Pergamon Progress in Nuclear Energy, Vol. 36. No. 4. PP. 379-385. 2000

0 2000 Elsevier Sctence Ltd All rights reserved. Printed in Great Britain

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PII: SO149-1970(00)00086-X

DYNAMIC SIMULATOR FOR THE MINIATURE NEUTRON SOURCE REACTOR

I. Khamis, M. B. Alsous, H. Haj Hassan, H. Jouhara

Department of Physics, Atomic Energy Commission P.O. Box 6091, Damascus, Syria

ABSRACT A dynamic simulator for the Syrian Miniature Neutron Source Reactor was developed and implemented on a desktop computer using C++ Builder. Mathematical models for the main physical phenomena of reactor such as heat transfer and neutronics were developed on the basis of the lumped parameter approach and real experimental data fitting. Point model equations of reactor kinetics was employed and solved using fourth order Runge-Kutta integration procedure. Simulation for training purposes of both real and accelerated time for normal and abnormal conditions can be accomplished with the model. The simulator is user friendly with operator. 0 2000 Elsevier Science Ltd. All rights reserved.

INTRODUCTION The Miniature Neutron Source Reactor (MNSR), similar to the Canadian SLOWPOKE, is a low temperature pool-type research reactor developed and manufactured by the Chinese Institute of Atomic Energy’. It is intended primarily as a neutron irradiation facility for training and neutron activation analysis. The core design embodies highly enriched uranium (89.97 percentage by weight U-235), light water moderator and beryllium reflectors. There are five inner and five outer irradiation sites in and around the annular beryllium reflector. The nominal thermal neutron flux, in any one of the inner irradiation sites, is 10” n. cme2 .s“ at 30 kW power. The core, cooled by natural convection, is located inside and near the bottom of a water-filled cylindrical container which is suspended in a large pool of water. The MNSR, an inherently safe reactor, has limited maximum installed excess reactivity (less than 4 mk), a highly negative moderator temperature coefficient of reactivity (approximately 0.1 rnk/“C at temperature range of 20-45”C), and a low critical mass (less than 1 kg of U-235). These characteristics limit achievable peak power levels following an accidental insertion of reactivity and assure the safety of the MNSR reactor under all conceivable accident conditions.

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MODEL OVERVIEW

I. Khamis et al.

For simulation purposes, the reactor system was divided into two models: neutron kinetics and thermal-hydraulics. In the first one, a ^

main point

kinetics model’- representing neutronics uses six groups of delayed neutrons and one group of photo-neutrons3. Nuclear fission process dynamics are governed by the excess reactivity in the core as well as the decay of delayed and photo neutrons. The neutron kinetics equations are

dm P-P -=-,+C~AiCi i = 1...7 dt I

dci -Pin+. 27-1 I1

i=1..6

where n = Neutron density p = Reactivity load ,8 = Total fraction of delayed and photo neutrons I = Neutron generation time Ai = Decay constant for the i-th precursor Ci = i-th group of delayed neutron i.e. precursor fi = the i-th fraction of delayed neutron

and group 7 represents the photoneutron source due to the presence of the beryllium reflector. All the kinetics equations are solved numerically using a readily-available fourth order Runge-Kutta scheme4. The thermal-hydraulic model consists of simple balance equations where heat-transfer phenomena take place. Heat generated in the fuel by nuclear fission is transferred to the coolant. Formulation of temperature equations for the incoming, outgoing, and in-core coolant is made using some real data collected during operation of the reactor. In addition, some empirical relationships have been devised for simulation purposes of the reactor. For example, the model incorporates a correlation that relates core temperature drop across the core to the operating power and coolant inlet temperature ‘, that is:

Dynamic simulator for the MNSR 381

where AT = temperature difference across the core [“Cl Hi,1 = hight of the coolant inlet nozzle to the core [6.0 mm] T,, = coolant temperature at the inlet of the core [“Cl P = nominal power [kW]

Moreover, the core excess reactivity is correlated to control rod worth as function of position using the following formula:

where per = control rod worth of reactivity [rnk]” x = position of control rod in the core [mm] ao, al... etc. = constants given as follows:

a0 = 1.0802871 10m5, al = 4.6260476 10e4, a2 = 2.644029 1O-4

a3 =3.237764 10m7, a4 =-8.1093212 10e9, a5 =1.6867707 10“’

As reactor power increases, the coolant density decreases, causing a decrease in core reactivity hence stabilizing the neutron flux. Due to its low value of available excess reactivity, changes in coolant temperature have a drastic negative effect on core reactivity and are expressed as a linear function as a, = 0.026445 x 1 Om3 - 0.0034752 x lo-‘T

where IX~ = moderator temperature coefficient of reactivity [mk/“C] T = core coolant temperature [“Cl

SYSTEM SOFTWARE HIGHLIGHTS A reactor model was assembled that includes major components and system control. Simulation of the model was realized with C++ Builder’. The Builder, a component-based tool, proved to be an efficient environment that can create robust and fast applications. It is widely known that visual programming allows the programmer to construct rather huge models out of small program codes using already existing libraries and components. The simulation program includes two forms (windows) consisting of an introductory and working spreadsheets. The working spreadsheet consists of schematic reactor drawing, 5 charts, and set of control buttons. The reactor drawing, consists of essential elements

t The unit of reactivity is expressed as 1 mk=1O-3 * p * $

382 I. Khamis et al.

such as annular and top and bottom reflectors as well as fuel and control rods with cooling channels in and around the core. Pointing the mouse pointer to each component, the drawing reveals the name of this element. The five charts show variation with time for neutron flux, core inlet and outlet temperatures, and the increase of temperature across the core. The neutron flux, due to its wide range of variation, was partitioned into two charts: local and global time variations. The local chart is represented on the basis of one-hundred- seconds’ duration. The global one, treated as the other three charts, is represented on the basis of one-hour duration. The control buttons are dedicated mainly for reactor operation i.e. the movement of the sole control rod, up or down. Pressing the up or down buttons of the control rods will direct the drawing of the control rod to move also in real time. In addition, there is other control buttons for shutdown and pause or speed up of the time scale of simulation. Colours of essential elements of the schematic drawing of the reactor, affected by temperature changes, are altered automatically pending the magnitude of change. The range of temperature increase was divided into different segments and a different colour is assigned to each segment. Main variables for colour change are fuel and the coolant channel across the core, i.e. at inlet, outlet, and across the core. Other characteristic variables such as power excess reactivity, gamma radiation doses . . .etc. are displayed numerically on real time basis on the screen. The software stores all variables as an ASCII data tile versus time in the RAM of the PC. At the end of each run, the program will ask whether the operator wants to store the results in such a file on the PC permanent memory.

RESULTS Once the simulator was implemented, it was used to simulate the reactor response for different aspects of operations. Results show that the simulator agrees very closely with real data. Figure 1 represents the simulator’s screen as well as some graphical data outputs. The screen consists of technical drawings of the main reactor components, 5 diagrams for variable representations, and control buttons for simulator operation. As can be seen from Figure 1, reactor response during normal operation starting from control rod withdrawal to shutdown can be observed as a function of real or accelarated time. Reactor variables, as function of time, are stored in a batch file as listed in Table 1. Selected variables to-be-stored in this file can be chosen from a menu that appears directly after the simulator was prepared for running.

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Time position 1 0 75 185 129 185 183 185 239 185 251 185 213 185 283 185 300 185 309 185 312 185 318 185 325 185 333 185 342 185 352 185 358 185 312 185 390 185 402 185 418 185 441 185 599 185 650 185 696 185 741 185 786 185 833 185 882 185 934 185 990 185 1052 185 1121 185 1201 185 1296 185 1416 185 1580 185 1840 185 2387 185 2546 171 2560 171 2571 166 2582 161 2591 151 2600 153 2608 150

I. Khamis eral.

inlet-T n flux POWER

22.00 0.00 22.00 0.00 22.01 0.03 22.04 0.28 22.08 2.71 22.09 4.22 22.14 8.85 22.17 11.92 22.25 18.51 22.30 22.54 22.31 23.95 22.35 26.81 22.41 30.21 22.41 34.08 22.56 38.34 22.61 42.84 22.74 45.38 22.91 50.80 23.17 56.61 23.35 59.77 23.60 63.16 23.99 66.60 26.66 66.49 21.41 63.29 28.02 60.25 28.56 51.34 29.04 54.59 29.49 51.94 29.90 49.43 30.29 47.06 30.64 44.80 30.98 42.65 31.30 40.61 31.60 38.66 31.88 36.82 32.14 35.05 32.40 33.31 32.63 31.77 32.88 30.26 32.92 28.85 32.91 21.31 32.90 26.07 32.89 24.79 32.87 23.46 32.85 22.32 32.83 21.15

l.zOOE+07 9.3573+07 9.7003+08 9.7433+09 9.372EtlO 1.458Etll 3.044Etll 4.089Etll 6.313Etll 7.669Etll 8.140Etll 9.098Etll 1.0233+12 1.1523+12 1.2933+12 1.4423+12 1.5263+12 1.7053+12 1.8973+12 2.000E+12 2.1123+12 2.2243+12 2.2143+12 2.1073+12 2.006Et12 1.909Et12 1.8183+12 1.7303+12 1.646Et12 1.5673+12 1.4923+12 1.4203+12 1.3533+12 1.2883+12 1.2273+12 l.l68E+12 l.l12E+12 1.0593+12 1.009Et12 9.566Etll 9.105Etll 8.6683+11 8.199Etll 7.791Etll 7.374Etll 7.017E+ll

Out T 22.01

T-Fuel 23.02

22.05 23.10 22.25 23.46 23.06 24.92 26.37 30.78 27.77 33.27 31.21 39.36 33.11 42.75 36.68 49.13 38.64 52.64 39.29 53.81 40.58 56.14 42.06 61.83 43.67 62.29 45.38 62.78 47.12 68.43 48.09 68.79 50.10 69.55 52.23 17.93 53.39 78.53 54.66 79.22 56.03 80.01 58.17 82.17 57.82 82.35 57.40 82.44 56.94 82.47 56.48 73.93 56.01 73.92 55.54 73.90 55.08 73.88 54.63 73.84 54.18 73.81 53.74 73.76 53.31 67.45 52.88 67.41 52.46 61.37 52.05 67.32 51.66 67.28 51.28 67.24 50.76 92.44 50.15 91.36 49.60 90.39 49.05 89.39 48.46 88.33 47.95 87.39 47.40 86.40

h3rana 0.00 0.00 0.05 0.55 5.51 8.62 18.25 24.66 38.50 41.03 50.01 56.10 63.34 71.64 80.81 90.55 96.09 108.00 120.90 128.00 135.80 144.00 148.20 142.20 136.30 130.60 125.10 119.70 114.50 109.70 105.00 100.60 96.38 92.39 88.65 85.15 81.96 79.24 77.45 74.34 70.55 67.23 63.97 60.55 57.63 54.61

Table 1. Stored data file of simulation

CONCLUSIONS

Dynamic simulator for the MNSR 385

It has been demonstrated in the present work that a model of the MNSR consisting of theoretical and empirical correlation can be implemented successfully to form an easy-to-use computer simulator. The model provides a very efficient tool to conduct scoping and parametric studies of the reactor. The simulator is useful for training in the fields of operation and some nuclear engineering studies. The C++ Builder proved to be an excellent tool for programming purposes and simulation at large with relatively little modeling and analysis expertise. Thereby, the C++ Builder saves considerable engineering and computer costs by reducing or even eliminating the need for mainframe or workstation computers.

ACKNOWLEDGMENT The authors are in great debts to Prof. I. Othman, the Director General of the Atomic Energy Commission of Syria, for the encouragement and continued support.

REFERENCES

1. The Syrian MNSR Safety Report, internal report, 1992. 2. Hetrick, D.L. “Dynamics of Nuclear Reactors”, Chicago Press Ltd., 197 1.

3. Lewins, J. “Nuclear Reactor Kinetics and Control”, Pergamon Press, Oxford,

1978. 4. Press, W.H.; Teukolsky, S.A.; Vetterling, W.T.; and Flannery, B.P.

“Numerical Recipes in C”, 2”d Ed., Cambridge Univ. Press, 1992. 5. Calve& C.

“Borland C++ Builder Unleashed”, Borland Press, 1998.