99mTc is a very useful radioisotope in medical diagnostic procedure. 99mTc is produced from 99Mo decay. Currently, most of 99Mo is produced by irradiating 235U in the nuclear reactor. 99Mo mostly results from the fission reaction of  235U targets with a fission yield about 6.1%. A small additional amount is created from 98Mo neutron activation. Actually 99Mo is also created in the reactor fuel, but usually we do not extract it. The fuel will become spent fuel which is a highly radioactive waste. 99Mo production system in the aqueous homogeneous reactor offers a better method, because all of the 99Mo can be extracted from the fuel solution. Fresh reactor fuel solution consists of uranyl nitrate dissolved in water. There is no separation of target and fuel in an aqueous homogeneous reactor where target and fuel become one liquid solution, and there is no spent fuel generated from this reactor. Simulation of the extraction process is performed while reactor in operation (without reactor shutdown). With an extraction flow rate of 3.6 L/h, after 43 hours of reactor operation the production of 99Mo is relatively constant at about 98.6 curie/hou

A. Isnaeni

M.S. Aljohani

S.I. Bhuiyan

T.G. Aboalfaraj

Atom Indonesia

99mTc is a very useful radioisotope in medical diagnostic procedure. 99mTc is produced from 99Mo decay. Currently, most of 99Mo is produced by irradiating 235U in the nuclear reactor. 99Mo mostly results from the fission reaction of  235U targets with a fission yield about 6.1%. A small additional amount is created from 98Mo neutron activation. Actually 99Mo is also created in the reactor fuel, but usually we do not extract it. The fuel will become spent fuel which is a highly radioactive waste. 99Mo production system in the aqueous homogeneous reactor offers a better method, because all of the 99Mo can be extracted from the fuel solution. Fresh reactor fuel solution consists of uranyl nitrate dissolved in water. There is no separation of target and fuel in an aqueous homogeneous reactor where target and fuel become one liquid solution, and there is no spent fuel generated from this reactor. Simulation of the extraction process is performed while reactor in operation (without reactor shutdown). With an extraction flow rate of 3.6 L/h, after 43 hours of reactor operation the production of 99Mo is relatively constant at about 98.6 curie/hour./hour.Received: 11 January 2014; Revised: 18 February 2014; Accepted: 28 February 201

Isnaeni, A. (A)

Aljohani, M. S. (M)

Aboalfaraj, T. G. (T)

Bhuiyan, S. I. (S)

Neliti

Analysis of 99Mo Production Capacity in Uranyl Nitrate Aqueous Homogeneous Reactor Using ORIGEN and MCNP

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A. Isnaeni, at al., / Atom Indonesia Vol. 40 No. 1 (2014)  Analysis of 99Mo Production Capacity in Uranyl Nitrate Aqueous Homogeneous Reactor using ORIGEN and MCNP  A. Isnaeni*, M.S. Aljohani, T.G. Aboalfaraj and S.I. Bhuiyan Nuclear Engineering Department, King Abdulaziz University  P.O. Box 80240 Jeddah, Kingdom of Saudi Arabia   A R T I C L E   I N F O  A B S T R A C T   Article history: Received  11 January 2014 Received in revised form  18 February 2014 Accepted 28 February 2014   Keywords: 99Mo Uranyl nitrate Homogeneous reactor MCNP ORIGEN   99mTc is a very useful radioisotope in medical diagnostic procedure. 99mTc is produced from 99Mo decay. Currently, most of 99Mo is produced by irradiating 235U in the nuclear reactor. 99Mo mostly results from the fission reaction of  235U targets with a fission yield about 6.1%. A small additional amount is created from 98Mo neutron activation. Actually 99Mo is also created in the reactor fuel, but usually we do not extract it. The fuel will become spent fuel which is a highly radioactive waste. 99Mo production system in the aqueous homogeneous reactor offers a better method, because all of the 99Mo can be extracted from the fuel solution. Fresh reactor fuel solution consists of uranyl nitrate dissolved in water. There is no separation of target and fuel in an aqueous homogeneous reactor where target and fuel become one liquid solution, and there is no spent fuel generated from this reactor. Simulation of the extraction process is performed while reactor in operation (without reactor shutdown). With an extraction flow rate of 3.6 L/h, after 43 hours of reactor operation the production of 99Mo is relatively constant at about 98.6 curie/hour.  © 2014 Atom Indonesia. All rights reserved   INTRODUCTION  99mTc is a very useful radioisotope in medical diagnostic procedure; it is used in nearly 80% of all nuclear medicine procedures [1]. 99mTc is produced from 99Mo decay. Because of the short half-life of 99mTc (6.0058 hours) it is not transported to hospitals around the world, but 99Mo which has a longer half life (65.94 hours) is delivered instead. The total production and consumption of 99Mo in the world is approximately 400 TBq/week [2]. Global demand for 99mTc is expected to grow at an average annual rate of 3–8% [3]. The reduction of uranium enrichment from ∼90% to ∼19.8% demands modifications on process operation to compensate for the resulting loss in output [4]. The reduction of uranium enrichment is due to nuclear non-proliferation. Currently most of 99Mo is being produced in research and test reactors by the irradiation of targets containing enriched fissile material 235U [5]. 99Mo is extracted using an acidic process [6]. This process generates waste. Figure 1 shows that 99Mo is mostly produced from the fission reaction of 235U targets with a fission yield about 6.1%, while a small additional amount is created from the neutron                                                  Corresponding author. E-mail address:arifisnaeni@gmail.com activation of 98Mo which is itself created from 235U target fission reaction. Actually 99Mo is also created in the reactor fuel, but it is not usually extracted. The fuel will instead become a highly radioactive waste known as spent fuel.   Fig. 1.99Mo production.  99Mo production system in the aqueous homogeneous reactor (AHR) offers a better method, because all of the 99Mo can be extracted from the reactor solution. One such aqueous homogeneous reactor was built in the past; it was operated almost daily as a neutron source from 1951 until its deactivation in 1974, after 23 years of safe, reliable operation [7]. Fresh reactor fuel solution consists of uranyl nitrate dissolved in water. The production of 99Mo in the AHR follows the same nuclear reactions shown in Fig. 1. However, there is no separation of target and fuel in the AHR, as the target and the fuel become one liquid solution. Therefore, there is no 98Mo Activation 99Mo 99mTc 99Tc 6.0058 h 65.94 h  uranium solution235U  Fission  Product Atom Indonesia Vol. 40 No. 1 (201 ) 40 - 43     40 A. Isnaeni, at al., / Atom Indonesia Vol. 40 No. 1 (2014)  spent fuel generated from this reactor after 99Mo extraction from the reactor solution; the remains of the extraction results will be returned to the reactor core as the fuel solution for further operation. The use of fuel solution reactors for the production of medical isotopes is potentially advantageous because of their characteristics such as: low cost; small critical mass (low power); simple fuel handling, processing and purification characteristics; and inherent passive safety [8].    The void volume created by bubbles in the solution core will introduce a strong negative reactivity feedback [9].   EXPERIMENTAL METHODS  We use MCNP5 for reactor criticality analysis and ORIGEN2.2 for nuclide inventory analysis. MCNP is a general-purpose Monte Carlo N–Particle code which can be used for neutron, photon, electron, or coupled neutron/photon/ electron transport and possesses the capability to calculate eigenvalues for critical systems [10]. ORIGEN is a computer code system for calculating the buildup, decay, and processing of radioactive materials [11]. We couple both MCNP5 and ORIGEN2.2. We started by making MCNP input which consists of fuel, reactor vessel and reflector, then ran the input. The fuel solution height was gradually increased until the reactor became supercritical (MCNP output). After the reactor became supercritical we started to extract 99Mo in a close loop. We extracted the 99Mo from the fuel and then returned the remains of the extraction in the reactor core. The 99Mo extraction was simulated in ORIGEN2.2. The simulation flow chart is shown   in Fig. 2.     Fig. 2. Simulation flow chart. The inputs of ORIGEN 2.2 are power (200 kW), time (one hour), and n % of fuel solution, where n % depends on the ratio of the fuel solution extracted per hour to all of the fuel solution; the ratio depends on the extraction flow rate and the volume of the fuel solution. The 99Mo extraction is capable of diverting 0.1 to 1.0 mL/second flow of uranyl nitrate solution [12]. The extraction flow rate that we used in this simulation is 1.0 mL/second (Table 1).  Table 1. Extraction flow rate.   Flow rate(mL/s) Flow rate(L/h) Extraction flow rate 1 3.6  The reactor core parameters are shown in Table 2. The fuel solution consists of 20% enrichment uranyl nitrate dissolved in water. The atom densities of the fuel solution are shown in Table 3. The reactor vessel atom densities are shown in Table 4.  Table 2. Reactor core parameters.  Parameter Value Reactor power (thermal) 200 kW  Fuel solution  Uranyl nitrate Enrichment 20% Inner core diameter (cm) 30 Reactor height (cm) 100 Reactor vessel Stainless steel-304 Vessel thickness (cm) 0.5 Reflector  (radial) Beryllium   Reflector thickness (cm) 30  Table 3. Atom densities in fresh fuel.  Isotope atom/barn.cm 235U 1.26504531144E-04 238U 5.07525204789E-04 16O 3.34878465916E-02 14N 1.26805947187E-03 1H 5.68312174084E-02  Table 4. Stainless steel-304 [13].  Nuclide atom/barn.cm Chromium  1.74 × 10-2 Manganese 1.52 × 10-3 Iron 5.81 × 10-2 Nickel 8.51 × 10-3  We used MCNP5 for criticality calculation.                   The geometrical model of the reactor on the Visual Editor of MCNP5 is shown in Fig. 3. MCNP5 input material (uranium solution), geometry Start k > 1 ORIGEN input n%  of fuel solution, power, time 99Mo ORIGEN output Non 99Mo radionuclide 99Mo results Finish  No,  increase  volume   No  (non   99Mo) Yes  ( 99Mo) 41 A. Isnaeni, at al., / Atom Indonesia Vol. 40 No. 1 (2014)    Fig. 3. Geometrical model of the reactor, from the side of the reactor (left) and from the top of the reactor (right). The reactor consists of uranyl nitrate solution (red), reactor vessel (blue) and beryllium reflector (yellow).   RESULTS AND DISCUSSION  The first step we performed was to obtain the condition where keff> 1 was barely achieved, or, in other words, to determine the height of the uranyl nitrate solution because keff is a function of fuel solution height. Using MCNP we made a variation of uranyl nitrate solution height. The result of the simulation is shown in Table 5.  Table 5. Fuel solution height vs keff.  No Fuel solution height (cm) keff 1 10 0.58154 2 15 0.75903 3 20 0.86390 4 25 0.93485 5 30 0.98914 6 35 1.02461 7 40 1.05394 8 45 1.06629 9 50 1.08568 10 55 1.10217 11 60 1.11266 12 65 1.11608 13 70 1.12535  From the result we decided to use keff = 1.02461 for which the height of uranyl nitrate solution in the core is 35 cm. Since the inner diameter of the core is 30 cm, the volume of the fuel solution is 24.7275 L. The extraction flow rate is 3.6 L/h. Therefore, the ratio of the fuel solution which is extracted per hour to all fuel solution in the core is:                           Thus, 14.59% of all fuel solution is extracted per hour. ORIGEN2.2 simulates the fuel extraction. Of the 99Mo in the fuel, 14.59% is extracted as the 99Mo result per hour, while the other nuclide and 85.41% of 99Mo remains in the core as the fuel solution. In this simulation this result becomes ORIGEN2.2 and MCNP5 input for the next hours. The resulting 99Mo has to be decayed for one hour because this extraction process takes one hour. The 99Mo result after the decay can be seen in Fig. 4.   Fig. 4. 99Mo production capacities per hour.  After 43 hours of reactor operation the production rate of 99Mo is relatively constant at about 98.6 Ci/h. The 235U in the fuel solution was consumed during reactor operation; the decreasing atom density of 235U in the fuel can be seen in Fig. 5.    Fig. 5. 235U atom density (atom/barn.cm).  Criticality analysis in MCNP5 shows that the reactor is supercritical for 96 hours of reactor operation. The keff during reactor operation is shown in Fig. 6.    Fig. 6. keff during reactor operation. 0204060801001200 20 40 60 80 100 99Mo product (Ci/h) Time (h) 1,2638E-041,2640E-041,2642E-041,2644E-041,2646E-041,2648E-041,2650E-041,2652E-040 20 40 60 80 100235U (atom/barn.cm) Time (hours) 0,900,920,940,960,981,001,021,040 20 40 60 80 100keff Time (h) 42 A. Isnaeni, at al., / Atom Indonesia Vol. 40 No. 1 (2014)  Figure 7 shows the accumulation or total of 99Mo product for 720 hours of reactor operation.   Fig. 7. Total 99Mo Product (Ci).  The product accumulation is increasing with reactor operating time, but after reaching about 400 hours it relatively constant. Because the accumulation result of 99Mo will decrease by decay process, it is better to extract the product as often   as possible.   CONCLUSION  Reactor simulation on MCNP5 shows that the reactor is supercritical during reactor operation.  The 99Mo extraction is performed during reactor operation with a flow rate of 3.6 L/h. After 43 hours of reactor operation the 99Mo production is relatively constant at about 98.6 Ci/h. Since the accumulation result of 99Mo will decrease by decay process, it is better to extract the result as often      as possible.   REFERENCES  1. A.J. Youker, S.D. Chemerisov, M. Kalensky   et al., Sci. Technol. Nucl. Install. 2013 (2013) 1.  2. B.L. Zhuikov, Appl. Radiat. Isot. 84 (2013) 48. 3. Anonymous, Non-HEU Production Tech-nologies for Molybdenum-99 and Technetium-99m, IAEA Nuclear Energy Series No. NF-T-5.4, IAEA (2013) 1.  4. A.H.A. Sameh, Sci. Technol. Nucl. Install. 2013 (2013) 1. 5. Tayyab Mahmood and Masood Iqbal,          Ann. Nucl. Energy 42 (2012) 175. 6. C.K.W. Cheung, E.(Lou)R. Vance, M.W.A. Stewart et al., Procedia Chem. 7 (2012) 548. 7. A.G. Buchan, C.C. Pain, A.J.H. Goddard et al., Ann. Nucl. Energy. 48 (2012) 68. 8. Anonymous, Homogeneous Aqueous Solution Nuclear Reactors for the Production of 99Mo and other Short Lived Radio Isotopes, IAEA – TECDOC–1601, IAEA (2008) 1. 9. Y. Li, H. Wu, L. Cao et al., Nucl. Eng. Des.  240 (2010) 763. 10. Anonymous, A General Monte Carlo N-Particle Transport Code, Version 5, Los Alamos National Laboratory, MCNP, California   (2003) 5. 11. UT-Battelle, RSICC Computer Code Collection Origen 2.2 Isotope Generation and Depletion Code Matrix Exponential Method, Oak Ridge National Laboratory, Tennessee (2002) 3. 12. Ball and M. Russell, Medical isotope production reactor, US Patents 5596611A (1997). 13. S.I. Bhuiyan, M. Musa, G. Ara et al., ANISN–A Multigroup Discrete Ordinates Transport Code with Anisotropic Scattering and Its Use in Reactor Physics, Institute of Nuclear Science and Technology Atomic Energy Research Establishment, Dhaka (1987) 23.   0100020003000400050006000700080009000100000 200 400 600Total 99Mo product (Ci) Time (h) 43 

Analysis of 99Mo Production Capacity in Uranyl Nitrate Aqueous Homogeneous Reactor using ORIGEN and MCNP

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Analysis of 99Mo Production Capacity in Uranyl Nitrate Aqueous Homogeneous Reactor using ORIGEN and MCNP

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