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Performance of UO2 ceramic fuel in low-power research reactors
Progress in Nuclear Energy 53 (2011) 73e75
Contents lists available at ScienceDirect
Progress in Nuclear Energy journal homepage: www.elsevier.com/locate/pnucene
Performance of UO2 ceramic fuel in low-power research reactors M. Albarhoum Department of Nuclear Engineering, Atomic Energy Commission, P. O. Box, 6091, Damascus, Syria
a r t i c l e i n f o
a b s t r a c t
Article history: Received 15 March 2010 Received in revised form 8 August 2010 Accepted 13 August 2010
The Low Enriched Uranium UO2 fuel performance in low-power research reactors is assessed in this paper. The usability of this fuel has been demonstrated in some research reactors in the world (SLOWPOKE-2). The fuel proved to be usable in the miniature neutron source low-power research reactors when about 50 fuel rods were substituted by as many dummy rods, while in SLOWPOKE reactors the number of fuel pins reduced by 98. About 3.8531 mk reactivity was rendered available at reactor start-up in MNSRs. The power of MNSRs needed to be increased by about 19%. Shut-down margin, effective shutdown margin, and control rod worth all decreased. Ó 2010 Elsevier Ltd. All rights reserved.
Keywords: Low-power reactors MNSR Fuel Initial excess reactivity Flux Power
1. Introduction Ceramic fuels have been used in Power Reactors since the beginning of the atomic era working for tens of years constantly at fairly high temperatures (higher than 300 C) (Cumo, 1986). They include UO2 (Rahn et al., 1984), UC (Glasstone and Sesonske, 1981), PuC (Bruzzi et al., 1981), PuO2 (Foster, 1983), etc fuels. Not all of these fuels were successfully used in nuclear reactors for their different nuclear and physical properties. The UO2 fuel proved to be a good fuel for many power reactors. In recent years it has been employed as a Low Enriched Uranium (LEU) fuel in Low-Power Research Reactors (LPRRs), intended here to have 30 kW power (Beeley et al., 1989). The advantages of this fuel consist mainly in the high density which allows reducing the enrichment of the fuel to the accepted levels (20%). Since the melting temperature of the uranium oxide is very high (2750 C) for the power LPRRs are operated at (w30 kW), the problem of fuel melting is eliminated. Sometimes this requires the cladding material to be changed (i.e. the use of zirconium alloys instead of aluminum alloys). The fuel in LPRRs (MNSRs as example) is arranged on 10 concentric circles (See Table 1). The total number of fuel rods is 347 in the actual configuration where Highly Enriched Uranium (HEU) fuel is used. Four tie rods are added in the 8th circle to fasten the upper and lower fuel ends. Three dummy elements (made of aluminum alloys) are inserted in the 10th circle.
E-mail address: pscientifi[email protected]. 0149-1970/$ e see front matter Ó 2010 Elsevier Ltd. All rights reserved. doi:10.1016/j.pnucene.2010.08.007
The main use of MNSRs is the Neutron Activation Analysis (NAA) (CIAE, 1993). Samples can be irradiated in 10 positions, 5 of which rely within the beryllium reflector. They are called Internal Irradiation Sites (IIS). Other 5 positions rely outside the reflector and are called External irradiation sites (EIS). Table 2 shows some of the general characteristics of these reactors (CIAE, 1993). These reactors use a HEU fuel which brings about 3.9400 mk Initial Excess Reactivity (IER) at reactor start-up when the fuel is fresh (See Table 3) (CIAE, 1993). The fuel is not highly consumed in this reactor (w1% only). The thermal neutron flux in these positions, both IIS and EIS, at rated power is about 1 1012 n/cm2 s, and 5 1011 n/cm2 s, respectively. These reactors have a good power/flux ratio. They have a high ratio of thermal/fast fluxes as well (4). The new LEU fuel is expected to keep these ratios possibly as high as in the case of the HEU fuel. Since the nuclear properties of the HEU and LEU fuels are different safety issues should be examined and addressed. Although this type of fuel has been considered in other works (Tayyab et al., 2008; Ghazi et al., 2009; Khattab and Sulieman, 2009) results are different for the different conditions authors used: 1 For the PARR-2 (Tayyab et al., 2008) the philosophy of reducing the enrichment of the fuel is different. The enrichment is reduced to 12.6% (instead of 19.75%) and the cladding material is changed to zircalloy-4. The core configuration is preserved. 2 For the Syrian MNSR (Ghazi et al., 2009) the analysis is directed primarily to the calculation of the feed back coefficients (in two
74
M. Albarhoum / Progress in Nuclear Energy 53 (2011) 73e75
Table 1 Lattice positions for the Fuel Cage for a generic MNSR. Four tie rods and three dummy elements add in circles 8 and 10, respectively (CIAE, 1993). Circle no.
Fuel Rods No.
Circle diameter (mm)
1. 2. 3. 4. 5. 6. 7. 8. 9. 10.
6 12 19 26 32 39 45 52 58 65
21.9 43.8 65.7 87.6 109.5 131.4 153.3 175.2 197.1 219.0
cases in which enrichments are 12% and 20%, respectively) which will, as the authors state, be used in the safety analyses related to the conversion of the reactor, while this paper is directed primly to the flux in the IIS and EIS and the related changes to the power values in addition to the IER. 3 For the case of the Syrian MNSR analyzed with the MCNP code (Khattab and Sulieman, 2009) very different configurations of the core were considered, while in this paper the core configuration was preserved as afore-mentioned. This paper therefore is directed to show when and where this LEU fuel will work satisfactorily in MNSRs preserving the core configuration completely.
2. Reactor calculations Neutronic calculations are executed using the BMAC system (Albarhoum, 2008). Using the models developed in other works (Albarhoum, 2005, 2009) and utilizing the BMAC system the parameters shown in Table 3 are found for the actual HEU fuel in a typical MNSR. The IER refers to the available reactivity at start-up and Control Rod Out (CRO) of the core. Flux values refer to 4 neutronic groups used in the calculation whose upper energy limits are: 10 MeV, 0.821 MeV, 5530 eV, and 0.625 eV, respectively. Table 2 Main properties of a generic MNSR (CIAE, 1993). Parameter
Description
Reactor type Rated thermal power Fuel Fuel pin inner radius without cladding (mm) Fuel pin outer radius without cladding (mm) Fuel pin length (mm) Cladding composition U-235 percentage in the fuel pin U-235 enrichment Uranium density in fuel pin Core Shape Core diameter Core height Fuel element shape Fuel elements number in the core Reactor continuous operating time at rated power Refueling period Burn-up Control rod length (mm) Absorber of the control rod Total number of irradiation sites Number of internal irradiation sites Thermal neutron flux in outer irradiation sites
Tank-in-pool w30 kW UeAl4 dispersed in Al matrix 4.3 5.5 230 Al 24.83 w90%
Reactor cooling mode Radial reflector thickness
Cylinder 23 cm 23 cm Thin rod 347 2.5 h 10 years w1% 260 Cd 10 5 1 1011 n/cm2 s at rated power Natural convection 10 cm
Table 3 Main Reactor Neutronic Characteristics for the UeAl4eAl fuel (CIAE, 1993). Item Initial Excess Rea. (IER) with CRO (mk) Neutron Flux in (I.I.S.) (1012 n/cm2 s) Neutron Flux (E.I.S) (1012 n/cm2 s) Control Rod Worth (mk) Shut-Down Margin (mk) Effective Shut-Down Margin (mk)
Calculated 3.9840 1.004370 475,096 7.142800 3.215300 1.303600
Measured 3.9400 0.02 1.000000 1% 0.500000 1% 7.000000 0.01 3.060000 0.02 e
Since the results fit fairly well with experimental data, the system is used to calculate the new properties of these reactors (i.e. the case of UO2 fuel). The physical properties of the UO2 fuel used in the calculation are shown in Table 4. The table makes a comparison between the same physical properties for both the actual and the LEU UO2 fuel. The enrichment is reduced to 19.75% instead of about 90%. Since the UO2 fuel is used in power reactors, a large international experience is available for its use in research reactors. The behavior of the fuel under irradiation is well known and requires no special precautions especially at these reduced burn-ups in LPRRs. 3. Results and discussion If no changes are made on both the dimensions and the materials of the fuel and the cladding, the data of Table 4 will lead to the results shown in Table 5. A typical MNSR will have more than 45 mk IER instead of the 4 mk permitted by regulations (CIAE, 1993). Reference fluxes are the fluxes for the UeAl4eAl fuel in the same positions. The IER is too high. Some fuel rods should be removed and substituted by dummy elements such that only about 4 mk are available in the core. This is done in Table 6. The new core configuration renders available about 3.8531 mk as IER which is acceptable for the regulations as mentioned above. The new characteristics of the reactor are summarized in Table 7. Fluxes are not reported in this table since they have been shown in Table 5. Fifty fuel rods were substituted by as many dummy rods not to disturb or modify the thermo-hydraulic characteristics of the core. Flux values in the IIS and EIS decreased by 1e18.55%, while the CR worth decreased by more than 25%. The effective Shut-Down Margin now is only 1.3358 mk. The safety of the reactor seems to be a bit compromised if the UO2 fuel is used, but the use of this fuel is still acceptable because the Effective Shut-Down Margin (ESDM) is still negative. The ESDM is a parameter which we use to evaluate exactly the safety of the
Table 4 A Comparison between the physical properties of the HEU and LEU fuels (CIAE, 1993). Fuel Properties
UeAl4eAl
UO2
Meat Density (g/cm3) Dispersed Phase Density (g/cm3) Wt% U in Dispersed Phase U Dens. in Dispersed Phase (g/cm3) U density in Meat (g/cm3) Volumetric. Fraction of Dispersed Phase % Porosity % Enrichment % Al content % in the meat U-235 fraction in meat % U-238 fraction in meat % Si fraction in meat % Mo fraction in meat % Oxygen fraction in meat %
3.456 5.70 64.0 3.70 0.955 26.20 1.20 89.87 72.310 24.831 2.80 e e e
10.6 10.6 88.0 9.30 9.342 39.0 0.0 19.75 e 17.405 70.72 e e 11.8669
M. Albarhoum / Progress in Nuclear Energy 53 (2011) 73e75 Table 5 A Comparison between the actual fuel and the UO2 fuel characteristics. Fuel Type
UeAl4eAl
UO2
Reactor Neutronic Characteristics Initial Excess Reactivity with CRO (mk) Variation from IER (mk) Neutron Flux in IIS (1012 n/cm2.s) Variation from Reference Flux (%) Neutron Flux in EIS (1012 n/cm2.s) Variation from Reference. Flux (%) Control Rod Worth (mk) Shut-Down Margin (mk)
3.9840 e 1.004370 e 0.475096 e 7.142800 3.158800
45.420800 41.436800 0.817976 18.558300 0.4143230 12.791700 4.566000 e
Table 6 Core configuration of a typical MNSR that uses LEU UO2 as fuel. Initial Excess Reactivity (mk)
No. of Dummy Elements (DE)
Circles occupied by DE
Fuel rods
Circles occupied by fuel rods
3.8531
53
10
294
1,2,3,4,5,6,7,8,9,10
Reactor Neutronic Characteristics
dimensions and the cladding material were modified, the latter being changed to zircaloy. The lifetime of the reactor was extended by about 100%, and 98 fuel rods were saved (See Table 8). The total mass of uranium increased, as expected from 0.9 kg to 5.6 kg. The volumes of water are almost the same in both cases (HEU and LEU fuels). These data confirms the usability of the LEU UO2 as a fuel for LPRRs. 4. Conclusions The UO2 ceramic fuel has been used in both power and LPRRs as a LEU fuel. The long international experience facilitates its use in MNSRs and SLOWPOKEs as LPRRs. The number of fuel rods is decreased by about 50 in the case of MNSRs and by about 100 in the case of SLOPOKEs, and the number of dummy elements is increased correspondingly. These reactors, from the neutronic stand-point are safer in the case of UeAl4eAl fuel. To keep the MNSRs performance in terms of neutron flux values in the IIS and EIS, the nominal power of these reactors should be increased to about 56 kW. However the control rod worth decreases by about 25%. Acknowledgment
Table 7 Reactor properties when the LEU fuel (UO2) is used. Fuel: UO2
Initial Excess Reactivity with CRO (mk) Variation from IER (mk) Control Rod Worth (mk) Shut-Down Margin (mk) Effective Shut-Down Margin (mk)
75
3.85310 0.13090 5.188900 1.335800 0.4706000
Table 8 Comparison of the HEU-fuelled and the LEU-fuelled cores of SLOWPOKE-2 (Bennett and Nielsen, 2002). Item
HEU-fuelled
LEU-fuelled
Core diameter (mm) Number of fuel pins Fuel pin diameter with cladding (mm) Fuel length (mm) Cladding Fuel Total mass of uranium (kg) Enrichment U-235 (%) Total mass of uranium 235 (kg) Volume of water in core (L)
228 296 5.23 225 Aluminum UeAl 28% alloy 0.9 93 0.82 7.8
234 198 5.26 234 Zircalloy-4 UO2 5.6 19.89 1.12 8.1
reactor. It is based on the normal Shut-Down Margin (SDM), but takes into consideration the effect of the initiating events mentioned in the Safety Analysis Report (SAR) of the reactor (i.e. the effects of flooding the IIS and EIS). The reduction of the flux values in both the IIS and EIS requires the power of the reactor to be increased correspondingly (by w19%). The new nominal power should be about 36 kW. The LEU UO2 has been already used for other LPPRs, the SLOWPOKE-2 reactor (Bennett and Nielsen, 2002). In that case both
The author thanks Professor I. Othman, Director General of the Atomic Energy Commission of Syria, for his encouragement and continued support. References Albarhoum, M., 2008. Automation of the modeling and some neutronic calculations of the Syrian miniature neutron source reactors. Annals of Nuclear Energy 35, 1760e1763. Albarhoum, M., 2005. A 3-D neutronics model for the calibration of the control rod of the Syrian MNSR. Progress in Nuclear Energy 46 (2), 159e164. Albarhoum, M., 2009. Optimization of the reflector design of the Syrian MNSR. Annals of Nuclear Energy 51 (6e7), 676e679. Beeley, P.A., Bennett, L.G.I., Kennedy, G.G., 1989. Comparison of the operational characteristics of the HEU and LEU fuelled SLOWPOKE-2 reactors. In: Proceedings, IAEA Symposium on Research Reactor Safety, Operation and Modifications, vol. 1, IAEA-SM-310/50P, Chalk River. Bennett, L.G.I., Nielsen, K.S., 2002. LEU-fuelled SLOWPOKE-2 research reactors: operational experience and utilization. In: Proceedings of the 2002 International Meeting on Reduced Enrichment for Research and Test Reactors, Bariloche, Argentina, November 3e8. Bruzzi, L., Cicognani, G., Domenici, G., 1981. Il Ciclo Del Combustibile Dei Reacttorii Nucleari. CNEN, Roma, Italy. CIAE, 1993. Safety Analysis Report (SAR) for the Syrian Miniature Neutron Source Reactor China. Cumo, M., 1986. Impianti Nucleari. UTET, Torino, Italy. Foster, A., 1983. Basic Nuclear Engineering. Allyn and Bacon, Inc, Boston, USA. Ghazi, N., Haj Hassan, H., Hainoun, A., 2009. Determination of major kinetic parameters of the Syrian MNSR for different fuel loading using Monte Carlo technique. Annals of Nuclear Energy 36 (11e12), 1663e1667. Glasstone, S., Sesonske, A., 1981. Nuclear Reactor Engineering. Van Nostrand Reinhold, New York, USA. Khattab, K., Sulieman, I., 2009. Assessment of fuel conversion from HEU to LEU in the Syrian MNSR reactor using the MCNP code. Progress in Nuclear Energy 51 (6e7), 727e730. Rahn, F., Adamantiades, A., Kenton, J., Braun, C., 1984. A Guide to Nuclear Power Technology. John Wiley & Sons, Inc, New York, USA. Tayyab, M., Showket, P., Masood, I., 2008. Neutronic analysis for core conversion (HEUeLEU) of Pakistan research reactor-2 (PARR-2). Annals of Nuclear Energy 35 (8), 1440e1446.
Contents lists available at ScienceDirect
Progress in Nuclear Energy journal homepage: www.elsevier.com/locate/pnucene
Performance of UO2 ceramic fuel in low-power research reactors M. Albarhoum Department of Nuclear Engineering, Atomic Energy Commission, P. O. Box, 6091, Damascus, Syria
a r t i c l e i n f o
a b s t r a c t
Article history: Received 15 March 2010 Received in revised form 8 August 2010 Accepted 13 August 2010
The Low Enriched Uranium UO2 fuel performance in low-power research reactors is assessed in this paper. The usability of this fuel has been demonstrated in some research reactors in the world (SLOWPOKE-2). The fuel proved to be usable in the miniature neutron source low-power research reactors when about 50 fuel rods were substituted by as many dummy rods, while in SLOWPOKE reactors the number of fuel pins reduced by 98. About 3.8531 mk reactivity was rendered available at reactor start-up in MNSRs. The power of MNSRs needed to be increased by about 19%. Shut-down margin, effective shutdown margin, and control rod worth all decreased. Ó 2010 Elsevier Ltd. All rights reserved.
Keywords: Low-power reactors MNSR Fuel Initial excess reactivity Flux Power
1. Introduction Ceramic fuels have been used in Power Reactors since the beginning of the atomic era working for tens of years constantly at fairly high temperatures (higher than 300 C) (Cumo, 1986). They include UO2 (Rahn et al., 1984), UC (Glasstone and Sesonske, 1981), PuC (Bruzzi et al., 1981), PuO2 (Foster, 1983), etc fuels. Not all of these fuels were successfully used in nuclear reactors for their different nuclear and physical properties. The UO2 fuel proved to be a good fuel for many power reactors. In recent years it has been employed as a Low Enriched Uranium (LEU) fuel in Low-Power Research Reactors (LPRRs), intended here to have 30 kW power (Beeley et al., 1989). The advantages of this fuel consist mainly in the high density which allows reducing the enrichment of the fuel to the accepted levels (20%). Since the melting temperature of the uranium oxide is very high (2750 C) for the power LPRRs are operated at (w30 kW), the problem of fuel melting is eliminated. Sometimes this requires the cladding material to be changed (i.e. the use of zirconium alloys instead of aluminum alloys). The fuel in LPRRs (MNSRs as example) is arranged on 10 concentric circles (See Table 1). The total number of fuel rods is 347 in the actual configuration where Highly Enriched Uranium (HEU) fuel is used. Four tie rods are added in the 8th circle to fasten the upper and lower fuel ends. Three dummy elements (made of aluminum alloys) are inserted in the 10th circle.
E-mail address: pscientifi[email protected]. 0149-1970/$ e see front matter Ó 2010 Elsevier Ltd. All rights reserved. doi:10.1016/j.pnucene.2010.08.007
The main use of MNSRs is the Neutron Activation Analysis (NAA) (CIAE, 1993). Samples can be irradiated in 10 positions, 5 of which rely within the beryllium reflector. They are called Internal Irradiation Sites (IIS). Other 5 positions rely outside the reflector and are called External irradiation sites (EIS). Table 2 shows some of the general characteristics of these reactors (CIAE, 1993). These reactors use a HEU fuel which brings about 3.9400 mk Initial Excess Reactivity (IER) at reactor start-up when the fuel is fresh (See Table 3) (CIAE, 1993). The fuel is not highly consumed in this reactor (w1% only). The thermal neutron flux in these positions, both IIS and EIS, at rated power is about 1 1012 n/cm2 s, and 5 1011 n/cm2 s, respectively. These reactors have a good power/flux ratio. They have a high ratio of thermal/fast fluxes as well (4). The new LEU fuel is expected to keep these ratios possibly as high as in the case of the HEU fuel. Since the nuclear properties of the HEU and LEU fuels are different safety issues should be examined and addressed. Although this type of fuel has been considered in other works (Tayyab et al., 2008; Ghazi et al., 2009; Khattab and Sulieman, 2009) results are different for the different conditions authors used: 1 For the PARR-2 (Tayyab et al., 2008) the philosophy of reducing the enrichment of the fuel is different. The enrichment is reduced to 12.6% (instead of 19.75%) and the cladding material is changed to zircalloy-4. The core configuration is preserved. 2 For the Syrian MNSR (Ghazi et al., 2009) the analysis is directed primarily to the calculation of the feed back coefficients (in two
74
M. Albarhoum / Progress in Nuclear Energy 53 (2011) 73e75
Table 1 Lattice positions for the Fuel Cage for a generic MNSR. Four tie rods and three dummy elements add in circles 8 and 10, respectively (CIAE, 1993). Circle no.
Fuel Rods No.
Circle diameter (mm)
1. 2. 3. 4. 5. 6. 7. 8. 9. 10.
6 12 19 26 32 39 45 52 58 65
21.9 43.8 65.7 87.6 109.5 131.4 153.3 175.2 197.1 219.0
cases in which enrichments are 12% and 20%, respectively) which will, as the authors state, be used in the safety analyses related to the conversion of the reactor, while this paper is directed primly to the flux in the IIS and EIS and the related changes to the power values in addition to the IER. 3 For the case of the Syrian MNSR analyzed with the MCNP code (Khattab and Sulieman, 2009) very different configurations of the core were considered, while in this paper the core configuration was preserved as afore-mentioned. This paper therefore is directed to show when and where this LEU fuel will work satisfactorily in MNSRs preserving the core configuration completely.
2. Reactor calculations Neutronic calculations are executed using the BMAC system (Albarhoum, 2008). Using the models developed in other works (Albarhoum, 2005, 2009) and utilizing the BMAC system the parameters shown in Table 3 are found for the actual HEU fuel in a typical MNSR. The IER refers to the available reactivity at start-up and Control Rod Out (CRO) of the core. Flux values refer to 4 neutronic groups used in the calculation whose upper energy limits are: 10 MeV, 0.821 MeV, 5530 eV, and 0.625 eV, respectively. Table 2 Main properties of a generic MNSR (CIAE, 1993). Parameter
Description
Reactor type Rated thermal power Fuel Fuel pin inner radius without cladding (mm) Fuel pin outer radius without cladding (mm) Fuel pin length (mm) Cladding composition U-235 percentage in the fuel pin U-235 enrichment Uranium density in fuel pin Core Shape Core diameter Core height Fuel element shape Fuel elements number in the core Reactor continuous operating time at rated power Refueling period Burn-up Control rod length (mm) Absorber of the control rod Total number of irradiation sites Number of internal irradiation sites Thermal neutron flux in outer irradiation sites
Tank-in-pool w30 kW UeAl4 dispersed in Al matrix 4.3 5.5 230 Al 24.83 w90%
Reactor cooling mode Radial reflector thickness
Cylinder 23 cm 23 cm Thin rod 347 2.5 h 10 years w1% 260 Cd 10 5 1 1011 n/cm2 s at rated power Natural convection 10 cm
Table 3 Main Reactor Neutronic Characteristics for the UeAl4eAl fuel (CIAE, 1993). Item Initial Excess Rea. (IER) with CRO (mk) Neutron Flux in (I.I.S.) (1012 n/cm2 s) Neutron Flux (E.I.S) (1012 n/cm2 s) Control Rod Worth (mk) Shut-Down Margin (mk) Effective Shut-Down Margin (mk)
Calculated 3.9840 1.004370 475,096 7.142800 3.215300 1.303600
Measured 3.9400 0.02 1.000000 1% 0.500000 1% 7.000000 0.01 3.060000 0.02 e
Since the results fit fairly well with experimental data, the system is used to calculate the new properties of these reactors (i.e. the case of UO2 fuel). The physical properties of the UO2 fuel used in the calculation are shown in Table 4. The table makes a comparison between the same physical properties for both the actual and the LEU UO2 fuel. The enrichment is reduced to 19.75% instead of about 90%. Since the UO2 fuel is used in power reactors, a large international experience is available for its use in research reactors. The behavior of the fuel under irradiation is well known and requires no special precautions especially at these reduced burn-ups in LPRRs. 3. Results and discussion If no changes are made on both the dimensions and the materials of the fuel and the cladding, the data of Table 4 will lead to the results shown in Table 5. A typical MNSR will have more than 45 mk IER instead of the 4 mk permitted by regulations (CIAE, 1993). Reference fluxes are the fluxes for the UeAl4eAl fuel in the same positions. The IER is too high. Some fuel rods should be removed and substituted by dummy elements such that only about 4 mk are available in the core. This is done in Table 6. The new core configuration renders available about 3.8531 mk as IER which is acceptable for the regulations as mentioned above. The new characteristics of the reactor are summarized in Table 7. Fluxes are not reported in this table since they have been shown in Table 5. Fifty fuel rods were substituted by as many dummy rods not to disturb or modify the thermo-hydraulic characteristics of the core. Flux values in the IIS and EIS decreased by 1e18.55%, while the CR worth decreased by more than 25%. The effective Shut-Down Margin now is only 1.3358 mk. The safety of the reactor seems to be a bit compromised if the UO2 fuel is used, but the use of this fuel is still acceptable because the Effective Shut-Down Margin (ESDM) is still negative. The ESDM is a parameter which we use to evaluate exactly the safety of the
Table 4 A Comparison between the physical properties of the HEU and LEU fuels (CIAE, 1993). Fuel Properties
UeAl4eAl
UO2
Meat Density (g/cm3) Dispersed Phase Density (g/cm3) Wt% U in Dispersed Phase U Dens. in Dispersed Phase (g/cm3) U density in Meat (g/cm3) Volumetric. Fraction of Dispersed Phase % Porosity % Enrichment % Al content % in the meat U-235 fraction in meat % U-238 fraction in meat % Si fraction in meat % Mo fraction in meat % Oxygen fraction in meat %
3.456 5.70 64.0 3.70 0.955 26.20 1.20 89.87 72.310 24.831 2.80 e e e
10.6 10.6 88.0 9.30 9.342 39.0 0.0 19.75 e 17.405 70.72 e e 11.8669
M. Albarhoum / Progress in Nuclear Energy 53 (2011) 73e75 Table 5 A Comparison between the actual fuel and the UO2 fuel characteristics. Fuel Type
UeAl4eAl
UO2
Reactor Neutronic Characteristics Initial Excess Reactivity with CRO (mk) Variation from IER (mk) Neutron Flux in IIS (1012 n/cm2.s) Variation from Reference Flux (%) Neutron Flux in EIS (1012 n/cm2.s) Variation from Reference. Flux (%) Control Rod Worth (mk) Shut-Down Margin (mk)
3.9840 e 1.004370 e 0.475096 e 7.142800 3.158800
45.420800 41.436800 0.817976 18.558300 0.4143230 12.791700 4.566000 e
Table 6 Core configuration of a typical MNSR that uses LEU UO2 as fuel. Initial Excess Reactivity (mk)
No. of Dummy Elements (DE)
Circles occupied by DE
Fuel rods
Circles occupied by fuel rods
3.8531
53
10
294
1,2,3,4,5,6,7,8,9,10
Reactor Neutronic Characteristics
dimensions and the cladding material were modified, the latter being changed to zircaloy. The lifetime of the reactor was extended by about 100%, and 98 fuel rods were saved (See Table 8). The total mass of uranium increased, as expected from 0.9 kg to 5.6 kg. The volumes of water are almost the same in both cases (HEU and LEU fuels). These data confirms the usability of the LEU UO2 as a fuel for LPRRs. 4. Conclusions The UO2 ceramic fuel has been used in both power and LPRRs as a LEU fuel. The long international experience facilitates its use in MNSRs and SLOWPOKEs as LPRRs. The number of fuel rods is decreased by about 50 in the case of MNSRs and by about 100 in the case of SLOPOKEs, and the number of dummy elements is increased correspondingly. These reactors, from the neutronic stand-point are safer in the case of UeAl4eAl fuel. To keep the MNSRs performance in terms of neutron flux values in the IIS and EIS, the nominal power of these reactors should be increased to about 56 kW. However the control rod worth decreases by about 25%. Acknowledgment
Table 7 Reactor properties when the LEU fuel (UO2) is used. Fuel: UO2
Initial Excess Reactivity with CRO (mk) Variation from IER (mk) Control Rod Worth (mk) Shut-Down Margin (mk) Effective Shut-Down Margin (mk)
75
3.85310 0.13090 5.188900 1.335800 0.4706000
Table 8 Comparison of the HEU-fuelled and the LEU-fuelled cores of SLOWPOKE-2 (Bennett and Nielsen, 2002). Item
HEU-fuelled
LEU-fuelled
Core diameter (mm) Number of fuel pins Fuel pin diameter with cladding (mm) Fuel length (mm) Cladding Fuel Total mass of uranium (kg) Enrichment U-235 (%) Total mass of uranium 235 (kg) Volume of water in core (L)
228 296 5.23 225 Aluminum UeAl 28% alloy 0.9 93 0.82 7.8
234 198 5.26 234 Zircalloy-4 UO2 5.6 19.89 1.12 8.1
reactor. It is based on the normal Shut-Down Margin (SDM), but takes into consideration the effect of the initiating events mentioned in the Safety Analysis Report (SAR) of the reactor (i.e. the effects of flooding the IIS and EIS). The reduction of the flux values in both the IIS and EIS requires the power of the reactor to be increased correspondingly (by w19%). The new nominal power should be about 36 kW. The LEU UO2 has been already used for other LPPRs, the SLOWPOKE-2 reactor (Bennett and Nielsen, 2002). In that case both
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