Complik: A versatile tool for automatizing the representation of material data in MCNP models - 14412
Creators
- 1. Paul Scherrer Institute - PSI, Villigen PSI (Switzerland)
- 2. ecole Polytechnique Federale de Lausanne - EPFL, Lausanne (Switzerland)
Description
At the Paul Scherrer Institut, the CASMO/SIMULATE code system is used as reference scheme for core analyses of the Swiss reactors. Besides, Monte Carlo (MC) based neutron transport codes are in parallel being established both for out-of-core applications and full core analyses, in order to provide higher fidelity solutions to problems beyond the modelling capabilities of the standard core modelling methods. In this perspective, a necessity is to be able to conduct the MC calculations at any operating conditions, requiring thus proper initialisation of the thermal-hydraulic feedback as well as at any burnup, prompting the need for a detailed modelling of all material zones in the irradiated fuel assemblies. To tackle the later challenge, the development of the COMPLINK tool, aimed at initialising large-scale MC models of burned core configurations, using as basis upstream 2-D assembly/ 3-D core analyses results, was initiated and is presented in this paper. A main feature is the capability to extend in an automatic manner, a user given MC model in order to represent material data with a higher resolution, keeping track of the location of such materials also within nested repeated structures. This capability prevents the user from the burden of developing specific computational framework for a target application. Since the same functionalities are required to analyse out-of-core spent fuel managements systems, a first verification of COMPLINK is illustrated on the basis of a criticality safety evaluation (CSE) with MCNP6.1 of a Swiss PWR wet storage pool loaded with fresh and burned fuel assemblies. Starting from CASMO-5 depletion calculations at each axial layer of the depleted fuel assemblies and reconstructed based on realistic irradiation histories, COMPLINK is then applied to load the isotopic compositions into the material zones of the corresponding MCNP6.1 assembly models. On this basis, the CSE analyses are conducted with focus on investigating axial and radial modelling effects. All in all, the obtained results are found to be well in line with the impact on CSE expected at the considered burnup levels, providing thus confidence in the correct implementation of the COMPLINK functionalities and confirming its effectiveness as well as efficiency for the initialization of all material zones in large-scale MC models of complex and strongly heterogeneous burned configurations. (authors)
Additional details
Publishing Information
- Publisher
- American Nuclear Society - ANS
- Imprint Place
- La Grange Park, IL (United States)
- ISBN
- 978-0-89448-723-1
- Imprint Pagination
- 10 p.
Conference
- Title
- 2015 International Conference on Nuclear Criticality Safety
- Acronym
- ICNC 2015
- Dates
- 13-17 Sep 2015
- Place
- Charlotte, NC (United States)
INIS
- Country of Publication
- United States
- Country of Input or Organization
- France
- INIS RN
- 53017360
- Subject category
- S42: ENGINEERING; S97: MATHEMATICAL METHODS AND COMPUTING;
- Resource subtype / Literary indicator
- Conference
- Descriptors DEI
- BURNUP; CRITICALITY; FUEL ASSEMBLIES; FUEL MANAGEMENT; ISOTOPE RATIO; MONTE CARLO METHOD; NEUTRON TRANSPORT; PWR TYPE REACTORS; SAFETY ANALYSIS; SPENT FUELS; THERMAL HYDRAULICS; WET STORAGE
- Descriptors DEC
- CALCULATION METHODS; DIMENSIONLESS NUMBERS; ENERGY SOURCES; ENRICHED URANIUM REACTORS; FLUID MECHANICS; FUELS; HYDRAULICS; MANAGEMENT; MATERIALS; MECHANICS; NEUTRAL-PARTICLE TRANSPORT; NUCLEAR FUELS; NUCLEAR MATERIALS MANAGEMENT; POWER REACTORS; RADIATION TRANSPORT; REACTOR MATERIALS; REACTORS; STORAGE; THERMAL REACTORS; WATER COOLED REACTORS; WATER MODERATED REACTORS
Optional Information
- Notes
- 11 refs.; available on CD Rom from American Nuclear Society - ANS, 555 North Kensington Avenue, La Grange Park, IL 60526 (US)