DYNAMIC PARAMETER FITTING OF THE ZED-2 REACTOR CORE USING A NEUTRON FLUX PERTURBATION EXPERIMENT

The energy in nuclear power plants is produced by thermal fission. It is extremely important to be able to monitor the processes in the reactor to ensure the safety and reliability of the power plant. One of the main traits of the reactor core is neutron flux. It changes in time and space therefore it is crucial to be able to simulate its changes with computer codes. In the research work a program code was established in the Matlab software with which the neutron flux of a one-dimensional zone can be simulated with homogenous and heterogenic zone parameters as well. The code is written using the one-group one-dimensional time- and space-dependent diffusion equation. The equation of an average delayed neutron group and xenon and iodine distributions was also included in the system to give a more precise look on the problem. The main innovation in the code is that numerical methods were used to solve the problem: the finite difference approach was applied for the place-dependent and for the time-dependent solution. The advantage of this code compared to other ones is that one-dimensional zones can be simulated in a really short time and it still gives a precise solution because of the complex numerical methods used.

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Effects of various types of reflectors on the 99Mo production in the VVER-Ts reactor targets

Nuclear Energy and Technology ◽

10.3897/nucet.6.54623 ◽

2020 ◽

Vol 6 (2) ◽

pp. 89-92

Author(s):

Oleg Yu. Kochnov ◽

Pavel A. Danilov

Keyword(s):

Neutron Flux ◽

Reactor Core ◽

Raw Material ◽

The Core ◽

Assessment Tasks ◽

Materials Used ◽

Core Characteristics ◽

Water Space ◽

Reactivity Margin ◽

Effective Multiplication

The effects from introducing various types of reflectors in the VVR-Ts reactor core on the 99Мо production were analyzed. Earlier the effects of only the beryllium reflector on the VVR-Ts reactor core characteristics, such as reactivity margin, neutron flux in experimental channels, and activity of the accumulated 99Мо, were calculated. The calculations are based on a generated precision model of the core which comprises one experimental channel where targets are irradiated for the 99Мо production. The model was built using the SCALE code. The code allows a fairly broad range of calculations to be performed, from criticality estimation to radiological assessment tasks. As the result of the computational analysis of the model, such characteristics were obtained as the effective multiplication factor, the power density in the 99Мо production targets, the neutron flux in the target raw material, and the quantity of the produced 99Мо after 120 hours of irradiation. The data was compared with the results of similar calculations of the VVR-Ts reactor core parameters. Further, the list of the materials used extensively as the reactor core reflector or moderator was formed based on reference literature. A number of models were obtained and analyzed on its basis, in which the water space on the core periphery was substituted for the investigated materials.

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Advanced Test Reactor: A National Scientific User Facility

Volume 4: Structural Integrity; Next Generation Systems; Safety and Security; Low Level Waste Management and Decommissioning; Near Term Deployment: Plant Designs, Licensing, Construction, Workforce and Public Acceptance ◽

10.1115/icone16-48426 ◽

2008 ◽

Author(s):

Clifford J. Stanley ◽

Frances M. Marshall

Keyword(s):

Neutron Flux ◽

National Laboratory ◽

Reactor Core ◽

Nuclear Research ◽

Reactor Power ◽

Department Of Energy ◽

Life Extension ◽

Test Environment ◽

User Facility ◽

Test Reactor

This presentation and associated paper provides an overview of the research and irradiation capabilities of the Advanced Test Reactor (ATR) located at the U.S. Department of Energy Idaho National Laboratory (INL). The ATR which has been designated by DOE as a National Scientific User Facility (NSUF) is operated by Battelle Energy Alliance, LLC. This paper will describe the ATR and discuss the research opportunities for university (faculty and students) and industry researchers to use this unique facility for nuclear fuels and materials experiments in support of advanced reactor development and life extension issues for currently operating nuclear reactors. The ATR is a pressurized, light-water moderated and cooled, beryllium-reflected nuclear research reactor with a maximum operating power of 250 MWth. The unique serpentine configuration (Fig. 1) of the fuel elements creates five main reactor power lobes (regions) and nine flux traps. In addition to these nine flux traps there are 68 additional irradiation positions in the reactor core reflector tank. There are also 34 low-flux irradiation positions in the irradiation tanks outside the core reflector tank. The ATR is designed to provide a test environment for the evaluation of the effects of intense radiation (neutron and gamma). Due to the unique serpentine core design each of the five lobes can be operated at different powers and controlled independently. Options exist for the individual test trains and assemblies to be either cooled by the ATR coolant (i.e., exposed to ATR coolant flow rates, pressures, temperatures, and neutron flux) or to be installed in their own independent test loops where such parameters as temperature, pressure, flow rate, neutron flux, and chemistry can be controlled per experimenter specifications. The full-power maximum thermal neutron flux is ∼1.0 x1015 n/cm2-sec with a maximum fast flux of ∼5.0 x1014 n/cm2-sec. The Advanced Test Reactor, now a National Scientific User Facility, is a versatile tool in which a variety of nuclear reactor, nuclear physics, reactor fuel, and structural material irradiation experiments can be conducted. The cumulative effects of years of irradiation in a normal power reactor can be duplicated in a few weeks or months in the ATR due to its unique design, power density, and operating flexibility.

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