ML20101G588
| ML20101G588 | |
| Person / Time | |
|---|---|
| Site: | 05200001 |
| Issue date: | 05/28/1992 |
| From: | Fox J GENERAL ELECTRIC CO. |
| To: | Jun Lee, Poslusny C, George Thomas NRC |
| References | |
| NUDOCS 9206260155 | |
| Download: ML20101G588 (7) | |
Text
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11AY 28 '92 01:3 #l1 G E f0CLEFA BLIG J P.2/7
' ABWR ux6i=^u Standard Elant firv. c Table 15.0 2
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RESULTS
SUMMARY
OF SYSTEM RESPONSE ANALYSIS TRANSIENT EVENTS Mas Core No.
Man. -
Man.
Average
.1 Durr r.
Man.
Mu Veuel Steam surface VaMs of l
Sub Nawiron Dorne -
Bottom line Heat ihm o
Fnq. Fint : D:owdowf Section Figure flus Pressure Pnssure Pressure
(% of in Cate-Blow.
(uconds) 2 2
2 U2.
L{h Desmr4
% NBR Qig!,(s g)
Ma!Cm g) gggjgg El d
Ull-EPff. 123 15.1 Decrease in core coolant tetr.pernrurs 15 1.1 Loss of feed.
112 8 73.1 75 9 71 6 112.4 0.07 - e 0
0 water heating 15.1.2 15.1 2 Runout of one 104.5 73 2 75.8 71.7 101 8 0.06 a
0 0
feedwater pump 139.0 83 3 84,9 81.8 10$ 9 0 10 a+
10 6
15.1 2 15.1phsoowsgN=o
,.u c. 4.. ts m
g;g. gu : %- % D e mat 15.1 3 15.1 4 opening of 102.1 711 75.6 71.6 100.0 a
0 0
onc Bypass VaM 15.1.3 15.1 5 Openi g of a0
- 102.0 80,4 81 8 00.1 100.0 e+
0 0
Control and Bypasa VaM,
15 1.4 Inadwttant open SEE
'!TXT ing of One SRV 15 1.6 Inadwrtent RNR SIDwJ C/csseg D[
SEE TEXT Shutdon1:Coohng 08 O'
I 15 2 lacnase in g p orPnasu 15.2 1 15.21E,Closurv olon 3ms"
.Mf Jhs Jef
.,Mief 0.10 _
a 0
0 Turbac Con v s.i{
g.1 91 (,
j). 'l lo j.6 VaM
.. 1 l 0. A i it.)-I b l'83 ID i7.3
'II 3
'#3'3 15.2.1 150 2 Pres. Regu.
1543 85.8 87.4
- 85.1 103.0
'9fM c
18 6
tatoe boenscale Fall 15 2.2 15.2-3 Generetot land 148.1 832 84.7 82.7 100J -
0.06 a
10 5
l RcM4 Bypass on Frequency definition is discussed in Subsection 150.4.1 Not limiting (See Subsection 150.4.5) a.
Moderate FrequelfVy
~ * *
- C P A-CHW:n dD *t 'ho1*f/*l)'*
b Infrequent c
Limiting Fault N/A Not applicable lhls event should be classified as a limitmgfault. However, criteria for moderatefrequent inciden:s ars.
+
conservatively applied.
15 0-7 Amendant 15
. F F Mf 4C?*?2 Elf 97 0p29-9:
c3 39 pg p g ;s
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' Standard.flant nirv c
^
expected for this tUadient.
15.1.2 Feedwater Contruller Failure..
Maximum Demand -
15.1.13 Core and System Performance 15.1.2.1 Identificatloa of Causes and 15.1.1J.1 input Iarameters aud Initial Frequency Classification Conditions 15.1.2.1.1 Identification of Causes The transient is simulated by programming a change in feedwater enthalpy corresponding to a This event is postulated on tbe basis of a 55.60C loss is feedwater heating. Aa..A cau single failure of a control device, specifically w:t4 $14 A T r4 h
- N t A m c 3,,g g y e,
one which can directly cause an increase in c
15.1.1 3.2 Results i!=/n %/g./.
coolant inventory by increasing the feedwater flow.
Because the power increase during this event is relatively slow, it can be treaud as a quasi The ABWR feedwater control system uses a steady state translent. The 3 D core >imulator, triplicated digital control system, insteed of a pgAcgA,has been used to evaluate this event for the single. channel analog system as used in current equilibrium cycle. The results art vammarized in BWR designs (DWR 2 6). The digital systems Tablgr15.12W (3 l. no.,
consist of a triplicated fault tolerant dieltal controller, the operttor control stations and The MCPR response of this event is small duc displays. The digital controller contains three to the mild thermal power increase with shifting parallel processing channels, each containing-l axial shape. The worst a CPR response is 0.07.
the microprocessor based hardware and associated software necessary to perform all the control No scram is initiated in this event. The calculations. The operator interface provides l lacreased core inlet subcooling aids thermal information regarding system status and the margins. Nuclear system premre does not change required control functions.
2 significantly (less than 0.4 Kg/Cm ) and consequ ntly, the reactor ' iolant pressure
- Redundant transmitters are provided for key boundary is not threatened.
process inputs, and input voting and validation are provided such that faults can be identified 15.1.1A Barrier Performance and isolated. Each system loput is t.iplicated internally and sent to the three processing As noted previously the consequences of this channels. (See Figure 15.11) The enannels -
event do not result in any temperature or will produce the same output during normal pressure transient in excess of the criteria for operation. Interprocessor communication which the fuel, pressure vessel or containment provides self diagnostic capability. A'two out-are designed; therefore, these barriers maintain of three voter compares the processor outputs to j their integrity and function as designed.
generate a validated output to the control actuator. A separate voter is provided for each 15.1.1.5 Radiological Consequences actuator. A 'ringback" fetture feeds back the final voter output to the processors. A voter Because this event does not result in any fuct failure will thereby be deter +d and alarmed.
failures or any release of primary coolant to In some cases a protection circuit will lock the either the secondary containment or to the' actuator into its existing position promptly environ:nent, there are no radiological after the failure is detected, consequences associated with this event.
Amendment 15 1114 FMM 408-925:5e7 05-2S-?:
07:M FM PC3
MAY 28 '92 01:35PM G E f0CLEFR BLDG J P.44 ABWR u-ptandar j Platit Prv c their integrity and function as Vped.
turbine control valves and bypass valves could 4
be fully opened. However, the probability of 15.1.2.5 Radiological Consequences this event to occur is extremely low (less than 7 x 10 6 failure per reactor year), and benee While the consequences of this event do not the event is considered as a limiting fault.W'4N result in any fuel failures, radioactivity is
& r.*it4v4 of. mode r n fye.f i.s s W,. f t A4 nevertheless discharged to the suppression pool 15.1J.1.2 Frequency Classificat on e,.,.,q w j,
, ft,g as a result of SRV actuation. However, the mass g Wgj input, and hence activity input, for this event 15.1J.1.2.1 Inadscrtent Opening of one is much less than those consequences identified Turbine Bypass Valve in Subscuion 15.2.4.5 for Type 2 events.
Therefore, the radiological exposures noted in This transient disturbance, estimated to Subsection 15.2.4.5 cover the consequences of occur less than 0.0088 times per year, is this event.
conservatively categorized as one of moderate j frequency.
15.1.3 Pressure Regulator Failure-Open 15.1J.1.2.2 Inadvertent Opening of all 15.13.1 Identification of Causes and frequency Turbine Control Valves and Bmass Vahes Classifications The frequency of occurance for this event is 15.1J.1.1 Identification of Causes estimated to be less than onec per 10000 years.
It should be classified as a limiting fault as The ABWR steam bypass and pressure control specified in Chapter 15 of Regulatory Guide system (SB&PCS) uses a triplicated digital 1.70. Nonetheless, since the consequence of control system instead of an analog system as this event has no significant impact on the used in current BWR designs (BWR 2 6). The operating CPR limit, the criteria of moderate SB&PCS centrols turbine control valves and frequent incidents are conservatively applied to turbine bypass valves to maintain reactor this event.
i pressure. As presented in Section 15.1.2.1.1, no credible single failure in the control system 15.1J.2 Sequence of Ennts and Systems will result in a maximum demand to all actuators Operation for all turbine control valves and bypass valves. A voter or actuator failure may result 15.1J.2.1 Sequence of Esents in an inadvertent opening of one turbine control valve or one turbine bypass valve. In this case, 15.1J.2.1.1 Inadvertent Opening of One the SB&PCS will sense the pressure change and Turbine Bypass Valve command the remaining control valves to close, and thereby automatically mitigate the transient Table 15.16 lists the sequence of events and maintain reactor power and pressure, for Figure 15.14.
Because the effeer of sudden opening of one 15.1J.2.1.2 Inadscrtent Opening of All b> pass vahe, which bypasses about 11% of rated Turbine Control Valves and Bypass Valves steam flow when full opened is more severe than sudden opening of one turbine control valve, Table 15.17 lists the sequence of events for t
which is almost wide open at rated power, it is Figure 15.15.
assumed for purposes of this transient analysis that a single failure causes a single bypass 15.1J.2.1J Identillcation of Operstor valve to fail open.
Actions As presented in Section 15.1.2.1.1, multiple 15.1J.2.1J.1 Inadvertent Opening of One failures might cause thr. SB&PCS to erroneously Turbine Bypass Valves issue a maximum demand to all turbine control vahes and bypass vahes Should this occur, all Because no scram occurs during this event, no Ameridment 1$
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t14Y 28 '92 01:37Pt1 G E TUCLEAR ELDG J P. 74
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ABWR R
Standard Plant __
15E.6 TRANSIENT RESPONSES t.he FMCRDs. The insertion of the control rods is su. cessful in bringing the reactor to hot shutdown.
For every event selected for analysis, three Peak valuer of key parameters are shown in Table cases were analyred. The first one shows the 15E.6.11 for the ARI case and Table 15E.6.12 for ATWS perfortnance 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 case, which uses FMCRD shutdown by automatic boron injection in about 19.4 run in, assurning a total failure of ARI, was minutes from the beginning of the event. The performed to show the backup capability of trausient behavior of this case is listed in Table FMCRD run.in. The third case was analyzed to 15E.6.14 The rea. tor system response is presented show the in-depth ATWS ruitigation capability of by Figures 15E.6.1-1 to 15E.6.1-4 for ARI activated, the ABWR. In this case, both ARI and FMCRD Figures 15E.6.15 to 15E.6.18 for FMCRD run-in run in are assumed to fail. Automatie boron case and Figures 15E.6.19 to 15E.6.112 as SLCS injaction with a 180 seconds delay, are rzl* d operating, respectively. The normalized axial power upon to mitigate the transient event.
shape change during FMCRD run-in are presented in
[W Figure 15E.6.1-13. The increase of the local power p+ p@c: nc &c grformance criteria / cur -.d [,
If the ARI and FMCRD run in f at the same rime, which has extremely logprobability of a %.QGT.f%N occurrence, the peak reactor would still be 15E.6.2 Loss of AC Power 3
controlled by the Recirculation runbsck and relief valves. However, the nuclear shutdown In this event, all scram signal paths, including valve wilI then re1y oa the automatic SLCS position, high flux, high pressure, low level, and all injection. The boron would reach the core 60 manual attempts have been assumed to fail.
3 seconds after the initiation. The operation of j
both SLCS pumps generate a 100 gpm volumetric The loss of AC power has the following effects:
Dow 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 ttirbine control valves to close.
15E.6.1 Main Steam Isolation Valve Closure (2) As a result of the load rejection, four of the ten recirculation pumps will trip.
This transient is considered an initiating ev-nt caused by either operator action c.
(3) Due to the loss of power to the condensate instrument failure. Scram signal paths thet 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 rnain ali snual attempts. A short time after the condenser vacuum.
MSIVs ha.: closed cornpletely, the ATWS high pressure setpoint u ehd. which initiates Figures ISE.6.2-1 to 15E.6.2-4 show the transient four,f the ten recirculation pumps to trip and beha"ior 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.6.212 for automatic SLCS, respectively.
and increases core voids, thereby reducing power generation which limits pressure increase The fast closure of the turbine control vahes causes a and steam discharge to the suppression 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 ART and the electric insertion of valves have closed. Because the four pumps have already tripned at this time on the load rejection i
i
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