• Energy Research
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Several TH-TH code coupling methods are used in the nuclear industry to model the core thermal–hydraulic conditions and the system behaviour. In some cases, the boundary conditions obtained by system codes are applied to sub-channel codes by table (off-line coupling). Even though this approach is in general considered valid, some boundary parameters will present inconsistencies. Alternatively, system and sub-channel codes are coupled using different coupling methods (semi-implicit coupling). Recent studies have shown a strong influence of the bound- ary conditions uncertainty on the sub-channel code results. The present study aims to evaluate the differences produced by the coupling methods by performing a best-estimate plus uncertainties (BEPU) comparison to the following cases: a complete loss of forced flow and a pressurizer relief valve opening. Results show that BEPU analysis presents good agreement with some discrepancies that can be explained and correlated to the boundary conditions deviations between codes. Peer Reviewed

The containment is the last barrier to prevent the release of radioactive materials into the environment in case of a severe reactor accident. Thus, it is important to understand the flow and transport processes of combustible gases and to develop strategies to maintain the integrity of the containment. A holistic OpenFOAM-based CFD software package, containmentFOAM, is developed for containment analysis. During the validation of the code with regard to free convection flows and mixing processes, it became apparent that the simulation results are fairly sensitive to the definition of initial and boundary conditions, which are often not precisely known from the experiment. Therefore, establishing the validity of the models is impossible without a quantification of uncertainties.

The cross section averaged over 235U thermal-neutron induced fission spectrum is a fundamental quantity that can be used in evaluation and validation of nuclear data. Many experiments focused on the determination of Spectrum Averaged Cross Sections (SACS) in 235U(nth,f). Prompt Fission Neutron Spectrum (PFNS) in light water reactors using enriched uranium fuel. In these reactors, there is already some amount of water moderator between the uranium fuel and the irradiated sample. Due to the decrease of hydrogen cross-section with neutron energy, the high energy tail of the reactor spectrum in cores with water moderator may be harder than the pure PFNS. This paper aims to compare the shape of the actual reactor spectrum in various core positions of a research light-water reactor differing each from other by the effective water thickness. The spectrum shape is determined both by calculations and experimentally using various high energy threshold reactions. The impact of the photo-nuclear reactions ({\gamma},n) competing with (n,2n) in production of the same residual nucleus was shown to be less than a percent for most of studied dosimeters. An important exception was found for 197Au(n,2n) 196Au dosimeter irradiated in the outcore channel where a notable photo-neutron contribution to the production of 196Au is caused by the neutron production from the high energy {\gamma}-rays from thermal-neutron capture in 54Fe. The corresponding ENDF/B-VIII.0 data turned out to underestimate such {\gamma}-yield by 40% in comparison with ENDF/B-VI.8. This has improved but however not resolved the disagreement between our measurement and calculations. The remaining deficiency was attributed to the underestimation of the evaluated cross section IAEA/PD-2019 for the 197Au({\gamma},n) cross section near the reaction threshold. The later was confirmed by comparison with existing measured data.

Abstract In this paper, we assessed the interchannel mixing models currently available in MATRA (Yoo et al., 1999), a subchannel analysis code modified based on COBRA-IV-I (Stewart et al., 1977), for application to two-phase flow conditions of both BWR and PWR fuel rod bundles. By recalculating selected test cases from the NUPEC BFBT benchmark (Neykov et al., 2006) and the ISPRA benchmark (Herkenrath et al., 1981), prediction accuracy of the equal-volume-exchange of turbulent mixing and void drift (EVVD) model implemented in MATRA was found to be not satisfied for application in the dispersed bubbly flow regime under PWR pressure level. Since the EVVD model is adopted in most state-of-the-art subchannel analysis codes in similar manners as in MATRA, we believe that an improvement of the EVVD model should be carried out in the future. Finally, based on results obtained in the current study, we discussed the drawbacks of the EVVD model and pointed out possible aspects that should be considered for improvement.

This paper proposes a new benchmark for VVER-1000 control rod ejection transient. The benchmark is designed for code-to-code comparison and its purpose is testing and verifications of time-dependent solvers, core models and cross section generation methodology. Four solutions obtained using different combinations of three lattice transport, two nodal neutronics, and two thermo-hydraulic codes are analyzed, demonstrating good agreement in predicting transient reactivity and power peaks and highlighting the impact of gas gap modeling and DNBR correlation on simulation results.

Weight window, a variance reduction tool, is often used to improve the performance of radiation shielding calculations. One common issue with the weight window is determining the optimal set of weight window parameters for solving deep penetration shielding problems. To address this issue, the recursive Monte-Carlo methodology has been used with the in-house TRAWEI code. The program is responsible for generating the optimal weight parameters in a single run with minimum computational time. This paper presents the results of a numerical test conducted using a simple reactor model to evaluate the performance of the developed weight generator program. The findings reveal that MCNP simulations utilizing TRAWEI-generated weight values exhibit significantly higher calculation efficiency compared to both analog simulation and MCNP simulation using weights generated by the existing MCNP weight window generator. Overall, the utilization of the RMC methodology has shown its potential to significantly contribute to deep penetration shielding calculations.