Proceedings of the 3rd Applied Science for Technology Application, ASTECHNOVA 2014
International Energy Conference
Yogyakarta, Indonesia, 13-14 August 2014
Development of New ORIGEN2 Library Sets for an Aqueous
Homogeneous Reactor (AHR) Dedicated for Mo-99/Tc-99m
Medical Radioisotope Production
Peng Hong Liem
Nippon Advanced Information Service (NAIS Co., Inc.)
416 Muramatsu, Tokaimura, Ibaraki, Japan
liemph@nais.ne.jp
Bakri Arbie
PT MOTAB Technology
Kedoya Elok Plaza Blok DA 12, Jl. Panjang, Kebun Jeruk, Jakarta Barat, Indonesia
dayatekn2@gmail.com
Tagor Malem Sembiring
Center for Reactor Technology and Nuclear Safety, National Nuclear Energy Agency (BATAN)
Kawasan Puspiptek Gedung No. 81, Serpong, Tangerang Selatan 15310, Indonesia
tagorms@batan.go.id
ABSTRACT
New ORIGEN2 library sets were developed based on JENDL-4.0 nuclear data for an aqueous
homogeneous reactor (AHR) dedicated for Mo-99/Tc-99m medical radioisotope production which uses
nitrate-based, low enriched uranium fuel solution.
KEYWORDS: ORIGEN2 library, JENDL4.0 nuclear data, aqueous homogeneous reactor, Mo-99/Tc99m, medical radioisotope
1
INTRODUCTION
ORIGEN2 (Croff, 1980, 1983) is a code developed at Oak Ridge National Laboratory for
calculating the buildup, decay and processing of radioactive materials in easily comprehensive form. The
materials commonly treated and characterized by the code cover in and out of core fuels (or spent fuels),
radioactive wastes, recovered elements such as uranium and plutonium, uranium and thorium ores, mill
tailings etc. The code provides a comprehensive result in terms of the radionuclide composition (by its
mass, fractional isotopic composition, radioactivity) and its characteristics such as toxicity (radioactive
and chemical ingestion and inhalation), neutronics (neutron absorption rate, fission rate, spontaneous
fission, (α,n) reaction rate) and photon emission (18 energy group of photon spectrum and total heat).
Another feature of the code worthily remarked is its fast computation time. In addition to simulating
nuclear fuel cycles, the code has been also used for much wider application areas such as fuel
reprocessing, spent fuel shipping, waste disposal, decommissioning, and some aspects of the analysis of
potential reactor accidents.
The coverage of reactor models in the ORIGEN2 data libraries (distributed with the code) evolves
from time to time (and released versions). Within the original ORIGEN2.1 (Croff, 1980) code package,
reactor models cover mainly typical reactors operated in the United States of America and Canada, i.e.
pressurized water reactors (PWR), boiling water reactors (BWR), CANDU reactors, liquid metal fast
breeder reactors (LMFBR) and the FFTF research reactor. Unfortunately, until recently there is no
publicly released ORIGEN2 data library set for aqueous homogeneous reactors (AHR) which mainly use
low enriched uranium (LEU, < 20 w/o U-235 enrichment) solution fuels and consequently the versatility
of ORIGEN2 code may not be fully benefited.
The AHR is now being considered as a potential candidate for the Mo-99/Tc-99m medical
radioisotope production. The detail of the conceptual reactor design and its optimization were discussed
thoroughly elsewhere (Liem et al., 2013a, 2013b). The ORIGEN-2 code with the proper library set
dedicated for the AHR is expected to be one of the important analytical tools for the AHR design. Here
we report the development of new ORIGEN-2 library sets dedicated for the AHR based on the latest
Japanese Nuclear Data JENDL-4.0 (Shibata et al., 2011).
2
DESCRIPTION OF THE TARGETED REACTOR MODEL (AHR)
The AHR considered here consist of a small modular core (thermal power < 300 kW/module)
containing nitrate-base uranium fuel solution (LEU, enrichment of less than 20 w/o, chemical form of
UO2(NO3)2). The main characteristics of the proposed AHR are listed in Table 1. When operated the
reactor module(s) will be submerged into a large water pool so that the pool water also functions as
reflector (cf. Figure 1). The core fuel solution is cooled by cooling coils. Other components include the
control and safety rods penetrating the core which may be made of B4C or cadmium. Our preliminary
optimization work indicated that the optimal uranium concentration is around 200 – 300 g U/L (Liem et
al., 2013b).
Figure 1: Configuration of the targeted aqueous homogeneous reactor (AHR)
The inherent safety features of the reactor are attributed to the low power, low power density,
atmospheric operational pressure, small excess reactivity and large negative reactivity coefficients. As
shown in the figure, since the reactor is submerged in a large water pool, the pool water functions both as
passive residual heat removing agent and biological shielding. The production capacity of the reactor can
be adjusted by varying the reactor power of a single reactor and/or by varying the number of reactor
(modularity feature).
Table 1. Main characteristics of the proposed AHR.
Design Parameter
Value
Constraint
Thermal power (kW) per module
Variable
< 300
Power density (kW/L)
1–2
< 2.5
Fuel solution
UO2(NO3)2
U-235 enrichment (w/o)
19.75
< 20.0
Uranium concentration (gU/L)
200 – 300
< 500
Concentration of HNO3 (mol/L)
0.1 – 0.5
Initial U-235 (kg)
approx. 3
Average fuel temperature (C)
approx. 80
< 100
Operation pressure (MPa)
atmospheric
-
3
METHODOLOGY
Based on these design parameters, we should develop new ORIGEN2 data library sets, each
dedicated to (1) fuel solution, (2) cooling coils, control rod guide tubes and other structural materials
inside the core, (3) core tank and (4) water reflector which may contain structural components
surrounding the vessel. The first library set is for core depletion (as well as for evaluating the Mo-99
production) and, therefore, the effective one-group cross sections are commonly generated with the
assumption that the neutron spectrum within the core of the targeted reactor is the most dominant and are
a function of burnup level. The latter 3 library sets are for non-fuel (i.e. non-depleted) regions which are
used for irradiation/activation analysis of the corresponding structural materials. The neutron spectra for
these regions may deviate from the core (fuel solution) and accurate neutron spectra must be prepared
separately for obtaining the effective one-group cross sections.
The methodology for developing the new library sets is described in detail (Liem and Sembiring,
2013c) and its validation results can be found in Liem et al. (2013d), therefore, we only review briefly
here. Figure 2 outlines the calculation procedure for creating the new sets of ORIGEN2 data library. In
Step 1 of the calculation procedure, a multigroup master library containing all nuclides of JENDL-4.0
nuclear data was generated using the NJOY99 code system (MacFarlane and Muir, 1994). The master
library was characterized by the fuel solution temperature (in this case 353.15 K) and the infinite dilute
background cross section. In the master library, multigroup (n,2n), (n,3n), (n, gamma), (n, fission), (n,
alpha) and (n, proton) cross sections were stored. Also in this step, the burnup dependent effective onegroup cross sections for actinides (13 nuclides: uranium-234, -235, -236, -238, neptunium-237,
plutonium-238, -239, -240, -241, -242, americium-241, -243, curium-242, -244) were generated for each
fuel type and uranium loading using SRAC2006 code system (Okumura, 2007). In Step 2 of the
calculation procedure, the obtained burnup dependent actinide one-group cross sections were edited and
hard-coded into the ORIGEN2 source code (and after that compiled to get the executables).
Also in this step, using the multigroup neutron spectra and the multigroup master library obtained
in Step 1, the effective one-group neutron cross sections were generated and over-written into the
PWR47J40 library for non-actinide nuclides. During this process, the nuclides (actinides and important
fission products) treated by SRAC2006 have already effective one group cross sections calculated,
therefore, these were used for over-writing the PWR47J40 library. In addition, it should be noted that
except the nuclides included in JENDL-4.0 nuclear data library, the cross sections for other nuclides were
remained unchanged.
JENDL-4.0
SRAC Library
Burnup
Dependent
Effective
One-Group
Actinide XS
Actinide XS
Editing Tools
XSEC Subroutine
for ORIGEN2
Code
SRAC2006
(Cell Burnup)
NJOY99
(Nuclear Data
Processing)
Multigroup
Neutron
Spectra
Multigroup
Master
Library
STEP 1
ORIGEN2
Library
Editing Tools
PWR47J40
Data Library
New Set of
ORIGEN2
Data Library
JENDL-4.0
Nuclear Data
STEP 2
Figure 2: Calculation procedure of ORIGEN2 data library generation for fuel region
The calculation procedure for ORIGEN2 data library generation for non-fuel regions is shown in
Figure 3. In Step 1 of the calculation procedure, a multigroup master library containing all nuclides of
JENDL-4.0 nuclear data was generated using the NJOY99 code system, and multigroup macroscopic
cross sections for fuel and non-fuel regions were prepared using the SRAC2006 code system. The
temperatures adopted in the calculations were changed to match with the temperatures of the non-fuel
regions. In Step 2 of the calculation procedure, the multigroup neutron spectra (in 107 groups) of non-fuel
regions were obtained from a full reactor 3-D multigroup neutron diffusion calculation using the
CITATION module where the macroscopic cross sections were already prepared in Step 1. Then, using
these multigroup neutron spectra and the master library obtained in Step 1, the effective one-group
neutron cross sections were generated and over-written into the PWR47J40 library set.
As an example, the calculated multigroup neutron spectra used in this work is shown in Figure 4.
The neutron spectra correspond to the optimal range of uranium concentration of 250 g U/L. It can be
observed that the neutron spectra are similar for fuel solution and vessel regions but significantly different
in the water reflector region. Therefore, the library set used for neutron activation analysis of structural
materials surrounding the reactor vessel should be generated using the neutron spectrum of the water
reflector in order to obtain accurate calculation results.
JENDL-4.0
SRAC Library
CITATION
(3-D Diffusion
Calculation)
SRAC2006
(Cell Burnup)
NJOY99
(Nuclear Data
Processing)
Multigroup
Macro XS
Multigroup
Master
Library
STEP 1
Multigroup
Neutron
Spectra
ORIGEN2
Library
Editing Tools
PWR47J40
Data Library
New Set of
ORIGEN2
Data Library
JENDL-4.0
Nuclear Data
STEP 2
Figure 3: Calculation procedure of ORIGEN2 data library generation for non-fuel region.
4
APPLICATION EXAMPLE
Here we give an application example of the new sets of ORIGEN2 library for the proposed AHR
depletion calculation. The calculation was conducted using the latest version of ORIGEN2 i.e. the
ORIGEN2.2UPJ code. The AHR with fresh fuel solution is assumed to be operated continuously for 5
days with a constant power density of 1 kW/L. After 5 days operation, the reactor is shutdown. The
composition of the fresh fuel solution is already given in Table 1.
The calculated time-dependent Mo-99 concentration in the fuel solution is shown in Figure 5. The
Mo-99 concentration reaches approximately 4.6×10-10 /cm3 barn after 5 days irradiation and then decays
to approximately 1.0×10-10 /cm3 barn at 6 days after the reactor is shutdown. During 5 days reactor
operation approximately 0.013 % of the initial U-235 is consumed. The information on the initial and
final compositions of both fuel and fission products were used for criticality calculations. The criticality
calculation result shows that the burnup reactivity loss for 5 days reactor operation (start from fresh fuel
solution) is approximately 85 pcm.
0.07
Fuel solution
Spectrum (1/ lethargy)
0.06
Vessel
Water reflector
0.05
0.04
0.03
0.02
0.01
0
1.0E-04
1.0E-02
1.0E+00 1.0E+02 1.0E+04 1.0E+06 1.0E+08
Energy (eV)
Figure 4: Region-wise neutron spectra used for the new ORIGEN2 library set generation
(uranium concentration 250 g U/L)
5
CONCLUDING REMARK
New sets of ORIGEN2 data library, as an extension of ORLIBJ40 sets were successfully developed
in order to meet the needs of such data library sets for design and analysis of aqueous homogeneous
reactors. Since the data library sets was generated using carefully prepared space-dependent neutron
spectra, the ORIGEN2 code will produce accurate depletion calculation results in fast and comprehensive
manner.
Nuclide Concentration (1/barn.cm)
6.0E-10
1.2
Mo-99 Concentration
Power Density (W/cc)
5.0E-10
1.0
4.0E-10
0.8
3.0E-10
0.6
2.0E-10
0.4
1.0E-10
0.2
-4.1E-25
0.0
0
1
2
3
4
5
6
7
8
9
10 11 12 13 14
Operation Time (days)
Figure5: ORIGEN2.2UPJ calculation result using the newly developed AHR dedicated library sets.
REFERENCES
Croff, A.G., 1980. A user’s manual for the ORIGEN2 computer code. ORNL/TM-7175 (CCC-371).
Croff, A.G., 1983. ORIGEN2: A versatile computer code for calculating the nuclide compositions and
characteristics of nuclear materials. Nuclear Technology 62, 335-352.
Liem, P.H. et al., 2013a. Conceptual Design of a New Homogeneous Reactor for Medical Radioisotope
Mo-99/Tc-99m Production. ICANSE-2013, Bandung Institute of Technology, Bandung, Indonesia.
Liem, P.H. et al., 2013b. Design Optimization of a New Homogeneous Reactor for Medical Radioisotope
Mo-99/Tc-99m Production. INES-4, Tokyo Institute of Technology, Tokyo, Japan.
Liem, P.H. and Sembiring T.M., 2013c. Development of New ORIGEN2 Data Library Sets for Research
Reactors with Light Water Cooled Oxide and Silicide LEU (20 w/o) Fuels Based on JENDL-3.3
Nuclear Data. Nuclear Engineering and Design, 262, 52-62.
Liem, P.H. et al., 2013d. Nondestructive burnup verification by gamma-ray spectroscopy of LEU silicide
fuel plates irradiated in the RSG GAS Multipurpose Reactor. Annals of Nuclear Energy 56, 57-65.
MacFarlane, R.E. and Muir, D.W., 1994. The NJOY nuclear data processing system, version 91. LA12740-M, Los Alamos National Laboratory.
Okumura, K. et al., 2007. SRAC2006: A comprehensive neutronics calculation code system. JAEAData/Code 2007-004.
Shibata, K. et al., 2011. JENDL-4.0: A new library for nuclear science and engineering. J. Nucl. Sci.
Technol. 48, 1-30.