RSICC CODE PACKAGE PSR-521

1. NAME AND TITLE

UNF: Code System to Calculate Multistep Compound Nucleus Neutron Cross-Sections and Spectra for Structural Materials, Version 2003-2004.

2. CONTRIBUTOR

The Chinese Institute of Atomic Energy, Beijing, People's Republic of China.

3. CODING LANGUAGE AND COMPUTER

Fortran 90; Pentium (P00521PC58600).

4. NATURE OF PROBLEM SOLVED

The UNF code (2003 version) calculates fast neutron reaction data of structure materials with incident energies from about 1 keV up to 20 MeV.

5. METHOD OF SOLUTION

The code consists of the spherical optical model, the unified Hauser-Feshbach and exciton model. The angular momentum dependent exciton model is established to describe the emissions from compound nucleus to the discrete levels of the residual nuclei in pre-equilibrium processes, while the equilibrium processes are described by the Hauser-Feshbach model with width fluctuation correction. The emissions to the discrete level in the multi-particle emissions for all opened channels are included. The double-differential cross sections of neutron and proton are calculated by the linear momentum dependent exciton state density. Since the improved pickup mechanism has been employed based on the Iwamoto-Harada model, the double-differential cross sections of alpha-particle, 3He, deuteron and triton can be calculated by using a new method based on the Fermi gas model. The recoil effects in multi-particle emissions from continuum state to discrete level as well as from continuum to continuum state are taken into account strictly, so the energy balance is held accurately in every reaction channels. If the calculated direct inelastic scattering data and the calculated direct reaction data of the outgoing charged particles are available from other codes, one can input them, so that the calculated results will included the effects of the direct reaction processes. To keep the energy balance, the recoil effects are taken into account for all of the reaction processes. The gamma-production data are also calculated. The calculated neutron reaction data can be output in the ENDF/B-6 format.

6. RESTRICTIONS OR LIMITATIONS

The code can handle decay sequence up to (n,3n) reaction channel. The total reaction channels are 14 (0:13). Note that the reaction channels (n,np) and (n,pn) as well as (n,na) and (n,an) should be treated as one channel, respectively. Thus the total reaction channels are 12 (0:11).

7. TYPICAL RUNNING TIME

In general for one isotope calculation the running time is about 3-4 minutes on a 1400 MHZ Pentium computer.

8. COMPUTER HARDWARE REQUIREMENTS

The code runs on personal computers.

9. COMPUTER SOFTWARE REQUIREMENTS

UNF is written in Fortran 90. An executable created on a Pentium IV 1400 MHZ under Windows 2000 with the Compaq Visual Fortran Version 6.6a compiler is included.

10. REFERENCES

a) included in document:

Jingshang Zhang, "User Manual of UNF Code," CNDC-0032 (December 2001).

b) background reference:

Jingshang Zhang, "UNF Code for Fast Neutron Reaction Data Calculations," Nuclear Science and Engineering: 142, pp. 207-219 (2002).

11. CONTENTS OF CODE PACKAGE

Included are the referenced document in 10.a on a CD which also includes the Fortran source, PC executable, and test case input and output.

12. DATE OF ABSTRACT

June 2003.

KEYWORDS: CROSS SECTION PROCESSING; ENDF FORMAT; KERMA; NUCLEAR MODELS