Unprotected Transients in a Small Scale Accelerator Driven System
X.-N. Chen, T. Suzuki, A. Rineiski, E. Wiegner, and W. Maschek Forschungszentrum Karlsruhe (FZK), Institute for Nuclear and Energy Technologies (IKET) Postfach 3640, D-76021 Karlsruhe, Germany,
[email protected] and M. Flad D.T.I. GmbH, Fritz-Erler-Straße 1-3, D-76133 Karlsruhe
Abstract — This paper deals with transients in a small scale accelerator driven system (ADS) caused by beam variations, reactivity changes and coolant blockages, namely, the beam trip (BT), unprotected transient over current (UTOC), protected and unprotected transient over power (PTOP and UTOP) and unprotected loss of flow (ULOF). The beam trip itself is of interest for investigating fuel temperature changes. The calculations of UTOC show that no severe fuel melting problems are encountered within the range of 100% increase of the source strength. The UTOP calculations predict that a 1$ reactivity increase does not cause any problem. For completeness also PTOP calculations have been performed. The ULOF calculations show that the natural convection remaining after the pump trip is sufficient to prevent any clad and fuel melting.
I. INTRODUCTION Accelerator Driven Systems (ADS), which combine a sub-critical reactor with a high energy proton accelerator and a spallation target, have a remarkable potential for transmutation of the long-lived nuclear waste. Currently the feasibility of an ADS is investigated worldwide. Within the 5th Framework Program of the European Union the so-called Preliminary Design Studies of an Experimental Accelerator-Driven System (PDS-XADS)1,2 are performed. This work is devoted both to a heavy metal cooled (Pb/Bi) and a gas (He) cooled option. In the first step this demonstrator is designed for using conventional fast reactor fuel and with a power level of 80 MWth and a sub-criticality level of keff = 0.97 at the beginning of life (BOL). At IKET of FZK safety analyses and investigations have been performed for Pb/Bi cooled designs3,4. Some benchmarking and new model developments for Pb/Bi multiphase flow conditions has been carried out5. In addition the SIMMER-III code used in this study has been adapted to the needs of describing a
sub-critical core with a strong external neutron source, the requirements of Pb/Bi cooling and other ADS specifics. In this paper results of various transients are displayed, analyzed for the PDS-XADS1, as the beam trip (BT), unprotected transient over current (UTOC), unprotected and protected transient over power (UTOP and PTOP) and unprotected loss of flow (ULOF). II. COMPUTER CODE The computer code SIMMER-III6 is developed by JNC (Japan Nuclear Cycle Development Institute, O-arai Engineering Center) in cooperation with Forschungszentrum Karlsruhe (FZK), CEA (Commissariat á l'Energie Atomique, CEN Grenoble and CE Cadarache) and IRSN (L’Institut de Radioprotection et de Sûreté Nucléaire). Recently PSI (Paul Scherrer Institute) and SCK.CEN (Studiecentrum voor Kernenergie-Centre de D’Etude de L’Energie Nucléaire) joined the SIMMER cooperation. SIMMER-III is a two-dimensional (R-Z or X-Y), three-velocity-field, multi-phase, multi-component,
Eulerian, fluid-dynamics code coupled with a structure model (fuel pins etc.) and a space-, time- and energydependent neutron dynamics model. SIMMER-III uses an elaborate scheme of equations of state functions for fuel, steel, coolant (light and heavy liquid metals, water and gas), absorber and simulation materials (e.g. alumina). In neutronics, the transient neutron flux distribution is calculated with the improved quasi-static method7. For the space dependent part, a TWODANT based flux shape calculation scheme has been implemented8. For ADS applications, an external space-, time- and energydependent neutron source has been implemented in the kinetics equations9,10. For the calculation of the different kinetics quantities different weighting functions can be chosen. The source effectiveness, indicating e.g. the position of the source relative to the bulk of fuel is transiently determined. A three-dimensional version of the code, SIMMER-IV is available and currently in its testing phase11.
material expansion processes, Doppler neutronic feedbacks etc.. With the given data the power level is ~ 80 MWth, the subcriticality level of the core is keff = 0.970 with βeff = 316 pcm. The total Doppler constant is about -600 pcm. The calculated steady state is in a good agreement with the investigated PDS-XADS design data1. The principle thermal hydraulic design data and calculated results are displayed in Tab. 1. The calculated axial and radial temperature distributions of coolant, clad, fuel surface and fuel interior in the core region at the steady state are shown in Fig. 2.
III. COMPUTATION MODELING AND RESULTS The present calculations are based on a cylindersymmetric geometric model. The configuration of the computation modeling is shown in Fig. 1. The core oxide fuel has a Pu enrichment of 23.25 %. In revised designs enrichment zoning is planned for the XADS. As the basic transient behavior should be studied in this ADS, no dedicated transmuter fuel10 with Minor Actinides (Np, Am, Cm) is loaded in this core in the first step. The SIMMER equations are solved on a fluid-dynamics and neutronics mesh, respectively. In the cases studied a fluiddynamics mesh with 10 radial and 29 axial cells has been chosen. The neutronics mesh is 46 times 64 with mesh cells smaller than 3 cm in core and blanket. For these first calculations the simple pin (SPIN) model in SIMMER has been chosen, which specifies a central and a surface fuel node. The effect of the central hole in the fuel pellet is taken into account. The elaborated DPIN2 model is planned to be used in the future. III.A. Steady State Steady state calculations serve as a basis for the following transient calculations. SIMMER is a transient code which requires detailed point-wise input data for densities, temperatures, pressures, velocities etc. These space and energy dependent data are usually not available. Therefore the calculations start from averaged initial values with given inlet temperature, overall power etc. and then iterates to balanced thermal hydraulic and neutronic steady state conditions. This includes local thermal
Fig. 1. The configuration of computation modeling.
1200
1100
1100
1000
1000
900
900
T[ K ]
T [ K]
1200
800
800
700
700
600
600 0.2
0.4
0.6
0.8
1
0.2
0.4
z [ m]
0.6
0.8
1
r [ m]
Fig. 2. Calculated axial (at the first ring subassembly in the core) and radial (at the mid-height of the core) temperature profiles at the steady state, where c, a, s and i denotes temperature of coolant, cladding, fuel surface and fuel interior, respectively and dots denote corresponding results of Ansaldo1.
Parameter Total thermal power Mean velocity of coolant
Design 80 0.42
Calculation 80 0.419
Unit [MW] [m/s]
Dynamic pressure drop
25
25
[Kpa]
Inlet coolant temperature
573
573
[K]
Outlet coolant temperature
673
677
[K]
Maximal clad temperature
753
741
[K]
Maximal fuel temperature
1169
1168
[K]
Table 1. Principal thermal hydraulic values in the design and the present calculation. III.B. Unprotected Beam Trips
and the other of ten seconds. Their results are presented in Figs. 3 and 4 respectively. After the beam trip, the power drops to about 10 % of nominal and decays further until the recovery of beam. The fuel temperature drop, which plays an important role for fuel pellet damage, depends on the duration of beam trip. The calculated clad and fuel temperature drops as function of the beam-trip duration at the hottest cell are shown in Table 2.
Since beam trips have to be expected during ADS operation, it is of importance to study their effect on the core, more precisely saying, on the fuel pellet. Because of limitation of the SPIN model we only study the temperature changes during beam trips without taking account of any fuel pellet thermal-mechanics. Two examples are displayed. One is a beam trip of one second
Duration of beam trip [s]
1
2
3
Maximal clad temperature drop [K]
6.78
19.3 34.3 49.6 63.9 76.8 88.2 98.3 107
114
Maximal fuel temperature drop [K]
84
158
470
222
4 277
5 324
6 364
7 397
8 426
9 450
Table 2. Maximal temperature drop in cladding and bulk fuel vs. duration of beam trip.
10
100
-2800
1200 -2900
Power Reactivity
20
-3000
10 -3100
7
Fuel Temperature [K]
Power [MW]
40
Reactivity [pcm]
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Middle
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1000 Bottom
900 800
4 -3200 0
5
10
15
20 25 Time [s]
30
35
40
700
45
5
15
35
25 Time [s]
45
Fig. 3. Results for a beam trip of one second: power and reactivity traces (left) and fuel interior temperatures of first ring subassembly (right) at the bottom, middle and top planes of the core. 100
-2800
1200
Power [MW]
40
-2900
20
-3000
10 -3100
7
Middle
1100
Fuel Temperature [K]
Power Reactivity
Reactivity [pcm]
70
Top
1000
Bottom
900 800 700
4
600
-3200
0
5
10
15
20 25 Time [s]
30
35
40
5
45
15
25 Time [s]
35
45
Fig. 4. Results for a beam trip of ten seconds: power and reactivity traces (left) and fuel interior temperatures of first ring subassembly (right) at the bottom, middle and top planes of the core. 200
-3050 Power Reactivity
-3150
100
-3200
90
Fuel Temperature [K]
Power [MW]
-3100
Reactivity [pcm]
1800
Middle
1600
Top 1400
Bottom
1200 1000 800
80
0
5
10
15
20 25 Time [s]
30
35
40
45
5
15
25 Time [s]
Fig. 5. Results for a 100 % increase of external neutron source at BOL condition of keff = 0.97.
35
45
III.C. Unprotected Transient Beam Over Currents
results are displayed in Fig. 6, where the over power is 10.2%, which is far away from the marginal value that leads to fuel melting. The maximal fuel temperature increase at the hottest cell is only 63 K. Independently from the transient calculations within the PDS-XADS project, hypothetical scenarios have been investigated to analyze margins of failure. The calculations with a 10$ reactivity jump reveal the expected sensitivity in the case where keff gets close to one. Up to a certain subcriticality level, e.g. by 9 $ increase, no fuel melting occurs. Slightly overshooting this subcriticality value leads to local fuel melting and fuel pin failure. In the case of 10$ reactivity increase the thermal power is elevated to a level of 600 MW, which results in an extended fuel and cladding melting. This fuel pin failure leads to a fuel sweep-out process and, subsequently, reactivity reduction due to the fuel removal with the coolant. Its results are shown in Fig. 7. The fuel sweep-out process is very effective in the SIMMER calculation, causing a deep drop in reactivity. Investigations are under way to better assess the sweep-out potentials for heavy liquid metal conditions. The impact of spacer grids on blockage formation is one open issue which has to be resolved. No severe power excursion takes place and the ADS is neutronically shut-down after the transient. Additional calculations with a 9$ reactivity increase have been performed for both BOL and EOL, where the subcriticality level is keff = 0.970 and 0.935 respectively. The radial power form factor at the EOL condition is larger than at the BOL condition. However, the over power caused by a reactivity increase, as well as the fuel temperature increase, at the EOL condition is significantly less than at the BOL condition. We show the comparison results in Figs. 8.
This scenario assumes that the beam power, which is variable to manage the power loss caused by fuel burn-up (no rods or burnable poisons are envisioned for this ADS), is incidentally increased. In the cases chosen, the neutron source is increased by 100 % at both beginning of life (BOL) and end of life (EOL) conditions, i.e. keff = 0.97 and 0.935 respectively. The power, reactivity and fuel temperature are shown for BOL in Fig. 5. The maximum increase of the fuel temperature is 588 K. The results show that the 100 % neutron source increase does not cause any fuel melting problems. For the EOL case similar results are obtained. The calculated asymptotic results of power are corroborated by the simple relation between proton beam power (Pbeam) and the core thermal power (Pcore) in dependence of the effective core multiplication factor (keff)1:
Pcore = Go Pbeam (1 − k eff ) where Go is the gain coefficient that mainly depends on the (p+, Xn) conversion. The temperature increases have been confirmed by a pin analysis for a given power. III.D. Unprotected Transient Over Power A UTOP might be caused by a reactivity perturbation in the core. As no scram rod system is provided in the design, the core must survive certain reactivity increases. The possible positive reactivity additions have been analyzed and identified1. The negative overall coolant void effect of the small core is of importance. At the BOL condition a reactivity jump of 1$ at a ramp rate 2 $/s, which is required in the design, is calculated. Its
-2700
1300
87.5
-3000 82.5 -3100 80.0
-3200 0
5
10
15
20 25 Time [s]
30
35
40
45
1200
Fuel Temperature [K]
-2900
85.0
Reactivity [pcm]
Power [MW]
Middle
-2800 Power Reactivity
1100
Top
1000
Bottom 900 800 700 5
15
25 Time [s]
35
45
Fig. 6. Results for a 1$ reactivity increase: power and reactivity traces (left) and fuel interior temperatures of first ring subassembly (right) at the bottom, middle and top planes of the core.
1000
0
Power Reactivity
600
-10000
100 60
-20000
Reactivity [pcm]
Power [MW]
300
Fuel Temperature [K]
3000
30
Top
Middle
2500 2000
Bottom
1500 1000 500
-30000 0
5
10
15
20 25 Time [s]
30
35
40
5
45
15
25 Time [s]
35
45
Fig. 7. Results for a 10$ reactivity increase: power and reactivity traces (left) and fuel interior temperatures of first ring subassembly (right) at the bottom, middle and top planes of the core. In the right figure only the pin fuel temperature is displayed. After fuel melting or clad failure the solid fuel is transferred into the liquid fuel or fuel particles. 0
200
-4000
400
200
-2000
-3000
-5000
-6000
100
Reactivity [pcm]
Power Reactivity
Power [MW]
Power [MW]
-1000
Reactivity [pcm]
Power Reactivity
90
100 80
80
-4000 0
5
10
15
20 25 Time [s]
30
35
40
45
-7000 0
5
10
15
20 25 Time [s]
30
35
40
45
Fig. 8. Power and reactivity traces in the cases of reactivity increase by 9 $ at the BOL condition (left) and the EOL condition (right). III.E. Protected Transient Over Power 100 70
Power [MW]
40
Power Reactivity
-2750
20 10
Reactivity [pcm]
-2500
-3000
7 4 0
5
10
15
20 25 Time [s]
30
35
40
45
Fig. 9. Power and reactivity traces in the PTOP case of 1$ reactivity increase.
In a protected transient over power the accelerator beam is immediately shut down after the reactivity increase is detected. One example with a 1$ reactivity increase is calculated and presented in Fig. 9. Since the 1$ increase is too small, the power trace behaves like in a beam-shut-down. III.F. Unprotected Loss of Flow Under nominal conditions, the coolant flow in the core region is driven partially by natural convection and partially by a gas lift pump that enhances the convection. In the present ULOF calculation, a coast down of the gas lift pump is simulated. For ULOF conditions the clad temperature is of importance. As shown in Fig. 10, the coolant flow rate under ULOF condition does approach 46% of nominal value by the natural convection after the
pump trip. The figure also reveals that the maximal clad temperature will never reach its melting point, i.e. the remaining natural convection is sufficient to prevent clad melting. Hence, it is predicted that no fuel-pin failure will occur under ULOF in the PDS-XADS.
2. L. Mansani, K. W. Burn, R. Tinti, B. Giraud, R. Sunderland & J. Cetnar, Proposed core configurations for a gas cooled and a lead-bismuth eutectic cooled ADS system, ENC2002 Scientific Seminar, Lille, France (2002). 3. W. Maschek, A. Rineiski, K. Morita, E. Kiefhaber, G. Buckel, M. Flad, P. Coste, S. Pigny, G. Rimpault, J. Louvet, T. Cadiou, S. Kondo, Y. Tobita, T. Suzuki, H. Yamano & S. Fujita, SIMMER-III, a Code for Analyzing Transients and Accidents in ADS, AccApp'00, Washington D.C., USA (2000). 4. W. Maschek, A. Rineiski, M. Flad, K. Morita & P. Coste, Analysis of Severe Accident Scenarios and Proposals for Safety Improvements for ADS Transmuters with Dedicated Fuel, Nucl. Technology, 141 (2003).
Fig. 10. Coolant flow rate and temperature of cladding steel for ULOF case. IV. CONCLUDING REMARKS Transient analyses of PDS-XADS with the SIMMERIII code are presented. Four types of transient problems, i.e. beam trip, beam over current, protected and unprotected transient over power, and unprotected loss of flow, are calculated. The beam trip transients are related mainly to operation, but they may cause thermal stress loads through large temperature drops. The calculations of beam over current transients show that within the range of a 100% increase of the source strength no fuel melting will occur. The same holds for the UTOP calculations for 1$ reactivity increase. In the ULOF calculations the natural convection after the pump trip is sufficient to prevent any pin failure. ACKNOWLEDGEMENTS This work has been partly funded by the EU Program PDS-XADS, Contract No: FIKW-CT-2001-00179. REFERENCES 1. PDS-XADS Preliminary Design Studies of an Experimental Accelerator-Driven System, EU Contract No. FIKW-CT-2001-00179 (2001).
5. T. Suzuki, Y. Tobita, S. Kondo, Y. Saito, K. Mishima, Analysis of gas-liquid metal two-phase flows using a reactor safety analysis code SIMMER-III, Nuclear Engineering and Design, 220 (2003), 207-223. 6. S. Kondo, K. Morita, Y. Tobita & N. Shirakawa, SIMMER-III: An advanced computer program for LMFBR severe accident analysis, ANP'92 Tokyo, Japan (1992). 7. K. O. Ott & R. J. Neuhold, Nuclear Reactor Dynamics, ANS, La Grange Park, USA (1986). 8. G. Buckel, E. Hesselschwerdt, E. Kiefhaber, S. Kleinheins & W. Maschek, A new SIMMER-III version with improved neutronics solution algorithms, FZKA 6290, Karlsruhe, Germany (1999). 9. R. Rineiski, B. Merk, & W. Maschek, ADS related extension of the neutronics module in the accident analysis code SIMMER-III, ADDTA'99, Praha, Czech Rep. (1999). 10. Rineiski, W. Maschek, & G. Rimpault, Performance of neutron kinetics models for ADS transient analyses, AccApp’01 & ADTTA’1, Reno (Nov. 2001). 11. Sa. Kondo, H. Yamano, Y. Tobita, S. Fujita, K. Morita, M. Mizuno, S. Hosono & T. Kondo, SIMMER-IV: A three-dimensional computer program for LMFR core disruptive accident analysis – Version 1.B model summary and program description, O-arai Engineering Center, Japan Nuclear Cycle Development Institute (2000).
