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NEA-1388 NORMA.

NORMA, Neutron & Thermo-Hydraulic Behaviour of LWR's by Coarse-Mesh Diffusion Methods

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1. NAME OR DESIGNATION OF PROGRAM:  NORMA
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2. COMPUTERS
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Program name Package id Status Status date
NORMA NEA-1388/01 Tested 27-OCT-1999

Machines used:

Package ID Orig. computer Test computer
NEA-1388/01 IBM PC PC Pentium III 500
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3. DESCRIPTION OF PROGRAM OR FUNCTION

The NORMA code is intended to solve multigroup diffusion problems or two-group diffusion-depletion problems in three dimensions.  In order to simulate the fuel depletion, the core lifetime is divided into suitable burnup steps in which the nuclear parameters, required by the spatial solution of the diffusion equation in steady-state critical condition, are computed as functions of local burnup, coolant density and burnup-weighted coolant density (or spectral ratio) and are corrected for dilute Boron, Sm and Xe poisoning and Doppler effects with the further provision for an automatic dilute Boron adjustment to achieve criticality.  Moreover, at the beginning of each burnup step, general directions for fuel management are provided.
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4. METHODS

The burnup model is implicit and the overall solution scheme at each burnup step is iterative.  The multigroup diffusion equation is approximated by a coarse-mesh fourth-order polynomial method.  More precisely,  the numerical approximation is represented by coarse-mesh finite-difference equations corrected by discontinuity factors internally computed so as to match the accuracy of a fourth-order polynomial approach.  Then, the finite-difference equations are solved for the nodal neutron fluxes by the usual inner/outer scheme.
The equivalent homogenised nuclear parameters including the discontinuity factors correcting for homogenisation errors, according to Henry's generalised equivalence theory, are updated by trilinear interpolation in input libraries of reference values.
The thermal-hydraulic model (an upgraded version of the COBRA-3C code called COBRA-EN) is based on three partial differential equations that describe the conservation of mass, energy and momentum for the water liquid/vapor mixture and the interaction of the two-phase coolant with the system structures.  Optionally, a fourth equation can be added which tracks the vapor mass separately and which, along with the correlations for vapor generation and slip ratio, replaces the subcooled quality and quality/void fraction correlations, needed by the homogeneous model.  In each coolant channel, the one-dimensional (z) fluid dynamics equations in the vertical direction as well as the one-dimensional (r) equation in the horizontal direction that models the heat transfer in solid structures are approximated by finite differences.  The resulting equations of hydrodynamic phenomena form a system of coupled nonlinear equations that are solved by the upflow scheme (allowed only when no reverse flow is predicted) or by a Newton-Raphson iteration procedure (needed when the vapor mass continuity equation is added).  The heat-transfer equations in the solid structures are treated implicitly.  Moreover, a full boiling curve is provided, comprising the basic heat-transfer regimes, each represented by a set of optional correlations for the heat-transfer coefficient between a solid surface and the coolant bulk.
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5. RESTRICTIONS ON THE COMPLEXITY OF THE PROBLEM

The data-dependent arrays are contained in the named Common block BLANK whos standard lenght of 4x10(6) bytes can be changed by modifying a PARAMETER statement in an include file (see the Installation Directions).
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6. TYPICAL RUNNING TIME

The cases of sample problem 1 which use a mesh grid of 9*9*23 nodes (i.e., one planar node per fuel assembly) require 2*4 min for each burnup step while the case using a mesh grid of 17*17*23 nodes (i.e., four planar nodes per fuel assembly) requires 7 min on a PC-486/100.
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7. UNUSUAL FEATURES
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8. RELATED OR AUXILIARY PROGRAMS

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 prepared for ENEL-ATN/GNUM, Milan, 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, Milan, September 1991
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9. STATUS
Package ID Status date Status
NEA-1388/01 27-OCT-1999 Tested at NEADB
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10. REFERENCES
NEA-1388/01, included references:
- E. Salina and E. Brega
The NORMA Program for Simulating the Long-Term Neutronic and Thermal-Hydraulic
Behaviour of Large LWR's by Three-Dimensional Coarse-Mesh Diffusion Methods.
Rep. No. 1034/2
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11. HARDWARE REQUIREMENTS

A Personal Computer with 486 or Pentium processor and at least 8 Mb of RAM
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12. PROGRAMMING LANGUAGE(S) USED
Package ID Computer language
NEA-1388/01 FORTRAN-77
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13. SOFTWARE REQUIREMENTS

DOS or WINDOWS provided with MS FORTRAN Power Station Compiler version 1.0 or higher.
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14. OTHER PROGRAMMING OR OPERATING INFORMATION OR RESTRICTIONS
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15. NAME AND ESTABLISHMENT OF AUTHORS

E. Brega
            ENEL SpA
            Via Pozzobonelli 6
            20162 Milano, Italy

            E. Salina
            Synthesis Srl
            Via B. Garofalo 10
            20133 Milano, Italy
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16. MATERIAL AVAILABLE
NEA-1388/01
Install.doc Installation procedure
Normadoc.zip Documentation
Normafor.zip Source code
Normasp1.zip Sample problem 1
Normasp2.zip Sample problem 2
Pkunzip.exe Unzipping software
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17. CATEGORIES
  • F. Space - Time Kinetics, Coupled Neutronics - Hydrodynamics - Thermodynamics

Keywords: LWR reactors, fuel rods, heat transfer, neutron flux, power distribution, reactivity, reactor cores, reactor safety, thermodynamics, three-dimensional, two-group theory, two-phase flow.