Neutronic Aspect of SAMOP Reactor

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International Conference on Advances in Nuclear Science and Engineering in Conjunction with LKSTN 2007 (23-26)

Neutronic aspect of Subcritical Assemby for Mo-99

Production (SAMOP) Reactor

Topan Setiadipura* , Elfrida Saragi

Computational Division PPIN-BATAN, Serpong, Indonesia

2

Affiliation, City, Country*E-mail:[email protected]

Abstract

NEUTRONIC ASPECT OF SUBCRITICAL ASSEMBLY FOR Mo-99 PRODUCTION (SAMOP)

REACTOR. Design of a subcritical assembly for Mo-99 production (SAMOP) is in progress at National Nuclear

Energy Agency. The main purpose of the project is to b able to produce Mo-99, which is a parent of Tc-99m an

important nuclide for nuclear medicine application. The major source of Mo-99 is from fission of U-235. The

conventional technique is by forming the uranium into targets and irradiated by neutrons from research or test

reactors, then the irradiated targets are dissolved and the Mo-99 is extracted from the solution. Another

technique to produce Mo-99 from U-235 is by using homogeneous reactor fueled with uranyl nitrate. This

method introduced by Ball in 1992 and some advantages compared with the conventional method. On the

SAMOP design, the low enriched uranyl nitrate is placed in the stainless steel container and irradiated by

neutron generator. In this paper, the neutronic aspect of the design will be reported. Including the criticality

analysis to secure the subcriticality of the design and the neutron flux distribution analysis, also the effect of the

graphite reflector. The neutronic analysis was using the general monte carlo code, MCNP.

Keywords: Mo-99, uranyl nitrate, subcritical, neutronic, monte carlo

1. Introduction

99Mo is used as a parent isotope of thewidely used medical radiotracer99mTc. It is estimated

that 99mTc is used in over 85%of all nuclear medicine

clinical studies in the world. The strong demand for99mTc has stimulated a search for reliable supplies

of 99Mo [1].

Basically, there are two process or reaction

to produce 99mTc, fission from U-235 or capturereaction of 98Mo as shown in the picture 1. The

fission yield of 99Mo is about 6.1%. The fission-

produced 99Mo has a high specific activity (~104 Ci99Mo /g Mo) which makes it the most important

source of 99Mo in the world. The fission technique,however, requires considerable capital investment

and produces large quantities of radioactive waste.The cross section of the 98Mo (n,) 99Mo reaction is

small (th ~ 0.14 barns) and only a small portion of

the 98Mo is converted to 99Mo. The resulting specificactivity (~Ci 99Mo /g Mo) is much lower than that of

the fission-produced99

Mo. Advanced generatortechnologies are required to produce high quality

99mTc generators from the capture-produced 99Mo.

For the fissioning process, the conventional

way is to form the U-235 into a target which thenirradiated by neutrons from research or test reactor.

These irradiated targets are dissolved and the fission

product, 99Mo, is extracted from the solutions. In

1992, Ball [2] introduced a method of "targetless"

production of fission product Mo-99 using anaqueous homogeneous reactor fueled with uranyl

nitrate. The design anticipated that the uranium salt

could be made with low enriched uraniurn [3,4].

Figure 1. Two different reaction to produce99Tc from 99Mo.

Subcritical Assembly for Mo-99 Production

(SAMOP) is designed at the PTAPB-BATAN basedon the Balls method. The core of the SAMOP is the

Uranyl Nitrate solution in the SS-304 tank which

irradiated by the neutrons from a D-T neutron

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Table of Contents
http://table_of_contents.pdf/


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generator. Geometrical data of the SAMOP design is

given in the table below

Table 1. Samop Geometrical Data

Parameter Value

Core tank (Inner tank)

Diameter 15.35cm

Height 35cmUranyl Nitrate Solution Height 30.7cm

SS-304 thick 0.3cm

Coolant Tank (Outer tank)

Diameter 80cm

Height 190cm

Distance Inner Tank Reflector 1cm

Distance Inner Tank Outer tankbased

40cm

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2. MethodologyThe neutronic calculation is done using

MCNP [6], a general purpose monte carlo code. To

do the calculation on the MCNP we modeled the

neutron source term from the neutron generator, thegeometry and material of the SAMOP reactor. The

geometrical model of the SAMOP reactor on the

Visual Editor of MCNP is shown in the picturebelow:

Figure 2. Geometrial model of the SAMOPreactor, from the side of the reactor (left) and

from the top of the reactor. (right).

Energy and intensity of the source neutron

as a function of the angle is modeled using several

probability distribution. The intensity distribution is

modeled using Source Information (SI) card for the

cosines of the angle where the intensity is known and

the Source Probability (SP) card for the relatedintensity. Dependent Source (DS) cards is used for

the energy distribution because the energy is depends

on the direction distribution. The intensity(normalized for the value of angle 90o degree) and

energy of the neutron source is given in the table

below [5]

Table 2. Neutron source characteristic

, degree IntensityEnergy

(MeV)

0 1,059 14.962

10 1,048 14.948

20 1,046 14.906

30 1,043 14.839

40 1,038 14.74850 1,031 14.637

60 1,024 14.509

70 1,017 14.368

80 1,008 14.220

90 1,000 14.069

100 0,992 13.919

110 0,984 13.776

120 0,976 13.642

130 0,969 13.523

140 0,964 13.421

150 0,959 13.339

160 0,956 13.278

170 0,954 13.241180 0,953 13.229

Uranyl Nitrate used in this calculation is enriched to

20%, and the Uranium concentration is 300 g/L. The

material composition of Uranyl Nitrate is given in thetable below

Table 3. Material composition of the UranylNitrate

NuclideAtomic Density

(atom/barn cm)

U-234 1.2977E-06U-235 1.5374E-04

U-238 6.0591E-04

N 1.5219E-03

H 5.5457E-02

O 3.3816E-02

KCODE card is used to calculate the effective

multiplication factor, k-eff, of the SAMOP core. And

the F4 tally is used to calculate the flux distributionof the SAMOP core. Using the F4 tally, the average

neutron flux is estimated by summing the neutrons

track length in the cells. The track length estimator is

generally quite reliable to because there arefrequently many tracks in the cell (compared to thenumber of the collisions), leading to many

contribution to this tally[6]. To calculate the flux of

different area of the core, F4 tally is applied to asmall spherical cells with different position that

represented the area of the core. To simplified the

tallying, the height of the cells position is divided

into three level, and in each level there are nine cells


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represented the center and the pheripery of the core at

that level. The detail cells configuration and

numbering is shown in the figure below,

Figure 3. Cell configuration for detail flux

calculation.

The above figure picturing the cells of theupper level, with the center of the cell height is

27.55cm (from the base of the core tank), cells 17-25

for the middle level, and cells 26-34 for the lower

level with the height of the cells center is 15.35 and

3.15 respectively.

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3. Results and DiscussionThe k-eff that represented the criticality

condition of the SAMOP reactor is calculated for

different width of the reflector. In this calculation wecalculate for two reflector material berrylium andgraphit, each with different width starting from zero

to 40cm.

Effect of diferent Be Reflector Width

0.86

0.88

0.9

0.92

0.94

0.96

0.98

1

1.02

1.04

0 5 1 0 1 5 20 2 5 2 6 2 7 2 8 29 3 0 3 1 3 2 3 3 34 3 5 3 6 3 7 3 8 39 4 0

tebal reflektor

k-eff

Figure 4. Effect of Diferent Be Reflector Thicknes

Effect of diferent Graphit Reflector Width

0.89

0.9

0.91

0.92

0.93

0.94

0.95

0.96

0.97

0.98

0.99

0 5 1 0 1 5 20 2 5 2 6 2 7 2 8 29 3 0 3 1 3 2 3 3 34 3 5 3 6 3 7 3 8 3 9 4 0

tebal reflektor (cm)

k-e

ff

Figure 5. Effect of Diferent Graphite Reflector

Thicknes

From the calculation, the k-eff of the core without the

reflector is 0.92319 which is to low for the

application. This k-eff can be increased by additional

(radial) reflector. The calculation shows that theincrease of the k-eff is more significant by using

berrylium reflector than using graphite reflector. The

results also shows that using any reflector material,there is a limitation of the k-eff, it can be higher even

the reflector thickness is increase.

The flux distribution inside the SAMOP

core for different reflector width is given in the figure

below

Flux Distribution

0

0.005

0.01

0.015

0.02

0.025

0.03

0.035

0.04

0.045

0.05

8

1

0

1

2

1

4

1

6

1

8

2

0

2

2

2

4

2

6

2

8

3

0

3

2

3

4

Cell Number

Norm

alized

Flux

sam_f3

sam_f4

sam_g6

Figure 6. Flux distribution of the SAMOP core

for different reflector width.

The above figure picturing the distribution for

reflector thickness 10cm, 5 cm and without reflector.

As the reflector getting thicker the flux is gettinghigher as expected. The results shows that the higher

flux is in the center of the core. The average flux in

the core for different reflector thickness are given in

the table below


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Table 4. Flux average of the SAMOP core

for different reflector thickness

Reflector thickness Average flux (n/cm2-s)

0 (non reflector) 4.21e+9

5cm 6.85e+9

10cm 9.27e+916cm 1.16e+10

4. ConclusionThe neutronic aspect of the SAMOP

reactor design is already done including the

criticality of the core and the neutron fluxdistribution on the core. Regarding the criticality, it

is confirmed that the core is subcritic and the level

of the subcriticality, the k-eff, can be increase close

to critical at least by using thicker reflector. And itis investigated that the Be reflector is much more

effective to increase the k-eff of the core. Theaverage flux in the core is about 1.16E+10 for the

graphite reflector thickness 16cm where the k-effvalue is between 0.97 0.98.

AcknowledgmentThe Authors would like to aknowledge

Prof. Sarip and the SAMOP Team at the PTAPB-

BATAN Yogyakarta for the opportunity to involvein the project.

References

1. S.C.Mo (1993), Production of 99Mo UsingLEU and Molybdenum Targets, RERTR

Meeting 1993.

2. Ball,R.M.(1992), Testimony Before theCongressional Committee on U.S. Resources

on the Production of Mo-99 with Aqueous

Homogeneous Reactors, Mike Synar,Chairman.

3. Ball, R.M. (1994), Use of LEU in the AqueousHomogeneuos Medical Isotope Production

Reacto, RERTR Meeting, Williamsburg,

Virginia 1994.4. Ball,R.M.(1995), The Mo-99 Solution, Nuclear

Engineering International.5. Slamet Santoso (2007), Neutron Yield and

Energy Calculation of thr Neutron Generator

for SAMOP, not published.

6. X-5 Monte Carlo Team (2003), MCNP-AGeneral Monte Carlo N-Particle Transport

Code, Version 5 Vol.1, LA-UR-03-1987

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Neutronic Aspect of SAMOP Reactor

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