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Burnup analysis and in-core fuel management study of the 3 MW TRIGA MARK II research reactor
Available online at www.sciencedirect.com annals of
NUCLEAR ENERGY Annals of Nuclear Energy 35 (2008) 141–147 www.elsevier.com/locate/anucene
Technical Note
Burnup analysis and in-core fuel management study of the 3 MW TRIGA MARK II research reactor M.Q. Huda a
a,*
, S.I. Bhuiyan a, T. Obara
b
Institute of Nuclear Science and Technology Atomic Energy Research Establishment Ganakbari, Savar, GPO Box 3787, Dhaka 1000, Bangladesh b Research Laboratory for Nuclear Reactors, Tokyo Institute of Technology, 2-12-1-N1-19 Ookayama, Meguro-ku, Tokyo 152-8550, Japan Received 16 October 2006; received in revised form 28 May 2007; accepted 29 May 2007 Available online 24 July 2007
Abstract The principal objective of this study is to formulate an effective optimal fuel management strategy for the TRIGA MARK II research reactor at AERE, Savar. The core management study has been performed by utilizing four basic types of information calculated for the reactor: criticality, power peaking, neutron flux and burnup calculation. This paper presents the results of the burnup calculations for TRIGA LEU fuel elements. The fuel element burnup for approximately 20 years of operation was calculated using the TRIGAP compute code. The calculation is performed in one-dimensional radial geometry in TRIGAP. Inter-comparison of TRIGAP results with other two calculations performed by MVP-BURN and MCNP4C-ORIGEN2.1 show very good agreement. Reshuffling at 20,000 MWh step provides the highest core lifetime of the reactor, which is 64,500 MWh. Besides, the study gives valuable insight into the behaviour of the reactor and will ensure better utilization and operation of the reactor in future. Ó 2007 Elsevier Ltd. All rights reserved.
1. Introduction The 3 MW TRIGA MARK II research reactor was commissioned at the Atomic Energy Research Establishment, Savar, Dhaka in 1986 and it went critical on 14th September, 1986. The reactor was designed to implement the various fields of basic nuclear research, manpower training and production of radioisotopes for its uses in agriculture, industry and medicine. The reactor is a light water cooled, graphite-reflected one, designed for continuous operation at a steady-state power level of 3000 kW (thermal). An outstanding feature of the TRIGA reactor is its proven safety, which stems from the large prompt negative temperature coefficient of reactivity of its U–ZrH fuel-moderator material. This study aims at establishing the applied technological know-how for burnup analysis and to formulate an optimal fuel management strategy for most effective utilization *
Corresponding author. Tel.: +880 2 7117050; fax: +880 2 8613051. E-mail address: [email protected] (M.Q. Huda).
0306-4549/$ - see front matter Ó 2007 Elsevier Ltd. All rights reserved. doi:10.1016/j.anucene.2007.05.013
of the fuel elements loaded initially in the TRIGA core. The basic objectives of core management are to achieve (i) high fuel discharge burnup as envisaged in the design and (ii) safe and efficient rated power operation. This requires planning of strategies for fuel reshuffling and/or reloading schemes and its safe implementation. The benchmark neutronic analyses have been performed only for the fresh core using different computer codes (Bhuiyan et al., 2000; Huda et al., 2004). At present, it is necessary to know the individual fuel burnup for the reshuffling of the core to ensure optimum utilization of fuel, which becomes very costly now a days. So it is believed that this study will contribute significantly to the safe and economic use of the TRIGA reactor. The core management study has been performed by utilizing four basic types of information: (i) criticality (keff) predictions, (ii) power peaking predictions, (iii) neutron flux and power distributions and (iv) fuel element burnup calculations. These involve the relationship of core as a function of burnup, which relates to the amount of energy produced by each fuel element in the core. In this context,
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M.Q. Huda et al. / Annals of Nuclear Energy 35 (2008) 141–147
fuel depletion or burnup calculations in developing schemes were used for core management. Various schemes for distribution of fuel rods have been considered in order to achieve high fuel burnup and reasonably flat power distribution. The TRIGAP (Mele and Ravnik, 1985) computer code that has already been used successfully elsewhere for the analysis of TRIGA research reactor (Ravnik et al., 1999) was applied for this analysis. The analyses which have been performed during this study are: (i) determination of individual fuel element burnup, (ii) calculation of core lifetime and (iii) formulation of fuel management strategy. 2. Tools and techniques The TRIGA core consists of 100 fuel elements arranged in a concentric hexagonal array within the core shroud. Fig. 1 shows the cross sectional view of the present core configuration of the reactor, which was achieved on October 9, 1986 during reactor start-up at full power operation. Elements are arranged in seven concentric rings and the spaces between the rods are filled with water that act as coolant and moderator. The burnup calculations were performed using TRIGAP computer program. TRIGAP was developed for research reactor calculations especially for TRIGA type reactors. The program assumes that the reactor has a cylindrical geometry and problems are supplied
with appropriate databases consisting mainly of the nuclear constants and the fuel operating history. The expected accuracy of the calculation is 0.5% for keff, 15% for power distribution and peaking factors and 10% for fuel burnup, which yields practically the same values as is conventionally done by standard power reactor codes (Mele and Ravnik, 1985; Ravnik et al., 1999). TRIGAP package is based on two-group diffusion equation (group boundary at 1 eV). It is solved in the finite differences approximation by fission density iteration method. A database for the TRIGAP code was generated for the Bangladesh TRIGA MARK II research reactor (Bhuiyan et al., 1992). The library was created using the WIMS-D/4 code (WIMS-D/4, 1983). The original WIMS cross-section library extended for TRIGA reactor specific nuclides (hydrogen bound in zirconium lattice, erbium) is used. According to the generally accepted approach the analysis starts from the basic reactor cell calculations. The cell is composed of a single LEU fuel consisting of a homogeneous mixture of Er–U–ZrH (20 wt% U, 19.7% 235U enrichment and 0.47 wt% Er), cladding (SS-304L) and Zr rod. The non-fuel units consist of dummy graphite element, water channel, void channel, central thimble (CT), pneumatic post and reflector. Normally one unit-cell represents a fuel rod in the centre surrounded by water. The homogenized unit-cell cross-sections are calculated for each fuel element separately as a function of buenup, fuel temperature, water
Fig. 1. Core configuration of the 3 MW TRIGA MARK II research reactor.
M.Q. Huda et al. / Annals of Nuclear Energy 35 (2008) 141–147
The primary thrust of this study was on the burnup calculations and their analysis. By burnup we mean the following changes in the core: (i) depletion of 235U, (ii) fission products buildup, (iii) spectral changes of flux, (iv) negligible plutonium production and (v) depletion of burnable poison. The criticality calculation gave confidence to perform a set of burnup calculations for the core loaded with fresh LEU fuel. Burnup calculations were performed with one-dimensional approximation for realistic operating conditions i.e. at a power of 1 MW. Some relatively smaller initial steps such as 50, 250, 300 and 400 MWh were considered. Rests of the steps were all 500 MWh to cover the reactor operating history of interest. This rule is followed throughout the calculation. The initial burnup steps in the calculation were taken smaller to consider the Xe and Sm buildup poisoning. It may be mentioned that TRIGAP calculates only the ring average fuel element burnup. A separate program was linked with TRIGAP to quantify individual fuel element burnup from the ring-averaged burnup. It was written on the basis of the individual power production by each fuel and fuel follower elements of the core (Huda et al., 2004). An essential aspect of developing any accurate reactor physics model is validation. Due to the lack of any measured data of individual fuel burnup, the TRIGAP results
6 TRIGAP MVP-BURN MCNP-ORIGEN
5
U burnup (%)
3. Burnup calculation
were compared with those performed by the 3D computer codes MVP (MVP/GMVP, 2003) and MCNP4C (Briesmeister, 1997) calculations. The neutronic analysis of the 3 MW TRIGA MARK II benchmark experiments at AERE, Savar were performed by the three-dimensional continuous energy Monte Carlo codes MVP and MCNP4C computer codes (Mahmood, 2006; Huda et al., 2004). All the neutronic parameters including the effective multiplication factor for the fresh core were evaluated. The analyzed models represent in detail all components of the core with literally no physical approximation. Continuous energy cross-section data from JENDL3.3 and ENDF/B-VI were used during the analysis. The Excess reactivity for the fresh core configuration calculated by MVP and MCNP codes were found to be 10.58$ and 9.90$, respectively. The experimental and TRIGAP calculated values for excess reactivity were found to be 10.27$ and 10.267$, respectively. All the calculated results are found to be in good agreement with the experimental value and seem to indicate that the comparison between the TRIGAP calculation and the calculations performed by MVP and MCNP codes are worthwhile. The individual fuel burnup was performed by the TRIGAP code and the results were compared with those performed by the MVP-BURN (MVP-BURN, 2003) and coupled MCNP4C-ORIGEN2.1 (ORIGEN2.1, 1991) calculations (Mahmood, 2006; Huda et al., 2007). The library used during the MVP calculation was derived from JENDL3.3 (JENDL-3.3, 2000) and the PWR library (ORIGEN2.1, 1991) was used during MCNP-ORIGEN calculation. The calculated burnup (% 235U depletion) at different locations of the TRIGA core (Fig. 1) during the years of 1986–2001 (total burnup was 5563.054 MWh) is shown in Fig. 2. An excellent agreement (less than ±5% difference) was observed between the TRIGAP result and the
235
temperature and xenon concentration. The group constants belonging to different unit-cells are homogenized in TRIGAP by volume weighting over each ring. The diffusion calculation is performed in one-dimensional radial geometry. At nominal power of 30 kW/element, 330 °C was used for the average fuel temperature, 121 °C for the average cladding temperature and 40 °C for the bulk water temperature. Cross-sections are calculated from zero burnup to 37% of initial 235U in 20 burnup steps. The TRIGAP input data consists of cell geometry, material composition and the generated cross-section library. More detailed specification of the fuel elements treated during the study may be found elsewhere (Huda et al., 2004). Using the zone averaged group constants the multiplication factor keff was calculated by direct solution of the diffusion equation. Each ring (A, B, C, D, E, F and G) has considered to be a separate zone and was smeared and homogenized. The created TRIGAP library was tested through practical calculations and was compared with experimental values or with the values in the safety analysis report (SAR) (Ahad et al., 1992). Excess reactivity of the fresh core configuration was measured and found to be 10.27$, while a value of 10.267$ was obtained using the generated library. The TRIGAP code with its new library was also used for calculating fast and thermal flux distributions which were found to be very close to the values reported in the SAR (Bhuiyan et al., 1992). Excellent agreement between the calculations and measurements establishes the validity of the library.
143
4
3
2
1
0 C4
C8
C10 G3 Locations
G34 Core average
Fig. 2. Comparison of the TRIGA fuel burnup (% 235U depletion) results calculated by TRIGAP at different locations of the core during the period of 1986–2001 (total burnup is 5563.054 MWh) with those of MVP-BURN and MCNP-ORIGEN calculations.
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Table 1 Individual TRIGA LEU fuel element burnup Depletion of
C1 C2 C4 C8 C10 D2 D3 D4 D5 D6 D7 D8 D9 D10 D11 D12 D13 D14 D15 D16 D17 D18 E1 E2 E3
6.832223 7.494614 7.817494 7.776865 7.912402 5.077292 5.317046 3.618824 5.220597 5.226909 3.663372 5.636665 5.638497 3.627453 5.219398 5.237113 3.823109 5.396029 5.454355 3.79074 6.043614 5.647886 4.478887 4.803263 5.055774
235
U (%)
Fuel ID
Depletion of
E4 E5 E6 E7 E8 E9 E10 E11 E12 E13 E14 E15 E16 E17 E18 E19 E20 E21 E22 E23 E24 F1 F2 F3 F4
4.531549 4.010598 4.358332 4.71893 4.38529 3.961605 4.314122 4.685726 4.324385 4.045845 4.507238 4.937471 4.658397 4.400801 4.721832 5.238534 4.635572 4.366761 4.670169 5.178094 4.815128 4.093133 4.109045 4.066116 3.963445
235
U (%)
calculations performed by the two most sophisticated advanced computer programs in nuclear industry. It provided the necessary confidence for further TRIGAP calculations. The individual burnup of TRIGA fuel and fuel-follower elements calculated by TRIGAP (% 235U depletion) during the period of 1986–2005 (total energy produced up to December 31, 2005 was 8585.594 MWh) are tabulated in Table 1. The core average burnup is found to be 4.4286%.
Fuel ID
Depletion of
F5 F6 F7 F8 F9 F10 F11 F12 F13 F14 F15 F16 F17 F18 F19 F20 F21 F22 F23 F24 F25 F26 F27 F28 F29
3.612635 3.446777 3.682604 3.80336 3.861872 3.894157 3.679575 3.520477 3.771162 3.798048 3.766685 3.671872 3.704859 3.969371 4.079943 4.072876 4.078055 3.977316 4.294085 4.189373 4.045156 3.989826 3.997859 4.299857 4.281707
235
U (%)
Fuel ID
Depletion of
F30 G3 G4 G5 G6 G9 G10 G11 G14 G15 G16 G17 G20 G21 G22 G23 G26 G27 G28 G29 G32 G33 G34 G35 G36
3.902233 4.183565 4.005969 3.604888 3.033704 3.68617 3.835633 3.932901 3.363811 3.506852 3.662109 3.788232 3.521861 3.96876 4.154594 4.32137 3.805243 4.131255 4.366002 4.407609 3.890201 4.248861 4.452008 4.193385 3.889503
235
U (%)
0.07
0.06
0.05
0.04
Δk
Fuel ID
0.03
4. Determination of core lifetime 0.02
The calculation of the keff and its relationship with core burnup is of primary importance to determine the core lifetime. First the excess reactivity for the beginning of the core life was calculated by TRIGAP at 1 MW reactor power and was found to be 8.9127$ (keff = 1.06654). Then the optimal time interval between two successive keff and power distribution calculation was determined. For burnup distribution calculation the power slope was considered constant in this interval. This time step was varied for fixed core calculation until it no longer influenced the core lifetime. The burnup calculation was performed for the TRIGA core without changing the loading pattern until the excess reactivity fell to zero. The variation of the keff according to average burnup in percent of initial 235U and also average burnup as a function of total thermal power produced are shown in Figs. 3 and 4, respectively. The excess reactivity becomes zero at an average burnup of 30.93% of 235U, which is equivalent to 60,500 MWh of reactor operating history. The depletion
0.01
0.00 0
5
10
15
20
25
30
Average Burnup (% 235U depletion)
Fig. 3. Multiplication factor as a function of average burnup in % depletion.
235
U
of 235U is found to be mostly a linear function of burnup as can be seen from Fig. 4. It should be noted, that the time steps used for the burnup calculations were reasonably small compared with fission product poisons to burn in. The initial time dependent sharp loss of reactivity is observed due to this effect and is shown in detail in Fig. 5. This is particularly due to the buildup of 133Xe and 149Sm. The concentration of these fission product poisons strongly influence the reactivity and eventually reaches equilibrium
M.Q. Huda et al. / Annals of Nuclear Energy 35 (2008) 141–147
D-ring and falls down linearly with the outer rings. Thermal flux peaks in the central thimble and falls sharply to 30–35% of its initial value in the D-ring and the fall continue up to F-ring. In the G-ring the flux has another small peak and then falls down gradually. The spatial distributions of ring averaged power density at BOL and EOL are also shown in Fig. 6. The power density remains almost analogous during the entire life of the reactor. The relative power produced by each fuel element over a power cycle as a function of burnup was determined and an inventory was maintained. These data were used to determine fuel depletion and isotopic buildup in each fuel rod.
35
Average Burnup (% 235U depletion)
30
25
20
15
10
5. In-core fuel management study
5
0 0
10000
20000 30000 40000 Core Burnup (MWh)
50000
60000
Fig. 4. Average burnup (%) as a function of core burnup (MWh).
9
8
Reactivity ($)
145
7
6
5 0
2000
4000
6000 8000 Core Burnup (MWh)
10000
12000
Fig. 5. Details of the initial fall of reactivity in relation to burnup at Xe and Sm equilibrium.
at about 0.5% burnup of initial 235U, as can be seen from Fig. 5. It is also observed from this figure that after 1000 MWh of core burnup the equilibrium is reached. During the first 1000 MWh, the excess reactivity drop is 2.35$. Whereas, the drop in excess reactivity during the next 1000 MWh is only 0.28$. This implies buildup of equilibrium Xe and Sm. This initial drop in multiplication factor also includes the reactivity loss due to heating of the fuel. Typical two-group zone average neutron radial flux distribution at the beginning of life (BOL) and end of life (EOL) of the core for 1 MW power were calculated and plotted in Fig. 6. The curves are for the calculated flux at the axial midplane of the core for neutron energies above and below 1 eV. It is observed that the fast flux peaks in
The strategies to determine the optimal in-core fuel management scheme for the best utilization of the fuel are already been described in the introduction. In order to perform the calculation, the core was burnt without changing the loading pattern until the excess reactivity fell to zero and the results are plotted in Fig. 7. It is observed from this figure that after initial few thousands MWhs the change in multiplication factor as a function of burnup is least and steady till it reaches around 20,000–25,000 MWh. From this point it is observed that the fall again becomes relatively sharper. This means that this is the nominal point for core life before the initial reloading step is needed. This point is found to correspond the 11–13% burnup of initial 235U. The GA projection to reach this point is 13% and 24,000 MWh. At this point the initial core requires added reactivity to remain operational at full power. A simple fuel management strategy was established in response to this findings. The plan was that the fuel elements were reshuffled after every 20,000 MWh according to their burnup, i.e. the most burnt fuel elements were replaced by least burnt elements and vice versa. Burnup was continued until excess reactivity again fell to zero. The results of the calculation are also shown in Fig. 7. The first observation is that the core lifetime depends on the burnup interval between two rearrangements. Second is that, it is possible to increase the core lifetime by around 7%, if the fuel is rearranged after certain period of operation. The best utilization of the fuel is obtained if the core is reshuffled after every 20,000 MWh that gives the maximum core life of 64,500 MWh. The rearrangement was done with extreme caution. Particularly the highest burnt elements were meticulously replaced by relatively least burnt elements in the core periphery. It has been observed that leakage increases with more burnt fuel in the periphery. The rate of reactivity reduction at end increases considerably as can be seen from Fig. 7. This is in good agreement with the characteristics of LEU Fuel. This fuel has 167Er and 166Er of high neutron absorption cross-section, so even small amounts have strong influence on fuel reactivity. After 1000 MWh of burnup the fuel temperature reactivity coefficient becomes less temperature dependent and smaller in magnitude than that for the initial clean core. This results
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M.Q. Huda et al. / Annals of Nuclear Energy 35 (2008) 141–147 0.5
0.020
BOL
0.016
EOL
0.3
0.012 Power density
0.2
0.008
Fast flux
Power Density (kW/cm3)
Relative Flux
0.4
Thermal flux
0.1
0.004
A
B
D
C
E
F
G
Reflector region
0.0
0.000 0
10
20
30 Radial Distance (cm)
40
50
Fig. 6. Radial distribution of relative flux and power density.
10
No reshuffling Reshuffling at 20,000 MWh Reshuffling at 40,000 MWh Reshuffling at 60,000 MWh
Reactivity ($)
8
6
4
2
0 0
10000
20000
30000 40000 Burnup (MWh)
50000
60000
70000
Fig. 7. Excess reactivity ($) of the reactor as a function of burnup (MWh) for different burnup intervals between reshuffling.
in increased transparency of the approximate 0.5 eV resonance region to thermal neutron. At 48,000 MWh this effect is more prominent. This temperature coefficient of reactivity in TRIGA core containing erbium is advantageous in that a minimum reactivity loss is incurred in reaching normal operating temperature. The reactivity reduction rate is obviously dependent on the reshuffling strategy especially at the bottom of the curve. 6. Conclusions The burnup of TRIGA fuel elements was investigated by reactor calculations using TRIGAP code. This study
describes the optimal in-core fuel management strategy of the TRIGA reactor that will contribute to safe operation and better utilization of the TRIGA fuel. Some specific and major points of interest are as follows: (i) Excess reactivity calculated at 50 W us 10.267$ compared to 10.27$ experimental value. (ii) Spatial flux and power density distributions are somewhat dependent on burnup. (iii) The keff is almost linear function of burnup after the Xe, Sm and temperature effects reach equilibrium. (vi) Xe, Sm equilibrium is reached at 1000 MWh i.e. at 0.5% burn up of the initial 235U.
M.Q. Huda et al. / Annals of Nuclear Energy 35 (2008) 141–147
(v) Reloading/reshuffling is needed to keep the reactor operational at full power at 25,000 MWh i.e. 11–13% burnup of the initial 235U. (vi) Reshuffling at 20,000 MWh step gives the longest core life of the reactor, which is 64,500 MWh. (vii) The highest average burnup that can be harnessed from this core in percent of 235U is 32.981%. This is the equilibrium burnup value for the core (GA projection is 25–30%). An effective optimal fuel management scheme for the TRIGA MARK II research reactor at AERE, Savar has been formulated in the study. Reshuffling at 20,000 MWh step will provide the highest core lifetime of the reactor, which is 64,500 MWh. The results from the TRIGAP calculation and those performed by MVP-BURN and MCNP4C-ORIGEN2.1 computer codes were found to be in very good agreement which establishes that the TRIGAP analysis might also be used as reference with confidence for TRIGA core configuration study. Acknowledgements The authors thank the members of the TRIGA Reactor Operation and Maintenance Unit of AERE for providing the operating history of the reactor. References Ahad, A.O.M.A., Ahmed, K., Ahmed, K.F., (Eds.), 1992. Final safety Analysis Report for the 3 MW TRIGA MARK-II Research Reactor at AERE, Savar, Dhaka, Bangladesh. Bhuiyan, S.I., Khan, A.R., Sarker, M.M., Rahman, M., Ara, Z.G., Musa, M., Mannan, M.A., Mele, I., 1992. Generation of a library for reactor
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calculations in core and safety parameter studies of the 3 MW TRIGA MARK II. Res. React. Nucl. Technol. 97, 253–263. Bhuiyan, S.I., Sarker, M.M., Rahman, M., Shahdatullah, M.S., Huda, M.Q., Chakrobortty, T.K., Khan, M.J.H., 2000. Criticality and safety parameter studies of a 3 MW TRIGA MARK II research reactor and validation of the generated cross-section library and computational method. Nucl. Technol. 130, 111–131. Briesmeister, J.F. (Ed.), 1997. MCNP – A General Monte Carlo NParticle Transport Code (Version 4C), RSIC#: CCC-700. Oak Ridge National Laboratory. Huda, M.Q., Rahman, M., Sarker, M.M., Bhuiyan, S.I., 2004. Benchmark Analysis of the TRIGA MARK II Research Reactor using Monte Carlo Techniques. Annal. Nucl. Energ. 31 (11), 1299–1313. Huda, M.Q., Bhuiyan, S.I., Obara T., 2007. Engineering analysis of the TRIGA research reactor using Monte Carlo technique. In: Proceedings of the International Conference on Nuclear Engineering, Paper ID#ICONE15-10449, Japan. JENDL-3.3, 2000. The Japanese Evaluated Nuclear Data Library, version 3.3, JAEA. Mahmood, M.S., 2006, Personal Communication, INST, Atomic Energy Research Establishment, Savar, Dhaka, Bangladesh. Mele, I., Ravnik, M., 1985. TRIGAP - A Computer Programme for Research Reactor Calculations. IJS-DP-4238. Josef Stephan Institute, Ljubljana, Slovenia. MVP/GMVP, 2003. General Purpose Monte Carlo Codes for Neutron and Photon Transport Calculations based on Continuous energy and Multigroup Methods, JAEA, Japan. MVP-BURN, 2003. A General Purpose Monte Carlo Code for Burnup Calculation, JAEA, Japan. ORIGEN2.1, 1991. Isotope Generation and Depletion Code – Matrix Exponential Method (Version 2.1), RSIC#: CCC-371, Oak Ridge National Laboratory. Ravnik, M., Zgar, T., Persic, A., 1999. Fuel element burnup determination in mixed TRIGA core using reactor calculations. Nucl. Technol. 128, 35–45. WIMS-D/4, 1983. A Neutronic Code for Standard Lattice Physics Analysis, distributed by OECD NEA Data Bank, Saclay, France.
NUCLEAR ENERGY Annals of Nuclear Energy 35 (2008) 141–147 www.elsevier.com/locate/anucene
Technical Note
Burnup analysis and in-core fuel management study of the 3 MW TRIGA MARK II research reactor M.Q. Huda a
a,*
, S.I. Bhuiyan a, T. Obara
b
Institute of Nuclear Science and Technology Atomic Energy Research Establishment Ganakbari, Savar, GPO Box 3787, Dhaka 1000, Bangladesh b Research Laboratory for Nuclear Reactors, Tokyo Institute of Technology, 2-12-1-N1-19 Ookayama, Meguro-ku, Tokyo 152-8550, Japan Received 16 October 2006; received in revised form 28 May 2007; accepted 29 May 2007 Available online 24 July 2007
Abstract The principal objective of this study is to formulate an effective optimal fuel management strategy for the TRIGA MARK II research reactor at AERE, Savar. The core management study has been performed by utilizing four basic types of information calculated for the reactor: criticality, power peaking, neutron flux and burnup calculation. This paper presents the results of the burnup calculations for TRIGA LEU fuel elements. The fuel element burnup for approximately 20 years of operation was calculated using the TRIGAP compute code. The calculation is performed in one-dimensional radial geometry in TRIGAP. Inter-comparison of TRIGAP results with other two calculations performed by MVP-BURN and MCNP4C-ORIGEN2.1 show very good agreement. Reshuffling at 20,000 MWh step provides the highest core lifetime of the reactor, which is 64,500 MWh. Besides, the study gives valuable insight into the behaviour of the reactor and will ensure better utilization and operation of the reactor in future. Ó 2007 Elsevier Ltd. All rights reserved.
1. Introduction The 3 MW TRIGA MARK II research reactor was commissioned at the Atomic Energy Research Establishment, Savar, Dhaka in 1986 and it went critical on 14th September, 1986. The reactor was designed to implement the various fields of basic nuclear research, manpower training and production of radioisotopes for its uses in agriculture, industry and medicine. The reactor is a light water cooled, graphite-reflected one, designed for continuous operation at a steady-state power level of 3000 kW (thermal). An outstanding feature of the TRIGA reactor is its proven safety, which stems from the large prompt negative temperature coefficient of reactivity of its U–ZrH fuel-moderator material. This study aims at establishing the applied technological know-how for burnup analysis and to formulate an optimal fuel management strategy for most effective utilization *
Corresponding author. Tel.: +880 2 7117050; fax: +880 2 8613051. E-mail address: [email protected] (M.Q. Huda).
0306-4549/$ - see front matter Ó 2007 Elsevier Ltd. All rights reserved. doi:10.1016/j.anucene.2007.05.013
of the fuel elements loaded initially in the TRIGA core. The basic objectives of core management are to achieve (i) high fuel discharge burnup as envisaged in the design and (ii) safe and efficient rated power operation. This requires planning of strategies for fuel reshuffling and/or reloading schemes and its safe implementation. The benchmark neutronic analyses have been performed only for the fresh core using different computer codes (Bhuiyan et al., 2000; Huda et al., 2004). At present, it is necessary to know the individual fuel burnup for the reshuffling of the core to ensure optimum utilization of fuel, which becomes very costly now a days. So it is believed that this study will contribute significantly to the safe and economic use of the TRIGA reactor. The core management study has been performed by utilizing four basic types of information: (i) criticality (keff) predictions, (ii) power peaking predictions, (iii) neutron flux and power distributions and (iv) fuel element burnup calculations. These involve the relationship of core as a function of burnup, which relates to the amount of energy produced by each fuel element in the core. In this context,
142
M.Q. Huda et al. / Annals of Nuclear Energy 35 (2008) 141–147
fuel depletion or burnup calculations in developing schemes were used for core management. Various schemes for distribution of fuel rods have been considered in order to achieve high fuel burnup and reasonably flat power distribution. The TRIGAP (Mele and Ravnik, 1985) computer code that has already been used successfully elsewhere for the analysis of TRIGA research reactor (Ravnik et al., 1999) was applied for this analysis. The analyses which have been performed during this study are: (i) determination of individual fuel element burnup, (ii) calculation of core lifetime and (iii) formulation of fuel management strategy. 2. Tools and techniques The TRIGA core consists of 100 fuel elements arranged in a concentric hexagonal array within the core shroud. Fig. 1 shows the cross sectional view of the present core configuration of the reactor, which was achieved on October 9, 1986 during reactor start-up at full power operation. Elements are arranged in seven concentric rings and the spaces between the rods are filled with water that act as coolant and moderator. The burnup calculations were performed using TRIGAP computer program. TRIGAP was developed for research reactor calculations especially for TRIGA type reactors. The program assumes that the reactor has a cylindrical geometry and problems are supplied
with appropriate databases consisting mainly of the nuclear constants and the fuel operating history. The expected accuracy of the calculation is 0.5% for keff, 15% for power distribution and peaking factors and 10% for fuel burnup, which yields practically the same values as is conventionally done by standard power reactor codes (Mele and Ravnik, 1985; Ravnik et al., 1999). TRIGAP package is based on two-group diffusion equation (group boundary at 1 eV). It is solved in the finite differences approximation by fission density iteration method. A database for the TRIGAP code was generated for the Bangladesh TRIGA MARK II research reactor (Bhuiyan et al., 1992). The library was created using the WIMS-D/4 code (WIMS-D/4, 1983). The original WIMS cross-section library extended for TRIGA reactor specific nuclides (hydrogen bound in zirconium lattice, erbium) is used. According to the generally accepted approach the analysis starts from the basic reactor cell calculations. The cell is composed of a single LEU fuel consisting of a homogeneous mixture of Er–U–ZrH (20 wt% U, 19.7% 235U enrichment and 0.47 wt% Er), cladding (SS-304L) and Zr rod. The non-fuel units consist of dummy graphite element, water channel, void channel, central thimble (CT), pneumatic post and reflector. Normally one unit-cell represents a fuel rod in the centre surrounded by water. The homogenized unit-cell cross-sections are calculated for each fuel element separately as a function of buenup, fuel temperature, water
Fig. 1. Core configuration of the 3 MW TRIGA MARK II research reactor.
M.Q. Huda et al. / Annals of Nuclear Energy 35 (2008) 141–147
The primary thrust of this study was on the burnup calculations and their analysis. By burnup we mean the following changes in the core: (i) depletion of 235U, (ii) fission products buildup, (iii) spectral changes of flux, (iv) negligible plutonium production and (v) depletion of burnable poison. The criticality calculation gave confidence to perform a set of burnup calculations for the core loaded with fresh LEU fuel. Burnup calculations were performed with one-dimensional approximation for realistic operating conditions i.e. at a power of 1 MW. Some relatively smaller initial steps such as 50, 250, 300 and 400 MWh were considered. Rests of the steps were all 500 MWh to cover the reactor operating history of interest. This rule is followed throughout the calculation. The initial burnup steps in the calculation were taken smaller to consider the Xe and Sm buildup poisoning. It may be mentioned that TRIGAP calculates only the ring average fuel element burnup. A separate program was linked with TRIGAP to quantify individual fuel element burnup from the ring-averaged burnup. It was written on the basis of the individual power production by each fuel and fuel follower elements of the core (Huda et al., 2004). An essential aspect of developing any accurate reactor physics model is validation. Due to the lack of any measured data of individual fuel burnup, the TRIGAP results
6 TRIGAP MVP-BURN MCNP-ORIGEN
5
U burnup (%)
3. Burnup calculation
were compared with those performed by the 3D computer codes MVP (MVP/GMVP, 2003) and MCNP4C (Briesmeister, 1997) calculations. The neutronic analysis of the 3 MW TRIGA MARK II benchmark experiments at AERE, Savar were performed by the three-dimensional continuous energy Monte Carlo codes MVP and MCNP4C computer codes (Mahmood, 2006; Huda et al., 2004). All the neutronic parameters including the effective multiplication factor for the fresh core were evaluated. The analyzed models represent in detail all components of the core with literally no physical approximation. Continuous energy cross-section data from JENDL3.3 and ENDF/B-VI were used during the analysis. The Excess reactivity for the fresh core configuration calculated by MVP and MCNP codes were found to be 10.58$ and 9.90$, respectively. The experimental and TRIGAP calculated values for excess reactivity were found to be 10.27$ and 10.267$, respectively. All the calculated results are found to be in good agreement with the experimental value and seem to indicate that the comparison between the TRIGAP calculation and the calculations performed by MVP and MCNP codes are worthwhile. The individual fuel burnup was performed by the TRIGAP code and the results were compared with those performed by the MVP-BURN (MVP-BURN, 2003) and coupled MCNP4C-ORIGEN2.1 (ORIGEN2.1, 1991) calculations (Mahmood, 2006; Huda et al., 2007). The library used during the MVP calculation was derived from JENDL3.3 (JENDL-3.3, 2000) and the PWR library (ORIGEN2.1, 1991) was used during MCNP-ORIGEN calculation. The calculated burnup (% 235U depletion) at different locations of the TRIGA core (Fig. 1) during the years of 1986–2001 (total burnup was 5563.054 MWh) is shown in Fig. 2. An excellent agreement (less than ±5% difference) was observed between the TRIGAP result and the
235
temperature and xenon concentration. The group constants belonging to different unit-cells are homogenized in TRIGAP by volume weighting over each ring. The diffusion calculation is performed in one-dimensional radial geometry. At nominal power of 30 kW/element, 330 °C was used for the average fuel temperature, 121 °C for the average cladding temperature and 40 °C for the bulk water temperature. Cross-sections are calculated from zero burnup to 37% of initial 235U in 20 burnup steps. The TRIGAP input data consists of cell geometry, material composition and the generated cross-section library. More detailed specification of the fuel elements treated during the study may be found elsewhere (Huda et al., 2004). Using the zone averaged group constants the multiplication factor keff was calculated by direct solution of the diffusion equation. Each ring (A, B, C, D, E, F and G) has considered to be a separate zone and was smeared and homogenized. The created TRIGAP library was tested through practical calculations and was compared with experimental values or with the values in the safety analysis report (SAR) (Ahad et al., 1992). Excess reactivity of the fresh core configuration was measured and found to be 10.27$, while a value of 10.267$ was obtained using the generated library. The TRIGAP code with its new library was also used for calculating fast and thermal flux distributions which were found to be very close to the values reported in the SAR (Bhuiyan et al., 1992). Excellent agreement between the calculations and measurements establishes the validity of the library.
143
4
3
2
1
0 C4
C8
C10 G3 Locations
G34 Core average
Fig. 2. Comparison of the TRIGA fuel burnup (% 235U depletion) results calculated by TRIGAP at different locations of the core during the period of 1986–2001 (total burnup is 5563.054 MWh) with those of MVP-BURN and MCNP-ORIGEN calculations.
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Table 1 Individual TRIGA LEU fuel element burnup Depletion of
C1 C2 C4 C8 C10 D2 D3 D4 D5 D6 D7 D8 D9 D10 D11 D12 D13 D14 D15 D16 D17 D18 E1 E2 E3
6.832223 7.494614 7.817494 7.776865 7.912402 5.077292 5.317046 3.618824 5.220597 5.226909 3.663372 5.636665 5.638497 3.627453 5.219398 5.237113 3.823109 5.396029 5.454355 3.79074 6.043614 5.647886 4.478887 4.803263 5.055774
235
U (%)
Fuel ID
Depletion of
E4 E5 E6 E7 E8 E9 E10 E11 E12 E13 E14 E15 E16 E17 E18 E19 E20 E21 E22 E23 E24 F1 F2 F3 F4
4.531549 4.010598 4.358332 4.71893 4.38529 3.961605 4.314122 4.685726 4.324385 4.045845 4.507238 4.937471 4.658397 4.400801 4.721832 5.238534 4.635572 4.366761 4.670169 5.178094 4.815128 4.093133 4.109045 4.066116 3.963445
235
U (%)
calculations performed by the two most sophisticated advanced computer programs in nuclear industry. It provided the necessary confidence for further TRIGAP calculations. The individual burnup of TRIGA fuel and fuel-follower elements calculated by TRIGAP (% 235U depletion) during the period of 1986–2005 (total energy produced up to December 31, 2005 was 8585.594 MWh) are tabulated in Table 1. The core average burnup is found to be 4.4286%.
Fuel ID
Depletion of
F5 F6 F7 F8 F9 F10 F11 F12 F13 F14 F15 F16 F17 F18 F19 F20 F21 F22 F23 F24 F25 F26 F27 F28 F29
3.612635 3.446777 3.682604 3.80336 3.861872 3.894157 3.679575 3.520477 3.771162 3.798048 3.766685 3.671872 3.704859 3.969371 4.079943 4.072876 4.078055 3.977316 4.294085 4.189373 4.045156 3.989826 3.997859 4.299857 4.281707
235
U (%)
Fuel ID
Depletion of
F30 G3 G4 G5 G6 G9 G10 G11 G14 G15 G16 G17 G20 G21 G22 G23 G26 G27 G28 G29 G32 G33 G34 G35 G36
3.902233 4.183565 4.005969 3.604888 3.033704 3.68617 3.835633 3.932901 3.363811 3.506852 3.662109 3.788232 3.521861 3.96876 4.154594 4.32137 3.805243 4.131255 4.366002 4.407609 3.890201 4.248861 4.452008 4.193385 3.889503
235
U (%)
0.07
0.06
0.05
0.04
Δk
Fuel ID
0.03
4. Determination of core lifetime 0.02
The calculation of the keff and its relationship with core burnup is of primary importance to determine the core lifetime. First the excess reactivity for the beginning of the core life was calculated by TRIGAP at 1 MW reactor power and was found to be 8.9127$ (keff = 1.06654). Then the optimal time interval between two successive keff and power distribution calculation was determined. For burnup distribution calculation the power slope was considered constant in this interval. This time step was varied for fixed core calculation until it no longer influenced the core lifetime. The burnup calculation was performed for the TRIGA core without changing the loading pattern until the excess reactivity fell to zero. The variation of the keff according to average burnup in percent of initial 235U and also average burnup as a function of total thermal power produced are shown in Figs. 3 and 4, respectively. The excess reactivity becomes zero at an average burnup of 30.93% of 235U, which is equivalent to 60,500 MWh of reactor operating history. The depletion
0.01
0.00 0
5
10
15
20
25
30
Average Burnup (% 235U depletion)
Fig. 3. Multiplication factor as a function of average burnup in % depletion.
235
U
of 235U is found to be mostly a linear function of burnup as can be seen from Fig. 4. It should be noted, that the time steps used for the burnup calculations were reasonably small compared with fission product poisons to burn in. The initial time dependent sharp loss of reactivity is observed due to this effect and is shown in detail in Fig. 5. This is particularly due to the buildup of 133Xe and 149Sm. The concentration of these fission product poisons strongly influence the reactivity and eventually reaches equilibrium
M.Q. Huda et al. / Annals of Nuclear Energy 35 (2008) 141–147
D-ring and falls down linearly with the outer rings. Thermal flux peaks in the central thimble and falls sharply to 30–35% of its initial value in the D-ring and the fall continue up to F-ring. In the G-ring the flux has another small peak and then falls down gradually. The spatial distributions of ring averaged power density at BOL and EOL are also shown in Fig. 6. The power density remains almost analogous during the entire life of the reactor. The relative power produced by each fuel element over a power cycle as a function of burnup was determined and an inventory was maintained. These data were used to determine fuel depletion and isotopic buildup in each fuel rod.
35
Average Burnup (% 235U depletion)
30
25
20
15
10
5. In-core fuel management study
5
0 0
10000
20000 30000 40000 Core Burnup (MWh)
50000
60000
Fig. 4. Average burnup (%) as a function of core burnup (MWh).
9
8
Reactivity ($)
145
7
6
5 0
2000
4000
6000 8000 Core Burnup (MWh)
10000
12000
Fig. 5. Details of the initial fall of reactivity in relation to burnup at Xe and Sm equilibrium.
at about 0.5% burnup of initial 235U, as can be seen from Fig. 5. It is also observed from this figure that after 1000 MWh of core burnup the equilibrium is reached. During the first 1000 MWh, the excess reactivity drop is 2.35$. Whereas, the drop in excess reactivity during the next 1000 MWh is only 0.28$. This implies buildup of equilibrium Xe and Sm. This initial drop in multiplication factor also includes the reactivity loss due to heating of the fuel. Typical two-group zone average neutron radial flux distribution at the beginning of life (BOL) and end of life (EOL) of the core for 1 MW power were calculated and plotted in Fig. 6. The curves are for the calculated flux at the axial midplane of the core for neutron energies above and below 1 eV. It is observed that the fast flux peaks in
The strategies to determine the optimal in-core fuel management scheme for the best utilization of the fuel are already been described in the introduction. In order to perform the calculation, the core was burnt without changing the loading pattern until the excess reactivity fell to zero and the results are plotted in Fig. 7. It is observed from this figure that after initial few thousands MWhs the change in multiplication factor as a function of burnup is least and steady till it reaches around 20,000–25,000 MWh. From this point it is observed that the fall again becomes relatively sharper. This means that this is the nominal point for core life before the initial reloading step is needed. This point is found to correspond the 11–13% burnup of initial 235U. The GA projection to reach this point is 13% and 24,000 MWh. At this point the initial core requires added reactivity to remain operational at full power. A simple fuel management strategy was established in response to this findings. The plan was that the fuel elements were reshuffled after every 20,000 MWh according to their burnup, i.e. the most burnt fuel elements were replaced by least burnt elements and vice versa. Burnup was continued until excess reactivity again fell to zero. The results of the calculation are also shown in Fig. 7. The first observation is that the core lifetime depends on the burnup interval between two rearrangements. Second is that, it is possible to increase the core lifetime by around 7%, if the fuel is rearranged after certain period of operation. The best utilization of the fuel is obtained if the core is reshuffled after every 20,000 MWh that gives the maximum core life of 64,500 MWh. The rearrangement was done with extreme caution. Particularly the highest burnt elements were meticulously replaced by relatively least burnt elements in the core periphery. It has been observed that leakage increases with more burnt fuel in the periphery. The rate of reactivity reduction at end increases considerably as can be seen from Fig. 7. This is in good agreement with the characteristics of LEU Fuel. This fuel has 167Er and 166Er of high neutron absorption cross-section, so even small amounts have strong influence on fuel reactivity. After 1000 MWh of burnup the fuel temperature reactivity coefficient becomes less temperature dependent and smaller in magnitude than that for the initial clean core. This results
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M.Q. Huda et al. / Annals of Nuclear Energy 35 (2008) 141–147 0.5
0.020
BOL
0.016
EOL
0.3
0.012 Power density
0.2
0.008
Fast flux
Power Density (kW/cm3)
Relative Flux
0.4
Thermal flux
0.1
0.004
A
B
D
C
E
F
G
Reflector region
0.0
0.000 0
10
20
30 Radial Distance (cm)
40
50
Fig. 6. Radial distribution of relative flux and power density.
10
No reshuffling Reshuffling at 20,000 MWh Reshuffling at 40,000 MWh Reshuffling at 60,000 MWh
Reactivity ($)
8
6
4
2
0 0
10000
20000
30000 40000 Burnup (MWh)
50000
60000
70000
Fig. 7. Excess reactivity ($) of the reactor as a function of burnup (MWh) for different burnup intervals between reshuffling.
in increased transparency of the approximate 0.5 eV resonance region to thermal neutron. At 48,000 MWh this effect is more prominent. This temperature coefficient of reactivity in TRIGA core containing erbium is advantageous in that a minimum reactivity loss is incurred in reaching normal operating temperature. The reactivity reduction rate is obviously dependent on the reshuffling strategy especially at the bottom of the curve. 6. Conclusions The burnup of TRIGA fuel elements was investigated by reactor calculations using TRIGAP code. This study
describes the optimal in-core fuel management strategy of the TRIGA reactor that will contribute to safe operation and better utilization of the TRIGA fuel. Some specific and major points of interest are as follows: (i) Excess reactivity calculated at 50 W us 10.267$ compared to 10.27$ experimental value. (ii) Spatial flux and power density distributions are somewhat dependent on burnup. (iii) The keff is almost linear function of burnup after the Xe, Sm and temperature effects reach equilibrium. (vi) Xe, Sm equilibrium is reached at 1000 MWh i.e. at 0.5% burn up of the initial 235U.
M.Q. Huda et al. / Annals of Nuclear Energy 35 (2008) 141–147
(v) Reloading/reshuffling is needed to keep the reactor operational at full power at 25,000 MWh i.e. 11–13% burnup of the initial 235U. (vi) Reshuffling at 20,000 MWh step gives the longest core life of the reactor, which is 64,500 MWh. (vii) The highest average burnup that can be harnessed from this core in percent of 235U is 32.981%. This is the equilibrium burnup value for the core (GA projection is 25–30%). An effective optimal fuel management scheme for the TRIGA MARK II research reactor at AERE, Savar has been formulated in the study. Reshuffling at 20,000 MWh step will provide the highest core lifetime of the reactor, which is 64,500 MWh. The results from the TRIGAP calculation and those performed by MVP-BURN and MCNP4C-ORIGEN2.1 computer codes were found to be in very good agreement which establishes that the TRIGAP analysis might also be used as reference with confidence for TRIGA core configuration study. Acknowledgements The authors thank the members of the TRIGA Reactor Operation and Maintenance Unit of AERE for providing the operating history of the reactor. References Ahad, A.O.M.A., Ahmed, K., Ahmed, K.F., (Eds.), 1992. Final safety Analysis Report for the 3 MW TRIGA MARK-II Research Reactor at AERE, Savar, Dhaka, Bangladesh. Bhuiyan, S.I., Khan, A.R., Sarker, M.M., Rahman, M., Ara, Z.G., Musa, M., Mannan, M.A., Mele, I., 1992. Generation of a library for reactor
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calculations in core and safety parameter studies of the 3 MW TRIGA MARK II. Res. React. Nucl. Technol. 97, 253–263. Bhuiyan, S.I., Sarker, M.M., Rahman, M., Shahdatullah, M.S., Huda, M.Q., Chakrobortty, T.K., Khan, M.J.H., 2000. Criticality and safety parameter studies of a 3 MW TRIGA MARK II research reactor and validation of the generated cross-section library and computational method. Nucl. Technol. 130, 111–131. Briesmeister, J.F. (Ed.), 1997. MCNP – A General Monte Carlo NParticle Transport Code (Version 4C), RSIC#: CCC-700. Oak Ridge National Laboratory. Huda, M.Q., Rahman, M., Sarker, M.M., Bhuiyan, S.I., 2004. Benchmark Analysis of the TRIGA MARK II Research Reactor using Monte Carlo Techniques. Annal. Nucl. Energ. 31 (11), 1299–1313. Huda, M.Q., Bhuiyan, S.I., Obara T., 2007. Engineering analysis of the TRIGA research reactor using Monte Carlo technique. In: Proceedings of the International Conference on Nuclear Engineering, Paper ID#ICONE15-10449, Japan. JENDL-3.3, 2000. The Japanese Evaluated Nuclear Data Library, version 3.3, JAEA. Mahmood, M.S., 2006, Personal Communication, INST, Atomic Energy Research Establishment, Savar, Dhaka, Bangladesh. Mele, I., Ravnik, M., 1985. TRIGAP - A Computer Programme for Research Reactor Calculations. IJS-DP-4238. Josef Stephan Institute, Ljubljana, Slovenia. MVP/GMVP, 2003. General Purpose Monte Carlo Codes for Neutron and Photon Transport Calculations based on Continuous energy and Multigroup Methods, JAEA, Japan. MVP-BURN, 2003. A General Purpose Monte Carlo Code for Burnup Calculation, JAEA, Japan. ORIGEN2.1, 1991. Isotope Generation and Depletion Code – Matrix Exponential Method (Version 2.1), RSIC#: CCC-371, Oak Ridge National Laboratory. Ravnik, M., Zgar, T., Persic, A., 1999. Fuel element burnup determination in mixed TRIGA core using reactor calculations. Nucl. Technol. 128, 35–45. WIMS-D/4, 1983. A Neutronic Code for Standard Lattice Physics Analysis, distributed by OECD NEA Data Bank, Saclay, France.