ML20050B558
| ML20050B558 | |
| Person / Time | |
|---|---|
| Site: | Midland |
| Issue date: | 03/30/1982 |
| From: | Jackie Cook CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.) |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| References | |
| 16591, NUDOCS 8204060044 | |
| Download: ML20050B558 (28) | |
Text
7,
oct.
l CORSum8fS Power James W Cook Vice President - Projects, Engineering and Construction General offices: 1945 West Pernell Road, Jackson, MI 49201 * (517) 798-0453 March 30, 1982 Q
liarold R Denton, Director E
4 Office of Nuclear Reactor Regulation
\\
US Nuclear Regulatory Commission
(
TG:Ci9 FED il Washington, DC 20555 APR5 1982es 2 MIDLAND PROJECT D
.]jQ97 d MIDLAND DOCKET NO 50-329, 50-330
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SAFETY EVALUATION REPORT INFORMATION
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FILE: 0505.16 SERIAL:
16591 N
7"
~
ENCLOSURE B&W Ltr CPCO 3561 Dated 1/26/82 Partial Pump Coastdown nalysis The enclosure provides information requested by the NRC Staff to facilitate their review of the Midland FSAR with respect to generating the Safety Evaluation Report.
This information will be contained in Amendmcat 44 of the Midland FSAR which is scheduled for distribution by June 1, 1982. The locked rotor analysis will be updated in the interim and included in Amendement 44.
JWC/JRW/fms CC RJCook, Midland Resident Inspector, w/a RHernan, USNRC, w/a RWHuston, Washington, w/a DBMiller, Midland, w/a I(
oc0382-0828a131 C204060044 820330 PDR ADDCK 05000329 E
+;
L-scM 3oGF SLC-l t ?G~S Babcoek & Wileox noei. r go..,2.n.r.iion civi. ion a McC.rmott company 3315 Old Fore,t Road P.O. Box 126C Lynchburg. Virginia 24505 1260 (804) 385-20'4 January 26, 1982 CPCO: 3561 File:
128/TI.2 Consumers Power Company 1945 Parnall Road Jackscr., MI 40201 Attention:
Dr. T. J. Sullivan Manager, Safety & Licensing Dept.
Subject:
Consumers Power Company Midland Flant, Units 1 and 2 FSAR SECTION 15.3 - PARTIAL PbMP C0AST00WN ANALYSIS dear Dr. Sullivani Attached is B&W's proposed revision to FSAR section 15.3 This revision incorporates the additional flow related transients for partial pump operation cormitted to NRC via J. W. Cook's letter to H. R. Denton, S-13750 dated August 31, 1981.
This letter also contains a discussion and results of one pump coast-down calculations as requested by your J. Uebb.
'Per receipt of this memo, H. Leech is requested to update the LCP to show Item 457-340 ' complete; and Bechtel is requested to initiate a Stil to incorporate this change.
Very truly eurs,l s
W 4
F.
. Levandoski Associate Project Manager For: D. F. Judd Senior Project Manager Attachment p
cc BAUMAN DB MILLER y
LH CURTIS
.i VH SPANGLER FEBE1 m42 ID GREEN 3
RL BAKEn P S PROJ M
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BABC0CK & iflLC0X - NPGD ENGINEERING If! FORMATION RECORD DOCUMENT IDENTIFIER 51-1130767-00 TITLE Section 15.3 Text Revision PREPARED BY R. J. Schomaker DATE January 15, 1982 REVIEWED BY R.M. E\\lison bNIb DATE L...-
- 2. /, /9P2 8
REMARKS:
Section 15.3 of the Midland FSAR has been revised to include:
- 1) 2/2 - 2/1 Coastdown 2-) 2/1 - 1/1 Coastdown I
3)
"% Pins in DNB" for Locked Rotor From 1/1 Operation l
1 t
i NOTE:
- 1) Tables have been added; corrected and renumbered.
- 2) Figures 15.3-12,13,14, and 15 have been added.
Note:
No review of FSAR material has been performed other than the specific material identified added by this transmittal.
{
1 Para
}
Of A4
51-1130767-00 15.3
.OE.CR EAS.E IU' FE?MTOh~ CCCLAMT S YST E M F LOM P A_T..E.
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M.3.1 SINGLE AUD MJf.TIPL7 RiLCTOR COOLT.UT FUMP TFIPS (LO3S-OF-CCCLANT YLCF) 15.3.1.1 l_dentification of Caus s and Frecuency Clattificatien The resc:or ccclant flowrate is reduced if cne er more of the reactor ecolant pumps fail.
A pumping failure can occur from mechanical fcilures or from a less of electric power.
With four independen pumps available, a nochaniecl failure in ene pump will not aff ect operation of the others.
Each reactor coclant pump receives electric power from one of two.
electrically separate buses, and one pump in each loop is cennected to each bus.
Faults in an individual pump motor or its power supply could cause a reducticn in ficw, but a complete loss of fcreed ficw, the mere cencervative chne, is extremely unlikely and would cccur only if all of feite power or both unit auniliary transformers were icst.
The less cf all.offsite power would cauce gravity insertion of control. reds independent of reactor trip frcm the plant protection system.
Even though this is an event of infrequent cecurrence, the nuclear unit can sustain such a failure without core damage.
.c' '
15.3.1.2 Secuence of Events and systet.1s operation 2
\\,...'
IYhe reactor is protected frcm the ccnscquences of reactor coolant pump failures by the reactor prctection cystem.
The reacter ic (2Agwpu:
' ripped ff insufficient reactor coclant flow exists for the pcweer 1.ev el.
TM prp mc:.urc n.r r ::c t _
~
W'g pr ir lert.
The ICS initiates a power reduction upcn pur.ip
.......c..~_
..s 6Ry
,tailure to prevent reactor scuer frca exceeding that permicsibic IN
/ for the available flew.
Even though ICS acticn is available to b
I prevent or mitigate this accident, the accident analysis wac made wit.hout ICS acticn.
Loss of pcwer to the RC pump 3 hac been analyzed for four, thr?e, and two pump operation.
Table 15.3-1 gives the sequence of events for the fcur pump cass.
Fcr the
'hree and two pump cases, 'he MDNBR cccurs at approximately the l
(,same time as the four pump case.
l i
15.3.1.3
,Cere and Svstem perfermance 15.3.1.3.1 Saf ety Evaluatien Criteria the safety e'raluaticn criterior for M:S c c.c:1.arce f.. t! -: t h r.
iainin.te 'X:b.atic shall not 29 le:. c
- nn 1.3.
r,
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I t.evic ion 2
- 13. r s gm;
[
5]-1130767-00 Insert A 15.3.1.2 Seauence of Events and Systems Ooeration The reactor is protected from the consequences of reactor coolant pump r
f trips by the /eactor /rotection gystem. The reactor is tripped if insufficient reactor coolant flow exists for the power generated. Protection from the one pump coastdown event from four pump operation is provided by the flux / flow trip function. Protection from other pump coastdown situations are provided by the power / pump monitor trip. The Integrated Control System (ICS) initiates a power reduction upon pump failure to prevent reactor power from exceeding that pemissable for the available flow.
Even though ICS action is available to mitigate this accident, the accident analysis was performed without ICS action.
Analyses has been perfomed for the loss of power to all the RC pumps from initial four, three, and two pump operation. An analysis has also been performed of the loss of one RC pump from four and three pump operation. The sequence of events for the four, three, and two pump cases are shown in Table 15.3-1.
The sequence of events for the loss of one pump from four and three pump operation is shown in Table 15.3-4
51-1130767-00 arcr.anD,ca-rsaR
' 15.3.1.3.2 Method of Analysis
,m I
The lese-of-ecolent-ficw cccident is analyted using a combination
\\. /
cf analog and digital computer programs.
The PUMPC1) analog simulaticn used to determine the reactor f1 curate following a loss of pumping pcwer includes descriptions cf fl.cw pressure drop relationships in the reactor ecolant loops.
Tump flow characteristics are determined frcm manufacturers' zone raps.
Flow speed, flow torque, and ficw head relaticnships are solved by pump affinity laws.
The power, inlet temperature, and system pressure are calculated Py a B6W digital ccde, CADD,(2) that includes a point kinetics nedel with a closed-lcop simulation.
The locp simulatien includes a pressurizer medel and a steam generator modsl, the whcle of which is ccnnected through time delays to the kinetics calculations.
The point kinetics medel used to determine the neutron pcwer during the coastdcwn and following reactor trip includes six delayed neutron groups, centrol rod worth and rod insertion characteristics, and trip delay times.
ESU hac developed the RADAR (3) digital ccmputer code to calculate fuel temperature, cladding temperature, and CNS ratio as a function of time fcr reactor system transients.
Input to the code censists of ficw, pcwer, inlet temperature, and system pressure as a f.uncticn of time. 'The core trancient analysis code sin.ulates the reacter core thrcu9h the use of a twc-channel l
J model.
Each channel consists of one fuel rod with its associated b
flew area and spacer grid gecmetry.
Given the necessary input as stated above, the ccde will calculate a prescure drop across a typical reactor channel (average channel) as a function of time.
This pressure drop is then imposed en the secend channel (usually a hot channel) to determine hot channel ficw and DNB ratio in addition to fuel and cladding temperatures.
Ccmpared to the average channel, the het channel has greater heat generation and reduced ficw areas, as well as statistical hot channel facters.
She analytical fuel pin acdel contains a transient response calculaticn, while the hydraulic mcdel considers the steady-state solutions of energy, mass, and momentum balances at each time step.
The transient res;cnse is cbtained by applying the changing ficw, power, inlet temperature, and system pressure to the initial conditiens of the average channel.
Tnis calculation yields the average channel pressure drcp as a function of time, alcng with the het channel pcuer, inist temperature, and cystem pressure.
12 c-
.L :- e.+ ;.;- c u ~ r. the m nrense or the hoi. cl nw 1 in t r:r.3 of c
C2f<
~:'.g
- ~<p:rOtU/;e I*.2 il
- 't.199!6%u.Tt.r ' UO O' A fFC'O iU cal.n:la t d e.s c : u :e:ica o f 2 :.: e.
The les s-cf-cc cian':- Cl.:w a nalvsis has been carried out in the a
power rance for coastdcan irem pcwer levele betusen the rated and s
the iasign overpcwer condi.icas.
The parau.cters used in the
'l
'nalysic sre giv n in Tab 30 15.3"3 Revision 2 12/77 15.3-2
51-1130767-00 m : m a,c x a w 15.3.1.3.3 Desults cf Analytic The results of th.s en..lysit. s hv.! that thc ice m all four rsactor ecclcn pumpc can SS sucrained from 102L rnetd poorer (2%
allowance for hest ba?.anco errer) without damage to the fuel.
The results of the evaluaticn are presented in Figures 15.3-1 fg9LA' through 15.3-4 Figure 15.3-1 shows the power and reactor coolant ficu as a function cf time af ter the lecu of all pump d'd N pcwer.
Figure 15.3-4 showc the RC system pressure and
, g3g4.I i
temperature.
The minimum DNO3 and the ma::imum 1ccal heat flux as a functicn of time are shewn in Figure 15.3-2.
Figure'15.3-8 2[
shows the mininum C:DR fer the three pump and ts.c pump cases, rigure 15. 3-5 shcwn the minimum DN3R which cccurs during coastdown from various initici poaer levels using the minimum tripped rod worth (as sumine 1% ak/k het, chutdo:n enrgin).
The degree of core protecticm during ccastdown is indicated by ccmparing the minimum DNER for the coactdown with the criterion g value of 1.3.
The RC cystem is capable of providing natural circulation flew after the pumps hnve stopped.
The natural circulation characteristics of the RC system have been calculated with conservative values for all recistance and fcrm loss factors.
Resistance and form less factors were deternined using the system design (ma ximum) resistancec with additicnal conaarvatism applied, including the use cf leched retor resistance: for all primary pumps.
l.)3 No voids are assumed to exit in the core or the recctor outlet f3 n #AA0A N-piping.
Table 15. 3-3 shews the natural circulation ficw ppd"It capability as a function of dscay heat generation.
Thaca flows-provide more than adequate heat transfer capatilit/ for core
/N S #dU~
ccoling and decay heat removal by the RC system.
g; v
15.3.1.4 Earrier Forformance The loss-of-coolant-flow accident dces not result in fuel damage or excessive pressure in the reactor coclant system since the l
core thernal pcwer does not c::cced 1127. cf rated power and the peak pressure never exceeds code allowable limits.
15.3.1.5 Padioloaical Consecuences This event will result in the release of steam from the secondcry side to the atmosphere.
It is shown in Subsecticn 15.3.1.3.3 that l33 fuel damag= will not cecur.
F0r this rocson there will be no i ncrease et radicactivity in + he reacter ccole.at or in the Ltosm.
T"'ctse th?
'.ccts cra n functirn of t.ho
.ccet c !. L t. c *m c 1... a A /.i.,
H e p0tanni:
- .e Sielc,:tcal ccm wa % ac s ci; 3.r..c t,.:.
" 11 %
ii c: 04.er.
- w. n u.e :... a e q m n e :e. d lou
.. c nen _f.c:g.:,;f ce rever to the teaticn (ace S etcecti en 15. 2. 6. 5).
l33
.,i
.,a Revi.sion 33 1 h,;- 3 y5i l
l b{
l 1
51-1130767-00 Insert B 15.3.1.3.3 Results of Analysis The results of the analysis show that the loss of all four reactor coolant pumps can be sustained from 102'.' rated power without damage to the fuel.
The results of the evaluation are presented in Figures 15.3-1 through 15.3-4.
Reactor power and coolant ficw are shown in Figure 15:3-1 as a function of time after the loss of all pump power. Maximum local heat flux and minimum DNBR vs.
time are shown in Figure 15.3-2.
The minimum DNBR which occurs during coast-down from various initial power levels is shown in Figure 15.3-3.
The degree of core protection during coastdown is indicated by comparing the minimum DNBR for the coastdown with the criterion value of 1.3 indicating that margin exists for all power levels examined. RC system pressure and average moderator temperature as a function of time are shown in Figure 15.3-4.
The loss of power to all RC pumps has also been evaluated for three -pump and two pump initial operation condi tions. Minimum DNBR versus time is shown in Figure 15.3-8 for loss of. power to the RC pumps from three and two pump operation.
The minimum DNBRs are summarized in Table 15.3-2 for the loss of power to all pumps ftm four, three, and two pump operation, indicating loss of pour from four pump operation as the
(
most limiting with a minimum DNBR of 2.0.
The minimum DNBR is well above the 1.3 l
minimum DNBR criterion.
The pump coastdown transient has also been evaluated for the loss of one reactor coolant pumo from initial four and. three pump operation. The accident initial conditions are specified in Table 15.3-5.
The loss of one RC pump from four l
pump operation is the limiting transient with respect to DNB since this transient 1
forms the basis for the determination of the setpoint for the flux / flow trip i
function of the Reactor Protection System. A limiting flux / flow setpoint is determined based on the core DNB response to the one pump coastdown from four pump operation using the minimum acceptable RC flow coastdown.
This procedure l
51:1130767-00 !
Insert B (continued) assures an acceptable DNBR response with respect to the 1.3 criterion. The minimum DNBR versus flux / flow setpoint (5Md is shown in Figure 15.3-14.
. Sn3
. The flux / flow setpoint determined in this analysis is subscquently adjusted for flow and instrumentation errors. Minimum CNBR versus time is shown in e
Figure 15.3-15 for the one pump coastdown from four pump operation using a flux / flow setpoint of 1.1335 as the limiting value for Midland Cycle 1.
The loss of power to one RC pump from three pump operation has been evaluated from an initial power level of 77% of rated power hesulting in a 2/1 to 1/1 change in pump status.
Reactor trip is provided by the power / pump monitor trip function. Reactor power, RC flow, RC system pressure and inlet enthalpy are s
shown as a function of time in Figure 15.3-12. Minimum DNBR versus time is shown in Figure 15.3-13 indicating a minimum DNBR of 2.28 which is well above the criterion limit. A comparison of the minimum DNBRs for the loss of one RC pump from four and three pump operation is provided in Table 15.3-6.
s
\\
T
~
51-1130767 '00 Insert C a
Natural /irculation flow capability as a function of decay heat generation is shown in Table 15.3-7.
These flows provide adequate heat transfer e
capability for core cooling and decay heat removal by the RC system.
o 1
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4 s
S
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e 1
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-,...., ~..,. _ -.... - -... _.....
o
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- MICLAND 162-FOAR
.a1 -11 3 0 3:,,67-00 15.3.2 I"JP. RECIRCUuTIO!! LCOP CONTFOLLER MALFUNCTION s.
1.ct applacEble to PWR
(-
15.3.3 REACTOR CCCLANT PUMP SHAFT SEIZURE (LCCKED KCTOB)
,15.3.3.1 Identification of Canser and 'Frecuenev Claseification The reactor ecclant flowrate is reduced if' one er more of the reactor ecolt.nt pumps fail.
A pamping failure can occur from v.echanical failures er frem a 1:s3 of electrt.c power.
With four indepandent pumps available, a mechanical failure in one pump will not affect cperation of the others.
The loss of ficw due to rechanical malfunction from any cause has been considered and analysed ac the locksd retor accident.
This is c nonmochanistically porzulated design basis e' vent class'ified as a limiting fault; therefore, only one pump is assumed to be utfected.
15.3.3.2 Sagg,ence of Evants and Systems Oceration Table 15.3-8 91voc the sequence of avants.
The reactor is protected fr:n the consequences of ro ctor coolant pump f ailure (c) by the reactor prctection system.
The reactor is tripped if insufficient reactor coclant tiew exists fer the power level.
The ICS initiates a power reducticn upon pump failure to (CN prevent reactor powcr frcm exceeding that permissible for the A./
available tiow.
Even tucugh ICS acticn is available to prevs;nt er mitigets this act:1 dent, the accident analysis was done without ICS action.
15.3.3.3 Core and Syst4m Performancos 15.3.3.3.1 Safety Evaluation Criteria The cafety evaluatica criterion for this accident is that no fuel cladding failure shall occur.
15.3.3.3.2 Method cf Ana.1.ysi c The reactor coelF.nt pump Ghaf t GeituIC Occident is analy2Od by using a combination of analog and digital cc. piter programs.
The PUMPC4' anale:- minulation used te., Oter;r.:.n th2 reccser D.owra to..:.119 r 4 m; e ic.:s of Pnreing ov. 2 u:c ;.u ' 9 d;2.;r;j? ions ci [1 -).2 V 'It'."tr 2. ' ; ' E '. ". C, i. 2
- n t. : ?.
.. r C ". r 1 coo.*.a n :
J. C.*; p n.
Fump flo
- 1. : u rf
.t. ca c;;c e;;: err. : ;.:.;,.an
.....J u ~.ur..:
- :,nu cr. ps.
Fles s;eed, fic..e cer y t c., and :'1:/. h.ed.ulat.ionships.re tolved by, u.r.:; atrintt; Jc s.
p..
f
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4 :. 3-c.
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51-1130767-00 MICIAND 102-FSAR The power, inlet temperature, and system preueure are calculated by a b w digitcl cede, C',00, 23 that includan a scint hinatics
' ")
mc9tl with a closed-1 cop circlatien.
The locp 'simulacien includes a precturiter model and a steam gencrator nedel, the whole of which is ccnnected through time delays to tha kinetics calculaticns.
The point hinetics medel used to determine the n=utren pcwer during the coastdcwn and folicwing reacter trip includes six delayed neutren groups, control re,d worth and rod insertion characteristics, and trip delay times.
E~N has developed the RADAR (3) digital computer code to calculate fuel temperature, cladding temperature, and CNB ratio as a functica of time fer reacter system transients.
Input to the code consists of ficw, pcuer, inlet temperatur0, and system 1'
pressure as a function cf time.
The core transient analycia code cimulates the reacter core through the use of a two-channel model.
Each channel cenaists of one fuel rod with-its associated flow area and spacer grid gocmetry.
Given the t'ecessary input as stated above, the cede will calculate. a pressure drop across a typical reacter channel (average channel) as a function of time.
This press.ure drop is then imposed cn the seccad channel (usually a hot channel) to determine hot channel ficw and DUB ratio, in addition to fuel and cladding temperatures.
ccmpared to the average channel, the hot channel has greater heat generation and reduced flew areas, as well as statistical het channal factors.
The analytical fuel pin model centains a transient response calculation, while the hydraulic model considers the steady-state solutions of energy, mass, and momentum balances at each time step.
The tranrient respense is chtained by applying the changing ficw, pcwer, inlet temperature, and system pressure tc the initial conditions of the averago channel.
This calculation yicida the average channel pressure drt; as a function cf time, alcng with the hot channel power, inlet temperature, and system pressure.
From these results, the response of the hot channel in terms of cladding temperature, fuel temperature, and CNB ratio is calculated as a function of time.
The locked retor accident is a rapid decrease in flow resulting frem the instantanscus seizure of a reactor coolant pump retor.
The '.nitial operating conditions are as given in Table 15.5-7 The neutron pouer rises slightly but quickly drops after rod insertion due to a power /ivbalance/ficw reactor trip.
The reactor syctem pressure posks after the neutren pcwcr peak and remains well bslow code limits.
The initial reactor ecolant pressure is adjucted -65 poi to account for uncertaintier.
The assumption that the ini-ial. inlet cmperature fe ne von than 2F theve the nor**C 3 Value.'.? hatAd on ecatrol cycr.sn soneitivity of elf and c.n incizument er' v of f.11.
t 1$.3-5 O
51-1130767-00 m m.AxD i n -r:AR 13.3.3.3.3 Fesul+.: of Analysis The rec.alts of tnic accicient chow that the reactor can sustain a I
e reacter ecolant purp chaft seizure during four pump and two pump operation (one pump in each loop) without damage to the fuel.
The four pump operation resulte are shown in Figures 15.3-5, ara chnv" '
h68 gan 15.3-6, and 15.3-7; two pun p operation resul te Ficure.s 15.3-9, 15.3-10, and 15.3-11 / Figures 15.3-5 and 15.3-9 Ng94 3
show we..m.ctor coolanc :Ac.,w a;.s neutron power after seizure, and Figure 15.3-6 and Figure 15.3-10 show reactor coolant system I jgsn Figures {
b) pre.couro cnd corc inlet enthalpy.
The peak ree.ctor coolant system pressure for both casen is given in Table 15.3-5.
1S And 15.3-11 show the ONB ratin a-d fuel and clad
_, /
ke.m. 3 - 7 perctures after the acinure.f The DNS ratio in both cascc cccps celow 1.3, a s....._..,
s,act fuel cladding curface tecpercture doec not exceed lim? ts.
The percentage of fuel pine
- hich would experience a DIiSR lesc than 1.3 in providea in 16 Table 15. 3-5.
15.3.3.4 Barrier Performance The locked rotor accident does not result in excessive reactor coolent syctem pressure and the cladding temperature does not enceed limits.
The integrity of the reactor vessel is maintain *d.
15.3.3.5 Radiolecical Consecuences
('.)
%.s This event will recult in the release of steam from the secondery side to the atmocphere.
It ic shown in Subsection 15.3.3.3 that 133 fuel camage will not occur.
For this reason thare will be no increase in radioactivity in the reactor coolant or in the steam.
Eeceuse the doses are a function of the amount of steem released, the pctential radiological consequences of this event will be less severe then the consequences of loss of nenemergency ac power to the station as discussed in Subsection 15.2.6.
15.3.4 REACTOR COOLANT FU;G SHAFT BREAK The lors of reactor coolant flow due to a pump shaft breakage or ct?.cr mechanicci talfunction hac been examined for its effects on l
core integrity.
The frequency cf occurrence of thic accident is expected to be the same ac thct for any grons mechanical failure if the primary cyLcem.
There fore, only one pump is assumed to be nilkCtCc.
Th; !fftete of nc loca of re.acr.cr coolant flow ciue to a pump
"'u f ;
. v.
- a. c b.' rc* 1C by the cannw7.: ncca of the. 1.r..ched ':^ tor a ce.t eent, discusac..i
.a suhcectit.n 2i.3.3.3.2.
Tb' # ' c' cc ntdm:n l33 resultAny from the brear.cge et' ene pump chaft in less rapid than that resulting from n shaft seizure situc' ion indic tive of the i<evicicn 33
(
15.J-6 4/01 11
51 - 1 13 0 T 6 '7 - 0 0 Insert 0 15.3.3.3.3 Results of Analysis Reactor coolant flow und neutron power after seizure are shown in Figures 15.3-5 and 15.3-9.
Reactor coolant system pressure and core inlet enthalpy are shown in Figures 15.3-6 and 15.3-10. The peak reactor coolant system ss pressure for oath cases.bre given in Table 15.3-5.
The minimum DNB ratio, fuel and clad temperature after seizure are shown in Figures 15.3-7 and 15.3-11.
.l 1
I l
l
51-1130767-00 momo u2-rsu i
locked-rotor accident.
In either case, the reactor is tripped if
/" '.
iarufficient reactor coolant flow c):ists for the pouer level, b..'
The taargin of core protection indicated by the C::SR during the coastdown is greater for the pump shaft breakage than for the case of shaft seizure since the flow decrease is not as rapid.
Thus the power / imbalance / flow trip function of the reactor protection system ensures adequate protection of core integrity.
E I
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5..I 15.3-7 lh
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-. ~... -.,
- - - = - - - - - -
51-1130767-oo rucuno 1c2-rSu REFERENCES C
1.
74;?tP - Ana
~
2.
,gADD - Cemeter Acolicatien to Direct Dioital Simulation of
_Trc.s en.s in Jarer IMactors, BAW-10076P-A, Rav. 2 (May 197 5),
B6i!
3.
ff.pM_- Fe a,gtor_ Therrj i a nd Hydraul3 c.ine
- yttin Duri ng Ra.=.ctee Floa coachcc:in, BAW-100 69A, Rev. 2 (April 1973), BGW h
e G
- f.,
i
\\.
15.3-0 M
51-1130767-00 TABLE 15.3-1 SEQUENCE OF EVENTS FOR LOSS OF POWER TO ALL REACTOR COOLANT PUMPS
/
(TIME,J9 Event 4 Pumo Operation 3 Pumo Ooeration 2 Pump Ooeration Pumps begin coasting down 0.0 0.0 0.0 Control rods begin to drop 0.62 0.62 0.62 Minimum DNBR occurs 1.8 1.45 1.25 TABLE 15.3-2
. MINIMUM DNBR FOLLOWING VARIOUS PUMP C0ASTDOWN TRANSIENTS Transient Minimum DNBR Trio Function 4 Pump Coastdown From 4 Pump Operation 2.0 Flux / Pump Monitor l
3 Pump Coastdown From 3 Pump Operation 2.5 Flux / Pump Monitor 2 Pump Coastdown l
From 2 Pump Operation 3.8 Flux / Pump Monitor
l 51-1130767-0(
MICI.A!!D 102-FSAn
'T A3T E 15. 3-3,
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MULTIPLE REACTOR COOLANT PUMP i,
I s.,
1 RIP ACCICERT PMAtlETERSC 1)
I Four Puno Ocaration f2 Steady-state design overpower, % of rated power 112 DNBE at. steady-state design everpower at full flow 2.36 tsaximum indicated power, %
100 Maximum real power, % of rated power 102 II Three Pumn Oceration Real power, % of 102% of rated power 75 BC flowrate, % of fdur pump design flowrate
.74+ 74,8 f
2 JII We Fumo onoration Peal power, % cf 102% of rated pcuer 50
~,
FC flowrate, % of four pump design flowrate 49.5
( 5 ) Analysis Assumptions:
a.
The initial pressure and inlet temperature are nominal values minus 65 psi and. +2F : *spectively.
b.
The trip delay time is '62Q mr.
The percent cf rated ne'utron pouer (at 'beginning of c.
core life) as a function of time af ter trip is shown in Figure 15.3-1 (including trip delay of 620 ms).
d.
This figure represents the shutdown curve using the minimum tripped rod worth, which is the worth required to ensure a minimum het shutdown margin of 1% ch/k with a c uck rod.
e.
The pump motor inertia is 70,000 lb-f ta.
s e
novision 2 12/77
51-1130767-00 TABLE 15.3-4 SEQUENCE OF EVENTS FOR LOSS OF POWER TO ONE REACTOR COOLANT PUMP sec I " 'I 4 Pump 3 Pump Event Ooeration Operation Pump Begins Coasting Down 0.0 0.0 i
Control Rods Begin to Drop 1.89
.620 Minimum DNBR Occurs 5.9 1.1 4
h
i 51-1130767-00 TABLE 15.3-5
{
SINGLE REACTOR COOLANT PUMP TRIP ACCIDENT PARAMETERS I.
Four Pumo Operation Real power, % of rated power 102%
RC flowrate, % of four pump design flowrate 100%
Initial Steady State DNBR 2.0 II. Three Pump Operation Real power, % of rated power 77%
RC flowrate, % of four pump design flowrate 74.8%
Initial Steady State ONBR 2.57 NOTES:
- 1. Pump Monitor Trip Delay is 620 M, Flux / Flow Trip Delay is 1.89 /
MS W
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5j-1130787-00 TABLE 15.3-6 MINIMUM DNBR FOLLOWING ONE PUMP COASTDOWN TRANSIEfiTS TRAAlsorAPT MoshMtM DN W WIP MTIM One Pump Coastdowa(0 1.43 Flux / Flow / Imbalance From Four Pump Operation One Pump Coastdown 2.28 Power / Pump Monitor From Three Pump Operation (1) This transient is defined as the limiting transient for the purpose of establishing the Flux / Flow trip setpoint value of the Technical Specifications. The Flux / Flow setpoint is chosen to allow a minimum DNBR of 1.43 (BAW-2 Limit of 1.3 plus 10% margin).
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51 -113 0 7 6 7 - M azotra sc2-rm TABLE 15.3, r.,
NATUP.AL C7EOUII.'fIOl C A F.3,E I L I T Y Natural Circulation Flow Time After Power Level (106 lb/hr)
Power Loss, s
("s of Full Pcser)
Recttired Ave.ilaisle 0.1 7
4.1 6.6 10 5
2.9
- 5. 8 150 3
1.8 4.7 700 2
- 1. 2 4.0 0,800 1
0.6 3.1 f%
f N
0 0
J
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1 51-1130767-00 aIce.ao ic2-rs^a TABLE 15.3 W8 tit'E CECU2 lice C7 EVE!;TS IC3
(.,,
DEACTOR COOLANT FUMP SHAFT SEIZURE ACCIDENT Evant Time, s I
Four Pumo Operation Peter en one pump locks 0
Fewer /ir. balance / flow trip setpcint is reached 0.40 Control rods begin to drcp 1.90 Peak system pressure occurs 3.20 II Two Pume Oceration Roter en ene pump iceks-0.0 2
rewer/ imbalance /ficw trip setpcint is reached 0.23 centrcl rods begin te drcp 1.73 Peak system pressure cccurs 3.40 4
N
51-1130767-00 MIDLAND 1&2-FSAR TABLE 15.3-[
s.,
Rh;',CN.h,.OOLM7 PUii2 J' IMET SCJ ZURS ACCIDLdT PARA 12T.2RS'"
I Cour Pur.o Oneration Initial power, % rated power 102 Initial flow, % design flow 100 Power / imbalance / flow trip delay time, s 1.5 Peak RC pressure, psia 2,240 Percent of pins with MDNBR<1.3
<1.0
/HAMMM CLAa SU254CB TGWWN> F toto II Two Pump Operation Initial power, % of 102% of rated power 50.0 Initial flow, % design flow 49.5 Power / imbalance / flow trip delny time, s
1.5 g3 Peal: P.C sy:: ten precsure, psia 2,235 A r~c e n t-Y :
$Ainnthnof Pms worm k hrf sit C 1.
9 CLAb M9AfflR&; 5
-S TW 06
"' System characteristics
'a.
The initial reactor coolant pressure was ascuued to be 65 psi lower than norms 1.
The reactor ecclant inlet temperature was accumed to be 2F higher than normal.
b.
Maximum design conditions were assumed for the thermal conditions.
The reactor is tripped by pcwcr/ imbalance / flow.
c.
fN q,, :
I e
t.2
.s ENTHALPY
- 1. 0
- 1. 0 -
- 1.0 1
POWER PRESSURE 2140 g
0.8 g
FLOW E
a M
f 5
E 0.9 -%
as E
5 S
E
- 0.6 S
3 a.
E E
4 S
e c
2 2
'd E
2 2
Mc
"- 0.4 2135 0.8._
l i
I I
0.0'
> q.
0-0.5
- 1. 0 1.5 2.0 2.5
. i.
Transient Time, Sec f
.,i
=i
.'0 REACTOR COOLANT SYSTEM PRESSURE. REACTOR COOLANT s>
FLOWRATE, CORE INLET ENIHALPY AND REACTOR POWER VS TIME AFTER 4 ONE PUMP C0AST00WN FROM 3 PUMP OPERAll0N-(2/l-1/I) AT 77% RATED POWER
't Figure 15.3-12 Js g
'l
51-1180767-00
- 2. 7
=-
- 2. 6 C.
5
- 2. 5 5
5 E= 2.4 2.3 1
I I
~
f' O
0.5 1.0 1.5 2.0 2.
Transient Time. Sec illNIMUM DEPARTURE FROM NUOLEATE BOILING Rail 0 VERSUS TIME FOR A ONE PUMP COAST 00WN FR0t! 3-PUMP OPERATION (2/1 1/1) AT 775 0F RATED POWER Flp,ure 15.3 13 l
e e
~
51-1130"'87-00 o
1.5 o
n 9
5 5
8 I
e.-
EE I
E
- 1. 4 l
1 I
l I
I l
l 1.1335 1.3 1.09 1.10 1.11-1.12 1.13 1.14 1.15 Flur/ Flow Trip Setpoint (S )
m i
MINIMUM ONSR VERSUS FLUX / FLOW TRIP SETPOINT FOR ONE PUMP C0AST001'!N FROM FOUR PUMP OPERATION Figu re 15.3-14 l
[
1
51 - 1 13 0 ? A 7 - 0 ()
2.0 1.9 1.8 C
E 1.7 E
E
- s 1.6 1.5 1.4 i
0 1
2 3
4 5
5 7
Transient Time, see MINIMUM ONBR VERSUS TIME FOLLOWING A ONE 80 PUMP C0AST00WN FROM FOUR PUMP OPERATION (S =1.1335) 3 Figure 15.3-15 I
m m
y
- - - -