Analysis of Mo Production Capacity in Uranyl Nitrate Aqueous
Total Page：16
File Type：pdf, Size：1020Kb
A. Isnaeni,Atom atIndonesia al., / Atom Vol. Indonesia 40 No. 1 Vol.(2014) 40 40No. - 143 (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: 99mTc is a very useful radioisotope in medical diagnostic procedure. 99mTc is Received 11 January 2014 produced from 99Mo decay. Currently, most of 99Mo is produced by irradiating 235U Received in revised form 18 February 2014 in the nuclear reactor. 99Mo mostly results from the fission reaction of 235U targets Accepted 28 February 2014 with a fission yield about 6.1%. A small additional amount is created from 98Mo 99 neutron activation. Actually Mo is also created in the reactor fuel, but usually we Keywords: 99 do not extract it. The fuel will become spent fuel which is a highly radioactive Mo waste. 99Mo production system in the aqueous homogeneous reactor offers a better Uranyl nitrate method, because all of the 99Mo can be extracted from the fuel solution. Fresh Homogeneous reactor reactor fuel solution consists of uranyl nitrate dissolved in water. There is no MCNP separation of target and fuel in an aqueous homogeneous reactor where target and ORIGEN 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 * 98 235 INTRODUCTION activation of Mo which is itself created from U 99 99m target fission reaction. Actually Mo is also created Tc is a very useful radioisotope in medical in the reactor fuel, but it is not usually extracted. diagnostic procedure; it is used in nearly 80% of all The fuel will instead become a highly radioactive nuclear medicine procedures [1]. 99mTc is produced 99 waste known as spent fuel. from Mo decay. Because of the short half-life of 99m uranium Tc (6.0058 hours) it is not transported to 235 99 solution U hospitals around the world, but Mo which has a Fission longer half life (65.94 hours) is delivered instead. Product 99 The total production and consumption of Mo in E- g the world is approximately 400 TBq/week [2]. 99Mo 99mTc 99Tc Global demand for 99mTc is expected to grow at an 65.94 h 6.0058 h average annual rate of 3±8% [3]. 98Mo The reduction of uranium enrichment from Activation ý90% to ý19.8% demands modifications on Fig. 1.99Mo production. process operation to compensate for the resulting loss in output [4]. The reduction of uranium 99Mo production system in the aqueous enrichment is due to nuclear non-proliferation. 99 homogeneous reactor (AHR) offers a better method, Currently most of Mo is being produced in because all of the 99Mo can be extracted from the research and test reactors by the irradiation of reactor solution. One such aqueous homogeneous targets containing enriched fissile material 235U [5]. 99 reactor was built in the past; it was operated almost Mo is extracted using an acidic process [6]. This daily as a neutron source from 1951 until its process generates waste. Figure 1 shows that 99Mo 235 deactivation in 1974, after 23 years of safe, reliable is mostly produced from the fission reaction of U operation [7]. Fresh reactor fuel solution consists of targets with a fission yield about 6.1%, while a uranyl nitrate dissolved in water. The production of small additional amount is created from the neutron 99Mo in the AHR follows the same nuclear reactions shown in Fig. 1. However, there is no separation of * Corresponding author. target and fuel in the AHR, as the target and the fuel E-mail address:[email protected] become one liquid solution. Therefore, there is no 40 A. Isnaeni, at al., / Atom Indonesia Vol. 40 No. 1 (2014) spent fuel generated from this reactor after 99Mo The inputs of ORIGEN 2.2 are power (200 kW), extraction from the reactor solution; the remains of time (one hour), and n % of fuel solution, where n the extraction results will be returned to the reactor % depends on the ratio of the fuel solution extracted core as the fuel solution for further operation. per hour to all of the fuel solution; the ratio depends The use of fuel solution reactors for the on the extraction flow rate and the volume of the production of medical isotopes is potentially fuel solution. advantageous because of their characteristics such The 99Mo extraction is capable of diverting as: low cost; small critical mass (low power); 0.1 to 1.0 mL/second flow of uranyl nitrate solution simple fuel handling, processing and purification [12]. The extraction flow rate that we used in this characteristics; and inherent passive safety [8]. simulation is 1.0 mL/second (Table 1). The void volume created by bubbles in the solution core will introduce a strong negative reactivity Table 1. Extraction flow rate. feedback [9]. Flow Flow rate(L/h) rate(mL/s) Extraction flow rate 1 3.6 EXPERIMENTAL METHODS We use MCNP5 for reactor criticality The reactor core parameters are shown in analysis and ORIGEN2.2 for nuclide inventory Table 2. The fuel solution consists of 20% analysis. MCNP is a general-purpose Monte Carlo enrichment uranyl nitrate dissolved in water. The N±Particle code which can be used for neutron, atom densities of the fuel solution are shown in photon, electron, or coupled neutron/photon/ Table 3. The reactor vessel atom densities are electron transport and possesses the capability to shown in Table 4. calculate eigenvalues for critical systems [10]. ORIGEN is a computer code system for calculating Table 2. Reactor core parameters. the buildup, decay, and processing of radioactive Parameter Value materials [11]. We couple both MCNP5 and ORIGEN2.2. We started by making MCNP input Reactor power (thermal) 200 kW which consists of fuel, reactor vessel and reflector, Fuel solution Uranyl nitrate then ran the input. The fuel solution height was Enrichment 20% gradually increased until the reactor became Inner core diameter (cm) 30 supercritical (MCNP output). After the reactor Reactor height (cm) 100 99 became supercritical we started to extract Mo in a Reactor vessel Stainless steel-304 close loop. We extracted the 99Mo from the fuel and Vessel thickness (cm) 0.5 then returned the remains of the extraction in the reactor core. The 99Mo extraction was simulated in Reflector (radial) Beryllium ORIGEN2.2. The simulation flow chart is shown Reflector thickness (cm) 30 in Fig. 2. Table 3. Atom densities in fresh fuel. Start Isotope atom/barn.cm 235 Non 99Mo U 1.26504531144E-04 radionuclide 238U 5.07525204789E-04 MCNP5 input 16O 3.34878465916E-02 material (uranium No (non 99Mo) solution), geometry 14N 1.26805947187E-03 1H 5.68312174084E-02 99Mo ORIGEN Table 4. Stainless steel-304 [13]. No, k > 1 output increase Nuclide atom/barn.cm volume Chromium 1.74 10-2 99 Yes ( Mo) Manganese 1.52 10-3 ORIGEN input n% -2 99 Iron 5.81 10 of fuel solution, Mo results power, time Nickel 8.51 10-3 Finish We used MCNP5 for criticality calculation. The geometrical model of the reactor on the Visual Fig. 2. Simulation flow chart. Editor of MCNP5 is shown in Fig. 3. 41 A. Isnaeni, at al., / Atom Indonesia Vol. 40 No. 1 (2014) 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. 120 100 80 Fig. 3. Geometrical model of the reactor, from the side of the reactor (left) and from the top of the reactor (right). The reactor 60 consists of uranyl nitrate solution (red), reactor vessel (blue) 40 and beryllium reflector (yellow). 99Mo product (Ci/h) (Ci/h) product 99Mo 20 0 0 20 40 60 80 100 RESULTS AND DISCUSSION Time (h) The first step we performed was to obtain the Fig. 4. 99Mo production capacities per hour. condition where keff> 1 was barely achieved, or, in other words, to determine the height of the uranyl After 43 hours of reactor operation the production nitrate solution because keff is a function of fuel rate of 99Mo is relatively constant at about 98.6 solution height. Using MCNP we made a variation Ci/h. The 235U in the fuel solution was consumed of uranyl nitrate solution height. The result of the during reactor operation; the decreasing atom simulation is shown in Table 5. 235 density of U in the fuel can be seen in Fig. 5. Table 5. Fuel solution height vs keff. 1,2652E-04 No Fuel solution height (cm) keff 1,2650E-04 1 10 0.58154 1,2648E-04 2 15 0.75903 1,2646E-04 3 20 0.86390 1,2644E-04 4 25 0.93485 U U (atom/barn.cm) 1,2642E-04 5 30 0.98914 235 6 35 1.02461 1,2640E-04 7 40 1.05394 1,2638E-04 8 45 1.06629 0 20 40 60 80 100 9 50 1.08568 Time (hours) 10 55 1.10217 Fig.