Thorium Power.

Haldilal

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Advanced Heavy Water Reactor(AHWR) is being designed by BARC for thorium utilisation and demonstrate all aspects of the thorium fuel cycle. The AHWR is a vertical pressure tube type thorium-based reactor cooled by boiling light water and moderated by heavy water. The equilibrium fuel cycle of AHWR is being optimised for a discharge burn-up of 40 GWd/Te with Plutonium as makeup fuel in a closed fuel cycle. The 233U content in the Uranium and fissile Plutonium content in the Plutonium have been assumed to be about 78% and 75% respectively. One of the important passive safety features of AHWR is heat removal through natural circulation which is governed by a strong neutronic and thermal hydraulic coupling. The neutronics design simulations were done for average coolant condition. In order to study the effect of natural circulation and its consequences on the coolant behaviour and the resulting power distribution, the neutronics and thermal–hydraulic coupled calculations were carried out. The neutronics calculations code (FEMFOL) and thermal–hydraulic calculations code (ARTHA) were externally coupled to estimate the coolant density distribution and the critical heat flux ratio (CHFR) for the fuel. This paper gives the details of the critical heat flux ratio estimation and the effect of density distribution on the core power distribution at some selected operating phases during core follow-up of AHWR.

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Haldilal

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Nuclear power programme of India envisages a large-scale utilization of thorium as it has large deposits of thorium while the availability of uranium is limited. A thorium based Advanced Heavy Water Reactor (AHWR) is being designed and developed in India for thorium utilization. AHWR is a 920 MWth / 300 MWe, vertical pressure tube type thorium-based reactor cooled by boiling light water and moderated by heavy water. It uses thorium in closed fuel cycle with 233U in a self-sustaining configuration employing plutonium as the external fissile feed and derives about two-third of its power from thorium fuel. It has several passive safety
features including negative coolant void reactivity coefficient and heat removal through natural circulation. All through the design development of AHWR use has been made of our experience in design, operation and safety aspects of Pressurized Heavy Water Reactors (PHWRs) and Boiling Water Reactors (BWRs) currently operating in India. In this paper, we give a brief account of the physics design features/ objectives and how they have been
achieved in the current design.

1. INTRODUCTION.

The AHWR is a 300MWe/ 920 MWth, vertical pressure tube type thorium-based reactor cooled by boiling light water and moderated by heavy water. The prime objective is to produce power utilizing thorium available abundantly in India from a relatively simple system with enhanced safety level [1-6]. It is endowed with several innovative safety features such as negative coolant void reactivity, heat removal through natural circulation and passive containment cooling. The development of reactor design has drawn heavily on the experience generated through design and operation of Pressurized Heavy Water Reactors (PHWR) and Boiling Water Reactor (BWR) in India. It was an opportunity to develop a reactor system using thorium-based fuel and gain some valuable experience. A non proliferate thorium/ 233U based closed fuel cycle is chosen for AHWR. Plutonium discharged from PHWRs is used as the fissile seed fuel with thorium for the generation of 233U and then as a top-up fuel in the equilibrium core along with self-sustaining 233U in the thorium matrix. The physics design has several challenges in achieving negative void reactivity, fuel cycle, spatial core control, on-line fuelling and minimization of inventory of plutonium fuel. It is difficult to achieve negative coolant void coefficient in a heavy water moderated pressure tube type reactor. For this a multi-pronged approach involving pitch reduction, heterogeneous cluster design and use of mild absorbers is chosen. Plutonium bearing fuel is located separately in the outer region of the cluster with self-sustaining 233U bearing fuel in the inner region of the cluster. A small amount of mild absorber if required can be placed in the multipurpose displacer located in the centre of the cluster . The void coefficient varies with burn up and it is a challenge to have it negative throughout the core. The state of nuclear data for the elements of interest and type of neutron spectrum in the reactor puts heavy demand on the calculation models and validation of reactivity coefficients to ensure safety. A critical facility has especially been designed to carry out various lattice experiments to validate calculation models and nuclear data. AHWR is a reactor with largely thermal spectrum and employs on-line fueling. Fuel cycle flexibility is its inherent characteristics and the current design works very well with the LEU fuel and also with thorium/ LEU fuel [9-10]. Equilibrium core is designed to run on self-sustaining closed fuel cycle of 233U with plutonium discharged from PHWRs added as top-up fuel to gain in fuel burn up. Two variants of the fuel cluster with different fraction of plutonium are used to achieve self-sustaining thorium/ 233U fuel cycle. It is seen that uranium fuel does not rapidly get degraded in the closed fuel cycle in AHWR and ensuring same inventory of 233U in the reprocessed uranium suffices to a large extent to run the next cycle. However, to arrive at equilibrium core configuration 233U must be generated in situ. For this purpose, initial and pre-equilibrium core is loaded largely with thorium plutonium fuel. To overcome the scarcity of fissile plutonium, uranium-plutonium fuel is also used in the initial core of AHWR.

AHWR is neutronically a large reactor in comparison to the currently operating PHWRs that makes it susceptible to xenon induced oscillations and other spatial instabilities. It is seen that the first azimuthal mode is close to unstable and requires spatial control and monitoring. It is difficult to provide a large number of control elements and in-core detectors at a relatively tight lattice pitch. A quadrant control scheme is chosen that matches with the thermal hydraulic design of the reactor. Thereare six control elements in each quarter to control the reactivity and power distribution. About 150 SPNDs are installed in the core for on-line flux mapping and core monitoring. Further analysis is being carried out to ascertain the need of in-core detectors for the safety purpose.

A Kumar.
 

Haldilal

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2. PHYSICS DESIGN FEATURES.

Heat removal by natural convection in normal operation and in accident condition is basic design feature of AHWR and dictates other characteristics. Effective and efficient utilization of thorium and the top up fuel plutonium with enhanced safety features is the prime objective from physics standpoint. Efficient utilization of thorium is possible in self sustaining mode in the closed fuel cycle only. As the plutonium that comes from the reprocessing of PHWR spent fuel is also scarce, its use too must be optimized to get high discharge fuel burn up. To achieve the above objectives, the physics design has evolved from a seed-blanket core design to a core consisting of a single type of cluster called D5 composite cluster containing both (Th-233U) and (Th-Pu) MOX fuel pins in the equilibrium core configuration. Some of the important physics design safety features are detailed below.

2.1 Reactivity coefficients / Fuel cluster.

The coolant void reactivity coefficient in a pressure tube system is normally positive as the system is over moderated. In AHWR coolant has been replaced by light water coolant and that would make it more positive. The cluster design is mainly dictated by the objective of achieving negative void reactivity coefficient. To achieve the above objectives, the physics design has evolved from a seed blanket core design with widely varying cluster types to a core consisting of a single type of cluster called D5 composite cluster. The void reactivity can be made negative with harder spectrum, which could be achieved either by changing the properties of the moderating medium or by decreasing the inventory of moderator. It is also possible to achieve negative void coefficient by using a burnable absorber in the fuel or in isolated pins in an inert matrix [6-7]. On voiding, burnable absorber is made to absorb more neutrons. In an earlier design of AHWR a burnable absorber was employed in the centre of the cluster which made the void coefficient of reactivity negative. The fuel utilization, however, suffered a lot due to large quantity of burnable absorber required. A radical modification in the cluster design was then carried out by putting fuel in the area of high neutron importance and relocating the ECCS in a multipurpose displacer in the centre of the cluster. With this change in the cluster design there was significant reduction in the void reactivity coefficient and increase in fuel utilization. Target fuel discharge burn up could now be increased to 36,000 MWd/T from 20,000 MWd/T. In the current design, lattice pitch is further reduced to 225 mm to obtain harder neutron spectrum and that helped in the elimination of dysprosium absorber [12-13]. Now only a small quantity of grey absorber Zr/ SS is needed to be used, if required, in the displacer region. Fuel reconstitution and usage of different fuel types has also become very easy with the introduction of multi-purpose displacer. Void coefficient of reactivity varies with burn up but core averaged void reactivity is always negative. Fuel temperature coefficient is always negative and increases slightly with burn up as plutonium burns. Power coefficient is also negative throughout the range of interest. The standard equilibrium core cluster is heterogeneous fuel assembly comprising (Th,233U)MOX and (Th, Pu)MOX both and is called the D5 composite cluster arranged in a circular array of 54 fuel pins.The inner and middle ring of 12 and 18 pins contain (Th, 233U) MOX and the outer 24 pins contain (Th, Pu)MOX. The inner ring of 12 pins has a 233U content of 3.0 wt% by weight and the middle 18 pins have 3.75 wt% 233U. The outer ring of (Th, Pu)MOX pins have an average of 3.25 wt% of total plutonium. The lower half of the active fuel will have 4.0 wt% plutonium and the upper part will have 2.5 wt% plutonium.

A Kumar.
 
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Haldilal

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2.2. 233U in self-sustaining configuration.

One of the major objectives is to operate AHWR in the closed fuel cycle. The 233U burns and is bred in situ in the cluster. It is possible only if system is self-sustaining in 233U. With irradiation, the 233U content would deplete in the inner (Th,233U)MOX pins and would gradually build-up in the outer (Th, Pu)MOX pins due to conversion from thorium. The production of 233U in the outer pins is however not sufficient to compensate for the burn up of 233U in the inner pins. An alternate cluster is designed for this purpose. The alternate cluster as shown in FIG.1 is basically same as the standard cluster except that the 12 fuel pins in the alternate cluster contain 4.0 wt% plutonium in place of 3.0 wt% 233U used in the standard cluster. Reactor is fuelled with both the clusters in equal proportion. It is now possible to achieve the self-sustaining characteristics of 233U with the annual plutonium consumption increasing by about 50 kg to 175 kg. The uranium from the discharged fuel would contain other isotopes of uranium like 234U, 235U and 236U. But unlike the problem associated with the recycling of MOX in PWRs, the situation here is simple as the rate of build-up of higher isotopes and their absorption cross-section is much lower. The even higher isotopes do absorb neutrons but their reactivity load is compensated by 235U. And it is possible to operate AHWR in successive cycles by employing the same enrichment of 233U.

2.3. Plutonium as top up fuel.

The usage and location of the plutonium in the cluster is important from the basic design objective of minimising the plutonium requirement. This is done by locating the plutonium pins in the outermost ring of the cluster, where it faces the highly thermalized flux. The plutonium used as top up fuel comes from the spent fuel of PHWRs. It has about 75% fissile plutonium. By virtue of being placed in the region of higher thermal flux, the plutonium leads to high fuel burn up. It however depletes at a faster rate that result in strong burn up dependence of some safety parameters such as void reactivity
coefficient.

2.4. Graded fuel enrichment/ power uprating.

The axial power profile is bottom-peaked in a BWR due to bulk boiling that leads to decreased
moderation in the upper part of the core. In AHWR, in the absence of bulk boiling, the axial flux
distribution is not enough distorted as there is not much change in the neutron moderation. The neutron flux with uniform fuel enrichment would therefore be nearly the same in the top and bottom portion of the core. One way to increase the power derived from the reactor is by decreasing the flux at core exit that increases minimum critical heat flux ratio (MCHFR) and hence thermal margins. An increased thermal margin translates directly into increased power. Desirable axial power distribution for this purpose is achieved by altering the plutonium content in the outer pins only; the lower half of the fuel assembly is loaded with 4.0 wt% plutonium and upper half with 2.5 wt% plutonium. With this the power derived from the reactor could be increased by about 20%.

2.5 Multi-purpose displacer.

The fuel cluster was redesigned by creating a region in the centre of cluster by displacing good amount of water, hence called displacer. ECCS was also shifted to this region and, in addition, it acts as an aid (support structure) for fuel reconstitution [7]. The region called multi-purpose displacer is basically a zircaloy-2 tube of 36 mm OD/ 30 mm ID housing ECCS with small holes radially for cooling fuel pins. To achieve desired negative coolant void reactivity characteristics a grey rod of an appropriate material may be located in the centre of displacer.

A Kumar.
 
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Haldilal

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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.
 

piKacHHu

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I am reading same thing for last 2 decade now. end result is that we are still building more CANDU reactors copies.have we operationalize the one we were building down south.
Not yet. It's Still on the drawing board.

CANDU design is more practical and well suited for Indian Nuclear program as of now. The new ones couldn't be called copies but enhanced Indianised CANDU.

Thorium cycle has several limitations vis-a-vis conventional Uranium fuel cycle.
 

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