RSICC CODE PACKAGE PSR-471

1. NAME AND TITLE

NORMA: Code System to Solve Burnup Dependent Neutron Diffusion Equations in Two and Three Dimensions.

2. CONTRIBUTORS

Synthesis Srl, Milano, Italy and ENEL SpA, Milano, Italy through the OECD Nuclear Energy Agency Data Bank, Issy-Les Molineaux, France.

3. CODING LANGUAGE AND COMPUTER

Fortran 77; PC 486 (P00471PC58600).

4. NATURE OF PROBLEM SOLVED

The NORMA program is designed to simulate the follow-up of a PWR core and can solve multigroup diffusion problems or two-group diffusion-depletion problems in three dimensions.

5. METHOD OF SOLUTION

The burnup model is implicit, and the diffusion equations are approximated by a coarse-mesh Polynomial Nodal Method, developed at MIT, and originally included in the CONQUEST core kinetics code. The equivalent homogenized nuclear parameters are supplemented by the discontinuity factors, according to Henry's generalized equivalence theory. These parameters are updated by trilinear interpolation in three-entry tables and are corrected for dilute boron, Sm and Xe poisoning and Doppler effects with the further provision for an automatic boron adjustment to achieve criticality.

In order to calculate the local coolant densities and temperatures, two thermal-hydraulic models can be applied optionally, the earlier CRTN model dealing with four zones along each channel and the last available version of the more sophisticated COBRA-EN model. NORMA also allows a separate detailed output of the results for each burnup step.

6. RESTRICTIONS OR LIMITATIONS

The data-dependent arrays are contained in the named common block BLANK whose standard length can be changed by modifying a PARAMETER statement in an include file (see the Installation Directions).

7. TYPICAL RUNNING TIME

Each test case ran in 10 seconds or less on a Pentium III processor of 500 MHZ under Windows NT.

8. COMPUTER HARDWARE REQUIREMENTS

NORMA runs on a Personal Computer with 486 or Pentium processor and at least 8 Mb of RAM.

9. COMPUTER SOFTWARE REQUIREMENTS

The code is written almost entirely in FORTRAN 77 and was developed with MS FORTRAN Power Station Compiler. The included PC executables were created with Digital Visual Fortran Version 6.0A and tested at RSICC on a Pentium/266 under Windows95.

10. REFERENCES

a) included in package:

E. Salina, E. Brega, "The NORMA Program for Simulating the Long-Term Neutronic and Thermal-Hydraulic Behavior of Large LWR's by Three-Dimensional Coarse-Mesh Diffusion Methods," Synthesis Srl, rep. 1034/2 (July 1995).

b) Background references:

E. Salina, G. Alloggio, E. Brega, "QUARK: A Computer Code for the Neutronic and Thermal-Hydraulic Space- and Time-Dependent Analysis of Light Water Reactor Cores by Advanced Nodal Techniques," Synthesis Srl, rep. 1034/1 (September 1994).

E. Brega, R. Fontana, E. Salina, "The NORMA-FP Program to Perform a Subchannel Analysis from Converged Coarse-Mesh Nodal Solutions (Rev. 3)," ENEL-DSR-CRTN-N5/91/05/MI (September 1991).

11. CONTENTS OF CODE PACKAGE

Included are the referenced document in (10.a) and two DS/HD diskettes with self-extracting compressed DOS files which contain the FORTRAN source files, PC executable, the documentation files (in WinWord) and the input/output files for the two sample problems described in User's Manual.

12. DATE OF ABSTRACT

June 2000.

KEYWORDS: HEAT TRANSFER; LWR; REACTOR SAFETY; THERMAL HYDRAULICS