O W
I-8
<E O'
L1-O M
40.
i 3
'a 3I
~
O
_1 l'-
1 3 h
r l *~
~,
3' l,f 2
3
>3, t
mnn'nn g~
0.
O. 5 1.
- 1. 5 2.=10'
O O
O t
Figure 15E.6.4-1. ABWR Ims' of Feedwater IIcate;, ARI i
AT3 REDYA02V ABWR SSAR ATWS LFWH ARI-RRBK f
- 1. 4
= 10'
' DOME PRES 3URE y
d S
I-
- 1. 2
' ' ' < ~ - -
3 g
O_
i v
1 Q'
D i
i i
g g
g, LLD
\\'
t O'
L I
1 e
i.
l 4
- 0. 8" ' ' ' ' I ' ' ' '
. O.
200.-
400.
600.
800.
_.-,-_,_...--.__-.-......,.....-_..,-.--...._.__,.._-m.
~.. - -.. -.
fcr
l l
l Figure 15E.6.4-2. ABWR Loss of Feedwater Heater, ARI AT3 REDYA03V 8
ABWR SSAR ATWS LFWH Arti-RRBK g2 NEUTRON F1 UX (%)
~
i 2 AVE SURFACE HEAT FLUX (%)
3 CORE INLET FLOW INLET SUBC00LIN TU/LB)
.g 3
3 3
3 3
100.
1 lI 50.
q 4
4 8
4
[
M 2
'I'I O.
O.
200.
400.
600.
800.
e.,,
O O
O l
Figure 15E.6.4-3. ABWR Imss of Feedwater IIcater, ARI AT3 REDYA02V ABWR SSAR ATWS 2
LFWH ARI-RRBK g" Y
LEVEL (FT.REF SEP-SKIRT 2 W R SENSED LEVEL (
3 NET REACTIVITY 2
ROD REACTI' QTY
)
(
r 2
3 2
2 2
')
pc a
I l'
I
-10.
r 6
6 I'
-20.
O.
200.
400.
600.
800.
. Figure 15E.6.4-4. ABWR loss of Feedwater Heater, ARI AT3 REDYA02V r
1 l
ABWR SSAR ATWS LFWH ARI-RRBK g4 120' FEEDWATER R OW i
RELIEF E FLOW VFSRFI'VALV 2
2 8
STFM ROU 2-A ECCS SCALED FLhw p
h.
tsj g.
<[
-O' w
I' O
h M
40.
3 3
O~
_s:
e
~
3 g
g a
8 8
8 8 3 8
8 -*
g
..~
..1. _L f l
. p' i i i f O.
200.
'400.
600.
800.
<--m
O O
O Figure 15E.6.4-5. ABWR Iess of Feedwater Heater, FMCRD Run-in
'AT3 REDYA02V ABWR SSAR ATWS LFWH FMCRD-RR8F fl
- 1. 4 DOME PRES 3URE
= 10' m
<C
- 1. 2 l
g l
O_
v d-E
/
i i
D W
1 g
W l~
1 1.
i Q.
pus i
9 s
I'
- 0. 8 O.
200.
400.
600.
800.
v..
.e enens yyy
rag F
O O
Figure 15E.6.4-6. ABWR Loss of Feedwater Ikater, FMCRD Run-in t.
~
AT3 REDYA02V ABWR SSAR ATWS LFWH FMCRD-RR8P
- 2 150.
E H{AT7.) FLUX (7.)
NEUTRON Fl UX 2 AVE SURFAC E
- CORE INLET FLOW (
INLET SUBC00 LING TU/LB)
=
=
2 s
2 2
-100.
. i r
4.
- 50.
4 4
4-4-
. v
'I'
1 O.
O.
- 200.
400.
600.
800.
s (m,
(~p.
L)
L./
Figure 15E.6.4-7. ABWR Loss of Feedwater IIc. iter, FMCRD Run-in AT3 REDYA02V l
ABWR SSAR ATWS LFWH FMCRD-RRBF g3 9
2 10.
LEVE FT REF SEP-SKIRT) i 2 WR ENSED LEVEL (Fi)
[-
2 NET REACTIVITY ROD REACTI'/ITY
)
2 2
2 2
i O.
4 3
l
-10.
e 6
6 m
I'
-20.
O.
200.
400.
600.
800.
n pY a)
Figure 15E6.4-8. ABWR Loss of Feedwater Heater. FMCRD Run-in AT3 REDYA02V ABWR SSAR ATWS LFWH FMCRD-RRBF g4 120.
FEEDWATER FLOW REUEF E FLOW VFSSFI,VALV 2
STFAM FtOW 2
3 ECCS SCALED FLhW g
80.
J O
Lt1 b
s tr
)
Lt.
o 40.
3
~
o ag
{3 M
g O.
200.
400.
600.
800.
n (r m. m rmy
~
'u J GI Figure 15E.6.4-9. ABWR Loss of Feedwater IIcater, SLCS AT3 REDYA02V ABWR SSAR ATWS LFWH SLCS-RR8K
- 1
- 1. 4 DOME PRESSURE
= 10' n<f 12 (D
O_
v h1 O'
I D
i (A
(A i
Ld O'
1.
i
=-
a
\\
l
- 0. ' 8 'I'
O.
O. 5 1.
- 1. 5 2.=10'
<-m
Figure 15E.6.4-10. ABWR Loss of Feedwater IIcater, SLCS AT3 REDYA02V ABWR SSAR ATWS LFWH SLCS-RRBK
- 2 150.
NEUTRON FL UX (%)
2 AVE SURFACE HEAT FLUX (%)
FLOW ((%)TU/LB) 2 CORE INLET INLET SUDC00 LING B a
,a 100.
i r
2 i
r 3
50.
g*>
_g M
~
y 4
[
j 0.
O.
O. S 1.
- 1. 5
- 2. = 10'
,..,c-tcr ns
O O
O l
Figure 15E.6.4-11. ABWR Loss of Fe(dwater lleater, SLCS AT3 REDYA02V ABWR SSAR ATWS LFWH SLCS-RRBK (3 10.
ENSED LEVEL (FT))
REF SEP-SKIRT
- LEVEL 2 WR 3 NET REACTi\\1TY ROD REACTPSTY,
)
4 I
e 2,
2 2
g I
[
-3 3
4 3
4' 3
4 3
4 3 O.
-~~~~ - w 5
-10.
'I'
-20.
~1; 1;5
- 2. = 10' 0.
O. S
gY
, G.16 Figure 15E.6.4-12. ABWR Loss of Feedwater Heater, SLCS AT3 REDYA02V ABWR SSAR ATWS LFWH SLCS-RRBK (4 120.
FEEDWATER ELOW 2 RELIEF VALVE FLOW 3
2 VESSEL STEAM FLOW ECCS SCALED FLOW
(
a p
Ld
- F--
<C i
3 7
Lt
\\
O N
x '"]
M g
d.
f 1
l :t qO y
tu i
~~
g~
(t
'I1 O.
O.
O. 5 _
1.
- 1. 5
- 2. = 10' n
O ABWR Loss of Feedwater Heater ATWS - FMCRD Run-in, with Run-back MLHGR, KW / FT 2o 18
/
/
14 12
/
{
O L/
4 2
-100
-70
-40
-10 20 50 00 110 180 170 200 Tune, seconds Figure 15E.6.4-13. ABWR Loss of Feedwater Heater, Max. LHGR nU 5
]
Cr Axial Node es N
e T = 127.44
/
y 4 T = 84.96 I
16 f
T = 42.48
/
/
/
~
(
L-T = o.
/
8
(
/
~
0 1
2 3
4 5
Core Average AxialPower-Figure 15E.6.414. ABWR Loss of Feedwater Heater, FMCRD Run-in 3
-l I
n
O O
O Figure 15E.6.5-1. ABWR Turbine Trip w/ Bypass, ARI AT3 REDYA02V i
- 1. 4
' DOME PRESSURE
= 10' e
- 1. 2 a.
f w.
1 j
m Dm i
W Ld M
1.
I a
i i
9 i
3 M
m
'I'
I
- 0. 8 0.
200.
400.
600.
800.
O' O
O Figure 15E.6.5-2. ABWR Turbine Trip w/ Bypass, ARI AT3. REDYA02V ABWR SSAR ATWS TT_BP-ARI J2 -
l-
-150.
- NEUTRON FL UX (7.)-
2 ' AVE ' SURFACE HEAT FLUX (7.)
FLOW ((7.)
CORE. INLET 3
00 LING BTU /LB)
INLET SUBC l
100.
E E
'b.
l 3
3 3
tt
'-N Qa m
80a
Figure 15E.6.5-3. ABWR Turbine Trip w/ Bypass, ARI-
~
AT3 REDYA02V ABWR SAR ATWS TT_BP FI #3 L
10.
p D LEVEL (FT))
LEVEL (FT.REr SEP-SKIRT i
2 W R SENSE ITY (($))
s NET REACIN STY $
ROD REACTP i,
>l O.
o g
-l Y-
)
s 2
V I
t 2'
3 s
-10.
6 m
q M
e
I'
-20.
600.
800.
O.
200.
400.
o, O ^'
Cr Figure 15E.6.5-4. ABWR Turbine Trip w/ Bypass, ARI AT3 REDYA02V ABWR SSAR ATWS TT_BP-ARI g4 120.
FEEDWATER FLOW i
2 REllEF VALVE FLOW 3 VESSEL STEAM FLOW ECCS SCALED FLOW s
e 80.
- O La 5
. F-4Q' tu I
O.
3 40.
n g-r 2
O
_J LL 4
4
.1 3
3 3
'O.
' 0.
200.
400.
600.
800.
Figure 15E.6.5-5. ABWR Turbine Trip w/ Bypass, FMCRD Run-in AT3 REDYA02V ABWR SSAR ATWS TT__BP-FMCRD #1
- 1. 4
= 10' DOME PRES 3URE G
e
- 1. 2 g
Q_
v U
i O'
D U) i V)
LJ
{
(
1.
i i
ep 1
I''
- 0. 8 O.
200.
400.
600.
800.
R O
O O
\\l Figure 15E.6.5-6. ABWR Turbine Trip w/ Bypass, FMCRD Run-in AT3 REDYA02V ABWR SSAR ATWS l-g l
TT__BP-FMCRD #2 150.
NEUTRON FL UX (%)
2 AVE SURFACE HEAT FLUX (%)
FLOW ((%)TU/LB) 2 CORE INLET 00 LING B INLET SUBC t,.
6 I
3 100.
2 3
1 b.
1 I
I 3
l w.
~
\\Q'c' t 1,
=
=
- * ~ ' ' ' ~
0.
O.
200.
400.
600 800.
O O
O I
l Figure 15E.6.5-7. ABWR Turbine Trip w/ Bypass, FMCRD Run-in AT3 REDY 02V A
ABWR SSAR ATWS TT__BP-FMCRD #3 10.
D LEVEL (FT))
LEVEL (FT.REr SEP-SKIRT i
2 W R SENSE MTY ($ITY ($))
3 NET REACTIV ROD REACT 1
2 5
/
/
1 _
f--
N'
~
t
(
[
-10._
r
-)
4
- -20. I ' ' ' ' I O.
200.
400.
600.
800.
r...e.
recor
O O
O i
t Figure 15E.6.5-8. ABWR Turbine Trip w/ Bypass, FMCRD Run-in i
AT3 REDYA02V ABWR SSAR ATWS IT_BP-FMCRD #4 120.
FEEDWA1ER FLOW i
2 REllEF VALVE-FLOW 2 VESSEL STEAM FLOW ECCS SCALED FLOW 3.
i-80.
O w-5 r
F-
<C CE Q-l O
5 2
M
'N i
40
> n
'r (
j g
O i
___I LL
~
4 4
l r
-3 i
3 3
3 8 ' I
^
O.
O.
200.
400.
600.
800.
/
+-
O O
O Figure 15E.6.5-9. ABWR Turbine Trip w/ Bypass, SLCS AT3 REDYA02V ABWR SSAR ATWS TT__BP-SLCS #1
- 1. 4
^
= 10'
' DOME PRESSURE G
- 1. 2 i
g O_
fI\\
f I
f v
i I
T
~3 m
W W
l
't 1.
6 i
g i
)
/
.r.
l
~
1 i
'I'
- 0. 8 O.
200.
400.
600,.
800.
T
O O
O I
Figure 15E.6.5-10. ABWR Turbine Trip w/ Bypass, SLCS AT3 REDYA02V ABWR SSAR ATWS g
TT_BP-SLCS #2 150.
NEUl PON FL UX j7.)
i 2 AVE J URFACE HtAT FLUX (7.)
' INLET, FLOW ((7.) /LB) 2 CORE i
INLEl SUBC00UNG BTU t
=
l l
1 100.
=
i 5
i y$ 'l 50.
t i
a.
a 2
5 2
l s
s
'I'
0.
600.
O.
- 200, 400.
800.
O' O'
O t
I Figure 15E.6.511. ABWR Turbine Trip w/ Bypass, SLCS
!AT3 REDYA02V ABWR SSAR ATWS TT_BP-SLCS #3 10.
-LEVha.(FT,RE k'
D LEVEL (FT))
F SEP-SKIRT i
r W R SENSE r
ITY (($))
i NET !
ACTIV i ROD CTr /ITY $
a s
1.
/
/
^ ^~
- 0. rv er' ""
1 3
i
/
l 1
y I
3 2
l 3
t 2
f
't
{
{
a
-10.
y i
1 l
i
~
e I'
-20.
8 ' ' '
0.
200.
400.
600.
800.
-,.ee
/ne n\\
,-,--,-.-,-,w,
~
l l
Figure 15E.6.5-12. ABWR Turbine Trip w/ Bypass, SLCS AT3 REDYA02V ABWR SSAR ATWS IT_8P-SLCS #4 120.
FEEDWATER FLOW i
^
2 REllEF VALVE FLOW 3 VESSEL STEAM FLOW ECCS SCALED FLOW 5-i I
1 80.
O Ld I,
F-it
<E l
I 5 ii ii g
t I
a ti-O
)
) )
/
O l
- 40.,i
,2 2
2 N
3 i
O Is
' s L
_I 5
s i
,n 3
1 r,;
- g
(
4 9
2 3
9 2
3 e
2 4
2 4
2 3 8'
23 3
~
'I'
^~
O 0.
200.
400.
600.
800.
Figure 15E.6.6-1. ABWR Loss of Condenser Vacuum, ARI AT3 REDYA02V ABWR SSAR ATWS LOSS _VACUM-ARI p 1
- 1. 4 DOME PRES 5URE x10' m
- 1. 2 O_
i t
V I
1 W
1 m
t t
t 6J M
1.
Q_
~
t
'I'
- 0. 8 O.
200.
400.
600.
800.
TI L A f-
[ C l-N cias otrtn
CP O
Opre 15E.6.6-2. ABWR Loss of Condenser Vacuum ARI AT3 REDYA02V ARWR CSAR ATWS i
LOSS _ VACUM-ARI v2 NEUTRON FL UX (7.)
l i
2 AVE SURFACE HEAT FLUX (7.)
I s CORE -INLET FL INLET SUBC00LIN U/LB) a n
r a
l 1
.100.
~~
~-
r
-t 2
Li
- 50. ')
4 4
a Ul 4:
-s J vg s.
s On 4.1 'l
- i %
v-it o.
.w-
.g.
O.
200.
.400.
.600.
800.
Tis e r
/crM ci,,, nic,n
e O'
O O
as a
g s
Figure 15E.6.6-3. ABWR less of Condenser Vacuum, ARI AT3 REDYA02V ABWR SSAR ATWS' LOSS _VACUM-ARI g3 10.
LEVEL FT.REF SEP-SKIRT 2.W-R P LEVEL (FT NET REACTISTY s
' ROD REACTP'/ITY ii 4
i.
k.
\\c w_)N 2
s 2) 3 hi.
O.
p; I
g 2 '
f s'
s
-10.
8
' ' ' I ' '
-20.
.600.
800.
- O.
200 400.
Tf t A F
/P RAT -
.re
...n.etn
O~
O~
CT Figure 15E.6.6-4. ABWR Loss of Condense Vacuum, ARI AT3 REDYA02V ABWR SSAR ATWS LOSS _VACUM-ARI v4 0.
FEEDWATER FLOW 1
2 RELIEF VALVE FLOW 3 VESSEL STEAM FLOW ECCS SCALED FLOW j
4 E
i 80.
g b
2 l-
.<C 0'
Lt l3
)
O 92 o"
l 40.
3 i
o v
-J LL r-I 4
.2 2
2 2
2
_s 3
1 38 3
3 et 4
i 3
2 5
2 8 1 3
2 4 1 3
. O.
200-400.
600.
800.
n i nr-
/ c c-M
. c.. n,cio
(^p
(~'y n (v~
kj'
'v Figure 15E.6.6-5. ABWR Loss of Condenser Vacuum, FMCRD Run-in AT3 REDYA02V ABWR SSAR ATWS LOSS _.VACUM-FMCF
- 7 l4 DOME PRES'iURE 3
i xi0 i
i m
<Cg
- 1. 2 i
i i
i a
v i
i t
Ld d
i M
s o
D-U)
V)
La T
1.
4 O_
.i
'I'
- 0. 8 O.
O.25
- 0. 5 0.75
- 1. = 10'
,.. e e-
/e r n \\
n
-s
O O
O Figure 15E.6.6-6. ABWR less of Condenser Vacuum, FMCRD Run-in AT3 REDYA02V ABWR SSAR ATWS i_
LOSS _VACUM-FMCF
- 8 I
150.
NEUTRON FL UX (7.)
i 2 AVE SURFACE HEAT FLUX (7.)
3 CORE INLET FLOW (
i
' INLET SUBC00 LING U/LB)
(
100.
2 L
7 50.
- 631, q
ll J
4 L
5 s
l i
h, h* l
\\ f' ~/ %
1
~
~
' ' M"
. 0.
O.
O.25 O. 5 O. 75
- 1. = 10' r. i a r-rct ns
[
Figure 15E.6.6-7. ABWR Loss of Condenser Vacuum FMCRD Run-in AT3 REDYA02V ABWR SSAR ATWS LOS VACUM-FMCF
- 9 LEVEL
.REF SEP-SKl i
2 3 WR JSED LEVEL (FT 2 NET REACTI\\ ITY ROD REACTivlTY I
i 3
3 2
1 0.
f 3
I
- - (
-10.
i
~
s 4
i 1'
-20.
- 0. -
O.25
- 0. 5 -
O. 75
' 1. # 10' Tii nr
/cre T-
<...,, rn<-
N
O O
O Figure 15E.6.6-8. ABWR Loss of Condenser Vacuum, FMCRD Run-in AT3 REDYA02V ABWR SSAR ATWS l
3 LOSS _VACUM-FMCF 10 i
120.
i EEDWATER FLOW 2 RELIEF VALVE FLOW 3 VESSEL STEAM FLOW '
l ECCS SCALED FLOW p.
~
l 80.
.g Lt.]
tC c
T Ls_
- 3 3
O g
M 40.
3 J'3 O
\\
t J
L1_
4 j lr 3
3 2 3 2 3 23' 23 23 23' 2
' I
]
O b
N i
I*I l
R I
l' 0.
O.
O.25 O. 5' O, 75 i. m 10'.
/cr rT r n n-nni nwns
~
w
,. ~
Y a.
v i
Figure 15E.6.6-9. ABWR Loss of Condenser Vacuum, SLCS AT3 REDYA02V ABWR SSAR ATWS LOSS._VACUM-SLCS
- 1
- 1. 4 x10' DOME PRESSURE A
s i
(A i
L.j O'
1.
s'f i
Q_
~
I''
O.
O. 5 1.
- 1. 5
- 2. = 10' I
- 0. 8 j
,_s
Figure 15E.6.6-10. ABWR Loss of Condenser Vacuum, SLCS AT3 REDYA02V ABWR SSAR ATWS LOSS _VACUM-SLCS
- 2 150.
NEUTRON Ft UX T
iln i
2 AVE SURFACEH FLUX (%)
2 CORE INLET-FLOW (
INLET SUBC00 LING TU/LB)
I 4
l' l
100.
5 i
I 5
1 3
50.
i 3
3 4
L J
3 6
3 3
3 h
l 3
0'. ' ' '
L -
~
~~ '
O O. 5 1.
- 1. 5
- 2. = 10' Tii n r
/croi o,c,-
(~%.
r3
(_-
U Figure 15E.6.6-11. ABWR Loss of Condenser Wcuum. SLCS AT3 REDYA02V
(
frBWR SSAR AT1 S LOSS VACUM-!1SS
- 3 i
10.
5 D LEVEL (FT))
LEVEL (FT.REF SEP-SKIRT 2
i 2 W R SENSE ITY (($))
s NET REACTIV i
/ITY $
(
ROD REACTI' a
I 2
b
[J.' r i
/L
_j 5
3 p
r w s
O.
3 3
f 2 i s
(
3 b')I 7
e f
s r
i
-10.
Y
'I'
-20.
O.
O. 5 1.
- 1. 5
- 2. = 10' r,-rns
Om o^
O^
Figure 15E.6.6-12. ABWR Loss of Condenser Vacuum, SLCS AT3 REDYA02V f
LOSS _VACUM-SLCS #4 120.
FEEDWATER FLOW i
5 2 RELIEF VALVE FLOW 3 VESSEL STEAM FLOW ECCS SCALED FLOW F
I 80.~
O LJ
}-
<C T
w r
O M
40.
3 i
~,
O 6
i
_J Lt_
4 1
2W
-~
?
O 3
2 3
2 3
2 R
R
'F.'
i, ',
i 0.
O.
O. 5 1.
- 1. 5
- 2. = 10' rhAF
[PIM en.. n,res n
O^
O^
On Figure 15E.6.7-1. Feedwater Controller Failure Maximum Demand, ARI AT3 REDYA02V ABWR SSAR ATWS FWCF_. MAX-ARI #1 1.- 4 x10'
' DOME PRESSURE 3
I
<t
- 1. 2 y) v, b
Q.
t
- -)
.U)
- y) :
12J
'l'
-1.
ia-i
/
~
L 3
t i.
' ' I l
.- 0. 8 0.
200.
'400.
600.
800.
= _
O' O^
C Figure 15E.6.7-2. Feedwater Controller Failure Maximum Demand, ARI AT3 REDYA02V ABWR SSAR ATWL FWCF_ MAX-ARI #2 150.
NEUTRON Fl UX (%)
i 2 AVE SURFACE HEAT FLUX (%)
FLOW ((%) /LB) s CORE INLET INLET SUBC00 LING CTU 100.
r I
E f
50.
- 1 e
I 1
3 4
N
\\
s,.
.c t
x,
=.
' ' ' ' X"d '
OL O.
200.
400.
600,.
800.
3 O'
GT M
Figure 15E.6.7-3. Feedwater Controller Failure Maximum Demand, ARI AT3 REDYA02V ABWR SS ATWS FWCF_M ARI g3 10.
D LEVEL (FT))
. LEVEL (FT.REr.SEP-SKIRT 2 W R SENSE
!TY- (($))
s NET REACTIV
/ITY $
ROD REACTl' l
4 I
2 g
f
/
=
~
O.
y 7
. i i
2 5
-10.
5 s
main
I''
-20.
O. -
200.-
400.
600.
CJ0.
..<-cm
ry r~~'
V' Figure 15E.6.7-4. Feedwater Controller Failure Maximum Demand, ARI AT3 REDYA02V ABWR SSAR ATWS FWCF_ MAX-ARI #4
~
1120.
FEEDWATER FLOW i
i 2-REllEF VALVE FLOW s VESSEL STEAM FLOW l
ECCS SCALED FLOW 3
a i
1 80.
O Ld i
k h
,3 I
4 j
k' h
O l o i
M l,
s 40.
p 3
[
.(
o b
~
2 i
3
~
l i
4 t
ig
^
3 s
s l '-
1 0.
0.
200.
400.
600.
800.
e u
1 Figure 15E.6.7-5. Feedwater Controller Failure Maximum Demand, FMCRD Run-in -
AT3 REDYA02V ABWR SSAR ATWS FWCF_ MAX-FMCRD #1:
- 1. 4
-,10 i DOME PRES iURE 5
L m
4g.
- 1. 2 a-.
v
. (r -
i D.
.M
(
in.
Ld (E
.1.
i
.e MWD-6 4
- 0. 8
-200.
400.
, 600 800.-
.....; O,J T.. e r
/cenY
-g
- O O
Figure 15E.6.7-6. Feedwater Controller Failure Maximum Demand, FMCilD Run-in AT3 REDYA021/
ABWR SSAR ATWS FWCF_ MAX-FMCRD #2 t '
150.
NEUTRON FL UX'(7.)
l.
2 AVE SURFACE HEAT FLUX (7.)
s CORE INLET FLOW
.I' INLET SUBC00 LING BTU /LB) a i
3 E
t 00.
?
3 50.
- l
~
2 3
3
(
h.
~
3 u
1 1
2
'L4 s
i s
i-3-
s
,.,,a
,3 g 0.'
-200<
400.
600.
800.
e O
O O
Figure 15E.6.7-7. Feedwater Controller Failure Maximum Demand, FMCRD Run-in AT3 REDYA02V ABWR SSAR ATWS l
FWCF__ MAX-FMCRD #3 10.
D LEVEL (FT))
LEVEL (FT.REF SEP-SKIRT l
2 W R SENSE ITY (_$
VITY ($))
. NET REACTIV ROD REACTI l
/
[1
/
%g 4
2
-10.
~
1 I
s 5
1 e
~
4
''I'L
-20.
O.
200.
400.
600.
800.
r. i a r-
/crnN c...
m,..,,
^
~~ '
K^
L l
e e
e 1
l l
Figure 15E.6.7-8. Feedwater Controller Failure Maximum Demand, FMCRD Run-in AT3 REDYA02V ABWR SSAR ATWS FWCF_ MAX-FMCRD #4 120. 9 FEEDWATER FLOW i
4 2 REUEF VALVE FLOW 3 VESSEL STEAM FLOW ECCS SCALED FLOW 4
/
l
' 80.
g
'2J l
H i
<C Z
h L1 O
'n\\
m 40.
1 3
[
N O
4 4
i r
- l, 3
3 3
3 i2 3
8 23 a
-g d
I i f i I I I Q
o 900 400.
600.
800.
f'y (3
r~p
'%J
'm//
Qt Figure 15E.6.7-9. Feedwater Controller Failure Maximum Demand, SLCS AT3 REDYA02V ABWR SSAR ATWS FWCF_ MAX-SLCS h i
- 1. 4 DOME PRESSURE x 10' i
G
- 1. 2 g
g ct v
i
!1$
l Dm
~
(A LJ E
1.
cL 4
m 6
6
I'
- 0. 8 O.
200.
400.
600.
800.
m..
. r--
ec r ns
,3 h,--3 V'
Figure 15E.6.7-10. Feedwater Controller Failure Maximum Demand, SLCS AT3 REDYA02V ABWR SSAR ATWS FWCF_ MAX-SLCS # 2 150.
i NEUTRON FL UX (%)
i 2 AVE SURFACE HEAT FLUX (%)
2 CORE INLET FLOW (
- INLET SUBC00 LING U/LB) h 100.
2 3
2 50.
-l 2
s
~
3 2-
~
ks 1 1
4
'J i " "
0.-
O.
200.
400.
600.-
800.
.=
O O'
CF' t
Figure 15E.6.7-11. Feerfwater Controller Failure Maximum Demand, SLCS AT3 REDYA02V ABWR. SSAR ATWS FWCF_ MAX-SLCS # 3
~
10.
LEVEL (FT RElD LEVEL. (FT))
SEP-SKIRT i
2 W R SENSE VITY ($))
ITY s NET REACTIV 4 ROD REACTI
. 0.
3.
3:/,'
.j 3.
3
)
2 1
J 2
-10.
2
' ' ' ' ! ' '~' '
--20.
O.
- 200.'
400.
600.'
800.
Egure 15E.6.7-12. Feedwater Controller Failure Maximum Demand, SLCS AT3 REDYA02V 3
ABWR SSAR ATWS FWCF_ MAX-SLCS # 4
~
120.
FEEDWATER FLOW i
2 REllEF VALVE FLOW 3
s VESSEL STEAM FLOW ECCS SCALED FLOW 4
j m
80, g
LiJ I
F-
<t
'8 m
L t,_
a O
N l
40.
3
[
O 5
_J i
LL 3
r k_
~
,1<i,sf,a s2 s
s 2
,i<_
o.
0.
200.
400.
600.
800.
.. -, s
(
ABWR Standard Plant (i
15E.7 CONCLUSION' Based upon the results of this analysis, the proposed ATWS design for the. ABWR is satisfactory in mitigating the consequences of an ATWS. p.ll performance criteria specified in Section 15E.2 are met.
1 1 It is also concluded from results of the above analysis that automatic boron injection could mitigate the most limiting 4TWS event with t margin (at least 1.1 kg/cm margin in peak containmeut pressure). Therefore, an automatic
{
~
SLCS injection as a backup for ATWS mitigation -
g is acceptable.
T
(
a
\\
l
):
I p
i 4-;
t
(
i l
h l
i.
.c.
Y
-(
4 h
a i
m
\\
s i
(>~
. s.
t k_ '
e Y
c
)
=.
x.
1 f..
~
m
i
\\
ABWR I
i ' ;f f Standard Plant
\\
15E.8 REFERENCE 3
1,
- 1. Assessment ofBHR Mitigation ofATH5, i
1 (NEDE.74222, September,1979),
l r
\\
.ji
'\\
(
i ix r
l l
l' l
I i
,