ML20203K848
| ML20203K848 | |
| Person / Time | |
|---|---|
| Site: | Vogtle  |
| Issue date: | 04/08/1986 |
| From: | Brigdon R, Swartzwelder J GEORGIA POWER CO. |
| To: | |
| Shared Package | |
| ML20203K798 | List: |
| References | |
| LO-LP-36101-, LO-LP-36101-00, NUDOCS 8608210378 | |
| Download: ML20203K848 (28) | |
Text
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TRAINING MATERIAL ROUTING go _ a w t o - oo.c.
3 Eth %ER g MATER
/ REVISED MATERIAL A
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AUTH0 REVISOR,. 1. i ;^ '
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DATE NEEDED BY
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Georgia Power "T-,-.w
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so VOGTLE ELECTRIC GENERATING
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^
TRAINING LESSON PLA q.M.. E d 1,
J MCD: CORE COOLING MECHANISMS TITLE; NUMBER:
.S.n Ln-073
- - e-we o. - u -c_
PROGRAM:
LICENSED OPERATOR M 4 i REVISION:
7O AUTHOR:
R. BRIGDON V
DATE:
3/25/8' APPROVED:
OjQh J TE:
g--g-. g INSTRUCTORGidDELINES: d STUDENT MATERIALS 4 w.e_3. ion-co-c. g g HANDOUT: 4A-He-ONM00t, "MCD CORE COOLING MECHANISMS" MMS 9 ' o.g PasPost ht r-we best ce casse.nuta g e-m n g ECCS ACCEPTANCE CRITERIA (10CFR50.46)
.5 *= 07" __.
s.e,Tp. w ioe " M ot 8!*
V SR-9-GM H IN CONTAINMENT vs. % METAL-WATER REACTION 2
i.o - t e -w i.. -*e 4, S: '"" 070 g SOLUBILITT LIMITS OF BORIC ACTD SOLUTIONS co -re-m o i - eo4 SR=TP-6Ni g BASIC NATURAL CIRCULATION m -,,... o i -.
00-H vi,, g TDH RELATIONSHIP TO PRESSURE & TEMPERATURE to -Te-m e i.
~
- r -07^, Q MODE 1: LOCA CORE COOLING MECHANICS w-Tv-us o n.,a i.
-05'""070,],
MODE 2: LOCA CORE COOLING MECHANICS co-rP-nio o.-c MODE 3: LOCA CORE COOLING MECHANICS (3)
.. q s.o-fr-w eos -..-6 "I ** M7? ^^^ MODE 3: LOCA CORE COOLING MECHANICS -
MC STOPS (4)
LO-Tf~h*s m et-TP=678' p MODE 3/4: LOCA CORE COOLING MECHANICS gh. to Qpea a w-c7, OM ^Q MODE 4: LOCA CORE 400te MECHANIC c es c.
, l E= Le = W 1 bt*l** 0-L 4 G.o 1-N MDDE 4: LOCA CORE MECHANICS (LOOP 083 Aq 74 SEAL baAIN)
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.,j.S' te-1f->M o t => <
d D-070-Oli MODE SA: LOCA CORE COOLING MECHANICS w -7f-ueor- $1 CaalaL.
l 85-D-976-0t4 MODE 55: LOCA CORE 40GI,Y-MECHANICS w-it-sses sm*?
OR-4P-GM REFLUXING DETAILS STEAM FLOW
(
MASTER COPY 1
mc _ o. s,.... - - c.
sR-he-on l.
PURPOSE STATEMENT:
THIS LESSON DESCRIBES VARIOUS CORE COOLING MECHANISMS AVAILABLE DURING AND AFTER AN ACCIDENT. TH ENT WILL LEARN THE ECCS ACCEPTANCE CRITERIA AND ITS RELATIONSHIP TO CORE
- COOLING, AFFECTING AND ENHANCING NATURAL CIRCULATION FLOW, AND THE MODES OF NATURAL i CIRCULATION FLOW DURING SMALL BREAK LOCA's.
II.
LIST OF OBJECTIVES:
1.
State the ECCS acceptance criteria of 10CFR50.46 and describe how each protects against core damage.
2.
Describe how exceeding the cladding oxidation limits, peak clad temperature and H 2
generation can result in degraded of core cooling.
3.
Describe what is meant by boron precipitation and how excessive core borom concentration can result in degraded core ecoling.
4.
Describe, in general terms, how boron precipitation problems can be minimised~during a LOCA.
5.
State the three conditions requiied to establish natural circulation flow.
6.
Describe thermal driving head (TDH).
7.
State the relationship between Mass flow rate (M), core delta T (AT), and heat transfer rate (i}) during natyrel circulation flow.
i 8.
List and Describe three factors that can be used by the operator to enhance natural l
circulation flow.
9.
List and Describe three separate indications of NC flow.
- 10. Describe how the formation of noncondensible gasses and/or steam can result in degradation of NC flow.
- 11. List several.(at?
sources of noncondensible gasses in the RCS.
Describe _phat,is. g -
" pool boiling".
12.
3,w Describe whifEC g s
' during small break LOCAs.
13.
- 14. Describe the modes of NC flow during small break LOCA.
2
.: 7. - --- 2
~.&
e - ct-3s,c, _co. c
" " "^
t
REFERENCES:
1.
" Poet Accident Cooling". Mitigating Core Damage, Chapter 1. Westinghouse i
Electric Corporation, Pittsburg, PA, Revision 1.
2.
"LOCA Core Cooling Mechanisms" Small Break LOCA's Chapter 2, Mitigating Core Damage. Westinghouse Electric Corporation, Pittsburg, PA, Revision 1.
3.
" Modes of Nataral Circulation", Mitigating Reactor Core Damage Chapter 4, GeneralPhysicsCorporation,198(
l on M4it (%Qb 2% (g d g h M4g (ggg N,
boa bob *I k b(4 W L SM S }
h Ms'ic 1 Cam. tem. si eq $p ee w
O
%t e
4 m
T=
v e
.c,
I l
3
i.e - LP - E c o - Cc - C-
.a-se es Ill.
LESSON OUTLINE:
NOTES I.
INTRODUCTION A.
This lesson discusses various aspects of core cooling in a post accident situation.
1.
The operator should know what cooling mechanisms are available (for should be available) during accident conditions.
2.
Implicit in this an understanding of the ECCS Acceptance Criteria, natural circulation, and factors affecting the effectiveness at core cooling during an accident.
06(ch M B.
This lesson also covers the various modes of natural circulation that occur during a small break LOCA.
II.
PRESENTATION A.
ECCS Acceptance Criteria and Post-Accident Cooling Relationships.
P Aw e-4L 1.
To ensure core integrity for all postulated LOCA's
.. Ni, =001-the ECCS must meet the, acceptance criteria of 10CFR50.46.
os3ecnv0 l a.
The primary goal of the criteria is to prevent core failure.
b.
Of all the limits, the Peak Clad Temperature Limit of 2200*F essentially determines the success with which the other four criteria are satisfied.
2.
Cladding oxidation m
W - ce2.
a.
Cladding oxidation is. limited to 17 percent of che original cladding thickness.
by.
1 react with water and steam to the following chemical reaction:
.e.
Er02 + 2R2
- R***
' G;'Y${ &
-- I7 'The rate of oxidation is strongly dependent on the clad temperature; as clad temperature increases the rate of oxidation increases exponentially.
4
?
l
~
l l*
i.o. i.e 54.o l - oc. c, :
= :.7,,8
' ~
Ill.
LESSON OUTLINE:
NOTES 2)
Clad oxidation results, a)
At normal operating temperatures, the accumulated oxidation is inconsequential.
b) - At greater than 1800*F, the oxidation rate becomes significant.
Example: At 2000*F, the 17 percent limit can be reached in I hour.
c)
At greater than 2200'F, the oxidation rate accelerates very rapidly.
3)
Note that heat is one of the products of the Zr-H O reaction:
2 a)
In situations where the oxidation rate is already high due to the internal heat source (fuel heat),
the heat produced by the reaction can result in the further elevation of clad, ding temperature.
b)
As cladding temperature approaches 2800*F, enough heat is produced by the Zr-E 0 reaction to keep the 2
ireactiongoing.
c)
Excessive clad oxidation can result in loss of clad strength and o%16cnM Z.
ductility.
(1) In the event of a LOCA, the fuel rods could undergo a severe thermal shock as the cold safety injection water p-comes in contact with them.
- 2) The resultant embrittlement g
could be so severe, that portions of the core could 7t collapse under its own weight.
(3) The resulting core rearrangement could result in:
(a) Loss of long term coolability.
5 l
t :
a..e - s.... - so - c.
M !.E-076
~
lil.
LESSON OUTLINE:
NOTES (b) Adverse change in the local core flux distri-bution.
(c) Reduction in the shutdown margin.
2.
Hydrogen Generation elH 9-478-001 W ccA a.
Hydrogen Generation is limited to less than 1 percent of the theorecical volume of H that 2
could be produced if all of the Zircaloy in the core were to oxidize.
b.
Sources of H in accident situation.
2 1)
Major source would be the Zr-H O reaction 2
previously discussed.
a)
There is approximately 45,000 lbs of Zircaloy in the Vogtle core;.
potential of 360,000 ft* of H2 *t STP.
b)
If rele'ased tio coatainment, assuming 6-containment free voltme is 2.5x10 s
fe, the dry air concentration would 48-38-076-002 be approximately 15 percent.
Te_goos Fol'owing a LOCA, by taking into c) 1 account the steam in containment, the concentration would decrease to about 10 percent.
l d)
Since lower f1semability limit of H2 is about 4 percent, excessive H generationwouldbeasignificadt
~
ard.
t
,,.y,,
n
[other potential sources of H2 ****
a t
~
,)sdiolysis of primary coolanc in the l4 Mp.Y.CS and the containment sump.
B b)
Expansion of the H Portion of the 2
pressuriser bubble with subsequent release to the containment.
c)
Release of dissolved H., from the coolant as the RCS is 8epressurized.
6
u. J - w. o e - c a., - c.
a-a-m Ill.
LESSON OUTLINE:
NOTES 3)
Other potential sources of H in g
containment a)
High temperature metal-water reaction l
between stainless steel and water.
l (1) Potentially more severe than Zr-H O reaction from stand-2 t
I point of energy produced per l
pound of metal reacted and the amount of stainless steel in the reactor.
(2) However several factors make this source of H i"*i *ifi""*t*
E 2
(a) Amour.t of gas produced is only significant when the steel particles are very small.
(b) Steel temperature will i
be significantly lower
~
~
than Zr since the steel.
is somewhat removed from the core.
b)
Aluminum inside containment
(
(1) Reacts with both water and NaOH (2) These reactions can produce significant enounts of H2 per poemd of Al==4=== at relatively low temperatures therefore containment ale =minne content is limited
. N, i ta less than 2000 lbs.
.y2,
- e,. f
-. (3) Aluminum components typically
'
- 1' %'%g found in containment.
L.+ #3
' ' Y. ).7 g k (a) Ele::trical connectors and related equipment.
(b) Miscellaneous components !
in large cosiponents such ~
as fans.
l I
I i
7 l
i
e.c -s P - 3 6 e c t-c c ~c N 76-i lil.
LESSON OUTLINE:
NOTES c.
Hazards associated with H2 1)
If hydrogen is released to atmosphere if can pose an explosive hazard.
en 'r=-e?S-cer r P. **C.
2)
In the RCS, H accumulation does not
- 3 9
exist as an eRplosive hazard, however the H accumulation can:
2 a)
Displace coolant from the core causing overheating.
(A i
b)
Collect in the SG)f-tubes impeding O S M C 7-natural circulation heat removal.
3.
Boron Frecipitation and Long-Term Core Effects a.
The fourth and fifth acceptance criteria 88 To n7s_nni require the ECCS to maintain the core in a TV-oc4 long term, coolable geometry.
1)
The limits do not specify what the g
06 @
geometry must be; only that it be coolable.
2)
For all practical purposes, the designers ensure that no charge in geometry is predicted for any FSAE accident. (It is
,~
extremely difficult to calculate ultimate core coolability for other than design geometry).
3)
If first three ECC3 acceptance criteria are maintained core geometry can usually be assured to remain in a coolable configuration.
b.
In the event of a large LOCA, the primary
~
ling may be boiling and ambient
, p.
. 'g.
1*tThig a boron concentration in the
'g to increase for the following
- f'd Q,3 1)
Borie Acid has a low volatility.
2)
As core cooling progresses, the lower et=TP=0P9=002 tamperatures caused by ECCS* operation 9,
will result in a decrease in borie oost seid solubility.
i l
8
.Z
a-J - s<a rc n-cc - c.
SE L" ^70 111.
LESSON OUTLINE:
NOTES c.
If no action is taken, boron precipitation will most likely start as core temperatures decrease.
1)
This buildup of boron on fuel rod surfaces can result in:
a)
A reduction in the he t transfer p ability of the clad to the added layers of boron.
b)
Clogging of flow passages between fuel rods.
2)
Both of these will result in a decrease in heat transfer from the fuel rods, and the fuel will star: to heatup.
a)
Some of the boron vill return to solution as heatup progresses.
b)
However, precipitation will exceed the amount of boron returning to solution.
c)
If heatup is allowed to persist, eventual core overheating and desage could result.
d.
To ensure boror precipitation does not become a problem for lang-term cooling, boron concentration in tht; core must be reduced g
44 at some point after acetdent initiation.
1)
This is accomplished by shifting to hot-les recirculation, 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> after t.cident initiation.
- 2).
}r[]hg,-.?
, g Recirculation flushes boron out
..b core via the break, improving
- fly, al capability.
w.
.a Yld-legbreaks.HotLesRacirculation 4.,' ) g Q b added benefit of quenchingT g Tete any F
core regions.
f B.
Natural Circulation 1.
Introduction 4
a.
At say time during normal operations, RCP's maybelostfordecayheat. removal [.
9
.- i
w. P-36 set-cc-c 111.
LESSON OUTLINE:
NOTES b.
The operator must understand the concept of natural circulation, how its existence (or lack of it) can be detected, and what phenomena can promote or degrade its effectiveness.
1)
This becomes even more inportant in an accident situation where continuous core cooling is essential to prevent further degradation on plant status.
2)
In the following discussion, each of the above concepts will be discussed with the emphasis on the practical rather than theoretical aspects of establishing, maintaining, and enhancing natural circulation.
3)
In addition, the effects of non-condensible gas accumulation on natural circulation effectiveness will be discussed.
2.
Natural Circulation Concept.
Conditions requifed for Natural Circulation N
a.
M -sen6 1)
There must be a heat source and a heat oor sink.
o6 Mat 6 2)
There must be a density difference.
3)
The heat sink must be elevated above the heat source.
b.
Thermal Driving Head (TDR) 1)
TDE = -g/g* (
,s) (i.e., TDE h T)
Os@A b the pressure differential due u
of fluids at different mW,
.s.
[h be shown that:
' hii8 i
l
- Me( tis ( T M ).
~
OW" 7 b)
M (k1/3 (Q M 8
c) kalT3/2 10 l
..c
)
a ~ < P - % a n -ce -c.
==m
'~
lil.
LESSON OUTLINE:
NOTES 3.
Factors Enhancing Natural Circulation Flow a.
Pressuriser Pressure greater or equal to g%, t, 6
2000 psia (1920 psig by E0P-19002-1) 1)
This ensures that the RCS is subcooled and void formation from steam and i
noncondensible gasses do not form.
2)
At100percentpower$isapproximately Te 620*F.
A pressure of greater than or equal to 2000 psia results in greater than or equal to 15'F subcooling at t2e core outlet.
b.
Pressuriser Level greater than or equal to 50 percent (25% by EOF 19002-1) 1).
This ensures adequate water inventory in the RCS to ensure ability to maintain pressurizar pressure control.
2)
Adequate lov'el coincident with adequate pressure ensures the pressurizer bubble is the only bubble in the BCS.
c.
SG level in the narrow range of at least one (greater than or equal to 4%)
1)
A heat sink is required for natural circulation.
2)
By maintaining SG 1evel above the top of the U-tubes of at least one S/G, i
the secondary heat sink is maintained.
4.
'Indicat Natural Circulation Heat Removal.
i
.; _ f t OW43 l
. "385 T full load Delta T.
i
- rI? %5
}
U-1)3 inum Delta T expected in NC should Q '4
~
- exceed the full load Delta T roximately 60'F.
2)
Typical NC Delta T should never exceed approximately 30'F and would decrease as core decay heat level decreases.
~
a)
Too high at Delta T indicates an interruption of NC cooling flow caused by excessive or rapid feeding of SC's.
11
o
. a,- 1..... -
m-tr= ors lil.
LESSON OUTLINE:
NOTES b)
No Delta T indicates no heat sink indicating a loss of SG inventory.
b.
RCS vse core exit thermocouple temperatures constant or decreasing.
1)
If WC is lost, core exit temperatures would rise due to the loss of heat sink.
2)
Keep in mind that when establishing NC, core exit temperatures must rise to help establish the TDH necessary to promote NC.
This is temporary and a decrease or leveling of the temperature should eventually occur.
c.
SG pressures are constant or decreasing at a rate equivalent to the rate of RCS temper-ature decrease with constant AFW flow.
1).
SG pressure would fall rapidly if NC flow stopped, due to excessive cooling f
by AFW with no heat input from the RCS.
SG level woEld al' o be seen to rise 2) s fairly rapidly.
5.
Rffect of Steam Accumulation in the Reactor Vessel Head.
a.
To ensure adeqeate subcooling during NC, a means of naintaining primary pressure must be established.
i 1)
Preferred means is 'ase of pressuriser heaters.
l 2)
Pressure may als'o be controlled by amans of a steam bubble in the RV
_: i ~,
- n
' 'a)f;@swaterincoreheats,itmay l'];
~# a'm ->r.vt. boil and the bubble may rise f e t) h. M" and collect in the head region.
1 J
St b)
This bubbles controls large density differencies and promotes N.C.
~
c)
As long as the steam space stays above the outlet nossles, NC flow will continue.
12
.,.,,.,n,-.,,,
w. s. n... - - c.
- - m Ill.
LESSON OUTLINE:
NOTES b.
There is concern that sufficient voiding could occur to regard NC flow.
1)
Remember that voids will rise to system highpoints.
a)
RV head b)
Pressurizer COM N
c)
SG U-Tube bundle 2)
Voids in the head or pressuriser are of no concern since they don't exist in the NC flowpath.
3)
For SG tube bundle voids, continued addition of AFW to keep the tube bundle covered should condense the steam on the primary side of the tubes, allowing NC to continue.
c.
Formation of a RV head steam bubble r' enders l
pressuriser level indication useless as j
means of evaluatiing RCS inventory.
1)
It is important that a RV head bubble not be allowed to form during cooldown unless deemed necessary by operating condition.
2)
To ensure a bubbla is not formed, a cooldown rate of 50*F in any hour should not be exceeded.
a)
This will allow the upper heat time to cool as t'ha RCS is slowly cooled preventing the Rf head from becoming
.a w.~,
hottest region in the system.
49 in mind that although an hr, rational limit of 50*F in any x
42 1;'
r exists, conditions could ist that could restrict the Idown rate to. a lower valve.
c)
On the other hand, if conditions warrant a acree rapid cooldown.
j the operator must realize that a i
bubble will be drawn in the head.
13
- - ~
~
i c -LP - 16 e o s - cc - c, lli.
LESSON OUTLINE:
NOTES (1) As long as indications of effective NC cooldown are maintained and as long as an effective heat sink is maintained the cooldown
$ycontinue.
(2)
If bubble expansion appears to degrade the effectiveness of core cooling by NC, the cooldown must be either slowed or stopped.
Any situation requiring a rapid cooldown should not be complicated by a Loss of Decay Heat Removal capability also.
6.
Available NC Driving Head 6
- -y_ s,.y.
a.
Curve shows available driving head as a 6
function of primary pressure and cold leg temperature.
1)
If flow is reduced the natural system response is to heatup so that Th approaches saturation (PlotassumesTh=T,,g) 2)
The increased Delta T results in a larger TDH.
b.
Example T = 300*F 1) teen excess. System Pressure = 800 psia driving head = 200 psf.
A.
M ae conditions of T and Pressure.
T I h/ h usuld actually be much*1ess than T,,g.
j ?
t impedance to flow is formed (col-
[i[4:di.n @he~TBE would increase in response to of noncondensibles in the U-tubes)
' ' ' ' t the hot leg heatup to overcome the impedient.
In this example 200 psf of*TDH would be' available before boiling would commence.
14
1 w -J. hs.c o -ecx ae=
ll1.
LESSON OUTLINE:
NOTES 7.
Non-condensable Gas Formation and Effects a.
During the TM1-2 accident, the presence of large volumes of noncondensible gasses in the RCS loops severly affected the ability of the plant to utilize NC cooldown.
As a direct result considerable study was conducted on the formation of nondensable i
gasses in a Westinghouse PWR.
l b.
SG are required to remove decay heat for a significant period of time following small break LOCA's.
1) 2" break - (must limiting as far as noncondensible formation is concerned)
- SG's required for 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.
2) 1" break - SG's required for 1 day.
After the times given, the combination of the break and ECCS flow remove all of the decay heat.
c.
Sources of Non-Condensible Gasses 665 cD IL 8 8 f
1)
Dissolution of H2 a)
As mass is lost through the break, RCS pressure decreases allowing dissolved H to be released from 2
the coolant i
b)
E., concentration may be as high e5 50 ce/kg.
2)
Radiolysis RCS coolant I concentration of
,radiolysisdayresultinthe sociation of water molecules.
0 2 p.252+O2 b)
Esey in mind that this effect is not considered unless coolant I concentrationexceedsapprod.Sec/kg.
3)
Pressuriser Vapor Space H i
2 a)
Approximately 600 sef of H2 I
15
.\\
w a.1,. -. - ~ - e
u-
III.
LESSON OUTLINE:
NOTES b)
Assumed to expand into the RCS after the P3r empties following the ideal gas laws.
4)
Zr-H O reaction 2
a)
- Amount of H,, produced dependent upon the level of core temperature.
5)
Accumulator are typically pressurized to 650 psig with N2.
b)
All or part of the N2 may be ad,(jected into the RCS during a LOCA.
6)
Fission Gasses and Helium a)
Assumed to be released only if a fuel rod bursts.
7)
Gasses from, Injection Water a)
Gas content is usually low.
~
b)
For analyses is usually assumed to be approximately 20 cc/kg of various noncondensibles.
d.
Effect of non-condensible gasses of heat removal capability 1)
Model F Steam generators have a primary side tube bundle volume of approximately 760 ft*, (approximately 100 fts in the tube bend region)
... d. 'p(
- s.,' 39 worst case small break LOCA,
'y, b ut of noncondensibles released
- M4he coolant would block approximately g = :..
eat of the available tube bundle AN."** %.A yath.
7.
u This does not seriously affect the NC flow, therefore core cooling will be maintained.
NOTE: For the worst case break considered the accumulator do not inject and no fuel rods burst.
16
a.s s....~. c.
SS '". 07" lil.
LESSON OUTLINE:
NOTES i
3)
For larger breaks where accumulators do inject, the potential will exist for a loss of NC flow i; re--'-te M2
- dad fo Ma. Accum.w,9
- 1=*ar in the SG tube bundle.
4)
It is important to -maintain adequate SG' level to condense any steam existing in the tube bundle region to minimize the size of any blockage.
5)
If blockage does occur, the TDH will increase as Th increases to force the blockage from the U-tubes.
C.
LOCA Core Cooling Mechanisms (Small Breaks) 1.
The physical layout of the RCS (SG above core) ensures the requirements for NC are met.
i s.
During LOCA, the flowing coolant may be subcooled, two-phase or steam vapor.
b.
Flow induced by high pressure coolant l
venting from a break may be an effective asens cooling but is not considered i
natural circulation flow.
2.
For large break LOCA's, the flow out of the
~
break, in conjunction with injection flow establishes effective core cooling.
3.
For small break LOCA's the primary means of heat t
removal is natural circulation.
l a.
" Pool Boiling" is the mechanisu 17 which o MEerUc. t"L heat is removed from the core.
1)
This heat transfer process is one where i
ively stagnant water undergoes
/MA.
and then moves upward away g-hot surface, carrying away
- ';c..,w. *.
^
W$*$.K.
fluid moves in to the take pla'ce of fluid.
b.
Because of the nature of ses11 breaks.
oGTEc d !- 0 Natural Circulation terminates during the i
~
accident.
l i
1)
The introduction rate of vapor into the SC from the core exceeds the rate of condensation in the U-tubes.
17
3 A~LP*344e1=W "C
-sa-a-we lil.
LESSON OUTLINE:
NOTES a)
If all the heat added to the coolant by the core were removed in the SG, the mass loss from the break would continuously lower RCS temperature and pressure.
b)
- Therefore the liquid mass in the U-tubes decreases over time.
2)
As mass lowers, the interface level existing between the liquid and saturated mixture rises up the hot les portion of the tubes.
a)
M en the interface transfers to the cold side of the tubes NC flow begins to degrade, b)
The high density fluid on the cold les side acting as the driving force for inlet flow diminishes
~
and inlet flow degrades along with outlet flow.
c)
Break f' low remains unchanged.
d) steam forms and expands at the tops of the U-tubes.
3)
Een NC terminates, the core will still be covered with a two-phase mixture of steam and water, assuring the core is being cooled.
a)
As long as liquid water exists sbove the top elevation, the core can et i
uncover, since water in the loo?s gravity drain to the low points the system, such as the core.
(
n
,T ual core uncovery will eventually 1>.'.'
cur because of the net loss of Ac~
water inventory through the.
?sg 'h esk.
w.A,.,
4)
It should be noted that NC is not needed for long term decay heat removal and core cooldown.
a)
On the larger small breaks, injection and break flow will eventually reach equilibrium.
18
. --.~
gi-6A 5 6s o i - M - f.,
I". I." 070 lli.
LESSON OUTLINE:
NOTES j
b)
Upper regions of the RCS are actually void of water-4.
The various modes of NC that occur during a small LOCA aust be understood to appreciate the core cooling mechanisms present.
These were described by Mr. Carl Michelson in his report, " Decay Heat Removal Associated with Recovery from a Small Breair. LOCA for CE System 80 PWR", May 1977.
b.
e 6
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19
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ll1.
LESSON OUTLINE:
NOTES MODE 1 - Decay Heat Removal by Natural Circulation with the Pressurizer Controlling Pressure S8 3D-434-006-TF- **-6, The break in the RCS occurs and water begins draining from ce>7 the peak. The reactor trips and safety injection begins.
The following events are taking place:
1.
The pressuriser steam bubble pressure is above the saturation pressure corresponding to the reactor vessel outlet plenum liquid temperature. The reactor coolant system is subcooled, and plant pressure is being cor' >11ed by the pressuriser.
2.
With flow out the break exceeding injection flow, mass is being lost from the system. The pressurizer begins to drain and consequently pressuriser level decreases.
3.
The pressuriser liquid begins flashing to stems as pressure drops below the saturation value e,orre-sponding to the current liquid temperature. This flashing process requires energy. Energy is
~
obtained from the stored thermal energy in the pressuriser liquid. Thus the pressuriser fluid begins cooling, exhibiting a natural tendency to remain close to saturation. The flashing process is 4
[~
insufficient to terminate the pressure drop caused by steam bubble expansion; consequently, pressure i
continues to decrease.
4.
The walls of the pressuriser are now hotter than the saturation temperaturs estating in the pressuriser.
Heat is transferred from the pressuriser walls to the steen, raising T above The steam bubble g
9 superhests.
5.
- us hons ether than the pressuriser metal, is-ompensate for thermal energy lost ing of water to steam. The re continues to drop rapidly.
M 6.
1, the pressuriser empties and e
trol pressure in the RCS. RCS pressure drops rapidly to the saturation pressure corresponding to the temperature in the outlet plenum of the reactor. Bubbles begin to form in the hot less and the core. Subcooling he's been lost.
~
20
Lo -LP - M,, t o t - oc - C.
111.
LESSDN OUTLINE:
NOTES 7.
Flow exists in the loops as a result of coasting down reactor coolant pumps, and the natural circulation caused by the density gradient between the steam generators and the core. Bubbles formed in the core and hot les are condensed as they flow through the steam generator tubes. The cold legs are still i
experiencing single phase flow.
8.
As the bubble begins to form in the core outlet, RCS pressure control switches from the pressuriser to the vessel outlet. Note that the pressuriser drains for all LOCAs unless the break is so small that injection prevents a.
its, or b.
the break is in the vapor space of the pressuriser.
In all cases, except the vapor space break in the top of the pressuriser, pressuriser level is a measure of the liquid mass in the RCS.
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111.
LESSON OUTLINE:
NOTES t
MODE 2 - Decay Heat Removal by Natural Circulation with the Reactor Vessel Controlling Pressure (Figure 2)
By the end of mode 1 RCS subcooling has been lost and a steam bubble has formed in the reactor vessel outlet plenum.
The following events occurs in mode 2:
1.
Mass is still being lost from the system since break flow exceeds injectior. flow. The reactor vessel outlet plenum liquid levels begins to decrease.
2.
Additional reactor vessel outlet plenum liquid is flashing to steam in an attempt to maintain constant pressure. This same effects escurs in the pres-surizer on an outsurge. Cota cooling is being provided by the formation of steam bubbles and their removal to the area above the core liquid level.
By the end of mode 2, considerable voiding has occurred in the hot legs and the vessel. As lonC as the steam generator secondary water level is being naintained by auxilisry feed.
steam bubbles can still be condensed at the steam geserator.
System mass is dropping throughout this mode, and by its and the vessel liquid level has dro p d to the top of the hot leg.
~
It is important to note that the core coutsiains a mixture of saturated liquid and v. spor, a large density gradioat exists between the hot and c.11d legs. The boiling process is
~
actually enhancing the driving head for natural circulation.
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LESSON OUTLINE:
NOTES MODE 3 - Decay Heat Removal During the Transition from Natural Circulation to Core Boiling SNor SE=IE=G38-009 TP $$$ or Fluid loss out the break is still in excess of injection flow.
7 Pressure is hoing maintained by the reactor vessel steam bubble The atable pressure is a result of the mismatch in thermal energy removal (i.e., vapor formation rate greater than vapor condensation rate) being exactly compensated for by the volume flow ste mismatch (i.e., break flow greater than safety injection flow). The steam generator safety valves are providing a heat sink. The pressure which exists is above the 1200 psia lif t pressure of the lowest setpoint safety valve.
i The following events are occurring:
l 1.
The reactor vessel level, which reached the top of the hot les by the end of mode 2, is now below the top of the hot leg. The amount of steam vapor in the U tubes is increasing.
2.
The steam generators are able to condense a portion of the steam which reaches them. For the moment l
natural circulation continues.
3.
As system pressure drops to the 1200 pais steam i
generator safety valve lift pressure, the primary side water temperature approaches the secondary j*
side temperature. The ability of the staam generator to condense steam on the primary side of the U tubes drop drastically, further increasing the vapor content on the U tubes.
4.
As steam continues to form in the U tubes, the driving head for natural circulation decreases.
5.
The steam mixture continues to flow to the steam generator where some of it' is condensed. This
'contiames the generator is completely drained.
- w. pg 6.
e in the loop seal between the steam reactor coolant pump.
7.
in the core now has no place to fleefvet tasate the loop liquid seal blocks its passage to the break. The steam space above the core begins to pressurise.
8.
Vessel level and loop seal level are'still decreasing.
9.
Natural circulation has stopped.
It is important to note that the core is still covered.
l 23 i
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sa-hr-m lli.
LESSON OUTLINE:
NOTES MODE 4 - Decay Heat Removal by Core Boiling 1.
The reactor vessel level continues to decrease due
" " 07" 010 to fluid loss through the break.
SAN 43N3-023 0M 1
2.
Boiling is taking place in the core. The boiling h
process is removing energy from the core and trans-g o,,
I porting it to the steam bubble above the core.
(f-Olb 3.
Any liquid which is produced from condensation inside p
og I
the steam generator tubes is returning to the core via gravity counterflow along the bottom of each l
partially filled hot les pipe. This phenomenon is called reflux flow. The cold les side of the U tubes is draining to the loop seal.
4.
Eventually, the decay heat level drops to the point where the steam generator safety valves are no longer needed as a heat sink. The exact point in time at which this occurs is dependent on the decay heat level and the break size. The larger the break, the l
sooner this event will happen. As soon as the steam l
generator saturation p,ressure drops below the safety valve setpoint, the safety valves go shut. Decay heat is then removed only by heat lose through the break sad by heat loss to the environment.
5.
Plant pressure is now controlled soley by the steam l
bubble which exists above the core. As the decay l
heat level drops without a corresponding drop in j
heat removal, the system departure decreases.
6.
As system pressure drops, the driving force for flow out the break decreases. At the same time, the lower system pressure allows injection flow to increase. This occurs when decay heat level falls to within the capabilities of the emergency core
' cooling By the end k heat is being removed by the break.
s have become a heat source rather than a hee '.
1 g;
It should alesi t the operator can cause system -
i pressure to drop below steam generator safety valve lift pressure earlier in the transient by using the steam dump.
This action is not necessary to protect the cor,s.
l l
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' 24 i
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sa-ut-o;te
~
~
~
lli.
LESSON OUTLINE:
NOTES MODE 5 - Decay Heat Removal During Transition from Core c"-T* ^7" 013-Boiling to Natural Circulation Mode
'" '"" 070 014' W
1.
With injection flow exceeding break flow, reactor u,om vessel level continues to increase to the top of the 4 gg hot les nossles.
2.
Prior to this point, a steam vent path existed from the top of the reactor vessel to the break. The venting process provided efficient heat removal and at the same time prevented pressure build-up in the i
reactor vessel outlet plenuni.
3.
As vessel level increases toward the top of the hot les nozzles, slugs of water begin to flow down the hot 1ess to the steam generators.
4.
The steam generator inlets become flooded by the slug flow, disrupting the flow of steam through the steam generator tubes to the break.
5.
The disruption of steam flow to the break reduces heat removal capability. The reactor core exit temperature and the corresponding reactor vessel
~
steam bubble pressure starts to increase in response to the disruption.
6.
As steam pressure increases, the two-phase fluid existing in the core compresses and the reactor vessel mixture level dropa.
7.
When level drops to the tops of the hot les nossles, slug flow through th hot legs stops.
8.
The compression affect on the core level is trans-mitted to the dowacomer. back throush the cold leg, and up into the cold-leg s$de of the U tubes.
9.
"If f
ter had been initially present in i
and if the rate of vapor con-hot leg side of the U tubes can 6(
rate of vapor introduction,
. 8 can lead to a refilling of
. Skis occurs, the density gradients for astural' circulation will be estab-lished. If large quantities of noncondensable gases, particularly dissolved hydrogen and oxygen,
released during the depressurisation, have accumu -
j isted in the steam generator tubes, the reestablish-l ment of natural circulation may be prevented. More will be said about noncondensable gases later.
25
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~
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-c - s e. 5e, c o - ce - c.
d' H Th A *R h_
Ill.
LESSON OUTLINE:
NOTES
- 10. If natural circulation is not reestablished (as a result of noncondensable gases), coolant system pressure continues to increase. Eventually, the setpoints of the reactor coolant PORVs is reached.
- 11. As water is expelled from the reactor coolant system by the PORVs (remember that the high system pressure is above the shutoff head of the safety injection pumps). core level drops to the point where the level is below the top of the hot leg nozzles.
12.
Once this happens, steam can flow from the vessel outlet plenum to the steam generators where it can be condensed. This reflux cooling adequately cools the core and once again lowers core temperature and pressure.
Mode 5 is a very complex mode so several points should be reviewed before leaving it.
First, mode 5 is a stable mode.
Decay heat is being removed primarily by the break, and the core is remaining. covered. The mode can last several hours.
For large breaks, it may in fact be the final mode since no refill is expected if the break is large. If this is the case, an equilibrium is establis,hed where the following occurs:
1.
Break flow equals injection flow
~
2.
Decay heat production equals decay heat removal by the break.
Secondly, suppose that little heat can be removed from the system due to noncondensable gas formation in the U tubes.
As stated above, the system would repressurize to the PORY setpoint. With the FORTS open, the system would drain again. The sequence of modes 3. 4 and 5 would then be repeated only less severely. This cannot and will not jeo-pardise core cooling as long as safety injection is not terminated. The cosa essentially covered at all times. ' 9Wy,,v g l
. : ty;4
-l
('j,fff ny
- c
~t n
26 l
l 4
f-s uo -up-ssa c t -oc - c \\
2: tr-:::
lil.
LESSON OUTLINE:
NOTES HODE 6 -- Decay Heat Removal af ter Reestablishing Natural Circulation with the Reactor Vessel Controlling Pressure Assume the natural circulation has been reestablished by the end of mode 5.
The reactor vessel level will be above the hot legs and the steam generator tubes will be full of water.
Also assume that the break under discussion is small enough to allow refill to occur.
If it is not, then mode 5 will be the final mode of the system.
1.
Reactor vessel level continues to increase as injection flow exceeds break flow. The size of the vessel plenum steam bubble is thus decreasing.
2.
The pressurizer surge line begins to fill with liquid. The pressurizer contains a trapped steam bubble which is initially in equilibrium with the reactor vessel steam bubble.
3.
The pressurizer steam bubble condenses as stesa is condensed on the walls of the pressuriser due to heat losses to the environment.
j 4.
The pressurizer eventu' ally becomes filled with liquid.
As system pressure and pressurizer level increase, eventually the point is reached where safety injection could be termi.
nated according to the termination criteria. In actuality, injection flow could be adjusted to reestablish the steam bubble in the pressuriser. The core could then be cooled by fully developed, subcooled natural circulation with the pressurizer controlling pressure. A plant cooldown could then be conducted to begin preparations for repairing the break.
III. SUM ARY l
Review Leesom ha },
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^.jfgx; 27