Multidimensional Deterministic Nuclear Data Sensitivity and Uncertainty Code System: Method and Application

Kodeli, I.

doi:10.13182/nse00-43

Cited by 65 publications

(61 citation statements)

References 12 publications

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“…The code is a modular computer program package designed to solve the timeindependent or dependent multigroup discrete ordinates form of the Boltzmann transport equation in several different geometries. The modular construction of the package separates the input processing, the transport equation solving, and the post-processing (or edit) functions into distinct code modules: the (3) Nuclear data sensitivity and uncertainty calculations (code SUSD3D [10,11]): the development of SUSD3D started in the early 1990s in the scope of the French pressure vessel surveillance programme and the EC fusion project. XSUN-2017 includes the latest improved and extended version of the SUSD3D multidimensional nuclear cross-section sensitivity and uncertainty code, based on the first-order generalized perturbation theory.…”

Section: Xsun Computer Code System Descriptionmentioning

confidence: 99%

“…(i) Use of flux angular moments instead of much more voluminous angular fluxes [10,27] (ii) Demonstration of the equivalence of the constrained sensitivity method and covariance matrix fix-up in the calculation of the sensitivity to the secondary energy distributions, such as fission spectra sensitivities [21][22][23] (iii) Sensitivity and uncertainty calculations of the effective delayed neutron fraction ( ) [28][29][30].…”

Section: Special Features Developed In Susd3dmentioning

confidence: 99%

“…Since the sensitivity of eff would require the calculation of the second derivative ( eff being already equal the 1st derivative) another approach was developed in the SUSD3D code [10,11] based on the derivation of the Bretscher's prompt -ratio method:…”

Section: Beta-effective Sensitivity and Uncertainty Analysismentioning

confidence: 99%

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SUSD3D Computer Code as Part of the XSUN-2017 Windows Interface Environment for Deterministic Radiation Transport and Cross-Section Sensitivity-Uncertainty Analysis

Kodeli

Slavic

2017

Science and Technology of Nuclear Installations

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A Windows interface XSUN-2017 facilitating the deterministic radiation transport and cross-section sensitivity-uncertainty (S/U) calculation is presented. The package was developed to assist the users in the preparation of input cards, rapid modification, and execution of the complete chain of codes including TRANSX, PARTISN, and SUSD3D, all available from the OECD/NEA Data Bank and RSICC. The objective of this work was to make the input and output handling for these codes as user-friendly as possible, passing information among codes internally. XSUN-2017 allows a user-friendly viewing of results obtained from the PARTISN and SUSD3D programs. The first version of the Windows interface XSUN-2013 was developed in 2013 and submitted to OECD/NEA Data Bank Computer Code Collection and RSICC in early 2014. An updated version, XSUN-2017, will be released in 2017. The package includes also the new version of the SUSD3D code. The XSUN-2017 and SUSD3D code systems and recent improvements and updates are described. Examples of the use and validation are presented, including the S/U intercomparison exercise using the SNEAK-7 benchmarks involving the XSUN-2017 code system comparison with the codes such as TSUNAMI, SERPENT, and MCNP6, and the S/U analysis of the eff and eff parameters for the MYRRHA accelerator driven system (ADS).

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Section: Xsun Computer Code System Descriptionmentioning

confidence: 99%

Section: Special Features Developed In Susd3dmentioning

confidence: 99%

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SUSD3D Computer Code as Part of the XSUN-2017 Windows Interface Environment for Deterministic Radiation Transport and Cross-Section Sensitivity-Uncertainty Analysis

Kodeli

Slavic

2017

Science and Technology of Nuclear Installations

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“…In particular, the calculations have been performed with the SCALE6.2 system [7], in its third beta release, the MCNP6.1 code [8] and the SUSD3D code [9], part of the XSUN-2013 system [10]. Evaluated neutron-induced data from ENDF/B-VII.0 [11] have been used by SCALE and MCNP, and from ENDF/B-VII.1 [12] by SUSD3D.…”

Section: Sensitivity and Uncertainty Analysismentioning

confidence: 99%

Neutron-induced nuclear data for the MYRRHA fast spectrum facility

et al. 2017

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Abstract. The MYRRHA (Multi-purpose hYbrid Research Reactor for High-tech Applications) concept is a flexible experimental lead-bismuth cooled and mixed-oxide (MOX) fueled fast spectrum facility designed to operate both in sub-critical (accelerator driven) and critical modes. One of the key issues for the safe operation of the reactor is the uncertainty assessment during the design works. The main objective of the European project CHANDA (solving CHAllenges in Nuclear DAta) Work Package 10 is to improve MYRRHA relevant nuclear data in order to reduce the reactor parameter uncertainties derived from them. In order to achieve this goal, several tasks have been undertaken. First, a sensitivity study of MYRRHA integral parameters, such as energy dependent cross sections, fission spectra and neutron multiplicities, to nuclear data has been conducted resulting in a list of MYRRHA relevant quantities (nuclides and reactions). On the second task, an analysis of the existing experimental data and evaluations for the quantities included in the list has been carried out. In this framework, the impact on the multiplication factor of quantities from different nuclear data libraries for different nuclides, reactions and energy regions has been investigated on the MYRRHA MOX critical core model. As the next step, new experiments and evaluations will be performed in order to improve existing nuclear data libraries.

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“…Compared to deterministic uncertainty propagation [1], random sampling is less time efficient, however may be more accurate and less cumbersome.…”

Section: Introductionmentioning

confidence: 99%

Random sampling and validation of covariance matrices of resonance parameters

Plevnik

Žerovnik

2017

EPJ Web Conf.

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Abstract. Analytically exact methods for random sampling of arbitrary correlated parameters are presented. Emphasis is given on one hand on the possible inconsistencies in the covariance data, concentrating on the positive semi-definiteness and consistent sampling of correlated inherently positive parameters, and on the other hand on optimization of the implementation of the methods itself. The methods have been applied in the program ENDSAM, written in the Fortran language, which from a file from a nuclear data library of a chosen isotope in ENDF-6 format produces an arbitrary number of new files in ENDF-6 format which contain values of random samples of resonance parameters (in accordance with corresponding covariance matrices) in places of original values.The source code for the program ENDSAM is available from the OECD/NEA Data Bank. The program works in the following steps: reads resonance parameters and their covariance data from nuclear data library, checks whether the covariance data is consistent, and produces random samples of resonance parameters. The code has been validated with both realistic and artificial data to show that the produced samples are statistically consistent. Additionally, the code was used to validate covariance data in existing nuclear data libraries. A list of inconsistencies, observed in covariance data of resonance parameters in ENDF-VII.1, JEFF-3.2 and JENDL-4.0 is presented. For now, the work has been limited to resonance parameters, however the methods presented are general and can in principle be extended to sampling and validation of any nuclear data.

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Multidimensional Deterministic Nuclear Data Sensitivity and Uncertainty Code System: Method and Application

Cited by 65 publications

References 12 publications

SUSD3D Computer Code as Part of the XSUN-2017 Windows Interface Environment for Deterministic Radiation Transport and Cross-Section Sensitivity-Uncertainty Analysis

SUSD3D Computer Code as Part of the XSUN-2017 Windows Interface Environment for Deterministic Radiation Transport and Cross-Section Sensitivity-Uncertainty Analysis

Neutron-induced nuclear data for the MYRRHA fast spectrum facility

Random sampling and validation of covariance matrices of resonance parameters

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