A. Background

Using the Monte Carlo neutron transport code (MCNP) is widely recognized as more precise and mature than any other neutron transport program code in reactor design and analysis. The point-wise continuous-energy cross section data in ACE format is suitable for use by the MCNP family of Monte-Carlo codes and other codes that can read the same format library.

When using MCNP to calculate temperature-related parameters in the design of a rector core, nuclear data from corresponding temperatures needed to be produced to ensure its accuracy. That was quite a waste of time. A nuclear data processing package SIGACE [1] was released by IAEA. It could broaden the ACE format data at merely 300 K to higher temperatures based on FENDL2.1 nuclear data. But it would produce errors when dealing with some heavy nuclei, such as the fallacious values of the nuclear data’s location parameter JXS array [2].

B. Doppler-broadening

In the neutron calculations, the effective cross section for a material at temperature T is defined as the cross section that gives the same reaction rate for stationary target nuclei as the real cross section gives for moving nuclei. In many cases of interest, the target motion is isotropic and the distribution of velocities can be described by the Maxwell-Boltzmann function [2]. Therefore, the standard form of the Doppler-broadened cross section is as follows:

where, α=M/(2kT), V=|v-v’|, k is Boltzmann’s constant, M is the target mass, v is the velocity of the incident particles, v’ is the velocity of the target, and σ(T, v) is the cross section for nuclides.

C. The preparation of the MCNP input file

Writing the information for radionuclide species and radionuclide temperatures that needed in the MCNP neutron transport calculations in the MCNP input file without interfering with the normal operation of MCNP. For example:

M1 82208.83c 3.91128E-3 $ t=356.9 K

This is the description of the material card of the MCNP input file. The original MCNP input file is on the left of the character '$’ and the comment is on the right. '82208’ stands for the radionuclide types of elements. The first two digits are for the number of protons and the last three digits are for the mass of the nuclides. '83c’ is for the database identifier. It can be defined according to the user’s own needs. In order to distinguish a normal radionuclide from a transient radionuclide, the transient radionuclide was defaulted to '*1.c’ deliberately. 't=356.9 K’ is the nuclides temperature. All of the letters are not case sensitive.

D. The programming of AMTND

This work completes the production of the database library and the neutron transport calculations in the process. The flow chart of AMTND is given in Fig. 1.

Fig. 1.Full-screen View

The flow chart of AMTND.

1. The AMTND reads radionuclide species information, database identifiers, and radionuclide temperature from the material card of the MCNP input file. Meanwhile, AMTND will merge the identical information of all the nuclides automatically and store it in a temporary structure array.

2. AMTND sealed the radionuclide species information, radionuclide identification (ID) number, and radionuclide evaluation database name (408 nuclides come from HENL-ADS) in a structure array. Using the radionuclide species information obtained from MCNP input file and the radionuclide ID number, a radionuclide evaluation database name could be found in the structure array above.

3. Using the information on the radionuclide species, radionuclide ID number, database identifiers, and the temperature obtained above, AMTND generated a Doppler-broaden input file automatically.

4. AMTND calls for Doppler-broaden modules. It reads the Doppler input file and the evaluation database files and then generates ACE format files and corresponding index files.

5. Merge the ACE format files and the corresponding index files into the original database. The new database can be used by MCNP.

A. Critical benchmark experiments

Critical benchmarks testing mainly deals with some thermal and fast reactor data [3-8]. In this test, 10 fast reactor experiments, which were released by Organization for Economic Cooperation and Development (OECD), were selected from the "International Handbook of Evaluated Criticality Safety Benchmark Experiments". The evaluations of the radionuclide come from HENDL-ADS [9] and the temperature is 293.6 K.

Based on criticality calculations using MCNP with the data produced by AMTND, the overall agreement of k_eff between experimentally measured and computed shown in Table 1 was very good. Some of the deviation between the calculation results and the experiment values was less than that produced by NJOY.

B. Time-consumed by generating nuclear data

As it is shown in Table 2, AMTND has been greatly improved by NJOY99 in generating an actinide nuclei data library. The average speedup ratio is 1.37 and the maximum speedup ratio is 1.61.

C. SEFOR Doppler benchmark

The SEFOR reactor [11] was designed to provide a Doppler measurement in an environment that is representative of an operating LMFBR with respect to the neutron spectrum, fuel temperature range, reactor composition, and the fuel microstructure. Standard fuel for SEFOR was mixed oxide (20% PuO₂, 80% UO₂), in which the Pu contained a minimum amount of ²⁴⁰Pu (∼8% of the Pu) and the U was depleted in ²³⁵U. A one-dimensional spherical model is analyzed in the present study. The 1-D SEFOR Doppler benchmark spherical model is shown in Fig. 2 and the spherical model region composition is shown in Table 3. More details can be found in Ref. [11]. Using the Doppler Calculation Model, the isothermal Doppler coefficient is computed as described below:

Fig. 2.Full-screen View

1-D SEFOR Doppler benchmark spherical model (in cm).

T₁ = 677 K; average fuel temperature at zero power;

T₂ =1365 K; average fuel temperature at 20 MW;

k₁ =neutron multiplication factor with the fuel at T₁;

k₂ =neutron multiplication factor with the fuel at T₂.

The radionuclide of the SEFOR model adopted is the evaluation documents of HENDL-ADS/MC [9]. The results are shown in Table 4. A correction factor for modeling of -0.0005 was obtained due to the resonance heterogeneity, subassembly heterogeneity and other factors [11]. The calculated value of the Doppler coefficients were -0.0081±0.0011 using NJOY-MCNP and -0.0080±0.0012 using AMTND-MCNP. These results compared well with the experimental result of -0.0080±0.0010.