ML19257B538
| ML19257B538 | |
| Person / Time | |
|---|---|
| Site: | Crane, Davis Besse  |
| Issue date: | 12/21/1978 |
| From: | BABCOCK & WILCOX CO. |
| To: | |
| References | |
| TASK-TF, TASK-TMR NUDOCS 8001160814 | |
| Download: ML19257B538 (19) | |
Text
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DRAFT STEAM GE.'iERATOR LE'/EL EFFECTS ON PLRIT OPERATION PREPARED BY E. SWANSON, B.M. DUNN BABC0CK AND WILCOX FOR TOLEDO EDISON 12/21/78 DRAFT 1917 116 9
8001160-6/
0
TABLE OF CD::TE"TS I.
INTRODUCTION II.
C0"0LUSIOUS III.
DISCUSSION A.
Relationship with Events Presented in the SAR B.
Loss of Feedwater and Loss of Off-Site Power - Introduction B.1 -
Loss of Off-Site Power B.2 -
Loss of Feedwater APPENDIX - Bounding Analysis of Loss of Feedwater Event With Failure of Operator to Control feedwater at 35" TABLE 1 -
Steam and Feedwater Line Rupture Control System (SFRCS) - Actuation Parameters I9l~7 117
t 1.
INTPCOUCT!01 The Davis-2 esse Unit 1 5::am ar.d Feed.15:r Line Ructure Control System (SFRCS) design objectives are to prevent the release of hign energy stac=, to autcccticcily start auxiliary fccc. eater ( AFE), and to provice acequate AFJ, via essential steam generator level control, to remove decay heat during anticipated and design basis events when AP.! is required. Table 1 correlates the station variables and accident con-ditions for which AFW actuation is required. iFor all actuation ~ signals, the SFRCS initiates and controls AFW addition automatically _to maintain
~
, a 120" level (96" indicated on the startup range instrumentation) in the steam generators.
The recen't natural circulation test at Davis-Besse 1 (TF800.04) demonstrated that a 35-inch (indicated) steam generator level of AFW provides adequate natural circulation for decay heat removal.
The auto essential SG level control setpoint of 120-inches (g6-inch-indicated):is, thus in excess of minimum, SG level requirements.
Operating experience, such as the November ll,1977 incident, indicates that the addition of AFW at a rate of 800 gpm to each SG to achieve a 120-inch level produces a primary system cooldown for which indicated pressurizer level may be lost momentarily under certain conditions.
In response to these facts, operating instructions requiring manual
- control of steam generator level at 35-inches en the startuo range level indicators following non-LOCA events-wnere developed and used at D-3 Unit 1 pending installment of permanent design enanges to tne SFRCS.
Margin in maintenance of indicated pressuri:er level and assurance of acequate natural circulatien ca: ability will exist thru caerator 1917 118
intervention curing condi-icns unere AFW is required.
Sctsequen: to tne use of -he interim site operating procedures d:scribec a::ve'[f RC questions rein:ive to the c:nsequencas cf the ccerat:r failing to control AFW to a 35-inch during anticipa ed events such as tne less-of-normal feedwater or loss of off-site powerzhave arisen. The extent to which the primary system is cooled resulting from feeding the SG with cold AFW, the impact of over:coling the RCS on pressurizer level',
and the con. sequences.on core cooling are specific items which have been identified.
S&W has reviewed the NRC concerns and has concluded that
- failure of the operator, to comply with the present operating instruction will possibly result in a momentary loss of pressurizer level l and/or level indication under certain conditions ibut will not produce
' consequences which are non-reversible or detrimental to safe operation of the plant.
Provided below is further discussion to support this position.
1"9 1 7 l'1 9 2-
TABLE 1.
STEA" A.;0 FEEC'..'ATER LU;E RUPTURE CO.:TRC:. SYSTE.4 (SFRCS) ACTL'AT*C:: P ARAMET EF.S Ooerational Events A: tea-ion Parareter Station Variables Setooint
< 591.6 psig.2 Steam Line Break l
1.
Low Steam Line Feedwater Line Break Pressure I
2.
Low SG Level i 17 inches Loss of F/W l
3.
SG Pressure Minus
> 197.6 psi -
FWLB, LOMFW Main Feedwater Line Pressure 3
Loss of Off-Site Power 4.
Loss of ALL RC Pumps NOTES 1.
When actuated, SFRCS closes both main steam isolation valves, closes both main FW control and stop valves, initiates AFW and controls AFW to maintain a 120-inch level in the SG's.
2.
Aligr..ent of AFW to a pressuri:ed SG is provided for steam and feedwater line breaks.
3.
AFW initiation but not steam and feedwater line isolation occurs I917 120
C ~ '. ' '_' I C '; 5
- r SF:.C: a:: atier, and fill cf tne steam generat:rs to the auto-e::ential level con:r:1 :cir.t Of 120" withcut oper:::r acticn:
No Unreviewed Safety Question exists The loss of offsite power transient will not cause the cressuri:er to drain although a loss of pressurizer indicated level will occur.
The loss of feed'. tater transient may result in oressuri:e emptying however acceptance criteria for DNS will be met.
Steam bubbles which exist in the reactor coolant for a short time will be collapsed by HPI injection.
Pressurizer refilling by HPI will occur.
No return to oower will result in the long term.
III.
DISCUSSION The following section is divided into three segments:
Relationship with Events Presented in the DB-1 FSAR, Loss of Offsite Power, and Loss of Feedwater.
A.
Relationshio with events cresented in the SAR Addition of auxiliary feedwater at rates considerably greater than the decay heat generation rate will result in overcooling of the reactor coolant, contraction and a reduction cf oressuri:er level.
This secuence of events is typical of several transients cresented in the FSAR wnich have been submitted to the NRC and a::oreved as a part of the Licensing Process.
Overcocoling transients can be caused by a variety of circum-stances, failures, ccmbinatiens of c:erating ecuictent, and imor:cer operator interactions.
From a cractical view:cint each single discovera:le possible transient cannot be analy:ed and cresented as a cart of the riAF.
analysis, b;; a br:ad variety of transients have been selected.
Pis scecific transien fits within :nat br aa caiegory; each of tne F5AF, transients has been cemenstrated :c cr:auce ac:ectable results.
1917 121
Overcooling trancier.ts re:alting fr:: a varic*.y of causes are ccccrite:c in Section 15.2.10 "Eftessive Heat Re.: val due to Feeivater ".alfe::ier.s'.
This section describes a transient resulting from excessive main f=: caster addition which is similar to the specific transient of increased level addition by auxiliary feedwater.
Further information is presented in response to question 15.2.15 and 15.2.16.
The steam line break (see sections 15.4.4, 15.4.8, 15.4.1) is the most severe overcooling transient in that the reactor coolant system is decreased 50F in average core temperature over a 30 second time.
This is compared with the cooldown in question which takes a much longer time to achieve a similar temperature drop and system conditions.
During the SLB, system pressures are reduced fem 2200 psia to about 900 psi as system temoeratures are driven toward eouilibrium with the unaffected (pressuri:ed) steam generator attaining saturation temperature of about 530 F.
The cressurizer is near empty at about 20 seconds and thereafter loses its influence on the system thus permitting the uoper elevations of the coolant loop to approach saturation as cooldown continues toward 530 F.
HPI injection pumos are actuated on low RC pressure such that oressuri:er level will be restored.
As shown in. Figure 15.4.4-1 and 15.4.4-2 of the 03-1 FSAR, the rapid cooldown of RCS after reactor trip is limited by the pressure maintained in the pressuri:ed steam generator in much.the same fashion as anticipated for events such as the event of concern.
As the RCS approaches saturation, core cooling is not impeded.
Minimum 0.SR si.3 N
occurs just before reactor trip and subsequently increases with substantial margin throughout the remainder of the cooldown.
d.17 122 The close relaticnsnic ef the auxiliary feet,ater level increase 25 aa overcooling transient with these similar overcooling transients alicws us to draw the conclusion that no L'nreviev.ed Safety Question exists.
To shew
.E.
a com:,arison to the d2: ailed arMyses re:crted in the FSAP., we have cer-formed conservative counding analyses cf. o recresentative cases.
Tnese accroximate analyscs are cresented below.
3.
Less of Feecwater and Less cf Offsite c.ter Introduction o
We have briefly analy:ed two transients resulting from auxiliary feed-water addition and establishment of level above the site operating in-atruction 35" limit.
The two transients examined are a loss of'offsite ipower (reactor coolant pumps stop, makeup stops, main feedwater stops) and alloss of _feedwater (reactor coolant pumps continue, makeup continues).
Of these two transients the loss of feedwater results in the greater volumetric coolant contraction because the forced coolant flow (RC Pumps operating) causes a faster rate of heat rejection to the steam generator.
1.
Loss of Offsite Power Preliminary calculations for a reactor trio following a less of off-site power show that the pressuri:er loses indication but dees not empty. The assumptions used to derive this result included full runout auxiliary feedwater flow '(s2400.gpm) resulting in a fill time sto.120".of about 4 minutes. t No net mass change to the primary coolant l(no makeup, no letdown)'was considered even though the makeuo controls would respond to decreasing pressurizer level with increasing tne net input to about 200 gpm.. At the termination.of the transient the pressuri:er level is slightly above the outlet into. the surge line.
Reactor c clan: cressure remains above 1500 :si and hign pressure injecticn is not automatically initiated.
Althcugh the net makeup was nc: censidered, i: wculd in fact, cause One pressuri:er to refill to the ner 21 level.
At the same time compression of the steam would cause a cartial recressuri:atien cf jgj/
}23 the system ensuring that the cociant remains subccoled.
This trar.siert presents no safety concerns.
2.
Loss of Feedwater This transient has a greater reactor c5elant contraction than the less of offsite power case resulting iniemotying of the pressurizer.
Con-sequently it will be described in creater detail.
A brief sumary of the events is:
Reactor trip; time = 0 Makeuo control valve opens wide admitting full makeup to reactor coolant system time e 0 AFW initiated time :8 40 sec Pressurizer emoties; RC system pr... ure slightly greater than 1800 psi time =: 2 min HPI initiated by RSFAS; makeup isolated time e 2+
min Steam cenerator level = 10 ft; voids exist
.ime = 4 mm.
in reactor coolant HP1 inflow reolaces volume cccu:ied by voids; time 8 min oressurizer level begins to be restored The major concerns that evolve from this transient are the disposition of 'he steam voids and the approach to OMS.
Both of the concerns are ameliorated by tne reactor coolant cumns. Steam voids will nct collect in reactor cociant piping and no ficw bicckace will occur because of discersal and.ixing by the fer:ed flew.
The DNS acce:tance criterion limit will be met because tne ::wer output of the core 4?> at the decay hcat level and all reac:ce ou-:s are C er2:i":
maintaininc core heat removal.
We conclude that no safety nec:lem i exists.
]
l /
} 2/}
_7
A detailed discussion of this trar.sient based on an approxi. ate m
bounding analysis is presented in the accendix.
O 1917 125
-S-
APPENCIX E u-.faa Ar.alysis of Less of Fr.ec. *:er Evert '..'i' h Failure of C--ra :- to Cen rol Feec.ater Level at 2E" Intr:fuction:
The following bounding analysis conservatively predicts the events occurring within the primary reactor coolant system and reactor following a loss of main feedwater from 100% power for the Toledo NSS.
Auxiliary feedwater control has been assumed at 10 feet within both steam generators.
Results:
Because of the conservative, bounding, nature of this calculation, the overcooling of the primary system due to auxiliary feedwater injection causes a contraction of coolant volume sufficient to create steam within the primary system.
The steam is shown to be uniformly distributed within the RCS and the void fraction is 4%. The reactor coolant pumps maintain, full capability.
The CNS ratio is shown to exceed 2.0 and no return to criticality potential exists.
Thus, during the course of the incident, no core problems develop.
- Furthar, following the time of maximum contraction, the system recovers to full pressure, pressuri:er function is regained and the reactor coolant returns to a subccoled water configuration witncut operator action.
an=1..s4 :
The following assumptiens have been made to assure the bounding nature of the resul ts :
1917 126
.g_
H::ter F:...cr:
1 C.. ur. il L::il'ng ::c;s in tr.e ::c:r genera::rs; 0" after tha: t i *:.t.
TMs assu ; tion is ::nservative as c:re heat w:uic ::m;enstte fcr the cociing ::.:se:
by the auxiliary feedwater.
Initial Coolant Inventories Water:
3 P.CS = 11200 ft 3
Pressurizer = EG4 ft 3
RCS-Press = 10420 ft These assumptions are nominal operating values.
Initial Temperatures:
The whole system is taken to be at T
= 582 F.
average This assumption is a reasonable average.
Initial System Mass: s 500,000 lbm The mass is figured from the temperature and volumes above.
Makeup System:
No credit is taken for additional makeup flow which will occur as the pressuri:er loses level.
(In all likelibcod, the makeup system will contribute aporoximately 3
200 ft extra liquid volume).
Local Fewer (kw/ft):
15.4 kw/ft This value is ::Ler. as the m:ximu alle..ed by Technical Specifica:icns.
Secondary Side Volume At 10 Foo: Level 3
711 ft per cenerator, actual volume.
1917 127
Auxiltery Feet. '.or Fle.-:
LOU.5 ft / min. ;ce gene atcr ac:ual value.
Auxilicry Feec:ater Entr.al;y:
8 Stu/lbm icwer bound for maxirum cooling.
With the initiating event, loss of main feetwater, the reactor coolant system pressure will ste.rt to rise.
Reactor trip will occur on high RCS pressure.
Following trip, the RCS pressure will fall because core power has been reduced and boiling of residual main feedwater or auxiliary feedwater is occurring in the steam generators.
These events are almost identical to those which occur in a main feed line break and are analy:ed in detail in Section 15.2.8 of the FSAR.
In short order, the system will return to its initial configuration but, because the auxiliary feedwater heat absorption rate exceeds the decay heat generation rate, the RCS continues to depressurize.
During this phase, residual main feedwater and injected auxiliary feedwater will be boiled and vented through the steam generator safety relief valves.
The primary system average temperature will fall to the saturation temperature of water at the safety valve set pressure.
At this time, primary and secondary cencitions are expected to be approximately as follows:
Primary Secondary Pressure 1800 psia 9S0 psia Temperature 542 F 542 F 1917 128 Mass 503344 lbm 0 lbm 400 ft
N.A.
Liquid Vclure in Press.
g a
2 min.
- 2 min.
Time into Trar.sient g
g...
It is c:rservative
, assure c r..;lete L,il.; cf inc sec:,r.dary tic.e
.,*.cr ar.:
- .rrict: ecuilibrium :,c:..an prim;ry cnd :t::.ed:ry sides :s thc:e anum;;ic.:
lead :: tne maxir.ur fc'le., en injection of auxiliary feen;ater cr.d therefcrc, maxicum ccn:racticn.
D.CS pressure is held u; by the ste-= bubble ir. :.".c Ort:suri:er.
The time has been estimated by calculating the necessary energy loss by the primary system from its initial conditions, the cass of auxiliary feedwater required to remove this energy and then dividing by the auxiliary feecwater flow rate.
(586 - 542) 503344 s 54 sec.
time s
( 1194-c-) 433 52
=
=
Six seconds was used to estimate the initial pressuri:ation portion of the transient.
In performing the remainder of the evaluation 10 feet of cooled (40 F) auxiliary feedwater is placed in each steam generator and the thermal equilibrium condition calculated.
As after a 10 foot level is obtained auxiliary feedwater flow stops this conditicn represents the maximum contraction pessible.
The state variables resulting are:
primary Secondarv Pressure 560 psia 560 psia Temperature 478 F 478 F En:nal;y cf Water 462 Btu /lbm 462 Stu/lba 3
3 Sce:i fic '.'cl ur.e
.020 ft /lbm
.020 ft /lba 1917 129 _ _
Frc.- ti ; !; ecific volure, the t rirtr/ l i c ;i e.
- .: s ce to calci.15 c -
f = 10G52 f:
"ol u ;G As 1CCE2 is smaller than the RCS cinus pressuri:er vclu.' c, the rcrair.ing vcire must be filled with steam.
3 Y
= 10426 - 10032 = 374 ft y 400 ft st 3
400 ft corresponds to a system void fraction of 3.8% y 4%, and as will be shown later, is of no consequence as far as core heating or system performance is concerned.
This steam volume is larger than actually expected for two reasons:
- 1) seme temperature difference would always exist between the primary and secondary systems, and 2) the effect of core decay heat has been ignored.
Both of these would increase the primary side liquid temperature, thus increasing its volume and reducing the steam volume.
Following this state of maximum contraction, no further heat is removed from the RCS via the secondary side until the RCS rises.
in temperature due to decay heating; this will expand the licuid volume, compress the steam and repressuri:e the RCS. As no mass can be lost from the secondary system prior :
achievir.g 20 psia tre first reheating stage will end at a primary system 3
pressure, ter erature, and liquid volute of 9SO psia, 542 F,10332 ft, Subtracting 3
10426 fr:n 10232 shews that about 200 ft of fluid has been forced back into the pressuri:ar.
Fecssuri:er function would then be restored (if not directly, then, by either the makeup or HPI system), the RCS subcooled and the transient 1917 130 ended.
Several questier.s exist about the transiert:
di:;.c-rted witnin :ne primary :y: cm and can *.-ct volume I.
How is the C O ft2 collect in ene location?
From the au.viliary fete.;ater flow ra e over a minutes are recuired to fill the genera crs.
As the pressuri:er has 400 ft3 in it at 980 psia and ne RCS has 400 fd in it at maximum cor. raccinn approximately 2 minutes are used to eject steam from the pressuri:er to the RCS.
Because this steam will be superheated when it enters the RCS it will first desuperheat and then condense at a ra:e governed by its expanding pressure compared to the contraction of the liquid coolant.
In the time of 2 minutes the reactor coolant will have made about 8 complete circles of the primary system and the steam can be considered well mixed.
As the flow velocity in the RCS will remain normal about 25 ft/sec steam water separation will tend not to occur.
Some limited steam accumulation may occur in the upper head of the reactor vessel as in that specific location of the RC velocity is low.
~
II.
How well will the pumps work? Experiments performed on steam carry over capability show that for void fractions up to 10% no loss of pump capa-bility is observed. This is documented in Figure 5-47 of BAW-10104, "B&W's ECCS Evaluation Report With Specific A; plication to 177 FA Class Plants With Lower Loop Arrangement." Actually pump capability increases for the first 5", of void introduced into the system.
III. Will any return to power be. encountered because of the low RCS tempera-ture? A return to cower can occur fer a non-berated core at 49CF. This tem:erature includes the assum tien of the nest reactive rod stuck out of the core;if that rod were taken as insertec the critical temccrature would fall to at or below 45CF.
Althougn ne credit was taler. for HPI in calculating the RC steam volume below 16CO psia, the HFI will be injec:ing
[917 131
~
borated.-cter and, therefore, creventing any return to power condition.
If the orir,r.ry syste : were to stsblize at 1600 psia and thus prevent the HPI from providing bcrdn the RCS temcerature would be at least iE11F and, therefore, no return to power would be exce:ted.
IV. Will DMB be encountered in the core? The maximum contraction condition is again:
P = 560 psia T = 478F a = 4%,
and occurs at least 5 minutes after power shutdown (trip occurs very early within 10 seconds of main feedwater loss). At this time, the de-cay heat rate is less than 3.2% using ANS + 20% (the LOCA evaluation curve). 'As low pressure and high void and high power are conservative bounds a DNS evaluation was performed at:
P = 500 psia T = corresponding saturated value c = 85 power = 105 S
W = full volumetric flow.
The resultant DMSR was >I5 in the hottest channel with maximum design conditions assumed and well within acceptable values.
V.
Will any steam remain traoped in the primary system? Sene may be tranced for a short period of time in the upper head of the reactor vessel but this will be of no consecuence and will eventually be condensed by thermal conduction through the interfacing water.
1917 132
'.r,r.c u,i on :
Tne maximu : con:raction cf the P.C5 vin cr has ' cen cakulated taking n: creci:
for c.itigating systems (make flow or HP!) and ne credit for decay heating.
.';o adverse conse:;'.lences of the transient have been shown and, therefore, this transient poses no concerns to the safe operation of the plant.
1917 133.
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