- Email: [email protected]
Gamma irradiator dose mapping simulation using the MCNP code and benchmarking with dosimetry
Applied Radiation and Isotopes 57 (2002) 537–542
Gamma irradiator dose mapping simulation using the MCNP code and benchmarking with dosimetry M. Sohrabpoura,*, M. Hassanzadehb, M. Shahriarib, M. Sharifzadeha a
Gamma Irradiation Center Atomic Energy Organization of Iran, POB 11365-8486, Tehran, IR Iran b Physics Department, Amir Kabir University of Technology, Tehran, IR Iran Received 18 April 2001; received in revised form 11 April 2002; accepted 29 April 2002
Abstract The Monte Carlo transport code, MCNP, has been applied in simulating dose rate distribution in the IR-136 gamma irradiator system. Isodose curves, cumulative dose values, and system design data such as throughputs, over-doseratios, and efficiencies have been simulated as functions of product density. Simulated isodose curves, and cumulative dose values were compared with dosimetry values obtained using polymethyle-methacrylate, Fricke, ethanol-chlorobenzene, and potassium dichromate dosimeters. The produced system design data were also found to agree quite favorably with those of the system manufacturer’s data. MCNP has thus been found to be an effective transport code for handling of various dose mapping excercises for gamma irradiators. r 2002 Elsevier Science Ltd. All rights reserved. Keywords: IR-136 gamma irradiator; Monte Carlo; Dose mapping; MCNP code; Dosimetry; Benchmarking
1. Introduction Various radiation transport codes have been used routinely for the analysis of nuclear and or radiation producing systems. Among the different analytical techniques the so-called Monte Carlo method has performed best. Using this method it is possible to calculate the various parameters of flux, fluence, energy spectra, and absorbed dose with good accuracy, and precision in a given volume of matter when it is in interaction with a given source of ionizing radiation. In the field of radiation processing, system design engineers have typically made use of analytical techniques, while facility operators have perhaps relied more on routine dosimetry for process or quality control purposes. *Corresponding author. Tel.: +98-21-800-4065; fax: +9821-800-9054. E-mail address: [email protected] (M. Sohrabpour).
Availability of PC’s and mathematical codes (in the latter case being either home made or those issued by various institutes) have also allowed physicists or dosimetrists to extend routine dosimetry to also include dose mapping, process validation, and process optimization, etc. A number of benchmark studies using point kernel codes and comparison with dosimetry for Co-60 irradiators is found in Raisali et al. (1990), PinaVillalpando and Sloan (1995, 1998), Saylor and Jordan (2000). Monte Carlo application for process control and dose rate determinations in a gamma irradiation facility has been described by Oliveira et al. (2000a, b). Application of Monte Carlo code EGS4 for determination of gamma ray spectrum and dose rate distribution in Gamma cell 220 is also given by Raisali and Sohrabpour (1993). In this work application of the MCNP code for doserate mapping of a gamma irradiation system is reported, benchmarked with dosimetry data and also with the manufacturer’s design information.
0969-8043/02/$ - see front matter r 2002 Elsevier Science Ltd. All rights reserved. PII: S 0 9 6 9 - 8 0 4 3 ( 0 2 ) 0 0 1 3 0 - 6
538
M. Sohrabpour et al. / Applied Radiation and Isotopes 57 (2002) 537–542
Table 1 IR-136 gamma irradiator system specification Irradiator type
Carrier, Wet storage
No. of carriers Carrier material and effective thickness No. of rows
69 Al, 3.65 mm
No. of shelves in each carrier Product carton dimensions
4 44 44 44 cm3
Air gap around each carton
4 cm
6
2. IR-136 gamma irradiator system specification The IR-136 irradiator is a rectangular carrier type system having 69 aluminum carriers each holding four product cartons. Carrier movements take place in railed tracks under the action of pneumatic pistons. Each entering carton makes six passes at each shelf level for a total of 24 passes around the source rack. The general layout of the irradiator system has been depicted in Fig. 4. The system comprising six rows and four shelves. The source rack, also shown in this figure, consists of 10 trays arranged in two sets of five trays each. The two sets are superimposed in a vertical arrangement. Each source tray comprises of 42 receiving positions which are filled by either the source pencils or the dummies. The source rack is also symmetrically positioned inside the irradiator between the rows 3 and 4. In the vertical direction the two sets of trays are placed at the levels of the 2nd and the 3rd shelves, respectively. Other pertinent system specifications are given in Table 1.
3. System simulation in the MCNP medium MCNP is a general purpose Monte Carlo code used for calculating the time-dependent continuous energy transport of neutrons, photons, and electrons in three-dimensional geometries (Briesmeister, 1993). In definition of our problem each product carton was divided into cubical cells of 4 cm mesh size. Flux tallies F4 or also the track length estimates of the flux cell have been combined with the heating response function HðEÞ to give the dose tallies F6 as follows: Z Z Z dV ¼ Wvt=V ¼ WT1=V Fðr%; E; tÞ dE dt F4 ¼ V V t E
Minimum source to carrier distance Source type Source dimensions
10 cm Cylindrical rod 45 1.11 cm2
Clad material, thickness, and density Inter-row distance Product material and density range
Stainless Steel, 0.145 mm, 7.87 g/cm3 7 cm CH2 0.05–0.3 g/cm3
Other material densities
Co 8.9 g/cm3 Al 2.7 g/cm3
since ds ¼ v dt Z Z Z dV Fðr%; E; tÞ dt dE V V E t Z Z Z dV Nðr%; E; tÞ ds dE ; ¼ V V E S Z Z Z dV F6 ¼ ra =rg HðEÞFðr%; E; tÞ dE dt V V t E and heating response HðEÞ for photons is given by HðEÞ ¼ sT ðEÞHavg ðEÞ; Havg ðEÞ ¼
3 X
pi ðEÞ ðE E% out Þ;
i¼1
where Fðr%; E; tÞ is the energy-dependent flux, Nðr%; E; tÞ is the track length density, W is the weight, Tl is the track length estimate, V is the cell volume, ra is the atom density (atom/barn-cm), rg is the mass density (g/cm3), and HðEÞ is the energy or heating response. sT ðEÞ is the total energy dependent photon cross section and pi ðEÞ is the energy-dependent discrete probability density function which is related to energy depositions from pair production, Compton scattering, and the photoelectric effect, respectively. Simulation of the cobalt sources with exact activities, dimension, and coordinate positions was based on an analog volumetric source pencil model which produced isotropically scattered photons. Sampling of the source volume was done on a frequency which was related to the actual activity of each source rod. Each source rod length was assumed to represent a point source for distances exceeding three times the rod length. The active part of the source pencils (40 cm) in certain cases were also divided into four sections of 10 cm each. These 10 cm sections were used as point sources for the nearby target cells. One or two million photon histories were followed for each carton volume for measurement of the dose distribution. Relative
M. Sohrabpour et al. / Applied Radiation and Isotopes 57 (2002) 537–542
errors of the accumulated scores were o0.1% at the 1s confidence interval.
539
25
4. Simulated results and comparison with dosimetry
25.5 26 27 27.7 28.4 28.9 29.3
Fig. 2. Calculated isodose curves by the MCNP code in a product box of density r ¼ 0:13 g/cm3 in the IR-136 gamma irradiator system, in a plane perpendicular to the source frame.
400 2 Relative dose rate
Fig. 1 presents the isodose curves on a central plane in a product carton placed in a vertical position relative to the source rack. The doses were measured with clear PMMA dosimeters (Kazemi and Sohrabpour, 1992) placed in a matrix arrangement of 4 cm spacing. Conversely, Fig. 2 shows isodose distribution curves calculated by the MCNP code, demonstrating close agreement of these with the results depicted in Fig. 1. Fig. 3 shows the relative dose-rate distribution at a product density r ¼ 0:15 g/cm3 for a carrier traveling along the third row track and at all four shelf levels shown in Fig. 4. The indicated shelf nos. 2, 3, 1 and 4 correspond to the decreasing order of dose-rates at dosimetry point of the traveling cartons. Fig. 4 is a schematic diagram of the IR-136 irradiator system showing all of its 6 rows and 4 shelves. Each pair of row and shelf coordinates (a total of 24 pairs) were selected and the total cumulative dose values at the position of minimum dose (i.e. center of the top flap cover of each carton) was simulated through use of the MCNP code. Measurements were also made with dosimeters along the same travel path. The top rows of numbers in Fig. 4 are the sum of the calculated dose values obtained through use of the MCNP code while the bottom rows of numbers are the total dose values measured by a given dosimeter type traveling along each
300
3 1
200
4
100
S 0 25
25.4 25.8 26.6 27.2
0
100 200 300 Distance along the row (cm)
400
Fig. 3. Relative dose-rate distribution for a product density r ¼ 0:15 g/cm3 along the third row at various shelf positions. ‘‘S’’ represents the source position. Curves nos. 2, 3, 1 and 4 correspond to the decreasing order of dose-rates encountered by the product cartons at the source rack.
27.8 27.5 28.4 28.9
Fig. 1. Measured isodose curves in a product box of density r ¼ 0:13 g/cm3 in the IR-136 gamma irradiator system, in a plane perpendicular to the source frame.
of the 24 carrier travel paths. The product material used for this benchmark study consisted of polyethylene vials of 10 ml volume with a bulk density of 0.13 g/cm3, filling the volume of each carton, as well as, the volume of the entire irradiator. The simulated dose represents the average value over a cell volume of 4 4 4 cm3, containing within this volume several 10 ml polyethylene vials. No attempt was made to measure dose distribution inside a given bottle, it being apparent that inside and outside dose values for the small plastic vials should be very nearly the same due to the small thickness, low
M. Sohrabpour et al. / Applied Radiation and Isotopes 57 (2002) 537–542
540
Fig. 4. Comparison of the MCNP code simulation and measured dosimetry results for the IR-136 system for a product density r ¼ 0:13 g/cm3. The types of dosimetry system are indicated for every row, and the figures shown are the cumulative dose values along the indicated row and shelf numbers.
3500
3
2500 Dose (Gy)
MCNP Measurement
Shelf No. 2
3000
2000 1
1500 1000
4
500 0 R=1
R=2
R=3
R=4
R=5
R=6
Row No.
Fig. 5. Graphical comparison of cumulative doses along the six carrier rows of the IR-136 system, comparison being made between MCNP and dosimetry results.
density, and the low-atomic number values of the vial material. Due to the variable dose rates encountered at different distances from the source rack various dosimeter types with suitable dose response ranges were selected and used. These were Fricke dosimeters at rows l and 6 and at all four shelf positions, ethanol-chlorobenzene dosimeters, ECB (Kovacs et al., 1985), at rows 2 and 5 and also at all four shelf positions and finally, potassium
dichromate dosimeters (Sharpe et al., 1985), utilized at rows 3 and 4 nearest to the source rack and also at all four shelf levels. Fig. 5 details the simulated and measured dose values of Fig. 4 in a single graphical representation. Figs. 6 and 7 show the rates of throughput (m3/100 kCi/h), and efficiencies (%) as functions of product densities, calculated by MCNP and compared with the manufacturer’s data, MDS Nordion (1983). The method of
M. Sohrabpour et al. / Applied Radiation and Isotopes 57 (2002) 537–542
541
0.5 Throughput (m3/100 KCi/hr)
0.45 0.4 0.35 0.3 0.25
MCNP NORDION
0.2 0.15 0.1 0.05 0 0.05
0
0.1
0.15
0.2
0.25
0.3
0.35
Density (g/cm3)
Fig. 6. The calculated throughput rate (m3/100 kCi/h) as a function of product density comparison being made between MCNP code results and the system manufacturer’s data.
45 40 35 Efficiency
30 25 20
MCNP
15
NORDION
10 5 0 0
0.05
0.1
0.15
0.2
0.25
0.3
0.35
3
Density (g/cm )
Fig. 7. The calculated percent efficiencies as a function of product density, comparison being made between MCNP code values and the system manufacturer’s data.
1.35 MCNP NORDION Measured
1.30 Over-dose-ratio
calculation of the latter is believed to be point kernel technique. Fig. 8 also compares MCNP simulated and the manufacturer’s over-dose-ratios as functions of product density. The over-dose-ratio, measured experimentally at a single density of 0.13 g/cm3, is also plotted in this figure. Of note is that this point correlates more closely with the MCNP simulation than with the Nordion data.
1.25 1.20 1.15 1.10
5. Benchmark data and discussion 1.05
The qualitative agreement between the isodose curves obtained through use of MCNP and dosimetry, as illustrated in Figs. 1 and 2, are supported by over-doseratios for these two curves are approximately 1% only. Significant number of dosimetries was conducted in order to obtain the broad range of comparisons between
0
0.05
0.1
0.15
0.2
0.25
0.3
0.35
Density (g/cm3)
Fig. 8. Over-dose-ratios as a function of product density calculated by the MCNP code and compared with the system manufacturer’s data and also with a single measured value at a density of r ¼ 0:13 g/cm3.
542
M. Sohrabpour et al. / Applied Radiation and Isotopes 57 (2002) 537–542
calculation and measured values shown in Fig. 4. The observed variations between the simulated and measured data were: MCNP-Fricke (3%); MCNP-ECB (6%); and MCNP-dichromate (2%). The largest percent variation, between the MCNP and ECB data set was largely due to the estimation of the lower scale reading on the oscillotitrator reader for the ECB dosimeters. In general, only reference standards, or transfer standard dosimeters were used in this benchmark exercise in order to reduce the measurement uncertainties. The mean difference of 3% between measured and simulated total dose values, however, may also be partly attributed to the inhomogeneous packing of the product cartons, small random displacement of the carriers and also the cartons about their supposed resting positions, etc.
6. Conclusions The MCNP code system has been used for simulation of transport of gamma rays within the IR-136 system. It has been used to simulate dose rate variation, to generate isodose curves and cumulative system dose values, and in examining the design parameters of the throughput rates, the over-dose-ratios and also the system efficiencies as functions of the product densities. The MCNP generated system parameters were compared with the manufacturer’s data, producing excellent agreement of o1% variation between the two sets of data values. The differences between the calculated and the system cumulative dosimetry values obtained with the reference standard dosimeters were in the range of 2–6%. In the case of simulated vs. experimental isodose determinations the over-dose-ratios agreed to within 1%, the actual isodose curves also showing qualitative agreement with each other. The observed differences are thought to be associated in part with the displacement of the carrier positions vs. the simulated positions, nonhomogeneous product loading within the cartons as opposed to the uniform loading assumed in the simulation, etc. Based on the benchmarked data for the IR-136 it can be ascertained that the Monte Carlo code, MCNP, can
be effectively utilized for dose rate mapping of a gamma irradiator system.
Acknowledgements This work is based on the results of the M.Sc. thesis of one of the authors (MH).
References Briesmeister, J.F., 1993. MCNP: a general Monte Carlo Nparticle transport code, Version 4A. Los Alamos National Laboratory Report LA-12625. Kazemi, A.A., Sohrabpour, M., 1993. Dosimetric properties of commercial grade plexiglass as a possible routine dosimeter for radiation processing. Radiat. Phys. Chem. 42 (4–6), 793–797. Kovacs, A., Stenger, V., Foldiak, G., Legeza, I., 1985. Evaluation of irradiated ethanol-mono-chlorobenzene by conductivity method. Proceedings of the Symposium High Dose Dosimetry, IAEA, Vienna, STI/PVB/671, pp. 135–142. MDS Nordion, 1983. IR-136 project proposal, private communication. Oliveira, C., Salgado, J., Botelho, M.L., Ferreira, L.M., 2000a. Dose determination by Monte Carlo, a useful tool in gamma radiation process. Phys. Chem. 57, 667–670. Oliveira, C., Salgado, J., Carvalho, A.F., 2000b. Dose rate determinations in the Portuguese gamma irradiation facility: Monte Carlo simulation and measurements. Radiat. Phys. Chem. 58, 279–285. Pina-Villalpando, G., Sloan, D.P., 1995. Use of computer code for dose distribution studies in a 60Co industrial irradiator. Radiat. Phys. Chem. 46, 1385–1390. Pina-Villalpando, G., Sloan, D.P., 1998. Dose distribution studies of a gamma industrial irradiator using a PC code. Radiat. Phys. Chem. 52 (1–6), 563–567. Raisali, G.R., Sohrabpour, M., 1993. Application of EGS4 computer code for determination of gamma ray spectrum and dose rate distribution in Gamma cell 220. Radiat. Phys. Chem. 42 (4–6), 799–805. Raisali, G.R., Sohrabpour, M., Hadjinia, A., 1990. A computer code for dose rate mapping of gamma irradiators. Radiat. Phys. Chem. 35 (4–6), 831–835. Sharpe, P.H.G., Barrett, J.H., Berkely, A.M., 1985. Acidic aqueous dichromate solutions as reference dosimeters in the 10–40 kGy range. Appl. Radiat. Isot. 36, 647–652.
Gamma irradiator dose mapping simulation using the MCNP code and benchmarking with dosimetry M. Sohrabpoura,*, M. Hassanzadehb, M. Shahriarib, M. Sharifzadeha a
Gamma Irradiation Center Atomic Energy Organization of Iran, POB 11365-8486, Tehran, IR Iran b Physics Department, Amir Kabir University of Technology, Tehran, IR Iran Received 18 April 2001; received in revised form 11 April 2002; accepted 29 April 2002
Abstract The Monte Carlo transport code, MCNP, has been applied in simulating dose rate distribution in the IR-136 gamma irradiator system. Isodose curves, cumulative dose values, and system design data such as throughputs, over-doseratios, and efficiencies have been simulated as functions of product density. Simulated isodose curves, and cumulative dose values were compared with dosimetry values obtained using polymethyle-methacrylate, Fricke, ethanol-chlorobenzene, and potassium dichromate dosimeters. The produced system design data were also found to agree quite favorably with those of the system manufacturer’s data. MCNP has thus been found to be an effective transport code for handling of various dose mapping excercises for gamma irradiators. r 2002 Elsevier Science Ltd. All rights reserved. Keywords: IR-136 gamma irradiator; Monte Carlo; Dose mapping; MCNP code; Dosimetry; Benchmarking
1. Introduction Various radiation transport codes have been used routinely for the analysis of nuclear and or radiation producing systems. Among the different analytical techniques the so-called Monte Carlo method has performed best. Using this method it is possible to calculate the various parameters of flux, fluence, energy spectra, and absorbed dose with good accuracy, and precision in a given volume of matter when it is in interaction with a given source of ionizing radiation. In the field of radiation processing, system design engineers have typically made use of analytical techniques, while facility operators have perhaps relied more on routine dosimetry for process or quality control purposes. *Corresponding author. Tel.: +98-21-800-4065; fax: +9821-800-9054. E-mail address: [email protected] (M. Sohrabpour).
Availability of PC’s and mathematical codes (in the latter case being either home made or those issued by various institutes) have also allowed physicists or dosimetrists to extend routine dosimetry to also include dose mapping, process validation, and process optimization, etc. A number of benchmark studies using point kernel codes and comparison with dosimetry for Co-60 irradiators is found in Raisali et al. (1990), PinaVillalpando and Sloan (1995, 1998), Saylor and Jordan (2000). Monte Carlo application for process control and dose rate determinations in a gamma irradiation facility has been described by Oliveira et al. (2000a, b). Application of Monte Carlo code EGS4 for determination of gamma ray spectrum and dose rate distribution in Gamma cell 220 is also given by Raisali and Sohrabpour (1993). In this work application of the MCNP code for doserate mapping of a gamma irradiation system is reported, benchmarked with dosimetry data and also with the manufacturer’s design information.
0969-8043/02/$ - see front matter r 2002 Elsevier Science Ltd. All rights reserved. PII: S 0 9 6 9 - 8 0 4 3 ( 0 2 ) 0 0 1 3 0 - 6
538
M. Sohrabpour et al. / Applied Radiation and Isotopes 57 (2002) 537–542
Table 1 IR-136 gamma irradiator system specification Irradiator type
Carrier, Wet storage
No. of carriers Carrier material and effective thickness No. of rows
69 Al, 3.65 mm
No. of shelves in each carrier Product carton dimensions
4 44 44 44 cm3
Air gap around each carton
4 cm
6
2. IR-136 gamma irradiator system specification The IR-136 irradiator is a rectangular carrier type system having 69 aluminum carriers each holding four product cartons. Carrier movements take place in railed tracks under the action of pneumatic pistons. Each entering carton makes six passes at each shelf level for a total of 24 passes around the source rack. The general layout of the irradiator system has been depicted in Fig. 4. The system comprising six rows and four shelves. The source rack, also shown in this figure, consists of 10 trays arranged in two sets of five trays each. The two sets are superimposed in a vertical arrangement. Each source tray comprises of 42 receiving positions which are filled by either the source pencils or the dummies. The source rack is also symmetrically positioned inside the irradiator between the rows 3 and 4. In the vertical direction the two sets of trays are placed at the levels of the 2nd and the 3rd shelves, respectively. Other pertinent system specifications are given in Table 1.
3. System simulation in the MCNP medium MCNP is a general purpose Monte Carlo code used for calculating the time-dependent continuous energy transport of neutrons, photons, and electrons in three-dimensional geometries (Briesmeister, 1993). In definition of our problem each product carton was divided into cubical cells of 4 cm mesh size. Flux tallies F4 or also the track length estimates of the flux cell have been combined with the heating response function HðEÞ to give the dose tallies F6 as follows: Z Z Z dV ¼ Wvt=V ¼ WT1=V Fðr%; E; tÞ dE dt F4 ¼ V V t E
Minimum source to carrier distance Source type Source dimensions
10 cm Cylindrical rod 45 1.11 cm2
Clad material, thickness, and density Inter-row distance Product material and density range
Stainless Steel, 0.145 mm, 7.87 g/cm3 7 cm CH2 0.05–0.3 g/cm3
Other material densities
Co 8.9 g/cm3 Al 2.7 g/cm3
since ds ¼ v dt Z Z Z dV Fðr%; E; tÞ dt dE V V E t Z Z Z dV Nðr%; E; tÞ ds dE ; ¼ V V E S Z Z Z dV F6 ¼ ra =rg HðEÞFðr%; E; tÞ dE dt V V t E and heating response HðEÞ for photons is given by HðEÞ ¼ sT ðEÞHavg ðEÞ; Havg ðEÞ ¼
3 X
pi ðEÞ ðE E% out Þ;
i¼1
where Fðr%; E; tÞ is the energy-dependent flux, Nðr%; E; tÞ is the track length density, W is the weight, Tl is the track length estimate, V is the cell volume, ra is the atom density (atom/barn-cm), rg is the mass density (g/cm3), and HðEÞ is the energy or heating response. sT ðEÞ is the total energy dependent photon cross section and pi ðEÞ is the energy-dependent discrete probability density function which is related to energy depositions from pair production, Compton scattering, and the photoelectric effect, respectively. Simulation of the cobalt sources with exact activities, dimension, and coordinate positions was based on an analog volumetric source pencil model which produced isotropically scattered photons. Sampling of the source volume was done on a frequency which was related to the actual activity of each source rod. Each source rod length was assumed to represent a point source for distances exceeding three times the rod length. The active part of the source pencils (40 cm) in certain cases were also divided into four sections of 10 cm each. These 10 cm sections were used as point sources for the nearby target cells. One or two million photon histories were followed for each carton volume for measurement of the dose distribution. Relative
M. Sohrabpour et al. / Applied Radiation and Isotopes 57 (2002) 537–542
errors of the accumulated scores were o0.1% at the 1s confidence interval.
539
25
4. Simulated results and comparison with dosimetry
25.5 26 27 27.7 28.4 28.9 29.3
Fig. 2. Calculated isodose curves by the MCNP code in a product box of density r ¼ 0:13 g/cm3 in the IR-136 gamma irradiator system, in a plane perpendicular to the source frame.
400 2 Relative dose rate
Fig. 1 presents the isodose curves on a central plane in a product carton placed in a vertical position relative to the source rack. The doses were measured with clear PMMA dosimeters (Kazemi and Sohrabpour, 1992) placed in a matrix arrangement of 4 cm spacing. Conversely, Fig. 2 shows isodose distribution curves calculated by the MCNP code, demonstrating close agreement of these with the results depicted in Fig. 1. Fig. 3 shows the relative dose-rate distribution at a product density r ¼ 0:15 g/cm3 for a carrier traveling along the third row track and at all four shelf levels shown in Fig. 4. The indicated shelf nos. 2, 3, 1 and 4 correspond to the decreasing order of dose-rates at dosimetry point of the traveling cartons. Fig. 4 is a schematic diagram of the IR-136 irradiator system showing all of its 6 rows and 4 shelves. Each pair of row and shelf coordinates (a total of 24 pairs) were selected and the total cumulative dose values at the position of minimum dose (i.e. center of the top flap cover of each carton) was simulated through use of the MCNP code. Measurements were also made with dosimeters along the same travel path. The top rows of numbers in Fig. 4 are the sum of the calculated dose values obtained through use of the MCNP code while the bottom rows of numbers are the total dose values measured by a given dosimeter type traveling along each
300
3 1
200
4
100
S 0 25
25.4 25.8 26.6 27.2
0
100 200 300 Distance along the row (cm)
400
Fig. 3. Relative dose-rate distribution for a product density r ¼ 0:15 g/cm3 along the third row at various shelf positions. ‘‘S’’ represents the source position. Curves nos. 2, 3, 1 and 4 correspond to the decreasing order of dose-rates encountered by the product cartons at the source rack.
27.8 27.5 28.4 28.9
Fig. 1. Measured isodose curves in a product box of density r ¼ 0:13 g/cm3 in the IR-136 gamma irradiator system, in a plane perpendicular to the source frame.
of the 24 carrier travel paths. The product material used for this benchmark study consisted of polyethylene vials of 10 ml volume with a bulk density of 0.13 g/cm3, filling the volume of each carton, as well as, the volume of the entire irradiator. The simulated dose represents the average value over a cell volume of 4 4 4 cm3, containing within this volume several 10 ml polyethylene vials. No attempt was made to measure dose distribution inside a given bottle, it being apparent that inside and outside dose values for the small plastic vials should be very nearly the same due to the small thickness, low
M. Sohrabpour et al. / Applied Radiation and Isotopes 57 (2002) 537–542
540
Fig. 4. Comparison of the MCNP code simulation and measured dosimetry results for the IR-136 system for a product density r ¼ 0:13 g/cm3. The types of dosimetry system are indicated for every row, and the figures shown are the cumulative dose values along the indicated row and shelf numbers.
3500
3
2500 Dose (Gy)
MCNP Measurement
Shelf No. 2
3000
2000 1
1500 1000
4
500 0 R=1
R=2
R=3
R=4
R=5
R=6
Row No.
Fig. 5. Graphical comparison of cumulative doses along the six carrier rows of the IR-136 system, comparison being made between MCNP and dosimetry results.
density, and the low-atomic number values of the vial material. Due to the variable dose rates encountered at different distances from the source rack various dosimeter types with suitable dose response ranges were selected and used. These were Fricke dosimeters at rows l and 6 and at all four shelf positions, ethanol-chlorobenzene dosimeters, ECB (Kovacs et al., 1985), at rows 2 and 5 and also at all four shelf positions and finally, potassium
dichromate dosimeters (Sharpe et al., 1985), utilized at rows 3 and 4 nearest to the source rack and also at all four shelf levels. Fig. 5 details the simulated and measured dose values of Fig. 4 in a single graphical representation. Figs. 6 and 7 show the rates of throughput (m3/100 kCi/h), and efficiencies (%) as functions of product densities, calculated by MCNP and compared with the manufacturer’s data, MDS Nordion (1983). The method of
M. Sohrabpour et al. / Applied Radiation and Isotopes 57 (2002) 537–542
541
0.5 Throughput (m3/100 KCi/hr)
0.45 0.4 0.35 0.3 0.25
MCNP NORDION
0.2 0.15 0.1 0.05 0 0.05
0
0.1
0.15
0.2
0.25
0.3
0.35
Density (g/cm3)
Fig. 6. The calculated throughput rate (m3/100 kCi/h) as a function of product density comparison being made between MCNP code results and the system manufacturer’s data.
45 40 35 Efficiency
30 25 20
MCNP
15
NORDION
10 5 0 0
0.05
0.1
0.15
0.2
0.25
0.3
0.35
3
Density (g/cm )
Fig. 7. The calculated percent efficiencies as a function of product density, comparison being made between MCNP code values and the system manufacturer’s data.
1.35 MCNP NORDION Measured
1.30 Over-dose-ratio
calculation of the latter is believed to be point kernel technique. Fig. 8 also compares MCNP simulated and the manufacturer’s over-dose-ratios as functions of product density. The over-dose-ratio, measured experimentally at a single density of 0.13 g/cm3, is also plotted in this figure. Of note is that this point correlates more closely with the MCNP simulation than with the Nordion data.
1.25 1.20 1.15 1.10
5. Benchmark data and discussion 1.05
The qualitative agreement between the isodose curves obtained through use of MCNP and dosimetry, as illustrated in Figs. 1 and 2, are supported by over-doseratios for these two curves are approximately 1% only. Significant number of dosimetries was conducted in order to obtain the broad range of comparisons between
0
0.05
0.1
0.15
0.2
0.25
0.3
0.35
Density (g/cm3)
Fig. 8. Over-dose-ratios as a function of product density calculated by the MCNP code and compared with the system manufacturer’s data and also with a single measured value at a density of r ¼ 0:13 g/cm3.
542
M. Sohrabpour et al. / Applied Radiation and Isotopes 57 (2002) 537–542
calculation and measured values shown in Fig. 4. The observed variations between the simulated and measured data were: MCNP-Fricke (3%); MCNP-ECB (6%); and MCNP-dichromate (2%). The largest percent variation, between the MCNP and ECB data set was largely due to the estimation of the lower scale reading on the oscillotitrator reader for the ECB dosimeters. In general, only reference standards, or transfer standard dosimeters were used in this benchmark exercise in order to reduce the measurement uncertainties. The mean difference of 3% between measured and simulated total dose values, however, may also be partly attributed to the inhomogeneous packing of the product cartons, small random displacement of the carriers and also the cartons about their supposed resting positions, etc.
6. Conclusions The MCNP code system has been used for simulation of transport of gamma rays within the IR-136 system. It has been used to simulate dose rate variation, to generate isodose curves and cumulative system dose values, and in examining the design parameters of the throughput rates, the over-dose-ratios and also the system efficiencies as functions of the product densities. The MCNP generated system parameters were compared with the manufacturer’s data, producing excellent agreement of o1% variation between the two sets of data values. The differences between the calculated and the system cumulative dosimetry values obtained with the reference standard dosimeters were in the range of 2–6%. In the case of simulated vs. experimental isodose determinations the over-dose-ratios agreed to within 1%, the actual isodose curves also showing qualitative agreement with each other. The observed differences are thought to be associated in part with the displacement of the carrier positions vs. the simulated positions, nonhomogeneous product loading within the cartons as opposed to the uniform loading assumed in the simulation, etc. Based on the benchmarked data for the IR-136 it can be ascertained that the Monte Carlo code, MCNP, can
be effectively utilized for dose rate mapping of a gamma irradiator system.
Acknowledgements This work is based on the results of the M.Sc. thesis of one of the authors (MH).
References Briesmeister, J.F., 1993. MCNP: a general Monte Carlo Nparticle transport code, Version 4A. Los Alamos National Laboratory Report LA-12625. Kazemi, A.A., Sohrabpour, M., 1993. Dosimetric properties of commercial grade plexiglass as a possible routine dosimeter for radiation processing. Radiat. Phys. Chem. 42 (4–6), 793–797. Kovacs, A., Stenger, V., Foldiak, G., Legeza, I., 1985. Evaluation of irradiated ethanol-mono-chlorobenzene by conductivity method. Proceedings of the Symposium High Dose Dosimetry, IAEA, Vienna, STI/PVB/671, pp. 135–142. MDS Nordion, 1983. IR-136 project proposal, private communication. Oliveira, C., Salgado, J., Botelho, M.L., Ferreira, L.M., 2000a. Dose determination by Monte Carlo, a useful tool in gamma radiation process. Phys. Chem. 57, 667–670. Oliveira, C., Salgado, J., Carvalho, A.F., 2000b. Dose rate determinations in the Portuguese gamma irradiation facility: Monte Carlo simulation and measurements. Radiat. Phys. Chem. 58, 279–285. Pina-Villalpando, G., Sloan, D.P., 1995. Use of computer code for dose distribution studies in a 60Co industrial irradiator. Radiat. Phys. Chem. 46, 1385–1390. Pina-Villalpando, G., Sloan, D.P., 1998. Dose distribution studies of a gamma industrial irradiator using a PC code. Radiat. Phys. Chem. 52 (1–6), 563–567. Raisali, G.R., Sohrabpour, M., 1993. Application of EGS4 computer code for determination of gamma ray spectrum and dose rate distribution in Gamma cell 220. Radiat. Phys. Chem. 42 (4–6), 799–805. Raisali, G.R., Sohrabpour, M., Hadjinia, A., 1990. A computer code for dose rate mapping of gamma irradiators. Radiat. Phys. Chem. 35 (4–6), 831–835. Sharpe, P.H.G., Barrett, J.H., Berkely, A.M., 1985. Acidic aqueous dichromate solutions as reference dosimeters in the 10–40 kGy range. Appl. Radiat. Isot. 36, 647–652.