ML19093B068
| ML19093B068 | |
| Person / Time | |
|---|---|
| Site: | Surry  |
| Issue date: | 04/28/1978 |
| From: | Stallings C Virginia Electric & Power Co (VEPCO) |
| To: | Case E, Schwencer A Office of Nuclear Reactor Regulation |
| References | |
| Download: ML19093B068 (20) | |
Text
VIRGINIA ELECTRIC AND POWER COMPANY l.}
RICHMOND,VIRGIN.lA 23261 April 28, 1978'. *
'~ **,
ti-' ... )
Mr. Edson G. Case, Acting Director Serial No. 069B/0131J8 Huclear Reactor Regulation PO&M/ALH:das U. S. Nuclear Regulatory Commission Docket Nos. 50-280 Washington, D. C. 20555 50-281 License Nos. DPR-32 Attention: Mr. Albert Schwencer, Chief DPR-37 Operating Reactors Branch 1
Dear Mr. Case:
Your letter of March 16, 1978 requested certain information regaraing our November 22, 1977 submittal for the permanent solution of the low head safety injection and recirculation spray pumps net positive suction head problems. We provided some of the requested information in our letter of April 14, 1978. Re-sponses to request nos. 3.8, 4.0, 8.0, 8.1 and 8.2 are attached. You should now have our responses to requests 1.0, 3.0, 3.1, 3.8, 4.0, 8._0, 8.1, 8.2, 9.0, and 12.0. As mentioned previously, we will continue to forward responses to the re-maining requests as the information becomes available from our A-E.
Very truly yours,
~~
C. M. Stallings -,t---
Vice President - Power Supply and Production Operations cc: Mr. J. P. O'Reilly
----- - - - - - ~
i---- I I '
7B1240014
e Question 3.8 The justifications and assumptions made regarding the time at which the refueling water storage tank empties
Response
The refueling water storage tank (RWST) is never assumed to be emptied.
Instead,it is asswned,for analysis purposes that the containment spray pumps stop with 19,050 gal remaining in the'tank (el 29 ft-3 1/4 in).
In fact, the operator will not stop the containment spray pumps until nfter the 11 Stop Containment Spray Pumps 11 alarm (el 29 ft) is activated.
Thus, .the assmned value for the analysis is conservative because less water is added to the containment.
gtiestion I+, 0 Describe and justify the analytical procedure used to conservatively determine the minimum available net positive suction head (NPSH) for the containment recirculation spray and low head safety injection pmnps.
Response
A conservative calculation of mJ.n1mum available NPSH for the containment recirculation spray and low head safety pumps is obt~ined through the following assumptions:
- 1. Minimizing the mass and energy release rates to the containment atmosphere and maximizing the mass and energy release rates to the containment floor as described in Section 8.2
- 2. 100 percent initial reactor power (2,490 HWth) maximizes the energy release *
.3. Absolute minimum air partial pressure allowed by the technical specification, i.e., 9.0 psia minimizes the containment pressure.
4, 'Jhe max:i.nmm. containment dry bulb temperature which, along with minimum air partial pressure minimizes containment atmosphere density
- 5. Perform sensitivity studies for the limiting extremes of service water temperature, i.e. , 25 to 95°F.
6~ Operate the RWST at its minimurn level perm.i tted by the technical specification, i.e., 386,200 gal, in order to minimize the total water addition to the containment at the time of low head safety injection (LHSI) recirculation from the sump.
- 7. 100 percent spray thermal effectiveness transfers more energy to the containment floor from the containment atmosphere, thereby reducing containment pressure and increasing sump temperature. Both of these effects* reduce NPSHAo 8 *. Perform sensitivity studies for the most limiting single active failure.
- 9. Perform sensitivity studies for the most limiting break location, i.e., PS or HL.
- 10. Perform sensitivity studies for the most limiting break size.
- 11. Utilize the earliest possible.time of LHSI switchove:..-- to the sump.
This is based on the .maximwn positive error for the level instrumenta-tion and minimum valve operating times.
- e. e 120 HaY.imum RWST temperature (1+5°F)
- increases the energy _addition to the containment and reduces the effect of subcoolant injection to the ORS pumps.
- 13.
- Investi~ate the effect of increased condensation heat transfer to the passive heat sinks by increasing the peak Tagami value by a factor of four.
Question 8~0 e
Provide a detailed description of the methods 6 assumptions and justifications used to calculate mass *and. energy release rates for blowdovm 8 reflood, and post reflocd phases of the accident.
Response
The amounts of energy and steam and/or water released to the containment structure are time dependent variables and also depend on the pipe break sizen After the rupture occurs/!' the reactor coolant flows out of the break and flashes,. raising the temperature and pressure inside the contairunent structure.
Sensible heat energy stored in the hot meta1 reactor vessel, piping,. and core, the fission product qecay heat 6 anq the power coastdown heat are transferred to the coolant and tnence into the containment atmosphere.
8~0.1 Fission Product Decay Heat and Power Coastdown Heat The LOCTIC program interpolates a curve representing the decay heat generated as a fraction of full operating power versus time after shutdown. The curve is calculated from the correlations presented in Reference 8-1 (BTP 9-2).
The fission product decay heat.generated as a fraction of full operating power versus time aft~r shutdown is shown in Figure 8 .. 0-,1. This curve is based on an equivalent of 1,.095 full power days of operation prior to shutdown and includes the contribution of heat released by the. heavy elementso For conservatism, and in order to allow for uncertainties, the following margins are added to the values. calculated from "
Reference 8--1 :
Time Interval, secs Margin Fission products 0-103 +20%
103-10 7 +10%
after 107 +25%
Coastdown heat arises from the neutron flux coastdown; the energy release mechanism is _that of nuclear fissiono The amount of coastdown heat deposited in the fuel during each interval is determined by using the coastdown heat rate curve. This curve is characteristic of the particular reactor system and pipe rupture (or accident) being analyzed and is obtained from Westinghouse.
Conservatism The curve of power coastdown is generated using assumptions that minimize flow rate and maximize the coastdown power. The curve
.of decay heat g_eneration is based on the worst case possible, a core with a 3 yr life .. BTP 9-2 suggests a* 16,000 hr operating history (1m83 yr) as acceptable.
1
- 8. 0 .2
. e .
Core Sensible Heat The core contains considerable heat at a temperc1ture*above the average reactor coolant temperature.. Thus the LOC'.l'IC . program
- computes the transfer of heat from the core to the reactor coolant as a function of the amount of water in the core and*
whether or not boiling or convective heat transfer is occurring.
This relationship is conservative from *the standpoint of transferring more heat to the contain.~ent atmosphere than would be expected to occurD The fission product decay heat and the power coastdown heat are added to the core sensible heat.
8-0b3 Hot Metal Heat Sources During a loss-of-coolant accidentu the pressure. and temperature of the coolant are reduced very rapidlyo As a result of the coolant temperature drop, a temperature difference will exist between the.coolant and metallic components in contact with the coolant. This temperature difference is the diving force for heat transfer from the components to the coolant. Typical sources include the core barrelr the control rod drives, and the reactor vessel, and primary piping and valves.
The one characteristic which determines the treatment of a source at a given time in the program execution is whether the source is in contact with coolant liquid or vapor at the time~
As the primary coolant escapes through the rupture 11 the coolant
.interface in the primary system decreases, and more hot metal is nuncovered., 1P Thus 11 the heat receiver (liquid or vapor) for a source is determined by the location of the source with reference to* the interface ..
- For this determination the reactor vessel is divided into three regions: above*top of core, below bottom of core, and the core r~gion.
Whether the source is in contact with coolant liquid or vapor is
.of importance to the program for two reasons: (1) it influences the assignment of the heat released to either coolant liquid or to containment atmosphere, and (2) i t influences the heat transfer coefficient used.
Temperature profiles are calculated for each source by modeling it as a monolithic, one-dimensional slab. An *explicit 11 finite-difference technique is used .. One boundary is considered to be insulated, and the other has either a* finite or a virtually infinite heat transfer coefficient, as discussed above.
A hot metal source that is exposed to the coolant on both sides can be treated as insulated by considering one-half the thickness and considering the surface area of both faces ..
2
Co.nservatism e
In general, where simplified models were used ::..n place of ,texact 11 treatments. only conservative models were .used. Where a choice between assumptions (or models) was required in writing the program, the most conservative formulation was selected.
By '~conservativetN is meant that which results in a higher peak containment pressure .. For exampleg the faster a given amount of source heat is assigned to the containment, in the course of a loss-of-coolant accident, the more. *conservative is the calculation .. Also 6 the larger the given quantity of heat (or the larger the source mass), the more conservative the calculation,,
Several factors favor a rapid source *heat transfer:
a.. high heat transfer coefficient
- b. large heat transfer area (or small thickness for a given source mass)
Some of the conservative features of the_program 0 s treatment of hot metal sources are listed below:
- 1) In ~dealing with sources which have curved surfaces, t.i.'1.e a*ssumption of slab geometry is conservative because the source mass and heat transfer area are computed using the larger circumferenceo Higher source mass (and therefore higher source heat content) is conservative.
Also, the larger heat transfer area promotes a more rapid dissipation of heat to the coolant and hence to the containment.
- 2) When a source is in contact with liquid coolant, i t is assumed that there is 1rirtually no heat resistance.
between its surface and the coolant. This results in a more rapid heat transfer than it a finite heat transfer coefficient were imposed.
- 3) Heat released by *sources in contact with_coolant vapor is assigned directly to the containment in the same time stepr thus promoting rapid heat transfer.
- 4) The values of surface-to-vapor heat transfer coefficients are within the ranges given by McAdams, p .. 5 (Ref. 8-2) or higher.. High heat transfer coefficients are conservative.
8 .. 0.4 Coolant Discharge Model LOCTIC calculates the flow of reactor coolant from the reactor vessel through the piping, to the break and out the break. The break flow is assumed to be critical for all flow regimes .. In the subcooled regime, Darcyes equation is used to calculate the
e
. flow from the reactor vessel to the break o 1-'low out the break is based on the Zaloudek correlation (Ref a 8-3) o In the saturated regime, the Moody model with friction (Ref. 8-4) is employed to calculate flow from the vessel to and out the break. In the superheated regime, the flow is conservatively predicted with the Moody model as the saturated steam flow rate at the reactor vessel pressure.
LOCTIC is capable of calculating the flow into the containment in several modes (ebg., single or double ended primary coolant pipe breaku surge line rupture, etcb) b When multiple streams are analyzed, each flow path is calculated separately; the total flow to the containment is the sum of the flow streams.
'l'he initial state of the primary coolant, before rupture, is that of compressed liquid. At the instant of rupture, the primary system starts to depressurizeo This depressurization continues until the liquid reaches a saturated stateo The flow during this time is that of compressed liquid. Further depressurization results* in the flashing of. the saturated liquid, creating a two-phase mixturec As these bubbles have a finite rate of rise (about 2 ft per sec)" the result is a lower region consisting of a frothing rnb-..ture of steam and liquid and an upper region of steam., Before *blowdown is completedv and as long as the interface between the two regions is above the level of the rupture, a two-phase mixture will*flow through the breako When this interface is at or below the level of the* break" a single phase _(vapor) will_ flow through the break *.
During the time the reactor coolant is subcooled" flow from the reactor vessel to the break is proportional to the square root of the difference between the reactor vessel pressure and the pressure just* upstream from the break, called the upstream pressure (Darcy's equation). Flow out of the break is proportional to the square root of* the difference between the upstream pressure and the vapor pressure (Zaloudek Correlation) e Therefore,. eliminating the upstream pressure and assuming the flows are equal yields an expression for subcoo1ed flow G =
CI< ti) (Eq e 8-1) where:
G = mass flux, 1bm/ft2-sec P0 = stagnation pressure of reactor coolant, psf Pv = the upstream vapor pressure,, psf
. 4
e e
= a smoothing coefficient that merges the Zaloudek predicted mass flux to the Moody predicted mass flux at saturation pressure and zero quality CS = C S (P 0 " K)
Vr = specific volume of the reactor coolant, ft3/lbm K = pressure loss coefficient With the reactor coolant at saturated conditions, the ~ass .flux is calculated with the Moody model with friction~ Thus where h 0 = stagnation enthalpy of reactor coolant A value of loO is used for the Moody multiplier.
Conservatism The Zaloudek Correlation, on which Eq. 8-1 is based~ has been shown to predict mass fluxes in good agreement with experiment (Refse 8-3
- and 8-S)o However,, the good agreement is achieved through the use of a discharge coefficient <1 .. 0 and a smoothi~g coefficient of 1..,0o The L~CTIC Zaloudek equation uses a discharge coefficient of 1.0 and a smoothing coefficient <1o0 ..
Both of these steps increase the mass flux; thus L~CTIC predicts a conservatively high mass flux.
The use of the Moody model meets the acceptance criteria of Standard Review Plan Section 6a2~1.3 (Ref. 8-6) o .
8.0 .. 5 Reflood Hot Leg There is no mass and energy released through a hot leg DER during and after reflood other than that resulting from heatup of injection water as it flows to the break., All water injected into cold legs flows through the core and directly out the break in the hot leg,, and there is no flow through the steam generators.
Cold Leg Use of subroutine COLDER al.lows a cold leg double-e_nded rupture to be analyzed with LOCTIC. Basically., heat and mass balances are performed on from 4 to 6 nodesv depending on the the time period., within the reactor coolant system :for the purpose of determining the heat and mass flow of liquid and vapor through the break to the containmento
- These calculatiol are performed in COLDE,only*after blowdown transient is completed. The time at which blowdown ends is that time at which flow out the break is first calculated to be zero.
Various parameters concerning the core reflooding portion of the accident are *utilized, including the*transient *behavior of the driving head in the downcomer, _the core average heat transfer coefficient, the core carryover fraction; the fl.ow split _between broken and intact loops, and the core reflooding rate. This information is obtained from Westinghouse. The Westinghouse Core Reflooding model is discussed in Ref. 8-8.
At the end of blowdown, the .water level is assumed.to be at the bottom of the active core. The water in the lower plemL'1l is saturated liquid at the reactor vessel pressure. At this time, both the core and downcomer begin to fill. The water from the downcomer mizes with the water in the lower plenum.
The downcomer driving head, average core heat transfer coefficient, flow split between broken and intact loops, and core flooding rate* are all conrrolled by transient data specified as input to the progra~. The independent variable for these data is time after reflooding begins.,,
The heat and mass flows for the reflooding period with
.-accumulators working are illustrated in Figure 8 .0-2 and the symbols used there are shown in Table 8
- 0-1
- I:f the safety injection
- pumps come on before the 9-ccumulators empty, mixing is assumed to occur in the unbroken cold legs
.and/or the inlet plenum so that the liquid entering the downcomer is at a mixed mean enthalpy (HINL).
The heat stored in the fuel rods and generated by fission product and heavy element decay is transferred to the coolant passing through the core dl.'rring reflooding and is calculated with the
- input specified core heat transfer coefficient data.
The heat (DQINT) added to the two-phase flow between the core and the steam generator consists of hot metal sources from the vessel heat and other vessel parts above the nozzles, and thin metalg thick metal, and piping in the ruptured loops.
In the steam generator, heat (DQSTG) stored in the secondary liquid is transferred to the two-phase coolant coming from the
- -,core. Whatever liquid is present in the primary flow is vaporized, and superheated steam exists from the steam generator and flows to the break.
superheated* steam from the steam generators in unbroken loops must pass through cold legs and the reactor vessel inlet plenum in order to reach the rupture in the broken cold leg. In so doing i t passes the points at which hig;tl and low head safety injection water is introduced into the reactor coolant system.
6
- .1-***
'*.' ......_,..,.. ,. - -** T" . . . . . . . "'!; * * * * ,-r- ~1,i;,_.. .'/.*,""' *- ~)':.". - * - .-,,. **** _,_ - " * " * * * *- * . . - ,.* -* - * * . - * - .'":'l)l'#.! --
- . .~ - --- . ---- -- - . . \j -
No cre~it is tak.for mixing steam.
Reflooding ends when the water level in the core reaches the and the resultant quenching of 10 ft elevation (active core height= 12 ft). Use of the 10 ft elevation to terminate reflooding presu.rnes the core .ax.ial power shape to have been a cosine distribution prior to *the accident.
PiiR FLECHT tests, in which a cosine distribution is assumed, have shown the entire core to be cooled down, i.e., all of the stored beat to be removed by the time the water level reaches 10 ft (Ref. 8-7).
Conservatism Conservatism with respect to containment pressure is assured by using the core reflooding velocity reported by Westinghouse with
- the break at the puinp suction, but no spillage out of the break.
The core heat transfer coefficients provided by Westinghouse.were increased to a value of 1,000 for conservatism and input to the code. Also, reflooding, is conservatively assumed to continue until the water level in the core reaches the 1 O ft elevation, and the lower plenum refill time is assumed to be zero.
Although tests made by Westinghouse indi*cate that mixing of the superheated steam exiting the steam generators with the cold injection water in the inact loops generally results in a temperature ri.se in the injection fluid to the mixed mean temperature 11 the quenching effect was conservatively assumed to be zeroD 8.0.6 Post Reflood Releases LOCTIC requires as input the Westinghouse post reflood mass and energy release rates for the pump suction DERD Ref. 8.8 describes the basis and the model used to calculate these releases. These release data are based on a reference temperature for heat stored. in the steam generator metal and
- secondary-side fluid of saturation corresponding to a .reference containment back-pressure.. Additional mass and energy releases become available as the containment depressurizesD The steam generator energy is released .to the containment atmosphere in two stages referred to as the ~equilibration stagen and the ~aepressurization stage.n In the former, the energy sources above the reference pressure used in calculating the mass and energy releases are brought into equilibrium with the containment pressure& The rate for this phase is set by the Westinghouse froth calculation model. In the later, the sources give up additional energy as the containment pressure decreases ..
The rate for this stage is set by the containment depressurization rate.
7
.* r
- * *-** '""" * * *$*-- .. *:**--,~--------,-?,, ... .*.-o*""'"'-*"* ,, **. ,"1.~ * , * - . - . -. . . . ~ ...... ,-.S, ,.-,.--,-~***~;s ... ,.,,-~*t,,,'-9,,,-*w-.~,.-..-- ... ,*,~,~ ...... _ '!.: "f', .... -,,* ,*, ,Tr,.-,,?..,._.... ,-".-~:*:,......~, .. ,. ......~.*,<* t' '1° '.'< '* * ** * ~* '*" ', ;'*,,m"* *:
,s:*
The intact loo~team generators and met~nergies are lu.~ped to~ether for this calculation.
Br-oken Loop Stea:n Generator-:-Equilibration *Stage The mass and energy release rates into the containment during the time prior to broken loop steam generator equilibration (t*eq(b1)) are those supplied by Westinghouse.
At the time (t*eq(b1)) the broken loop equilibrates with the reference pressure (P*}, the calculated containrµent.pressure may be less than the*reference pressure assumed by_Westinghouse. If so, an extension *of' the broken loop equilibration stage is required ... The total additional energy which must be transferred from the broken loop at any time during the extension of the equilbration stage is:
L:1 Ee'l, (b1) where:
= B_roken loop steam generator secondary energy to be removed during the extension of the broken loop equilibration stage.
- Eqv (b1) = The total energy (Btu) available in the broken
- loop steam generator secondary at the reference pressures relative to the lowest pressure and temperature obtained post DEA.
P* = The containment reference pressure used in the mass and energy_release analysis (psia)
P10 = Minimum* containment pressure (psia)
- b. P.j = P* - P I* ; i = 1 A Pi =. P,*-1 - Pj ; i >1 and Et{n defines the i-th time interval following the reference broken loop equilibration time ( t*ei(b1)}.
At the end of each interval a new equilibration time is calculated by tci (b1) = t*et, (b1) + 'b,. E ~i Cb 1)
Where: i* (t,1) tei (b1) = Broken loop equilibration time based on contain.~ent pressure response.
t*e'i- (bl) = Reference equ*ilibration time in the broken loop
.'l* (b1) = Broken loop steam generator cool-down rate at t* (bl)
- n .,
' ' > ' ~ * .-. ' ' * * * ~*- *, * ',I.. ~ ,'X, ** , * , . ,_ ' *** '. * ** : * -*. *."'*- . .O:-""'- ' *.* ,*
- t_i * *- * : * * * . * " ~ - . ~;
. ff~ .
When t. ~'i- (b1) is.al~ulate~ to be equal to accident initiation. the broken loop le is
.current time after assumed* to be equilibrated. The mass and energy rates to the containment from the broken loop between t*ei(b1) and t~~(b1) are equal to the mass and energy release rates provided by *westinghouse at t*eq (bl)
- Broken Loop Steam Generator - Depressurization Stage During *the depressurization stage, the steam generator secondary is brought to the ambient conditions in the contain.~ent. The secondary energy which remains after equilibrati6n is:
E Jer (b1) where:
E = Energy which must be transferred from the broken dep (bl) loop steam generator secondary during the depressurization stage_
The energy release rates during this period are those supplied by Westinghouse plus the additional energy increment:
n_E =E dep (bl) av (bl) ( L>p!-p, c .)
where:
bE = the energy transferred from the steam generator dep (bl) secondary during the tL~e increment and bP = The change in containment pressure during the pr~vious time increment.
The additional mass increment is then calculated by:
bM =hE dep (bl)
.hf'j where. h~9 is the latent heat of vaporization at the current containment total pressure.
Intact Loop Stearn Generators-Equilibration Stage The same procedure is used for the intact loops. However. metal and core energy are lumped with the stea...~ generator energ 1r for.
this calculation.
The equa t:i.ons are the same as those for the broken loop except that the subscript 5 (il) u replaces the subscript 9 (bl) w 9
- a. . ~ .e .
Intact Loop Steam Generators-Depressurization Stage Again the procedure used is the same as the broken loop case except that decay heat, q C, ~'f is added to the heat addition rate~.
'11he additional mass*rate to be added is then MJcy q Jcy
- 8. O* 7 Long-_Term Releases
- The mass and energy releases for a pump suction break in the long-term are generated by decay heat boil-off based on Figure 8.0-1.
REFERENCES 8-1 .*
- -- e BTP 9 Residual Decay Energy for Light Water Reactors* for Long Term Cooling, November 24, 1975 *.
8-2 McAdaro.s, W. H., "Heat Transmission; 11 3rd Ed., McGraw-Hill, New York (1954).
8-3 Zaloudek, F. R., "Steam-Water Critical Flow from High Pressure*
System - Interim Report, 11 IDT-80535, January 1964.
8-4 Moody, F.' J., "Maximum Two-Phase Vessel Blowdown from Pipe, 11 APED-4827, April 20, 1965.
8-5 Zaloudek, F. R., "The Critical Flow of Hot Water Through Short Tubes, 11 l-n-T-77594, May 1, 1963.
8-6 i1Mass and Energy Release Analysis for Postulated Loss 'of Coolant Accidents," U.S. Nuclear Regulatory Commission Standard Review Plan, Section 6.2.1.J, February 1975.
8-7 11 PWR FLECHT Final Report," WCAP-7665, April 1971 8-8 Shepard, R. M., et al, "Topical Report - Westinghouse Mass and Energy Release Data for Containment Design," WCAP-8312, Rev. 1, March 1974.
., . , , . __ .,r.': ..* ,...:-,r-, *7'?,>q'T. *v~*?-~~w--,
.. e TABLE 8.0-1 e SYMBOLS USED IN FIGURE 8.0-2 Primary. cooling syst!3m node - no storage,* empty hr.,*~:*::..::*! Primary cooling system node - with storage liquid flow
- - ~ Vapor flow *
=-=-i Two-phase mixture flow CORCO = Mass at core exit
. CORFLO = Mass at core inlet DQOORE = Heat into core DQDC = Heat into downcomer DQINT = Heat added between core exit and steam generator inlet DQSTG = Heat into steam generator(s)
- DWOS = Mass of low head safety injection from RWST DWDC = Mass stored in down.comer
. DWSI = Mass of high head safety injection DWUB = l.fass from unbroken loops at b~aak in cold leg
.F = Fraction of core effluent flowing to unbroken loops FCS = Fraction of low head.safe~y injection mass into primary system FGA = Fraction of gas accumulator mass into primary system FSI = Fraction of high head safety injection mass into primary system HOO = Enthalpy at.core exit during reflooding HOS = Enthalpy of low head safety injection -water HEFF = Enthalpy at exit of steam generator.Cs)
HGA = Enthalpy of gas accumulator water HIN = Enthalpy* at do'Wllcomer exit HINL ::: Mixed mean enthalpy in i~tact cold legs and/or inlet.plenum HINT = Enthalpy at steam generator inlet
.HSI = Enthalpy of high head safety injection water
--t = Liquid Flow*
-'_,o 1 o-s 10-2 en
. ro r--
CD If.)
"<t"
~
w ('f')
3:
0 a_
_J "'
_J
- , (\J I
L1... 0 L1... en 0 ro r--
]: z (.0 0
1--1 If.)
r-u "<t" a:
~ ('f')
,.*, L1...
1
\
L
'*. ~
e
- L* b_,
- 5 1 a2 3 5 10 10~ 10 0 TIME AFTER SHUTDOWN.(SECONDSJ Fl GURE 8.0-1 DECAY HEAT GENERATION AFTER SHUTOOHN 6TP 8-Z SYR REACTOR CPERRTJNO HISTORY J
- ,it~$.~
FGA*DWGA e ( 1.-FGA) *IJ,lGA HGA FCS* HGA i (1.-FCS) (1.-FSI)"
DWCS *-FSI*
- *DWCS *DWSI DWSI HCS l
- INTACT 1 1 RSI*
~
HS:t COLD REFF CONTAINMENT LEGS A
HINL *DWUB CORFLO
+ DWDC I I
I (1-.-F)*
I -CORCO OOWNCOMER F*CORCO DQDC r HIN t
CORFLO I HEFF HEFF DQCORE F*DQilfT DQil~T*(1-F)
INTACT k1 . CORE BROKEN STE.AM - - - - - .... .. t * , * * * * * :-- - - - L ~ STEAM
. *, " . ** * ,' " . * . r GENERATOR GENERATORS HINT HCO *: _.: : * :: *: : ~ _.*: :. HCO HINT F*DQSTG DQSTG*(1.~F)
NOTE:
REFER TO TABLE 8. 0-1 FOR SYMBOLS FIGURE 8*.0-2
. HEAT AND MASS FLOWS DURING REFLOODING
' ..'. J
. . ._ - . . . , ..... -*** --* . . . . .' .. . ,. ..,......._ ,... . .....- . .* *~ .'"" . . . ... ,,.. ; .., .. .
- -T'A
e e Question 8.1 A discussion of the conservatism in tho model with respect to maxirrµzing the muss and energy release to the containnent for calculating the con-tainment second peak pressure and contailli~ent third peak pressure, respectively.
Response
A conservative calculation for the containment second peak pressure and containment third peak pressure is achieved by maximizing the mass and energy release to the containment atmosphere. Section 8.0 describes how that is achieved by overestimating the following:
- 1. Fission product decay heat and power coast-down heat
- 2. Core sensible heat
- 3. Hot metal sources
- 4. Break effluent during the Goolant blowdown, reflood, and post reflood periods.
In addition, the b~eak effluent is added in its entirety to the contain."I!ent atmosphere. This maximizes the energy input to the containment atmosphere.
'.['he water that cannot be supported as saturated vapor by the available energy in the containment atmosphere falls to the floor at the dew-point temperature *
.. ~* * -* - * .. - :**" ,. * * ~ , **"*:, .. * .. - -*r -~ *-*- _ "'- _,,......, __ .,,..._,7p *'-"**:--* .... *'-'" ..... * . - .~ .. - ........ ~..,,._~..-.-*_-:;,-.. -.-,,_-:-*,............ -*:.~,;**"'- . . -..: ..*1-*,-*,,,*.~*w**
. *~
e e
~estion 8.2 A discussion of the conservatism in the model with respect to minimizing the mass and energy release to the contairunent for calculating the minimum available NPSii for RS and LHSI pumps.
Response
A conservative calculation for available NPSH for the RS and LHSI pumps is achieved by minimizing the mass and energy release to the containment atmosphere and maximizing the mass and energy release to the containment floor. Thus the containment pressure is underestimated and the containment floor water vapor pressure is overestimated. Since containment pressure is a positive component of the NPSHA equation and the floor irater vapor pressure is a negative component, a conservative calculation of NPSH results.
Minimizing the mass and energ,J release to the containment atmosphere and
- maximizing the wass and energy release to the containment floor is achieved through the following assumptions:
- 1. The conservatively high mass and energy release rates calculated as described in Section 8.0 2o The break effluent experiences a pressure flash, i.e., it expands at constant enthalpy to the containment total pressure. The saturated vapor component goes to the containment atmosphere and the saturated liquid component goes-to.the sump unmixed with the co~tainment atmos-phere. This assumption conservatively neglects the evaporation cooling the liquid co:nponent most certainly will realize.
J. For a break in the cold leg, it is assumed that there is complete mixing at the break between the break effluent as calculated in Section 8.0 (either steam or two-phase) and the safety injection spillage (liquid). Thus, some or all of the steam is quenched resulting in more mass*and .energy release to the floor and less to the atmosphere.