Bal Raj Sehgal

Film Boiling Heat Transfer on a High Temperature Sphere in Nanofluid

Quenching experiments of a high temperature sphere in Al2 O3 nanofluids are conducted to investigate the characteristics of film boiling and compared to those in pure water tests. One stainless steel sphere of 10 mm in diameter at the initial temperatures of 1000∼1400 K was tested in the nanofluids of the volume concentrations from 5 to 20% and the degrees of subcooling from 20 to 80 K. The test results show that film boiling heat fluxes and heat transfer rates in nanofluids were lower than those in pure water. The differences of the film boiling heat transfer rates between pure water and nanofluids become larger when the liquid subcooling decreases. Those results suggest that the presence of nanoparticles in liquid enhances vaporization process during the film boiling. The effects of nanoparticle concentrations of more than 5 vol. % on film boiling appear to be insignificant. However, the minimum heat fluxes tend to decrease when the concentration increases. Direct quenching without film boiling was repeatedly observed when an unwashed sphere was employed for quenching tests in nanofluids. It suggests that nanoparticle deposition on the sphere surface prevents the sphere from forming film around the sphere, which consequently promotes the rapid quenching of the hot sphere.Copyright

In-Vessel Melt Retention of Pressurized Water Reactors : Historical Review and Future Research Needs

A historical review of in-vessel melt retention (IVR) is given, which is a severe accident mitigation measure extensively applied in Generation III pressurized water reactors (PWRs). The idea of IV ...

Safest Roadmap for Corium Experimental Research in Europe

Severe accident facilities for European safety targets (SAFEST) is a European project networking the European experimental laboratories focused on the investigation of a nuclear power plant (NPP) s ...

Feasibility and Desirability of Employing the Thorium Fuel Cycle for Power Generation

Use of the thorium fuel cycle for nuclear power generation has been considered since the very start of the nuclear power era. In spite of a large amount of research, experimentation, pilot-scale and prototypic-scale installations, thorium fuel has not been adopted for large-scale power generation (Loewenstein and Sehgal in Trans Am Nuclear Soc, 27:312, 1977 [1]; Sehgal in The application of thorium in fast breeder reactors and in cross-progeny fuel cycles, NARA, Japan, 1982 [2]). This paper reviews the developments over the years on the front and the back ends of the thorium fuel cycle and describes the pros and cons of employing the thorium fuel cycle for large-scale generation of nuclear power. It examines the feasibility and desirability of employing the thorium fuel cycle in concert with the uranium fuel cycle for power generation.

Analysis of VVER-1000 large and small break LOCA experiments with RELAP5/MOD3.2

A RELAP5 model for the analysis of the PSB-VVER test facility was developed by the EREC in Russia. The PSB-VVER is a large-scale integral test facility to model the VVER-1000 type NPP. The volume and power scale in this test facility is 1:300 and the elevation scale is 1:1, which corresponds to the elevation mark of the reactor prototype. At the Division of Nuclear Safety, RIT, Sweden, we have modified the PSB-VVER facility’s RELAP5 model in order to analyze two of the transient tests performed on the PSB-VVER facility, which serve as the validation matrix described by NEA/CSNI. The objective of the work conducted was to validate the results obtained from RELAP5’s calculation with the supplied experimental data from the PSB-VVER test facility. Two accident scenarios have been calculated and analyzed. After being verified against the “11% UP LOCA” test data, the RELAP5/MOD3.2 model was used for a so-called “blind” transient calculation of the test “2×25% HL LOCA” and the results obtained were compared with the experimental data provided after the calculation.Copyright

VIPRE02 Code Assesment for CHF and Post-CHF Heat Transfer Modes in Long Tubes

The paper presents validation of the CORETRAN-01 code thermal–hydraulic module VIPRE02 against experimental data, obtained at experimental facilities at EREC E108 (Russia) and RIT (Sweden). The validation was carried out in order to assess the VIPRE02 code predictions for CHF occurrence and post–CHF heat transfer in 7.0 m long tubes. The lengths of the test sections in the experiments correspond to the heated length of the RBMK reactor fuel assembly. The CHF and post-CHF (transition and film boiling) correlations, included in the standard version of VIPRE02 code package were tested. The validation results as well as VIPRE model and code modification description are the subject of this paper.Copyright

Validation of HELIOS Neutron Cross-Section Library for RBMK Reactors Against the Data From the Critical Facility Experiments

The RBMK reactors are channel type, water-cooled and graphite moderated reactors. The first RBMK type electricity production reactor was put on-line in 1973. Currently there are 13 operating reactors of this type. Two of the RBMK-1500 reactors are at the Ignalina NPP in Lithuania. Experimental Critical Facility for RBMK reactors, located at Kurchiatov Institute, Moscow was designed to carry out critical reactivity experiments on assemblies, which imitate parts of the RBMK reactor core. The facility is composed of Control and Protection Rods (CPR’s), fuel assemblies with different enrichment in U-235 and other elements, typical for RBMK reactor core loadings, e.g. additional absorber assemblies, CPR imitators, etc. A simulation of a set of the experiments, performed at the Experimental Critical Facility, was carried out at the Royal Institute of Technology (RIT), Nuclear Power Safety Division, using CORETRAN 3-D neutron dynamics code. The neutron cross sections for assemblies were calculated using HELIOS code. The aim of this work was to evaluate capabilities of the HELIOS code to provide correct cross section data for the RBMK reactor. The calculation results were compared to the similar CORETRAN calculations, when employing WIMS-D4 code generated cross section data. For some of the experiments, where calculation results with CASMO-4 code generated cross sections are available, the comparison is also performed against CASMO-4 results. Eleven different experiments were simulated. Experiments differ in size of the facility core (number of assemblies loaded): from simple core loadings, composed only of a few fuel assemblies, to complicated configurations, which represent a part of the RBMK reactor core. Diverse types of measurements were carried out during these experiments: reactivity, neutron flux distributions (both axial and radial), rod reactivity worth and the voiding effects. Results of the reactivity measurements and relative neutron flux distributions were given in the Experiment report [1] as parameters, to be obtained using static calculations, i.e. the reported results were already processed numerically using the facility equipment, e.g. the reactimeter. The reported measurement errors consist only of instrumentation errors, i.e. measurement method errors and the influence from the space–time effects were not included in the error evaluation.Copyright