Nuclear Society of Slovenia - Simulation of the Paks-2 ...201.6 Figure 6: Photos of damaged fuel...
9
International Conference Nuclear Energy for New Europe 2006 Portorož, Slovenia, September 18-21, 2006 http://www.djs.si/port2006 201.1 Simulation of the Paks-2 Incident: The CODEX-CT-1 Experiment Péter Windberg, Zoltán Hózer, Imre Nagy, András Vimi Hungarian Academy of Sciences KFKI Atomic Energy Research Institute H-1525 Budapest, Hungary [email protected], [email protected], [email protected], [email protected] ABSTRACT The Paks-2 cleaning tank incident was simulated with an electrically heated fuel bundle in the CODEX facility. The test conditions included seven hours of oxidation in hydrogen rich steam and final water quenching of the brittle fuel rods. The final state of the bundle showed similar picture that was observed after the incident at the power plant in 2003. 1 INTRODUCTION The cleaning tank incident at the Paks NPP on 10 th April 2003 resulted in severe fuel damage of 30 assemblies. The fuel rods heated up due to insufficient cooling. The zirconium components suffered heavy oxidation. The final quenching of the assembly’s lead to fragmentation of brittle materials. Several fuel failure mechanisms are known from accidents and experiments. However this incident differs from the commonly known and earlier investigated conditions. The temperature was not high enough for melting of fuel components, but the degradation was very severe. For the sake of better understanding of phenomena that took place during the Paks-2 incident AEKI launched a new experimental program to improve our knowledge on the behaviour of NPP fuel under accident conditions. The National Office for Research and Technology of Hungary supports the program. In the framework of the present project the conditions leading to severe fuel assembly damage at the Paks NPP in 2003 will be investigated. Experimental and computational activities are considered and the typical conditions of the incident (e.g. several hours of oxidation at relatively low temperature, hydrogen rich steam atmosphere) will be established in the tests. 2 TEST FACILITY Figure 1. shows the arrangement of the cleaning tank at the phase of overheating in “B” (stand-by) mode. The capacity of the pump was insufficient to remove the decay heat of the assemblies due to the geometrical configuration of the cleaning tank, and a by-pass flow occurred through holes of the assembly shrouds.
Transcript of Nuclear Society of Slovenia - Simulation of the Paks-2 ...201.6 Figure 6: Photos of damaged fuel...
International Conference
Nuclear Energy for New Europe 2006 Portorož, Slovenia, September 18-21, 2006
http://www.djs.si/port2006
201.1
Simulation of the Paks-2 Incident: The CODEX-CT-1 Experiment
Péter Windberg, Zoltán Hózer, Imre Nagy, András Vimi Hungarian Academy of Sciences KFKI Atomic Energy Research Institute
H-1525 Budapest, Hungary [email protected], [email protected], [email protected],
ABSTRACT The Paks-2 cleaning tank incident was simulated with an electrically heated fuel
bundle in the CODEX facility. The test conditions included seven hours of oxidation in hydrogen rich steam and final water quenching of the brittle fuel rods. The final state of the bundle showed similar picture that was observed after the incident at the power plant in 2003.
1 INTRODUCTION
The cleaning tank incident at the Paks NPP on 10th April 2003 resulted in severe fuel damage of 30 assemblies. The fuel rods heated up due to insufficient cooling. The zirconium components suffered heavy oxidation. The final quenching of the assembly’s lead to fragmentation of brittle materials.
Several fuel failure mechanisms are known from accidents and experiments. However this incident differs from the commonly known and earlier investigated conditions. The temperature was not high enough for melting of fuel components, but the degradation was very severe.
For the sake of better understanding of phenomena that took place during the Paks-2 incident AEKI launched a new experimental program to improve our knowledge on the behaviour of NPP fuel under accident conditions. The National Office for Research and Technology of Hungary supports the program.
In the framework of the present project the conditions leading to severe fuel assembly damage at the Paks NPP in 2003 will be investigated. Experimental and computational activities are considered and the typical conditions of the incident (e.g. several hours of oxidation at relatively low temperature, hydrogen rich steam atmosphere) will be established in the tests.
2 TEST FACILITY
Figure 1. shows the arrangement of the cleaning tank at the phase of overheating in “B” (stand-by) mode. The capacity of the pump was insufficient to remove the decay heat of the assemblies due to the geometrical configuration of the cleaning tank, and a by-pass flow occurred through holes of the assembly shrouds.
201.2
Figure1: Cleaning tank in service pit at unit No. 2 of Paks NPP [1]
The CODEX model of the cleaning tank was designed according to the following
requirements: • The model assemblies should be a 7-rod bundle • Length of bundle: 1000mm. • Max. temperature: 1200-1300oC. • Spacer grids close to that of original VVER-440 spacers. • Constant radial temperature profile. • Instead UO2, Alumina pellets • Rich instrumentation.
The main elements of the facility are the followings:
a) The cleaning tank model: • 7-rod VVER-440 type, fully zirconium bundle with alumina pellets. • Surrounding heaters with stainless steel cladding are used to reach high
temperature and to guarantee the constant temperature profile in the cleaning tank model.
• Inlet and outlet junctions are located in the bottom, perforations are used to establish bypass flow.
• Possibility to pressurize each of the fuel tubes with different pressure argon gas. • In the upper part of cleaning tank model a blow-down junction is placed to
simulate upper quenching. b) The expansion tank:
• Is connected to cleaning tank model to simulate the spent fuel storage pool. • The separate level measurement is applied.
c) The water pool condenser: • Condensates the steam released during the final quenching.
Proceedings of the International Conference Nuclear Energy for New Europe, 2006
201.3
• Empty volume to collect the non condensable gases after the water pool condenser
d) Hydrogen volume measurement system: • Measured the volume of hydrogen released from the cleaning tank in the course of
oxidation.
The CODEX CT facility is shown in Figure 2. The next Figures 3. and 4. show the cross and axial section and theirs highlights of cleaning tank model. The photos of cleaning tank components are also shown in Figure 5. The main characteristics of Paks-2 cleaning system and the CODEX CT facility are summarized and compared in Table 1.
~0,28MPa
0-10 V4000 A
P
Cle
ani n
g Ta
nk m
o del
Pump
Coo
ler
Fl
Pyro
met
er
cam
era
DP
P
DP
PCcontroller
cool
er
AR
GO
N
AR
GO
N
L
DPL
Safetyvalve
1 - 7
Hydrogen concentrationmeasurement
Hydrogen volume measurement system
L
Empt
y
P
Wat
er p
ool c
onde
nser
fi xed
leve
lve
ssel
Expa
nsio
nTa
n k
P
P
P
Figure 2: The schema of CODEX CT facility
Figure 3: Cross section of cleaning tank model
Proceedings of the International Conference Nuclear Energy for New Europe, 2006
201.4
To E
xpan
sion
Tan
k
Pyro
met
er
Cam
era
DP
leve
l
P. transducer
thermocouples
P
thermocouples
P. transducer
thermocouples
pressure loading
cooling inlet
cooling outlet
To b
ath
cond
ense
r7-rod bundle
Heated element
spacer
1000
(hea
t ed
leng
th)
Heat isolation
fixed
lev e
l-ves
s
35
Figure 4: Axial section of cleaning tank model
Shroud
7-rod bundle
Inner side of cleaning tank model
Figure 5: Photos of main components of cleaning tank model
Proceedings of the International Conference Nuclear Energy for New Europe, 2006
201.5
Table 1: Main parameters of the Paks cleaning tank and the CODEX-CT facility
Paks-2 cleaning tank CODEX-CT facility Number of fuel rod: 3870 (30X126) 7+12 (19) Flow rate/rod 5,6l/h 5,6 l/h Total flow rate 21t/h 106 l/h By-pass 43-48 cm2 3,81 cm2 (6XØ9) Volume of steam 4 m3 6 l Total volume Total volume: 6 m3 9,2 l Total surface of Zr 30 m2 0,44 m2
Ratio of total Zr surface/total volume 50 m-1 48 m-1
Total power (decay) 241 kW ~2,3 kW
The CODEX cleaning tank model was equipped with on-line data recording system that contained the following measurements:
• Temperature measurements: o 14 pcs. Thermocouples in fuel rods. o 1 pc. on heater surface o 5 pcs. On inside of outer tube, (in the gap) o 2 pcs. Inlet and outlet of cooling water o 1 pc. Water in expansion tank.
• Pressures: o In each fuel rod (7) o System pressure in the cleaning tank (1)
• Flow rate of cooling water • Water levels:
o Cleaning tank o Expansion tank
• Hydrogen concentration • Hydrogen volumetric flow • Video camera
3 THE PRELIMINARY TEST
A preliminary test experiment was carried out before the CODEX-CT-1 test. The full test scenario was performed. The differences between the CODEX-CT-1 test and preliminary test were the followings. In the preliminary test
• Empty cladding tubes were used. • Instead of hexagonal shroud Zr plates were used. • No quenching was applied.
The pre-test proved the operability of the facility. Heavy oxidation and brittle failure of cladding tubes and shroud plates was observed after the test (Fig.6.).
Proceedings of the International Conference Nuclear Energy for New Europe, 2006
201.6
Figure 6: Photos of damaged fuel tubes after the preliminary test
4 THE CODEX-CT-1 TEST
The CODEX-CT-1 test started with a tank filled up with water. During the first phase of the test the water evaporated from the upper part of the volume and a water level was established at the elevation of by-pass. The second phase included high temperature oxidation in H-rich steam atmosphere for seven hours. Finally the bundle was quenched from the bottom and the top.
Fig. 7. shows the main parameters of the CODEX-CT-1 experiment. In the test the power was constant. The temperature continuously increased up to 1280 oC. After reaching the saturation temperature a steam volume was formed in the tank and part of the water moved to the expansion tank. The high temperature oxidation took place for 7 hours.
Fig. 8. shows the pressure histories of fuel rods: Each rod was initially filled with different pressure argon gas (6,2-30,5 bar). Burst of fuel rods happened between 750 oC and 850 oC at 14-42 bars. The last rod filled with 6,2 bar did not burst until quenching. The other rods had higher initial pressure. The curves of the pressure history show that ballooning was accompanied with large volume change, probably long deformation of the tube. This will be investigated during the post-test analysis of the bundle.
The hydrogen production during the test was more than 100 l. This value corresponded to ~15% oxidation of the total Zr mass in the bundle. It is very probable that the oxidation ratio was higher and significant mass of the produced hydrogen was picked-up by the Zr components. The H content of the Zr cladding and shroud will be determined after the completion of post-test examinations.
Proceedings of the International Conference Nuclear Energy for New Europe, 2006
201.7
0 3600 7200 10800 14400 18000 21600 25200 28800 32400 36000 396000
5001000150020002500
elapsed time (sec)
Pow
er (W
)
0 3600 7200 10800 14400 18000 21600 25200 28800 32400 36000 39600
0400800
1200
Tem
pera
ture
at le
vel
760m
m o C
0 3600 7200 10800 14400 18000 21600 25200 28800 32400 36000 39600-40-20
0
20
expa
nsion
tank
le
vel (
cm)
0 3600 7200 10800 14400 18000 21600 25200 28800 32400 36000 39600-40-20020406080100120
Cle
aning
tank
le
vel (
cm)
Figure 7: The main parameters of the CODEX-CT-1 test
0 3600 7200 10800 14400 18000 21600 25200 28800 32400 360000
10
20
30
40
50
Pres
sure
(bar
)
Elapsed time (sec)
30.5
24.0
16,4
12,4
10,056,2
Initial pressures(bar)
Burst pressures (bar)
32,8136,32
27.01
22,96
15,5
11,51
not bursted
Figure 8: The pressure histories of fuel rods.
Proceedings of the International Conference Nuclear Energy for New Europe, 2006
201.8
Broken 7-rod bundle X-ray record of upper side of
damaged bundle The upper broken part of
bundle.
Figure 9: Photos of damaged bundle
Proceedings of the International Conference Nuclear Energy for New Europe, 2006
201.9
5 RESULTS
After the experiment the fuel bundle rods looked like some damaged fuel rods
observed at the Paks NPP after the incident (Fig. 9.). The upper part of the rods was almost fully oxidized, while the bottom part – cooled by water during the experiment – remained intact. The middle and upper sections were broken into several pieces. The material was very brittle, further fragmentation took place during the handling of fuel rods. The post-test examination indicated very high hydrogen content (several thousands ppm) in the Zr components (cladding and shroud). 6 CONCLUSIONS
The simulation of the cleaning tank incident provided detailed information on the
probable scenario of the real incident. The data will be used for model development purposes. Many similarities were found in bundle state with the assemblies of Paks-2 incident.
The test characterized the high temperature oxidation and embrittlement of Zr in hydrogen rich steam. The effect of quenching of heavily oxidized and hydrided fuel assembly was demonstrated.
It is very probable that the main parameters of the Paks-2 incident were close to that of CODEX-CT-1 test.
REFERENCES
[1] Bajsz József, Gadó János: Ex –core Fuel Demage Event at Paks. Causes, Consequences and Lessons Learned, International Conference NENE 2004 Portoros. Slovenia September 6-9.
Proceedings of the International Conference Nuclear Energy for New Europe, 2006
