NEA-0547 HASSAN.
HASSAN, Time-Dependent Temperature Distribution and Stress and Strain in HTR Fuel Pins
NAME OR DESIGNATION OF PROGRAM, COMPUTER, NATURE OF PHYSICAL PROBLEM SOLVED, METHOD OF SOLUTION, RESTRICTIONS ON THE COMPLEXITY OF THE PROBLEM, TYPICAL RUNNING TIME, FEATURES, RELATED AND AUXILIARY PROGRAMS, STATUS, REFERENCES, MACHINE REQUIREMENTS, LANGUAGE, OPERATING SYSTEM OR MONITOR UNDER WHICH PROGRAM IS EXECUTED, ANY OTHER PROGRAMMING OR OPERATING INFORMATION OR RESTRICTIONS, NAME AND ESTABLISHMENT OF AUTHOR, MATERIAL, CATEGORIES
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| Program name | Package id | Status | Status date |
|---|---|---|---|
| HASSAN | NEA-0547/02 | Tested | 01-JAN-1978 |
Machines used:
| Package ID | Orig. computer | Test computer |
|---|---|---|
| NEA-0547/02 | IBM 370 series | IBM 370 series |
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3. NATURE OF PHYSICAL PROBLEM SOLVED
The programme calculates simultaneously, stepwise in time, the temperature, stress and strain distributions in high temperature reactor (HTR) fuel pins of tubular and hollow rod types. Any sequence of ramp and step changes of power may be followed. At each time step, the programme computes the operating stresses, also the residual stresses which develop if, at that point in time, the reactor is shut down and the fuel pin cools to a uniform temperature. Variations of material properties with temperature and neutron dose may be taken into account. In the structural analysis, axial symmetry and plane strain conditions are assumed. The programme cannot therefore give the stresses with sufficient accuracy in the region of pin ends or in regions of cross-pin flux or temperature tilt. In the thermal analysis, the fuel is assumed to be distributed uniformly along the fuelled length of the channel. The calculated temperature distributions are therefore not accurate for the pin end regions.
The programme calculates simultaneously, stepwise in time, the temperature, stress and strain distributions in high temperature reactor (HTR) fuel pins of tubular and hollow rod types. Any sequence of ramp and step changes of power may be followed. At each time step, the programme computes the operating stresses, also the residual stresses which develop if, at that point in time, the reactor is shut down and the fuel pin cools to a uniform temperature. Variations of material properties with temperature and neutron dose may be taken into account. In the structural analysis, axial symmetry and plane strain conditions are assumed. The programme cannot therefore give the stresses with sufficient accuracy in the region of pin ends or in regions of cross-pin flux or temperature tilt. In the thermal analysis, the fuel is assumed to be distributed uniformly along the fuelled length of the channel. The calculated temperature distributions are therefore not accurate for the pin end regions.
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4. METHOD OF SOLUTION
At each time step, the energy equation is integrated numerically along the channel to obtain the coolant axial temperature distribution(s). In the case of a tubular fuel pin, the radius of the adiabatic surface, which determines the heat fluxes to the inner and outer channels, is found by iteration. Radiation and convection heat transfer mechanisms are assumed, using appropriate empirical convective heat transfer correlations. The one-dimensional Fourier equation is used to determine the fuel pin internal temperature distribution. Heat transfer across internal (helium filled) gaps is assumed to be by conduction and radiation. The neutron-irradiation-induced dimensional changes (Wigner strains) are determined from sets of polynominals fitted to experimental data.
Incremental creep strains are calculated iteratively assuming a Maxwell model with dose and temperature dependent parameters. Total creep strains are obtained by summation of the incremental creep strains and the stresses are calculated using closed-form expressions. Radial displacements are calculated and checked for interaction between components. In the case of interaction, the interface pressure is found by iteration.
At each time step, the energy equation is integrated numerically along the channel to obtain the coolant axial temperature distribution(s). In the case of a tubular fuel pin, the radius of the adiabatic surface, which determines the heat fluxes to the inner and outer channels, is found by iteration. Radiation and convection heat transfer mechanisms are assumed, using appropriate empirical convective heat transfer correlations. The one-dimensional Fourier equation is used to determine the fuel pin internal temperature distribution. Heat transfer across internal (helium filled) gaps is assumed to be by conduction and radiation. The neutron-irradiation-induced dimensional changes (Wigner strains) are determined from sets of polynominals fitted to experimental data.
Incremental creep strains are calculated iteratively assuming a Maxwell model with dose and temperature dependent parameters. Total creep strains are obtained by summation of the incremental creep strains and the stresses are calculated using closed-form expressions. Radial displacements are calculated and checked for interaction between components. In the case of interaction, the interface pressure is found by iteration.
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5. RESTRICTIONS ON THE COMPLEXITY OF THE PROBLEM
The present version of HASSAN is restricted to the following maximum dimensions:
number of axial positions for stress/strain analysis = 10
number of axial positions for temperature calculation = 21
number of radial mesh points in each component (tube) = 20
The present version of HASSAN is restricted to the following maximum dimensions:
number of axial positions for stress/strain analysis = 10
number of axial positions for temperature calculation = 21
number of radial mesh points in each component (tube) = 20
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10. REFERENCES
- A. Jezernik, J.L.Head:
"Stresses in the Graphite Fuel Tubes of a High Temperature Gas Cooled Reactor"
D.P. Report 662 (January 1970).
- U. Weicht, W. Mueller:
"TAPIR-COCO-DELTAP, Three Computer Programmes for the Thermal Analysis of a Reactor Core Containing Internally and/or Externally Gas Cooled Prismatic Fuel Elements"
D.P. Report 748 (March 1971).
- A. Jezernik, J.L.Head:
"Stresses in the Graphite Fuel Tubes of a High Temperature Gas Cooled Reactor"
D.P. Report 662 (January 1970).
- U. Weicht, W. Mueller:
"TAPIR-COCO-DELTAP, Three Computer Programmes for the Thermal Analysis of a Reactor Core Containing Internally and/or Externally Gas Cooled Prismatic Fuel Elements"
D.P. Report 748 (March 1971).
NEA-0547/02, included references:
- A. Alujevic, J.L. Head and J.C. Uhry:HASSAN - A Computer Programme for Thermal and Structural Analysis
of Prismatic HTR Fuel Elements
DP report 826 (June 1974).
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NEA-0547/02
| File name | File description | Records |
|---|---|---|
| NEA0547_02.001 | INFORMATION | 2 |
| NEA0547_02.002 | SOURCE PROGRAM F4-EBCDIC | 2384 |
| NEA0547_02.003 | SAMPLE PROBLEM DATA | 88 |
| NEA0547_02.004 | SAMPLE PROBLEM PRINTED OUTPUT | 3921 |
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- F. Space - Time Kinetics, Coupled Neutronics - Hydrodynamics - Thermodynamics
- H. Heat Transfer and Fluid Flow
- I. Deformation and Stress Distributions, Structural Analysis and Engineering Design Studies
Keywords: Fourier transformation, HTR reactors, fuel elements, graphite moderated reactors, heat transfer, strains, stress analysis.