- Email: [email protected]
The use of WIMS and CITATION codes in fuel loading required for the conversion of HEU MNSR core to LEU
Annals of Nuclear Energy 37 (2010) 1223–1228
Contents lists available at ScienceDirect
Annals of Nuclear Energy journal homepage: www.elsevier.com/locate/anucene
The use of WIMS and CITATION codes in fuel loading required for the conversion of HEU MNSR core to LEU T.S. Azande a,*, G.I. Balogun b, A.S. Ajuji b, S.A. Jonah b, Y.A. Ahmed b a b
Department of Physics, Ahmadu Bello University, Zaria, Nigeria Centre for Energy Research and Training, Ahmadu Bello University, Zaria, Nigeria
a r t i c l e
i n f o
Article history: Received 3 October 2009 Received in revised form 31 March 2010 Accepted 16 April 2010 Available online 2 June 2010 Keywords: 21 Specific nuclear reactors and associated plants Performance testing Reactivity insertions Steady-state conditions WIMS and CITATION Neutronics parameters
a b s t r a c t The Nigerian Research Reactor-1 (NIRR-1) falls in the category of Miniature Neutron Source Reactors (MNSR) using a fuel of 90% HEU. It is therefore desirable to convert it from this enrichment to LEU (less than 20%) in conformity with the new global trend of making research reactor fuel as unattractive as possible to groups that may be interested in using such highly enriched cores for non-peaceful purposes. In this work, we have developed a computational scheme based on WIMS and CITATION that would theoretically achieve this objective as easily as possible. The scheme systematically reduces the enrichment from 90% (or any other initial values) to less than 20% in steps of 5% or any desired percentage variation. Two fuel types (UAl4 and UO2) are considered in here, while maintaining the size and geometry of the core as well as the excess reactivity (between 3.5 and 4 mk). Our results show that the U-235 loading increases sharply as enrichment decreases. It has also been noticed that at 5% enrichment the fuel loading for both types is 2505 g. However, at 90% enrichment, the loading drops sharply to 998 g for UAl4 fuel and 946 g for UO2 fuel. Below the enrichment of 5%, the operation of NIRR-1 with both fuel types can be considered unrealistic as this requires structural adjustment which the work tries to maintain constant. Ó 2010 Elsevier Ltd. All rights reserved.
1. Introduction The international Reduced Enrichment for Research and Test Reactors (RERTR) program has recently advanced a cause for conversion of reactors with highly enriched uranium (HEU) to low enriched uranium (LEU). The reason for this global trend is to make test and research reactor fuel as unattractive as possible to groups that may be interested in using such highly enriched cores for nonpeaceful purposes. Miniature neutron source reactor (MNSR) is one of the many research reactors in the world that uses HEU as fuel. The Nigerian Research Reactor-1 (NIRR-1) falls in the category of HEUs, being 90% enriched in the fissile U-235. The neutronic parameters of MNSRs were carefully studied in the past by various operators using difference codes and programs. This include NIRR1 (Balogun, 2003 and Jonah et al., 2007), GHARR-1 (Sampong et al., 2006), PARR-2 (Waqah et al., 2008; Mahmood et al., 2008) and Syrian Reactor (Khamis and Khattab, 1999; Albarhoum, 2008). The differences observed in their approach and methodology has prompted the desire to embark on comprehensive conversion studies of the NIRR-1 in order to convert the reactor from HEU to LEU (less than 20%) in conformity with the growing global trend (see Table 1). * Corresponding author. Tel.: +234 70 3914 7585. E-mail address: [email protected] (T.S. Azande). 0306-4549/$ - see front matter Ó 2010 Elsevier Ltd. All rights reserved. doi:10.1016/j.anucene.2010.04.013
Earlier, intensive work was done by our group on thermal hydraulic and other reactor parameters (Balogun, 2004; Ahmed et al., 2008) and on automation of the neutronic calculation of MNSR (Balogun, 2003). In the neutronic calculation work, the reactor core parameters were modified interactively using the thermal reactor neutronics code WIMS (Askew et al., 1966) and core calculation code CITATION (Fowler et al., 1989). However, experience has shown that the input requirements for applications of these codes can be demanding and difficult to setup (Albarhoum, 2008; Balogun, 2003). To simplify the user interface to specific calculations like U-235 loading and fuel enrichment, we developed a simple user interface, where the major choices allow the user with the much simpler task of specifying basic geometrical, material and operational data and the program runs with ease. The provision of sensible defaults and ability to select from pre-checked libraries of such data has further aid WIMS user and made the program friendly (see Table 2). The scheme developed, which is an interactive program, prompts the user to give the names of data files which will be used in the calculation. Some default files are provided by the scheme and may be used unchanged. Others can be created or modified locally for the particular application being modeled and several variants may exist from which the user will be invited to make a selection.
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T.S. Azande et al. / Annals of Nuclear Energy 37 (2010) 1223–1228
Table 1 Results for UAl4 and UO2 fuel types from low to high enrichments. Enrichment (%)
U235 core loading (g) for UAl4
U235 core loading (g) for UO2
05.00 10.00 15.00 20.00 25.00 30.00 35.00 40.00 45.00 50.00 55.00 60.00 65.00 70.00 75.00 80.00 85.00 90.00
2505 1448 1274 1201 1155 1125 1107 1090 1074 1063 1052 1043 1034 1027 1023 1012 1004 0998
2505 1389 1216 1141 1098 1070 1049 1033 1020 1009 0999 0990 0982 0975 0968 0960 0953 0946
Table 2 Comparison of results obtained on HEU to LEU conversion for other MNSR with this work. Works
Fuel type
Enrichment (%)
Excess reactivity (mk)
Khamis and Khattab (1999) Sampong et al. (2006)
UO2
20.00
4.5790
UO2 U3Si2 U9Mo
12.60 19.75 19.75
4.5195 4.0436 4.3014
Waqah et al. (2008)
UO2 U3Si2 U9Mo
11.20 20.70 14.25
4.3351 4.3014 4.0733
This work
UO2
11.20 12.60 19.75 11.20 19.75
3.7616 3.9799 3.5762 3.5702 3.7664
UAl4
Fig. 1. Reactor dimensions and R–Z model used for neutronic analysis of NIRR-1.
T.S. Azande et al. / Annals of Nuclear Energy 37 (2010) 1223–1228
Recently, an automated software (BMAC) was developed in Syria by Albarhoum (2008) for modeling and performing neutronic calculations of MNSRs. However, the program is not a stand-alone for calculating the reactor parameters unless coupled with WIMSD-4 and CITATION codes. Also the BMAC software was not used to study conversion of HEU to LEU of the MNSRs due to its inability to import preliminary CITATION input data required for fuel conversion studies. In this work, we have developed a scheme that takes care of all these tedium by allowing the user to modify input options as desired and at any interval without initial procedures once one CITATION base input file is prepared. The desired calculations can be carried out by clicking the button for such a calculation. In the case of power failure or a pause at user’s discretion, the scheme automatically saves the last file it worked with so that whenever the code is invoked, the scheme resumes from where it stopped except if the user elects to abandon the old run. In a situation where there is no convergence, the scheme adjusts at least one input parameter as predetermined by the user until a convergence is arrived at. The adjustment can be upwards or downwards depending on which direction favors convergence as deemed by the scheme. The scheme stops automatically if there is no convergence after performing 550 iterations.
2. Materials and implementation The reactor was divided into a coarse R–Z mesh on which the fuel burn-up has been followed (Fig. 1). Sixty-nine group energy cross sections were averaged to obtain four groups. A flux volume weighting was used to reduce all the materials into a homogeneous material. A simplified scheme for WIMS and CITATION was made and 2D CITATION input file in R–Z geometry (Fig. 1) was pre-
1225
pared. All the group constants associated with the homogenized regions representing NIRR-1, including the control rod and the channel through which the control rod travels were properly defined in the input file. On invocation, the code prompts the user to supply some information from which it is able to locate the control rod approximately by automatically modifying the CITATION input data before proceeding with each CITATION run (Fig. 2). The user has the option to change fuel matrices and the input file to his/her CITATION input. Thereafter, the code displays a vertical cross section of a typical MNSR as depicted by Fig. 1 below: A pull-down menu captioned ‘Automated Calculations’ containing various manipulations such as ‘‘Test Mockup” and ‘‘fuel loading” are also presented to the user. When any of these calculations is going on, DOS runs in the back ground by default except elected by the user to have the DOS window displayed while the DOS application is running. This implies that only Computers with DOS compatible operating system can be used to run the scheme. Text boxes allowing the user to make changes to the default values that are applicable are displayed upon invocation of the code (Fig. 2). The values that can be changed by the user include: CITATION input file. The integer value representing the group constant data for the control rod as provided in section 008 of the CITATION input file. The length and worth of the control rod. The fuel matrices. The columns and rows occupied by the annular beryllium reflector and the thickness of annular beryllium reflector. The columns and rows occupied by the bottom beryllium reflector and the thickness of bottom beryllium reflector.
Fig. 2. A vertical cross section of MNSR and menu of various neutronics analysis that the user may carry out automatically.
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T.S. Azande et al. / Annals of Nuclear Energy 37 (2010) 1223–1228
Fig. 3. Display window for changing core enrichment of NIRR-1.
UO2
2600 2400 2200 2000
U-235 Loading (g)
1800 1600 1400 1200 1000 800 600 400 200 0 5
10
15
20
25
30
35
40
45
50
55
60
65
Enrichment (%) Fig. 4. U-235 loading against enrichment for UO2.
70
75
80
85
90
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T.S. Azande et al. / Annals of Nuclear Energy 37 (2010) 1223–1228
UAl4 2600 2400 2200 2000
U-235 Loading (g)
1800 1600 1400 1200 1000 800 600 400 200 0 5
10
15
20
25
30
35
40
45
50
55
60
65
70
75
80
85
90
Enrichment (%) Fig. 5. U-235 loading against enrichment for UAl4.
3. Results and discussion The movement of the control rod from one location to another is reflected on the monitor as well as a rotating moon with the message ‘‘CITATION is running” to warn the user from interrupting CITATION while it is running. At the end of each CITATION run, the scheme examines the output and extracts the calculated effective multiplication factor, keff. Two options are available to change enrichment which are by either fixing the U-235 loading in the core or varying the fuel enrichment in specified steps of reduction/increment or by fixing the desired enrichment and varying the U-235 loading in the core. This is accomplished by checking or un-checking a box provided for this purpose. When the box provided is checked, a desired enrichment is predefined and the interval for changing U-235 loading in the core is specified. When un-checked, the desired U-235 loading is predefined and the steps of increase or decrease in enrichment are specified (Fig. 2). Each time a calculation is unsuccessful, the fuel enrichment is adjusted either upwards or downwards depending on which direction favors convergence and the number density of U-235 is re-evaluated before proceeding with the main calculation. Once the U-235 loading and fuel enrichment in the core is known, the scheme can automatically determine the number density. Each time that there is a change in U-235 loading; a new number density is evaluated before proceeding to carry out other calculations (Fig. 3). When the ‘‘TEST MOCKUP” menu option is clicked, two critical tests are performed. First, the control rod is fully withdrawn and the effective multiplication factor keff is calculated. If a value greater than unity is returned, the message ‘‘reactor has passed first test” is displayed and the second test follows. Otherwise, ‘‘reactor has failed first test” is displayed. The control rod is then fully inserted and again CITATION is invoked to compute keff whose value is expected to be less than unity. Again, the appropriate message is
displayed as to whether the reactor has passed or failed the second test. Failure of any of these two tests clearly indicates that the reactor that has been put together is unreasonable hence the Analyst/ Designer must go back to the drawing board. Otherwise, if the design passes both tests, then the design process can be carried forward and other tests may follow. It was observed that the U-235 loading increases sharply as enrichment decreases. The behavior of U-235 loading with fuel enrichment for UO2 and UAl4 fuel types were shown in Figs. 4 and 5 respectively. The result indicates that at 5% enrichment the fuel loading for both types is 2505 g. However, at 90% enrichment, the loading drops to 998 g for UAl4 fuel and 946 g for UO2 fuel. Below 5% enrichment, the operation of NIRR-1 with both fuel types becomes impracticable as this requires structural adjustment. 4. Conclusion This work shows that to convert MNSRs core (NIRR-1) from HEU to LEU using WIMS and CITATION codes, the U-235 loading of the core has to be increased from its present HEU value of 998 g to about 1201 g for UAl4 fuel. For UO2 fuel, the loading is required to move from 946 g earlier simulated for HEU to about 1141 g for LEU. These two findings indicate that to achieve enrichment of less than 20% and maintain the reactor’s excess reactivity between 3.5 and 4 mk, a drastic review of U-235 loading and fuel enrichment is required. This is in agreement with earlier findings by workers in core conversion studies for Ghanaian, Syrian and Pakistani MNSRs. Acknowledgement The authors gratefully acknowledge the Management and the staff of the Reactor Engineering Section, Center for Energy Research and Training (CERT) Zaria, Nigeria for their various contributions.
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T.S. Azande et al. / Annals of Nuclear Energy 37 (2010) 1223–1228
References Ahmed, Y.A., Balogun, G.I., Jonah, S.A., Funtua, I.I., 2008. The behavior of reactor power and flux resulting from changes in core-coolant temperature for a miniature neutron source reactor. Annals of Nuclear Energy 35, 2417–2419. Albarhoum, M., 2008. Automation of the modeling and some neutronic calculations of the Syrian miniature neutron source reactor. Annals of Nuclear Energy 35, 1760–1763. Askew, J.R., Fayer, F.J., Kemshell, P.B., 1966. A general description of lattice code WIMSD. Journal of the British Nuclear Energy Society. Balogun, G.I., 2003. Automating some analysis and design calculations of miniature neutron source reactor at CERT (I). Annals of Nuclear Energy 30, 81–92. Balogun, G.I., 2004. Temperature dependence of safety reactivity factor and implications for limiting maximum NIRR-1 core temperature. Nigerian Journal of Physics, NIP 16, 55–59. Fowler, T.B., Vondy, D.R., Cunnigham, G.M., 1989. Nuclear reactor analysis code CITATION. ORNL-TM-2496.
Jonah, S.A., Liaw, J.R., Matos, J.E., 2007. Monte Carlo simulation of core physics parameters of the Nigeria research reactor-1 (NIRR-1). Annals of Nuclear Energy 34, 953–957. Khamis, I., Khattab, K., 1999. Lowering the enrichment of the Syrian miniature neutron source reactor. Annals of Nuclear Energy 26, 1031–1036. Mahmood, T., Pervez, S., Iqbal, M., 2008. Neutronic analysis for core conversion (HEU–LEU) of the Pakistan Research Reactor-2 (PARR-2). Annals of Nuclear Energy 35, 1440–1446. Sampong, S., Maakuu, B., Akahoo, E., Andan, A., Liaw, J., Matos, J., 2006. Progress in the neutronic core conversion (HEU–LEU) analysis of Ghana Research Reactor-1. In: Proceedings of the 28th International Meeting on RERTR Program, Cape town-South Africa, October 29-November 2, 2006. Waqah, S., Mirza, S., Mirza, N., Asad, T., 2008. A comparative neutronic study of the standard HEU core and various potential LEU alternatives for MNSR system. Nuclear Engineering and Design 238, 2302–2307.
Contents lists available at ScienceDirect
Annals of Nuclear Energy journal homepage: www.elsevier.com/locate/anucene
The use of WIMS and CITATION codes in fuel loading required for the conversion of HEU MNSR core to LEU T.S. Azande a,*, G.I. Balogun b, A.S. Ajuji b, S.A. Jonah b, Y.A. Ahmed b a b
Department of Physics, Ahmadu Bello University, Zaria, Nigeria Centre for Energy Research and Training, Ahmadu Bello University, Zaria, Nigeria
a r t i c l e
i n f o
Article history: Received 3 October 2009 Received in revised form 31 March 2010 Accepted 16 April 2010 Available online 2 June 2010 Keywords: 21 Specific nuclear reactors and associated plants Performance testing Reactivity insertions Steady-state conditions WIMS and CITATION Neutronics parameters
a b s t r a c t The Nigerian Research Reactor-1 (NIRR-1) falls in the category of Miniature Neutron Source Reactors (MNSR) using a fuel of 90% HEU. It is therefore desirable to convert it from this enrichment to LEU (less than 20%) in conformity with the new global trend of making research reactor fuel as unattractive as possible to groups that may be interested in using such highly enriched cores for non-peaceful purposes. In this work, we have developed a computational scheme based on WIMS and CITATION that would theoretically achieve this objective as easily as possible. The scheme systematically reduces the enrichment from 90% (or any other initial values) to less than 20% in steps of 5% or any desired percentage variation. Two fuel types (UAl4 and UO2) are considered in here, while maintaining the size and geometry of the core as well as the excess reactivity (between 3.5 and 4 mk). Our results show that the U-235 loading increases sharply as enrichment decreases. It has also been noticed that at 5% enrichment the fuel loading for both types is 2505 g. However, at 90% enrichment, the loading drops sharply to 998 g for UAl4 fuel and 946 g for UO2 fuel. Below the enrichment of 5%, the operation of NIRR-1 with both fuel types can be considered unrealistic as this requires structural adjustment which the work tries to maintain constant. Ó 2010 Elsevier Ltd. All rights reserved.
1. Introduction The international Reduced Enrichment for Research and Test Reactors (RERTR) program has recently advanced a cause for conversion of reactors with highly enriched uranium (HEU) to low enriched uranium (LEU). The reason for this global trend is to make test and research reactor fuel as unattractive as possible to groups that may be interested in using such highly enriched cores for nonpeaceful purposes. Miniature neutron source reactor (MNSR) is one of the many research reactors in the world that uses HEU as fuel. The Nigerian Research Reactor-1 (NIRR-1) falls in the category of HEUs, being 90% enriched in the fissile U-235. The neutronic parameters of MNSRs were carefully studied in the past by various operators using difference codes and programs. This include NIRR1 (Balogun, 2003 and Jonah et al., 2007), GHARR-1 (Sampong et al., 2006), PARR-2 (Waqah et al., 2008; Mahmood et al., 2008) and Syrian Reactor (Khamis and Khattab, 1999; Albarhoum, 2008). The differences observed in their approach and methodology has prompted the desire to embark on comprehensive conversion studies of the NIRR-1 in order to convert the reactor from HEU to LEU (less than 20%) in conformity with the growing global trend (see Table 1). * Corresponding author. Tel.: +234 70 3914 7585. E-mail address: [email protected] (T.S. Azande). 0306-4549/$ - see front matter Ó 2010 Elsevier Ltd. All rights reserved. doi:10.1016/j.anucene.2010.04.013
Earlier, intensive work was done by our group on thermal hydraulic and other reactor parameters (Balogun, 2004; Ahmed et al., 2008) and on automation of the neutronic calculation of MNSR (Balogun, 2003). In the neutronic calculation work, the reactor core parameters were modified interactively using the thermal reactor neutronics code WIMS (Askew et al., 1966) and core calculation code CITATION (Fowler et al., 1989). However, experience has shown that the input requirements for applications of these codes can be demanding and difficult to setup (Albarhoum, 2008; Balogun, 2003). To simplify the user interface to specific calculations like U-235 loading and fuel enrichment, we developed a simple user interface, where the major choices allow the user with the much simpler task of specifying basic geometrical, material and operational data and the program runs with ease. The provision of sensible defaults and ability to select from pre-checked libraries of such data has further aid WIMS user and made the program friendly (see Table 2). The scheme developed, which is an interactive program, prompts the user to give the names of data files which will be used in the calculation. Some default files are provided by the scheme and may be used unchanged. Others can be created or modified locally for the particular application being modeled and several variants may exist from which the user will be invited to make a selection.
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T.S. Azande et al. / Annals of Nuclear Energy 37 (2010) 1223–1228
Table 1 Results for UAl4 and UO2 fuel types from low to high enrichments. Enrichment (%)
U235 core loading (g) for UAl4
U235 core loading (g) for UO2
05.00 10.00 15.00 20.00 25.00 30.00 35.00 40.00 45.00 50.00 55.00 60.00 65.00 70.00 75.00 80.00 85.00 90.00
2505 1448 1274 1201 1155 1125 1107 1090 1074 1063 1052 1043 1034 1027 1023 1012 1004 0998
2505 1389 1216 1141 1098 1070 1049 1033 1020 1009 0999 0990 0982 0975 0968 0960 0953 0946
Table 2 Comparison of results obtained on HEU to LEU conversion for other MNSR with this work. Works
Fuel type
Enrichment (%)
Excess reactivity (mk)
Khamis and Khattab (1999) Sampong et al. (2006)
UO2
20.00
4.5790
UO2 U3Si2 U9Mo
12.60 19.75 19.75
4.5195 4.0436 4.3014
Waqah et al. (2008)
UO2 U3Si2 U9Mo
11.20 20.70 14.25
4.3351 4.3014 4.0733
This work
UO2
11.20 12.60 19.75 11.20 19.75
3.7616 3.9799 3.5762 3.5702 3.7664
UAl4
Fig. 1. Reactor dimensions and R–Z model used for neutronic analysis of NIRR-1.
T.S. Azande et al. / Annals of Nuclear Energy 37 (2010) 1223–1228
Recently, an automated software (BMAC) was developed in Syria by Albarhoum (2008) for modeling and performing neutronic calculations of MNSRs. However, the program is not a stand-alone for calculating the reactor parameters unless coupled with WIMSD-4 and CITATION codes. Also the BMAC software was not used to study conversion of HEU to LEU of the MNSRs due to its inability to import preliminary CITATION input data required for fuel conversion studies. In this work, we have developed a scheme that takes care of all these tedium by allowing the user to modify input options as desired and at any interval without initial procedures once one CITATION base input file is prepared. The desired calculations can be carried out by clicking the button for such a calculation. In the case of power failure or a pause at user’s discretion, the scheme automatically saves the last file it worked with so that whenever the code is invoked, the scheme resumes from where it stopped except if the user elects to abandon the old run. In a situation where there is no convergence, the scheme adjusts at least one input parameter as predetermined by the user until a convergence is arrived at. The adjustment can be upwards or downwards depending on which direction favors convergence as deemed by the scheme. The scheme stops automatically if there is no convergence after performing 550 iterations.
2. Materials and implementation The reactor was divided into a coarse R–Z mesh on which the fuel burn-up has been followed (Fig. 1). Sixty-nine group energy cross sections were averaged to obtain four groups. A flux volume weighting was used to reduce all the materials into a homogeneous material. A simplified scheme for WIMS and CITATION was made and 2D CITATION input file in R–Z geometry (Fig. 1) was pre-
1225
pared. All the group constants associated with the homogenized regions representing NIRR-1, including the control rod and the channel through which the control rod travels were properly defined in the input file. On invocation, the code prompts the user to supply some information from which it is able to locate the control rod approximately by automatically modifying the CITATION input data before proceeding with each CITATION run (Fig. 2). The user has the option to change fuel matrices and the input file to his/her CITATION input. Thereafter, the code displays a vertical cross section of a typical MNSR as depicted by Fig. 1 below: A pull-down menu captioned ‘Automated Calculations’ containing various manipulations such as ‘‘Test Mockup” and ‘‘fuel loading” are also presented to the user. When any of these calculations is going on, DOS runs in the back ground by default except elected by the user to have the DOS window displayed while the DOS application is running. This implies that only Computers with DOS compatible operating system can be used to run the scheme. Text boxes allowing the user to make changes to the default values that are applicable are displayed upon invocation of the code (Fig. 2). The values that can be changed by the user include: CITATION input file. The integer value representing the group constant data for the control rod as provided in section 008 of the CITATION input file. The length and worth of the control rod. The fuel matrices. The columns and rows occupied by the annular beryllium reflector and the thickness of annular beryllium reflector. The columns and rows occupied by the bottom beryllium reflector and the thickness of bottom beryllium reflector.
Fig. 2. A vertical cross section of MNSR and menu of various neutronics analysis that the user may carry out automatically.
1226
T.S. Azande et al. / Annals of Nuclear Energy 37 (2010) 1223–1228
Fig. 3. Display window for changing core enrichment of NIRR-1.
UO2
2600 2400 2200 2000
U-235 Loading (g)
1800 1600 1400 1200 1000 800 600 400 200 0 5
10
15
20
25
30
35
40
45
50
55
60
65
Enrichment (%) Fig. 4. U-235 loading against enrichment for UO2.
70
75
80
85
90
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T.S. Azande et al. / Annals of Nuclear Energy 37 (2010) 1223–1228
UAl4 2600 2400 2200 2000
U-235 Loading (g)
1800 1600 1400 1200 1000 800 600 400 200 0 5
10
15
20
25
30
35
40
45
50
55
60
65
70
75
80
85
90
Enrichment (%) Fig. 5. U-235 loading against enrichment for UAl4.
3. Results and discussion The movement of the control rod from one location to another is reflected on the monitor as well as a rotating moon with the message ‘‘CITATION is running” to warn the user from interrupting CITATION while it is running. At the end of each CITATION run, the scheme examines the output and extracts the calculated effective multiplication factor, keff. Two options are available to change enrichment which are by either fixing the U-235 loading in the core or varying the fuel enrichment in specified steps of reduction/increment or by fixing the desired enrichment and varying the U-235 loading in the core. This is accomplished by checking or un-checking a box provided for this purpose. When the box provided is checked, a desired enrichment is predefined and the interval for changing U-235 loading in the core is specified. When un-checked, the desired U-235 loading is predefined and the steps of increase or decrease in enrichment are specified (Fig. 2). Each time a calculation is unsuccessful, the fuel enrichment is adjusted either upwards or downwards depending on which direction favors convergence and the number density of U-235 is re-evaluated before proceeding with the main calculation. Once the U-235 loading and fuel enrichment in the core is known, the scheme can automatically determine the number density. Each time that there is a change in U-235 loading; a new number density is evaluated before proceeding to carry out other calculations (Fig. 3). When the ‘‘TEST MOCKUP” menu option is clicked, two critical tests are performed. First, the control rod is fully withdrawn and the effective multiplication factor keff is calculated. If a value greater than unity is returned, the message ‘‘reactor has passed first test” is displayed and the second test follows. Otherwise, ‘‘reactor has failed first test” is displayed. The control rod is then fully inserted and again CITATION is invoked to compute keff whose value is expected to be less than unity. Again, the appropriate message is
displayed as to whether the reactor has passed or failed the second test. Failure of any of these two tests clearly indicates that the reactor that has been put together is unreasonable hence the Analyst/ Designer must go back to the drawing board. Otherwise, if the design passes both tests, then the design process can be carried forward and other tests may follow. It was observed that the U-235 loading increases sharply as enrichment decreases. The behavior of U-235 loading with fuel enrichment for UO2 and UAl4 fuel types were shown in Figs. 4 and 5 respectively. The result indicates that at 5% enrichment the fuel loading for both types is 2505 g. However, at 90% enrichment, the loading drops to 998 g for UAl4 fuel and 946 g for UO2 fuel. Below 5% enrichment, the operation of NIRR-1 with both fuel types becomes impracticable as this requires structural adjustment. 4. Conclusion This work shows that to convert MNSRs core (NIRR-1) from HEU to LEU using WIMS and CITATION codes, the U-235 loading of the core has to be increased from its present HEU value of 998 g to about 1201 g for UAl4 fuel. For UO2 fuel, the loading is required to move from 946 g earlier simulated for HEU to about 1141 g for LEU. These two findings indicate that to achieve enrichment of less than 20% and maintain the reactor’s excess reactivity between 3.5 and 4 mk, a drastic review of U-235 loading and fuel enrichment is required. This is in agreement with earlier findings by workers in core conversion studies for Ghanaian, Syrian and Pakistani MNSRs. Acknowledgement The authors gratefully acknowledge the Management and the staff of the Reactor Engineering Section, Center for Energy Research and Training (CERT) Zaria, Nigeria for their various contributions.
1228
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References Ahmed, Y.A., Balogun, G.I., Jonah, S.A., Funtua, I.I., 2008. The behavior of reactor power and flux resulting from changes in core-coolant temperature for a miniature neutron source reactor. Annals of Nuclear Energy 35, 2417–2419. Albarhoum, M., 2008. Automation of the modeling and some neutronic calculations of the Syrian miniature neutron source reactor. Annals of Nuclear Energy 35, 1760–1763. Askew, J.R., Fayer, F.J., Kemshell, P.B., 1966. A general description of lattice code WIMSD. Journal of the British Nuclear Energy Society. Balogun, G.I., 2003. Automating some analysis and design calculations of miniature neutron source reactor at CERT (I). Annals of Nuclear Energy 30, 81–92. Balogun, G.I., 2004. Temperature dependence of safety reactivity factor and implications for limiting maximum NIRR-1 core temperature. Nigerian Journal of Physics, NIP 16, 55–59. Fowler, T.B., Vondy, D.R., Cunnigham, G.M., 1989. Nuclear reactor analysis code CITATION. ORNL-TM-2496.
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