DYNAMICAL TWO-PHASE FLOW ANALYSIS
5.6 Comparisons Between Base Case Transient and Some Variations
5.6.5 Reactor Coolant Pump Not Tripped
A calculation was made without the tripping the pump, in order to see its effect on the break mass flow rate and the pressure in the system. The pump keeps the water level in the fuel elements high, and, therefore, water is leaking out for a longer period than in the base case.
This means that the pressure, and the mass flow rate is kept at a higher level. If the pumps would have been kept on during the incident, the water level in the reactor vessel would become lower (compared to the base case) after the equilibrium is reached between the containment and reactor vessel.
In any case, the pump could not be operated during the whole transient, because of the voiding, and subsequent cavitation. Furthermore, loss of off-site power is assumed for this hypothetical scenario, and there were not enough power generated by the diesel generators to run the main coolant circulation pumps.
5.7 Discussion
The RELAP5/MOD3.1 code has wide range of validation with experiments, and is gen-erally a reliable code. In the present application, however, with a complicated break-flow geometry having very thin pipes with an hydraulic diameters between 1.5 - 51 mm, not much validation work has been reported. The question whether two phase critical flow in small pipes having large pressure drop at high two-phase flow velocities is well-predicted by RELAP5, might have an impact on the results. Validation of the critical flow model used can be found in the manual of RELAP5/MOD3.1. The diameters in those experi-ments are not as small as in this particular model.
The modeling of the location of the break flow geometry inlet in the RELAP5/MOD3.1 model of Agesta is a source of uncertainty, which has a considerable impact on the final collapsed water level in the reactor vessel. Some analyses of the impact of the break location has been made in this report and it indicates varying break mass flow rates for different break locations. The junction, which connects the top of the moderator with the top of the fuel channel has an impact on the mass flow rate of the break as well, because of its equalizing effect on the pressure and water level for the moderator and for the coolant in the fuel channel.
The base case may not be the most realistic course of events in this incident. At the time of the incident, the plant was flooded with water, and the control system was not in perfect functioning mode. The time duration between the hypothetical break and the SCRAM are assumed to be short.
After the flow ceases to be critical, break flow can be calculated by the TEENAGE model described in Chapter 4 of this report. This calculation has been performed in order to estimate the progression of the transient after 6000 s. Of main interest is the time elapsed until pressure equilibrium is reached between the reactor vessel and the containment, and the amount of water that has escaped into the containment by that time. The collapsed level in the vessel after the blowdown can then be calculated.
59
Table 5.1: Heavy and Light Water Properties.
I Temperature=100 (°C) |
H2O
| Temperature=200 (°C) |
H2O j Temperature=240 (°C) |
H2O
Table 5.2: Primary system condition in steady state.
Mass flow rates (kg/s) Part of the plant
Main Pump Fuel element
Junction 315 (see Figure 5.1)
According to [1] | Steady state run 1200.0
Part of the plant Main Piping (225) Main Piping (275)
SG-tubes Moderator Central Element
According to [1] | Steady state run 7.0
Part of the plant Bottom, inlet Water Spreader (in) Fuel Elements (out)
Moderator (top) Moderator (bottom)
Bottom, outlet Pump, inlet Pump, outlet
According to [1] | Steady state run 36.1
410tmdpvol
405 trpvlv (SRV)
761 tmdpvol
699 snglvol
698 tipvlv
605 trpvlv
Figure 5.1: Nodalization of Agesta for RELAP5/MOD3.
61
Table 5.3: Steady State Conditions of the Primary and Secondary Side.
J Reactor Power (MW): 65.0 ||
Pressures (bar) Primary side (Prcssurizer)
34.0 Total Mass (kg)
Primary side 63,200
Secondary side 33,100 Mass flow rates (kg/s)
Primary side Part of the plant Main Pump Part of the plant
Feed water in Steam out
Boiling region (Bottom) Downcomer
Part of the plant Main Piping (275) Main Piping (225) SG-tubes Part of the plant
Feed water in Steam out
Boiling region (Bottom) Boiling region (Top) Steam dome Part of the plant Bottom, inlet Fuel Elements (out) Moderator Part of the plant
Feed water in Downcomer (top) Boiling region (bottom) Separator
Part of the plant | Steady state Bottom, inlet
Water Spreader Fuel Elements (out) Moderator (top) Part of the plant
Feed water in Downcomer (top) Boiling region (bottom) Separator
RELAP5/AGESTA
Pressures during blowdown
RELAP5/AGESTA
Velocities in break flow geometry.
20
10
"~™'™ Pressurizcr
——- Steam generator
—-— Containment
400
j
Sonic Liquid Gas
n iT
2000 4000 6000 0 2000
Time (s) Time(i)
Figure 5.2: RELAP5: Pressure during the Figure 5.4: RELAP5: Velocities in the blowdown. break flow geometry.
RELAP5/AGESTA
Mass flow rate of the Mowdown
I M«a flow rate!
1.0
0J
~ 0.6
RELAP5/AGESTA
Volumetric void fraction in break flow geometry.
2000 4000 Time (s)
0.0
- M e t
• Outlet
2000 4000 Time (s)
Figure 5.3: RELAP5: Mass flow rate of the Figure 5.5: RELAP5: Void fraction in blowdown. break flow geometry.
63
RELAP5/AGESTA
Pressure In break flow geometry
30
k
\l —
i Inlet Outlet
—-— Containment
— . 550
RELAP5/AGESTA
Cladding temperatures
450
2000
Timed) Tuned)
Figure 5.6: RELAP5: Pressure in break Figure 5.8: RELAP5: Cladding Tempera-flow geometry. tures
Temperatures in the steam generator. RELAP5/AGESTA
Cladding temperatures
/
1 1 _J
Primary side. Inlet Primary side, outlet Secondary side
—— - Top cladding Top (water added) Top-2
Top -1 (water added)
500 1000 Timed)
1500 2000 4000
Time (s)
Figure 5.7: RELAP5: Primary and Sec- Figure 5.9: Cladding temperature when wa-ondary Steam Generator Temperatures ter ls added.
RELAP5/AGESTA
Break location: Pressure in Prcssurizer. RELAP5/AGESTA
Pressures in Break flow geometry, 5 inch break
250 500 750 1250
Figure 5.10: RELAP5: Comparison of pres- _,. _ ,_
.,, ,.a , , , , ,. Figure 5.12:
sure with different break locations. , ,
and base case
Timed)
: Pressure of 5 break
to
RELAP5/AGESTA
Break location: Comparison of massflow.
RELAP5/AGESTA
Comparison between 5 inch break and base case.
r.mew
Figure 5.11: RELAP5: Mass flow rates with
different break locations. Figure 5.13: RELAP5: Mass flow rates of 5" break and base case
65
RELAP5/AGESTA
Mass flow rates.
Time (5)
RELAP5flTEENAGE/AGESTA
Pressure during blowdown (bar)
Pressuriier (TEENAGE Pressuriier (HELAP5) Containment (RELAPS)
— — - Containment (TEENAC i)
4000 8000
Time(s)
12000 I6OOO
Figure 5.14: RELAP5: Mass flow rates with Figure 5.16: Extension of RELAP5: Pres-and without junction 319 sures
RELAP5/AGESTA
Mass flow rates.
30.0
25.0
20.0
With31»
Without 319
\
1000.0 Time (s)
5 "
. C
30.0
20.0
10.0
RELAP5/TEENAGE/AGESTA
Mass flow rate duriug blowdown (bar)
TEENAGE RELAPS
8000.0 Time (s)
12000.0 16OO0.0
Figure 5.15: RELAP5: Pressures with and Figure 5.17: Extension of RELAP5: Break without junction 319 mass flow rates
Chapter 6 SUMMARY
In order to get an overall picture of the hypothetical accident, the different parts of the analysis can be combined. The RELAP5/MOD3.1 calculation will serve as the best estimate calculation for the initial phase of the transient, when two phase critical flow prevalent, and heat transfer from secondary side to primary side occurs in the steam generator. When the break flow no longer is restricted by critical flow, the model in Chapter 4 (TEENAGE) can be used. This model is used in order to estimate the time elapsed, and the level in the reactor vessel, until pressure equilibrium is reached between the reactor vessel and the containment. After this, coolant boil-off takes place, which is a very slow process; especially in this case, where the volume ratio of the moderator and fuel is very large, and the power of the core is relatively low.
After the first 6000 seconds of the transient, the normalized collapsed level of the water in the reactor tank is 0.76 (where 1.0 denotes the top of the core, and 0.0 denotes the bottom). After 16,000 seconds, when the pressure in the containment and the vessel are equal, the level is 0.73. The level is between 0.44 - 0.52, 24 hours later, respectively, for the decay heat levels of 1.0% and 0.67%. The time to reach a normalized level of 0.1 is 87.2 hours (3.6 days) and 128 hours (5.3 days) depending on decay heat level.
We believe that the system for filling and emptying the primary side of its water at low pressure could be functioning with an external light water source. Therefore, it is highly unlikely that Zircaloy oxidation, and subsequent temperature increase would take place in this incident.
Figures 6.1 and 6.2, show the collapsed level as a function of time after the break.
67
PRIMARY SYSTEM LEVELS
Time 0 s.
Time 2980 s.
Level = 0.86
Time 6000 s.
Level = 0.76
Time 16000 s Level = 0.73
Figure 6.1: Collapsed level in Agesta before boil-off.
Boil-off: Collapsed level.
0.0
DK>1.0%
D K - 0 . 6 7 *
100000 200000 300000 400000 500000 Time after break (s)
Figure 6.2: Boil-off: Collapsed level.
Bibliography
[1] B. Lilliehook; Agesta Kraftvdrmeverk, Sammanstallning av Tekniska Data och Beskrivningar mm For Reaktordelen; AB Atomenergi; Stockholm; 1964. (In Swedish) [2] B. McHugh et al.; The Agesta Nuclear Power Station, a Staff Report by AB
Atom-energi; AB AtomAtom-energi; Stockholm; 1964.
[3] K. Karlsson; Vdrmetekniska Berdkningar for RS/ADAMs Katastrofkylanldggning;
PM A-49/59; Statens Vattenfallsverk; Stockholm; 21 Oct 1959. (In Swedish)
[4] W. Hasling; Agesta Kraftvdrmeverk Nodkylsystem P214 Funktionsanalys; PM V-79/62; Staten Vattenfallsverk; Stockholm; 28 May 1962. (In Swedish)
[5] L. 0 . Wredberg; R3/Adam — Katastrofkylanldggning; PM A-B5/59; Statens Vat-tenfallsverk; Stockholm; 2 Oct 1959. (In Swedish)
[6] H. Hoglund; Forslag Till Ombyggnad av Sprinklersystemet for Reaktortank; Arbet-srapport 108/1968 (reg no. 8242); Vattenfall; Stockholm; 14 June 1968. (In Swedish) [7] E. Ericsson et al.; Agesta Kraftvdrmeverk — Haveri pa kylsystem P210 den 1/5
1969; Arbetsrapport 156/1969; Vattenfall; Stockholm;12 May 1969. (In Swedish) [8] K. E. Sandstedt et al.; Agesta kraftvdrmeverk — Felfunktioner och skador pa
elutrustning i samband med haveri pa kylsystem P210 den 1/5 1969; Arbetsrapport 158/1969; Vattenfall; Stockholm; 30 May 1969. (In Swedish)
[9] AB Atomenergi; Beskrivning av Reaktorstommen for Agesta Kraftvdrmeverk; Arbet-srapport R3-198; AB Atomenergi; Stockholm; 1963. (In Swedish)
[10] K Persson et al.; Agesta kraftvdrmeverk — Arsrapport for tiden 1 januari - 31 december 1969; Arbetsrapport 207/1970 Vattenfall; Stockholm; 1970. (In Swedish) [11] P. Osterman; Svdr Reaktorolycka i Agesta; Dagens Nyheter, 13th April 1993. (In
Swedish)
[12] T. Okkonen; A Synthesis Of Hydrogen Behaviour In Severe Reactor Accidents; Re-port STUK-YTO-TR-59; STUK (Finnish Center For Radiation and Nuclear Safety);
1993; Helsinki, Finland.
69
[13] D. F. Fletcher, B. D. Turland, and S. P. A. Lawrence; A Review of Hydrogen Pro-ductions During Melt/Water Interaction in LWRs; Nuclear Safety, Vol 33, No 4, pp 514-534; 1992;
[14] W.J. Garland and B.J. Hand; Simple Functions for the Fast Approximation of Light Water Thermodynamic Properties; Nuclear Engineering and Design 113, 1989 (pp 21-34).
[15] C. D. Fletcher, R.R. Schultz; RELAP5/M0DS User's Guidelines; ISBN 0-16-036098-6; 1992; Washington D.C., USA
[16] I. E. Idelchick; Handbook of Hydraulic Resistance, 2nd ed; Hemishpere Pub. Corp.;
1986; Washington, D.C., USA
[17] Ya. Z. Kazavchinskii; Heavy Water Thermophysical Properties; Israel Program for Scientific Translations; 1971; Jerusalem, Israel.
[18] L. Wester; Tabeller och Diagram for Energitekniska Berdkningar, 1990; Vasteras, Sweden.