Safety Analysis of Leu Target Plate Irradiation at Parr-1 for Fission Molybdenum-99 Production
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SAFETY ANALYSIS OF LEU TARGET PLATE IRRADIATION AT PARR-1 FOR FISSION MOLYBDENUM-99 PRODUCTION T. Mahmood, I. H. Bokhari, A. Muhammad and M. Iqbal Nuclear Engineering Division, Directorate of Systems and Services, PINSTECH, Nilore, Islamabad Keeping in view the reservations of international community on utilization of high enriched uranium (HEU) fuel in civil nuclear reactors, it has been planned to design and irradiate low enriched uranium (LEU) fuel plates for Molybdenum-99 production at Pakistan Research Reactor-1 (PARR-1). Safety analysis for the proposed LEU target fuel plate, its irradiation in the core and transportation to the processing plant was performed. Neutronic analyses of the target holder bearing three fuel plates at equal distance from each other, was performed. The activity of the target plates was calculated. Effect of irradiation was studied by placing this holder at different axial positions [1-3]. It was concluded that the criteria of avoiding Onset of Nucleate Boiling (ONB) is fulfilled for irradiation of the target plates at any vertical position in the water box positioned at C-4. With such arrangement, the results showed that target holder irradiation had safe Departure from Nucleate Boiling Ratio (DNBR) which was greater than 2. The maximum temperature achieved was 104.7 ºC, which is about 21 ºC below the clad surface temperature that can initiate nucleate boiling. Fig. 1 shows the maximum activities of waste and 99Mo after irradiation. Fig. 1 Maximum activities of waste and 99Mo after irradiation References 1. T. Hamid, Reactor Kinetics Parameters as a Function of Fuel Burnup, PIEAS-445 (1999). 2. M. Iqba and T. Mahmood, Fission Moly Production at PARR-1, PINSTECH-192 (2005). 3. I. H. Bokhari, M. Israr and S. Pervez, “Analysis of reactivity induced accidents at Pakistan Research Reactor-1’, Ann. Nucl. Energy 29 (2002) 2225. 1 STEADY STATE ANALYSIS FOR IRRADIATION OF LOCALLY FABRICATED NATURAL URANIUM ELEMENT AT PARR-1 T. Mahmood, I. H. Bokhari and M. Israr Nuclear Engineering Division, Directorate of Systems and Services, PINSTECH, Nilore, Islamabad Efforts are underway at PINSTECH to fabricate the fuel elements locally, to be used in future cores of PARR-1. Neutronic and thermal hydraulic analyses were carried out for a proposed core comprising of the existing 34 imported LEU fuel elements and a locally fabricated natural uranium fuel element. Calculated neutronic parameters included excess reactivity, shut down margin, control rod worth, peak power density location, peaking factors, neutron flux in the fuel elements and neutron flux at irradiation sites in the core [1-3]. Calculated thermal hydraulic parameters included steady-state temperatures, peak temperatures for fuel centerline, clad surface & water coolant, and safety margins to Onset of Nucleate Boiling (ONB), Onset of Flow Instability (OFI) and Departure from Nucleate Boiling (DNB). Fig. 1 shows the axial thermal neutron flux profile with and without natural uranium element. From the analysis, it was concluded that present core can be operated safely at steady state power level of 10 MW along with the newly fabricated natural uranium standard fuel element placed at grid position E-3 or D-3. Core will have sufficient safety margins against onset of nucleate boiling, onset of flow instability and departure from nucleate boiling. Fig. 1 Axial thermal neutron flux profile with and without natural uranium element References 1 International Atomic Energy Agency, Research Reactor Core Conversion from the Use of Highly Enriched Uranium to the Use of Low Enriched Uranium Fuels, IAEA- TECDOC-233 (1980). 2 T. Hayat et al, Final Safety Analysis Report (FSAR) of PARR-1, (2001). 3 M.M. El-Wakil, Nuclear Heat Transport, International Textbook Company (1971). 2 OPTIMIZATION STUDY OF LEU FUELLED HOMOGENEOUS AQUEOUS SOLUTION NUCLEAR REACTORS FOR SHORT LIVED ISOTOPES PRODUCTION T. Mahmood and M. Iqbal Nuclear Engineering Division, Directorate of Systems and Services, PINSTECH, Nilore, Islamabad Low enriched uranium (LEU) based solution reactors possess the potential to meet increasing demand of 99Mo and other short lived fission product radioisotopes being used in medical field. In the current study, optimization and neutronic design calculations were carried out for LEU homogenous aqueous solution reactor with uranyl nitrate as a solution fuel. Lattice calculations and core modeling were performed employing available standard nuclear reactor codes WIMSD and CITATION. Calculation procedure was verified using experimental published data and simulated criticality results were compared with the published data [1-3]. After verification of the calculation methodology, optimization and design calculations were performed for four different uranium enrichments (5%, 10%, 15%, and 19.99%) of the uranyl nitrate solution. Keeping in view the restraints on peak power density of solution reactors, annular geometry is suited as compared to the cylindrical geometry. Among the four considered enrichments in the analysis, 19.99% enriched uranyl nitrate solution with annular geometry presents optimum design parameters for the homogeneous aqueous solution reactor. Critical fissile mass profile as a function of uranium concentration for annular geometry is shown in Fig. 1. Fig. 1 Critical fissile mass profile as a function of uranium concentration for annular geometry References 1 International Atomic Energy Agency, Homogeneous Aqueous Solution Nuclear Reactors for the Production of Mo-99 and other Short Lived Radioisotopes, IAEA- TECDOC-1601, Vienna (2008). 2 S. Sakurai, M. Miyauchi and S. Tachimori, J. Nucl. Sci. Tech. 24 (1987) 415 3 Y. Yamane, Y. Miyoshi, S. Watanabe and T. Yamamoto, Nucl. Tech. 141 (2003) 221. 3 ANALYSIS FOR CORE CONVERSION (HEU-LEU) OF PARR-2 T. Mahmood, I. H. Bokhari and M. Iqbal Nuclear Engineering Division, Directorate of Systems and Services, PINSTECH, Nilore, Islamabad Calculational methodology for conversion of Miniature Neutron Source Reactor (MNSR) from HEU to LEU was validated by doing analysis of HEU fuel (90.2% enriched). On the basis of HEU based reactor model, analysis of LEU (UO2 fuel) core gives results, which qualify the UO2 fuel for future LEU core of MNSR. However for LEU fuel, neutron flux at irradiation sites is slightly lower for the reactor operating at 30 kW power. Therefore reactor power will have to be increased to a level of 33 kW to get the same thermal flux values as obtained for HEU core. Use of the same control rod as being used in the current HEU core gives lower values of shut down margin and control rod worth. But the slightly increased diameter of control rod improves shut down margin to a value that is comparable to the corresponding value for HEU core. LEU (UO2 fuelled) core with following characteristics provides replica of the currently operating HEU core: Enrichment: 12.46% Guide tube and grid plate material: Zr-4 Reactor power: 3.3kW Cladding material of fuel pin: Zr-4 Control rod absorber (cadmium) thickness: 4.5mm All other materials and structures have been assumed to be same as are being used in the presently operating HEU core. There is no significant difference between the dose values for HEU and prospected LEU fuel. Therefore existing HEU core and prospected LEU core of MNSR are considered to be safe for the public even in case of an accident releasing radioactive gases from the fuel [1-3]. Core characteristics are shown in Table 1. Table 1: Core characteristics Amount Criticality Control Thermal Flux Excess Shut Down Fuel Material / of 235U Position Rod at Inner Sites U- Density Reactivity Margin Enrichment (g) in (Rod cm Worth at 30 kW (mk) (mk) Core out) (mk) (#/cm2-sec) U-Al alloy / 0.92g/cm3 995 9.0 4.046 -2.344 -6.39 1.02E+12 90.2% enriched UO fuel / 2 9.35g/cm3 1353 7.0 4.007 -1.43 -5.437 9.36E+11 12.6% enriched UO fuel / 2 9.35g/cm3 1264 7.0 4.160 -1.498 -5.658 9.41E+11 12.3% enriched UO fuel / 2 9.35g/cm3 1339 8.5 4.012 -2.375 -6.387 9.16E+11 12.46% enriched References 1. M. K. Qazi, M. Israr and A. Karim, Revised Final Safety Analysis Report (FSAR) on Pakistan Research Reactor-2 (1994). 2. J. S. A. Jonah, J. R. Liaw, A. Olson and J. E. Matos, “Criticality calculations and transient analysis of the Nigeria MNSR (NIRR-1) for conversion to LEU”, (2006). 3. S. S. Raza and M. Iqbal, Ann. Nucl. Energy 32 (2005) 1157. 4 BENCHMARK NEUTRONIC ANALYSIS OF SYRIAN MNSR T. Mahmood and M. Iqbal Nuclear Engineering Division, Directorate of Systems and Services, PINSTECH, Nilore, Islamabad Benchmark neutronic calculations were carried out for the Syrian Miniature Neutron Source Reactor (MNSR). Ten energy group lattice constant calculations were performed employing WIMSD code while entire core with beryllium reflector and irradiation sites was modeled in CITAION. Calculated neutronic parameters included core excess reactivity, control rod worth, thermal neutron flux at irradiation sites, peaking factors, feed back moderator temperature coefficient of reactivity, and axial & radial profiles for power densities [1-3]. Calculated values of these parameters were compared with the quoted/published data. Values of calculated excess reactivity, control rod worth and neutron flux at irradiation sites match reasonably well with the published data. However there is some difference between calculated and quoted values of peak power density and moderator temperature coefficient of reactivity. Axial power distribution for different control rod positions is shown in Fig. 1. Fig. 1 Axial power distribution for different control rod positions References 1. M. J. Halsall, “A Summary of WIMSD4 Input Options”, AEEW-M 1327 (1980). 2. T. B. Fowler, D. R. Vondy and G. W. Cunningham, “Nuclear Reactor Core Analyses Code CITATION”, ORNL – TM – 2496, Rev. 2. (1971). 3. E. Alhassan, E. H. K. Akaho, B. J. B. Nyarko, N. A. Adoo, V. Y. Agbodemegbe, C.