Search results for: PHWR

Authors: Bharat Kumar, Deepak Kumar, Vijay Chaudhry

Abstract:

Zirconium alloys are widely used for the core components of a pressurized heavy water reactor (PHWR) or Canada deuterium (CANDU) reactor due to their low neutron absorption cross-section and excellent mechanical properties. The components made of Zirconium alloys are subjected to flow-induced vibrations, resulting in fretting wear at the interface of; pressure tubes and bearing pads, pressure tubes and calandria tubes, and calandria tubes and Liquid injection shutdown system (LISS) nozzles. There is a need to explore the tribological response under such conditions. Present work simulates the contact between calandria tube and LISS nozzle of PHWR/CANDU reactor as cylinder-on-cylinder contact configuration. Reciprocating tribo-tests were conducted on Zircaloy-4 (Zr-4) under the self-mated condition at varying amplitude, frequency, and sliding time. To understand the active wear mechanism, worn surfaces were analyzed using Scanning Electron Microscopy (SEM) and Energy Dispersive Spectroscopy (EDS). The change in amplitude severely affects the wear than other factors. The wear mechanism transits from adhesion to abrasion with increasing test amplitude. The dominant wear mechanisms are micro-cutting and micro-plowing followed by delamination in some areas. However, the coefficient of friction has indifferent behaviors.

Keywords: zircaloy-4, tribology, calandria tube, LISS nozzle, PHWR

Procedia PDF Downloads 170

Authors: D. Ramajo, S. Corzo, M. Nigro

Abstract:

A Multi-dimensional computational fluid dynamics (CFD) two-phase model was developed with the aim to simulate the in-core coolant circuit of a pressurized heavy water reactor (PHWR) of a commercial nuclear power plant (NPP). Due to the fact that this PHWR is a Reactor Pressure Vessel type (RPV), three-dimensional (3D) detailed modelling of the large reservoirs of the RPV (the upper and lower plenums and the downcomer) were coupled with an in-house finite volume one-dimensional (1D) code in order to model the 451 coolant channels housing the nuclear fuel. Regarding the 1D code, suitable empirical correlations for taking into account the in-channel distributed (friction losses) and concentrated (spacer grids, inlet and outlet throttles) pressure losses were used. A local power distribution at each one of the coolant channels was also taken into account. The heat transfer between the coolant and the surrounding moderator was accurately calculated using a two-dimensional theoretical model. The implementation of subcooled boiling and condensation models in the 1D code along with the use of functions for representing the thermal and dynamic properties of the coolant and moderator (heavy water) allow to have estimations of the in-core steam generation under nominal flow conditions for a generic fission power distribution. The in-core mass flow distribution results for steady state nominal conditions are in agreement with the expected from design, thus getting a first assessment of the coupled 1/3D model. Results for nominal condition were compared with those obtained with a previous 1/3D single-phase model getting more realistic temperature patterns, also allowing visualize low values of void fraction inside the upper plenum. It must be mentioned that the current results were obtained by imposing prescribed fission power functions from literature. Therefore, results are showed with the aim of point out the potentiality of the developed model.

Keywords: PHWR, CFD, thermo-hydraulic, two-phase flow

Procedia PDF Downloads 438

Authors: Muhammad Minhaj Khan, Jaemin Lee, Suhong Lee, Jinyoung Chung, Johoo Whang

Abstract:

The Republic of Korea has been facing strong storage crisis from nuclear waste generation as At Reactor (AR) temporary storage sites are about to reach saturation. Since the country is densely populated with a rate of 491.78 persons per square kilometer, Construction of High-level waste repository will not be a feasible option. In order to tackle the storage waste generation problem which is increasing at a rate of 350 tHM/Yr. and 380 tHM/Yr. in case of 20 PWRs and 4 PHWRs respectively, the study strongly focuses on the advancement of current nuclear power plants to GEN-IV sustainable and ecological nuclear systems by burning TRUs (Pu, MAs). First, Calculations has made to estimate the generation of SNF including Pu and MA from PWR and PHWR NPPS by using the IAEA code Nuclear Fuel Cycle Simulation System (NFCSS) for the period of 2016, 2030 (including the saturation period of each site from 2024~2028), 2089 and 2109 as the number of NPPS will increase due to high import cost of non-nuclear energy sources. 2ndly, in order to produce environmentally sustainable nuclear energy systems, 4 scenarios to burnout the Plutonium and MAs are analyzed with the concentration on burning of MA only, MA and Pu together by utilizing SFR, LFR and KALIMER-600 burner reactor after recycling the spent oxide fuel from PWR through pyro processing technology developed by Korea Atomic Energy Research Institute (KAERI) which shows promising and sustainable future benefits by minimizing the HLW generation with regard to waste amount, decay heat, and activity. Finally, With the concentration on front and back end fuel cycles for open and closed fuel cycles of PWR and Pyro-SFR respectively, an overall assessment has been made which evaluates the quantitative as well as economical combativeness of SFR metallic fuel against PWR once through nuclear fuel cycle.

Keywords: GEN IV nuclear fuel cycle, nuclear waste, waste sustainability, transmutation

Procedia PDF Downloads 322