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Transcript of LEADER, Task 5.5 ETDR Transient Analyses with SPECTRA Code LEADER Project JRC, Petten, February 26,...
LEADER, Task 5.5 ETDR Transient Analyses with
SPECTRA Code
LEADER Project
JRC, Petten, February 26, 2013
M.M. [email protected]
NRG-22694/13.118781
Content
• SPECTRA Model 3- SPECTRA/RELAP Model and Steady State Results 5- SPECTRA Model - Heat Transfer Correlations 17- SPECTRA Model - Reactivity Feedback 18- SPECTRA Model - SCRAM Signals 20
• Analyzed Transients 21- TR-4, Reactivity insertion, 250 pcm in 2 s
22- T-DEC1, Loss of all primary pumps, reactor trip fails 27- T-DEC5, Partial blockage of hottest fuel assembly 32
• Conclusions 34
• References 35
• Appendix A: Liquid Lead Properties 36
2NRG-22694/13.118781
SPECTRA Model
• A model of the ETDR, ALFRED reactor design [1] was prepared for the SPECTRA code [2].
• Nodalization of the SPECTRA model was assumed very similar to the nodalization applied for RELAP analyses at ENEA [3]. Some simplifications in the number of nodes were made whenever possible.
• The model consists of:- Primary system (liquid lead)- Steam Generators and secondary system loops (8 steam/water loops)- Isolation Condensers
• The EOC conditions were assumed. For modelling the gap, fuel swelling of 0.149 mm was assumed (initial gap size 0.150 mm), following RELAP model.
3NRG-22694/13.118781
SPECTRA Model
• The model was prepared such that the 8 loops can be combined into one or split into several (up to 8) loops, if needed.
- This is done using # and $, for example:* Multiplicity
102#21 $.0 * No. of loops
- Automatic replacement of # → loop No. and $ → number of identical loops, creates the desired model version.
• The model was tested by running steady state calculations and comparing results with the resuts obtained at ENEA using RELAP5 [3].
• Comparison of SPECTRA and RELAP results is given below. A good agreement is obtained.
4NRG-22694/13.118781
Fuel ElementsSPECTRA Model and Steady State Results
5NRG-22694/13.118781
Fuel ElementsRELAP Model and Steady State Results
6NRG-22694/13.118781
Reactor CoreSPECTRA Model and Steady State Results
7NRG-22694/13.118781
Reactor CoreRELAP Model and Steady State Results
8NRG-22694/13.118781
Primary SystemSPECTRA Model and Steady State Results
9NRG-22694/13.118781
Primary SystemRELAP Model and Steady State Results
10NRG-22694/13.118781
Steam GeneratorSPECTRA Model and Steady State Results
11NRG-22694/13.118781
Steam GeneratorRELAP Model and Steady State Results
12NRG-22694/13.118781
Secondary LoopSPECTRA Model and Steady State Results
13NRG-22694/13.118781
Secondary LoopRELAP Model and Steady State Results
14NRG-22694/13.118781
Isolation CondenserSPECTRA Model and Steady State Results
15NRG-22694/13.118781
Isolation CondenserRELAP Model and Steady State Results
16NRG-22694/13.118781
SPECTRA Model - Heat Transfer Coefficient Correlations
• If a liquid metal is to be applied in SPECTRA calculations, the HTC correlations must be defined in input. The following correlations have been used:
- Ushakov correlation - reference [4]:
here x = P/D
- Reactor Core: Ushakov, with P/D=1.32- Steam Generator: Ushakov, with P/D=1.4182
17NRG-22694/13.118781
)19.056.0(213
041.02055.7 xeP
xxx Nu
SPECTRA Model - Reactivity Feedback
• The reactivity feedback includes:
- Doppler reactivity effect:
- Axial fuel expansion:
- Coolant density:
- Cladding expansion:
- Wrapper expansion:
- Diagrid expansion:
- Pad expansion:
- Control rod:
)/ln( 0,fuelfuelDD TTKR
)( 0,coolcoolcoolcool TTcR
)( 0,cladcladcladclad TTcR
)( 0,diadiadiadia TTcR
)( 0,padpadpadpad TTcR
)( 0,rodrodrodrod TTcR
18NRG-22694/13.118781
)( 0,exp fuelfuelfuelfuel TTcR
)( 0,wrapwrapwrapwrap TTcR
SPECTRA Model - Reactivity Feedback
• The constants in the reactivity feedback are:
Component
(slides 8, 9)
- Doppler reactivity effect: KD = -566.0 SC-044/-055
- Axial fuel expansion: cfuel = -0.155 pcm/K SC-044/-055
- Coolant density: ccool = -0.268 pcm/K CV-044/-055
- Cladding expansion: cclad = +0.050 pcm/K SC-044/-055
- Wrapper expansion: cwrap = +0.026 pcm/K SC-064/-075
- Diagrid expansion: cdia = -0.152 pcm/K CV-020
- Pad expansion: cpad = -0.430 pcm/K CV-057
- Control rod, prompt: crod = -0.218 pcm/K CV-021
- Control rod, delayed: neglected
19NRG-22694/13.118781
SPECTRA Model - SCRAM Signals
• SCRAM signals incorporated into the model:
• Neutron flux > 120%• Average assembly ΔT > 1.2×nominal• Hot assembly ΔT > 1.2×nominal• Low primary floe W < 90%
20NRG-22694/13.118781
Analyzed Transients
• Transients:1. TR-4 Reactivity insertion, 250 pcm in 2 s.
Model: 8 identical loops combined into one, no IC
2. TO-1, TO-3Loss of FW pre-heater on 1 loop (TO-3: +all primary pumps stop)
Model: 1+3+4 identical loops, IC working on 4 loops
3. TO-4, TO-620% increase of FW flow (TO-6: +all primary pumps stop)
Model: 8 identical loops combined into one, no IC
4. T-DEC1 Loss of all primary pumps. Reactor trip fails.
Model: 8 identical loops combined into one, no IC
5. T-DEC3 Loss of SCS. Reactor trip fails.
Model: 3+5 identical loops, IC working on 3 loops
6. T-DEC-4 Loss of off-site power. Reactor trip fails.
Model: 3+5 identical loops, IC working on 3 loops
7. T-DEC5 Partial blockage of hottest assembly.
Model: 8 identical loops combined into one, no IC
8. T-DEC6 SCS failure
Model: 8 identical loops combined into one, no IC
green: done
red: still to be done
21NRG-22694/13.118781
TR-4 Reactivity insertion, 250 pcm in 2 s
• Scenario: - Reactivity of 250 pcm (0.8375 $) is inserted in 2 seconds.- Reactor trip (SCRAM signal) is disabled.
• Core power reaches 970 MW and decreases to about 500 MW. Corresponding peak in RELAP5 is 870 MW, with decrease to about 450 MW.
Reactor power, TR-4, SPECTRA Reactor power, TR-4, RELAP5 [3]
22NRG-22694/13.118781
0
200
400
600
800
1000
0 5 10 15 20 25 30
Pow
er (
MW
)
Time (s)
Core power
ETDR, ALFRED, SPECTRA
CF-900-Valu-0000 f(x)
Time302520151050
Tot
al c
ore
pow
er, [
MW
]
1,000
800
600
400
200
0
TR-4 Reactivity insertion, 250 pcm in 2 s
• Long term core power behavior:• After the initial transient the core power slowly reduces and stabilizes slightly below 400
MW, with the same power removed by SG-s. • SG power increases slowly due to temperature increase at SG inlet on the primary side.
Steam outlet temperature increases on the secondary side (constant FW flow rate).
Reactor and SG power, TR-4, SPECTRA Reactor and SG power, TR-4, RELAP5 [3]
23NRG-22694/13.118781
0
200
400
600
800
1000
0 300 600 900 1200 1500 1800
Pow
er (
MW
)
Time (s)
Core power
SG power
IC power
ETDR, ALFRED, SPECTRA
CF-900, reactor core CF-141, Steam Generators
Time1,8001,5001,2009006003000
Tot
al c
ore
pow
er, [
MW
]
1,000
800
600
400
200
0
TR-4 Reactivity insertion, 250 pcm in 2 s
• Fuel temperatures:• The fuel peak temperature reaches a maximum value close to 2700°C (2600°C in RELAP)
in the initial part of the transient and then slowly decreases to about 2400°C.• Maximum fuel temperature is higher in Spectra and exceeds for a short period the melting
temperature (MOX melting point ~2673°C). This is a consequence of higher peak power and the SPECTRA/RELAP difference will be investigated in the future.
Fuel temperatures, TR-4, SPECTRA Fuel temperatures, TR-4, RELAP5 [3]
24NRG-22694/13.118781
1000
1400
1800
2200
2600
3000
0 300 600 900 1200 1500 1800
Tem
per
atu
re (
°C)
Time (s)
T fuel peak
T fuel average
ETDR, ALFRED, SPECTRA
CF-110-Valu-0000, Fuel maximum RK-000-Tfue-0000, Fuel average
Time, [s]1,8001,5001,2009006003000
Tem
pera
ture
[C]
3,000
2,800
2,600
2,400
2,200
2,000
1,800
1,600
1,400
1,200
1,000
TR-4 Reactivity insertion, 250 pcm in 2 s
• Coolant temperatures:• After an initial jump of about 40 °C the core outlet temperature slowly increases following
the temperature increase at core inlet.• The maximum core outlet temperature stabilizes at about 620°C (about 600°C in RELAP).
Coolant temperatures, TR-4, SPECTRA Coolant temperatures, TR-4, RELAP5 [3]
25NRG-22694/13.118781
350
400
450
500
550
600
650
0 300 600 900 1200 1500 1800
Tem
per
atu
re (
°C)
Time (s)
T core in
T core out max
T core out ave
ETDR, ALFRED, SPECTRA
CV-020-Temp-pool, Core inletCV-017-Temp-pool, Core outlet, maximumCV-057-Temp-pool, Core outlet, average
Time, [s]1,8001,5001,2009006003000
Tem
pera
ture
[C]
650
600
550
500
450
400
350
TR-4 Reactivity insertion, 250 pcm in 2 s
• Reactivities:• The inserted reactivity is mainly counterbalanced by negative Doppler and fuel expansion
feedbacks induced by fuel temperature increase• Total reactivity reaches a maximum of about 190 pcm (175 pcm in RELAP) at 2 s and then
reduces according to negative feedbacks.
Reactivities, TR-4, SPECTRA Reactivities, TR-4, RELAP5 [3]
26NRG-22694/13.118781
-150
-100
-50
0
50
100
150
200
0 5 10 15 20 25 30
Rea
ctiv
ity (
pcm
)
Time (s)
Rea doppler
Rea fuel exp
Rea clad exp
Rea cool exp
Rea diagrid
Rea pads
Rea c.rods
Rea total
ETDR, ALFRED, SPECTRA
Total reactivity CF-901, DopplerCF-902, Fuel exp CF-903, Coolant densityCF-904, Clad exp. CF-908, Control rod
Time, [s]302520151050
Rea
ctiv
ity, [
-]
0.002
0.0015
0.001
0.0005
0
-0.0005
-0.001
-0.0015
-0.002
T-DEC1 Loss of All Primary Pumps
• Scenario: - Coastdown of all primary pumps.- The secondary circuits remain in operation in forced circulation- Reactor trip (SCRAM signal) is disabled.
• After an initial small core flow rate undershot natural circulation stabilizes in the primary circuit a little above 5000 kg/s.
Core inlet flow, T-DEC1, SPECTRA Core inlet flow, T-DEC1, RELAP5 [3]
27NRG-22694/13.118781
ETDR, ALFRED, SPECTRA
JN-059-Wtot-0000, Primary
Time, [s]1,8001,5001,2009006003000
Mas
s flo
w, [
kg/s
]
2.5E04
2.0E04
1.5E04
1.0E04
5.0E03
0
5000
10000
15000
20000
25000
0 300 600 900 1200 1500 1800
Mas
s flo
w r
ate
(kg
/s)
Time (s)
Core flow
T-DEC1 Loss of All Primary Pumps
• The core power initially reduces due to negative reactivity feedbacks and then stabilizes at about 240 MW (about 210 MW in RELAP5), in equilibrium with SG power.
• The SG power initially decreases due to reduced primary flow and then increases with the lead temperature increase at the SG inlet.
Reactor power, T-DEC1, SPECTRA Reactor power, T-DEC1, RELAP5 [3]
28NRG-22694/13.118781
0
50
100
150
200
250
300
0 300 600 900 1200 1500 1800
Pow
er (
MW
)
Time (s)
Core power
SG power
IC power
ETDR, ALFRED, SPECTRA
CF-900-Valu-0000 f(x) CF-141-Valu-0000
Time, [s]1,8001,5001,2009006003000
Pow
er [M
W]
300
250
200
150
100
50
0
T-DEC1 Loss of All Primary Pumps
• Fuel temperatures:• Peak and average fuel temperatures reduce according to the decrease of core power
level.• The maximum fuel temperature stabilizes at about 1700˚C (1400˚C in RELAP).
Fuel temperatures, TDEC1, SPECTRA Fuel temperatures, T-DEC1, RELAP5 [3]
29NRG-22694/13.118781
400
800
1200
1600
2000
0 300 600 900 1200 1500 1800
Tem
per
atu
re (
°C)
Time (s)
T fuel peak
T fuel average
ETDR, ALFRED, SPECTRA
CF-110-Valu-0000, Fuel maximum RK-000-Tfue-0000, Fuel average
Time, [s]1,8001,5001,2009006003000
Tem
pera
ture
[C]
2,000
1,600
1,200
800
400
T-DEC1 Loss of All Primary Pumps
• Coolant temperatures:• Initial lead temperature increase at core outlet max calculated value near 700°C at 15 s• Max core outlet temperature stabilizes just above 600 °C • The core inlet temperature slowly decreases and stabilizes at about 340°C
Coolant temperatures, TR-4, SPECTRA Coolant temperatures, TR-4, RELAP5 [3]
30NRG-22694/13.118781
300
350
400
450
500
550
600
650
700
0 300 600 900 1200 1500 1800
Tem
per
atu
re (
°C)
Time (s)
T core in
T core out max
T core out ave
ETDR, ALFRED, SPECTRA
CV-020-Temp-pool, Core inletCV-017-Temp-pool, Core outlet, maximumCV-057-Temp-pool, Core outlet, average
Time, [s]1,8001,5001,2009006003000
Tem
pera
ture
[C]
700
600
500
400
300
T-DEC1 Loss of All Primary Pumps
• Reactivities:• The inserted reactivity is mainly counterbalanced by negative Doppler and fuel expansion
feedbacks induced by fuel temperature increase• Total reactivity reaches a maximum of about 190 pcm (175 pcm in RELAP) at 2 s and then
reduces according to negative feedbacks.
Reactivities, TR-4, SPECTRA Reactivities, TR-4, RELAP5 [3]
31NRG-22694/13.118781
-120
-80
-40
0
40
80
120
0 300 600 900 1200 1500 1800
Rea
ctiv
ity (
pcm
)
Time (s)
Rea doppler
Rea fuel exp
Rea clad exp
Rea cool exp
Rea diagrid
Rea pads
Rea c.rods
Rea total
ETDR, ALFRED, SPECTRA
Total reactivity CF-901, DopplerCF-902, Fuel exp CF-903, Coolant densityCF-904, Clad exp. CF-908, Control rod
Time, [s]1,8001,5001,2009006003000
Rea
ctiv
ity, [
-]
0.001
0.0008
0.0006
0.0004
0.0002
0
-0.0002
-0.0004
-0.0006
-0.0008
-0.001
T-DEC5 Partial Blockage of Hottest Fuel Assembly
• Scenario: - Partial blockage of the hottest fuel assembly.- Inlet junction (JN-001, slide 7) assumed to be blocked- Blockages considered:
- 50%- 60%
- 70% Coolant temperatures, T-DEC5, SPECTRA
- 80%- 90%
- Reactor trip (SCRAM signal) is disabled.
32NRG-22694/13.118781
ETDR, T-DEC5, Blockage of Hottest Assembly
700.0
800.0
900.0
1,000.0
1,100.0
0 20 40 60 80 100
Time, [s]
Max
imum
Coo
lant
Tem
pera
ture
, [K
]
50% Blockage60% Blockage
70% Blockage80% Blockage
90% Blockage
T-DEC5 Partial Blockage of Hottest Fuel Assembly
• With 90% blockage (decrease of inlet flow area by a factor of 10, or increase of resistance factor by a factor of 100):
• maximum fuel temperature is ~2430 K, (~2160˚C)• maximum clad temperature is ~940 K, (~670˚C)• coolant exit temperature is ~1060 K (790˚C)
Fuel temperatures, T-DEC5, SPECTRA Cladding temperatures, T-DEC5, SPECTRA
33NRG-22694/13.118781
ETDR, T-DEC5, Blockage of Hottest Assembly
700.0
750.0
800.0
850.0
900.0
950.0
0 20 40 60 80 100
Time, [s]
Max
imum
Cla
ddin
g T
empe
ratu
re, [
K]
50% Blockage60% Blockage
70% Blockage80% Blockage
90% Blockage
ETDR, T-DEC5, Blockage of Hottest Assembly
2,250.0
2,300.0
2,350.0
2,400.0
2,450.0
0 20 40 60 80 100
Time, [s]
Max
imum
Fue
l Tem
pera
ture
, [K
]
50% Blockage60% Blockage
70% Blockage80% Blockage
90% Blockage
Conclusions
Results of several transients analyzed for the ETDR, ALFRED reactor design were shown and compared to the results of RELAP calculations from ENEA.
Steady state results obtained with SPECTRA and RELAP are in very good agreement.
Some discrepancies are observed for transient simulations. These discrepancies will be investigated in the future.
34NRG-22694/13.118781
References
[1] E. Bubelis, K. Mikityuk, "PLANT DATA FOR THE SAFETY ANALYSIS OF THE ETDR (ALFRED)", TEC058-2012, Revision: 0 (Draft), Issued by PSI/KIT (including contributions from ANSALDO, ENEA, EA, CEA, SRS), 30.04.2012.
[2] M.M. Stempniewicz, “SPECTRA Sophisticated Plant Evaluation Code for Thermal-Hydraulic Response Assessment, Version 3.60, August 2009, Volume 1 – Program Description, Volume 2 – User’s Guide, Volume 3 – Subroutine Description, Volume 4 - Verification and Validation”, NRG K5024/10.101640, Arnhem, April 24, 2009.
[3] G. Bandini, “Design and safety analysis of ALFRED - Accident Analyses Overview”, 3rd LEADER International Workshop, Bologna, 6-th - 7-th September.
[4] P.A. Ushakov, A.V. Zhukov, M.M. Matyukhin, “Heat transfer to liquid metals in regular arrays of fuel elements”, High Temperature, 15, pp. 868-873, 1977.
35NRG-22694/13.118781
Appendix A: Liquid Lead Properties
• If a liquid metal is to be applied in SPECTRA calculations, the properties of liquid metal must be supplied by the user. The properties of liquid lead were obtained flow:
“Handbook on Lead-bismuth Eutectic Alloy and Lead Properties, Materials Compatibility, Thermal-hydraulics and Technologies”, OECD/NEA Nuclear Science Committee. ISBN 978-92-64-99002-9, 2007
• The required properties include:- Saturation pressure
- Liquid properties, including:
- Density
- Specific heat
- Thermal conductivity
- Viscosity
- Speed of sound
- Vapor properties are not defined, i.e. sodium vapor cannot be encountered in calculations with the present model.
36NRG-22694/13.118781
Liquid Lead Properties, Psat(T), h(T)
37NRG-22694/13.118781
(a) Above: values tabulated for SPECTRA
(b) Below: source data
Liquid Lead Properties, ϱ(T), cp(T)
38NRG-22694/13.118781
(a) Above: values tabulated for SPECTRA
(b) Below: source data
Liquid Lead Properties, k(T), μ(T)
39NRG-22694/13.118781
(a) Above: values tabulated for SPECTRA
(b) Below: source data
Liquid Lead Properties, σ(T), c(T)
40NRG-22694/13.118781
(a) Above: values tabulated for SPECTRA
(b) Below: source data