Verification of the stationary module of the ShIPR software system for modelling experiments of the ASTRA HTGR type critical facility
-
M. N. Zizin
Abstract
Coupled neutronic and thermal hydraulic calculation codes are verified for calculating the design of modern and prospective types of nuclear reactors. This verification is done by comparing experimental and calculated results for stationary and transient conditions. This paper presents ShIPR (Shell of Intelligent Package for Reactor) Integrated Development Environment with automatic generation of head programs based on the chain of computational modules. The aim of this study is to find the reason of a discrepancy in the modelling of sub-critical states that was found in previous work. The comparison of ShIPR stationary module with Monte-Carlo code (MCU) and experimental results on ASTRA HTGR critical facility was presented in the paper. To compare the detector readings and MCU calculation with the ShIPR module, the interpolation cross-section procedure was performed. This procedure allows simulating 235U fission reaction rates (detector readings) in the complicate, annular core, using the micro-cross sections prepared by the cell-lattice code. We found that the calculation accuracy of stationary ShIPR module is on an acceptable level but the macro constants for control rods need to be prepared independently, given with surroundings.
© 2020 by Walter de Gruyter Berlin/Boston
Articles in the same Issue
- CONTENTS
- CALENDAR OF EVENTS · VERANSTALTUNGSKALENDER
- Verification of the stationary module of the ShIPR software system for modelling experiments of the ASTRA HTGR type critical facility
- Appraisal of new density coefficient on integrated-nanoparticles concrete in nuclear protection
- Comparison of back propagation network and radial basis function network in Departure from Nucleate Boiling Ratio (DNBR) calculation
- Developing a futuristic multi-objective optimization of the fuel management problems for the nuclear research reactors
- Sensitivity of the neutron multiplication factor to water ingress into a spent fuel cask
- Evaluation of single heated channel and subchannel modeling of a nuclear once through steam generator (OTSG)
- Note · Imprint
Articles in the same Issue
- CONTENTS
- CALENDAR OF EVENTS · VERANSTALTUNGSKALENDER
- Verification of the stationary module of the ShIPR software system for modelling experiments of the ASTRA HTGR type critical facility
- Appraisal of new density coefficient on integrated-nanoparticles concrete in nuclear protection
- Comparison of back propagation network and radial basis function network in Departure from Nucleate Boiling Ratio (DNBR) calculation
- Developing a futuristic multi-objective optimization of the fuel management problems for the nuclear research reactors
- Sensitivity of the neutron multiplication factor to water ingress into a spent fuel cask
- Evaluation of single heated channel and subchannel modeling of a nuclear once through steam generator (OTSG)
- Note · Imprint
