The discussion revolves around the Supercritical Water Reactor (SCWR), a Generation IV reactor technology. Participants explore its design, operational characteristics, and the neutronic calculations associated with it, including the codes used for modeling and analysis.
Participants do not reach a consensus on the specifics of neutronic calculations or the applicability of existing codes, indicating multiple competing views and ongoing exploration of the topic.
Limitations include the dependence on specific design parameters and assumptions about the behavior of supercritical water under varying conditions, as well as unresolved details regarding the modeling challenges posed by the reactor's operational characteristics.
Putting this in perspective, PWRs operate with pressure ~15.5-15.7 MPa and Tout ~310-330°C (590-626°F). BWR cores operate under saturated conditions at ~7.2 MPa and ~286°C (547°F). The exit temperature of a BWR is approximately (or slightly less than) the inlet temperature of a PWR.The Supercritical-Water-Cooled Reactor (SCWR) system is a high-temperature, high-pressure water-cooled reactor that operates above the thermodynamic critical point of water (374 degrees Celsius, 22.1 MPa, or 705 degrees Fahrenheit, 3208 psia).
The supercritical water coolant enables a thermal efficiency about one-third higher than current light-water reactors, as well as simplification in the balance of plant. The balance of plant is considerably simplified because the coolant does not change phase in the reactor and is directly coupled to the energy conversion equipment. The reference system is 1,700 MWe with an operating pressure of 25 MPa, and a reactor outlet temperature of 510 degrees Celsius, possibly ranging up to 550 degrees Celsius. The fuel is uranium oxide. Passive safety features are incorporated similar to those of simplified boiling water reactors.
chivasorn said:can you explain about neutronic calculations and the used codes for SCWR?
Whether the existing codes can be used for this type of reactor is?
Abstract
The HPLWR (High-Performance Light Water Reactor) is the European version of the SCWR (Supercritical-pressure Water Cooled Reactor) concept, which is one of the Generation IV reactors. In this reactor the primary water enters the core of the HPLWR under supercritical-pressure condition (25 MPa) at a temperature of 280 °C and leaves it at a temperature of up to 510 °C. Due to the significant changes in the physical properties of water at supercritical-pressure the system is susceptible to local temperature, density and power oscillations. This inclination is increased by the pseudocritical transformation of the supercritical-pressure water used as primary coolant.
At the Budapest University of Technology a coupled neutronics–thermohydraulics program system has been developed which is capable of determining the steady-state equilibrium parameters and calculating the related power, temperature, etc. distributions for the above-mentioned reactor. The program system can be used to optimize the axial enrichment profile of the fuel rods (e.g. in order to obtain a uniform power distribution).
Since the power distribution will change with burn-up, which causes a change in the temperature and density distributions, a coupled neutronics – thermohydraulics – burn-up calculation is required. Therefore, the program system has been extended with the ORIGEN-S burn-up calculator.
In the paper the developed program system and its features are presented. Parametric studies for different operating conditions were carried out and the obtained results are discussed.
A Korean group has used Helios.Abstract
A state-of-the-art study was performed to investigate the operational conditions for in-core and out-of-core materials in a high performance light water reactor (HPLWR) and to evaluate the potential of existing structural materials for application in fuel elements, core structures and out-of-core components. In the conventional parts of a HPLWR-plant the approved materials of supercritical fossil power plants (SCFPP) can be used for given temperatures (600 °C) and pressures (˜250 bar). These are either commercial ferritic/martensitic or austenitic stainless steels. Taking the conditions of existing light water reactors (LWR) into account an assessment of potential cladding materials was made, based on existing creep-rupture data, an extensive analysis of the corrosion in conventional steam power plants and available information on material behaviour under irradiation. As a major result it is shown that for an assumed maximum temperature of 650 °C not only Ni-alloys, but also austenitic stainless steels can be used as cladding materials.
INEEL Neutronic Analyses. We have used MOCUP, a combination of MCNP4B and Origen2, to model the consumption of fissile material and the buildup of fission products and actinides during irradiation in a prototypical thermal spectrum SCWR with solid moderator material. The model consists of 40 axial fuel zones along the 427 cm rod height. The model includes the axial variation in the coolant density and has a two-zone variation in the fuel enrichment, 4.0 wt % 235U in the lower half and 4.2 wt % in the upper half of the rod.