This paper presents the development of a three-dimensional space-time neutronic kinetic model of a Canadian deuterium uranium (CANDU) reactor using a modal method. In this method, the reactor space-time neutron flux is synthesized by a time-weighted series of precalculated neutron flux modes. The modes are eigenfunctions of the governing neutron diffusion equation during reference steady-state operation. The xenon effect has also been considered. The reactor model is then implemented within a simulation platform of a CANDU6 reactor regulating system in MATLAB/SIMULINK. A nondimensionalized SIMULINK representation of the reactor kinetic model is established. The behavior of the reactor during load following transients has been simulated using the developed reactor-modeling module. The simulation results prove the efficiency of the model. A three-dimensional neutron flux distribution during transients is represented.

1.
Frogner
,
B.
, and
Rao
,
H. S.
, 1978, “
Control of Nuclear Power Plants
,”
IEEE Trans. Autom. Control
0018-9286,
23
(
3
), pp.
405
417
.
2.
Cherchas
,
D. B.
, and
Mewdell
,
C. G.
, 1978, “
A Control Algorithm for Reactor Spatial Control During Nuclear Station Load Cycling
,”
ASME J. Dyn. Syst., Meas., Control
0022-0434,
100
, pp.
219
226
.
3.
Cherchas
,
D. B.
, and
Ng
,
S. S.
, 1978, “
Optimum Control of Neutron Flux During Nuclear Station Load Following
,”
Automatica
0005-1098,
14
, pp.
533
546
.
4.
Yorke
,
G. L.
, and
Cherchas
,
D. B.
, 1981, “
An Algorithm for Non-Linear Space-Time Nuclear Reactor Control
,”
Automatica
0005-1098,
17
, pp.
471
482
.
5.
Rouben
,
B.
, 2002, “
RFSP-IST, The Industry Standard Tool Computer Program for CANDU Reactor Core Design and Analysis
,”
Proceedings of the 13th Pacific Basin Nuclear Conference
, Shenzhen, China, p.
75
.
6.
Javidnia
,
H.
, and
Jiang
,
J.
, 2009, “
Modeling and Simulation of a CANDU Reactor for Control System Design and Analysis
,”
Nucl. Technol.
0029-5450,
165
, pp.
190
199
.
7.
Tiwari
,
A. P.
, 1999, “
Modeling and Control of a Large Pressurized Heavy Water Reactor
,” Ph.D. thesis, Indian Institute of Technology, Bombay, India.
8.
Luxat
,
J. C.
, and
Frescura
,
G. M.
, 1979, “
Space-Time Neutronic Analysis of Postulated Loss-of-Coolant Accidents in CANDU Reactors
,”
Nucl. Technol.
0029-5450,
46
, pp.
507
516
.
9.
Gold
,
M.
,
Wight
,
A. L.
, and
Frescura
,
G. M.
,1990, “
SMOKIN—A Family of Codes for Reactor Space-Time Neutronics Calculations Based on Modal Kinetics—Theory Manual
,” Report No. 90133.
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