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In order to assess the accuracy and validity of subchannel, system, and computational fluid dynamics codes, the Paul Scherrer Institut has participated in the OECD/NRC PSBT benchmark with the thermal-hydraulic system code TRACE5.0 developed by US NRC, the subchannel code FLICA4 developed by CEA, and the computational fluid dynamic code STAR-CD deve...
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The purpose of the presented research in this paper was to perform an uncertainty evaluation of a Main Steam Line Break (MSLB) transient scenario occurring in a PWR applying the CIAU-TN methodology (Code with -the capability -of Internal Assessment of Uncertainty for Thermal-hydraulics/Neutronics coupled codes). The work has been carried out within...
Incorporating full three-dimensional models of the reactor core into system transient codes allows for a "best-estimate" calculation of interactions between the core behavior and plant dynamics. Considerable efforts have been made in various countries and organizations on the development of coupled thermal-hydraulic and neutronics codes. Appropriat...
Accurate prediction of steam volume fraction and of the boiling crisis (either DNB or dryout) occurrence is a key safety-relevant issue. Decades of experience have been built up both in experimental investigation and code development and qualification; however, there is still a large margin to improve and refine the modelling approaches. The qualif...
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... Three heat structures were used: One for the pins at full power facing the inner channel (red), one for the pins at full power facing the outer channel (yellow) and a third for the fuel pins at reduced power (orange). As explained in Kim et al. (2012), to minimise the need for computational resources the symmetry of the subchannel was exploited. The size of the STAR-CCM+ model was reduced by a factor of 4 in both standalone and coupled simulations. ...
... PSBT test section cross section geometry -Adapted fromRubin et al. (2010) As explained inKim et al. (2012), to minimise the need for computational resources the symmetry of the subchannel was exploited. As a result, it was possible to reduce the size of the model by a factor of 8. ...
The thermal hydraulic analysis of nuclear reactors is largely performed by what are known as system codes. These codes predict the flows in the complex network of pipes, pumps, vessels and heat exchangers that together form the thermal hydraulic systems of a nuclear reactor.
These codes have been used for many decades and are now very well established. Given this long process of refinement, they are able to produce remarkably accurate predictions of plant behaviour under both steady and transient conditions. Modern CFD is able to produce high-quality predictions of flows in complex geometries, but only with the use of large computing resources. It would be impractical to build a CFD model of, for example, the entire primary circuit of a PWR. However, it is possible to model with adequate fidelity much of the primary circuit using a cheaper one-dimensional system code, and it may only be in a limited part of the circuit that full three-dimensional effects are important.
A coupling scheme was developed to couple the CFD software STAR-CCM+ and the system code RELAP5-3D. The structure of the scheme is presented, together with validations for single phase flow in smooth pipes in both transient and steady state cases. Attention was also given to the problem of reconstructing the flow profile at the inlet of the CFD model under the hypothesis of fully developed flow. This problem arises when flow data has to be passed from the one-dimensional system code to the three-dimensional CFD software.
The coupling scheme was then modified to be able to perform multiphase simulations. The PWR subchannel and bundle test (PSBT) benchmark was used to validate the multiphase coupling methodology.
The OpenFOAM-based GeN-Foam solver has recently been extended to allow simulating water boiling based on a porous-medium approach. This paper describes such extension and preliminarily validates it based on the OECD/NRC PSBT benchmark. Three benchmark exercises are included: steady-state single sub-channel exercise; steady-state rod bundle exercise; and transient rod bundle exercise. Model sensitivity analysis is conducted, with a simple empirical turbulent model that is developed and calibrated to account for the mixing effect caused by the interaction of the two phases. The calculated results are compared with the experimental data, and most of the predictions fall within the ± 2σexp error bar. The results indicate that the solver can simulate water boiling heat transfer process and predict void fraction distribution accurately, at least for the cases of the PSBT benchmark.
Dedicated subchannel analysis codes have been employed to subchannel analyses. However, the recent development of thermal-hydraulic system codes with advanced two-phase flow models and three-dimensional components has resulted in more accurate and precise prediction of multiphase phenomena in nuclear reactor systems. The accurate prediction of the void fraction is one of the most important factors in subchannel analyses. This paper aims at assessing the applicability of three different system codes to the prediction of the void fraction against NUPEC experiment used for OECD/NRC PSBT benchmark. In this study, all cases for single channels and bundles in the benchmark database have been analyzed by using TRACE V5.0 Patch 5, MARS-KS 1.4, and RELAP5/MOD3.3 Patch 5, respectively. It is concluded that the applied system codes have a systematic difficulty in void fraction prediction under low void fraction conditions and additional assessment should be performed for low void fraction conditions including subcooled boiling regime.
The development of subchannel models for fuel assemblies and reactor cores requires accurate information on flow distribution, wall friction and loss coefficients in order to
accurately predict the pressure, temperature and flow distribution on a subchannel level. This paper discusses the use of Computational Fluid Dynamics (CFD) simulations as a
practical tool for characterising inlet velocity boundary conditions, an approximation of wall friction factor and spacer grid pressure loss coefficients, which are of fundamental
importance to correctly generate a consistent subchannel model of a given assembly system. The geometry of the simplified PWR assembly presented here is based on the NUPEC PWR subchannel and bundle tests. Comparison of the derived friction factors and grid pressure loss coefficient with published and recommended values are reported. Discrepancies are also explained using additional calculations. A comparison of the
overall system pressure drop, by comparing numerical and analytical solutions, and the local axial pressure distribution at subchannel level are presented. To make a one-to-one
comparison between CFD and subchannel solutions, volumeaveraging is applied to the CFD results according to the chosen subchannel nodalization.
The obtained results show a perfect agreement between the two codes. This outcome reflects the correct approach employed to build two consistent numerical models by properly carrying important information from the high-resolution models (CFD) to the low-resolution models (subchannel code). Furthermore, it has been found that the large discrepancies recorded in the CFD prediction of the grid pressure loss coefficients suggested in the benchmark specifications are mainly because the suggested benchmark values do not take into account the presence of the bounding channel that was
present in the experimental facility.
In this study, a numerical framework, comprising of a two-phase flow subchannel solver module and a Departure from Nucleate Boiling (DNB) evaluation module, was developed to mechanistically predict DNB in rod bundles of Pressurized Water Reactor (PWR). In this regard, the liquid sublayer dryout model was adapted as the Critical Heat Flux (CHF) triggering mechanism to reduce the dependency of the model on empirical correlations in the DNB evaluation module. To predict local flow boiling processes, a three-dimensional two-fluid formalism coupled with heat conduction was selected as the basic tool for the development of the two-phase flow subchannel analysis solver. Evaluation of the DNB modeling approach was performed against OECD/NRC NUPEC PWR Bundle tests (PSBT Benchmark) which supplied an extensive database for the development of truly mechanistic and consistent models for boiling transition and CHF. The results of the analyses demonstrated the need for additional assessment of the subcooled boiling model and the bulk condensation model implemented in the two-phase flow solver module. The proposed model slightly under-predicts the DNB power in comparison with the ones obtained from steady-state benchmark measurements. However, this prediction is acceptable compared with other codes. Another point about the DNB prediction model is that it has a conservative behavior. Examination of the axial and radial position of the first detected DNB using code-to-code comparisons on the basis of PSBT data indicated that the our simulation is reasonably accurate. DNB prediction in the power increase test case showed that the proposed model predicts the appearance of DNB before it happens; nevertheless, better agreement was obtained in comparison with the data from other codes.
