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Title: Validation of simulation codes for future systems : motivations, approach, and the role of nuclear data.

Abstract

The validation of advanced simulation tools will still play a very significant role in several areas of reactor system analysis. This is the case of reactor physics and neutronics, where nuclear data uncertainties still play a crucial role for many core and fuel cycle parameters. The present paper gives a summary of validation motivations, objectives and approach. A validation effort is in particular necessary in the frame of advanced (e.g. Generation-IV or GNEP) reactors and associated fuel cycles assessment and design. Validation of simulation codes is complementary to the 'verification' process. In fact, 'verification' addresses the question 'are we solving the equations correctly' while validation addresses the question 'are we solving the correct equations with the correct parameters'. Verification implies comparisons with 'reference' equation solutions or with analytical solutions, when they exist. Most of what is called 'numerical validation' falls in this category. Validation strategies differ according to the relative weight of the methods and of the parameters that enter into the simulation tools. Most validation is based on experiments, and the field of neutronics where a 'robust' physics description model exists and which is function of 'input' parameters not fully known, will be the focus of this paper. Inmore » fact, in the case of reactor core, shielding and fuel cycle physics the model (theory) is well established (the Boltzmann and Bateman equations) and the parameters are the nuclear cross-sections, decay data etc. Two types of validation approaches can and have been used: (a) Mock-up experiments ('global' validation): need for a very close experimental simulation of a reference configuration. Bias factors cannot be extrapolated beyond reference configuration; (b) Use of 'clean', 'representative' integral experiments ('bias factor and adjustment' method). Allows to define bias factors, uncertainties and can be used for a wide range of applications. It also allows to define 'adjusted' application libraries or even 'adjusted' data files. The use of this last approach has been particularly successful in the design of SUPERPHENIX. In fact the prediction of the critical mass has been remarkably close to the experimental value observed at reactor start up (discrepancy of {approx}3 out of {approx}400 core sub-assemblies).« less

Authors:
; ; ; ; ;
Publication Date:
Research Org.:
Argonne National Lab. (ANL), Argonne, IL (United States)
Sponsoring Org.:
USDOE Office of Science (SC)
OSTI Identifier:
992043
Report Number(s):
ANL/NE/CP-60290
TRN: US1007602
DOE Contract Number:
DE-AC02-06CH11357
Resource Type:
Conference
Resource Relation:
Conference: NEMEA-4 Workshop; Oct. 16, 2007 - Oct. 18, 2007; Prague, Czech Republic
Country of Publication:
United States
Language:
ENGLISH
Subject:
21 SPECIFIC NUCLEAR REACTORS AND ASSOCIATED PLANTS; ANALYTICAL SOLUTION; CONFIGURATION; CRITICAL MASS; CROSS SECTIONS; DECAY; DESIGN; FORECASTING; FUEL CYCLE; PHYSICS; REACTOR CORES; REACTOR PHYSICS; REACTOR START-UP; SHIELDING; SIMULATION; VALIDATION; VERIFICATION

Citation Formats

Palmiotti, G., Salvatores, M., Aliberti, G., Nuclear Engineering Division, INL, and CEA Cadarache. Validation of simulation codes for future systems : motivations, approach, and the role of nuclear data.. United States: N. p., 2007. Web.
Palmiotti, G., Salvatores, M., Aliberti, G., Nuclear Engineering Division, INL, & CEA Cadarache. Validation of simulation codes for future systems : motivations, approach, and the role of nuclear data.. United States.
Palmiotti, G., Salvatores, M., Aliberti, G., Nuclear Engineering Division, INL, and CEA Cadarache. Mon . "Validation of simulation codes for future systems : motivations, approach, and the role of nuclear data.". United States. doi:.
@article{osti_992043,
title = {Validation of simulation codes for future systems : motivations, approach, and the role of nuclear data.},
author = {Palmiotti, G. and Salvatores, M. and Aliberti, G. and Nuclear Engineering Division and INL and CEA Cadarache},
abstractNote = {The validation of advanced simulation tools will still play a very significant role in several areas of reactor system analysis. This is the case of reactor physics and neutronics, where nuclear data uncertainties still play a crucial role for many core and fuel cycle parameters. The present paper gives a summary of validation motivations, objectives and approach. A validation effort is in particular necessary in the frame of advanced (e.g. Generation-IV or GNEP) reactors and associated fuel cycles assessment and design. Validation of simulation codes is complementary to the 'verification' process. In fact, 'verification' addresses the question 'are we solving the equations correctly' while validation addresses the question 'are we solving the correct equations with the correct parameters'. Verification implies comparisons with 'reference' equation solutions or with analytical solutions, when they exist. Most of what is called 'numerical validation' falls in this category. Validation strategies differ according to the relative weight of the methods and of the parameters that enter into the simulation tools. Most validation is based on experiments, and the field of neutronics where a 'robust' physics description model exists and which is function of 'input' parameters not fully known, will be the focus of this paper. In fact, in the case of reactor core, shielding and fuel cycle physics the model (theory) is well established (the Boltzmann and Bateman equations) and the parameters are the nuclear cross-sections, decay data etc. Two types of validation approaches can and have been used: (a) Mock-up experiments ('global' validation): need for a very close experimental simulation of a reference configuration. Bias factors cannot be extrapolated beyond reference configuration; (b) Use of 'clean', 'representative' integral experiments ('bias factor and adjustment' method). Allows to define bias factors, uncertainties and can be used for a wide range of applications. It also allows to define 'adjusted' application libraries or even 'adjusted' data files. The use of this last approach has been particularly successful in the design of SUPERPHENIX. In fact the prediction of the critical mass has been remarkably close to the experimental value observed at reactor start up (discrepancy of {approx}3 out of {approx}400 core sub-assemblies).},
doi = {},
journal = {},
number = ,
volume = ,
place = {United States},
year = {Mon Jan 01 00:00:00 EST 2007},
month = {Mon Jan 01 00:00:00 EST 2007}
}

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  • The validation of advanced simulation tools will still play a very significant role in several areas of reactor system analysis. This is the case of reactor physics and neutronics, where nuclear data uncertainties still play a crucial role for many core and fuel cycle parameters. The present paper gives a summary of validation motivations, objectives and approach. A validation effort is in particular necessary in the frame of advanced (e.g. Generation-IV or GNEP) reactors and associated fuel cycles assessment and design.
  • This paper presents a global approach to the validation of the parameters that enter into the neutronics simulation tools for advanced fast reactors with the objective to reduce the uncertainties associated to crucial design parameters. This global approach makes use of sensitivity/uncertainty methods; statistical data adjustments; integral experiment selection, analysis and “representativity” quantification with respect to a reference system; scientifically based cross section covariance data and appropriate methods for their use in multigroup calculations. This global approach has been applied to the uncertainty reduction on the criticality of the Advanced Burner Reactor, (both metal and oxide core versions) presently investigatedmore » in the frame of the GNEP initiative. The results obtained are very encouraging and allow to indicate some possible improvements of the ENDF/B-VII data file.« less
  • This paper discusses the generation and use of a data package for the validation of a computer model of a horizontally fired, cylindrical, research tunnel furnace. The combustion experiments discussed in this paper involved the combustion of either coal or oil in eleven separate experiments. The volatile content of the coal varied from 21 to 35 percent by weight and the swirl number of the flow varied from 0.0 to 0.4. The measurement scheme included the axial variation of the total and radiant heat flux from the flame, and a mapping of the gas temperature and major species concentration profiles.more » To complete the measurements inside the furnace a mapping of the gas velocity profiles was made at two total air flow rates and seven swirl settings ranging from a swirl number of 0.0 to 0.9. The use of these data to validate a combustion simulation code is demonstrated in this paper. The computer code used for this study is called 'MODTUN' and was developed at Imperial College. This computer program is based on the 'TEACH' code and incorporates the necessary sub routines to model the devolatilization and combustion of pulverized coal. A reasonable measure of success was obtained in comparing the measured to predicted data. By using this data package not only was it possible to determine the appropriate models for the physical processes involved but empirical parameters for these models were also identified. The most significant of these is the amount of volatile material released relative to that predicted by proximate analysis.« less