NAME OR DESIGNATION OF PROGRAM, COMPUTER, DESCRIPTION OF PROBLEM OR FUNCTION, METHOD OF SOLUTION, RESTRICTIONS ON THE COMPLEXITY OF THE PROBLEM, TYPICAL RUNNING TIME, UNUSUAL FEATURES OF THE PROGRAM, RELATED AND AUXILIARY PROGRAMS, STATUS, REFERENCES, MACHINE REQUIREMENTS, LANGUAGE, OPERATING SYSTEM UNDER WHICH PROGRAM IS EXECUTED, 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 |
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SPLOSH-3 | NEA-0609/01 | Tested | 02-DEC-1983 |

Machines used:

Package ID | Orig. computer | Test computer |
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NEA-0609/01 | IBM 3081 | IBM 3081 |

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4. METHOD OF SOLUTION

An axial single channel model with equally spaced mesh intervals is used to represent the neutronics of the re- actor core. Two-group cross section data are normally supplied by means of file handling procedures that are automatically linked to lattice codes. The initial steady state calculation may be for a subcritical core with a fixed source or to determine critical poison concentration or moderator height. The neutronics solution has a number of unusual features to provide accuracy and rapid conver- gence.

A radial finite difference model is used for heat conduction through the fuel pin, gas gap and can. Appropriate convective, boiling or post-dryout heat transfer correlations are used at the can-coolant interface.

The hydraulics model includes the important features of the SGHWR primary loop including 'slave' channels in parallel with the 'mean' channel. Standard mass, energy and momentum equations are solved explicitly with backward values in time or space being used as appropriate. Circuit features modelled include pumps, spray cool- ing and the SGHWR steam drum. Steady state convergence is based on a search for the channel inlet coolant velocity that gives a circuit pressure balance; this is always very rapid and reliable.

Perturbations to almost any feature of the circuit model may be specified by the user although blowdown calculations resulting in critical or reversed flows are not permitted. Automatic reactor trips may be defined and the ensuing actions of moderator dumping and rod firing can be specified.

Two basic steady state calculations are possible. For the study of startup transients a fixed source distribution may be input, or a photoneutron source distribution calculated and used to determine the neutron flux in an initially subcritical core. The user speci- fies a required value of K-effective (less than unity) and the code simultaneously determines an appropriate multiplier on the fission yield cross sections to meet this requirement. The other type of steady state calculation is for a reactor at a specified power level (where fixed sources are normally ignored) and a critical search is made for the boron concentration or moderator height that gives K-effective equal to unity.

An axial single channel model with equally spaced mesh intervals is used to represent the neutronics of the re- actor core. Two-group cross section data are normally supplied by means of file handling procedures that are automatically linked to lattice codes. The initial steady state calculation may be for a subcritical core with a fixed source or to determine critical poison concentration or moderator height. The neutronics solution has a number of unusual features to provide accuracy and rapid conver- gence.

A radial finite difference model is used for heat conduction through the fuel pin, gas gap and can. Appropriate convective, boiling or post-dryout heat transfer correlations are used at the can-coolant interface.

The hydraulics model includes the important features of the SGHWR primary loop including 'slave' channels in parallel with the 'mean' channel. Standard mass, energy and momentum equations are solved explicitly with backward values in time or space being used as appropriate. Circuit features modelled include pumps, spray cool- ing and the SGHWR steam drum. Steady state convergence is based on a search for the channel inlet coolant velocity that gives a circuit pressure balance; this is always very rapid and reliable.

Perturbations to almost any feature of the circuit model may be specified by the user although blowdown calculations resulting in critical or reversed flows are not permitted. Automatic reactor trips may be defined and the ensuing actions of moderator dumping and rod firing can be specified.

Two basic steady state calculations are possible. For the study of startup transients a fixed source distribution may be input, or a photoneutron source distribution calculated and used to determine the neutron flux in an initially subcritical core. The user speci- fies a required value of K-effective (less than unity) and the code simultaneously determines an appropriate multiplier on the fission yield cross sections to meet this requirement. The other type of steady state calculation is for a reactor at a specified power level (where fixed sources are normally ignored) and a critical search is made for the boron concentration or moderator height that gives K-effective equal to unity.

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7. UNUSUAL FEATURES OF THE PROGRAM

The axial neutronics solution has a number of notable features. At each steady state iteration and time step during a transient, a local analytic solution of the dif- fusion equation, allowing for the presence of moderator and control rod discontinuities, is used to determine neutron current boundary conditions for each mesh interval and also the ratio of spatially integrated to mesh point values of the flux. This ratio is used as a weighting factor on all neutron cross sections so that the final steady state flux solution is an exact solution of the diffusion equations, independent of the number and size of mesh intervals.

There are therefore no discontinuities in reactivity or rate of re- activity insertion as moderator level or rod positions cross mesh boundaries.

The axial neutronics solution has a number of notable features. At each steady state iteration and time step during a transient, a local analytic solution of the dif- fusion equation, allowing for the presence of moderator and control rod discontinuities, is used to determine neutron current boundary conditions for each mesh interval and also the ratio of spatially integrated to mesh point values of the flux. This ratio is used as a weighting factor on all neutron cross sections so that the final steady state flux solution is an exact solution of the diffusion equations, independent of the number and size of mesh intervals.

There are therefore no discontinuities in reactivity or rate of re- activity insertion as moderator level or rod positions cross mesh boundaries.

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8. RELATED AND AUXILIARY PROGRAMS

The two group cross section data may be read in from the WIMS lattice codes output.

SPLOSH-3 has been developed from the code SPLOSH-2 so as to provide a complete representation of the primary circuit of SGHWR's. New features include pressure dependent fluid properties, options to specify the primary circuit as a closed loop instead of as an open loop with boundary conditions, and a sophisticated treatment of the neutronics.

The two group cross section data may be read in from the WIMS lattice codes output.

SPLOSH-3 has been developed from the code SPLOSH-2 so as to provide a complete representation of the primary circuit of SGHWR's. New features include pressure dependent fluid properties, options to specify the primary circuit as a closed loop instead of as an open loop with boundary conditions, and a sophisticated treatment of the neutronics.

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NEA-0609/01

File name | File description | Records |
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NEA0609_01.001 | SPLOSH-3 INFORMATION FILE | 75 |

NEA0609_01.002 | SPLOSH-3 JOB CONTROL INFORMATION | 146 |

NEA0609_01.003 | SPLOSH-3 SOURCE PROGRAM (FORTRAN-H) | 16412 |

NEA0609_01.004 | SPLOSH-3 SUBROUTINES (FORTRAN-G1) | 234 |

NEA0609_01.005 | SPLOSH-3 SUBROUTINES (ASSEMBLER) | 768 |

NEA0609_01.006 | BINARY X-SEC LIB IN FACADE FORMAT | 631 |

NEA0609_01.007 | SPLOSH-3 INPUT FOR TEST CASE 1 | 39 |

NEA0609_01.008 | SPLOSH-3 INPUT FOR TEST CASE 2 | 155 |

NEA0609_01.009 | SPLOSH-3 INPUT FOR TEST CASE 3 | 69 |

NEA0609_01.010 | SPLOSH-3 PRINTED OUTPUT OF TEST CASE 1 | 351 |

NEA0609_01.011 | SPLOSH-3 PRINTED OUTPUT OF TEST CASE 2 | 3624 |

NEA0609_01.012 | SPLOSH-3 PRINTED OUTPUT OF TEST CASE 3 | 3124 |

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- F. Space - Time Kinetics, Coupled Neutronics - Hydrodynamics - Thermodynamics

Keywords: finite difference method, fuel elements, heat transfer, kinetics, moderators, one-dimensional, poisoning, reactivity, reactor cores, thermodynamics, time dependence.