scholarly journals Spatially-dependent nuclear reactor kinetic calculations with the explicit fission product model

2019 ◽  
Vol 133 ◽  
pp. 202-208
Author(s):  
Koji Katagiri ◽  
Go Chiba
1996 ◽  
Vol 23 (6) ◽  
pp. 517-532 ◽  
Author(s):  
B. Montagnini ◽  
P. Raffaelli ◽  
M. Sumini ◽  
D.M. Zardini

1982 ◽  
Vol 19 (2) ◽  
pp. 96-106 ◽  
Author(s):  
Shungo IIJIMA ◽  
Tadashi YOSHIDA ◽  
Tohru YAMAMOTO

2014 ◽  
Vol 3 (2) ◽  
pp. 83-90 ◽  
Author(s):  
M. Seydaliev ◽  
D. Caswell

There is a growing international interest in using coupled, multidisciplinary computer simulations for a variety of purposes, including nuclear reactor safety analysis. Reactor behaviour can be modeled using a suite of computer programs simulating phenomena or predicting parameters that can be categorized into disciplines such as Thermalhydraulics, Neutronics, Fuel, Fuel Channels, Fission Product Release and Transport, Containment and Atmospheric Dispersion, and Severe Accident Analysis. Traditionally, simulations used for safety analysis individually addressed only the behaviour within a single discipline, based upon static input data from other simulation programs. The limitation of using a suite of stand-alone simulations is that phenomenological interdependencies or temporal feedback between the parameters calculated within individual simulations cannot be adequately captured. To remove this shortcoming, multiple computer simulations for different disciplines must exchange data during runtime to address these interdependencies. This article describes the concept of a new framework, which we refer to as the “Backbone,” to provide the necessary runtime exchange of data. The Backbone, currently under development at AECL for a preliminary feasibility study, is a hybrid design using features taken from the Common Object Request Broker Architecture (CORBA), a standard defined by the Object Management Group, and the Message Passing Interface (MPI), a standard developed by a group of researchers from academia and industry. Both have well-tested and efficient implementations, including some that are freely available under the GNU public licenses. The CORBA component enables individual programs written in different languages and running on different platforms within a network to exchange data with each other, thus behaving like a single application. MPI provides the process-to-process intercommunication between these programs. This paper outlines the different CORBA and MPI configurations examined to date, as well as the preliminary configuration selected for coupling 2 existing safety analysis programs used for modeling thermal–mechanical fuel behavior and fission product behavior respectively. In addition, preliminary work in hosting both the Backbone and the associated safety analysis programs in a cluster environment are discussed.


1998 ◽  
Vol 35 (8) ◽  
pp. 527-537 ◽  
Author(s):  
Tadashi IKEHARA ◽  
Yoshihira ANDO ◽  
Munenari YAMAMOTO

2015 ◽  
Author(s):  
◽  
Lukas Michael Carter

High-temperature gas-cooled reactors (HTGRs) are one of the candidates being considered for the replacement of current nuclear reactor designs. Diffusion coefficients for fission products in HTGR graphite are required for estimation of fission product release rates from such reactors. We developed a method for analysis of fission product of fission product surrogate release rates from heated graphite samples. The graphite samples were infused with fission product surrogate material, and material which diffused from the graphite samples was transported via a carbon aerosol laden He jet system to an online inductively coupled plasma mass spectrometer for quantification of the release rate. Diffusion coefficients for cesium in IG-110 and NBG-18 grade nuclear graphites are reported.


2014 ◽  
Vol 3 (01) ◽  
pp. 29-36 ◽  
Author(s):  
Steve Livingstone

Fission product concentration in reactor primary heat transport systems is a common diagnostic indicator for assessing reactor core condition and determining the presence, size, power, location, residence time, burnup, etc., of defected fuel. Typically, diagnostic assessment assumes a priori that measured data (activity concentration measurements and reactor parameters) are accurate; however, this is not always a valid assumption. A set of novel methods has been developed for detecting minor discrepancies in fission product concentration measurements and reactor parameters (such as issues with transit times, purification, and spectral analysis). A variety of techniques are discussed and applied to a variety of reactor types (mainly commercial power plant designs); these techniques and concepts can be modified and applied for research and (or) commercial applications.


1968 ◽  
Vol 32 (3) ◽  
pp. 417-420 ◽  
Author(s):  
J. L. Russell ◽  
W. Rotter

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