| Full-Core VVER-440 extended calculation benchmark |
3 |
| Development of CASMO5 for VVER-1000/1200 analysis and preliminary validation using critical experiments |
2 |
| SIMULATE5-HEX extension for VVER analyses |
2 |
| Investigations on potential methods for the long-term monitoring of the state of fuel elements in dry storage casks |
2 |
| Computation of gamma radioactivity of natural rocks in the vicinity of Antalya province and its effect on health |
2 |
| Natural radioactive risk assessment in top soil and possible health effect in Minim and Martap villages, Cameroon: using radioactive risk index and statistical analysis |
2 |
| Validation of the AC(2 )Codes ATHLET and ATHLET-CD |
2 |
| Validation and Application of the AC(2) Code COCOSYS |
1 |
| ATHLET extensions for the simulation of supercritical carbon dioxide driven power cycles |
1 |
| Thermal-hydraulic insights during a main steam line break in a generic PWR KONVOI reactor with ATHLET 3.1A |
1 |
| ATWS severe accident analysis in the loss of flow scenario using the MELCOR code in Bushehr nuclear Power Plant |
1 |
| Assessment of void fraction predictability of system codes in subchannels |
1 |
| Review on using nanofluids for heat transfer enhancement in nuclear power plants |
1 |
| Analysis of the impact of different scenarios on the simulation results of unauthorized dilution of boric acid in the coolant of the primary circuit of the NPP-2006 |
1 |
| New research reactor protection system |
1 |
| Application of discontinuity factors and group constants generated by SERPENT in the KIKO3 DMG code |
1 |
| Investigations of the hydrogen diffusion and distribution in Zirconium by means of Neutron Imaging |
1 |
| Investigation of the excitation functions for the (n, 2n) reactions on the structural fusion material Ni-Ni-60-Ni-64 |
1 |
| Mathematical model for prediction of droplet sizes and distribution associated with impact of liquid-containing projectile |
1 |
| Abnormal control rod withdrawal analysis for innovative research reactor using PARET-ANL codes |
1 |
| Core loading optimisation in Slovak VVER-440 reactors |
1 |
| Applied study on optimizing the final disposal of Loviisa NPP spent fuel assemblies |
1 |
| Revision of a CHF correlation for PWR under low pressure conditions with only dimensionless parameters as independent variables |
1 |
| Criteria and comparison of thermal stratification between PRHR HX heating and ADS spraying process in IRWST based on a down-scaled experimental facility |
1 |
| Proposal of a novel CHF correlation for PWR under low pressure conditions based on stepwise regression method |
1 |
| Mechanistic prediction of post dryout heat transfer and rewetting |
1 |
| Calculations of spent fuel isotopic composition for fuel rod from VVER-440 fuel assembly benchmark using several evaluated nuclear data libraries |
1 |
| Considerations on spent fuel behavior for transport after extended storage |
1 |
| Physical startup tests calculations for Dukovany NPP using MOBY-DICK macrocode |
0 |
| Renewing the refueling neutron monitoring and reactivity measurement systems at Paks NPP |
0 |
| Contribution to the validation of the VVER-1000 Temelin NPP computing model for the ATHLET/DYN3D coupled codes |
0 |
| Hot channel calculation methodologies in case of VVER-1000/1200 reactors |
0 |
| Investigation of fuel cycles containing Generation IV reactors and VVER-1200 reactors |
0 |
| Study of neutron-physical characteristics of VVER-1200 considering feedbacks using MCU Monte Carlo code |
0 |
| Advantages of VVER-440 fuel cycles with new fuel assemblies |
0 |
| A neutronics feasibility study on utilization of a thinned cladding fuel design at Loviisa NPP |
0 |
| CIEMAT response to challenges on fuel safety research during dry storage |
0 |
| Open questions on the road to reliable predictions of cladding integrity |
0 |
| Research activities at GRS on fuel rod behaviour during extended dry storage |
0 |
| Power transient calculations with VERONA |
0 |
| Simulation of standard temperature control indications at the outlet of a fuel assembly of VVER1000 reactor of Rostov NPP unit No. 2 |
0 |
| Extension of nodal diffusion solver of Ants to hexagonal geometry |
0 |
| C-PORCA 7: a nodal diffusion reactor calculation code to support off-line and on-line core analysis at Paks nuclear power plant |
0 |
| Adaptation of the gas gap simplified model in DYN3D code to new types of fuel |
0 |
| A procedure for verification of Studsvik's spent nuclear fuel code SNF |
0 |
| Thermal hydraulic characteristics of two-phase natural circulation for secondary side passive residual heat removal system |
0 |
| TRACE simulations of IIST cooldown experiments |
0 |
| Flow and heat transfer characteristics in rough micro-channels |
0 |
| Benchmark study of OPAL research reactor using MCNP codes |
0 |
| CFD code-to-code benchmarking on simulation of fire and smoke propagation for nuclear applications |
0 |
