Neutronic and thermal-hydraulic analysis of fission molybdenum-99 production at Tehran Research Reactor using LEU plate targets

Applied Radiation and Isotopes 118 (2016) 160–166

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Neutronic and thermal-hydraulic analysis of fission molybdenum-99 production at Tehran Research Reactor using LEU plate targets

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Ebrahim Abedi , Marzieh Ebrahimkhani, Amin Davari, Seyed Mohammad Mirvakili, Mohsen Tabasi, Mohammad Ghannadi Maragheh Nuclear Science and Technology Research Institute, Tehran, Iran

A R T I C L E I N F O

A BS T RAC T

Keywords: Fission-moly Tehran research reactor LEU target Activity calculation Thermal-hydraulic CFD analysis

Efficient and safe production of molybdenum-99 (99Mo) radiopharmaceutical at Tehran Research Reactor (TRR) via fission of LEU targets is studied. Neutronic calculations are performed to evaluate produced 99Mo activity, core neutronic safety parameters and also the power deposition values in target plates during a 7 days irradiation interval. Thermal-hydraulic analysis has been also carried out to obtain thermal behavior of these plates. Using Thermal-hydraulic analysis, it can be concluded that the safety parameters are satisfied in the current study. Consequently, the present neutronic and thermal-hydraulic calculations show efficient 99Mo production is accessible at significant activity values in TRR current core configuration.

1. Introduction Molybdenum-99, with a half-life of 66 h, is a parent isotope of Technetium-99m with a relative short half-life of 6 h, which has been widely used in clinical nuclear imaging procedures (Cho and Kim, 1999). 99mTc is used in approximately 20–25 million procedures of medical diagnosis each year, that representing about 80% of all the nuclear medicine procedures in the world (Domingos et al., 2011). In Iran, more than 800,000 99mTc based procedures are applied annually. It is a unique isotope that can be incorporated into a number of radiopharmaceuticals to assist the diagnosis of problems in different parts of the human body including, brain, heart, liver, thyroid gland, lungs, kidneys and bone. The popularity of 99mTc is also due to its low cost, low radiation exposure to patients, high quality imaging and reliable availability in the form of 99Mo/99mTc generators (Mushtaq et al., 2008). 99Mo can be produced in research reactors via two different approaches. The first method is transmutation of 98Mo by absorption of a neutron (98Mo (n, γ) 99Mo), and the second method is production with fission of U-235 by the absorption of a neutron (Jo et al., 2014). Activation method is rather simple and inexpensive but gives low specific activity. On the other hand, fission method is complex and expensive but gives high specific activity (Mohammad et al., 2009). In Iran, during years 2007–2010, 99Mo radioisotope was produced using 98Mo transmutation method at Tehran Research Reactor (TRR). The whole activity of produced 99mTc was 100 Ci which was extracted by neutron transmutation of 200 g 98Mo oxide in TRR core for 12 days at 5 MW. Today the country 99Mo demands are estimated about

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100 Ci/3700 GBq (6 days Ci) entirely have supplied via imports by PARSISOTOPE Co. mostly from Russia. To decrease import dependency and also overcome to its difficulties due to transport and establishing local availability, a program has been launched to evaluation of 99Mo production feasibility and efficiency estimation via fissionMoly Method using LEU targets at TRR. In this method, in addition to 99 Mo, two other radioisotopes including Iodine-131, with a half-life of

Fig. 1. TRR Standard Fuel Element (SFE) cross section view and fuel plate dimensions.

Corresponding author. E-mail address: [email protected] (E. Abedi).

http://dx.doi.org/10.1016/j.apradiso.2016.09.011 Received 14 April 2016; Received in revised form 7 September 2016; Accepted 9 September 2016 Available online 10 September 2016 0969-8043/ © 2016 Elsevier Ltd. All rights reserved.

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9

IR-BOX

GRBOX

GRBOX

GRBOX

IR-BOX

GRBOX

8

SFE

CFE

SFE

SFE

SFE

SFE

7

SFE

SFE

SFE

SFE

CFE

SFE

6

SFE

CFE

SFE

IR-BOX

SFE

SFE

5

SFE

SFE

SFE

SFE

CFE

SFE

4

SFE

SFE

CFE

SFE

SFE

SFE

3

IR-BOX

SFE

SFE

SFE

SFE

IR-BOX

2

GRBOX

IR-BOX

IR-BOX

GRBOX

GRBOX

GRBOX

1

GRBOX

GRBOX

GRBOX

GRBOX

GRBOX

GRBOX

A

B

C

D

E

F

Table 1 Properties of designed mini-plates as the fission moly targets. Item No.

Feature

Quantity

Unit

1 2 3 4 5 6 7

U-Enrichment Chemical form Mini-plate Dimension Meat Dimension U-235 Mass Mass of plate Meat density

19.7 ± 0.2 U3O8 110×43×1.5 84.5×31.15×0.7 1.18 16.18 5.15

% – mm mm g g g/cm3

2. Tehran Research Reactor (TRR) TRR is a 5 MW thermal power pool type MTR research reactor operating by Nuclear Science and Technology Research Institute (NSTRI). The reactor first criticality took place at 1967 with HEU fuel elements. The core conversion to LEU fuel was performed by INVAP Co. in 1993 (Zaker, 1995). Since that time, TRR has been operated regularly for a vast range of services such as training, medicine and industrial radioisotopes production, BNCT and neutron radiography researches. One of the most important radioisotopes that have been produced in this period is 99Mo, but its production was abandoned due to low specific activity. Now, in the present work consideration is given to the neutronic and thermal-hydraulic investigations for a new program to produce significant country demand of 99Mo via fissionmoly approach.

Fig. 2. TRR core configuration that used for fission Moly targets irradiation.

8 days, and Xe-133, with a half-life of 5.25 days, are produced. So, the production rates and activity of these two radioisotopes are calculated in this work.

Fig. 3. (a)Target holder device (b) Target holder filled by mini-plates and their loading in IR-box isometric view (c) vertical view of two rows mini-plates and their numbering.

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Thermal Neutron Flux (n/cm2.s)

1.E+14

Termal Neutron Flux[n/cm2.s]

Fig. 4. (a)Target holder device vertical dimensions inside IR-box. (b) holder coolant channels dimensions.

D6 A9 C2 E9 F3

1.E+14 8.E+13 6.E+13 4.E+13 2.E+13 0.E+00

0

10

20

30

40

50

1.50E+14 1.30E+14 1.10E+14 9.00E+13

Water(row 1)

7.00E+13

Water(row 2) Meat(row 1)

5.00E+13

Meat(row 2)

3.00E+13 1.00E+13

0

2

4

6

8

10

12

14

Number of Target/plate

60

Distance from top of core(cm)

Fig. 6. Thermal neutron flux with and without 24 mini plates in 2 rows of twelve in D6 location.

Fig. 5. Axial thermal neutron flux distribution in irradiation boxes.

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Table 2 Neutronic safety parameters.

Mo-99 MCNPX I-131 MCNPX Xe-133 MCNPX Mo-99 ORIGEN2.1 I-131 ORIGEN2.1 Xe-133 ORIGEN2.1

Before loading value

After loading value

Allowed value

Effective Multiplication Factor (Keff) Core Excess Reactivity (pcm) Shutdown Margin (pcm) Safety Reactivity Factor (SRF) Peaking Factor (pmax/ pave)

1.032

1.048

1.062

4581

5460

–

10112 2.20

9084 1.66

> 3000[8] > 1.5[8]

1.54

1.58

–

ACTIVITY(CI)

2000

parameter

1500

1000

500

0 7

8

9

10

11

12

13

14

15

TIME(DAY)

Fig. 9. A comparison between activity values of three favorite radioisotopes calculated by MCNPX and ORIGIN codes.

2.50E+03

Xe-133

ACTIVITY(CI)

2.00E+03

I-131

1.50E+03

Mo-99

1.00E+03 5.00E+02 0.00E+00 0

2

4

6

8

10

12

14

16

TIME(DAY)

Fig. 7. Activity of three important radioisotopes versus time during 7 days irradiation, 2 days cooling and chemical process and 6 days utilizing. 3.50E+04

Actnides

ACTIVITY(CI)

3.00E+04

Fission Products

2.50E+04 2.00E+04 1.50E+04 1.00E+04 5.00E+03 0.00E+00

0

5

10

15

20

TIME(DAY) Fig. 8. Actinides and all fission products activity versus time for 7 days irradiation and 8 days decaying. Table 3 Core hot channel Safety parameters before and after mini-plates loading. Parameter

Before loading value

After loading value

Allowed value

Max coolant velocity(m/ s) Maximum heat Flux(W/ cm2) Max. coolant Temperature(℃) Max clad temperature (℃) Max. meat temperature (℃) ONB margin DNBR margin Flow redistribution margin

1.18

1.14

Vcr=15

33.76

33.76

–

59.2

59.9

< 120

86.1

87.4

< 650

105.5

106.8

< 650

1.93 5.75 3.31

1.87 5.6 3.2

> 1.3 >2 >2

2.1. Reactor core Fig. 10. Velocity counter inside holder devices and mini-plates cooling channels.

TRR core utilizes by plate type fuel assemblies. Fuel material is 20% enriched uranium oxide mixed by nuclear grade aluminum that covered by 0.4 mm thickness aluminum clad. The core consists of 163

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Fig. 11. Temperature contour in the mini-plates channels.

supplied by Iranian Nuclear Reactor Fuel Company. The meat material is a combination of U3O8 and an Aluminum alloy. The meat has been covered by a thin Aluminum clad with thickness of 0.4 mm, which is similar to the standard fuel plates of the reactor core. Each target plate contains 1.18 g U-235. The thickness of target plates is equal to the standard fuel plates of the core (1.5 mm). Table 1 shows the miniplates properties.

two type fuel assemblies, Standard Fuel Elements (SFE) and Control Fuel Elements (CFE). Fig. 1 depicts the SFE cross section view with fuel plate and cooling channel dimensions. SFE has 19 fuel plates while CFE has 14 fuel plates along with guide channels for fork type absorbers settlement (Atomic Energy Organization of Iran, 2009). 2.2. Core configuration Core fuel assemblies are fixed within 127 mm thick aluminum grid plate. The grid plate has 54 positions to mounting various types of core components such as fuel elements, empty boxes for samples irradiation purposes that called IR-boxes and graphite blocks as the core reflector. The core could be arranged by various compositions of mentioned components and total number of core fuel elements would be varied base on fuel management strategy. Fig. 2 shows one of the resent TRR core configurations which has been used in the present study. This core configuration consists of 28 SFEs and 5 CFEs. Also, as it can be seen in Fig. 2, several IR-boxes have been positioned almost in outer layers of the core and there is also one inside IR-box in D6 position.

3.1. Targets holder device To loading the mini plates inside the core, a target holder device has been designed. This holder is made by nuclear grade Aluminum alloy and it has a maximum capacity of 12 mini-plates (Fig. 3a). Considering national needs of 99Tc radioisotope and also relative low values of U235 in each mini-plate, it is concluded that more than 12 mini-plates are required to irradiation. Therefore, Designation of holder device is so that it could be loaded more than one holder in the one IR-box. Fig. 3b and c illustrate how these two holders are embedded at the bottom of IR-box so that one of them is set vertically above the other one. Fig. 4 depicts the vertical distances of the holders and their mini plates relative to core bottom level.

3. Mini-plate targets A specific type of domestic miniature target plates (mini-plates) has been designed for fission-Moly production purpose. Their U-235 enrichment is equal to 19.7% with a tolerance of ± 0.2% which is

4. Neutronic calculation At first step, a calculation is carried out to choose the best position 164

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of core for mini-plates irradiation. The most important parameter to do this is the thermal neutron flux which would be achievable at each IRbox. So, A Monte Carlo MCNPX code (Denise, 2008) model is used to calculate the axial thermal neutron flux in the various core IR-boxes. As seen in Fig. 5, D6 IR-box has the highest thermal neutron flux in compare with other core IR-boxes. The maximum neutron flux in this IR-box was calculated about 1.1E+14. Also, thermal neutron flux diminishing due to control rods insertion effect at the upper part of core can be seen in D6 flux curve. However, it could be concluded that the best position of TRR current core for Fission moly production will be D6 IR-box. In addition, the effect of mini-plates irradiation loading on D6 flux is studied. Thermal neutron flux with and without mini-plates inside D6 IR-box have been depicted in Fig. 6. The mini-plates are loaded in 2 rows with 12 plates at each row inside D6 IR-box with the heights of 170 mm and 20 mm from core bottom, respectively (Fig. 4a). As shown in Fig. 6 the average thermal neutron flux without mini-plates in row 1 (upper row) and row 2 (lower row) are 1.23E14 n/cm2 s and 1.13E14 n/cm2 s, whereas this flux is 0.558E14 and 0.492E14 n/ cm2 s within mini-plates, respectively. At the second step, the effects of mini-plates irradiation on the core neutronic safety parameters are studied. These parameters consist of reactivity (ρ), shutdown margin, safety reactivity factor and so on. The core configuration depicted in Fig. 2 is considered for this calculation. Also, it has been assumed that the reactor operates at 5 MW thermal power with 80% control rods off. Calculations are performed using Monte Carlo MCNPX code. Table 2 shows the core neutronic safety parameters before and after mini-plates loading. Results show that, Keff increases from 1.032 to 1.048, core excess reactivity increases from 4581 to 5460 pcm and the peaking factor value increases from 1.54 to 1.58 while safety reactivity factor (SRF) decreases from 2.20 to 1.58. Although, simulation results indicate a slight difference between neutronic safety parameters before and after mini-plates loading, however safety limits are fulfilled in both cases. In the third neutronic analysis step, the activity calculations are performed with both MCNPX and ORIGEN2.1 computer codes. Using the target power calculation by MCNPX, The depletion capability is limited to criticality problems in MCNPX2.6 code. The CINDER.dat file is required for depletion calculations. This library file contains decay, fission yield, and 63-group cross-section data (Denise, 2008). ORIGEN2.1 is a versatile point-depletion and decay computer code for use in simulation of nuclear fuel cycles and calculating the nuclide compositions. This code and its associated decay constants, cross sections, and photon libraries were developed in the late-1960s for use in generic fuel cycle studies (Croff, 1980). Fig. 7 has presented 99 Mo, 131I, 133Xe activities during 7 days irradiation, one day cooling, one day chemical processing and 6 days decay. As shown in Fig. 3 99Mo activity 6 d after processing is 290 Ci that supplies the country's demand. Fig. 8 shows actinides and all fission products activity during 7 days irradiation and 8 days decaying. These values are important for mini-plates’ transport and hot cell shielding considerations.

Fig. 12. Temperature contour on the mini-plates clad surfaces.

ANSYS Fluent code are used. MTR-PC is a special package for MTR type reactors that presented by INVAP Company. This package comprises various neutronic, thermo-hydraulic and shielding codes. Here, only thermal-hydraulic part, particularly CAUDVAP and TERMIC codes are used. CAUDVAP is a code to calculate the steady state velocity distribution of a system with different types of parallel channels with forced circulating flow of liquid water, single phase (Abbate and Mazufri, 1998). Gravity induced flow is distributed among each channel of TRR core according to the hydraulic resistance so that in the presence of fuel rod sample, coolant velocity through the plate type fuel assembly decreases. Using obtained velocities, thermal behavior of the fuel plates is analyzed using TERMIC code MODE 4.1 under the hot channel condition. TERMIC, as a part of MTR-PC package, is a one-dimensional code developed to perform core thermal hydraulic design using plate type fuel assemblies (Abbate, 2003). Table 3 illustrates various safety parameters of the core hot channel before and after mini-plates loading. The results show that even after mini-plates loading, all the safety parameters remain at conditions that satisfy the core safety margins. (Fig. 9) At the second part of thermal-hydraulic analysis, to analyze the thermal behavior of mini-plates, whole the mini-plates irradiation box is simulated by a 3D CFD model using ANSYS Fluent. ANSYS Fluent is a widely used CFD code, which solves Navier-Stockes equations with

5. Thermal-hydraulic analysis After insuring that neutronic parameters of core and target holder composition could satisfy enough 99Mo (and other favorite isotopes i.e. 133 Xe and 131I) production from neutronic calculations, here it has to be guaranteed an appropriate and trustable safety margin values would be supplied during fission moly targets irradiation period from thermal-hydraulic point of view. So, a complete core thermal-hydraulic analysis has been accomplished. This section's calculations contain two parts; first part includes investigation of whole core safety parameters except fission moly targets loaded IR-box. The second part is calculation of safety parameters for the target loaded Box specially. To fulfill these two purposes, MTR-PC package and a CFD model by 165

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possible to use larger plates for Mo-99 production at TRR, relatively small target plates have been employed in this analysis. Supplying the country demand of 99Mo radioisotope leads to load more than one mini-plate row inside the IR-box. Although, neutronic results show that the neutron flux of two mini-plates rows differ slightly, but this difference doesn’t have significant effects on final activity of targets. Thermal-hydraulic analysis of the core and mini-plates irradiation channel also showed that all safety parameters will be fulfilled during 7 days irradiation of 24 mini-plates loaded in D6 channel. Acknowledgments The present work was supported by Iranian Nuclear Science and Technology Research Institute (NSTRI). The authors also appreciate Tehran research reactor management and staffs for their technical contribution. References Fig. 13. Axial temperatures of mini-plate fuel and clad.

Atomic Energy Organization of Iran, 2009. Safety Analysis Report for Tehran Research Reactor (TRR). vol. 1, chapter 5. ANSYS, 2014. ANSYS-fluent User's Guide. Release 15.0. Domingos, D.B., Silva, A.T., Silva, J.E.R., João, T.G., 2011. Neutronic analysis for production of fission molybdenum-99 at IEA-R1 and RMB research reactors. INISBE–11N0031, Belgium. Abbate, P., Mazufri, C., 1998. CAUDVAP MODE 3.60: A computer program for flow distribution and pressure drop calculation in a MTR type core. INVAP Co. Abbate, P., 2003. TERMIC/MODE 401: A program for the thermal-hydraulic analysis of a MTR core in forced convection. INVAP Co.. Cho, D.K., Kim, M.H., 1999. Nuclear design methodology of fission moly target for research reactor. J. Korean Nucl. Soc. 31, 365–374. Croff, Allen G., 1980. A User's Manual for the ORIGEN2 Computer Code. ORNL/TM7175 (CCC-371). Oak Ridge National Laboratory. Denise, B.P., 2008. MCNPX User's Manual Version 2.6.0. LA-CP-07-1473. Los Alamos National Laboratory. Jo, D., Kim, H.C., Lee, K.H., Park, J., Chae, H., 2014. Neutronic and thermal-hydraulic analyses of irradiated fuel plates for molybdenum-99 production. Ann. Nucl. Energy 71, 467–474. Mohammad, A., Mahmood, T., Iqbal, M., 2009. Fission moly production at PARR-1 using LEU plate type target. Nucl. Eng. Des. Vol. 239 (3), 521–525. Mushtaq, A., Iqbal, M., Bokhari, I.H., Mahmood, T., Mahmood, T., Ahmad, Z., Zaman, Q., 2008. Neutronic and thermal-hydraulic analysis for production of fission molybdenum-99 at Pakistan Research Reactor-1. Ann. Nucl. Energy 35 (2), 345–352. Zaker, M., 1995. Conversion and start-up of Tehran Research Reactor with LEU fuel. 18th international meeting on reduced enrichment for research and test reactors. Iaea, Paris (France), 17–21.

finite volume method (ANSYS, 2014). For all the dependent variables, second-order upwind schemes are considered and the velocity and pressure fields are linked by the SIMPLE algorithm. Flow is assumed to be steady, incompressible and turbulent with constant fluid properties. Simulation results have been depicted in Figs. 10–13. Fig. 10 depicts the magnitude velocity contour inside the irradiation box from upper holder inlet to outlet nozzle of lower holder including mini plate cooling channels. Maximum velocity is equal to 3.1 m/s, which is more than core channels velocity (Table 3). Fig. 11 shows the temperature distribution at both upper and lower mini-plates rows in which maximum temperature occurs in the fuel and takes the value of 351.9 K(78.8 ℃). Furthermore, Fig. 12 shows the temperature distribution of mini-plates clads, which its maximum value is equal to 344.9 K (71.8 ℃). CFD results show that all safety margins are fulfilled in mini-plate irradiation channel appropriately. 6. Conclusions According to presented results, efficient production of 99Mo radiopharmaceutical isotope would be accessible in the current TRR core. Since, due to some fabrication and shielding limitations it is not

166