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Channel-Type Reactor with Coolant of Supercritical Parameters

#3213


Development of the Design Code Package to Analyze Thermohydraulic and Neutronic Characteristics of the Channel-Type Water-Cooled Water-Moderated Reactor with Supercritical Parameters of the Coolant

Tech Area / Field

  • FIR-ENG/Reactor Engineering and NPP/Fission Reactors

Status
8 Project completed

Registration date
14.03.2005

Completion date
09.04.2012

Senior Project Manager
Tocheny L V

Leading Institute
Federal State Unitary Enterprise Research and Development Institute of Power Engineering named after N.A.Dollezhal, Russia, Moscow

Collaborators

  • Atomic Energy Canada Limited, Canada, ON, Chalk River\nSchool of Energy Systems and Nuclear Science, University of Ontario Institute of Technology, Canada, Ontario

Project summary

The Project objective is to create a package of design codes accounting design features and specific characteristics as well as neutronic processes running in the channel-type reactor cooled by light water of supercritical parameters and moderated by heavy water (CT SC lW – hW R). The design code package under development will allow for design and test computation of stationary and transient neutronic and thermohydraulic processes running in the reactor cores of the above type to be performed. Consideration of design features and specific characteristics of thermohydraulic and neutronic processes being representative of the given type of the reactor will permit efficient computation analysis to be performed.

The principle objectives for creating CT SC lW - hWR are as follows:

  • substantial increase in economy, including efficiency, reduction in the terms of building and expences during construction;
  • enhancement of safety properties and performances;
  • preservation of advantages proved in time, which are distinctive of the channel-type reactor family.

The merits of the channel line in reactor building are:
  • modularity of the reactor design that eliminates restrictions with regard to unit power and does not require a large-size reactor pressure vessel;
  • fuel recharge can be performed while on-power that increases efficiency of the power use;
  • reduction of coolant leak intensity in case of leak tightness failure due to smaller scale of the core cooling circuit components.

The major advantages in application of the coolant with supercritical parameters are:
  • increase in efficiency that provides reduction in fuel consumption (reduction in prime cost of the fuel component) and heat releases to environment per installed unit power;
  • reduction in overall dimensions of thermal and mechanical equipment (reduction in the capital component of the prime cost);
  • reduction in intensity of deposit formation and corrosion impact upon fuel elements due to lack of phase transient (enhanced operating environment and higher reliability).

The international forum Generation-IV has set as its own task to offer to the world society by 2030 the power production systems suitable for commercialization. There have already been selected six systems approved to be most prospective. These are nuclear systems using fast and thermal neutrons, among them the water-cooled reactors having supercritical parameters.

It is worth noting the activities carried out by scientists and engineers in different countries aimed at creating a nuclear power source with supercritical parameters of the coolant, different approaches and various specific engineering parameters related to increased parameters of the coolant being proposed.

The interest in developing nuclear power sources with supercritical parameters of coolant and persity in approaches to and engineering solutions on such power sources indicate urgent character of the studies and optimization of both common approaches to creation of nuclear power plants with supercritical parameters and specific engineering solutions.

Since it is not real to get a direct experimental confirmation for effectiveness of any engineering solution, the only way out is to analyze the proposed engineering approaches and specific solutions with the help of computer codes. Thus, creations of a design code package with allowance for specific features of CT SC lW - hWR is an urgent task. The required set of such codes depends on engineering solutions under study and may involve the codes simulating thermohydraulic and neutronic phenomena and processes.

For the purpose of attaining the above Project objectives the following tasks are supposed to be solved:

  • develop the system of thermohydraulic completing relations with regard to thermal and physical properties, heat exchange, irreversible pressure losses and also heat and mass exchange between inpidual cells in a fuel assembly for the coolant flow of supercritical parameters on the basis of experimental data available and solution of 3D hydrodynamic task for a coolant flow in a process channel of the given reactor type;
  • develop the mathematical model and computer code for cell-by-cell computation of a fuel assembly of the given reactor type;
  • develop the mathematical model and computer code for combined neutronic and thermohydraulic computation of the given reactor core type;
  • carry out preliminary testing of the codes under development.

Development of a package of the design codes taking into account design features and specific characteristics of thermohydraulic and neutronic processes running in CT SC lW - hWR is supposed to be carried out on the basis of the NIKIET experience. This refers to creation and application of computer codes intended for selecting and substantiation of thermohydraulic and neutronic characteristics of the given reactor type, including channel-type water-cooled reactors. NIKIET has computer codes possible for use as prototypes of the codes to be newly developed.

The novelty of the computer codes to be created consists in that they will take into consideration the design features and specific characteristics of thermohydraulic and neutronic processes running in CT SC lW - hWR. The codes will observe the structure of core and fuel assemblies of the given reactor type. Thermal and physical properties of the water having supercritical parameters as well as the system of completing relations with regard to heat exchange, irreversible pressure losses and also heat and mass exchange between inpidual cells in a fuel assembly under supercritical parameters will be introduced into the codes.

Implementation of the Project will allow for the following concrete results to be obtained:

  • the package of design codes intended for selecting and substantiation of thermohydraulic and neutronic charactristics of CT SC lW - hWR will be created;
  • the code package will allow to perform the following computations:
    • local thermohydraulic parameters of fuel assemblies including temperature of structure components taking into consideration heat and mass exchange between inpidual cells in a fuel assembly (the computations will serve as a basis for selecting and substantiation of the core thermal and engineering reliability);
    • temperature distribution of fuel and other core components as well as thermohydraulic parameters of the coolant in channels of different power;
    • 3D fields of neutron flux densities and power density distribution with regard to both the reactor fuel assemblies and each fuel element taking into account feedbacks with respect to thermohydraulic parameters;
    • effective neutron multiplication factor;
    • reactivity effects and coefficients,
    • efficiency of CPS absorbing elements;
    • nuclear fuel recharge including replacement of fuel assemblies;
  • the above computations will ensure implementation of the design works on the new prospective CT SC lW - hWR;
  • the tasks on further experimental substantiation of such a reactor will be determined.

The given Project realizes to the full extent the ISTC objectives through providing 21 scientists and engineers connected to development of weapons technologies with the possibilities to:
  • realize their abilities in peaceful activities;
  • be actively integrated in solving international scientific and engineering problems;
  • further the development of applied studies in the area of nuclear safety and protection of environment.

The duration of the proposed Project is 36 months.

The estimated effort makes 749.05 men/months.

The leading institution, i.e. NIKIET, is one of the largest scientific and research centers of nuclear engineering and technology that performs large-scale scientific studies in nuclear power production. NIKIET has experience in cooperation with ISTC.

It is supposed that in the course of the Project development there will be used the NIKIET engineering approaches and methodology on:

  • designing, operation and decommissioning of nuclear installations computation studies of thermohydraulic and neutronic characteristics of nuclear reactors;
  • development and application of computer codes used for selecting and substantiation of thermohydraulic and neutronic characteristics of nuclear reactors;
  • development of completing relations based on computation as well as theoretical and experimental studies on decisive processes and phenomena.


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The International Science and Technology Center (ISTC) is an intergovernmental organization connecting scientists from Kazakhstan, Armenia, Tajikistan, Kyrgyzstan, and Georgia with their peers and research organizations in the EU, Japan, Republic of Korea, Norway and the United States.

 

ISTC facilitates international science projects and assists the global scientific and business community to source and engage with CIS and Georgian institutes that develop or possess an excellence of scientific know-how.

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