Invariability of neutron flux in internal and external irradiation sites in MNSRs for both HEU and LEU cores

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1 Indian Journal of Pure & Applied Physics Vol. 49, February 2011, pp Invariability of neutron flux in internal and external irradiation sites in MNSRs for both HEU and LEU cores M Albarhoum Department of Nuclear Engineering, Atomic Energy Commission, P. O. Box, 6091, Damascus- Syria Received 24 August 2010; revised 3 January 2011; accepted 11 January 2011 The effect of the control rod position on the values of the neutronic flux in the internal and external irradiation sites for the actual high enriched uranium UAl 4 -Al fuel and U 3 Si 2, U 3 Si, UO 2, and U-9Mo low enriched uranium fuels has been investigated by using a 3-D model for MNSRs. Results show that the values of the neutron flux in these sites depend only on the radial distance of these sites from the control rod axis, but not on the control rod vertical position. The standard deviation and the variance coefficient are all less than E12 and 0.812%, respectively except for the UO 2 fuel for which these two parameters are and 1.938%, respectively. These calculated values depend on the fuel type and enrichment for the same power of the core and not on the thickness of the annulus reflector. The invariability of the neutronic flux values with the vertical position of the control rod renders, MNSRs optimum tools for neutron activation analysis. Keywords: Fuel, MNSR, Initial excess reactivity, Control rod, LEU 1 Introduction Miniature Neutron Source Reactors (MNSRs) are research reactors of the tank-in-pool type using Highly Enriched Uranium (HEU) fuel (~90% w/o U 235 ). Their nominal power is about 30 kw and have a fuel cage formed of 10 circles and one central control rod for the regulation and shutdown of the reactor 1,2. Their primary use is the Neutronic Activation Analysis (NAA). This activity uses 10 sites for sample irradiation, five of which are called Internal Irradiation Sites (IIS) (or inner sites) and located inside the Annulus Reflector (AR) and five are called External Irradiation Sites (EIS) (or outer sites), and located outside the annulus reflector as shown in Fig. 1. Maximum values for the thermal neutron fluxes in these sites are: 1.E12 n/cm 2.s, and 0.5E12 n/cm 2.s, respectively at rated power 3. A noticeable physical characteristic of MNSRs is that the increase in temperature and the variation of the Control Rod (CR) position during reactor operation do not affect appreciably the neutron flux distribution in the whole reactor, IIS and EIS included, except in the space surrounding the CR position itself 5. This property (the invariability of the neutron flux in the IIS and EIS) allows to MNSRs to be a stable and reliable tool for NAA. Now since the distribution of the neutron flux in the whole reactor is not so important for the NAA, but only in the IIS and EIS where samples are irradiated, it may be a good question whether this property of MNSRs depend on the fuel type, or on its enrichment (where four low enriched uranium fuels were examines) from one side, and whether this depends on the thickness of reflector or even the presence of the IIS within the AR or not from the other side. This characteristic has been investigated for the aforementioned 4 fuel types namely: U 3 Si 2, U 3 Si, U-9Mo, and UO 2. 2 Methodology The neutronics calculations were performed using a detailed three-dimensional model implemented in the core calculation s code 1-4 CITATION coupled with the lattice cell code WIMSD4 through the BMAC package. The BMAC package was used to generate the Group Constants (GC) through the WIMSD4 code and then it was used to transfer the calculated GC to CITATION code which was used to calculate the neutronic flux and power distributions in the reactor. Four neutronic groups were used for the model. Their upper energy limits were: 10 MeV, MeV, 5530 ev, and ev, respectively. The GC for the 1 st fuel circle are presented in Table 1 for the 4 fuels. The criticality calculations for one configuration of the reactor were performed and the Initial Excess reactivity (IER) was determined. The position of the CR was then varied vertically along its axis for the whole CR path (23 cm), and the corresponding flux values in the IIS and EIS were found. This was repeated for all LEU fuel types in addition to the

2 84 INDIAN J PURE & APPL PHYS, VOL 49, FEBRUARY 2011 Fig. 1 Location of inner and outer irradiation sites of PARR-2(Ref.6) Table 1 Calculated group constants for the reflectors of the Syrian MNSR UAl 4 -Al Group 1 Group 2 Group 3 Group 4 Diffusion (cm) Absorption (cm 1 ) U 3 Si 2 Group 1 Group 2 Group 3 Group 4 Diffusion (cm) Absorption (cm 1 ) U 3 Si Group 1 Group 2 Group 3 Group 4 Diffusion (cm) Absorption (cm 1 ) U-9Mo Group 1 Group 2 Group 3 Group 4 Diffusion (cm) Absorption (cm 1 ) UO 2 Group 1 Group 2 Group 3 Group 4 Diffusion (cm) Absorption (cm 1 ) original HEU fuel and for all CR positions. Some statistical parameters (such as the average value of the thermal neutron flux (ANF), the Standard Deviation (STDEV) and the Variance Coefficient (VC)) indicating the variability of the obtained data around their average value were then calculated for each group and compared so as to draw conclusions about the argument. All calculations were made for the nominal power of 30 kw. Calculations were also made for the reflector thickness, which was in turn diminished and new Fuel Rods (FR) were added so that the IIS became outside the AR and in the core of the reactor. The IIS were thus transferred from the AR to the core of the reactor to see the effect of the thickness of the AR on the values of the ANF in the IIS and EIS.

3 ALBARHOUM: INVARIABILITY OF NEUTRON FLUX IN MNSRs 85 3 Results and Discussion 3.1 UAl 4 -Al Core The thermal (4 th group) neutron fluxes in both the IIS and EIS for the actual UAl 4 -Al HEU fuel are presented in Table 2 and shown in Fig. 2. The average Table 2 Calculated thermal neutron flux values in the IIS and EIS at different CR positions for the UAl 4 -Al fuel CR position Flux (IIS) Flux (EIS) (cm) (n/cm 2.s) (n/cm 2.s) E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E12 value of the thermal Neutron Flux (ANF) at the IIS (the sum of all flux values in column 2 of Table 2 divided by their number which is 23) was E12 n/cm 2.s and the STDEV of its values was equal to E12, and the Variance Coefficient (VC) was 0.301%, while the ANF, STDEV, and the VC at the EIS were equal to E12 n/cm 2.s, E12, and 0.350%, respectively. Hence, the flux value in both the IIS and EIS is practically constant since both the STDEV and the VC are very small. The values of the thermal neutron flux for the case of the U 3 Si 2 and U 3 Si LEU fuels were presented in Table U 3 Si 2 LEU Core From Table 3, the values of ANF, STDEV and VC for the U 3 Si 2 LEU fuel in the IIS were: E12 n/cm 2.s, E12 and 0.280%, respectively. The same parameters for the EIS were: E12 n/cm 2.s, E12 and 0.319%, respectively. The case of the U 3 Si 2 LEU fuel was also similar to the UAl 4 -Al HEU fuel. The obtained values for the ANF, STDEV and VC indicate that for this fuel the reactor property was valid as well. 3.3 U 3 Si LEU Core For the case of the U 3 Si LEU fuel, the values of ANF, STDEV and VC in the IIS were: E12 n/cm 2.s, E12 and 0.288%, respectively (Table 3). The same calculated parameters for the EIS were: E12 n/cm 2.s, E12 and Fig. 2 Distribution of the thermal neutron flux in the IIS and EIS versus CR position in the core for the UAl 4 -Al HEU fuel

4 86 INDIAN J PURE & APPL PHYS, VOL 49, FEBRUARY 2011 Table 3 Thermal neutron flux values in the IIS and EIS at different CR positions for the U 3 Si 2 and U 3 Si LEU fuels Fuel type U 3 Si 2 U 3 Si CR position Flux (IIS) Flux (EIS) Flux (IIS) Flux (EIS) (cm) (n/cm 2.s) (n/cm 2.s) (n/cm 2.s) (n/cm 2.s) E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E %, respectively. The STDEV and VC values were so small to consider the CR position not affecting the ANF in both the IIS and EIS. 3.4 UO 2 and U-9Mo LEU Cores The case of the UO 2 LEU and U-9Mo fuels is presented in Table 4. The ANF, STDEV and VC values for the case of the UO 2 LEU fuel in the IIS were: E12 n/cm 2.s, E12 and 0.262%, respectively as presented in Table 4, while the same parameters for the EIS were: E12 n/cm 2.s, E12 and %, respectively. For this fuel in the EIS, the VC was about 2%, while it was less than unity for all the preceding fuels. However, this value is still small not to make a significant conclusion about this fuel type. From the other side, the ANF, STDEV and VC for the case of the U-9Mo LEU fuel in the IIS were: E12 n/cm 2.s, E12 and 0.812%, respectively, while the same parameters for the EIS were: E12 n/cm 2.s, E12 and %, respectively. Although the absolute value of the neutron flux in both the IIS and EIS is about 20% less than that of theual 4 -Al HEU fuel, the values of the STDEV and VC are so small that this fuel can be treated like the other LEU fuel types except for the UO 2 fuel type. The results obtained for the different LEU fuels for the IIS are shown in Fig. 3. It can be concluded that the effect of the CR position on the thermal neutron flux value in the IIS and EIS does not depend on the fuel type. 3.5 Annulus Reflector Thickness effect The effect of the AR thickness on this property has been investigated. This will be done for the actual HEU fuel only. The Internal Radius (IR) of the AR was varied from cm (the actual IR) to cm and the values of the ANF in the IIS and EIS were calculated. The results are presented in Table 5 and shown in Fig. 4. The values of the statistical parameters, i.e. the ANF, STDEV and VC for the UAl 4 -Al HEU fuel are presented in Table 6. The ANF in the IIS was a bit varying with the internal radius of the AR, but the STDEV and the VC indicated that these flux values were really not affected by the CR position (Table 6). In Table 7, the values of the same

5 ALBARHOUM: INVARIABILITY OF NEUTRON FLUX IN MNSRs 87 Table 4 Thermal neutron flux values in the IIS and EIS at different control rod positions for the UO 2 and U-9Mo fuels Fuel type UO 2 U-9Mo CR position Flux (IIS) Flux (EIS) Flux (IIS) Flux (EIS) (cm) (n/cm 2.s) (n/cm 2.s) (n/cm 2.s) (n/cm 2.s) E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E E12 Fig. 3 Comparison among the actual HEU and the different LEU fuels distribution of the thermal neutron flux in the IIS and EIS versus CR position in the core

6 88 INDIAN J PURE & APPL PHYS, VOL 49, FEBRUARY 2011 Table 5 Thermal neutron flux values in the IIS at different control rod positions and increasing values for the internal diameter of the annulus reflector X (cm) Internal radius of the annulus reflector(cm) Fig. 4 Variation of the thermal neutron flux in the IIS versus CR position for different internal radii of the annulus reflector

7 ALBARHOUM: INVARIABILITY OF NEUTRON FLUX IN MNSRs 89 Table 6 Values of some statistical parameters for the neutron thermal flux as a function of the CR position in the IIS at different control rod positions and increasing values for the internal diameter of the AR Item Internal radius of the annulus reflector(cm) ANF (n/cm 2.s) E STDEV VC (%) Table 7 Values of some statistical parameters for the neutron thermal flux as a function of the CR position in the EIS at different CR positions and increasing values for the internal diameter of the AR Item Internal radius of the annulus reflector(cm) ANF (n/cm 2.s) E STDEV VC (%) statistical parameters for the EIS are presented. The decrease of the IR of the AR will cause the IER to decrease as well. The addition of new Fuel Rods (FR) will be necessary to recover the lost IER. Now, if the FR were increased to have an acceptable IER it can be shown that 228 FR in the 10 th circle would be required. In this case, the values of the neutron flux in the IIS and EIS will be those as presented in Table 8 with the presence of 3 Dummy Rods (DR) which are made of aluminium and have the same dimensions of the FR. In Table 8, X indicate the CR position. For this configuration of the core the values of the ANF, STDEV and VC were: E12 n/cm 2.s, E12 and 0.387%, respectively. It can be noticed that the IIS are now in the core of the reactor (and not in the AR) the thermal neutron flux was practically constant with respect to the variation of the CR position. The main neutronic parameters of the reactor are presented in Table 9. Although the fluxes in the IIS and EIS reduced by about 27%, the new core conserved a negative Shut Down Margin (SDM) and an even better Effective Shut Down margin (ESDM). Since this core was acceptable from the neutronic point of view, for the thermal hydraulics it needed some DR to be added to have an acceptable sub-channel shape. This can be satisfied if other 183 FR were added besides to other 369 new DR. This lead to the neutronic characteristics as presented in Table 10. This final configuration satisfied all safety requirements, but had a value of the neutron flux decreased by about 42% with respect to the original reference value (the value of the UAl 4 -Al HEU fuel). The core contained now about 50% more 235 U than the original one. For this configuration, the values of Table 8 Thermal neutron flux values in the IIS and EIS at different control rod positions when 8 cm of Be are substituted by fuel rods X Flux in IIS Flux in EIS (cm) (n/cm2.s) (n/cm2.s) the NF versus CR position are presented in Table 11. Since the ANF, STDEV and VC were: E12 n/cm 2.s, 0.002E12 and 0.387%, respectively for the IIS, and E12 n/cm 2.s, 0.002E12 and 0.499%, respectively for the EIS, this proves that this property is independent of the fuel type and of the reflector

8 90 INDIAN J PURE & APPL PHYS, VOL 49, FEBRUARY 2011 Table 9 Main neutronic characteristics of the new and old cores for 228 FR and 3 DR Item Calculated Measured Initial Excess Rea. (IER) with CRO (mk) ± 0.02 Neutron Flux in (I.I.S.) (10 12 n/cm 2.s) ± 1% Neutron Flux(E.I.S) (10 12 n/cm 2.s) ± 1% Control Rod Worth (mk) ± 0.01 Shut-Down Margin (mk) ± 0.02 Flooding the IIS with water (mk) Flooding the EIS with water (mk) Effective Shut-Down Margin (mk) Mass of 235 U in core (kg) Table 10 Main neutronic characteristics of the new and old cores for 245 FR and 372 DR Item Calculated Measured Initial Excess Rea. (IER) with CRO (mk) ± 0.02 Neutron Flux in (I.I.S.) (10 12 n/cm 2.s) ± 1% Neutron Flux(E.I.S) (10 12 n/cm 2.s) ± 1% Control Rod Worth (mk) ± 0.01 Shut-Down Margin (mk) ± 0.02 Flooding the IIS with water (mk) Flooding the EIS with water (mk) Effective Shut-Down Margin (mk) Mass of 235 U in core (kg) Table 11 Thermal neutron flux values in the IIS and EIS at different control rod positions when 8 cm of Be are substituted by fuel rods X Flux in IIS Flux in EIS (cm) (n/cm 2.s) (n/cm 2.s) thickness and type. The CR position does not affect the values of the neutron flux in the IIS and EIS anyway because the positions of these sites are far from the CR axis although the values of the neutron flux depend on the fuel type for the same power (30 kw). 4 Conclusions The control rod vertical position along its path does not affect the value of the thermal neutron flux in the IIS and EIS because of the considerable distance of these sites from the CR axis, but the value of the neutron flux in these sites is affected by the fuel type and enrichment. The thickness of the reflector does not affect the ANF in both the IIS and EIS as well. The maximum variation of the ANF values occur in the case of the UO 2 LEU fuel. Acknowledgement The author thanks Prof I Othman, Director General of the Atomic Energy Commission of Syria, for his encouragement and continued support. References 1 Albarhoum M, Annals of Nuclear Energy,35 (2008) Askew J R, Fayers F J & Kemshell P B, J British Nucl Energy Soc, 5 (1966) CIAE Safety Analysis Report (SAR) for the Syrian Miniature Neutron Source Reactor, China (1993). 4 Fowler T B, Vondy D R & Cunningham G W, Nuclear Reactor Core Analysis Code: CITATION ORNL-TM-2496, Rev 2, July (1971). 5 Khattab K, Omar H & Ghazi N, Nucl Engineering and Design, 236 (23) (2006) Tayyab M, Showket P & Masood, I Annals of Nucl Energy (35) (2008) 1440.