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3. LATTICE AND CORE CALCULATIONS/ COMPUTER CODES/ SOURCES OF
UNCERTAINTY.
The lattice evaluations have been carried out by WIMS-D/4 code and employ the 69 group library based on the ENDF-B/VI.8 dataset obtained from IAEA. This library comprises 14 fast, 13 resonance and 42 thermal groups. The D5 cluster has been modelled as a circular array cluster using the WIMSD code system. The methodology adopted was a heterogeneous infinite lattice cell calculation followed by a homogenous leakage calculation. The sequence of calculation is a detailed flux spectrum in the 69 groups for each of the principal regions of the lattice and then a detailed geometrical representation for a more accurate spatial solution using transport theory methods. The burn-up calculations have been performed with critical flux spectrum and operating temperatures. The results have been tested with the other code CLUB developed in BARC and widely used for PHWR simulations. The burn up, temperature and void/ coolant density dependent two energy group parameters are generated by the lattice calculations. These are used in the core simulation. The core calculations are carried out by 3D code FEMINA based on the higher order nodal expansion method. The advantage of this method is that meshes as large as fuel assembly size can be taken without any loss in accuracy. Heterogeneity within the core too is simulated accurately. As AHWR is a thermal system two energy group formulations appears sufficient for the core simulation. The code has been tested with several LWR/ PHWR benchmarks and has been used for PHWR calculations.
Heat removal through natural convection is an important feature of this reactor, which would require a uniform coolant flow and a low power density. In order to have good thermal hydraulic and neutronic coupling, a uniform radial power distribution is preferred. It also requires the height of the active core to be kept small with respect to the diameter of the core. The core height has been chosen to be 3.5 m and the vessel Calandria ID is 6.9 m from this point of view. Hence the number of lattice locations channels is large at 513 in the core, out of which 452 locations are occupied by fuel and the rest by reactivity devices. Main core design and safety features are given in Table II. Layout of AHWR equilibrium core configuration self-sustaining in 233U. The core simulations presented here have been done for the equilibrium core configuration. The time averaged simulations are done to get optimum discharge burn up and flattened channel power distribution for the equilibrium core configuration, whereas homogeneous simulations were used to calculate worth of reactivity devices. The limits on channel power are governed by the minimum critical heat flux in the channel. For this, physics and thermal hydraulic iterations are necessary. For this purpose, the neutronics code and the thermal hydraulic code has been combined to run in tandem.
Starting with the same average coolant density and iteratively modifying the coolant density with respect to power distribution. These iterations are continued till a convergence of <0.1% in the mesh powers is obtained. The channel power converges in 3 to 4 iterations. The converged power distribution in quarter core for the equilibrium core self-sustaining in 233U.
It can be seen that overall reactivity swing from cold critical to hot standby, excluding reactivity changes coming from fission products, is quite low at about 8 mk. Though the coolant temperature coefficient is positive with a total swing of about 11.0 mk, the channel temperature coefficient is about 6.0 mk positive and can easily be controlled by Reactor Regulating System (RRS). Moderator temperature coefficient, too, is low positive and is within the range of regulating system. All reactivity swings from cold critical to hot zero power and hot zero power to full power are well within the range of reactor regulating system. The reactor start up is slow operation carried out under supervision of RRS. There is enough reactivity depth in RRS to take care of reactivity changes during start-up. Though the coolant temperature coefficient is positive, fuel temperature coefficient is prompt and negative which is of prime importance during a transient. Equilibrium xenon is about 22 mk, which can be controlled either by boron or by gadolinium in moderator. Transient load of xenon following reactor shut down is only 9.0 mk due to considerably low levels of thermal flux. As stated above, computer codes used in the calculations are well tested and benchmarked. Major source of uncertainty, however, comes from the nuclear data used. Main physics safety parameters coolant temperature and coolant void reactivity coefficients are highly susceptible to the multigroup library used for the lattice calculations. Fuel temperature coefficient is, however, high negative throughout the burn up range and does not depend much on the multigroup library/ dataset used. A series of experiments are planned in the low power critical facility for this purpose.
A Kumar.