Volume 83, Issue 5

The reactor safety research of the Helmholtz Association is an integral part of the national provident research. It focuses on the safety of nuclear power plants in Germany and abroad as well as on safety aspects of internationally developed innovative reactor concepts. The research in the three involved Helmholtz centers Forschungszentrum Jülich, Helmholtz-Zentrum Dresden-Rossendorf and Karlsruhe Institute of Technology covers important areas of design basis and beyond design basis accidents. A unique combination of code and model development supported by own large-scale experiments ensures the active preservation of the reactor safety competences. The research that is embedded in a strong international co-operation will be continued after the completion of the national phase-out from the use of nuclear energy for electricity production in 2022.

Unauthorized dilution of boric acid in the reactor coolant due to improper operation of the chemical and volume control system (KBA) is a project mode with the violation of normal operation. The operator is obliged to identify unauthorized dilution within half an hour and stop feeding with pure condensate. The power unit has automation, which stops feeding on EP (emergency protection) and ATWS (Anticipated transient without scram) signals. The situation without the intervention of the operator and the failure of automation to stop dilution can be considered as a beyond design basis accident. The article presents an analysis of unauthorized boric acid dilution process in the primary circuit coolant of the NPP-2006 power unit up to the controlled state, either considering the operator's intervention after a 30-minute period of process, i. e. with a forced cessation of dilution, or without the operator's intervention until the tanks with pure condensate are completely emptied.

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.

Cooling down fuel rods is the critical technical challenge in nuclear reactors. In the past decades conventional fluids (like water, ethylene glycol, oil) were used for this purpose. Conventional heat transfer fluids extract less heat due to their lower thermal conductivity. Currently improving the heat transfer using advanced fluids – known as nanofluids – is investigated. A fluid which contains particles with sizes in nano-meters (known as nanoparticles) is called nanofluid. Solid metals have larger thermal conductivity when it is nano in size. The nanofluids which exhibit nanoparticles are usually made of oxides, metals, carbon nanotubes or carbides. Nanofluid is the new challenge in thermal sciences which generate from the concept of suspension of nanoparticles in base fluids for improving cooling phenomenon in nuclear reactors. Nanofluids have unique features or properties which are totally different from conventional solid-liquid mixtures that make them potentially useful for heat transfer enhancement in nuclear reactors. The research work on improvement in heat transfer using nanofluids is still in its primary stage. This paper presents a current review survey in this area with emphasis on the enhancement of heat transfer studies of nanofluids.

Many integrated pressurized water reactor (IPWR) designs using natural circulation operation mainly to enhance their inherent safety. The operating characteristics of primary coolant are completely different without mechanical pumps. The designs and safety analysis of forced circulation reactors are widely researched, but the natural circulation characteristics of IPWR have not been well studied by literatures. The present work discussed the thermal-hydraulic characteristics of IPWR under natural circulation conditions by using the best estimate codes RELAP5. And the effect of system parameters on natural circulation characteristics of IPWR is also studied. The results show that, the primary coolant average temperature and steam pressure are two key parameters that affect the natural circulation stable operating load. The set value of primary coolant average temperature effects the core outlet temperature and the steam temperature, but the primary coolant flow is basically the same under different primary coolant average temperature but same load conditions. The smaller steam pressure is more conducive to produce superheated steam, but there is risk of two phase flow instability in OTSG secondary side. The rapid load change process under natural circulation indicating that the reactor has a good load tracking characteristics under natural circulation, but the rapid change of primary coolant temperature will cause oscillations in system parameters.

Broad efforts have been ended in the last decade for the development of spacer and its effect on droplets that present in the vapour core. The spacers operate as droplet collectors and accumulate considerable amount of the droplet flux in an emblematic spacer section. This development paved the way for engineers and researchers to come up with spacer that exhibit better inter subchannel mixing in the reactor. The review is based on the correlation developed by different researchers for the spacer modelling, grey areas and challenges for further improvement in the prediction and some of the good models have been recommended. The effect of spacer on liquid films and droplet depositions before spacer and near spacer have been studied and finally based on this investigation a proposed LDV measurement techniques have been applied on 1/12 th segment of Advanced Heavy Water Reactor for further analysis of spacer effect on annular flow. Efforts are made to uncover the investigated gaps from the precedent literatures, as an effect of which it is concluded that there is extremely inadequate work published in the field of thermal-hydraulics incorporating spacer effect such as turbulent mixing, droplet deposition and liquid film thickness in the case of AHWR.