LD-84-069, Forwards marked-up CESSAR FSAR Pages Reflecting Review of Tech Specs Re First Sys 80 Plant.Rev Involves Interface Requirements for MSIV & Main Feedwater Isolation Valve Closure Time
| ML20100K493 | |
| Person / Time | |
|---|---|
| Site: | 05000470 |
| Issue date: | 12/05/1984 |
| From: | Scherer A ABB COMBUSTION ENGINEERING NUCLEAR FUEL (FORMERLY |
| To: | Eisenhut D Office of Nuclear Reactor Regulation |
| References | |
| LD-84-069, LD-84-69, NUDOCS 8412110140 | |
| Download: ML20100K493 (101) | |
Text
7 C-E Power Systems Tel. 203/688-1911 Combustion Engineering. Inc. Telex: 99297 1000 Prospect Hill Road Windsor, Connecticut 06095
. POWER SYSTEMS STN 50-470r December 5, 1984 LD-84-069 Darrell G. Eisenhut, Director Division of Licensing U.S. Nuclear Regulatory Commission Washington, D.C. 20555
Subject:
CESSAR Consistency Review Changes
Dear Mr. Eisenhut:
As a result of a review of the technical specifications for the first System 80" plant, several CESSAR-F changes have been found to be necessary. These changes provide clarification and te:hnical consistency of information given in CESSAR-F. The marked-up affected pnges of the CESSAR FSAR are attached.
The attachment provides clarification and consistency of the response times
=
used in the CESSAR safety analyses. These changes do not in any way affect the results and conclusions of the safety analyses, which reinain valid. The following information is provided.
(1) A revision of the CESSAR interftce requirements for MSIV and MFIV closure time to ensure consistency with the safety analysis assumptions.
(2) A revision of the CESSAP. Chapter 6 and 15 Sequence of Event tables, and supporting text (including Section 7.2) to avoid the appearance of any ..
inconsistency with Technical Specifications and interface requirements.
If you have any questions or comments concerning these changes, please feel free to contact me or Mr. G. A. Davis of my staff at (203) 285-5207.
Very truly yours, COMBUSTION ENGINEERING, INC.
84121 g 0g M hPDR o li. 5t/Tcherer PDR A Director A Nuclear Licensing AES:las Attach, cc: P. Mo-iette
[Si I L
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m y
e 4
4 CESSAR INTERFACE REQUIREMENTS AND SEQUENCE OF EVENTS CHANGES TO CLARIFY RESPONSE TIMES NOTES:
~ 1. Whenever-the Sequence of Events Table shows. response time to equal 0.0 seconds, the minimum response time is the conservative direction for the analyses; .
k r-
- 4. The full open to close stroke time of each MSIV and MSIV bypass l
valve,shallbegecondsorlessuponreceiptofanMSIS.
S. The ADV's shall be fail close and shall be capable of being remote manually positioned to control the plant cooldown rate.
- 6. The ADV's shall be provided with manual operators such that the valves may be hand operated from the control room and remote shutdown panel in.the event of a loss of normal power supply.'
- 7. In the combined event of either a steam line break or steam generator tube rupture and the loss of power operation of the ADV's, personnel access to the manual operators of the intact valves on the other steam generator shall be possible.
i
- 8. A MSIS actuation signal shall close the MSIV's, MSIV bypass valve, MFIV's and the steam generator blowdown valves.
j 9. Redundant feedwater system isolation valving shall be provided in both the economizer feedlines and the downcomer feedlines.such that the following criteria are met when the effects of single failure criteria are imposed:
, a. Complete termination of forward feedwater flow is assumed l
within)4eIseconds after receipt of an MSIS.
, b. Afarupt complete termination of reverse feedwater flow with the existance of a reverse flow condition. Check valves are considered to be an acceptable means of achieving the above.
- 10. The economizer and downcomer feedwater line isolation valves (MFIV's) in each main feedwater line shall be remote-operated and be capable of maintaining leak rate of less than 1000 cc/hr under i the main feedwater line pressure, temperature and flow resulting from the transient conditions associated with a pipe break on either side of the valves.
- 11. The Emergency Feedwater System shall be controllable in a post-accident environment from either the control room or a remote shutdown station.
- 12. The Emergency Feedwater System shall be controllable such that post accident operation will not result in overfilling the intact steam generator (s).
- 13. If the Emergency Feedwater System is used as an auxiliary feedwater system, the emergency feedwater pumps shall be designed for operation when steam generator pressure is negligible and not result in damage to the pumps or effect the ability of the system to deliver the required emergency feedwater flow. Such a condition can exist during startup or shutdown operation subsequent to an EFAS which starts the amergency feedwater pumps and fully opens the system isolation and control valves.
l 5.1-12 i
r.
I -. _._.-...__ _._ , _ ._ _ _ _,_ _ _ _,._ ,,. _ _ _ ,_. _ _ , _ __ _ _ _ _ _ _ _
W 5.4.5.2 System Design 5.4.5.2.1 General Description Each of the four main steam lines is provided with a power-actuated main steam isolation valve designed to stop flow from either direction when it is tripped closed. Each valve is located outside containment _and is provided with mean:: of actuation from the engineered safety features actuation system, meeting the reouirements of IEEE Standard 279.
The logic circuitry required to isoiate the main steam lines is discussed in Section 7.3. The main steam system valves and arrangement will be discussed in the Applicant's SAR.
5.4.5.2.2 Component Description The main steam isolation system consists of the main steam isolation valves and their associated controls and instrumentation. The main steam isolation 1
valves are remotely operated valves designed to either fail closed or be guaranteed to close upon receipt of Main Steam Isolation Signal. The main steam isolation valves can be monitored and controlled locally and in the control room.
5.4.5.2.3 System Operation The main steam _ isolation valves are designed to isolate the main steam
( lines and the steam generators as required during operation and under accident conditions.
A steam line break inside containment would result in a pressure rise in the containment. Reverse flow protection is also achieved through the. main steam isolation valves. To achieve reverse flow protection in the case of the main steam pipe rupture, the valve is fully closed within 5 seconds from receipt of the initiating signal.
The main steam line isolation system components are qualified to serve in the environment specified in Section 3.11.
5.4.5.3 Design Evaluation Design evaluations are listed to correspond with the design bases listing.
A. The main steam isola ton valves are capable of isolating the steam
- generators within gi Tconds after receiving a signal from the engineered safety features actuation system. In the event of a steam line break, l this action prevents continuous uncontrolled steam release from more l than one steam generator. Protection is offered for breaks inside or l - outside the containment.
B. The main steam isolation valves, their operators, and associated circui,ry are Seismic Category I, and are protected against missiles and the effect of high-energy ifne breaks.
l l-5.4-15
4 For MSLB cases with small break areas, steam can escape fast enough from the two ' phase region of the affected steam generator so that the level
. swell does not reach the steam line nozzle. A pure steam blowdown results.
Because of the pressure reducing effects of active and passive containment heat sinks, the highest peak containment pressure resulting from a MSLB for a given set of initial steam generator conditions ~ occurs for that case where the break area is the maximum at which a pure steam blowdown can
~
occur. The potential for steam generator two phase level swell following a MSLB increases as power level decreases; therefore, a spectrum of power :
levels must be analyzed to determine which one results in the peak MSLB )
containment pressures.
I The feedwater distribution box is below the steam generator water level; therefore, MFLB cases always result in two phase blowdowns and do not produce peak containment pressures as severe as MSLB cases.
To permit a determination of the effect of MSLB upon containment pressure, analyses are performed with SGNIII (described in Appendix 6B of Reference
- 1) at 102, 75, 50, 25, and 0 percent power. The largest slot and guillotine breaks at which a pure steam blowdown can occur are determitied. The breaks
- are conservatively assumed to be at the nozzle of one of the steam generators.
l The cases analyzed are' listed in Table 6.2.1-1.
The System 80 plants have integral flow restrictors in the nozzles of the steam generators. Credit for the flow restrictors is taken in the analysis.
In the planti, the main steam isolation signal (MSIS) of the engineered safety features actuation system (ESFAS) closes the MSIV's, MFIV's and the emergency feedwater isolation valves. MSIS is generated either by a steam generator low pressure signal or a containment high pressure sign.al. The MSIV'scloseingseconds. The valve closures have been considered in the l analysis. h The main steam line isolation interface requirements are discussed in Section S.1.4. The main feedwater line isolation interface requirements are discussed in Section 5.1.4. The emergency feedwater line isolation interface requirements are discussed in Section 5.1.4.
The emergency feedwater system functions automatically during MSLB to ensure that a heat sink is always available to the reactor coolant system by supplying cold feedwater to maintain an cdequate water inventory in the unaffected steam generator. The affected steam generator is identified and isolated while a controlled flow path is provided to the unaffected steam generator. No credit for emergency feedwater flow to the unaffected steam
[ generator is taken in the MSLB analysis.
I Interface requirements on the maximum steam line and feedwater line volumes l- are discussed in Section 5.1.4. The total volume of fluid between the MSIV's and each steam generator is assumed to be 2000 cubic feet (total for two steam lines). The volume of fluid between the MSIV's and the turbine stop valves is assumed to be 14000 cubic feet maximum. The maximum volumes l
l l 6.2-15 l
TABLE 6.2.1-11 DATA FOR CONTAINMENT PEAK PRESSURE / TEMPERATURE ANALYSES 102% POWER / SLOT /8.78 SQ. FT./ LOSS OF CONTAINMENT COOLING Sheet 7 of 7 ,
I PART D. Accident Chronology ,
Event 'h i ? -
yy T013 (Seconds)
\'I
-~
0.00 Break Occurs ,
3.80 Reactor Trip Signal gd ph 3.80 Main Steam Isolation Signal i
3.80 Main Feedwater Isolation Signal Turbine Admission Valve Closed I 4.70 \
j 4.70 Reactor Trip Begins
- 4.70 Main Steam Isolation Valves Start
'. To Close 4.70 Main Feedwater Isolation Valves Start
'q To Close 9.70 Main Steam Isolation Valves Closed i
! 9.70 Main Feedwater Isolation Valves Closed I
A* Containment Spray Actuation Signal A* Peak Containment Temperature A* Peak Containment Pressure 170.00 End Of Blowdown l
- See Applicant's SAR l-t 1
}&$
~
3.80 (' J J T m _u 77. ~ % L.oa
~
^:& T4 i
- L _
18'o (JATm_ G u .
N 5 M. _ '
( e i n s l h ._ _
('-J M PL w M ~ l
~
U ~
Qsrs aw
.. . _ _ ', & .Ym'. %._ }_m_.o m_ - . . . _ _
G M nn +w % ck.a4 . -
s I
I
- - _ - - -- _---_ - _ -- _ s
1 4.
TABLE 6.2.1-12
? DATA FOR CONTADMENT PEAK PRESSURE / TEMPERATURE ANALYSES
- . 1025 POWER /GUILQTINE/8.78SQ.FT./LOSSOFCONTAIt"KMTCOOLING s Sheet 9 of 9 .
s PART D:' Accident Chronology Event .
h (Seconds) N,,
~ '
0.00 Break Occurs .
/fu m
- - _,/ \ \
g
'~, m 5.25 Reactor Trip Signal 5.25 Main Steam Isolation Signal i p Main Feedwater Isolation Signal V' i 5.25 6.15 Turbine Admisssion Valve Closed ,
6.15 Reactor Trip Begins .;
6.15 Main Steam Isolation Valves Start
. To Close , H l 6.15 Main Feedwater Isolation Va'lves Start / -
To Close ,-
.. 11.15 Ma),. Steam Isolation Valves Closed 11.15 Main Feedwater Isolation Valves Closed A* Containment Spray Actuation Signal A* Peak Containment Temperature r-- 3
- A* Peak Containment Pressure 175.00 End Of Slowdown ( .,
~
" See Applicant's SAR i
'i a
- -e 's .j
==
e B
e es ---
5.15 t" J J L ~ . Tt. A 4.o -ic
%K ' '
T4 I be Y MY
- A^dk/Le_ bA.fSb I' ..
h Me o%421, -
J ~
J'-
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we m
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e
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-66 W
- O O O O Os. ( d' Tit .M Kw &~
~
~
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a.nlM$$$
. ... __ .LN . ._. e _ . OA '
6-@ T& W4 % @ Q l l
-- , . a .
v e eea ee g
e *
- h e#8* Ome + m eme . . ee _ e e- o eme .
e .m M .N M
e eee= oe g,u
'*N e -em
-e--r-- - - .,-----,,,--i.me . .-.%.-, ,7-- ,,w._,y , . - - - .,,,,.,,,w--,-,-- ~ - , . , - p, w, ,-.,,..--w.,,
9 TABLE 6.2.1 .
--s, DATA FOR CONTAIMENT PEAK PRESSURE / TEMPERATURE ANALYSES 755 POWER / SLOT /8.78 SQ. FT./ LOSS OF CONTAINMENT COOLING Sheet 7 of 7 PART D: Accident Chronology Event Time (Seconds) 0.00 Break Occurs 3.70 Reactor Trip Signal 3.70 Main Steam Isolation Signal I- k 3.70 Main Feedwater Isolation Signal f l
4.60 Turbine Admission Valve Closed 4.60 Reactor Trip Begins l
i 4.60 Main Steam Isolation Valves Start To Close.
l i 4.60 C MainFeedwaterIsolationValvesStart)
To Close 9.60 Main Steam Isol. Valves Closed 9.60 Main Feedwater Isol. Valves Closed A*
Containment Spray Actuation Signal Peak Containment Temperature A*
A*
Peak Containment Pressure 185.00 End Of Blowdown
- See Applicant's SAR f
~ * "' ----*-e.-r,-.- , . ,
L C
2 7D 0 Tw_e a E=nSv t..a L
~
' 0
^: & T m' w
(.
3JO (dM7A.c449/u hd .A o.. ..
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n
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4 -.
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9
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=
eue...- .e m 6 gy., . . , . . -
9
TABLE 6.2.1-14
' DATAFORCONTAINMENTPEAKPRESSURE/TEMPERATUREANALYSES 75% POWER / GUILLOTINE /8~.78 SQ. FT./ LOSS OF CONTAINMENT CO Sheet 9 cf 9 L
PART D: Accident Chronology Event Time (Seconds)
A
,7>
0.00 Break Occurs - '
Reactor Trip Signal 5.15 Main Steam Isolation Signal g)
Main Feedwater Isolation Signal I
- 5.15 6.05 Turbine Admission Valve Closed 6.05 Reactor Trip Begins 6.05 Main Steam Isolation Valves Start To Close
( W 6.05 MainFeedwaterIsolationValvesStart)
- .Tq t1_qs.e
- ~ ~ ~ ]
11'.05 Main Steam Isol. Valves Closed 11.05 Main Feedwater Isol. Valves Closed A*
Containment Spray Actuation Signal A*
Peak Containment Temperature A* Peak Containment Pressure 190.0? End of Blowdown
- See Applicant's SAR
- e 3
6 W
K.15 ('JJ 7- . F.~Lu 6.o s
_ ^ & T4 '
_Jf 'L s.1.5 WATm G - s.9 n o%4A. '
._ Dnns _. 6 esem_,p6 ee m- e -e GW 666 .6 D e we
~
G.I5
. =
rditPL_W' 4_ .. .e_. . . . .. __. . . . _ .___..
/ a.no M S.1s
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% cw .. O/A v.
. . ... w y
M e e .ee, H
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t
-e.bG*
6 66 .eMgg
TABLE 6.2.1-15
~~
DATA FOR CONTAINMENT PEAK PRESSURE / TEMPERATURE ANALYSES 505 POWER / SLOT /8.78 SQ. FT./ LOSS OF CONTAINMENT COOLING Sheet 7 of 7 J-PART.D: Accident Chronology Time (Seconds) Event F
0.00 Break Occurs .
3.55 Reactor Trip Signal // -
l Main Steam Isolation Signal AV 3.55 3.55 Main Feedwater Isolation Signal 4.45 Turbine Admission Valve Closed 4.45 Reactor Trip Begins
! 4.45 Main Steam Isolation Valves Start l
To Close i 4.45 Main Feedwater Isolation Valves Start i
To Close /
9.45 Main Steam Isol. Valves Closed' 9.45 Main Feedwater Isol. Valves Closed A* Containment Spray Actuation Signal A* Peak Containment Temperature A* Peak Containment Pressure 215.00 End of Blowdown
- See Applicant's SAR G
. C 3.55 (' J J T _ . L Lu i.o ,
~
fL
. TV Ta i
- 55 M_
h.D0
_ u. . _ .
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4 10 TM nt: . .
l
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l L 4
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l -
TA8LE 6.2.1-16 i DATA FOR CONTAIPMENT PEAK PRESSURE / TEMPERATURE ANALYSES 505 POWER / GUILLOTINE /8.78 SQ. FT./ LOSS OF CONT. COOLING Sheet 9 of 9 PART D': Accident Chronology
. Time (Seconds) Event 0.00 Break Occurs _
\ l).)
- - _ . - ~
y 5.0~0 Reactor Trip Signal _Nf (f
5.00 Main Steam Isolatfun Signal y
Main Feedwater Isolation Signal i, 5.00 5.90 Turbine Admission Valve Closed I
5.90 Reactor Trip Begins 5.90 Main Steam Isolation Valves Start To Close
- 1 5.90 Main Feedwater Isolation Valves Start
( To Close 10.90 Main Steam Isol. Valves Closed
- 10.90 Main Feedwater Isol. Valves Closed A* Containment Spray Actuation Signal .
A* Peak Containment Temperature A* Peak Containment Pressure
220.000 End of Blowdown l
I l
- fu Applicant's SAR 1
MF
- g. oo b. x A L n u. Lu x-( a ra- ww i
( V
- 2
- 5. OD [ddb_t44tuu h> M La un_
J ~ '"
\ %a12~~ ehAA _
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/
1 i
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ene e 1 .
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TABLE 6.2.1-17 DAT5FORCONTAINMENTPEAiPRESSURE/TEMPERATUREANALYSES 255 POWER / SLOT /8.78 SQ. FT./ LOSS OF CONTAINMENT COOLING Sheet 7 of 7 ,
PART D: Accident Chronology N -
Event Time (Seconds) A.9-0.00 Break Occurs
/ 3.45 s
Reactor Trip Sign N ,. ,
]k ;
N '
i 3.45 Main Steam Isolation Signal 3.45 Main Feedwater Isolation Signal 4.35 Turbine Admission Valve Closed 4.35 Rs. actor Trip Begins 4.35 i4ain Steam Isolation Valves Start To C1ose 4.35 Main Feedwater Isolation Valves Start
_ _ To C1ose- _...
)
9.35 Main Steam Isol. Valves closed 9.35 Main Feedwater Isol. Valves Closed A*
Containment Spray Actuation Signal A* Peak Containment Temperature A* Peak Containment Pressure 315.8 End of Blowdown
- See Applicant's SAR
-1 eemvW-- =__a m -_*-*-_m - -___--
(7-3U 6#6+1 Tho m. 72, m) , , f.w,
's W - -
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-e,----.-.,---,--.,%.4--.---4 --._,,%_4,- ,,yw..w,,.-. -.,-.m-.-y-g-.g..
TABLE 6.2.1-18 DATA FOR CONTAI M ENT PEAK PRESSURE / TEMPERATURE ANALYSES 25% POWER / GUILLOTINE /8.78 SQ. FT./ LOSS OF CONTAIMENT COOLING Sheet 9 of 9 PART D: Accident Chronology Time (Seconds) Event 0.00 Break Occurs J P4 h L
4.86 Reactor Trip Signal M s 4.86 Main Steam Isolation Signal 4.86 Main Feedwater Isolation Signal 5.76 Turbine Admission Valve Closed
- 5.76 Reactor Trip Begins 5.76 Main Steam Isolation Valves Start To Close l
[ 5.76 Main Feedwater Isolation Valves St) art la C10se 10.76 Main Steam Isol. Valves Closed 10.76 Main Feedwater Isol. Valves Closed A* Containment Spray Actuation Signal A* Peak Containment Temperature A* Peak Containment Pressure
, 315.80 End of Blowdown
- See Appifcant's SAR 1
't S
- , - ,, - - w.--. - ,.,--..--..-e.-- . . . - . -- -..,---,-.e-,, ..,,,-,- _ -- --,,,,, . ve -,%...-v.,-- -e,-,,,,...-,w.,..- e---
H i
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TABLE 6.7. 1-19 -
DATAFORCONTAINMENTPEAEPRESSURE/TEMPERATUREANALYSES 0% POWER / SLOT /4.00 SQ. FT./ LOSS OF CONTAINMENT COOLING Sheet 7 of 7 PART D: Accident Chronology .
Event Time (Seconds) I 0.00 Break Occurs
- 1 4.5 9 ignal g A Main Steam Isolation Signal y 4.55 Main Feedwater Isolation Signal 4.55 5.45 Turbine Admission Valve Closed 5.45 Reactor Trip Begins j 5.45 Main S*.eam Isolation Valves Start To Close I 5.45 MainFeedwaterIsolationValvesStar)t To Close _
10.45 Main Steam Isol. Valves Closed 10.45 Main Feedwater Isol. Valves Closed Containment Spray Actuation Signal A*
Peak Containment Temperature A*
A* Peak Containment Pressure 210.00~ End of Blowdown
- See Applicant's SAR
e e
=
Vr M 4DA L
- hA5 T M Tm s k Jani? -
-'V
. 1
( v
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A .
g.b4 _
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b w O A.
6".70 0 7%A * - .
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e e-ee O
es M
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. -, . , , , - . .. , . --, . _ _ _ - , . - p y_-__ _ ,. - _ - - - , . _ , , . . , _ - - --,_.,-,,-r_,,_ .-
- TAELE 6.2.1-20 DATA FOR CONTAINMENT PEAK PRESSURE / TEMPERATURE ANALYS 05 POWER / GUILLOTINE /8.78 SQ. FT./ LOSS OF CONTAINMENT CO Sheet 9 of 9 PART D: Accident Chronology Event _
Time (Seconds) 0.00 Break Occurs ' ;
-- - / ,
,y -
4.75 Reactor Trip Signal ,
J 4.75 Main Steam Isolation Signal 4.75 Main Feedwater Isolation Signal [
5.65 Turbine. Admission Valve Closed
/
5.65 Reactor Trip Begins i 5.65 Main Steam Isolation Valves Start To Close 5.65 Main Feedwater Isolation Valves Start
.... . g Clase---- .
10.65 Main Steam Isol. Valves Closed 10.65 Main Feedwater Isol. Valves Closed A*
Containment Spray Actuation Signal
- A*
Peak Containment Temperature A* Peak Containment Pressure 210.00 End of Blowdown
- See Applicant's SAR
--- --+- --__ .____, ,_ ,,_ ,_
. d 4,].5 ffd kmm 77 A t s a a.
T V Tm i M V
'f 75 , (JMTm LA 4Y,_
% .i A s-aAL -
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6
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& 93m8
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. v 4
6 9W&
me
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- . . . - ---4 - . . - - --
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i
. l TABLE 6.3.3.3-6 _
b . TIMES OF INTEREST FOR SMALL BREAKS (Seconds)
MPf2 f** gpgg purup s.r wons
/***#ArtemHot Spoti
/ kans J A N # # /se ,J yr a g er s Peak Clad ft M W.'.' N redef ea 5LJanine t a Acs Break {fze en Temp. Occurs (ft ) W # 4 ump-en 2 158.0 142.0 160.0 0.50 ft /PD 4'6.5 2 248.0 204.0 235.0 0.35 ft /PD 50.0 2 a. 400.0 442.0 0.20 ft /PD 62.0 2 a. b. 2010.0 0.05 ft /PO 208.0 2 a. b. 437.0 0.02 ft /PO l 2 a. b. 540.0 0.03 ft /HL 585.0 (a) Calculation terminated before time of LPSI pump activation.
(b) Calculation terminated before initiation of SI tank discharge.
@)7'hof f/**$ tucte 34$ A Se 7deauD *DesA Y fde n f f.s h oett. ,
+hnr the yestratoseg pressasse RfA<ord5 TW6 Laa yoe.uaao e s< yo-*2susee f.tn* A"*"v s/We'" t-OS N 'Nivst. wbLn ik 3I pum A- QNbE4Y l b5 A f ""/Wh. f lhW
,,-~,wea--,~-ncv-,,e,---.-w.-.,e-m _---,--------n-
t I TABLE 6.3.3.5-1~
(Sheet 1 of 2) l' -
SE3)ENCE OF EVENTS FOR REPRESENTATIVE _LARGE AND SMALL BREAK L 2
Large Break (0.8 DEG/PDj _ _
Small Break (0.02 ft ) ,
Setpoint Success Setpoint Or Value Time, Seconds ' Path or Value Time Seconds Event 0.0 0.0 i
i Break occurs 0.15 105% 96.0 117%
Core peak power Reactivity
.rws(W y 9.43 1600 psia 456.0 g_ _ ....-.,.i..,-.
_ .. .g
^
1600 psia Control 1 W 24acreo, fe ar Ae o w.ts Mempsie-es"1.*
-466 6 Reactivity l 4800= pres & 43 Contro)
Sufaty injection actuation ,
l
~ signals qsv#dArds Reactivity 607.7 psia 7500 607.7 psia 16.1 Control SIT discharge begins NA 37.7 j Reflood begins Sec. Sys.
1295 psia 456.0 J NA Integrity Main steam safety valves begin to open 1340 psia 194.0 Maximum secondaij pressure 1239 psia
- Reactivity 68.2 HPSI y N' T $$$ '
Control tb &
NA 66.2 SITS empty Reactivity NA
- TfL/ 68.2 Control LPSI pumpsf .fLocJ.rt i, a ; . .YfA.gp Nd AC $ ,
t
~
.. _1.ormer.i r_w.__6me- d.X 1.5-t. - -
AfsrXpas Wdd .3RettaAg-.2Enera43 m.E*4_rsed.e ..
8l##' .
. . . _ duM.)cret . FAT" Mis /#.1.? ,
- - =
- 8 l
t I
t t
1 I
. _ _ . - . - = - - . . - . - - -
- - = - - . - -
J t . .-. _ - _ .--...... ._ -
-- - --- - -. c
-- =-... _ _ . . . . ..
. . es.
. . - -. g ..g . , . .. _- mM-
. -*.*4. *- ..* .- . . - ....a
- ' - ,w - <
.- E 4
f.
The'systinisdesignedtodeterminethefollowinggenerating during Limiting Faults:
- 1. Core power;
- 2. RCS pressure;
- 3. Steam generator pressure; and
- 4. Containment pressure.
- g. The system is designed to monitor ali generating station variables that are needed to assure adequate date.'mination of the conditions given in listings e. and f. above, overThe thefull entire range power of normal nominal values and
- operation and transient conditions.the maximum and minimum values t The type, number, and location plant variable are given in Table 7.2-2.of.the sensors provided to mo 3.
- h. The system is designed to alert the operator when any monitored plant condition is approaching a condition that would initiate protective action.
- f. The system is designed so that protective action will not be initiated due to normal operation of the ginerating station.
Nominal full power values of monitored conditions and their corresponding protective action (trip) setpoints are given in Table 7.2-4.
The selection of these trip setpoints is such that adequate protection is provided when all sensor and processing time delays and inaccuracies Q fa- 'T are taken into account.S Acepense times and analysis setpoints used in thedafgty,analy'sj e are given in CM;t:r !5.0. T7sMe /OO NftPC C 'T W 4 The' trip delay timest and analysis setpoints ;c;;._d in Chapter 15.0 are representative or the manner in which the RPS and associated sient These quant Ltig,j gngd in ty hr TME deesnie. dadana' instrumentation will operate.Alone in Chaoter ts,end 15.0.9 delay Actual RPSMM'fta r vs <
y g times w'l' be obtained from calculations and tests performed on the
, The verified systes uncertainties l RPS and associated instrumentation.
are factored into all RPS settings and/or setpoints to assure that the system adequately performs its intended function when the errors and uncertainties combine in an adverse manner. .
! j. All . system components are qualified for environmental and seismic conditions in accordance with IEEE Standard 323-1974, and IEEE Standard 344-1971. Compliance is addressed in Section 3.11 and in CENPD-255,
" Qualification of Combustion Engineering Class IE Instrumentation",
i (Reference 3); and in Section 3.10 and CENP0-182, " Seismic Qualification L
7.2-17
..Twek. O. p. 7.2- n . _ . _ . _.
. . .reas.%c pnk ka ns%m.._seooc-. <eseam<-. b.me9 ceacht--
Iawek b .. .. _. p_. l. ? - n __ _ . _. . _ . . . _ . _ . . _ . _ . . . . _ _ .
M.&= . .
4k k. _ kke. . ceacbe h.e otela3 .hes _.skown -
in .TAL ts. o - 4 J. 4.+ ;4 e.t u ae &ke. se%.c respoose - +1ee s . . . .
Leet E p. 7.a-#7 _._ . ._ _
yesponse hame.S M etacho r he'i p I
k
.- -e M
-edw -= m .-4 .G e M.
t Manual operations performed on a given system or component are indicated by placing an "M" in the lower left-hand corner of the system block. . When a I manual action is required, the sensed variables necessary to perform the action '
are shown as taputs and the location of the input signal is shown above the input signal circle.
The system setpoint values assumed in the transient analysis, e.g., trip signal setpoints, will be noted along the success path. Time delays or the time required to perform an action are shown as a number with square brackets.
All events presented in Sequence of Events Diagrams (SED) in this chapter are shown from event initiation to achievement of the Cold Shutdown operating mode (see Chapter 16). Not all events require that the plant be taken to Cold Shutdown. The SED's only demonstrate that for any event presented here it is possible to take.them to Cold Shutdown by means of the safety actions indicated.
15.0.2 SYSTEMS OPERATION 1
During the course of any event various systems may be called upon to function.
Some of these systems are described in Chapter 7 and include those electrical, instrumentation, and control systems designed to perform a safety function (i.e., those systems which must operate during an event to mitigate the consequences) and those systems not required to perform a safety function (see
- Sections 7.2 through 7.6 and 7.7, respectively).
The Reactor Protection System (RPS) is described in Section 7.2. Table 15.0-4 lists the RPS trips for which credit is taken in the analyses discussed in this )
section, including the setpoints and the trip delay times associated with each trip. The analyses take into consideration the response times of actusted devices after the '.ci; ::tt' ; i: r:'.:d. Imt A r
EQUALS ORyable tietedS b hudt8- The reactor trip delay time) 3.. - 15.0-4)epe defined as the elapsed i
time from the time the sensor output c;;;'.:: the trip setpoint to the time the r;: r ': r f:hd O 2:f ; t' t-r:':-
r e.ujera.trip breaker open.
, .a a .. a_ a.
'h: r r-
.._a...- .
The interval between trip breaker opening and the time at which the magnetic flux of the Control Element Assembly (CEA) holding coils has decayed enough to allow CEA motion is conservatively assumed to be 0.34 seconds. Finally, a conservative value of 3.66 seconds is assumed for CEA insertion, defined as the elapsed time from the beginning of CEA motion to the time of 90% insertion of the CEAs in the reactor core.
The Engineered Safety Feature Actuation Systems (ESFAS) and electrical, instrumentation, and control systems required for safe shutdown are described in Sections 7.3 and 7.4, respectively. The manner in which these systems function during events is discussed in each event description. The instrumentation which is required to be available to the operator in order to assist him in evaluating the nature of the event and determining required action is described in Section 7.5. The use of this instrumentation by the operator is discussed in each event description.
4 15.0-4
I Towc4 A p. is.o - 4 . _ _ _ . . _ . . __
. _ . . . . VMue ke mev0_ho. red...pp.eame. / _ ab_._. . be 5e n ao.e _. . . .
. _ _ _ _ _ _ . . ta a I5 ae e.xc ee d.5 .be,. hC'.p_. Echc.%k. .. __
16 seek f. IT.D - 9 . . . _ .
l Ye .reAc.be p robe c.h;Ve 5gshee response. e la _ _ . . ...
f Ne sum o ke sensor .respande...h;me._ 0,62 he ..
ceae.%r ke.e Je\ay k;~ e . T*ke s e n soc _ rese.ons e. .
4;me is J.Leel as &Le be fr.- aken L .
value oL FLc monilocek par. mete r af A seasoc.
or . ,xceeAs A c e.c.6- p ro h c. k .. s y s h e 32.is
-hc; p sakpotak ank\ L sensor oakeak ap s. t oc
- exceeAs L +c.e selpo
- nf . Tke s enso< respon s e is moJ te) l3 usin3 a +cansCee 9unc6 be A parhremi e seaso< use).
f TABLE 15.0-4 REACTOR PROTECTION SYSTEM TRIPS USED IN THE SAFETY ANALYSIS R*-8e Analysi )2 Trip
~. RPS Setpoint Delay Time (C-)
Event High logarithmic Power Level 2% 550 ms .
Variable Overpower 17% or 130%(a)
High Pressurizer Pressure 2450 psia 550 ms Low Pressurizer Pressure 1580 psia 550 ms Events not Low Steam Generator Pressure 820 psia Low Steam Generator Water Level 40% wide range (b) 550 550 ms ms Mentioned Below High Steam Generator Water Level 99% nqrtow 550 ms rangeles 1.19 150 ms 21 kw/ft N.
Low DNBR High Local Power Density 150 ms seem G===.r-a or APLeo F/w 9e ./. (g) (a) (p) l Variable Overpower 17% or 130%
Feedwater and High Pressurizer Pressure 2475 psia 550 ms Steam Line Breaks low Pressurizer Pressure 1600 psia 550 ms Low Steam Generator Pressure, 810 psia Low Steam Generator Water Level 35% wide range (b) 550 550 ms ms High Steam Generator Water Level 99% ngrrow 550 ms rangete)
Low DNBR 1.19 150 ms High Local Power Density 21 kw/ft (d) 150 ms
- a. See discussion in Section 7.2.
i
- b. Percent of distance between the wide range instrument taps above the lower tap. See Chapter 5 for details,
- c. The G fdN ay times are3discussed in Section 7.2,r d
- -d 9 ti; r' : d r r ^- l 1: % .n,.
- d. Setpoint value is set below the value at which fuel centerline melting would occur. See Section 4.4.
- e. Percent of distance between the r. arrow range instrument taps above the lower tap. See Chapter 5 for details.
.F. %rn a. MTc.r i5 anQ se.s ct.eu.mel- rriere. ce(s=.rvar*ue ser pok-ts & .,p #< c. e.u c s .
c3. PereE eV luit. leg -Vlow.
- h. /.o 9. M b rn h eV ou.u rc=e- oV l~ -W
.tr;y c.o&c: en ux c;I ck reator- cr*> beaka.rs e %en.
o
TABLE 15.0-5 INITIAL CONDITIONS
( .
i-Parameter Units Range Core Power 1 of 38_00 Mwt 0 - 102
- Radial 1-pin peaking - 1.40 to 1.63 factor (with uncertainty)-
Axial Shape Index 0) -0.3 1 ASI 1 + 0.3 '
Reactor Vessel Inlet % of 445600 gpa 95 - 116 Coolant Flowrate Pressurizer Water 1 distance between 26 to 60 Level upper tap and lower
- tap above lower tap Core Inlet Coolant F 500 - 500 (2)
Temperature Reactor Coolant System psia , 1785 - 2400 Pressure 4
Steam Generator Water % distance between 40 - 88 C tevei upper tap and iower tap above lower tap area under axial shape in lower half of core (1) ASI = - area under axial shape in upper half of core total area under axial shape (2) Additional restrictions were apglied to: Section 15.2.3, minimum core inlet coolant temperature j pquals 5600F; and Section 15.1.5, maximum core {
, inlet coolant temperature equals 570 F.
O i6 k
l l
B. Input Parameters and Initial Conditions Table 15.1.4-3 lists the assumptions and initial conditions used for these analyses in addition to those discussed in section 15.0. Conditions were chosen such that the overpower condition caused by the increase in steam flow results in the closest approach to the specified acceptable fuel design limits (SAFDL) without causing a reactor trip. If core power increases more than the 11% due to the increasing steam flow, the Core Protection Calculators (CPC) i will initiate a reactor trip and there will be no further degradation in i thermal margin. For transients initiated at other sets of initial conditions, l a trip may or may not be required depending on whether the initial thermal margin is as low as for the combination of conditions used in these analyses.
C. Results Case 1:/ Inadvertent Opening of a Steam Generator Atmospheric Dump Valve
- ~
i!0SGADV)
/
The dynamic behavior of the salient NSSS parameters following the 10SGADV is presented in Figures 15.1.4-1.1 to 15.1.4-1.15. Table 15.1.4-1 summarizes the major events, times and results for this transient.
! The opening of an ADY increases the rate of heat removal by the steam generators causing cooldown of the RCS. Due to the negative moderator reactivity coefficient, core power increases from 102% of rated core power. reaching a new, stabilized value of 113% after approximately 30 seconds.
The feedwater control system, which is assumed to be in the automatic mode
- supplies feedwater to the steam generators such that the steam generator water l ,
levels are maintained.
! During the IOSGADV transient the minimum transient DNBR_ of 1.19 first occurs /#8 9
! at approximately 30 seconds and . remains there until Qgy seconds when the '
operator manually trips the reactor. At 1850.55 seconds the trip breakers l
l open. Aft;c ; 0_?' :::: d ;;f' i;., i?;, if.; GCA'; b;;f r 0; J. e., int; % l cere et iS50.S9 :::::ds. At this point, both the local and core average power decrease rapidly and DNBR increases. From 1858 seconds to 1886 seconds the
' MSSV's release steam. g 149,4 AtQg@ seconds the steam generator pressure drops below the Mi!S setpoint of 820 psia. The MSIS initiates closure of the MSIV's and MFIV'4 The MFIV's and MSIV's close by 2155 seconds. The affected steam generator dries out at 2650 seconds. At 3000 seconds the operator manually closes the open ADV. The oper,ator initiates plant cooldown at 3600 seconds.
Case 2: Inadvertent Opening of a Steam Generator Atmospheric Dump Valve with Loss of Offsite Power after Turbine Trip (IOSGADY + LOP)
The dynamic behavior of the salient NSSS parameters following IOSGADV with loss of offsite power is presented in Figures 15.1.4-2.1 to 15.1.4-2.15. Table t
I 15.1.4-2 summarizes the major events, times and results for this transient.
The opening of an ADV increases the rate of heat removal by the steam generators causing cooldown of the RCS. Due to the negative moderator reactivity coefficient core power increases from 102% of rated core 15.1-7 r
power, reaching a new, stabilized value of 1135 after approximately 30 secondsI ~~ ^
The feedwater control system, which is assumed to be in the automatic mode, l supplies feedwater to the steam generators such that the steam generator water '
levels are maintainbd until the time of loss offsite power.
During the 10SGADV + LOP transient the minimum transient DN8R of.1.195 first occurs at approximately 30 seconds and remains there until the assumed turbinetrip followed by loss of offsite. power at 45 seconds. Due to decreasing.
core flow following the loss of power to the reactor coolant pumps, conditions existfor a low DNBR trip. At 45.6 seconds a low DN8R trip signal is initiated by the core protection calculators. The reactor trip breakers open at 45.75 seconds M e*+-- : 0.00 ..;e..d een d.;; thy tM CE?' k-f = tr dr;,; ';.6 tM :;c; et 00.0^ w eide. At 46.1 seconds the minimum transient DNBR of 1.05 is calculated to occur, after which DNBR rapidly increases as shown by Figure 15.1.4-2.15. "y 50.5 ;;;;;d; th; 0:".': :r: ""? !; 'r.;;ct;d. At 52 seconds the MSSV's open and release steam until 81 seconds. Voids begin to form in the
- upper head of the reactor vessel at 74 seconds. g g ,9 setpoint of At 820@ia.
ps seconds The MSIStheinitiates steam closure generator pressure of the drops MSIV's and belowThe MFIV's, theMFIV's M and i the MSIV's close by 318 seconds. At 1150 seconds the affected steam generator dries out.
At 1800 seconds the operator manually closes the open ADV. The operator initiates plant cooldown at 3600 seconds.
Due to the coastdown of the reactor coolant flow a reduction of DNBR below 1.19 is calculated to occur. Approximately 8% of the fuel pins are predicted to i
experience DNB. However, within 3 seconds of reactor trip, the local and average core heat flux have decreased enough such that no pins remain in DNB.
15.1.4.4 Conclusions l The 10SGADV event results in a DNBR greater than 1.19 throughout the transient. The event in combination with a loss of off-site power (IOSGADV +
s LOP) results in a small fraction of the fuel pins being predicted to be in DNB for a few secondt. Thus at the most a limited number of fuel rod cladding perforations could occur for the 10SGADV + LOP event. For both cases, the RCS pressure remains well below 2750 psia, ensuring that the integrity of the RCS is maintained.
l -
l l
l 15.1-8
i TA8t.E 15.1.4-1 SEQUENCE OF EVENT 5 PUR FULL POWER INADVERIENI OPENING OF A 5TEMI GENERATdR i ATMO5PHERIC OtEP VALVE (105GADV)
Setnoint Time (sec_) or value
~ Event 1.0 One atmospheric dump valve opens fully Steady-state hot channel .ONBR achieved 1.19 30.0 1850, Operator initiates manual trip denen. --
l 1850.55 Trip breakers open
- l
-1850.00 Cia s vy;.. Le dc;;;
Main steam safety valves open, psia 1282 1858 Main steam safety valves close, psia 1219 1886 1872 Void begins to form in RV upper head F
Main steam isolation signal _ M[4.#' :tes 2150.L}
2155 MFIV's close completely 2155 MSIV's close completely 2650 Affected steam generator dries out 3000 Operator manually closes ADY 3600 Operator initiates plants cooldown s
2199 4 sh Csta P4 f r~e
- s..Iwmsip H5rs) r& s2o
! 4 sa y e,r u
TABLE 15.1.4-2 .
SEQUENCE OF EVENTS POK PULL P0wtR INADVERTENT OPENING OF A 5IEAM GENERATOR ATMU5PHERIG DUMP VALVE WITH ---
l 0 LD55 0F OFF51TE POWER AFTER TURBINE TRIP Setpoint Time (sec)
~ Event or Value .
0.0 One atmospheric dump valve opens fully Steady state hot channel DNBR achieved 1.19 30.0 Turbine trips 45.0 45.0 Loss of offsite power occurs 45.6 Low ONBR trip @ S Qe.ne.N~WA 45.75 Trip breakers open
~
- /', .. . :: fc:;
" 00 1.05 46.1 Minimum transient DNBR 48 Hot channel ONBR increases above 1.19'S
- 0.;
- :".' ; '.11; ' 9::rt:d 9
Main steam safety valves open, osia 1282 52 Main steam safety valves close, psia 1218 81 74 Void begin5to form in RV upper head Main steam isolation signal SR M 313,9 318 MFIV's close completely 318 MSIV's close completely 1150 Affected steam generator dries out --
1800 Operator manually closes ADV 3600 Operator initiates plant cooldown 3 17_. 9 ih C4 SND hre.ssW Pa.- < be $ g A sh- N .l.ct:>. - v g (Hsis) % sis re'C' a.%t jP5fa- .
1 i
~
of offsito power (Cas2 2) th? most adysrsa effcct is caus:d by failure cf o '
MSIV on one of the steam lines on the intact generator to close following MSIS. Consequently for this case steam is assumed to continue to be released y from the intact steam generator after MSIS at a rate consistent with the interface requirement of a maximum of 11% design steam flow rate non-isolable steam flow. Th open flow path is represented by an effective flow area for steam blowdown the intact steam generator of 0.2556 square feet. For case 5 (SLBFPD) there is no single failure which increases the potential for degradation in fuel cladding perft#mance or which increases the offsite dose.
However the failure of a MSIV was used in the analysis to be consistent with Case 2 (SLBFP).
The sequence of events for Cases 1 through 5 above are presented in Tables 15.1.5-1 through 5, respectively.- The sequence of events for Case 6 is the same as for Case 3.
15.1.5.3 Analysis of Effects and Consequences A. Mathematical Models The mathematical models and data transfer between codes used in the SLB analysis are presented in Appendix C.
B. Input Parameters and Initial Conditions The initial conditions assumed in the analysis of the NSSS response to Cases 1 through 5 are presented in Tables 15.1.5-6 through 10, respectively, The initial conditions for Case 6 are the same as those for Case 3. Justification l
of the selection of initial conditions and input parameters is presented in Appendix C.
1 C. Results l Case 1: Large Steam Line Break During Full Power 0 ration with Concurrent Loss of Offsite Power (SLBFPLOP x .
The dynamic behavior of the salient NSSS parameters following the SLBFPLOP is presented in Figures 15.1.5-1.1 through 15.1.5-1.16. Table 15.1.5-1 sumarizes
- the major events, times, and results for this transient.
l l Concurrent with the steam line break, a loss of offsite power occurs, At this i time an actuation signal for the emergency diesel generators is initiated. Due to decreasing core flow following loss of power to the reactor coolant pumps, conditions exist for a low DNBR trip. At 0.6 second a low DNBR trip signal is initiated by the core protection calculators. At 0.75 second the reactor trip I breakers
..._...._-.u open. ATW . G.X
. . ~ 2. _ , ,o .
seq,jyg Ph,, ,.,...
, QCitp;,,,
.. ..... ,,,w....
uivy ory r had a' h reac+~ rm . At@ seconde the steam Generator pressure droDs below the MSIS setpoint of 810~ psia. fThe MSIS initiates closure of the MSIVs and MFIVs The MFIVs and MSIVs close by 13.3 seconds. EFW is automatically initiated to the intact steam generator, assuming no delay after the EFAS signal on low level in the intact st enerator, at 13.3 seconds. At 120
) seconds the pressurizer empties. At seconds the pgssurizer oressure hat 2.
dropped below 1600 psia and initiates IA Within 9 secenas of SIAS the operable HPSI pump is loaded on the diesels nd reaches full speed and the HPSI valves are fully open. At 237 seconds the affected steam generator empties.
- d /77.4 M d oCl f/) Stad 15.1-12 r l
i of th mJrv.aveal. 4 a _ 4
i At 259 seconds the maximum core reactivity (+ 0.09 Sao ) occurs. Safety i injection boron begins to reach the core at 280 seconds. As shown by Figure- ..
15.1.5-1.16, the values of DNBR remain above those for which fuel damage would 'l ,
I be indicated. A a maximum of 30 minutes the operator, via the appropriate emergency proc , initiates plant cooldown by manual control of the atmospheric dungsfvalves, assuming that offsite power has 0 not been restored.
Shutdown cooling is initiated when the RCS reaches 350 F ar.d 400 psia.
j Case 2: Large Steam Line Break During Full Power Operation with Offsite Power Available (SLBFP)
The knamic behavior of the salient NSSS parameters following the SLBFP is presented in Figures 15.1.5-2.1 through 15.1.5-2.15. Table 15.1.5-2 summarizes
- the major events, tiines, and results for this transient.
t At 6.95 seconds after the initiation of the steam line break a trip signal is initiated by the core protection calculators on a projected DNBR of 1.19'. At 7.1' seconds the reactor trip breakers open. ^fter : 0.X m.;;d ;;i! i;:7 de 4 , th: CEfa L;M., w m y int; th: 7 ; et 7.00 r rt. At 11.9 sec voids begin to form in the upper head of the reactor vessel. At seconds IQ the steam generator pressure drops below the MSIS setpoint of 810 ia. N dWW MSIS initiates closure of the MSIVs and MFIVsp The MFIVs and the operable I MSIVs close by 18.5 seconds. EFW is automatically initiated to the intact steam generator, assuming no delay after the EFAS signal on low level in the i #\ intact steam nerator, at 18.5 seconds. At 67 seconds the pressurizer
! Sj,4 empties. At seconds theJressurizer pressure drops below 1600 osia and s N.b itiates a SI . Within Q9/ seconds of SIAS the HPSI pumps reach ful' speed l nd the HPSI valves are fully open. At 149 seconds the affected steam -
M 984 nerator empties. At 151 seconds the maximum core reactivity (-0.18% Ao) }
occurs. Safety injection boron begins to reach the core at 160 seconds. The values of DNBR remain above 10 during the post-trip approach-to-criticality portion of this transient. At a maximum of 30 minutes the operator, via the appropriate emergency procedure, initiates plant cooldown by manual control of the turbine bypass valves. Shutdown cooling is initiated when the RCS reaches 3500 F and 400 psia.
,, [ Case 3: Large Steam Line Break During Zero Power Operation with Concurrent
(-
Loss of Offsite Power The dynamic behavior of the salient NSSS parameters following the SLBZPLOP is presented in Figures 15.1.5-3.1 through 15.1.5-3.15. Table 15.1.5-3 summarizes the major events, times, and results for this transient.
Concurrent with the steam line break, a loss of offsite power occurs. At this time an actuation signal for the emergency diesel generators is initiated. Due to decreasing core flow following loss of power to the reactor coolant pur.ps, conditions exist for a low DNBR trip. At 0.6 second a low DNBR trip signti is initiated by the core protection calculators. At 0.75 second the reactor trip J.\ breakers open. d
--81 hay dal y , th: MA_t 50 '- t: l ei. - l 1
6.0 '-t?-tM *: 7 et Aftes'a 0 14 ==re M seconds the steam generator pressure e 1.00 !^'a-'t. At drops below the MSIS setpoint of 810 psia. The M S initiates closure of I The MFIVs and MSIVs close by seconds. EFW is /a.6 MSIVs and MFIV)itiated to the intact steam generator, assuming n delay after automatica11[n the EFAS signal on low level in the intact steam generator, at seconds.
.l }
10.6 .
sht 6.0 hds 15.1-13
i ap).fr b pressuriz:;r pressure drops below 1600 psic and initiater o At$ye SIA s:conds seconds of SIAS the operabla HPSI pump is loaded on the Within
( dietels and reaches full speed and the HPSI valves are fully open. At 55 At 59 ggseconds heconds voidsthe begin to formempties.
pressurizer in the upper head Safety of theboron injection reactor vessel.
begins to reach the core at 120 s . At 189 seconds the maximum core reactivity (-0.06% ao) occurs. At 12 seconds the affected steam generator empties. The values of DNBR remain above 10 during this transient. At a maximum of 30 minutes the operator, via the appropriate emergency procedure, initiates plant cooldown by manual control of the atmospheric dump valves, assuming that offsite power has not been restored. Shutdown cooling is initiated when the RCS reaches 350'F and 40 psia.
Case 4: Large Steam Line Break Zero Power Operation with Offsite Power Available(SLBZP) l The dynamic behavior of the salient NSSS parameters following the SLBZP is presented in Figures 15.1.5-4.1 through 15.1.5-4.15. . Table 15.1.5-4 sumarizes the major events, times, and results of this transiens 5.W G H.1 mWdr At $44 seconds after initiation of the steam line break, the steam generator l pressure drops below the low steam generator pressure trip and MSIS setpoint of 810 psia. At 6.79 seconds the reactor trip breakers open x . A't: e a u ueand san daeay thy, the ES WM iv e up iniv Uie uv . n 7. O m aade. The S initiates closure of the MSIVs and MFIV @ The MFIVs and MSIVs close by 11 2 seconds. EFW is automatically initiated to the intact steam generator, g ssumino no delay after the EFAS signal on low level in the intact steam ti,y generator, atg seconds. At W seconds the ressurizer pressure drops s O-[
~
below 1600 psia and initiates a SIAS Within seconos of SIAS the operable 79.4 PSI pump reaches full speed and the HPSI valves are fully open. At 48 seconds voids begin to form in the upper head of the reactor vessel, At 52 seconds the
{s pressurizer empties. Safety injection boron begins to reach the core at 110 seconds, At 310 seconds the maximum core reactivity (-0.02% ao) occurs. At g 418 seconds the affected steam generator empties. The values of DNBR remain above 10 for this transient. At a maximum of 30 minutes the crerator, via the yl,lo appropriate emergency procedure, initiates plant cooldown by manual control of the MSIV bypass valves associated with the unaffected steam generator and M turbine bypass valves. Shutdown cooling is initiated when the RCS reaches 350*F and 400 psia.
. Case 5: Steam Line Break Outside Containment During Full Power Operation with Offsite Power available (SiBFPD)
The dynamic behavior of the salient NSSS parameters following a typical limiting SLBFPD is presented in Figures 15.1.5-5.1 through 15.1.5-5.8. Table 15.1.5-5 sumarizes the major events, times and results for this transient.
The consequences of this transient -- fraction of fuel rods predicted to
! experience DNB -- are nearly the same as those for SLBFPDs for a spectrum of break sizes, due to the protective action of the core protection calculators (CPCs). See the discussion in Section 15C.3.2 and Figure 15C-1 of Appendix '
15C. The largest break size yields the minimum DNBR. Therefore the transient presented here is that which results from the double ended break of a main steam line.
.e Not later than 5.85 seconds after initiation of the steam line break, a trip signal is initiated by the CPCs on a projected DNBR of 1.19. At 6.00 seconds 15.1-14
_ _ - _ _ . _ .____2_________..____._ _ . . _ _ _
the reactor-trip breakdrs operr .u+w a u e ran a g e-g %
CCM h -i- +a d- 94-+k-c- MSE xx-2. At 7.49 seconds a minimune is calculated to occur, after which DNBR rapidly- .
transient unon of seconds voids begin to fo %
flc increases,as'sh Figure 15.1.5-5.9. At Teconds the steam generator /2. A I
' '3 j
in the upper head the reactor vessel. At pressure drops bel the MSIS setpoint of 810 ia. The MSIS initiates closure of the MSI and MFIVg The MFIVs and the operable MSIVs close by l 17.8 seconds. ggyg, Subsequently,(theeventsofthistransientfollowasequencesimilartothose of the SLBFP Case 2), Since the cooldown is less severe the potential for post-trip degradation in fuel cladding perfonnance is less for this case (SLBFPD) than for Case 2 (SLBFP). At a maximum of 30 minutes the operator, using the appropriate emergency procedure, initiates plant cooldown by manual control of the turbine bypass valves. Shutd wn cooling is initiated when the RCS reaches 350*F and 400 psia.
At the point of the minimum transient DNBR no more than 0.4% of the fuel rods are predicted to experience DNB. However, as a bounding assumption 0.7% of the fuel pins are assumed to fail. All of the activity in the fuel gap for fuel rods that are assumed to fail is assumed to be uniformly mixed with the reactor coolant. The activity in the fuel clad gap is assumed to be 10% of l
the iodines and 10% of the noble gases accumulated in the fuel at the end of core life, assuming continuous full power operation, This results in a primary coolant activity of 618 pCi/gm. Assuming one gpm steam generator tube leakage, during a period of two hours after initiation of the SLBFPD the integral
. leakage from the RCS through the affected steam generator is 720 lbm, which is l
assumed to be released to the atmosphere with a DF of 1. This mass release results in a contribution to the inhalation thyroid dose at the Exclusion Area Boundary (EAB) of not more than 220 rem. I The total steam released from the af'ected f steam generator is 153,000 lbm.
The affected steam generator will empty in two hours; therefore all the massThe release from the affected steam generator to the atmosphere has a DF of 1.
calculated inhalation thyroid dose is not more than 9.8 rem for the blowdown originating from the secondary system fluid discharge from the affected steam generator.
- Less than 86,000 lbm of steam from the unaffected steam generator will be l
released through the steam line break. During the SLBFPD the MSIVs will isolate ,
' the unaffected steam generator from the break and prevent it from emptying.
Therefore, a DF of 100 is assumed in calculating iodine activity released from the unaffected steam generator. The resulting contribution to the inhalation thyroid dose at the EAB is less than 0.1 rem. Should condensor vacuum be lost l during this transient, up to an additional ~ lbm of steam from the unaffected steam generator would be released to he atmosphere through the This would result i an additional contribution atmosphericsteamdumpvalves.9) rem. 6oco l I to the dose of not more than4 AT The foregoing doses are calculated by the methods outlined in Section 15.0.4.
Table 15.1.5-11 presents the major assumptions, parameters, and radiological consequences for this transient. _
In summary, the total two-hour inhalation thyroid dose at the EAB as a consequence of the SSLBFP is no more than 231 rem.
Oh 15.1-15
Case 6: Large Steam Line Break Outside Contatuant from Zero Power Operation with Loss of Offsite Power (SL8ZPLOPO)
Case 6 is i uded in Case 3, since the break of the latter can be either inside or . de of containment. The Figures, Tables, and Discussion for Case 3 apply to Case 6.
Assuming one gym steam generator tube leakage, curing a period of two hours after initiation of the SL8ZPLOPD the integral leakage from the RCS through the affected steam generator is 720 lba, which is assumed to be released to the atmosphere with a OF of 1. This mass release results in a contribution to the inhalation thyroid doses at the EA8 of:
(a) 1.6 rem, assuming technical specification primary coolant actii vity; (b) 20.1 rem, assuming a pre-existing fodine spike; or (c) 41.5 res, assuming an event-induced iodine spike.
The total steam released from the affected steam generator is 300,000 lba, l
- which is the total initial mass inventory. The affected steam generator will empty in two hours; therefore all the mass release from the affected steam generator to atmosphere has a DF of 1. The calculated inhalation thyroid 2
gose is IES rem for the blowdown steam originating from the initial steam l jg, generator mass inventory.
Less than 850,000 lbs of steam from the unaffected steam generator will be released through the atmospheric steam dump valves and through the steam line
' break within two hours. During the SL8ZPLOPD the MSIVs will isolate the unaffected steam generator and prevent it from emptying. Therefore, a DF o"
- 100 is assumed in calculating f odine activity released from the unaffected
- steam generator. The resulting contribution to the inhalation thyroid dose at the EA8 is 0.4 rem.
The foregoing doses are calculated by the methods outlined in Section 15.0.4.
Table 15.1.5-11 presents the major assumptions, parameters, and radiological consequences for this transient.
- In summary, the total two-hour inhalation thyroid dose at the EA8 as a consequence of the SL8ZPLOPD is no more than 56 rem.
15.1.5.4 Conclusion For the large steam line break in combination with a single failure and stuck CEA, with or without a loss of offsite power, fission power remains sufficiently low following reactor trip to preclude fuel damage as a result of post-trip return to power.
For a large steam line break during zero power operation in combination with a loss of offsite power and technical specification tube leakage the two-hour inhalation thyroid dose at the EA8 is weil within 10CFR100 guidelines:
(a) 16 rem, assuming technical specification primary coolant activity; (b) 35 ren, assuming a pre-existing iodine spike; or (c) rem, assuming an event-induced iodine spike.
15.1-16
TABLE 15.1.5 1 1 .1 '
~
SEQUENCE OF EVENTS FOR A LAFd 5ILAM LINE 8 REAR DURING FU optaATtum Wtin cdNCURRENT LD53 0F OPP 51It POWER (3LuPPLOP)
Event- Setooint or Value Time (SecQ 0.0 Steam Line Break and Loss of Offsite Power Occur Low ON8R Trip f: df ti - 0 : r") 1 19 0.6 Projected ON8R '-sc p t Ga w .Drd ,
Trip Breakers Open g g g >0.75 ~
l "A" -- _ -.
8.0 Voids Begin to Form in RV Upper Head
.ste- l ser 9.7 Main Steam Isolation Signa 1@
13.3 MFIVs Close Completely l --
13.3 MSIVs (. lose fompletely g g,7 .. --
13.35 EFW Init1ated to Intact Steam Generator 120 Pressurizer Empties
/ USER.T~ -_ m.
" g" .LW/ 76N Safety Injection Actuation Signal @ N 208 Safety Injection Flow Begins 237 Affected Steam Generator Empties
+0.09 259 Max { mum Transient Reactivity, 10~ ao Minimum Post-Trip DN8R 2.7 277 280 Safety Injection Baron Begins to l
Reach Reactor Core 1800 Operator Initiates Cooldown A
T' 151.5-1 74,7 N** "
$4K. M f'- k ST4.fv U"Arkeg ka.4K st= = fs./at .n sip (.(Hsrsh AWys;5 SeKseiw..C")]> sin S/C B rr1.+i%ssu.rh<r h.ssu.ce. he.L.<.s %%'tf-rm;ecr:.m Acw t:.w 5, g GsrA h A 4 45 600 Secho; z ,_wsia 1
i
'h . 53 hm G m n L r- 6 ,2eA
.yg. F,#-s A ct.-ej.~ @
e hf6 CccpK~C, l f A :
2s I
TA8LE 15.1.5-2 SEQUENCE OF EvtllTS OktAAT10m FOR WLIM QPr3 LIE A LARGE PQWER AVAILABLE STEAft (5LBPPI LINE BREA
"" Event- _
Setooint or Valut Time (Sec_)
0.0 Steam Line Broek occurs 6.95 Low DN8R Trip Mg 1.19 Projected DNGR g; 7.10 Trip Breakers Open l
- ."; : ;f- t: 0 :;
7.M Voids Begin to Form in RV Upper 11.9 Head
/NSEET M i
.13413,9 Main Steam Isolation Signal "A' Gamanitsi --
MFIVs Close Completely 18.5 MSIVs close Completely 18.5 j g gy --
"8 18.'5 EFW Initiated to Intact Steam Generator Pressurizer Empties 67 jgy- -
Safety Injection Actuation Signalg
-Mer f t #9o.4 OG a.a.ca(Cd l
l O --
Safety injection Flow Begins 120 Affected Steam Generator Empties 149
-0.18 151 Maximus Transiegt
' Reactivity,10~ .to 26 Minimum Post-Trip ON8R 151 160 Safety injection Boron Begins to Reach Reactor Core 1800 Operator Initiates Cooldown l
/ $, . E ,
< fi i z.s sw-~6 -rew L em pk kace 8k 9,A / 4 m w p ei U
(.
is.c st Lu& Lua.u&d 2c baA Cepw$ FaA.A huu C& N Jpk s:ehas,1wc.f wwu.
- y
<e ,.
-MN n n x b w % fm '
p,oo l 6 */it5 Q W ic L L '
%& lfkGr);if pa I
l
TA8LE 15.1.5-3 LOPD).
O _
Event _
Setooint or Value .
i TimeJ!ac) 0.0 Steam Line Break.and Loss of Offsite Power Occur 1.19 0.6 Low DN8R TripCConhuna Projected DNSR L.5,g%l Gusta: cad) 0.75 Trip Breakers Open
- .'.; ::;P ;; :.
jygg 1.00 _
4h>
Main Steam Isolation Signal "N ho MFIVs Close Completely 19er'/o.le M /4.le MSIVs Close Completety
, g --
- g" 10#ho,4, EFW Initiated to Intact Steam Generator
- -ader J M ETLT~ At" Safety Injection Actuation Signa
, "c" & GeseedtL Voids Begin to Form in RV Upper Head SS 59 Pressurizer Engties t
-- l 75,1 Safety Injection Flow Begins 120 Safety Injection Boron Begins to Reach Reactor core
-0.06 189 Max { sus Transient Reactivity.
10~ Ao Affected Steam Generator Empties 1240 1800 Operator Initiates Cooldown t
l l
l f
1 ,
k 0
6 t..
T /C.t.c-3 "A " 4.o it=a.1!! A fm m 0: A: gto AL,:<l==Tsa m sty?
/ 40Ai f s $4)>as#, f sin.
[g /0. 6 f4w M /.4./d-
,r4& Q AA-*m, GL,s-E e -" = A hauge cap ,R CQ % uis .~ p (c) vv.c b- %--- O /6 a rgf a . --
w $4 ge 0 ="-
/Aen npog'mfp g- G g
. 41 %
i TA8LE 15.1.5-4 l NE BREAK OURING Z'r.Rf1 Pf*ER SEQUENCE OPdvani5 FOR A l ARGE STEAM E ( 3L54P L .
2,;=_ _=mTIUm WITM UPr31It PUWLF, AVAILAeL Setpoint or Value Event Time (Sec) __
0.0 Steam Line areak Occurs fuse.pT- -
.. A # Low Steam Generator Pressure M .dter i
- Trip and Main Steam Isolation Sf gnal@ G west,,ed.
6.79 Trip Breakers Open
- 1
'" ::7.; ? :; ' M a ;
,Lk3" //,2 MFIVs Close Completely M51Ys Close Completely g Mrff //.2 --
JH 1. '2. EFW Initiated to Intact Steam
8" Generator teos
/AMEMT # r t//k Safety Injection Actuation Signate A
c- g aw.1 Voids 8egin to Form in RV Upper Head 48 52 Pressurizer Empties b
Safety injection Flow Begins
.X 7/. 2. --
110 Safety Injection Boron Begins to Reach Reactor Core
-0.02 310 Maxjous Transient Reactivity.
- 10~ ao I --
418 Affected Steam Generator Emoties 1800 Operator Initiates Cooldown l
l l
i 0
TilAs IC;I c _
s+ =-> h. - s Pe=- a % ,A n eio
.c.se -
f'"seL% rp D -+ -45.,
d 4 *-- -
4(o x@ 5 +-
/nh C+w %m ka Af A&ctN y f &2 .se f a r Se,u $ .
P^I - ..
, ,E ,,
s+n = = .bneswi%ko ae ll % g ra d A 9-J ave smyL s
hA "? )
% wha 4
'I f/
- e. e.
% = < P ,=2., ia w Pm%g:..%
a s,
. A%2p Ad4+, .s&-~_ .
p.m .
TA8LE 15.1.5-5 ' I SEQUENCEGE EVENTT FOR A STEAM LINE BREAK OUTSIDE
- ' OURInts Pt--M= UranArton WIIn orP31Tt Powta Setpoint or AVAILABLE Vafuer (i
M Event Time (Sec P --
0.0 Steam Line Broek Occurs 1.19 .
Low ON8R Trip 5.85 M Projected SigM Gaureted --
6.00 Trip Breakers Open
^
^
,~
M C-'_ _,. .. ..,,
1.11 7.49 Minimum Transient DN8R 8.94 Voids Begin to Form in RV l*?W- A Upper Head Main Steam Isolation Signale Gwrnmrl 6 M l3.2.
/ MSEtt-T -
EFW Initiated to Intact Steam
~
" B" - 17.8 Generator MFIVs Close Completely 17.8 MSIVs Close Completely
-- 17.8 _
Safety In ection Actuation M
c," 5 6 5 1. SignalC1@ Ge m.cers#
(
l 1.92 75 MaximumPost-tr{pTransient Reactivity, 10' ao Safety Injection Flow Begins _
a Sr 95.2 Affected Steam Generator Empties _
100 200 Safety injection Baron Begins to Reach Reactor Core
-2.06 430 SecondaryPost-trip,jransient Reactivity Peak, 10 ao 1800 Operator Initiates Cooldown 1
'l T I S. I. S- S" ;
M
- a-i n.i.x;. og " Ami ts:s 54* % ?*8IC .
w.re.r-5 9Cm Goem't'ar"La.9 e.l Pu, L, , gp, an_
17.8 Feedu. rot.e - Ac.D -tiea T 0h 1 A4 sis s.4t'o.', wt. ,7.ce.. r oc<m;a me I
25 t It bh '
("G 54 L4, t' t,f.,f'- Pe 6TkN #^"hA,6 b Age.t_t:,w Aer%r w s:p %' (sIAQ Aut gs:s sez. 's. wt , pstc~ neo i
i e
\
l 15.2.3.3 Analysis cf Effects and Cras qui.nc0s
- - A. Mathematical Model The NSSS res to a LOCV was simulated using the CESEC-II computer program i described in S ion 1E.0. The initial DNSR was calculated using the TORC computer code (see Section 15.0.3.1.6) which uses the CE-1 CHF correlation
- described in Reference 19 of Section 15.0.
! B. Input P.arameters and Initial Conditions i
The input parameters and initial conditions used to analyze the NSSS response to a LOCV are discussed in Section 15.0. Table 15.2.3-4 contains the initial conditions and assumptions used for this event. The initial conditions for the principal process variables were varied within the ranges given in Table 15.0-5 to determine the set of initial conditions that would produce the most adverse
' consequences following a LOCV. Various combinations of initial core inlet
) temperature, core inlet flow, pressurizer pressure, steam generator level and l
pressurizer water level were considered in order to evaluate the effects on peak reactor coolant system (RCS) pressure.
l .
! Decreasing the initial core inlet temperature reduces the initial steam l generator pressure, thereby delaying the heat removal associated with the
- opening of the main steam safety valves. However, the initial 0 inlet temperature for this event was restricted to a minimum of 560 F. Decreasing i
the initial inlet temperature (as well as increasing the initial core flow rate) also minimizes the core average coolant temperature which results in the
! most positive moderator temperature coefficient.
l Reduction of the initial pressurizer pressure delays the occurrence of reactor trip on high pressurizer pressure and allows the maximum reduction in steam generator heat removal prior to and following trip. As a result maximum RCS overpressurization occurs, provided that the delay does not allow the main steam safety valves to open prior to reaching the peak pressure condition.
Decreasing the initial pressurizer water level produces similar trip delays.
C. Results The dynamic behavior of important NSSS parameters following the loss of condenser vacuum is presented in Figures 15.2.3-2 to 15.2.3-14.
The sudden reduction of steam flow, caused by the LOCV leads to a reduction of the primary-to-secondary heat transfer. The moderator reactivity increases slightly prior to the reactor trip due to a positive MTC as the average core temperature increases from the initial conditions. This added reactivity causes the core power to reach a maximum at 6.8 seconds. The rapid heatup of d the reactor coolant results in a high pressurizer pressur trip condition at g, O Mthe seconds. TheGw -d; eg;;m " :t ::: re' at . seconds and ifmit maximum core power to lun of it i power. W (,,99]
(frg ,,, Ea r t e fp br M ope o The pressurizer safety valves open at 6.9 seconds and the manfmum RCS pressure of 2742 psia is reached at 8.6 seconds. The main steam safety valves open at
@ seconds and the maximum secondary pressure of 1353 psia is reached at 14.0 l seconds.
15.2-5
44.1 The RCS pressure decreases rapidly due to the combined effects of reactor trip and primary and main steam safety valves. The pressurizer safety valves close at 12.0 seconds and the main steam safety v 1ves close at 346.0 seconds.
Emergency feekster flow automatically begins at . seconds and continues to {.
At Igl@ fill theseconds stam g6nerators until a safety injection a normal actuation signal islevel is reached generated when the at 1408l secon pressurizer pressure decreases below 1580 psia. Borated water enters the RCS at 1150.0 seconds from the high pressure injection pumps. Thirty minutes after initiation of the events, the operator commences a cooldown using the atmospheric dump valves to release steam.
The DN8R during the event, remains above its initial value of 1.4; therefore, DN8 does not occur.
i D. Single Failures i l
The LOCV event is assumed to abruptly and completely terminate both main steam i
and feedwater flow. Considering peak pressure criteria, the only mechanisms
! for mitigation of the reactor coolant system (RCS) pressurization are the pressurizer safety valves, the reactor coolant flow and main steam safety valves. The last two influence the RCS-to-steam generator heat transfer rate.
j There are no credible failures which can degrade pressurizer safety valve or i
main steam safety valve capacity. A decrease in RCS-to-steam generator heat transfer due to reactor coolant flow coastdown can only be caused by a failure to fast transfer (FFT) to offsite power or a loss of offsite power (LOP) following turbine trip (i.e., two or four pump coastdown, respectively). The two and four pump coastdowns result in an immediate reactor trip, generated by I the Core Protection Calculators (CPC's). Due to the rapid reactor trip, both i of these failures reduce the peak pressure relative to the LOCV itself.
With regard to fuel performance, decreased coolant flow is the only parameter which can significantly reduce the minimum DNBR during the LOCV event. FFT and LOP are the only single failures which impact coolant flow. LOCV by itself, however, produces an increasing DN8R (see Figure 15.2.3-2). This results in a greater thermal margin than is required to preclude a DNBR below 1.19 for either single failure. Consequently neither will cause fuel failure. LOP, however, because of the more rapid flow coastdown, causes a greater degradation in DNBR and hence is more limiting. The decrease in DNBR is shown in Figure 15.3.1-9.
15.2.3.4 Conclusions For both the loss of condenser vacuum event, and LOCV with a single failure, the maximum RCS pressure remains below 2750 psia, thus ensuring primary system integri ty. The minimum DNBR remains above 1.19, thus ensuring fuel cladding integri ty.
15.2-6 r
TABLE 15_.2.3-1 SEQUEMCE OF EVENTS FOR THE LOCV
.: Setpoint Success Time T.' or Value Path (g) Event 0.0 Loss of Condenser Vacuum f I U ,,g 5 E.E T 2M Reactivity 6 4 High Pressurizer Pressure Trip Control 6.94 Signal @ Gau_ra tat 1282 Secondary 6.7 Main Steam Safety Valves Open System psia Integrity 7
p.. . = . . _-i_ . ,. .
3 p ,_e._,_~ ,
C,eeenet s - ,
--t :' " e nr. ;;
102 Reactivity 6.8 Maximus Core Power, T. of Design Contro)
Power 2525 Priamry 6.9 Pressurizer Safety Valves, Integri ty Open, psia System g,qq ~ Trip BrmkttsO p Reactivi ty 3.+ cf *' " ;' _ :. .,,
Control Maximus RCS Pressure, psia 2742 8.6 2462 Primary 12.0 Pressurizer Safety Valves Close, System psia Integrity 14.0 Maximum Steam Generator Pressure, 1353 s
i /95EgT psia
.. c. , ,_
Emergency Feedwater Actuation S
.3 ave' M 'l Signal (74rcent of wide range]d.
b Cr e.a.= r :r e. Se:ondary Emergency Feedwater Flow 875
.Aho' System l
i Wl Initiated.gpm Integrity j pf,E6'"
1219 Secondary Main Safety Valves Close, psia i
.. D. 346.0 System
{ Integrity tion Actuation 6 Reactor Heat S6&re' Safety I Removal 9W. I Signal -
G ae.e.mDr.J.
Primary lC^' .P Safety Injection Flow Initiated System 993.7 Integri ty 4
_.--<--,m.
s u-- _ a
- s. 2. 2 - 1 "A" bes-r;w %stre. > us g, t;r 5.B'f Tr-i% A m%ies Teg ;C) psia._
2.wo e.,c.e3 Ce a rel Lf
- % bhG.Pa r- k Y r== f/,
- c. 7 7_erm Trip Aqs;s sqL; -
- >ee e.C .? u. ride bya_.
qT e G e. e d or U.ht. er- Le.ue.L km2,3 "c ' '
Emgweq FeeA der- A tut;e% c;ig sercest ,F 33 ! .
Aqds q,.;e, GAe wqa-5 s
hb k MD hhk j 1^E'- A"' * " Tf" c'63 1 A A s',s Aa % L-u ,p u --
I6 bD hiG k.ed-
%d
TABLE 15.2 2.,(unt'd) .
SEQUENCE OF EVENTS F0li THE LOCV
- Setpoint Success Time or Value Path Event (Sec)
Reactivity 4 1150.0 Borated HPSI Flow Enters the Core Control
~
~ '
- ary 1408.0 av Ade) 80
. Integrity Operator Initiates Plant Cooldown Reactor Heat 1800.0 Removal T e ($ e_w.ar.rn5.4 I N #
r, A # EFAs 12_ esc Am%5I5 i 645.iC, peceE 8 A FawgE 1
i b
i l
t
~~.
!! , iii i l,i
-i :I l':l
.n L9 i
y i.[.
g l 1.
g 4:
(?1
- 5l i
llui !! : g
_t_ : 4 i i t ,_L i "
I e , s-
_ i,g
.I
- i
, . e ji . _.
v
.i i =
iil i - s _,
g;l, 15 t
la s li .j .
(, I ff !!
.i
!,, .. .= . : :
~
ik b i i <<
! li i u
l il igi $' ll i, I
! I;r j i ill 8
I 131 a
!! J_
- L.
sii f it I !!!!j il [*i!]i] O*@il i: J:I <t i n', ! ' ' ': i l Il il ;
Il's:-.1-lAl4
- I :: : : : i :
- o_ll.,% ,l; l
- l. s Y
- - i .
. i ;
- - I -
spL 111, I
i ll 7
11:
i!!!
8H' lil li '
=.
38
[ji l1!
e4
!I
-j es lt y 3)
- g!
I it !3 li 8
I.
12 J.:' t
-ll i i'
is .
25
__. jf G
Is
- . I i i iI
TABLE 15.3.1-1 SEQUENCE OF CVENTS FOR TOTAL LOSS OF REACTOR COOLANT FLOW Time Setpoint success
~'
(Sec) Event Or Value Path O.0 Loss of Offs 1te t Power
. Diesel Generator Starting Signal
- Reactor Coolant Pumps Coast Down L..= r 7 = = = = : >7 N(essehMaa g
O i' O.6 Low DNBR Trip Signal 1.19 @ Reactivity Control 0.77 Generated /9' maid JN4 Asaet %'4y (ae, s l l
1.09 7;.,p's Begin to Dropg Q og CEA Reactivitf Control 2.6 Minimum Transient DNBR 1.19 4.3 Pressurizer Safety Valves 2525 Primary System l
Open, psia Integrity i
5.3 Maximum RCS Pressure, psia 2576 Ma,'., dee ,
, 5.4 .n; _ . . . . - = Sa fe ty 1282 Secondary System i Valves Open, psia Integrity i
11.7 Maximum Steam Generator 1338
! Pressure, psia 12.2 Pressurizer Safety Valves 2463 Primary System Closed, psia Integrity I
! 1800.0 Operator Initiates Plant Cooldown O
s
( .. . .
l
^
.--,-,-,---.---.-r,-.
1.s s i!! ==
i i
,l .l. I l>i lii '[ .ii -
0 l!
i !!! l li !.
J. J. J_:' 4 ii
}Q;i '
!1 -. 2 ir, 2:;-. . :
tji
- =
i
![il -i (V jj ei i
i 1 5
m < _s gg "
g
,ll' tt ,. .t -@
1, fr - -<
If 3l
- 5E
!!! 11 ! 8 ;E I It I
[j g6 sa 5E 1
- is
!! II
!! Il1 I!!'
l' _t t:
, j!- - 5 i l N
'N d'
k=s n! es;n
! I i ll _
8 I
[ di e ;
l n{
b
. Il C.A !l1 I il '-
-+
I!!,i is, s!ii i
j,ggg 1135
/ en li s -
11 1 i:
3 !
11 il i io i t
- li I I II! i Il ij:! l' !! 'j' :5
!! ! !!!!h !I
, ].
I J:i jv- . i _f:IT
_1 91 : . 4 i! :1 tli I i15
~
T 2-- + : : gll3 "
_ s e e i l[ -+ [ 8 3
- js 15 a 1 5 a --
! It l t !] : :. I is il g 11 i
. \.\.
,- n. ,,n __ _ . . -
- . . - ~ . _ - , ,.-e-. . . - ., n ,
Table 15.3.3-1 (Sheet 1of2) .
SM 0F EVOf73 FOR THE SINGLE REACTOR COOLANT P
~
. Ruivu 3EIIUREFRQM WLin t.033 QF QPr5ITa PUWER Rt3ULTINii TUR51Mt TRIP _
Total
- Satpoint Integrated
. Safety Valve Success or Path Time Value Flow (1be) _
(sec.)_ Event 0.0 Seizure of a Single ~
Reactor Coolant Pump Reactivity Low ON8R Trip Signal 1.19 _
Control 0.76 Generated, projected Reactivity
' CEAs Begin to Orop _ , Control 1.25 Into the Core
"""~
~
Turbine Trip / Generator ~
1.25 Trip 0.967 _
- Secondary Main Steam Safety 1250 _
System 4.1 Integrity Valves Open, l Unaffected Loop, psia Maximus RCS Pressure, 2387 -
4.2 psia i
Loss of Offsite Power s 4.25 Occurs Secondary
- Main Steam Safety Valves 1280 _
System 4.5 Integrity Open. Affected Loop, psia 3,492 ~~
1347
! 6.8 Maximum Steam Generator Pressure Unaffected Loop, psia 5,451 -
, 1340 7.4 Maximum Steam Generator i
' Pressure, Affected Loop, psia l
}7_esACiVIk 7 enss~ Tr-), -
a ,,g
- 0,91 BreKes ofe.n 1
r
Table 15.3.3-1 (Continued)
(Sheet 2 of 2)
- REACTOR C00UM PUMP l
SEQUENCE OF EVENTS FOR THE SINGL illt PUWER RE3ULTIMrs Rgivn 5tI4URE WLIM LQ33 0F OPP, i
FROM TURBINE l' RIP _
ggg Total
! "A~ Setpoint Integrated Safety Valve Success or Path Time Value Flow Libm)
(Sec.) M Secondary' S 85,679 218~ G ater Level EFAS' System
.,1 Set int Rea nerator, in Integrity 5
j
- 5 ne ent of w e rangej l 91,407 Secondary 119 253 Emergency Feedwater System Begins Entering Steam Integrity Generator, Unaffected
- Loop, Ibm /sec 115,189 Secondary 697
~
w Water Level EF ]
5 Systen Se t Rea in - Integrity i
E'h6 the Ste Aff 4 Loop, p stor.
t I ,
' M"g*
- wide rance J Emergency Feedwater 119 Begins Entering the .
Affected Loop. Ibs/sec 120,398 Secondary 1218 821 Steam Generator System Sa~ety Valves Close, Integrity Affected and Unaf-facted Loop, psia 120,398 Secondary At:noscheric Oumo
-100.0 System f 1800 l
Valves Opened to Integrity l
Initiate Plant Cooldown,'F/ hour.
/
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1,128,7.93 ~ -
7200 Total Steam Release to Atmosphere, iba
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corresponds to the largest insertion rate expected from the secuential -
withdrawal of the CEA groups with 40% overlap at the maximum speed of 30 in./ minute.
C. Results The cynamic behavior of important NSSS parameters following a CEA withdrawal from low power conditions is presented in Figures 15.4.1-1 through 15.a.18 The withdrawal of CEA's from' low power conditions (1 MWt power) adds reactivity to the reactor core, causing both the core power level and the core heat flux to increase. The power transient causes increasing temperature and pressure transients, which together with a top peaked axial power distribution, produce the closest approach to the specified acceptable fuel design limit on OHBR.
Since the transient is initiated at low power levels, one of the normal reactor feedback mechanisms, moderator feedback._does t contribute to any appreciable A seconds into the transient, l 23,75'Wtent to the power excursion transient. s becin drocoina into the core a variable overpower trip is actuated. The C
- anal +-<6F24.2 seconds whicm terminates the transientpith a hot channel minirnum ON8R1 l If the maximum rod radial peaking factor occurs in the region of the l
<tTr 5.47.
axial power peak, the peak linear heat generation rate during the transient l reaches KW/ft.
15.4.1.4 Conclusions 1 The uncontrolled CEA withdrawal from a subcritical or low power condition event meets general design criteria 2F and 20. These criteria reouire that the specified acceptable fuel desica limits are not exceeded and the protection system action is initiated automatically. The withdrawal of CEA's from low power conditions meets the following fuel design limits which serve as the acceptance criteria for this svent: the transient terminates with a hot channel minimum DNBR greater than or equal to 1.19 and the peak linear heat generation rate during the transient is less than 21 KW/ft.
6fch.nn.L veaded during th hsient s h y mini m 6 N at.,
is 4sB4 dt py x sec. sed.s.
w 15.4-3
. TABLE 15.4.1-1 SEQUENCE OF EVENTS FOR THE SEQUENI LAL CEA WlIMDRAWAL EVENT Setpoint Success Event or Value Path Time (sec)
Withdrawal of CEA's - -- Reactivity 0.00 Control Initiating Event Variable Overpower i+:t Reactivity
- " 0::t;- Control 23 75
. Trip.
i newer Soyund GenemM l C:"" 77].P.;;r !:;;'y -- Reactivity 23.90 .
Control Breakers Open
" ::ti ci ty-
-G4,4 CE?': B:;fr t: Dre; --
C:-t 1 j Maximum Core Power, 45'9 ggg % of Design Power Maximum Core Average
!?+53
, M&S Heat Flux, % of Full Power Heat Flux 2.7.0 Minimum DNBR 4+84 Maximum Pressurizer 1894
( 35.2d Pressure, psia I
17.e Reub'h 33*5S* care Pm Reeles Cen h l l
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TABLE 15.4.1-4 ASSUMPTIONS AND INITIAL CONDITION FOR THE LOW POWER CEA WITHDRAWAL ANALYSIS
- Parameter Value Initial core power level, MWt 1 Core inlet coolant temperature, F 564.5 Core mass flowrate,10 6 lb,/h h IdY Reactor coolant system pressure, psia 1785 One pin radial peaking factor, with uncertainty 2.53 Steam generator pressure, psia 1178.
Moderator temperature coefficient,10-4 ao / 0F +0.5 Doppler coefficient multiplier .85 CEA reactivity addition rate,10-4 ao/ Osec 2.5 CEA Worth on trip,10-2 ao -. f. 4 Steam bypass control system Manual
< a. The scram worth used in this analysis does not take credit for the additional worth available from the withdrawn CEA's and is therefore considered conservative. Furthermore, the worth assumed is less negative than that calculated or expected.
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Other input parameters which ara imp rtant to this analysis ara th2 Moderator Temperature Coefficient (MTC) and Fuel Temperature Coefficient (FTC) of reactivi ty.
A moderator temperature coefficient was assumed in this analysis which corresponds to beginning-of-life core conditions. This MTC has the smallest icpact on retarding the rate of change of power, coolant temperature, and DNBR.
A fuel temperature coefficient corresponding to beginning-of-life corditions was used in the analysis, since this FTC causes the least amount of negative reactivity change for mitigating the transient increases The uncertainty in coreon the power, heat flux, and the reactor coolant temperatures.
fuel temperature coefficients used in the analyses is listed in Table 15.".2-4.
The regulating CEA position from which the CEA withdrawal This is initiated particular corresponds to 25% insertion of the first regulating bank.
insertion was selected based on the calculated CEA worth and A corresponding associated maximum uncertainties to oroduce the worst transient.
differential worth of 0.011, ao per inch of rod motion was conservatively This corresponds to a maximum reactivity assumed in the present analygis.ao per second based on the maximum CEA withdrawal rate of 0.5 x 10-withdrawal speed of 30 inches per minute, including all uncertainties.
All the control systems listed in Table 15.a.2-2, except the steam bypass control system, were assumed to be in the automatic mode sinca theseThe systems stean have no impact on the mininum DNBR obtained during the transient.
bypass control system is assumed to be in manual mode because this ninimizes DNBR during the transient.
C. Results The dynamic behavior of important NSSS parameters following an uncontrolled CEA group withdrawal are presented in Figures 15.4.2p2.to 15.4.2p. 12.
l The withdrawal of CEA's'causes a positive reactivity change, resultino in ane reac increase in the core power and heat flux. As a consequence seconds after I coolant temperature and pressurizer pressure increase. At initiation of the transient, a reactor trip on low DHRR is ac uated. At @ *n l seconds the trio breakers are opened. The CEA's begin dropping into the core The minimum DNRR reached during OMO S0.0 seconos whic]9 terminates the transient.If the maximum rod" radial peaking
% the transient is 1.19 at 11.0# seconds.
factor occurs in the region of the axial power Deak, the peak linear neneration rate during the transient reaches 16.7 KW/f t. Table 15.4.2-1 lists the sequence of events for the limiting DNBR case.
15.4.2.4 Conclusions The uncontrolled CEA withdrawal event meets general design criteria 25 and 20.
These criteria require that the specified acceptable fuel design limits The are not exceeded and the protection system action is initiated automatically.
withdrawal of CEA's from full power conditions meets meets the following the fuel design limits which serve as the acceptance criteria for this event:
transient terminates with a hot channel minimum DNRR greater than or poual to 1.19 and the peak linear heat generation rate during the transient is less than 21 KW/ft.
15.4-6
.. .,_- - .- - _ - - - __ _ - - -.--- -- - a
~
TABLE 15.4.2-1
- SEQUENCE OF EVENTS FOR THE SEQUENTIAL CEA WITHDRAWAL EVENT SETPOINT SUCCESS OR VALUE PATH TIME (sec) Event Reactivi ty .
0.0 Withdrawal of CEA's - --
Initiating Event Control Low DNBR Trip Si~ l 1.19 Reactivi ty 95/ bW9- Contro1 GenwLtal,ftei CE0" ":x:r h ;;ly -- Reactivity cfg 7tif Breakers Open Control Reactivi ty 10.0 CCA'5 Cegi.. te 0.ep --
Control Maximum Core Power, 108.2 10.1
- . of Design Power Minimum DNBR 1.19 11.0 Maximum Core Average 105.6 11.4 Heat Flux, % of Full Power Heat Flux Maximum Pressurizer 2363 12.3 Pressure, psia s
! /
TABLE 15.4.8-1 isneet 1 of z) f SEQUENCE OF EVENTS FOR THE CEA EJEGHON EVENT _
Setpoint Success
- Time '
or Value_ Path Event (sg) 0.0 Mechanical Failure of CEDM Causes CEA to Eject
~%uis.oure. <r 1&~ Reactivity LC ^ _.c Trio Si%
G syn ~r*L Contro1 00 W ==
0.05 CEA Fully Ejected
' Maximus Core power, 138.3 O.08 i
! l % of design power ity f --
hf'kicipopeM heact
- ' g i
Secondary Integrity
' OU Turbine Trip Occurs 1282 Secondary Main Steam Safaty System 2.53 Valves Open,-
' ' Integrity psia l
936 2.6 Maximus Clad Surface Temperature in the.
Hot Node, F
~
3779 3.8 Maximus Fuel Centerifne Temperature in the Hot Node, F 2525 Primary 3.9 a . Pressurizer Safety Valves System Open, psia Integri ty
- o. o a&T- zas n, z-wg brale. cues-@w .r- e c_ m.1 go I Gr- Te;> h % s s Ggfe1% pN*:T~aY tkesiy [owtr
TABLE 15.4.8-1 (Cont'd) (Sheet 2 of 2)
SEQUENCE OF EVENTS FOR THE CEA EJECTION EVENT Time Setpoint Success Event or Value Path (Sec) 3.9 Maximum Pressurizer 2525 Pressure, psia 4.7 Pressurizer Safety Valves 2462 Primary Closed, psia System Integrity Maximum Steam Generator 1348 rf9 Pressure, psia 4+ CC".; M'; I ::cted, A cm. om o_.->_-.
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f 40.2 Safety Injection Actua- 1!!ISW Reactor Heat tion SignalW Removal g
h L Geriera3"Wd.
Secondary 850 /NSE1tT' Main Steam Safety 1250 "g* Valves Closed, psia System Integri ty 1800 Operator begins plant -- Secondary cooldown System Integrity
- 12230 Shutdown cooling 400/350 Reactor Heat Removal initiated, RCg pressure, temperature, F l
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15.4.8-1 contains the sequence of events that occur during a CEA Ejection initiated from full power BOC initial conditions.
Ejection of a CEA causes the core power to increase rapidly due to the almost -
instantaneous addition of positive reactivity. However, the rapid increase in core power is te.minated by a combination of Doppler feedback and delayed neutron effects. This increase in power results in a high power trip and the reactor power begins to decrease as the CEA's enter the core. Reactivity ef fects are shown in Figure 15.4.8-7.
In the hot channel, the increase in heat flux is such that DNB is calculated to occur, resulting in:
- 1. A rapid decrease in the surface heat transfer coefficient.
- 2. A rapid decrease in heat flux.
- 3. A rapid increase in clad temperature.
The rapid increase in clad temperature is sufficient to override the decreased surface heat transfer coefficient, resulting in a second peak in the hot channel heat flux. At this time the CEAs are nearly fully inserted, resulting in a rapid reduction in the core power level. The heat flux continues to decrease for the remainder of the transient.
Initial RCS pressure for calculation of the limiting fuel performance and radiological release event was 2200 psia. RCS oressure vs. time for this case is given on Figure 15.4.8-8. The long term RCS pressure response is shown on Figure 15.4.8-10. Initial RCS pressure for the limiting peak pressure case is 2400 psia. RCS pressure vs. time for this case is given on Figure 15.4.8-9 i
Steam generator pressures, and steam generator safety valve flow rate following a FPBOC CEA ejection with a postulated loss of offsite power following turbine trip are shown in Figures 15.4.8-11 through 15.4.8-13.
i
! The transient behavior of the NSSS following a postulated CEA Ejection is as follows. The steam generator pressure increases rapidly due to the closure of
( s the turbine control valve following reactor and turbine trip. The steam bypass control system is inoperable on loss of offsite power and therefore is I unavailable. The steam generator pressure reaching a maximum of 1348 psia at seconds. The pressurizer pressure increases to a maximum of 2525 psia at l 3.9 seconds due to the decreased heat removal of the steam generators.
Subsequently, the reduced reactor power following the reactor trip, in addition to the postulated break in the primary system, cause the RCS pressure and temperature to decrease.
l
! - The steam generator pressure decreases slowly until the main steam safety valves close. Tne total released through the safety valves is approximately 136,800 lbm.
l Following a postulated CEA Ejection Event, 9.8% of the fuel is calculated to experience DNB. Regulatory Guide 1.77 recommends that the onset of DNS be used as the basis for predicting clad tailure. C-E does not cauate oaset of DNB with cladding failure. Nevertheless, this criterion was used to determine the percentage of pins that suffer clad failure.
15.4-20
TABLE 15.5.2-1 SEQUENCE OF EVENTS FOR THE PLCS MALFUNCTION WlIM A LD55 0F Utt51It POWER AT IURBINE TRIP Time Setpoint Success Event or Value Path ,, o (Sec) .
O Charging Flow Maximized --
gget,Ac E W
& Letdown Flow Minimized g High Pressurizer Pressure 2450 Reactivity 250.7 Trip and Loss of A.C. Control at the Time of Turbine Trip, psia Pressurizer Safety 2525 Primary 1252.7 Valves open, psia System Integrity 1253.2 Maximum Pressurizer 2561 Pressure, psia Pressurizer Safety 2525 Secondary 1262.3 System Valves Close, psia Integrity Main Steam Safety 1282 Secondary 1265.5 System Valves Open, psia Integrity 1270.3 Maximum Steam Generator 1298 l Pressure, psia Operator Initiates -- Reactor Heat 1800.0 Plant Cooldown Removal k
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charging pumps is decreastd significantly. Tner:fere, the most negativa valu2 of MTC was selected to maximize the positivo reactivity addition from injecticn of cold makeup water.
Total charging flow due to all three pumps is 132 GPM. Considering 16 GPM for the control bleed takeoff and 30 GPM for the minimum letdown flow, net flow increase to the RCS is 86 GPM. The Pressurizer Pressure Control System (PPCS) is assumed to be in the manual mode w!th the proport4 pal sprays off preventing 4
the PPCS from suppressing the resulting pressure traauent.
C. Results The dynamic behavior of NSSS parameters following PLCS ralfunction with loss et offsite power at turbine trip is presented in Figures 15.5.2-2 to 15.5.2-11.
Failure of the Pressurizer Level Control System (PLCS) causes an increase in reactor coolant sysem inventory initiated by the startup of the third charging pump coupled with the decrease in letdown flow to its minimum. With the PPCS in the manual mode and. the proportional sprays turned off, increase in RCS inventory results in a pressurizer ptessure ; increase to the "" - *;;.co trip Ands's
'M^ ' .nrD c'" Cec;ntr setpoint of 2450 psia %t ndl 'ii5.IseuWs. TL ; trip h.umus er4~ a4 ari.tr5eds.
Since the steam bypass control system is in the manual mode and the rate of closure of the turbine stop valves is faster than the rate of control rod insertion, pressurizer pressure increases to 2561 psia which opens the primary safety valves. Decreasing core heat flux and the opening of the primary safety valves causes the pressure to drop; however, the decrease in primary to secondary heat transfer due to four pump loss of flow causes pressurizer i pressure to again increase, reaching a peak value of 2480 psia.
- The unavailability of the steam bypass valves causes the steam generator l
pressure to increase, causing the main steam safety valves to open at 1265.5 i
seconds. The decreasing core power and the safety valves function to ifmit the
! steam generator pressure to 1298 psia.
The 796.5 lbs of steam discharged by the pressurizer safety valve is contained I in the quench tank with no releases to the atmosphere. The main steam safety valves discharge 22,714 lbs of steam to the atmosphere prior to 1800 seconds.
s At 1800 seconds, the operator stabilizes the plant and initiates plant cooldown, using steam dump valves.
15.5.2.4 Conicusion The peak pressurizer pressure reached during the Pressurizer Level Control System malfunction with a loss of offsite power at turbine trip is 2561 psia and is less than 110% of the design pressure. Since this transient causes an increase in RCS pressure due to an increase in primary coolant inventory the DNBR increases. Therefore, the acceptance criterion regarding fuel performance is met.
15.5-6 .
r
For a double-ended rupture, the primary to secondary leak rate exceeds the capacity of the charging pumps. .As a result, the pressurizer
'- pressure gradually decreases from an initial value of 2400 psia. The
- primary to secondary leak rate and drop in pressurizer water level
~
- causes-the third CVCS charging pump to turn on. Even with all three CVCS charging pumps on line the pressurizer pressure and level continue to drop. Thisresultsinthepressurizerheatersbeinade-energized /EySj; at 560 seconds. At .LMe'Teconds a reactor trip signal is generatec
~
due to exceeding the CPC low pressure boundary cim'ippet-path. Th s p.O pressurizer empties at approximately 1151 seconds. At onds a
- safety injection actuation signal is generated, and ( '_^^' x x .i the safety injection flow is initiated. After the pressurizer empties, the reactor vessel upper head begins to behave like a pressurizer, and controls the reactor coolant system pressure until the pressurizer begins to' refill at approximately 1447 seconds. Due to flashing r caused by the depressurization, and the boiloff due to metal structure to coolant heat transfer, small amounts of voids form in the reactor vessel upper head at about 1151 seconds. Consequently, the RCS pressure begins to decay at a lower rate at this time. However,' under the i combined action of safety injection and charging flows, and reduced
- primary to secondary leakage, the upper head voids completely collapse
- at about 1447 seconds. Prior to this time, the RCS pressure begins to slowly increase helping to collapse the reactor vessel upper head i voids. The pressurizer water level is reestablished at about the same
- time due to the net mass influx which increase the RCS inventory.
Following reactor trip and with turbine bypass assumed to be unavailable (i.e., in the manual mode), the main' steam system pressure increases until the main steam safety valves open at 1209 seconds to control the
- s. main steam system pressure. A maximum main steam system pressure of 1283 psia occurs at 0.1 seconds after the MSSVs open. Subsequent to *
! this peak in the pressure, the main steam system pressure decreases, resulting in the closure of the main steam safety valves at 1316 seconds.
Prior to reactor trip, the feedwater control system is assumed to be in the automatic mode and supplies feedwater to the steam generators-
' such that steam generator water levels are maintained. Following reactor trip, the feedwater flow decreases to approximately 5% of the full power flow rate. Since the steam flow out of the steam generators is less than this feedwater flow, the liquid inventory in the steam generators gradually increases. At 1690 seconds a HLO mode terminates feedwater flow to the damaged steam generator. At 1778 seconds a HLO mode terminates feedwater flow to the intact steam generator. I After 1800 seconds, the operator identifies and isolate the affected steam generator by closing the main steam isolation valves and by l
, securing the reactor coolant pumps in the affected loop. The operator -
i then initiates an orderly cooldown via the steam bypass system and the condenser, and with manually-controlled feedwater flow to the unaffected stear. generator. After the pressure and temperature of the reactor coolant are reduced to 400 psia and 350*F respectively, the operator l
l activates the shutdown cooling system and isolates the unaffected l .- steam generator.
1 ~
l 15.6-11 4
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TABLE 15.6.3-1 (Sheet 1 of 2)
SEQUENCE OF EVENTS FOR THE STEAM GENERATOR TU8E RUPTURE Time Setpoint success Event or Value Path (Sec)_
0.0 Tube Rupture Occurs Third Charging Pump Started, feet -0.75 Primary 30.0 below program level System Integrity
-0.75 Primary 30.0 Letdown Control Valve ~5rottled System Back to Minimum Flow, feet below program level Backup Heaters Energized, psia 2360 Primary 53.8 System Integrity Pressurizer Heaters Oe-energized 400 560.0 due to Low Pressurizer Liquid 3
Volume, ft 1148. 3 CPC Low Pressure Boundary Trip m Reactivity Signal & G6 Control Feedwater Flow Starts Ramp Down to
/M6EF T ,A, 5% of Initial Full power Flow g 7 ces. e-_i_ -
7 _
- Turbine Trip: Stop Valves Start --
Control g 99 -- Secondary to Close System Integrity 1151 Pressurizer Empties Turbine Stop Valves Closed -- Secondary 1152 System fpggy- m Integrity 1
, g. , -
Reactivity Safety I njection Actuation Si~gnalg @
1181,$ g G Q Control and a
lW ,, _
,,b -- Primary 1181 8 Letdown Isolation valves Closed on System SIAS Integrity i
TABLE 15.6.3-1 (Cont'd.) (Sheet 2 of 2)
SEQUENCE OF EVENTS FOR THE - ---
STEAM GENERATOR TUBE RUPTURE Setpoint Success Time or Value Path Event (Sec) 1282 . Secondary 1209 Main Steam Safety Valves Open, psia System Integrity Maximum Steam Generator Pressure, 1283 1210 psia
d --
l
!::' _"' c ':r. ',t-- : ti:te 1218 Secondary 1316 Main Steam Safety Valves Close, psia System Integrity 1447 Pressurizer begins to refill 80 Secondary 1690 HLO Mode Terminates Feedwater Flow System to Damaged Steam Generator, % wide Integrity range 80 Secondary 1778 HLO Mode Terminates Feedwater Flow System to Intact Steam Generator, % wide Integrity range
-- Reactor Heat 1800 Operator Isolates the Damaged Steam Removal Generator and Initiates Plant Cooldown at 100 F/hr for the 1.5 hour5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> time period s
Shutdown Cooling Entry Canditions 400/350 Reactor Heat 28,800 Removal are Assumed to be reached, RCS Pressure, psia /RCS Temperature, F
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~15.6.3.2.3 Analysis of Effects and Consequences 15.6.3.2.3.1 Core and System Performance A. Mathematical Model The mathematical used for evaluation of core and system performance is identical to that described in Section 15.6.3.1.3.1.
B. Input Parameters and Initial Conditions The input parameters and initial conditions used for the evaluation of core and systems performance are similar to those described in Section 15.6.3.1.3 and are given in Table 15.6.3-9. Both the initial core
! mass flow rates and the one pin radial peaking factor were chosen to:
(1) maximize the primary-to-secondary integrated leak, and the steam releases through the main steam safety valves, and (2) at the same time, obtain a simultaneous reactor trip on a low DNBR (=1.19) as well as a low pressurizer pressure. Consequently, a slightly lower core mass flow rate (104% instead of 116%) as well as a slightly lower radial peaking factor (1.53 instead of 1.55) were employed in the analysis.
l C. Results The dynamic behavior of important.NSSS parameters following a steam generator tube rupture with a loss of normal ac power are presented in Figures 15.6.3-19 through 15.6.3-34.
M g.cbiPM steamPrior to reactor trip, the dynamic behavior of the NSSS following a generator tube rupture with a loss of offsite power is similar tp#M, gI to that following a steam generator tube rupture without a loss of /g6 7 At offsite power which is described in Section 15.6.3.1.3.
N " #c.ceA seconds after the initiation of the. tube rupture *the CPC low pressure boundary; ' 5 p Hie . ", -- '.i; ' - - :t:: '-ip -% A t
' Subsequent to the reactor trip, the RCS pressure begins to decrease rapidly, and the pressurizer empties at about 1201 seconds due to the continued primary-to-secondary leak. After the pressurizer empties, l
- the reactor vessel upper head begins to behave like a pressurizer and controls the RCS pressure response. Due to the loss of offsite power, the reactor coolant pumps begin to coast down reducing the core coolant l
flow rate, and the mass flow into the upper head region. This region l becomes thermalhydraulically decoupled from the rest of the RCS, and due to flashing caused by the depressurization and boiloff from the metal structure to coolant heat transfer, voids form in this region at about 1196 seconds. The void formation is enhanced by the decoupling effect, since the RCS pressure reduction due to primary system cooling is felt in this region, while the RCS temperature reduction is not.
The significant impact of voids in the upper head region is a slower '
R'CS pressure decay resulting ir..the ceneration of the safety iniectio ' '
actuation signal (SIAS) at 1H'Fseconosag T P "igb Dressu m SafMy In F +4an (uo m numos hogin delivgr_y gf_j u ftty 4at t4 a fl ui d +'r-the (M 14 a Ln itM** fA4 Sapfg yeGm 15.6-17 c
R" ' -
-^ '
-% th. P agil$r a result, the upper l head voids begin to collapse at about 1677 seconds.
i Following turbine trip and loss of offsite power, the main steam system pressure increases until the main steam safety valves open at about 1197 seconds to control the main steam system pressure. A maximum main steam system pressure of 1310 psia occurs at about 1205 seconds. Subsequent to this peak in pressure, the main steam system pressure decreases resulting in the closure of the safety valves at 1721 seconds.
Prior to turbine trip, the feedwater control system is in the automatic mode, and supplies feedwater to the steam generators to match the steam flow through the turbine. Following turbine trip and loss of offsite power, the feedwater flow ramps down to zero. Consequently the steam generator water levels decrease due to the steam flow out f Q through the main steam safety valves, and a low steam generator level u signal is generated at about".Dt3 seconds. Subsequently, at abedt }
g59-(n hiite" seconds, emergency feedwater flow is initiated, and the steam generator water levels begin to recover.
After 1800 seconds, the operator identifies and isolates the affected steam generator by closing the main steam isolation valves. The operator than initiates an orderly cooldown by means of the atmospheric l dump valves and emergency feedwater flow to the unaffected steam generator. After the pressure and temperature are reduced to 400 psia and 350 F, respectively, the operator activates the shutdown cooling system and isolates the unaffected steam generator.
l The reduction in the RCS pressure due to the loss of primary coolant I through the ruptured steam generator tube results in a reduction in
.the thermal margin to DNB (see Figure 15.6.3-34). The transient minimum DNBR of 1.19 occurs at the time of reactor trip. The DNBR shows an increasing trend after reactor trip due to the rapidly decreasing heat flux. The RCPs do not begin their normal coastdown until after the loss of offsite power three seconds after turbine trip. However,
' there is a slight decrease in the core flow during the three seconds immediately after turbine trip and prior to the loss of offsite power due to decreasing pump speed caused by frequency degradation (approxi-mately 1 Hertz /second) of the electrical grid. The resultant calculation demonstrates that no violation of the fuel thermal limits occurs, since the minimum DNBR stays above the value of 1.19 throughout the transient.
The maximum RCS and secondary pressures do not exceed 110% of design pressure following a steam generator tube rupture event with a concurrent loss of offsite power, thus, assuring the integrity of the RCS and the main steam system.
Figure 15.6.3-29 gives the main steam safety valve integrated flow rates versus time for the steam generator tube rupture event with a loss of offsite power. At 1800 seconds, when operator action is assumed, no more than 54,936 lbm of steam from the damaged steam l
generator and 54,730 lbm from the intact steam generator are discharged I !
15.6-18
- I
t TABLE 15.6.3-6 (Sheet 1 of 2)
SEQUENCE OF EVENTS FOR A STEAM GENERATOR TUBE RUPTURE WITH A LOSS OF OFF5ITE POWER Setpoint Success Time Path Event or Value (Sec) 0.0 Tube Rupture Occurs Third Charging Pump Started, -0.75 Primary System 30.0 Integrity feet below program level Primary System 30.0 Letdown Control Valve Throttled Integrity Back to Minimum Flow, feet below program level -0.75 Backup Heaters Energized,' psia 2360 Primary System 53.8 Integrity Pressurizer Heaters De-energized 400 560.0 due to Low Pressurizer Liquid Volume, ft 3 gyggnr 11%M5 CPC Low Pressure Boundary Trip M Reactivity (g W o
Signal C - G erterm. .eA. T Control Turbine / Generator TripiB aurep -- Secondary System 1188 Integrity
%sammuttmagasmictsee m-L_ni t :_..;, #
' .':... ; , " ' :: Ci;d _
2- - -
- y :,:t:-
I ii:1 -- _
Loss of Offsite Power
-- Integrity
);97 LH Main Steam Safety Valves 1282 Secondary System 1197 open, psia Integrity RH Main Steam Safety Valves 1282 Secondary Sytem 1197 Integrity open, psia 1201 Pressurizer Empties Maximum Steam Generator Pressures 1310 1205 Both Steam Generator, psia i
5 Reactivity Control
! Safety Injection Actuation Signal, .LiiM8 (1563.2. O Crexe. rat. ed-1UGEltT
'b
' /5 (,3 2. S. % b J ec k N I d ' A -- -
L ,
TABLE 15.6.3-6 (Sheet 2 of 2)
SEQUENCE OF EVENTS FOR THE
^
STEAM GENERATOR TUBE RUPTURE WITH A LO55 0F OFF5ITE POWER Setpoint Success Time or Value Path Event (SecJ
-- Primary System IMyr Letdown Isolation Valves Closed Integrity
/%s2 on SIAS
~5aien Bj;; tic c'c" I-f ti;ted
-~
neacd. :t/ caatent C "eacter ::;;t 9r: cal 1714 to Emergency Feedwater Actuation - M Secondary System Integrity f
1218 Secondary System Main Steam Safety Values Closed, 1721 Integrity psia
-- Secondary System 1759,(,, Emergency Feedwata Flow Begins Integrity
-- Reactor Heat 1600 Operator Isolates the Damaged Removal Steam Generator and Initiates
! Plant Cooldown Shutdown Cooling Entry Conditions 400/350 Reactor Heat 28,800 Removal are Assumed to be Reached, RCS Pressure, psia / Temperature, *F l
l s
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1562 2 ~Pressa.s ur- Peme ys, ,, re..p;,;
C % t r. ,(
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4 159.4 RESULTS An example limiting analysis of the LFI transient suggested by Reference 1 '
was performed applying,the conservative methods with the most adverse set of initial plant conditions and transient parameters discussed above.
Table 158-1 lists the assumptions utilized in this worst case. The sequence of events and the dynamic response of the important NSSS parameters are provided in Table 158-2 and Figures 158-13 through 158-30. respectively.
f A 0.2 ft 2 crack in the main feedwater line is assumed to instantaneously terminate feedwater ficw to both steam generators and establish critical flow (~2000 lbm/see of saturated liquic) from the generator nearest the break. The absence of subcooled feedwater flow causes a constant heatup _ y* g and pressurization of the steam generators during the first @ seconas ,
which reduces the primary-to-secondary heat transfer rate. Rising reactor coolant temperatures and pressure result. Due to temperature reactivity feedback during this period the core power decreases slightly from 102 percent to 98 percent of design full power.
%3.%2 l
' loseAt@ll a heat transfer capability due to total depletion of its liquid in-se ventory by boil-off and the break discharge flow.
This initiates a rapid heatup and pressurization of the reactor coolant system and depressurization p.p of the steam generators. Once emptied, credit is taken for a low water level trip condit*on in the ruptured steam generator which leads to a
{
reactDr trip signal at W4. seconds simultaneous with a high pres'surizer pressure trip signal. The rate of reactor coolant system pressurization is l Closure of the turbine . leaves the pipe
' Aurther aggravated atT3J seconds.
C -
break as the only steam relief path, thereby reducW the energy flow from Y )T the intact steam generator below that of the primary-to-secondary heat
' ' transfer rate. The resulting steam generator pressurization reduces the In addition, the loss of primary-to-secondary temperature difference.
reactor coolant flow following the loss of electrical power decreases the heat transfer coefficient of the coolant in the steam generator tubes.
- A significant heat transf_er reduction occurs.
Compression of the pressurizer steam volume due to the high insurge flow raises the pressure to the safety valve setpoint at 34.6 seconds. Thereafter l
every increase in the surge flow causes a slight pressurization which cpens the safety valves such that their volumetr.ic discharge rate matches that of the insurge. The reactor coolant system pressure continues to increase to a maximum of 2843 psia at 38.2 seconds. At that time the increased pressure establishes a surge line prassure gradient which provides sufficient flow to allow the re, actor coolant to expand under the existing heatup with no further pressurization. Pressurizer pressure and sarge line flow are also at their maxic.a of 2537 psis and 2205 lba/sec, respectively.
The rate of heatup decreases s.bteaue.t to core heat flux decay causi w "the primary pressares to arca. I a.5 ncones the ;.' i stesm safet.y val.es open tnus stabilizing tne samry side te.nparature and al!cwing tne 158-7
f .
rising primary coolant temperature to develop greater heat trans'fer to the intact steam generator. The intact generator is forced to a maximum of _
1318 psia before the heat transfer begins to decrease. However, the care-to-steam generator heat rate mismatch is reduced sufficiently by 45.4 seconds to allow closure of the pressurizer sa'fety valves and by 45.8 !
seconds the reactor coolant system enters a cooldown Under the influence -
of. steam blowdown through the ruptured steam generator to the break, the cooldown proceeds even after the steam generator safety . valves close at 73.8 seconds. lido O !
A' main steam isolation- signal is generated at 65.6 seconds on low steam -
generator pressure which closes the main steam so ation valves decoupling the intact steam generator from the ruptured steam generator and the break..
The intact' steam generator repressurizes, thereby reducing its heat transfer and eventually causing a primary system heatup by 300 seconds. With the main steam safety valves open by 314.2 seconds, the primary-to-secondary heat inbalance is eliminated by approximately 600 seconds. Thereafter the NSSS enters into a quasi-steady state with a very gradual cooldown and -
depressurization due to decreasing core decay heat and.with emergency
.feedwater flow which was initiated atO89.6 seconds caintaining an adequate l
' liquid inventory within the intact steam 7 generator for heat removal. By
'1800 seconds the operator initiates a controlled /cooldown to shutdown
-cooling utilizing the atmospheric dump valves. L- 90. 0 The minimum DNBR vs. Time as shown on Figure 15B-30 remains above 1.19 throughout' the transient. .
During the first 30 minutes following the initiation of this LFI event mass
releases from the = system amount to 2970 lba of steam from the pressurizer safety valves to the reactor drain tank, 79,700 lba of steam from the main ',
steam safety valves to atmosphere, and 69,200 lbm of liquid and 34,200 lbm Steam release to of steam from the feedwater line break to containment.
the reactor drain tank may burst the tank's rupture disc discharging its ,
l l
' contents to containment.
i During tnis event, two sources of radioactivity contribute to the sit'e boundary f dose, the initial activity in the steam generator inventory, and the activity associated with primary to secondary leakage from the steam generator tubes which are assumed to be at the technical scecification , limits of 0.1 uCi/gm and 4.6 pct /gm dose equivalent I-131 respectively. During the first two hours of th's evant.,
the total activity from the steam generators includes 8.9 Ci fron.the affecteo' steam generator to the containment building including 1.6 Ci associated with technical specification tube leakage (1 gpm) and 0.33 Ci total activity released Assumingfrom all the
-unaffected steam genefator to the containment and atmosphere.
radioactivity is released to the atmosphere, the offsite cose due to feedwater line break with loss of offsite power results in no more than 9.5 rem"two' hour -
inhalation thyroid dose at exclusion area boundary.
158-8
6 4
TA8LE 158-2
% (Sheet 1 of 2)
- SEQUENCE OF EVENTS FOR THE LIMITING CASE LOSS OF FEEDWATER INVENTORY EVENT Satpoint l- Time or Value Event
.c,( Sect z<
Break in the Main Feedwater Line , F 0.2 M 0.0 0.0 Instantaneous Loss of All Feedwater
- - Flow to Both Steam Generators .
0.0 Instantaneous Development of Critical Flow from the Ruptured Steam Generator to the Break
) 33.8 2, IN66R.TInstantaneous "A" Loss of All Heat Transfer Q the Ruptured Steam Generator W
[.Jeir 554.M
'Luw Water Level Trip Signal deemeheA G eM.e.<ed.
7.r... ,,2 ::;_:- : ;. :::r M 34,61 - Emergency Feedwater Actuation C:::r:r-Signal Geu.umT.ed !!mpup f c;.: th. .".;;;r:1 "r-
, h l
34e* W M ' High Pressurizer Pressure Trip Si alg Pressurizer Safety Valves Open, ps. e.xe. rat. ird. _2525 W ly y l 34.6.
& 34.97 Trip Breakers Open m , ___
f
'35.8 Instantaneous Closure of the Turbine f '
Stop Valves l ,
l 35.8 Loss of Nomal On-Site and Off-Site
/A>(citi- Electrical Power
~
- ' aide LE "
- t:- L:::: Tci; ::,..J T,um 1
~B' f ' + 3EF m : n;;; ::::- w -rm -r c 3 ,,,nru-
- :::r 9, v 2843-pwe
- 'O. .18.2 Maximum Reactor Coolant Pressure , T*
Maximum Pressurizer Pressure, psa 2587 mais.
Maximum Pressurizer Surge Line Flow, Ib ,/sec 2206 2hmi$sse
.Y Y
TABLE 158-2 (Cont'd.) (Sheet 2 of 2)
SEQUENCE OF EVENTS FOR THE LIMITING CASE LOSS 0F FEEDWATER INVENTORY EVENT Setpoint Time or Value Event (Sec)_
Main Steam Safety Valves Open, 9 64 12824ede 40.5 luSGRX Emergency Feedwater Actuation Signal Gamardt24. N
- ' c " M ' :.- :... :..... ::...
' C. . .. . . :,; r M
t/S.o -
1318.ps4a Maximum Steam Generator Pressure , psia 44.8 45.4 Pressurizer Safety Valves Close ,fd
- 2525 vs*e Minimum Pressurizer Steam Volume , M 138 M 45.8 1218 pe+s 73.8 Main Steam Safety Valves Close , pria TN. :n , u e ced E',=
74 4- E.:rg ,., raj . m.
., . w . gy. . a o . 2,, c . ,, . . w 875 y agg $0 Emergency Feedwater Flow Initiated to the Intact Steam Generator,9p+
fpqgig.T :- +tensene.
'c. * :::
- _ :'; :' ' :r ?e D 123::6 IL..... ...:~.._.- ^^
69:6 /lo6,0 Main Steam Isolation Signal Ge.w.et a Ord.
W
- 2 3100 -t::m 170.6 Minimum Intact 5 team Generator Liquid Mass , (b.=
' -- . 1 - . -. .. . __ a . . . g , _
- n-___.
. ..r Main Steam Safety Valves Open , psi.
1292 W 314.2 1800.0 Coerator Ocens the Atmospheric Steam Dump Valves to Segin Plant Cooldown to Shutdewn Ccoling k no.6 Ra stam .lselecem V.Jue_s closd
- - - - - ,,--+% -,m y,-_-- ,y,.- , - _,-.-,,y-.,,,w--www=,.w+,w ,,w ,w- , , , , . , . ..,--w-, .em mw-,.,wrwy,wy-www
l I
i /59 -2. Q) l A
srw Ge>.ewbr Luc =c Leu I p-%
v3. a 2.
y_,,A T r ;;e Aw%+is SQ;Z lx Ila Aptg y_4tia-ea 6+_we.rator 3,9 7., Steam 6exextac- %Dter- Le.ve.{ Mg k E x q m ap S d L~Ga AcTu t.i,n 4;g u(
Aqs:s seq 6:xi%.ter.4. 6 de s.ez Fre+5m'w I're.55#e b 'l'*5 b^ '
- 24'15 T~r '. > h% sis CN"*' *) l" *'
y V Wu N'4 .
pr. i,.;) AQis Gefi~~C I' b g gn G *cr=' tor; l'*r* C
- l u.ria t'a-4E*_
l
- s. oo l
I r
(
4 '/, o srem 6 m e_ra te r- uxter Le.o e.( R.- J s ja Emerg.awg Fanc9.watter A# A ;aw s:N d 5;S C.e.t =e; C L w t Lc T & it- Ge.w<.rator, yerew A A r.-ge.
i, i.
D 16 5'. o * -- 6 =.ar=1zr N.ee P~ l .s e 81o R 4 SCea
- Isolah 5.pl A4 sis serp.ta ,7srm
A 0.20 ft 2rupture in the main feedwater line is assumed to instantaneously terminate feedwater flow to both steam generators, and establish critical flow from the generator nearest tne break at an initial rate of 1979 lbm/sec. This causes a decrease in steam generator liquid mass as shown by Figure 158-39.
The break discharge enthalpy is assumed to remain that of saturated liquid until the ruptured steam generator empties, at which time saturated vapor enthalpy is assumed. g The absence of subcooled feedwater flow causes a constant heatup and pressurization of the steam generators during the first ,
seconds l
'which reduces the primary-to-secondary heat transfer rate. Rising primary coolant temperatures and pressures result. Due to the temperature reactivity feedback during this period core power is reduced from an initial value of 102% to 99.8% at Q l 24 .9 8 2 seconds.
S.9 8 At M econ;,ds the ruptured steam generator produces a low water l level reactor trip signal. This reactor trip signal is coincident r
1 with a high pressurizer pressure trip signal. LE:: M ttt tR,) l heat transfer in the ruptured steam generator begins to de race due AC 2QS Sem to insufficient inventory. This degradation initiates a id heat c7,13 up and pressurization of the reactor coolant system. At 7 seconds the reactor trio breakers open followed by an assumed C instantaneous turbine tri5 Immediately following turbine trip, the [
.u_ 23M 7 sac.ds failure to fast transfer to offsite power occurs, resultins in the coastdown of two reactor coolant pumps. These occurrences further q.
aggravate the primary pressurization.
Closure of the turbine leaves the pipe break as the only steam i relief path, thereby reducing the energy flow from the intact steam generator below that of the primary-to-secondary heat transfer rate.
l The resulting steam generator pressurization reduces the the loss In addition, primary-to-secondary temperature difference.
( of reactor coolant flow following the loss of electrical power to s
two pumps decreases the heat transfer coefficient of the coolant in the steam generator tubes. A significant heat transfer reduction l
occurs.
Compression of the pressurizer steam volume due to the high insurge flow raises the pressure to the safety valve setpoint at 28.3 seconds. Thereafter, every increase in the surge flow causes a slight pressurization which opens the safety valves such that their volumetric discharge rate matches that of the insurge. At 30.2 seconds, the surge line flow reaches its maximum value of 1458 lbm/sec.
At this point in time, the reactor coolant system pressure is at a
(
' maximum of 2712 psia. Also, the increased pressure establishes a surge line pressure gradient which provides sufficient flow to allow i
158-13 m
TAPLE 158-4 SEQUENCE OF EVENTS FOR THE REANALYSIS OF THE LIMI -
LOSS OF FEEDWATER INVENTORY EVENT Setpoint Time Event-or Value g ~
2 0.20 Rupture in the Main Feedwater Line, ft 0.0 Complete Loss ~ of Feedwater to Both Steam ----
0.0 Generators
,1979
< 0.0 Initial Steam Generator Break Flow Ibm /sec teigh Pressurizer Pressureg Trip Gandstaan A>tl r's Mre' 7.5.98 /TeuA4, Reiter T N @ 4 g5I* 2475 N
Mrfr gg,9g igh Pressurizer Pressure Trip Signal Generated
^ ----
Low Level Trip Signal % Gam,. ____.__ CG 34,4 7.6.18 A uaer-t te.J. _
p Heat Transfer Degradation in Ruptured SG 8egins 25.98 P:t 27,15 Reactor Trip Breakers open P:S ~21.97 Turbine Trip on Reactor Trip
( E 27 D Failure to Fast Transfer - Two Reactor Coolant ----
l Pumps Coast Ocwn I -
, 2525 28.3 Pressurizar Safety Valves, psia l 1282
' Main Steam Safety Valves Open 30.0 1458
- Maximum Surge Line Ficw, Ibm /sec 30.2 2712
' Maximum RCS Pressure, psia 30.2 1342 Maximum Steam Generator Pressure, psia 33.8 36.8 Ruptured SG Ories Out l 2523 37.4 Primary Safety Valves Close, psia l
3FM us.io ece Gme.eme wen.c Leve.l l
Twd.e4 tr ~ m e. T H p A 4 sis SetpoW iw tA.a bared. G urotor, S,,, IjdA rwa l
n
core coolant flow, and maximum core coolant inlet temperature. This combination of initial conditions results in an early generation of a reactor. trip signal due to exceeding the CPC hot leg saturation tempera-ture range limit. .
C. Results .
The dynamic behavior of important NSSS parameters following a steam generator tube rupture is presented. in Figures 150-1 to 150-15.
For a double-ended rupture, the primary to secondary leak rate exceeds the capacity of the charging pumps. As a result, the pressurizer pressure gradually decreases from an initial value of 2100 psia. The primary to secondary leak rate and drop in pressurizer water level causes the second and third CVCS charging pumps to turn on. Even with all three CVCS charging pumps on line the pressurizer pressure and level continue to
- drop. At 47 seconds a reactor trip signal is generated due to exceeding the CPC hot leg saturation temperature range limit. The pressurizer empties at approximately 546 seconds (Figure 150-5). At 570 seconds a y safety injection actuation signal is generated, and-tpISSEUWWuns the [
safety injection flow is initiated. After the pressurizer empties, the reactor vessel upper head begins to behave like a pressurizer, and controls the reactor coolant system pressure until the pressurizer begins to refill at approximately 4020 seconds. Due to flashing caused by the depressurization, and the boil off due to the metal structure to coolant t heat transfer, the reactor vessel upper head begins to void at about 77 seconds (Figure 150-6). Consequently, the RCS pressure (Figure 150-2)
' begins to decrease at a lower rate at this time. .
Following reactor trip and with turbine bypass unavailable, the main i
steam system pressure increases until the MSSVs open at 52 seconds to l
control the main steam system pressure. A maximum main steam system
! pressure of 1330 psia occurs at 56 seconds. Subsequent to this peak in l the pressure, the main steam system pressure decreases, resulting in the closure of the main steam safety valves at 95 seconds. The MSSVs cycle twice more in this manner until the operator takes control of the plant.
Prior to reactor trip, the main feedwater control system is assumed to be in the automatic mode and supplies feedwater to the steam generators such that steam generator water levels are maintained. Following reactor trip, the main feedwater flow is terminated due to the loss of offsite power. As the level in the steam generators decrease an Emergency Feed-water Actuation Signal (EFAS) is generated resulting in, auxiliary feedwater flow which acts to restore the SG 1evel.
At 460 seconds the operator takes control of the plant and opens one ADV -
on each SG to cool down the plant. This is consistent with the EPGs. At i 2100 seconds the RCS has been cooled to 550*F. The operator isolates the auxiliary feedwater to the affected generator,. closes the main steam isolation valves of both steam generators, and attempts to close the ADV of the affected generator. The operator recognizes that the ADV did not close and has the appropriate block valve closed within 20 minutes. The l_ operator then initiates an orderly cooldown by means of the atmospheric j
l 150-6 I
TA8LE 150-1 SEQUENCE OF EVENTS FOR A STEAM GENERATOR TUBE RUPTURE W.,TN A LQ55 0F OFF5 lit PCWER AND STUCK OPEN ADV, Setpoint Time or Value Success Path (Sec) Event
! 0.0 Tube Rupture Occurs
-0.75 Primary System Integrity .
40 Third Charging Pump Started, feet below program levet l
-0.75 Primary Systen Integrity 40 Letdown Control Valve Throttled f Back to Minimum Flow, feet below j
j program level Reactivity Control M4 CPC Hot Leg Saturation Trip Signal ---
Secondary System Integrity (48
' Turbine / Generator Trip %
\
M --- Reactivity Control cza*4mpnemmiiGman l ' : " ;' '; - ' ._ W E Pr5: n:; '!9:x "':::f ---
5/ Loss of Offsite Power 1265 secondary System Integrity 52 LH Main Steam Safety Valves open.
psia 1265 Secondary System Integrity SZ RH Main Steam Safety Valves open, s psia 1330 56 Maximum Steam Generator Pressures Both Steam Generator, psia 1218 Secondary System Integrity 95 Main Steam Safety Valves Closed, -
-m psia' Secondary Syst rity l 19.76
! I16 Auxiliary Feedwater Actuation on Generator Level Trip Signal, n 3eamGenerator, feet above tube -
I ation on is. -
Secondary Systen Integrity i
Auxiliary Feedwat 177 l
! Low St r Level Trip
! S1 .
uptured Steam Generator, -
eet above tube sheet
( o /;
l
> W L
- --_- __ _ _.-_ _,- -. - - -. _ _ _.__... - ,,. ,__,__.,_.,__ m.__ ,,_ . _ -,,,- _w ,,, _ - .w - ,.-_- p_ --,.,--y , - .
TASLE 150-1 (Cont'd.)
s.
Setpoint Time or Value Success Path (Sec) Event Operator Initiates Plant Cooldown
-- Reactor Heat Removal 460 by Opening One ADV on each SE 546 Pressurizer Empties 1 570 Safety Injection Actuation Signalg W Reactivity Control .
. ? - Grwl ok Safety Injection Flow Initiated --- Reactivity Control 550 Secondary System Integrity 2100 Operator Attempts to Isolata the Damaged Generator, RCS Tem. , 'F Secondary Systen Integrity 3900 Operator Closes the ADV Block Valve Operator Initiatas Auxiliary Primary Systen Inventory 4020 Spray Flow Operator Centrols Auxiliary 20 Primary System Integrity 4500 Spray Flow, Backup Pressurizer .
- Heater Output, and HPSI Flow to Reduce RCS Pressure and Control Subcooling,'F
}
Shutdown Cooling Entry Conditions 400/350 Reactor Heat Removal 28,800 Reached. RCS Pressure, psia /
Temperature. OF 5 70 }res56;ut by , g.p pg A L e.s -QR gQ A<_i la / g (.srAs)
A 4 s,s -
ect.t> w 0
V
. _ . . . . . _ . . . _ . . . _ _ _ _ , . _ . . _ _ . _ . _ . . _ _ . . _ _ . _ _ . . _ _ . . _ _ _ _ . . . _ _ _ , . . . . _ _ . . _ , . _ _ _ . _ _ _ . _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _ . ~ . _ , , , . . . _ _ _ .
T /5D-l a ie A
q,, i s- Te:,y ke.~(e.rs QuA - 12~'tW%
Cemeo I ;
I 4 // '
I a 1.0 W C-6e W& 2S Lee 9. EeA Eneq% SmArn ggy r--o+m A&n Cg~L g- '
[6EAS)dk;5 % &
- i. w I L . d u N T A- C M era '
ger-e.n A e 9e_,
! I# *
- EFAG GeanTeD --
l31. 0 9Gn Gw% U%.G Q' DS Les2. % ewxs sym A f % C#/eO ~~.A. D~Hy U:4-Oya. %%
p e re.e e t 4 u ra. g e.
13 2,*o*
@ A 6 6 e+ e.tr M _ __
- &g &E c auMPnon W"a
_ c ..
g[Q
.o.rb E we g 1 "" g g - -
SycrR-
/NO b ' L _r. J L N a H 5.c/*4 g4 c; w b_a ,t. -
TABLE 150-4 ASSUMPTIONS AND INITIAL CONDITIONS FOR THE STEAM GENERATOR TUBE RUPTURE WITH A LOSS OF OFF5ITE POWER AND STUCK OPEN ADV Assumed Parameter Value
'3876 Core Power Level MWt Core Inlet Coolant Temperature, 'F 570 Reactor Coolant System Pressure, psia 2100 6 155 Core Mass Flow Rate,10 lbm/hr One Pin Integrated Radial Peaking Factor, with Uncertainty 1.53 Steam Generator Pressure, psia 1126 Moderator Temperature Coefficient,10'4 ap/*F -bb Doppler Coefficient Multiplier iyMP-je0 CEA Worth at Trip, % ao (most reactive CEA fully
-10.0 withdrawn)
/