ML20202B388
| ML20202B388 | |
| Person / Time | |
|---|---|
| Site: | Mcguire, McGuire  |
| Issue date: | 06/30/1986 |
| From: | DUKE POWER CO. |
| To: | |
| Shared Package | |
| ML20149C108 | List: |
| References | |
| TAC-61512, TAC-61513, NUDOCS 8607100291 | |
| Download: ML20202B388 (218) | |
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. T_able 15.0.3-2 (Page 2 of 2)
Summary of Accidents Analyzed With Computer Codes , s . . . FSAR Transient' . Section. Number Description'of Transient Summary of Cases Analyzed ' ! 15.4.1 12 Zero Power Rod Bank' ,
. Withdraw 1 ,
.* - t l
l 15.4.2 13a .
' Full Power Rod Bank 1) Fast react,1vity insertion -
, . 13b Withdraw 1 ,
- 2) Slow reactivity insertion
.*. s
~
- 15.4.'3 14a Control rod misoperation '
- 1) Dropped rod (s)" - .
, 14b
- 2) Dropped rod bank
- 14c - . 3) Misaligned rod
. 14d 4) Single rod
- withdrawl< -
I
- s e ,
l 15.4 4 15 . Cold Water Accident . 4 - 15.4.7 16a- . Misloaded Assembly 1) Regio ~n 1 % Regjon 3 . .
. 16b .
. 2) Region 1
- Rggion 2 , ,
16c '
- 3) Region 2 in center -
16d *
- 4) Region 2 o'n periphery ,,
15.4.8 17a > Rod. Ejection -
- 1) BOL, full power '
= 17b .
- 2) BOL, zero power -
, 17c. ,. -
.3) EOL, full power .
17d* , 4) EOL, zero power - y 15.5.1- 18 Spurious.SI' Actuation {* 15.6.1 19 - Accidental RCS * ' Depressurization . . . 2 15.6.5 - 20 i " Loss of Coolant ' Accident - '
- 1) 4 inch SBLOCA ,
20b , -
- 2) 6 inch SBLOCA , ,
20c . 3) 8 inch SBLOCA t . I , 20di ' '
'. 4) DECLG C =0 A , Perfect Mixing *'
. 20e ' '
- 5) DECLG C =0.6, Perfect Mixing '
20f .
. - 6) DECLG C =0.8, Perfect Mixing '
, 20g ,7) BECLG C =0,6, Imperfect Mixing -
860710o291 860630 PDR , ' ' P ADOCK 05000369 .
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0 Table'15.0.3-2 (Page 2 of 2) .
- Summary of Accidents Analyzed With computer Codes 3 .
FSAR Transient" ' ' Section Number Description of Transient Summary of Cases Analy' zed' e * *
~
15.4-1 12 Zero Power flod Bank . . : Withdrawl -
* ** . t 15.4.2 . 13a
- L Full Power Rod Back
- 1) ' Fast reactivity insertion
- l' '
13b Withdraw 1 , ,
- 2) Slow reactivity insertion 3 s ,.
l . 15.4.3 * , 14a Control. rod misoperR. ion 1) Dropped rod (s) - 14b .
~'
- 2) Dropped rod bank ,
'. 14c .
- '3) Misaligned rod *
. . 14d, ,4) Single red withdraw ,, 1- ,[ .
. . a. ~
~15.4.4 .
15 -
" Cold Water Accident . .
15'.4.7 , 16a Misloaded Assembly 1) Region 1
- Region 3 .-
~
~
. 16b - . 2) Region 1
- Regio'n 2 .
~
16c
- 3) Ifegion 2'in center -
,16d , ,
- 4) Region 2 on periphery ,
~
Rod. Ejection 1) BOL, fu11' power
~
15.~ 4. 8 - 17a , 17b -
. 2) BOL, zero power -
17c .
* -# 3) EOL, full power .
17d. 4) E0L, zero power - G ', -' g 15.5.'1 18 ~ Spurious SI Acttistion , ( , c - 1 15'.6.1 19 AccidentalRQ , Depressurizarlon a, ,
. 9 -
~
15.6.5 20a Loss of Coolant' Accident
~
.c 1) 4 inch SBLOCA . .
20b
' '.2) 6 inch SBLOCA -
, 20c. - 3) 8* inch SBl,0CA ,g .
- 4) DECLG C =0.4, Perfect Mix'ing
- 20d .
- .20e :- -
5 DECLG C =0.6, Perfect Mixing e 20f
' J6)) DECLG)C =0.8, Pegfect Mixing.
20g 7y DECLG.C =0.6,. Imperfect Mixing - 8607100291 PDR O .
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-' a Table 15.0.'3-3 s~ .
SummaryofComputerCodesan'dMethodologieh'Used , in Accident Analyses . . Computer Code . Transient Numbers
- Analyzed with . . .
or .Methodo19tly ,
. 'that Computer Code. or Methodology .
LOFTRAN 1-11 (all cases), 13a, 13b, 14a, .
.- -14b, 15,.18, 19 ,
FACTRAN , 9-12, 17a-17d
.THINC a
. . ,,9,10,12,'i4a-14d . .. '.
TW' INKLE -
*12, 14a'14d *
.WFLASH - -
20a-20c -- e
$ATAN-VI . .
20d-20g .
~ '
WREFLOOD 20d-20g - .
- POWLOCTA .
20.d-20g , LOCTA.I.V , 20a-20g , _ LOTIC .. 20d-20g . 20d-20g
.9 , BART' . .
O'
- o PAD - 20d-20g. .
.,3, ITOP 1a,1b, 2a-2d, Sa-5d, 9,10, J614 *
- (all cases), 18, 19 WRB-1 .. la,1b, Za-2b, Sa-5d, 9' 10, 12-14 ,
~
. (all cases), 18, 19 ,
W-3 3", 4a,'4b ,
*
- Transients are numbered according to the cases list,ed in Table 15.0.3-2. ,
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' '8 TABLE 15.0.3-4 (Page 1 of 5) ,
. Summary of Input Parameters for. Accident Analyses
. Initial Moderator Moderator -NSSS RCS Vessel Pzr Pzr Water Eeedwater Transient Temperatbre Density Output flow Press. Volume Temperature T,yg .
Number * -(pcm/*F) (LW/k/g/cc) Doppler (Mwt) (gpm) (*F) (psia) (fts) (op) , ,
'la -NA 0.43' Minimum.
3425 Note 1. 590.1 2250 1092 440 Ib -
- NA 0.43' . Minimum '0 Note.1 557 2250 484 70 2a NA 0.0 Maximum 3425 ' Note 1 590.1 . 2250 1092 44,0 ,
.2b NA 0.43 .
- Minia'un 3425 Note 1 590.1 2250 1092 44b Ic NA 0.0 Maximum 3425 . Note 1 590.1 2250 1092 440 2d NA 0.43 Minimum 3425 Note 1 .590.1 . 2250 1092 440 i
1 377, o0o . 3 NA Note 2 Note 3 Of. diete 4 557 2250 484 , 80 .
? 377.000 4a NA , Note 2'* Note 3 Ot die 4e 4 557 2250 484 80 377.000 . ,
4b NA Note 2 Note 3 Ot ele &es.4 557 2250 484 ,80 388,880 - Sa 4 +7 NA Minimum 3425 44eto 4 - 590.1 2250- 1175 - 440, g - 388,820 , e Sb NA 0.43 Maximum 3425 Note-4 590.~ 1 - 2250 1175 440 c 388,'880 A Sc 4 +7 NA Minimum 3425 6 590.1 2250 1175 >440 sair,rgo -
$ '0.43 Maximum 3425 46ete 4 590.1 2250 1175 440' Sd NA -
- Transients are numbered according t[o the c,ases listed in Table 15.d.3-2. .
.t Subcr*itical .
.i .
. . 5 .
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' s .* p f - i '
. r fm
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TABLE 15.0.3-4 (Page 2 of 5) .
'* Summary of ~ Input. Parameters for. Accident Analyses
- _ ' ~
-~
. Initial .
Moderator Moderator NSSS RCS Vessel 'Pzr Pzr Water Feedwdter
. Transient Temperatur.e Density Output Flow. .
.T avg Press. - ' Volume Temperature ,
3 Number * (pcm/*F.) (%ak/k/g/cc) Doppler (Mwt) (gpm), ( F) (psia) . (ft ) - (*F) ., 377 .
@,000 '
6 , 4+7 NA Maxim 0m ~3651 59,4.3 2280 1175 ,. 445 377,000 7 '
=*& +7 '-
NA . Maximum ~ 3651 44e6o-4 594.3 ' 2280 1175 445 3 7 7,'0 0 0
- 8a =*& + 7 NA
- Maximum- 3651 6 594.3 2280 1175 445 377,000 8b =9+7 A .
Maximum 3651 41e&a-4 594.3 ,2280 - 1175 445
- 388,880 - -
9 . * & <+ 7- liA Maximum 3425 ,
'N 590.1 2250 1092 440
*
- 338,530 ,
10 .*fr + 7 'NA Maximum 3425 . tie &esa 590.1 - 2250 -1092.'
- 440
. 377,000 . ,
"^' ' 440 11 ,
*> + 7 N,A Maximum 3494' 594.1 2280 109,2-17 ( , d2. , 'e '
12 Mote 5 Note 5 Note'6 0 .Mi&pl60 ' 557 2250 NA NA 4 e 3gg,ggo 13a . - NA 0.0 Minimus Note 7 eiste=4 Note 8 2250 Note 9- Note 10 - 3FF,2fo . . 13b NA , 0.43 - Maxim'un N6te 7 41ste4= Note,8 2250 Note 9 Note 10 14'a NA NA . NA 3425 Note 1 590.1 2250 NA . NA- ,
.' 14b NA 'NA NA 3425- Note 1 590.1 2250-NA NA c - i
~
14c NA NA NA 3425 Note 1 590.1 2250 . NA ' NA ' . 1 ~ ,
. %* 3425 Note 1 590.1.
2250 NA NA 14d NA NA NA -
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- Transients are numbered cac'ording to the cases listed in Table:15.0.3-2. < . .
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O, , TABLE 15:0.3-4 (Page 3 of 5)
~
Summary of Input Parameters for Accident Analyses. - . u E ' - Initial f Moderator Moderator , NSSS RCS_ Vessel Pzr Pzr Water 'Feedwater g Transient Temperature Density Output , ' Flow T,yg Press. Volume Temperature 3 Number * (pcm/*F) (%ak/k/g/cc) Ooppler ,(Mwt) , (gpm) -(*F) (psia) (ft ) (*F) 909' ( 15 NA O.43' Minimum 23'98 - 282,300 580.3 2250 4,00 16a . NA NA .NA 3494 NA NA .NA NA- ,NA ,
= .
16b NA t4A NA . 3494 NA NA NA' NA. NA
., 16c NA NA .
NA . 3493 'NA NA ~ . NA NA NA ,
. . e 16d. - NA NA NA 3493 NA
~
NA, NA , NA . NA .
,~
- 4 377,ooo NA..'
17a Note 11_ Note 11 . Note 6 3493 46ete4- 594.1 , 2220 ' NA ' o
- . 173,+2a '
, 17b ' Note 11 Note 11 - Note 6 0 4inih;MS 557 .2220 , NA >
NA, ,
. 377.000
.17c . Note 11 Note 11 Note 6 3493 N 594.1 2220 NA . NA .
173,fzo *
, 17d Note 11 Note 11 ' Note 6 0 .,
9 Pihy M G 557 '2220 ,. . NA NA , 18 NA ' O.0 Minimus' 3425 Note l' 590.1 (250 J092 440 388,880 , 440 19
- ek + 7 NA- Maximum 3425 N 590.1 '2250 1092 h' ,2Sa ,NA tjote 12 , Note.12 Note 13 364,300' 594.'0" 2280 1172 440;
?, .1 '
Nots 13 '364,300 594.0 '*1172 44O ,- c' 20b NA
. Note 12 Note 12 2280 X - - . . .
364,300 2280 1172 440
% 20c . NA Note 12 Note 12 Note 13~ #94.0 J - '
- e -
+ t f ,
.
- Transients are numbered accordir?g to the cases listed in Table 15.0.i3-2. +
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= .
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+,
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TABLE 15.0.3-4 (Page 5 of 5) , y, Summary of Input Parameters for Accident Analyses
. Notes
- 1. The results provided correspond to an assumed initial reactor coolant ,
flow of 393,600 gpm. Technical Specifications require a measured flow of 388,880 gpm.. The difference between assumed and required flow has been justified and approved based upon the DNBR margin available and based upo'n an evaluation of.DNBR sensitivity to flow submitted with the McGuire Unit 2 Cycle 2 and Unit 1 Cycle 3 reload desigr1 transmittals. .
- 2. The reactivity coefficient assumed is a function of moderator density as described in sections 15.1'.4.2 and 15.1.5.2 (Figure 15.1.4-1).
,. 3. Refer to Figure 15.1.5-1.
The results prov d correspond an assumed i al reactor coo t flow of 386, The Tec cal Specific ns require a m ured flow M*fe4 "*f 'pm (gpm. non-ITOP valent flow 82,000 gpm). uations o repd 6. of transi388,88 sensitivitie nd justificat for the reduce equired me red fl ates were su ted and appro with McGuire U 2 Cycle 2 d t 1 Cycle 3 ~r oad design'tr ittals.
- 5. R'efer to Section 15.1.4.2.
- 6. , Consistent.with the lower limit shown on Figure 15.0.4-1.
- 7. In addition to the full power (i.425 MWt) case, sensitivity studies were
- Tun for partial power levels of 60% (2055 MWt) and 10% (343 MWt).4
- 8. In addition to the full power (590.1*F) case, sensitivity studies were '
run for partial power levels of 60% (576.0*F) and 10%'(560.2 F). , { 9. "In addition to the full power (1092 ft )3 case, sensitivity studies were run for partial power levels of 60% (848 ft )3 and 10% (544 fta), O
- 10. In addition to the full power (440*F) case, sensitivity studies were run for partial power levels of 60% (360*F).and 10% (241*F).
- 11. Refer to Section 15.4.8.2.2. -
- 12. The moderator density and doppler effects on reactivity during LOCA transients are accounted for in the evaluation models as described in Section 15.6.5.2 and the associated references.
- 13. The hydraulic analysis was performed using 102% of the_ Engineered
~
Safeguards Design Rating.. The core thermal transient analysis is performed using 102% of the NSSS licensed power. l l . . l - 1984 Update
9 s . Table 15.0.3-6 Values of Reactor Coolant Fjow Utilized In The Accident Analyses With . Without Improve,d Improved
' -Thermal Thermal Design Design
- Procedure Proceduro -
Reactor coolant flow per loop (gpm) , 98,400 46,666 94250 , . Total. Reactor Coolant flow (10 sib /hr) 146.1 N /40.0 i
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e - s . 3 .. Table 15.0.6-1 -
. Trip Points and Time Delays Ta Trip 1 Assumed In Accident Analyses l.imiting Trip Trip Point Assumed
- Time Delays Function . in Analysis (Seconds)
Power range high neutron .
~
flux, high setting 118% 0.5 , c. Power. range.hi'gh neutron 0.5 flux, low setting 35% High neutron flux, P-9 85% 0.5
'vertemperature O AT Variable see 4c68 /0.0 Figure 15.0.3-1 - . .
Ov'erpower AT Variable see . er# /0.0 ' Figure 15.0.3'1 ' High pressurizer pressure 2410 psig 2.0 Low pressurizer pressure
- 1835 psig 2. 0 <
. /" ' Low reactor cool' ant flow
(from loop flow detectors) -
87% loop flow 1.0 Undervoltage trip , 68% nominal 1. 5 e . Turbine trip Not applicable 2.0 Low-low steam generator 28.0% of narrow range 4.0 " 'l level level span High steam generator level , 87.4% of narrow range " 2.0 feedwater pump trip, . level span , feedwater isolation, .. and turbine trip .
** Total time delay (including ^d .
,; x : '-->"'-----'"';"-' ,
qm e , io. g -- u _.i
--.--i+r. RTD time response, and trip circuit, channel electronics delay) from the time the temperature difference in the coosant loops exceeds the' trip'setpoint until the rods are free to fall.
Asm lyses ? kco prah . me delay on this trip functioil was lengthened effective w' re 1
- Cycle 2 Cycle 2. The 4.0 second va.lue e new delay and was used in a reana.y ' transients y on this trip: Loss of -
,Non-emergency AC Power to the iaries (15.2.6), Loss of Normal )
Feedwater (15.2.7), er System Pip 5.2.8). The reanalyses with the n' are not presented here but,were foun acceptable res license amendments 42 (Unit 1) and 23 (Unit 2). 1984 Update
- 2. The primary to secondary leak rate is evenly distributed to the four )
steam generators, i.e., 0.25 gpm per generator.
- 3. The plant is operating with 1 perceilt d,efect.ive fuel. '.
- 4. Secondary' side activity as a result of primary to secon'dary leakage ~has
- reached an equilibrium when the ' accident occurs.
- 5. AllnobleTaseswhichleaktothesecond'arysidearereleased.
- 6. The steam generator blowdown rate'is 50 gpm.
- 7. The. turbine building steam leak rate is 1700 lbm/hr.
- 8. The condensate leak rate is 7510 lbm/hr.
The gland seal iodine partition factor and.the condenser air ejector ; '
~
- 9. -
. iodine partition factor are each 0.15.,
- 10. Ste'aifreleases, which are the only significant releases to the environment, terminate in 8 hours.
11.. The steam' generator iodine.partitioh factor during the accident is 0.1.
- 12. Water density at 180*F is used to convert condenser volumetric flow to mass flow. -
- 13. Other assumptions as detailed.in Table 15.2.2-1. - -
Based on the model in Appendix 15A, the thyroid and witole body doses are
- calculated at the exclusion area boundary an,d the low population zone. The
- results are-presented in Table 15.2.2-1 and are within the limits of 10 CFR 100, 15.2.2.4 Conclusions
~
Based on results obtained for the turbine, trip event (Section 15.2.3) and con-siderations described in Section 15.2.2.1, the applicable acceptance criteria for a loss'of external load event are met. ' 1 . 15.2.3 TURBINE TRIP
.15.2.3.1 Identification of Causes and Accident Description .
For a turbine trip ' event, the turbine stop valves close rapidly (typically.0.1 , sec.)onlossoftripfluidpressureactuatedbyoneofanumberofpossible; turbine trip signals as described in Sect, ion 10.2.2. , Upon initiation of'stop valve closure,. steam flow to the turbine stops abruptly. Sensors on the stop valves' detect the turbine trip'and initiate steam dump. The loss of steam flow results in an almost immediate rise in secondary system temperature and pressure with a resultant primary system transient as described in Section.15.2.2.1 for the loss of external loaq event. A more severe tran'sient occurs for the turbine trip' event due to the more rapid loss of steam flow w I l caused by the more rapid valve. closure.
. 15.2-4 1984 Update .
8" . . _ _ - - _ , _ . _ _ _ g
Y ( The automatic steam dump system would normally accomodate the excess steam generation. Reactor coolant temperatures and pressure do not significantly increase if the steam dump system and pressurizer pressure control system are - functioning properly. . If the turbine condenser were not available, the excess steam generation would be dumped to the atmosphere and main feedwater flow , would be lost. For this situation feedwater flow would be maintained by the Auxiliary Feedwater System to ensure adequate residual and decay heat removal capability. Should the steam dump system fail to operate, the steam generator safety valves may lift to provide pressure control. See 15.2.2.1 for a.further discussion of the transient. A turbine trip event is the most limiting of loss of external load, loss of condenser vacuum, and other turbine trip events. As such, this event has been' analyzed in detail.' Results and discussion of the analysis are presented in Section 15.2.3.2.. A turbine trip is cla'ssified as an ANS Condition II event, a fault of moderate frequency. See'Section 15.0.1 fon a discussion of Condition II events.
- 15. 2. 3. 2 -g . Analysis of Effects and Consequences
~~
Method of Analysis . In this ' analysis, the behavior of the unit is evaluated for a complete loss of , steam load from 100 percen.t of full power primarily to show the adequacy of- the ~ pressure relieving devices and also to demonstrate core protection margins. The reactor is not tripped until conditions in the RCS result in a trip. No O. credit is taken for steam dump. Main feedwater flow is terminated at the time of turbine. trip,~ with.no credit taken for auxiliary feedwater to . mitigate the , l consequences of the transient. l The turbine trip tran'sients are analyzed by employing the detailed digital com- - puter program LOFTRAN (Reference 3). The program simulates the neutron kinetics, l RCS, pressurizer, pressurizer relief and safety valves, pressurizer spray, l steam generator, and steam generator safety valves. The program computes pertinent' plant . variables ' including , temperatures, pressures, and' power level. -
- This acci' dent is analyzed with the Improved Thermal Design Pro'cedure as described .
l in Reference 4. Plant characteristics and initial conditions are discussed in
~ ~ Section 15.0.3. ~
Major assumptions are summarized below: > -
- 1. Initial operating conditions - initial reactor power, pressure arid RC5 temperatures are assumed to be at their nomin'al values. Uncertainties in initial conditions are included in the limit DNBR as described in 3 l Reference 4. .
- 2. Moderator and Doppler Coefficients of Reactivity 'the turbine trip is analyzed with both maiimum and minimum reactivity feedback. The maximum feedback cases' assume a large positive moderator density coefficient and the most negative Doppler power coefficient. The minimum feedback
~
g cases assume a positive moderator temperature coefficient and the least negative Doppler coefficient (See Figure 15.0.4-1). c . l 15.2-5 . 1984 Update
r .
- 3. .-
- e . .-: ,_,=
1 .
- s. _
Reactor Control - from the standpoint of the maximum pressure ~s~ attained < 3. it is conservative to assume that the re. actor i's in manual control. If e7 ) the reactor were in automatic control, the control rod banks would move ' prior to trip and reduce the severity of the transient.' ', ,
- 4. Steam Release - no credit is taken for th'e operatior) of the steam dump system or s' team generator power-operatgd rel.ief valves. The steam gen-erator pressure rises to the safety valve setpoint where st,eam release. "
- c through safety valves lipits' secondary steam pressure at. the setpoint
. value. .
- 5. Pressurizer Spray and Power.0pe g ted Relief Val.ves - two cases for both minimum and max ^imum reactivity Medback are analyzed: ,
, W
- a. Full credit is taken for the effect 'o'f pressurizer spray and power-operated' relief valves in reducing or limiting the coolant pies,sure.
Safety valves are also available.
- l ,
b.. No credit'is taken for the effect of pressurizet sprgy.and power. Tperated relief valves .in reducing or limiting the coolant pressure. ~ Safety valves Are operable. ,
- 6. Feedwater Flow
- main feedwater flow to t'he steam generators is assumed to ". -
be lost at the time of turbine trip. Noscredit is taken for auxiliary . feedwatet flow since a stabilized plant condition will.be' reached before~
~
auxiliary feedwater ' initiation is normal *ly assumed to occur. However, the
- auxiliary feedwater pumps 'would be expected to. start on a trip o,f the main feedwater pumps. The auxiliary feedwater flow would remove core decay
~ '
heat following plant stab,ilization. -
. e
- 7. Reactor trip is actuated by the first Reactor. Protection System trip set .
point reached. Trip signals are expected due to high' pressurizer pressure, -
- overtemperattire AT, high pr,essurizer water level, and low-low. steam '
generator water level.. No cr. edit is assumedlin* this analys'is for'a
.I ' '
reactor trip due to a turbine trip. ,, Plant characteristics and initia1 conditions are further discussed-' in'Section 15.0.3. .
~
Except as discussed above, normal reactor coolant. system and Engineered Safety - ,.
' Systems are not required. to function. 'Several cases are presented in which, '
pressurizer spray and power operated relief valves are assumed, but th.e more limitingcaseswhere3hesefunctionsarenot'assumedareal,sopresented. The Reactor Protec' tion System may be' required to function fo1 Towing a turbine trip. Pressurizer safety valves.and/oi steain generator safety valves may be required to open to maintain system pres'sures*below allowable limits. No single active failure will prevent ope, ration of any system required lto function. A discuss' ion of ATWT considerations is presented in Reference 2.
,s
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~ ' '
15.2-6 ._ - 1984 UP datt-' ,
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e . _ _ _ _ _ _ _ _ . _ . ,_ __.m ~ . . , , . , . _ , . . , _ _ , . _ _ . _ . , . _ -
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e f .. . Re s ul t's.. . . . ... The transien't tesponses for a turbine trip from 100 p g entaof full power . operation are.shown for four cases: two case's for minTmum reactivity feedback and two cases for maximum reactivity feedback (Figures 15.2.3-1 through 15.2.3-8).. . The calculated sequence of events for th,e accident is shown in Table 19.2.3-1. s ,
. , Figures 15.2.3-1 and 15.2.3-2 show the transient' responses'for the total loss ~
of, steam load with,'a positive moderator ' temperature coefficient ass,uming full +
~
credit for the pressurizer spiay and pressurizer power-operated relief valves. *
. !!o credit.is t.aken for the steam dump. The reactor is tripped by the high h
pressurizer pressure tWp' charinel. The minimum DNBR remains well above the s limit valye. *The pressuriger safety valves are actuated for this case and -
, maintain system pressure trelow 110 percent of the design value. .The steam -
~
- generator safety salves limit the secondacy stfam conditions to saturation at ~ .'
. the safety valve setpoint. .
** , s e . .
Figures 15.2.3-3 andh15.2.3-4 show the response for the total loss of st'eam
~
load with a large negative moderator temperature coefficient. All other plant paramete.rT are the,saae as the above. The DNBR increase's throughoLit the tran- .
., sientaand never. drops below its initial value. Pressurizer. power-operated .
relief valves and steam gener.ator safety valves preventGverpressiurization.in , , o primary a'nd second-ary systems, resp,ectively. The press.urizer safety valves , are not actuated for titis case.
- In t * '
~ ~forathis case, flow wouldthatcontinue under automatic control withterminated the reactor at~ ~
O- } re:he duced event power. -The operator feedwater would take action flowt is not erminate' the trapsdegt and.at the t . bring the pla'nt to a stabilized cond d. ion. If no ion'were taken by the op . ' ' erator the reduced power operation would continue until the condens'er hotwell .
,'was emptied. A low-low ste'am generator water level reactor trip would be gen-erated along with auxiliary feedwate'r, initiation sigrrais'. AuAiliary feedwater would then be used to remove decty h. eat with the.results less severe than those -
presented in Section 15.2.7, Loss of Normal %eedwater Flow.
~
4 The turbine trip accident was also studi'ed assuming the plant to be initially , operating at full p'ower with no credit tjken for the pressurizer spray, pres-surizer power-operated relief va'Ives,' of steam dump. The reactor.is tripped on *
- the high pressurizer pressure signal.' Figures 15.2.3-5 and 15.2.3-6 show the
- transients with a positive moderator temperature coefficien}. The* neutron flux remains es'sentjslly constant at ful1 power until- the reactor,is tripped. The DNBR, neven goes below the initial value throughout the transient. Inthiscasg.'
the pressurifer safety valves are actuated, and maintain system pressure below .
~
110 p'erchnt o,f the. design value. ',.*. . ,' ~ -
~
<. Pigures 15.2 3'-7 and 15.2.3-8 are.the transients with Inaximum reactivity feed- a back with the other* assumptions being the same a'!i in the preceding case.' .'
Again,.the DNBR increases thr,oughout-the transient and the. pressurizer safety
.- v,alves-are actuated 'to limit pr1. mary pressure. . .
e -
. Refer ~ence 1 presents' additional results of an analys'is fo'r a complete loss of .
J heat sink including ' loss of main feedwater. This. analysis shows the overprg)sure a ' protection that 'iscafforded by the pressurizer and steam generatdr safety valves. , , s 4 15.2-7 ' 1984 Update o . .
- s 8n n m a y
s - -
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~
Environmental' Consequences I'
. 15.2.3.3 . , .
<.- ,s . . e ical con'sequenees resulting, from a'tmospheric steam d' ump will be .
. Th'e radiolpg'than the steamline break event analyzed in Section 15cl.5.3 less severe - sin
- no-fueldandeispostulatedtooccur.'. '
.m n . .
+
15.2.3.4' Conclusions , g- . . e .! 3 . . . Results o'f the anal'yses, includingthose in Reference 1, show thatt'he plant
, design is ;such that a turtting trip presents no^ hazard to the integrity- of thef
~
RCS or the main steam system. Pressure relieving devices ' incorporated in tifb two systems are adequate ~to' limit the m,aximum pressures to within the degign
- s
. -limiter ~
gy , The integrity of the core is gain'tained by operation of the Reactor Protection , e DN8R will be-maintained above ,the limit value. The appligable - - System,,,i.e..,@i acceptancefr
- eria as listed in .Section 15. 1 have been met. The above s .
analysis d6monstrates the ability of the NS to safely withstand a full lodd ' rejection,3 . 15.2.4 . INADVERTENT CLOSURE OF MAIN STEAM' ISOLATION'. VALVES . Inadvertent' closuri of the main steam isolition valves wouliresuit its a' turbine trip. Turbine , trips 'are discussed in Sectioh 15,.2.3. . b +" 15'.2.5 LOSSOFCONDENSERVACUUhANDOTHEREVENTSCAUSINGATURBINETRIP , h Turbine trip init'iating events are described in Section 10.2.2.1.oss
~
~
A loss' of d'f conde
.. condenser. vacuum would preclude the us'e of steam dump t5'iihe cgndesser. .Howeper, since steam dump is assumed not to be available in the turbine trip analysis,~
no additional adverse effects kould result if the turbine trip were caused by loss of~ condenser vacuum. Therefore,- the analysis results pnd conclusions contained in Section 15.2.3 apply to -loss of condenser vacuum. In addition, J. g,,
- analyses for the otherJpossible causes of a , turbine trio, as listed 'ih' Section 10.2.2 are. covered by'Section 15.2.'3. Possible.overfrecuency effects cue .: a. ,,
y - ' turbine overspeed condition are-discussed in Section 15.2.2.1 and are not a.- , co'ncern for this type of event. " ' 25.2.6 LOSSOFNON-EMERGENCY.ACPOWERTOTHfSTATIONAUXIL.IARIES -
. 15.2.6.1 Id ntification of Causes and Accident Descript'i'on g
.. A complete loss of'iion-emergency AC power may result in the , loss of al.1 power '
. 'to the plant auxili' aries, i.e., the reactor coolantepumps, condensate' pumps, .
etg The loss of power may be caused by a compfete loss of the offsite grid
,' accompanied by a turpine generator trip at the station,'or by a loss of thE ,.
onsite AC dis.tribution system. ,' ,
' * ~
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f @ ', 1 * . 15.g-8 a 11984 Up.date. 4 s se
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..- - , e Thls transient is more severe than.the turbine trip ev.erlt analyzed in Section' 15.2.3 because for this case the decrease in heat removal by the secondary .c system is accompanied by a flow coastdown which further reduces the capacity-of tHe primary coolant to remove heat from the ctre. -The reactor will trip (1) upon reach.ing one of the trip t setpoints ir} the primary and. secondary systems .
as a result of the flow coastdown and decrease ~in secondary heat removal or (2) due to loss'of power,.to the control rod dr'ive nechanisms as a r_esult of the,
- loss of power to the plant. .
r '.
. Following a loss of AC' power with turbine a'nd reactor trips, the sequence des-
~
=
cribed.below will occur: '
- 1. Planh.vitalinstrumentsaresuppliedfromemergencyDCpowersources. ~
. t
- 2. As the steam system pressure rises followin( the trip, the steam generat.or .
*~
power-operated relief valves may be automatically ' opened to the atmosphere.
, The condenser is assumed not to be avpilable for steap dump. If the steam '-
flow rate through the relief valves is not'avdilable, the steam generator safety Valves may li'ft to dissipate the sensible heat of the fuel and - coolagt plus the residual decay heat produced in the reactor.
* ' 3 .' As the'no load temperature is approached, the steam generator power" ' '
operated relief. valves (or safety valves, if the relief valves are
. not available) are used to dissipate the residual decay heat and to '
maintain to plant at the hot shutdown condition. 4.- The standby diesel gen'erators, started on loss of voltage on the plant
~
emergency busses, begi,n to' supply piant vital loads'.
~
- The Auxiliary Feedwater System is started. automatically as follows: ,
Two motor-driven a.uxiliary feedwater pumps are started on iny of the fo'llowing:
, a. Low-low narrow range level in any st'am e generator .
% (.
"I I b. Any safety injection signa'l ,
- c. Loss of offsite power , .
r .
- d. Tcip of s11 main feedwater pumps s ,
- e. -
Manual actuation - The one turbine-driven auxiliary feedwater pump is started on any of the . i follo#ing: .. .
- a. . Low-low level in any*two steam,cjenerators. *
- b. Lq,ss of offsite power ,
l *
- l
- c. Manual act,uati6n .
+-
l .: m e , t 1 - 15.2-9 198!. Epdate. i.
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- p. ,
= , ,
The motor-driven auxil,iary feedwateripumps are su'pplied poser by the. diesels ) and the turbine-dr'iven pump utilizes steam from the secondary system. Both
. type pumps are designed to supply , rated flgw within one minute of the initiating .
signal even if.a. loss of all non . emergency AC power occurs simultaneously with. loss'of normal feedwater. The t'urbine exhausts the secondary steam to the atmosphere. .The pumps,normally take suction from the upper surge tank foi-
" delivery to the stgam generato,rs, although they c'an be realigned to take ,
l suction from the auxiliary feedwater condensate storage tank if necessary. . Upon the los's .of power to the reactor coolant pumps, coolant flow necessary for
. core cooling and the' removal of residual'he'at is maintained by natural circu-
, lation in the reactpr coolant' loops. "
~
This ev,ent, as described above, is a more-limiting event than the turbine trip initiated decrease in secondary heat removal without loss of AC power, which was analyzed in Sectioii15.2.3. However, a loss of AC power to the plant auxiliaries as postulated above could result in a' loss of normal feedwater if the condensate pumps lose their power sOpply'. , Followi'ngT he reactor coolant pump coastdown caused by the loss of AC power, , the natural circulgtion capability of the RCS will remove residual and decay .
- heat from the core,' aided by auxiliary fe'edwater in the secondary system. An analysis-is gresented here to show.that the. natural circulation flow in the RCS ,
fol. lowing a* Toss of AC power event is sufficient to rgmove residual heat fro.m the core. A loss of non-emergency AC power to the station auxiliaries ~is classified as an ANS Condition II event, a fault of moderate . frequency. See Section 15.0.1'for
~
- i. Ta discussion of Condition II evegts.
~* '
l , . 'l , . l 15.2.6.2 Ahalysis of Ef.fects and Consequences , Method of Analysis' '
'A detailed ana'ysis l using the LOFTRAN Code (Reference 3) is perform'ed to obtain o the natural. circulation . flow following a station blackout. The simulation o ,
describes the plant thermal kinetics, Reactor Coolant System (RCS) including natural citculation, the pressurizer, steam generators and fsedwater system..
, The digital p'regram computes pertinent variables including the steam generat6r level, pressurizef water level, and reactor. co,olant ~ average temperature.
a ' The assumptions used in the analysis are as follows: , ,, , e
- 1. The plant is initially operating at 102 percent of the engineered safety ~
e features design rating. - e
- 2. A' conservative. core ' residual heat generation is assumed based upon long term operation at the initial power level preceding the trib. ,
. 3. A heat transfer coefficie'nt in the steam generator associated with RCS natural circulation following reactor coolant pump cogstdown.
. )
- a s 15.2 10 -
198,'. Update y . . , !- p .
. 1
- e. -
- 4. Reactor trip occurs on* steam ' generator low-low narrow range level. ,No
- credit'is taken for ilipediate release of the control rod drive mechanisms l caused by a loss.of offsite power. ,
- 5. ' The worst single failure in the auxiliary feedwater system occurs.
- 6. Auxiliary f'eedwater is delivered to' four steam generators.
- 7. Secondary ' system steam relief is ai:hieved through the steam generator safety valves. .
. .The assumptions used in the dnalysis are essentially identical to the loss of normal feedwater flow incident (Section 15.2.7) except that power is assumed to be lost to the reactor coolant'pi,1,mps at the time of reactor trip. .
Plant characteristics and initial conditions are further discussed in Section
~
15.O.3. , A Results~ t .
~
, The transient response of the RCS following a loss of .AC power is shown in Figures 15:2.6-1 and 15.2.6-2. The calculated sequence of events for this event is listed in Table 15.2.3-1.
, p The first few seconds after the -loss of power to the reactor ggolant pumps will .
closely resemble a simulation of the complete" loss of flow in5? dent,(see .
, ' Section 15.3.2)7 i.e.( core damage due to rapidly increasing core temperatures is prevented tfy promptly tripping the reactor. After the reactor trip, stored
** an4 residual decay heat must be remo,ved.to prevent damage.to eithe.r the RCS cr the core.
The LOFTRAN code results show that the natural circulation flow available is sufficient to provide adequate core decay heat removal following reactor trip
. and reactor coolant pump coastdown. -
r s . 15.2.6.3 Environmental Consequences . - l Since steam dump to the condenser is assumed to be lost, ' heat removal fr'om the j secondary system would occur through the steam generator power-operated relief j valves or safety valves *. Since~no fuel damage is p,ostulated to occur, radiolo-l , gical consequences' resulting from this transient would be less tevere than the l- , steamline break accident analyzed in Section 15.1.5.3. \ , 15.2 4.4 Conclusions- , Analysis of the natural cir'culation capability of the Reactor Coolant System has demonstrated that sufficient heat removal capability exist's following -
'. reactor coolant pump coastdown to prevent fuel or, clad damage. , ,
l t . t .
- . . e 15,2-11 . 1984 Update
~~ -
, , 4 15.2.7 , LOSS OF NORMAL FEEDWAT.ER FLO.W- )
Identification of Causes and Accident Description i
~15.2.7.1 A loss of normal feedwater (from pump failures, ' valve malfunctions, or loss of .
offsite AC power) results in a reduction in capability of the secondary system , to remove the heat generated in the reactor core. Ifanalternativesupplyo{ ' feedwater were not supplied'ta the giant, core residual heat following reactor trip would heat the primary system water to the point where water relief from . the pressurizer would occur, resulting in a substantial. loss of water from the Reactor Coolant System (RCS). ,Since the plant is tripped well before the steam
- generator heat transfer capability is' reduced, the primary system variables,.
- never appioach a DNB condition.
The following occur upon loss of normal feedwater (assuming main'feedwater pump failures or valve malfunctions): ,
- 1. As the steam system pressure rises following the trip, the steam generator
, power-operated relief valves are. automatically opened to the at',mosphere.
Steam J ump to the condenser is assumed not to be available. If the ste,am flow rate through the relief valves is not available, the steam gen'erator
- saf6ty valves ma.y lift to dissipate the sensible he'at of the fuel and
~
1 coolant plus the residual decay heat produced in the reactor.
- 2. As the no load temperatum is approached, the steam generator power- ,
'Sperated relief valves (or safety valves, if the relief valves are not available) are used to dissipate.the residual decay heat and to. maintain the plant at the .hi)t shutdown c,ondition'. ,
The Reactor trip on low-low narrow range level in any steam generator * provi' des the necessary protection against a loss of normal feedwater. The Auxiliary Feedwater System is started a.utomatically as discussed in Section 15.2.6.1'. ,The turbine-dPiven auxiliary feedwater. pump utilizes steam from the secondary system and exhausts to the atmosphere. The . motor-driven auxiliary - . , feedwater. pumps are supplied by power'from the diesel generators. The pumps take suction dirpctly from the, upper surge tank for delivery to the steam generators, but .can be realigned to the auxiliary feecwater condensate storage , tank if.necessary. r .
- An analysis of the system transient is presented below to show th'at followin~c, a '
loss of normal feedwater, .the Auxiliary Feedwater System is cap'able of removing.
,the stored and residual heat, thus preventing either overpressurization of the ,
RCS or lo,ss of water from the reactor core, and returning the plant to a stabilized, condition. , L .s A loss of. normal feedwater is classified as an ANS Condition II event, a fault of moderate freq0ency. See Section 15.0.1 for a discussion of Condition II events. -
)
..
- A 15.2-12
- 1984 Update "
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( *
, 15.2.7.2 Method of Analysis-Analysis of Effects and Consequences
~
A detailed analysic using the LOFTRAN' Code (R'eference 3) is performed in order to obtain the plant transient following a loss of normal feedwater. The.simu-lation deseribes the plant thermal ki,netics, RCS including'the nat, ural circ,u- , lation, pressurizer, steam generators dnd feedwater system. The digital , program computes pertinent variables including the steam generator level, , pressurizer water level, and reactor coolant average temperature. AssumptNnsmadein-theanalysisare: ,
. 1. The plant is initially operating at 102 percent of the' engineered afety s features design. rating. .
- 2. A conservative core residual heat generation is assumed based upon long term operation at the initi-al poyer level preceding the trip. -
- 3.
- Rea'dtor trip occurs on steam gene'rator low-lo'w narrow range '1'evel.
- 4. The worst' single failure .in the auxiliary feedwater system occurs. - -
Auxiliary feedwater is delivered to four steam generators.
- 5. .
- 6. Secondary system steam relief is achieved through the steam ggnerator safety.valvess '
j + The-loss.of normal feedwater analysis is performed to demonstr~ ate the adequacy !. of the reactor protection and engineered safe' guards systems (e.g.,-the ,Auxi.liary Feedwater System) in removing long term decay heat and preventing exc'essive heatup of the RCS with possible resultant .RCS overp'ressurization or loss of RCS - water. As such, the assupption's used in this analysis are desigrfed to minimize
. the energy removal capability of the' system and to maximize the.ptssibility of.
water relief from the coolant system by' maximizing the coolant system expansion. - as noted in tMe assumptions listqd above. ' For the loss of normal feedwater transient,'the r'eactor coolant volumetric flow '* remains at its normal value and the reactor trips via the low-low steam genera, tor*
., nar' row range level trip. The reactor coolant pumps may be manually tripped at i
, some later time to reduce , heat addition to the RCS.
P iAn additional assumption made for the loss of normal feedwater evaluation is that only the pressurizer safety valves are assumed to function normally. Operation of the valves maintains peak RCS. pressure at or below the actuation "" setpoint(2500' psia)throughoutthetransient. Plant characteristics and initial cond'itions are further discussed in Section
~
15.D.3. , Normal reactor control systems. are 'not requir.ed .to fun'ction. The teactor . i ,, Protection System is required to' function following a loss 'o.f normal feedwater e l' '
,as anal yzed here. The Auxiliary Feedwater. System is required'to deliven a ,
? ,,, . .s - y* $
" ~
15.2-13 1984-Update
- o
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' ~
O g o'* - minimum a,uxiliary feedwater flowrate. No single active failure will prevent ) > operat' ion of any system. required to function. A
- discussion of ATWT consid-erations is grisented in Reference 2.
Results . ,'
- e. . .
Figures,15.2.7-1 and 15.2.7-2 show the significant p,lant parameters fo,llowing a loss of normal feedwater. - e - , Following the reactor and turbipe trip from full load, the water level in the
, steam generators will fall due to the
- reduction of ste~am generator void fract-tion and because steam flow through the safety valves corltinues to dissipate the s,tored and decay' heat. One minute following the. initiation of the low-
- 4 1pw level trip,.at least two auxiliary fledwater pumps.are automatically star- , . , ted, ' reducing the rate of water. level decrease.' -
The capacity o,f the auxiliary feedwater system is such that; the water level in *
. . the steam generators being fed does not' recede below the lowest level at which
~
sufficient heat transfer are% is available to dissipati core residual heat
-withoyt gter relief from the RCS s'afety valves.- Figure 15.2.7-1 shows that at no time is there water relief fr,om the pressurizer. -
"Thecalculatedsequenceofeventsforf.hiIaccidentislistedinTable15.2.3-1.
As sh_own in Figyres.15.2.7.-1 and 15.2.7'-2, the plant approaches a stabilized" condition following reattor trip and auxiliary feedwater initiation. Plant procedures may be followed to further cool down the plant; -
~
15.2.7.3 - Environmental Consequences
,Since steam dump to the condenser is assumed to be lost, heat removal from'the secondary system would occur thfough the steam generator power-operated relief valves or safe'ty valves. Since no fuel damage i.s postulated to occur, radiolo- ,
gical consequences resulting from this transient would be less severe than the srteamline break accident analyzed in Section 15.1.5.3. , 15.2.7 4 ' Conclus' ions . Results of the analysis show thaf a loss of normal feedwater does not adversely affect the, core, th'e RCS, or t'he. a steam system sirice the auxiliary feedwater. .
. capacity is-such that reactor coolant water is not relieved ifrom the pressuriz,er relief or safe,ty vjtives.
15.2.8 FEEDWATER SYSTEM PIPE B'REAK , 15.2.8.1 Identification of Causes and ; Accident Description 4 A ' major feedwater line rupture is defined as a break in a' feedwater line large enough to prevent the addition _of sufficient feedwater to the steam generators to maintain shell side fluid inventory in the steam generators. If the break . . is postulated in.a feedline b'etween the check valve and the steam generator, fluid from the' steam generator may also be_ discharged thro. ugh the break. Fur-ther, a break in this location could preclude the subsequent addition of aux- ~ c iliary feedwater to the affected steam generator. '(A break upstream of the g feedline check valve would*affe,ct the Nuclear Steam Supply System only as a. loss of feedwater. This' case is covered by the evaluation in Section 15.2.7). 15.2-14 - 1984 Update n 1
.. '.o .
,1
' ~ . .
~
k - Depending upon'the size of the break and the plant' operating conditigns a't the time of the break, the break could cause either a RCS cooldown (by excessive
- energy discharge thr,ough the break) or a RCS heatup. Potential RCS coo 1down .' ,
resulting fr'mo a se*condary pipe rupture is avaluated in Se'ction 15.1.5. Therefore, ohly the RCS heatup effects are evaluated for a feedwater line. - rupture. . A feedwater line rupture reduces the ability to remove h' eat generated.by
- the core from the RCS for the following reasons:
~
- 1. Feedwater flow to the steam generators is reduced. Since'feedwater is subcooled, its loss may cause reactor coolant temperatures to increase '
prior to" reactor trip. , Fluid in the steam generator inay be discharged through the break, and 2.
~
woQld then not be available for decay heat removal after tripr+ -
- 3. The break may be large'enough to preven} the addition of any main feed-.
wate,r "af ter, trip. , , s
. An Auxiliary Feedwater System is provided to ensure that adequate feedwater will be available such that: g
- 1. No substantial overpressurization of the'RCS shall occ' ur.- -
'2. Sufficient liquid in th'e RCS sh'all be maintained in order to provide
',O-' adequate decay heat removal.
~
. Secti%n 10.,4.7 contains a description of the Auxiliary <Feedwater System
? .
' interfaces.
The sever'ity of the feedwater line rupture transient depends on 5' number of ~ a system parameters including break' size, initial reactor power, a'~d n credit.'~
-taken for the functioning of various control and safety systems. A number of cases of feedwater line break have.been analyzed. Based on these analyses, it has been shown th A the'most limiting feedwater line ruptures are the double '
ended rugture of the . largest feedwater line, occurring at full power with and
. without Toss.of.offsite power, with.no' credit
- taken.for pregsurizer control. *
,These cases are analyzed below. . -
~
Thee fellow'ing provides the necessary protection for a main feedwater rupture:
- 1. A reacto.r trip on any of the folTowing conditio'ns: .
> . t . , Hi.gh pressurizer pressure.
. a. < -
- Overtemperature AT.
- b. .
.. 4 p .
- c. Lo.walow narrow range level in any steam generato'r. ,
e
- d. Safetf injection siQnals .from any ok the afollowing: ,
s
- a e .
15.2-15 + 1984 Update r . f m _
O r
~ c .
9
- c l
l 2/3 low steam line pressure in any loop.' ' 1.
)..
2/3 h'igh containment pressure i
-- ,.2.
* %t _
'(Chapter 7 contains a description of the actuation system).
' 2. An Aviliary Feedwater System to provide an, ass'ured source of feedwater to '
the steam generators for decay heat removal. . (Section 10.4.7 contains a description of the Auxiliary Feedwater. System)~. , 4 A major feedwater line rupture is classified as an ANS Condition IV eveht, a. limiting fault. See'Section 15.0.1 for a discussion of Condition.IV events.
-15.2.8.2 Analysis of Effects and Consequences .
Method of Analysis .
- e .,
,m._. A detailed analysis using the LOFTRAN Code (Reffrence 3} is performed in order ~
- to determine the plant transient following a feedwater line rupture. The code describes the plant thehal ginetics, RCS including nat0ral circulation, pres.-
surizer,'Tseamgeneratorsandfeedwater: system,andcomputespertinentvariables , including' the pressurizer pressure,,
- pressurizer water level, a,qd reactor. -
coolant.averag,e temperature.
.The cays? analyzed assume a double ended rupture of'the largest feedwater pipe .
at ful' power. -Major assumptions riiade in the an'alyses are as fall.ows: l'T" The plant is initiarly operat.ing af. 102 percent of engin ered sifety s features design rating. ,, Initialreactorcoolant'averaghtemperaturesis 4.0'F above the nominal
~
. '2.
- value, and the initial presstarizer pressure is 30 psi abotte -its ' nominal value.
- e .
s
- 3. No credi.t.is'.taken for the pressurizer power-operated relihf vilve's -or ;
pressurizer spray. -
= . ~
- 4. Initial pressurizer level is at the nominal programmed value plus 5 per - .
cent (error); initial steam generator water' level is at the nominal value . plus 5 percent in the faulted ' steam generator and at the nominal valve *, minus 5 percent in the intact' steam generators.
~
- 5. No credit is taken for the high pressurizer pressure reactor trip.
.. e, , ,
6'. ' Main feedwater flow to all steam generators is4 assumed to,be lost at the <
-time the break occurs (all. main feedwater spills.out through the break). ,
- 7. The worst' possiWe break ' area is assumed. This maxinlires the blewdown
' dischange rate fgliowing'tihe time of trip, which maximizes the resultant h'eatup of the reaqtor dbolant. ,
- 8. A conservative feedline break discharge quility'is,, assumed: prior to the. -
time the reactor trip occurs, thereby maximizing ttie time unti1-the trip ~
, ay setpoint is reached. After the trip occtrs, a'. saturated liquid discharge *
. -15.2-16 -
1984 Update. *
-
- c ~ -
g
** ^ - -
e .
-4 is assumed.until all the water inventory is discharged from the' affected steam generator. This' minimizes the heat removal capability of the a, m i affected steam generator. .
. .; . v. . . ~
- 9. -Reactor trip is assumed to be initiated at'10 percent of narrow range span bel h the low-low level trip setpoint in'the faulted steam geherator.,
- 10. The ' Auxiliary Feedwater System is actuated.by the low-low steam genera' tor *
.. water level signaf. lte Auxiliary 4Feedwater System i's'assu~med to s'u pply a
. total of 450 gallons per minute (gpm) to Wy three unaffected steam gen-erators, including allowance for possible spillage.thrqugh the main feed- -
water line break. A 60 second delay was assumed following the low-low - level signal to allow time for startup of the emergency diesel generators and the auxiliary feedwater , pumps. An additional 132 seconds was as'sumed before the feedwater lines were purged and the relat,ively cold (IleiF) .- o -appiliary feedwater entered the unaffected steam generators. s . .
~
- 11. No credit is taken for. heat energy de* posited in RCS' metal during the RCS .
heatup. .
.J ,
's -
No credit, is taken for charging or Tetdown.
,12. , ,
. ~
- 13. Steam generator heat transfer area is assumed to decre'ase as the.shell - -
side liquid. inventory decreases. , Genteaunene Eore residual heat generation is assumed based upon !r; t-- 14.
^ ^ ^ -" i L
-t t'r #- Mf:! ;E r .! n :! ; :--"4-- tt: t-fp. -
~* -
~
- 15. .No credit is taken for the following potential protection lo 'c signals to :
mitigate the consequenc'es of the accicent: ,
- = a. '
High pressurize'r p*ressur.e. "
- b. Overtemperature AT. -
%e l9W vershm of AMS -5.l '
(Reference 5). ANSI
~
- c. High pressurizer level.
's a conseinfiv.e, re'/ANS-5.1-19 preseda.fb
.d.
Hig4Containmentpress@e.
*~ .
' of de clecay ^ energy re(edse rufes.
~
Receipt of a low-low st'eam* generator na'rrow# range. level signal in at'least one ~ 4 team generator starts the motor-driven auxiliary feedwater pumps, which then deliver auxiliary feedwater flow to the steam generators. The turbine.dtiven
~ auxiliary feedwater. pump is initiated if the low-low s. team generptor water *
.sigRal is reached in at least two steam generators'. Similarly, receipt of a l'ow steam line pressure signal in at least one steam line initiates a steam
- . line isoldtion signal which closes'the main steam line isolat' ion valves in all steam lines. This signal also gives afsafsty injection signal which initiatqPs, 4 -
flow of. borated water into the RCS. The amount of safety injection flow is a' ~
. func-tion o,f RCS pr' essure. ,
~
., Emergency operating.proced@es following a secondary system line rupture > call
,fpr the following actions to.be taken by the reactor operator:
- 1. Isolate f'eedwater f.lew spillfng out the break of ruptured steim generator
. and align system so level in intact steam generators recovers,
O ~ 3 . . . 15.2-17 , 1984 Update
. , s
+ w -
m . . .
r .
- 2. Stop the. safety injection pumps if: -
' ,) -
, a. . Wide. range' reactor coolJnt pressure-is. increasing, ,
w . P - b. Pressurizerwateflevelisonspan, . ,
'c. The RCS is adequately subcooled, and . -<
. - d. Steam generator narrow range level indication exists .in at least one feedwaterisbeinginjected
- + ,into steam generator or sufficient the steam generators to provide an auxil.iary' adequate heat s.in,k. '
m t . c Sub. sequent to recovery of-level in the intact steam generators,' the plailt , oper.at,ing procedures will be followed in cooling the plant to' hot shutdown conditions. Plant cNaracteristics and initial conditions are further discussed in-Section -
.15.0.3. ,
No reactor control sy' stems are assumed to function. The Reactor Protection
" System is fequired to functioh following a feedwater line rupture as analyzed
. here. N.o single active failure will prevent oper'ation of this system. '
A p c discussion of ATWT consideratio'ns is presented in Reference 2. . The engineered safety systems assumed to function are the Auxiliary Feedwater System and the Safety Injec. tion System. For-the Auxiliary Feedwater System, , theworst, casec,onfigurationhasbeenused,j.e.,threeintactst'eamgenerators~ . ,
- b. . receive auxi.liary feedwater following the break. The turbine-dt.iven, auxiliary <
feedwater pbap has-been assumed to fail; the two'inotor-driven pumps together a deliver.450 gpm to the three intact steam generato'rs allowing for.. spillage out ' r of the break. Only one train of safety injection has been assumed to be available. A detailed <lescription and analysis of_ the ' Safety Irijection System is provided in Sect'lon 6.3. The Auxiliary' Feedwater System ' is described 'in ' Section 10.4.7. For'the case without offsite power there will bh a flow coastdown until flow in ( . the loops reaches the'natusal circulation value'. The natural circul'atfon cap- -
** ability"of the RCS 'has been shown in Section 15.2.6, 'for the loss off,,,AC ngn- .
L emergenc power tr.an'sient, to be sufficient' t'o remove core decay heat' folT,owing
. reactor trip. Pump coastdown characteristics are3d emonstrated'in Sections
- 15.3.1 and 15.3.2 for single and multiple reactor coolant pump trips, respec-
~
tively. .
, Results ,
- i _
l Caltulated plant parameters following a majoi feedwater line rupture are shown *
- in* Figures 15.2.8-1 through:15.2.8-8.. Results f6) the case with offsite power ,
available are presented in Figures 15.2.8-1 through'15.2.-8-4. Results for-the o- case where'offsite power is lost are presented in Figures 15.2.8-5 through 15.2.8-8. The calculated sequence of events for .both cases analyzed are 14ted
-in TEble.15.2.3-1. .,
. The , system response following the feedwater 1,ine rupture is similar for both cases analyzed. RFsults presented in Figures 15.2.8-2 and 15.2.8-4 (with off- -)
site power available) and Figures 15.2.8-6 and 15.228-8,(without offsite power) l ,
~
15.2-18 ' 1984 Update - sp .. -4
s . ( - show th'ta pressures in the RCS and main steam ' system re' main below 110 percent ~ of the respective design pressures. Pressurizer pressure increases until re-actor trip occurs on low-low steam generato'r narrow range level. Pressure then decreases, due to the los's bf heat input, until the Safety Injection System is ' actuated on low steam.line pressure .in the ruptured loop. Coolant expansion '
." occurs du.e'to reduced heat transfer capability in the steam generators. The
~
pressurizer. safety valves open to maintain primary pressure at an acceptable
. value. Addition of the safety injection . flow aids in tooling down the pri, mary '
and helps to ensure that sufficient fluid exists to keep the core covered with water. Figures 15.2.8-1 and l'5.2.815 show .that folloying reactor trip,. the plant -
' remains subcritical. -
' t .
RCS press 6re will,be maintained'at the safety valve setpoint until safety in- - jection flow is terminated by the. operator, as mentioned in.Section 15.2.8.2. b .The reictor-core. remains covered with water throughout the transient, as water
.re' lief due to thermal expansion is limited.by the heat removal capabil~ity of -
the Auxiliary Feedwater Syst,em and makeup is'proVided by the Safety Injection ,
. System.'1 ..
The major difference between t'he two. cases analyzed can be seen,in the plots of - hot and cold. leg temperatures, Figure 15.2.8-3 (with offsite power available) and Figure 15.2.8-7 (without off. site power). It is apparent from the initia'l -
. portion of the transient (*200 seconds), that?the case without offsite power .
results in higher temperatures in the hot leg. For longer times, however, the - _ case with offsite power results in a more severe rise'in temperature unti.),the O~ : coolant pumps are turned off 'and,the Auxiliary Feedwater System is realig~ned. The pressurizer fills for the case with power dde to the increased coolant expansion resulting from the pump heat addition. Hence, water is relieved for the case with power. As previously. stated .however, the core remains cgvered
'with watdr for b6th casds. I 15.2.8.3 '
Environmental Consequences . . . o .. . , The feedwater. line bre~ak
- with the most significant consequences would be one -
that occurred intide the'containme'nt between a steam generator'and t'he feed- ' line check valve. In this case, the contents of the steam generator would be . released to the containment. Since no fuel failures are postulated, the .
. radioactivity. released would be less than that for the steamline break, as- -
analyzed i'hSection 15.1.5. Furthermore, automatic isolation of the contain , ment would furth~er. reduce any radiolqgi. cal consequences from this postulated , r accident. - *
~ '
o 15.2,8.'4 Conclusions . Results of the analyses show that for the postujated feedwater line rupture, a l the assumed Aux-iliary Feedwater System capacity is' adequate to r.emove decay heat, to prevent'overpressurizing the RCS,' and to prevent uncovering"the re- , actor core. Radio' activity doses fro'm the postulated feedwater line rupture are
. le'ss' than -those previously prpsented for the postulated steam line break. Afl
- , applicable acceptance criteria are therefore met. ,
~
(
- 1984 Update.
9- . 15.2-19 ,
~
t . .
. v, l
REFERENCES FOR SECTION 15.2 . N
, 1. Mangan, 4 A., " Overpressure Protect. ion for Wes.tinghouse Pressurized . I Water Reactors", WCAP-7769,-October 1971. /" ,,, , -
- 2. " Westinghouse Anticipated Trangients Without Trip Analysis", .
WCAP-8330, Au' gust 1974.- , 3'. ""---++ ' " ' , :t ' ,
"' 0F~" 2;d; 0;;; !;ti;r", "Cf ?$7[
. t
- w. ,
- 4. Chelemer, H., Boman, L. H., Sharp, D. R., " Improved Thermal, Design '-
Prbcedures",WCAP-8567, Jug 1975, Natida.1 Studard for ' Decay
- 5. , ANSI / ANS -!5 .1- 19H ,, Americci Heat Po wer i n 13 3 Lt Wa tee Reac.tes " Augaf -i9 79 e ,
Burnett, T. W. T., et.al., "LOFl'RAN Code Description," WCAP-7907-P-A (Proprietary),
- WCAP-7907-A (Non-Proprietary), Apr.il 1984.
)
. e .
e . er s ,
'b
~
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= v ,
e - - a
. + '.
ci . 15.2-20 198.4 Update . 0 0
. . .4 o
~
9
, . - Table 15.2.3-1 (Page 1 of 5) '
. Time Sequence Of Events For Incidents Which Cause A* Decrease In Heat Removal By The Secortdary Systems .
r Accident - Event Time (sec) Turbine, Trip- ,- , . -*+ ,.
- 1. With pressurizer Turb.ine trip, loss. 0.0 -
control (minimum of main feed, flow , reactivity feedback) , ' High pressurizer pressure .J 3 , reactor trip setpoint - { . reached..
\fInitiationofstea'm -7 r,.. 7,0, a
, , i release from steam gen-(eratordafetyvalves Rods begin to drop 9. 3 -
l , Peak p.ressurizer - 11.0 pressure occurs - M.inimum DNBR. occiars 11.5
~
' ~ . .
- 2. With pressurizer ,e Turb.ine trip, loss -
0.0 of main feed flow
~
control (maximum rea'ctivity , feedback) Peak pressurizer 7.0 .
* . pressure occurs *
, Initiation of steam W 7.0 release from steam gen-l erator safety. valves ,
- Low-low steam gen'- -
64,0 h2,6 '. l. l erator* level reactor trip setpoint reached ~, - Rods.b.egin to. drop ~ 644,.66.6
~ '
Minimum DNBR occurs., (1) s o l s * *
. (1) ONBR doer not decrease below its initial value. ,
, ,1984 Update
- ~ .
a 4
- 2
,e y ,
- Table 15.2.'3-1-(Page 2 of 5) ". .
Time, SeSJence Of Events For Incidents Which Cause A Decrease 5 - ,s . In' Heat Remov'al By The Secondary System .
- s. .
~
Accident
- Event Time (Sec)- .
n Turbine Trip .-
. 3. Without press'urizer Turbine trip, loss 0.0 control (minimum of main feed flew ,
reactivity feedback) ' High pressurizer pressure 3.7 '
. reactor t, rip setpoint , ,
reached
- w. .
. Rods begin to drop - ,
.5.7 -
Initiation of steam , 4 s 6.5 release frem steam. gen- . erator safety valves -
# Peak pressurizer M 7.0 pressure occurs .
Minimum.0NBR occurs (1)
. . . s .
s - . v , .
- 4. Without pressurizer Turbine trip, loss
- 0.0' of main feed flow
, control (maximum .
= re, activity feedback) ,
.Hi'gh pressurizer pressure 3.7 - *
. rgactor trip setpoint
, reached ,
~ ' ~
Rods begin to drop . 5.7 . Initiation of steam - M 6.5 . release from steam gen-erator safety valves -
- Peak pressurizer '
'M- 6.5 2 .
.. pressure occurs .
, Minimum DNBR occurs. ,
(1) - 3 .
- u. .
(1) ONBR does not decrease be]ow its initial a value. *
.e . * , <-
- .. 198.'. 'L'pda t e g* 4
- e, m
e _, . , s Table 15.*2.3-1 (Page 3 of 5) .
)
Time. Sequence Of Events For Incidents Which Cause A Decrease
~
In Hdat Removal By The Secondary System Accident , Event.' , Time (sec)
. Loss of>Non-Emergency ' Main feedwater flow stops 10 .
AC Power- .. .
~
<C Low-lowsteam'generaYor 8 6a- 5+ ~
*~
. 1 & el reactor trip
- Rods begin to drop. 66 58
,c
- Reac6ar coolant pumps 'iirs - M
. begin to'coastdown .
5 Peak 9ater level in - 6J. 6 l pressurizer occurs c 499-
~
rrr :'-- ;: : :t: : fl5
. L ,,,, ^.; ;;;i.; ; .f; ,
Flow hfrom two motor , - dri'ven auxiliary feedwater '
. pumps is afarted. -
Core decay heat' decreases ~ tee .125 . " to auxiliary feedwater heat r.emoval capacity l Loss of Normal . Mainfeedwaterflowstops ' 10 e , o I Feedwater Flow ! ' ! , . Low-low steam generator ' 60- 5 4.
# . level reactor trip 3
-6&'5I-l ,
Rods begin to drop .,
~
~
Peak. water level in 9P 61, d' pressu,rize'r occurs .
": r :ter- ------' ": 49ik 115 3; . - .,. . n , - _ . . , , , .
F lew *.gaw +e, from two motor- -
- driv *en auxiliar feedwater- "
l I Feededer im.es are a ,
- > pumps is afart ' '
2 51 ; . P9td dLeid cold auxiliasy ,- Core decay heat'plus pump ~ 1000 / -
. heat decreases to auxil,iary. -
I hedwO i.s delivered feedwater heat removal l
'l b b y aftame p era:fors ,
capacity , , I N J
~
1984dpdate ~ e
! <e O*
. .- ~
l l e e . Table 15.2.3-1 (Page 4 of'5) ,' - s - Time Sequence'Of Eve'n'ts For Incidents Which Cause A Decrease .
,,' In Heat Removal By The Secondary System
~
,. Accident , Event" -
.fime(sec)- -
s . , Feedwater System Pipe Break .
~
s . . .
- 1. 'With offsite power Main feedline rupture occurs . 10
, avdilable , .
Low-low' steam generator level 27
. e ,. reactor trip set.p6 int reached ~
in ruptured steam generator
. ~ Rods begin'to drop -
iM .,31 tdr M .49 IS' Auxili
. i:T '::,ary :: feedwater
- bt=2 is s
[ ^' teedwater lines are 220 C^^ 72 7 2
. P M a" d colcl ' tow steam line pressure se+ 342.'
O.L(X i h'ary [ded eter setpoint. reached in rup-is c}c3 icered Eo inYac.t tured steam generator ' 3 steam genera. ors . All main ste'am line 494r 343 isolation valves close-
'{ -
Steam generator safety 844 QSS . y a ' valve setpoint reached in intact steam genera' tors &, Pressurizer wa.ter relief -4066 2006 .
....begins , , c Core decay heat' plus pump - 60GG 4000 heat decreases to. auxiliary -
o feedwater heat removal - . , , capacity ., e o , r . . c 4 o s ,%.
. 1984 Update
. O g-u
. a
*s > ,
*~
. . Table 15.2.3-1 (Page,5 of 5) . a-
. . Time Sequence Of-Events For Incidents Which Cause A Decrease ,
In Heat Removal Ely The Secondar9 System -*
. ~
A'ccident . Event . Time (s'ec) , Feedwater System. Pipe Break .
.. 2. Without offsite power ' Main feedline rupture occurs 10 4
~ e '
Low-low steam-generator 27 ,
, level reacto'r trip set- '*
point reached in ruptured
.' steam generator. -
Rods begin to drop, ^~^- -es 34 - r- ,_.. . .u_ ____ -
--,~~
33
~~~'
Readon coolyt gs. J; m,ac - - kefn to coast down -> .
.~
~
t s - Auxiliary feedwater is s&rM ~ -W II -
. 1: :r:d *: 3-t::t .
. :t:: ;:.: :t::: .
Feededkr lines a're anyd 220 * . g [ "
.g " '
Low steam 1,1.ne pressure 46+* 420
.'"#V '
setpoint r ached in rup- " ,
- - IG , delivered to inYdrE tured steam gendrator , ,
deam 9e.nerators-
- All main steam line' Me427-C , isolation valves close ',
-Mee 1100^
, - . Steam genergt'or safety "
- valve setpoint reached -
in intact steam generators ,., .
, Core decay heat decreases asse 1700
', to auxiliary feedwater ~
. heat removal c,apaci,ty ..
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. TURBINE TRIP EVENT WITH I *~
PRESSURIZER SPRAY AND POWER OPERATED RELIEF VALVES, i { MINIMLH REACTIVITY FEEDBACK ri u e is 2.3-1
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40 60 - 80 1 00 F 0 20 ~ TIE (SECONOS) TUi1BINE TRIP EVENT- WITH l PRESSURIZER SPRAY AND POWER l OPERATED RELIEF VALVES, i g {g MINIMUM REACTIVITY FEEDEACK McGUIRE NUCLEAR STATION Figure 15.2.3-2 1984 Update
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f
. PRESSURIZER SPRAY AND PONER OPERATED RELIEF VALVES, 4
b) Mb I MAXIMUM REACTIVITY FEEDBACK McGUIRE NUCLEAR' STATION , Figure 15.2.3-3 1984 Update
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.* - TURBINE TRIP EVENT WITH' PRESSURIZER SPRAY AND POWER
' OPERATED RELIEF VALVES',
d {j] . MAXIMLH REA,CTIVITY FEEDBACK . i McGUIRE n UCLEAR STATION
' Figtire 15.2.3-4 ,
. 1984 Update ,
., *e . y 6
. N./
e e e-9 4
- b
. a
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McGUIRE NUCLEAR STATION Figure 15.2.3-6
- 1984 Update -
6 *
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MAXIMUM REACTI,VITY FEEDBACK ,
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. 1984 Update 4
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1984 Update
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I eMcGUIRE NUCLEAR STATION S
~
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