Analysis of
99Mo Production Capacity in Uranyl Nitrate Aqueous Homogeneous Reactor using ORIGEN and MCNP
Arif Isnaeni*, Mohammad Subian Aljohani, T. G. Aboalfaraj, 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 AIJ use only:
Received date Revised date Accepted date Keywords:
99Mo uranyl nitrate homogeneous reactor MCNP
ORIGEN
99mTc is 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 is mostly results of 235U targets fission reaction with a fission yield about 6.1%. Small amount of it created from 98Mo neutron activation. Actually
99Mo also created in the reactor fuel, but usually we do not take it. The fuel will be spent fuel as 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 fuel reactor solution consists of uranyl nitrate dissolved in water. There is no separation of target and fuel in aqueous homogeneous reactor, target and fuel become one liquid solution, there is no spent fuel generated from this reactor. Extraction simulation is performed while reactor in operation (without reactor shutdown), extraction flow rate 3.6 liter/hour, after 43 hours reactor operation the production of 99Mo is relatively constant about 98.6 curie/hour.
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INTRODUCTION
99mTc is very useful radioisotope in medical diagnostic procedure, it used in nearly 80% of all nuclear medicine procedures [1]. 99mTc is produced from 99Mo decay, because of short half live of 99mTc (6.0058 hours) we do not transport 99mTc to hospitals around the world, but we deliver 99Mo that has longer half life (65.94 hours). The total production and consumption of 99Mo in the world is approximately 400 TBq/week [2]. Global demand for 99mTc will grow at an average annual rate of 3 – 8 % [3].
The uranium enrichment dropping from
∼90% to ∼19.8% demands modifications on process operation to compensate for the resulting loss in output [4].
Currently most of 99Mo is being produced in research and test reactors by the irradiation of targets containing enriched fissile material 235U [5].
99Mo was extracted using an acidic process [6], this process generate waste. 99Mo is mostly results of
Corresponding author.
E-mail address: [email protected]
235U targets fission reaction with a fission yield about 6.1 %. Small amount of it created from 98Mo neutron activation, 98Mo created from 235U target fission reaction. Actually 99Mo also created in the reactor fuel, but usually we do not take it. The fuel will be spent fuel as highly radioactive waste.
Figure 1. 99Mo production
99Mo production system in the aqueous homogeneous reactor offers a better method, because all of the 99Mo can be extracted from the reactor solution. One of aqueous homogeneous reactor was built in the past, it was operated almost
235U Fission Product
98Mo Activation
99Mo 99mTc 99Tc
6.0058 hours 65.94 hours
daily as a neutron source from 1951 until its deactivation in 1974, 23 years of safe, reliable operation [7]. Fresh fuel reactor solution is consist uranyl nitrate dissolved in water. 99Mo mostly results of 235U targets fission reaction with a fission yield about 6.1 %. Small amount of it created from
98Mo neutron activation, 98Mo created from 235U fission reaction. There is no separation of target and fuel in aqueous homogeneous reactor, target and fuel become one liquid solution, so there is no spent fuel generated from this reactor, after 99Mo extraction from the reactor solution, the rest of the extraction will be returned to the reactor core as the fuel solution.
The use of fuel solution reactors for the production of medical isotopes is potentially advantageous because of their 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 METHOD
We use MCNP5[10] to analyze the reactor criticality and ORIGEN2.2 [11] for nuclide inventory analysis. We couple both MCNP5 and ORIGEN2.2.
Figure 2. Simulation flow chart.
ORIGEN 2.2 input is power (200 kW), time (one hour), and n % of fuel solution, the n % is depend on the ratio of the fuel solution extracted per hour to all of the fuel solution, the ratio is depend on the
extraction flow rate and the volume of the fuel solution.
The 99Mo extraction capable for 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. Extraction flow rate
Flow rate (ml/s) Flow rate (liter/hour)
Extraction flow rate 1 3.6
Table 2. Reactor core parameter
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. Stainless steel-304 [13]
Nuclide atom/barn.cm
Chromium 1.74 x 10-2
Manganese 1.52 x 10-3
Iron 5.81 x 10-2
Nickel 8.51 x 10-3
Table 4. Atom densities in fresh fuel.
Istope atom/barn.cm
235U 1.26504531144E-04
238U 5.07525204789E-04
16O 3.34878465916E-02
14N 1.26805947187E-03
1H 5.68312174084E-02
We are using MCNP5 for criticality calculation.
The geometrical model of the reactor on the Visual Editor of MCNP5 is shown in the Figure 3 below:
Figure 3. Geometrical model of the reactor, from the side of the reactor (left) and from the top of the reactor (right). Consist of uranyl nitrate solution (red), reactor vessel (blue) and beryllium reflector (yellow).
MCNP5 input material (uranium solution), geometry Start
k >
1
ORIGEN input n% of fuel solution, power, time
99Mo ORIGE N output Non 99Mo radionuclid
e
99Mo results
Finish No,
increase volume uranium solution
No (non 99Mo)
Yes ( 99Mo)
RESULTS AND DISCUSSION
The first step we do is to get the k-eff > 1, in other word to determine the height of the uranyl nitrate solution, because k-eff is a function of fuel solution height. Using MCNP we made a variation of uranyl nitrate solution height.
Table 5. Fuel solution height vs k-eff.
No Fuel solution height (cm) k-eff
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 k-eff = 1.02461 and height of uranyl nitrate solution in the core is 35 cm, the inner diameter of the core is 30 cm, then we get the volume of the fuel solution is 24.7275 liter. The extraction flow rate is 3.6 liter/hour, it means in one hour we can extract 3.6 liter fuel solution, the ratio of the fuel solution that is extracted per hour to all fuel solution in the core:
Ratio ¿ 3.6
24.7275 ¿ 0.1458868972 There are 14.58868972 % of all fuel solution is extracted per hour. ORIGEN2.2 simulate the fuel extraction, there are 14.58868972% of all 99Mo in the fuel as the 99Mo result per hour, the other nuclide and 85.41131028 % of 99Mo are remain in the core as the fuel solution, in this simulation it becomes ORIGEN2.2 and MCNP5 input for the next hours. The 99Mo result has to be decayed for one hour because this extraction process takes time one hour. The 99Mo result after decayed can be seen in the Figure 4 below:
0 10 20 30 40 50 60 70 80 90 100 0
20 40 60 80 100 120
Time (hours)
99Mo product (curie/hour)
Figure 4. 99Mo production capacities per hour.
After 43 hours reactor operation the 99Mo is relatively constant about 98.6 curie/hour. The atom density of 235U can be seen in the Figure 5 below:
0 10 20 30 40 50 60 70 80 90 100 1.26E-4
1.26E-4 1.26E-4 1.26E-4 1.26E-4 1.26E-4 1.26E-4 1.26E-4 1.27E-4 1.27E-4
Time (hours)
235U (atom/barn.cm)
Figure 5. 235U atom density (atom/barn.cm) Criticality analysis in MCNP5 shows that the reactor is supercritical for 96 hours reactor operation.
0 10 20 30 40 50 60 70 80 90 100 0.90
0.92 0.94 0.96 0.98 1.00 1.02 1.04
Time (hours)
k-eff
Figure 6. k-eff during reactor operation.
The Figure 7. shows the accumulation or total of 99Mo product for 720 hours reactor operation.
0 100 200 300 400 500 600 700 0
1000 2000 3000 4000 5000 6000 7000 8000 9000 10000
Time (hours)
Total 99Mo product (curie)
Figure 7. Total 99Mo Product (Curie)
The accumulation product is increasing by reactor operating time, but after reach about 400 hours it relatively constant. Because the accumulation result of 99Mo will be decreased by decay process, it is better to take the result as often as possible.
CONCLUSION
Reactor simulation on MCNP5 shows that the reactor is supercritical during reactor operation.
99Mo extraction is performed during reactor operation with flow rate 3.6 liter/hours. After 43 hours reactor operation the 99Mo production is relatively constant about 98.6 curie/hour. The accumulation result of 99Mo will be decreased by decay process, it is better to take the result as often as possible.
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