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NEA-1611 NORMA-FP.

NORMA-FP, Perform Subchannel Analysis from Converged Coarse-Mesh Nodal Solutions

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1. NAME OR DESIGNATION OF PROGRAM:  NORMA-FP
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2. COMPUTERS
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Program name Package id Status Status date
NORMA-FP NEA-1611/01 Tested 04-NOV-1999

Machines used:

Package ID Orig. computer Test computer
NEA-1611/01 IBM (generic) PC Pentium II 300
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3. DESCRIPTION OF PROGRAM OR FUNCTION

NORMA-FP unfolds the detailed flux and power distribution inside a selected heterogeneous fuel assembly starting from a global nodal solution supplied by the three-dimensional code NORMA or QUARK. Optionally, the neutronic analysis can be followed by a thermal-hydraulic static or transient subchannel analysis. Moreover, given a batch of fuel assemblies (possibly, the whole core), the code can identify the channel containing the most rated fuel rod, for which the results of a detailed neutronic and thermal-hydraulic analysis are issued. The actual heterogeneous structure of a node is accounted by pin power factors computed as functions of local burnup, coolant density and burnup-weighted coolant density.
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4. METHODS

The global solution, comprising such information as node-averaged fluxes and face-averaged partial currents, can be used to construct local high-order interpolates providing accurate approximations of group fluxes and power density inside homogenised nodes. In more detail, the flux distribution inside a homogeneous node is assumed to be separable in the x-y plane and in the axial (z) direction. An analytical solution is used for the axial shape while a non separable biquartic polynomial with 21 coefficients (the terms x3y3, x3y4, x4y3 and x4y4 are missing) is constructed as an approximation for the radial flux shape in the node. Finally, the resulting power distribution in the homogenised node is modulated by an array of rod power factors featuring the actual heterogeneous structure of the node and computed from a library of input reference values on the basis of local burnup, coolant density and burnup-weighted coolant density.
The thermal-hydraulic model used for the subchannel analysis of an individual fuel rod bundle (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 for hydrodynamic phenomena form a system of coupled nonlinear equations that are solved by the an upflow scheme (allowed when no reverse flow is predicted) or by a Newton-Raphson iteration procedure (needed when the vapor continuity equation is involved). 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 whose standard length of 4*106 bytes can be changed by modifying a PARAMETER statement in the source file (see the Installation Directions).
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6. TYPICAL RUNNING TIME

The cases of sample problem 1 dealing with a block of four 17*17 fuel rod assemblies require a few minutes 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. 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 prepared for ENEL-ATN/GNUM, Milan, July 1995
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9. STATUS
Package ID Status date Status
NEA-1611/01 04-NOV-1999 Tested at NEADB
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10. REFERENCES
NEA-1611/01, included references:
- E.Brega, R. Fontana, E. Salina
The NORMA-FP Program to Perform a Subchannel Analysis from Converged Coarse-Mesh
Nodal solutions (Rev.3)
N5/91/05/MI
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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-1611/01 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. Salina
                              Synthesis Srl
                              Via B. Garofalo 10
                              20133 Milano, Italy

                              E. Brega
                              ENEL SpA
                              Via Pozzobonelli 6
                              20162 Milano, Italy
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16. MATERIAL AVAILABLE
NEA-1611/01
Install.doc  Installation file
Norfpdoc.zip Documentation file
Norfpfor.zip Source file
Norfpsp1.zip Sample Problem 1
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, reactor cores, thermodynamics, two-group theory, two-phase flow.