Nuclear Fuels

PWR fuels :

Pressurized water reactor (PWR) fuel elements are made of uranium oxide pellets sheathed in Zircaloy tubes of about 1 cm diameter. These fuel elements are arranged in 14x14 or 17x17 formation and are about 4 meters in length. The fuel cladding gap is filled with helium gas to improve the conduction of heat from the fuel to the cladding. There are about 179-264 fuel rods per fuel bundle and about 121 to 193 fuel bundles are loaded into a reactor core. The fuel bundles are usually enriched . The uranium oxide is dried before inserting into the tubes to try to eliminate moisture in the ceramic fuel that can lead to corrosion and hydrogen embrittlement. The Zircoloy tubes are pressurized with helium to try to minimize pellet cladding interaction (PCI) which can lead to fuel rod failure.

CANDU fuel :

The fuel bundles are about a half meter long and 10 cm dia. They consist of sintered (UO2) pellets in Zirconium alloy tubes, welded to Zirconium alloy end plates. Each bundle is roughly 20 kg, and a typical core loading is on the order of 4500-6500 bundles, depending on the design. The bundle has typically have 37 identical fuel pins radially arranged used. The CANFLEX bundle has 43 fuel elements, with two element sizes. It is also about 10 cm (four inches) in diameter, 0.5 m (20 inches long) and weighs about 20 kg (44 lbs) and replaces 37-pin standard bundle. It has been designed specifically to increase fuel performance by utilizing two different pin diameters. Current CANDU designs do not need enriched uranium to achieve criticality (due to their more efficient heavy water moderator), however, some newer concepts call for low enrichment to help reduce the size of the reactors.

BWR Fuel:
It is similar to PWR fuel except that it is canned to prevent density changes near the fuel as the same can affect the nuclear reactions and thermal hydraulics of the reactor.The no of fuel pin per assembly is of the order of 90's varying to design to design. The no of assemblies depend on the size of the core.

Magnox Fuel :The Metallic fuel is used In Magnox reactor which are gas cooled reactors operating in UK. The size varied from 50MWEe to ~500MWe. They were the precursors of the Advanced Gas cooled reactor. Unenriched Uranium is cladded with an alloy of Mg-Al and other metals in small amounts. The main disadvantage of this fuel is limit on max fuel pin temp hence the efficiency of the plant and reactivity of Magnesium with water prevents long term under water storage.

TRISO Fuel: It consists of a fuel kernel composed of uranium oxide (sometimes Uranium carbide or UCO), coated with four layers of three isotropic materials.

1. The first layer is a porous buffer layer made of carbon.
2. The second layer pyrolytic carbon (PyC).
3. The third ceramic layer of Siilicon Carbide retains the fission products and gives the TRISO particle structural integrity.
4. The outer layer of is of PyC.

TRISO fuel particles are designed not to crack at temperatures beyond 1600°C(due to differential thermal expansion or released fission gas pressure). They can contain the fuel in the worst accident scenario in a properly designed reactor. Two such reactor designs are pebble bed modular reactor (PBMR), in which thousands of TRISO fuel particles are dispersed into graphite pebbles, and a prismatic-block gas cooled reactorin which the TRISO fuel particles are fabricated into compacts and placed in a graphite block matrix. Both of these reactor designs are high-temperature gas-cooled reactors (HTGR), which is a type of very high temperature reactors (VHTR).

Types of Stainless Steels

Austenitic Steels: This comprises the 300 series of stainless steels and accounts for over 70% of total stainless steel production. Nickel is added to stabilise the austenite structure of iron, manganese can also be added to preserve the austenitic structure but at a lesser cost. Austenitic steels contain a maximum of 0.15% carbon, a minimum of 16% chromium and sufficient nickel and/or manganese to retain an austenitic structure at all temperatures from the cryogenic region to the melting point of the alloy.
Superaustenitic stainless steels are produced by adding higher amounts of manganese.
High Molybdenum causes greater resistance to chloride pitting and crevice corrosion if it has content (>6%).Higher nickel content ensures better resistance to stress-corrosion cracking over the 300 series. The 300 Series—austenitic chromium-nickel alloys:

Type 304—the most common grade of steel; the classic 18/8 stainless steel
Type 301—highly ductile, for formed products. Also hardens rapidly during mechanical working. Good weldability. Better wear resistance and fatigue strength than 304.
Type 302—same corrosion resistance as 304, with slightly higher strength due to additional carbon.
Type 303—easier machining version of 304 via addition of sulfur and phosphorus. Type 304—the most common grade; the classic 18/8 stainless steel.
Type 309— better temperature resistance than 304
Type 316—the second most common grade (after 304); for food and surgical stainless steel uses; Alloy addition of molybdenum prevents chloride attack and crevice corossion. 316 steel is used in the manufacture and handling of food and pharmaceutical products where it is often required in order to minimize metallic contamination. It is also known as "marine grade" stainless steel due to its increased resistance to chloride corrosion compared to type 304. SS316 is often used for building nuclear reprocessing plants. Most watches that are made of stainless steel are made of Type 316L; Rolex is an exception in that they use Type 904L.
Type 321 :similar to 304 but lower risk of weld decay due to addition of titanium.
Type 347:with addition of niobium for desensitization during welding.

Steel Gr	Composition
303	17-19 Cr, 8-10 Ni, 0.15 C, 2.0 Mn, 1.0 Si, 0.20 P, 0.15 S min, 0.60 Mo (optional)
304	18-20 Cr, 8-10.50 Ni, 0.08 C, 2.0 Mn, 0.75 Si, 0.045 P, 0.030 S, 0.10 N
304L	18-20 Cr, 8-12 Ni, 0.03 C, 2.0 Mn, 0.75 Si, 0.045 P, 0.030 S, 0.10 N
316	16-18 Cr, 10-14 Ni, 0.08 C, 2.0 Mn, 0.75 Si, 0.045 P, 0.030 S, 2.0-3.0 Mo, 0.10 N
316L	16-18 Cr, 10-14 Ni, 0.03 C, 2.0 Mn, 0.75 Si, 0.045 P, 0.030 S, 2.0-3.0 Mo, 0.10 N
309	22-24 Cr, 12-15 Ni, 0.20 C, 2.0 Mn, 1.0 Si, 0.045 P, 0.030 S
347	17-19 Cr, 9-13 Ni, 0.08 C, 2.0Mn, 0.75 Si, 0.045 P, 0.030 S (Nb +Ta, 10 xC min,1 max)
420	12-14 Cr, 0.15 C min, 1.0 Mn, 1.0 Si, 0.040 P, 0.030 S
440A	16-18 Cr, 0.60-0.75 C, 1.0 Mn, 1.0 Si, 0.040 P, 0.030 S, 0.75 Mo

MOX fuels and Alternate Fuel Cycles

Looking beyond natural uranium alone as fuel material, several strategies/approaches are under consideration for use in PHWRs. Use of reprocessed material will reduce volume of spent fuel material and disposable fuel waste. Consequently this will reduce overall fuel cycle costs. Short length fuel bundles and on-power refueling provision in PHWRs provides flexibility to use variety of fuel loading patterns and different fuel types and consequently permits optimum use of fuel in the reactor. Following paragraphs cover the alternative fuel designs and core loading concepts in use or under consideration for use in Indian PHWRs.
Thorium
As part of Indian long term fuel cycle strategy of using thorium, irradiation of thorium is planned present power reactors to gather some experience. In the 220 MWe PHWRs, 35 Thorium bundles have been used for flux flattening in the initial core such that the reactor can be operated at rated full power in the initial phase. These bundles are distributed throughout the core in different bundle locations, both in the high power and low power channels. This loading was `successfully demonstrated in KAPS-1 and subsequently adopted in the initial reactor loading of KAPS-2, KAIGA-1 & 2 and RAPP 2,3&4. So far 232ThO2 bundles have been successfully irradiated in different reactors. The thorium dioxide fuel bundle fabrication and irradiation has provided valuable experience.
It is now planned to irradiate thoria bundles to higher burnups with suitable modification in design. It is also planned to take up loading a few thorium bundles regularly during equilibrium reactor operation.
MOX-7
It is also proposed to load MOX fuel in one of the existing PHWRs. For this purpose, MOX-7 bundle design has been evolved, which is a 19-element cluster, with inner seven elements having MOX pellets consisting of 0.4 wt % Plutonium dioxide ( about 70% fissile) mixed in natural uranium dioxide and outer 12 elements having only natural uranium dioxide pellets.
Based on detailed studies, an optimised loading pattern and refuelling scheme has been evolved for loading the bundles in an existing operating reactor. The scheme evolved is to load MOX-7 bundles in outer burnup zone and retain natural UO2 fuel bundles in inner burnup zone. The present natural uranium core will be converted gradually to a mixed MOX - natural UO2 core in a span of about 3 years. The core average discharge burnup in equilibrium core increases to around 9000 MWd/TeHE with this scheme. Due to this the fuelling rate comes down by 25%.
Initially trial irradiation of 50 number of MOX-7 bundles in one of the KAPS reactors is being taken up this year. Regulatory review has been completed and permission has been obtained for this purpose. Special bundle transport package and storage racks have been developed such that subcriticality is assured. The 50 fuel bundles are currently under fabrication and loading of these in the reactor would commence by this year end.
The different advanced fuel cycles relevant to PHWRs were reviewed during the nineties by a Committee appointed by DAE (Ref.2). The committee is also of the view that recycling of plutonium in PHWRs could provide an elegant way of dealing with the available spent fuel inventories. The major fuel cycle cost is back end cost and is typically 45-65%. Hence the fuel cycle which reduces back end cost by having higher burnup tends to be cheaper. The other high contributor to the energy costs is fuel fabrication. For, the MOX fuel, the back end cost is less compared to Nat. U cycle due to higher burnup. However the MOX fuel fabrication cost is 6 times that of Nat. U. In view of this the fuelling cost for MOX can be made competitive only by going to high burn-ups.
Depleted Uranium
In earlier years, RAPS, MAPS and NAPS reactors were loaded with 384 to 550 depleted uranium fuel bundles as a part of initial core fuel loading for the purpose of flux flattening. Recently, schemes have been worked out for fresh PHWR cores to maximize the use of depleted uranium whereby 40% of the fuel bundles can be of depleted uranium with U235 content of around 0.6%. The fresh core of MAPS-2, after the en-masse coolant channel replacement, was loaded in this fashion, effecting significant savings in natural uranium requirement.
Similar reactor physics studies have been carried out for use of large of number depleted Uranium bundles as a part of initial fuel loading in 540 MWe Reactors coming up at Tarapur. Fuel loading in first of these units will be taken up around mid 2004. The loading scheme consists of loading of depleted uranium bundles with different uranium 235 contents.
Theoretical studies, to use depleted uranium in combination with natural uranium for regular refueling in some of the current operating 220 MWe PHWRs has been completed.
Use of depleted uranium results in significant savings in available natural uranium reserves. Assuming that the depleted material is free, the depleted uranium fuel bundle cost consists of fabrication cost and other levies and it works out to be 50% of Nat U bundle cost.
Schemes are also worked out to load slightly enriched uranium (SEU) fuel bundles in PHWRs with 0.85% U235. This gives maximum energy output per kg of natural U processed.

Indian PHWR Fuel

The fuel bundles of PHWRs India are short cylindrical assemblies. Each coolant channel has 12 or 13 such fuel bundles.
220 MWe Reactor fuel : 19-element fuel bundle design. Stack of cylindrical sintered natural uranium dioxide fuel pellets, inside a zircaloy fuel tube and sealed at both ends by end plugs. The fuel elements are arranged in concentric rings and are assembled together by welding them to an end plate on each side to form a bundle. The bundle length is 495 mm and the weight is 16 Kgs. The spacers and bearing pads are attached to fuel elements by spot welding. The inside surface of fuel sheath is graphite coated to decrease the fuel element failure rate due to power ramps.
540 MWe and 700 MWe Reactor fuel: 37 element fuel bundle design is an extension of the closed packed 19 element fuel bundle. One more ring of elements has been added. All the elements are of a small diameter of 13 mm. The bundle has been designed to generate a bundle power of about 1 MW.

On power bi-directional fuelling is done using two fuelling machines, one at either end of the coolant channel. Pressure tubes containing a string of short length fuel bundles and the on-power refueling permit flexibility in choosing fuel designs and in-core fuel management parameters to maximize fuel utilization. A defective fuel can also be identified and removed from the reactor while it is in operation.

The unit energy cost distribution for PHWRs:

The fuel consumption cost:- 18% to 25% of unit energy cost.

The Capital cost:- typically 27%, .

The operation and maintenancecost:-typically10%

The heavy water costs :- Typically 32% .

Paradox:
PHWRS are optimally designed from neutron economy considerations, uses natural uranium fuel economically in comparison to other types in terms of extracting maximum energy per gram of natural uranium . Still the fuel consumption cost is high due to high cost of indigenous raw meterial. MOX fuels

Indian Nuclear Power Facts

India is poorly endowed with uranium. Further, Indian uranium reserves are of extremely low grade. India is extracting uranium from less than 0.1% ores compared to ores with 12-14% uranium in certain resources abroad. This makes Indian nuclear fuel 3-4 times costlier than international supplies. The substantial thorium reserves can be used but that requires converting the fertile thorium into fissile material.
Following are the some Important facts about developed countries and Nuclear energy (Billion KW) and as percentage of total energy produced:

Nuclear Power Generation Country wise
Country	Billion KWh	% of Total Generation
France	428	78%
USA	787	12.19%
Lithuania	8.0	69%
Switzerland	26.4	37%
Japan	291.5	30%
UK	69.2	18%
India	15.6	2%

In this context, a three-stage nuclear power programme is envisaged. This programme consists of setting up of Pressurised Heavy Water Reactors (PHWRs) in the first stage, Fast Breeder Reactors (FBRs) in the second stage and reactors based on the Uranium 233-Thorium 232 cycle in the third stage. It is also envisaged that in the first stage of the programme, capacity addition will be supplemented by electricity generation through Light Water Reactors (LWR), in itially through Technology imports with long term objective of indigenisation. PHWR technology was selected for thr first stage, as these reactors are efficient users of natural uranium for yielding plutonium fuel required for the second stage FBR programme. The FBRs will be fuelled by plutonium and will also recycle spent uranium from the PHWR for breeding more plutonium fuel for electricity generation. Thorium s blanket material in FBRs will produce Uranium 233 to start the third stage.

The Approximate Potential Available in Nuclear Energy

Particulars	Amount	Thermal Energy		Electricity
		TWh	GWyr.	GWe-Yr	MWe
Uranium-Metal	61,000-T	-	-	-	-
in PHWR	-	7,992	913	330	10,000
In FBR	-	1,027,616	117,308	42,200	5,00,000
Thorium-Metal	2,25,000-T	-	-	-	-
In Breeders	-	3,783,886	431,950	1,50,000	Very large

( Data from draft report of the expert committee on integrated Energy policy-Planning Commission)

Indian Plutonium Stockpile estimates (Albright)

The most widely accepted estimates of India's plutonium production have been made by David Albright ([Albright et al 1997], [Albright 2000]). His most recent estimate (October 2000) was that by the end of 1999 India had available between 240 and 395 kg of weapon grade plutonium for weapons production, with a median value of 310 kg. He suggests that this is sufficient for 45 - 95 weapons (median estimate 65). The production of weapon grade plutonium has actually been greater, but about 130 kg of plutonium has been consumed - principally in fueling two plutonium reactors, but also in weapons tests. His estimate for India's holdings of less-than-weapons-grade plutonium (reactor or fuel grade plutonium) are 4200 kg of unsafe guarded plutonium (800 kg of this already separated) and 4100 kg of IAEA safeguarded plutonium (25 kg of this separated). This unsafeguarded quantity could be used to manufacture roughly 1000 nuclear weapons, if India so chose (which would give it the third largest arsenal in the world, behind only the U.S. and Russia).

Research Reactors at Trombay
Name	Year	Type
Apsara	August 1956	Swimming pool type
CIRUS	July 1960	Canadian reactor
Dhruva	August 1985	Heavy water cooled & Moderated

There has been conflicting statements regarding the lifetime capacity factors of the Dhruva and CIRUS reactors. U.S. officials, for example strongly feel that India's Cirus and Dhruva plutonium production reactors have a lifetime capacity factor of about 40 percent whereas. Indian officials have stated that the average capacity factor is significantly greater, as large as 60 percent. Thus for the estimate, the most likely choice was selected as 40 percent with values up to 60 percent having a diminished probability of occurring. On the other end, a lifetime capacity factor less than 30 percent was viewed as highly unlikely.

India's inventory was estimated by arriving at the total production of weapon-grade plutonium in the Cirus, Dhruva, and power reactors and thereafter subtracting the amounts used in nuclear testing, processing losses and civil uses of the plutonium in the power reactors. The median value, which is the value midway between the smallest and largest value, is about 310 kilograms of weapon-grade plutonium at the end of 1999. The estimated values range between 180 kilograms and 480 kilograms. But the values in the most probable 90% an 10% percentile range are 250 kilograms and 375 kilograms.

Latest update from Albright 2004 document...........
While making the latest estimates the lifetime capacity factor the Cirus reactor, covering the period from start up in 1960 until shutdown in 1997, is estimated as a triangular distribution in the Crystal Ball® calculations, with the most likely value as 50 percent and the minimum and maximum values as 30 and 70 percent have been used.

The capacity factor for the Dhruva reactor is estimated for two different periods. The first period stretches from 1985 through 1998, and the second period is from 1998 through 2004.
The capacity factor for the first period is represented as a triangular distribution with the most likely value as 40 percent and the minimum and maximum values as 30 and 60 percent, respectively.
For the second period, the multi-year capacity factor is represented as a uniform distribution with a minimum of 55 percent and a maximum of 75 percent.
The calculation for total plutonium production for the military program through 2004 gives a median value of about 575 kilograms of weapon-grade plutonium. The range is defined as all values between the 5th and 95th percentiles, which are 495 and 665 kilograms, respectively. One way to interpret the results is that there is a 90 percent certainty that the true value lies between 360 and 530 kilograms of weapon-grade plutonium, where the median value is about 450 kilograms.

The largest overall users of plutonium from these reactors have been civil reactors utilizing plutonium fuels, including the Fast Breeder Test reactor (FBTR), the Purnima reactor, the Zerlina reactor, and power reactors. Nuclear testing in 1974 and 1998 also used a portion of this plutonium.
As above, many of these draw downs had to be estimated and are represented by ranges in the calculation. The calculation for the total amount of draw downs has a median of about 130 kilograms and 5th and 95th percentiles of 110 and 150 kilograms, respectively.
At the end of 2004, the median value of the estimate of this inventory is 445 kilograms of plutonium, and the 5th and 95th percentiles are 360 and 530 kilograms, respectively.

Magnox Reactors

Magnox reactors are pressurised, carbon dioxide-cooled, graphite-moderated reactors using natural uranium (i.e. not enriched) as fuel and magnox alloy as fuel cladding. Boron-steel control rods were used.

On power fuelling was an economically essential part of the design, to maximise power station availability by eliminating refuelling downtime. This was particularly important for Magnox as the unenriched fuel had a low burn-up, requiring more frequent changes of fuel than most enriched uranium reactors.

Early reactors have steel pressure vessels, while later units (Oldbury and Wylfa) are of reinforced concrete; some are cylindrical in design, but most are spherical.

Technical Features:

Steam Quality: There is very little difference in the steam conditions between the American light water reactor and the European gas cooled reactors. Both produced saturated steam at approximately the same temperature and pressure.
In gas cooled and pressurized water reactors, the steam systems were separate, non-radioactive systems, a feature that was a good selling point to customers concerned about the unknown dangers of radioactive contamination.



1. there was a view that gas reactors would eventually provide better steam conditions as material knowledge improved and as inert gas coolants like helium became more available.


2. Construction Costs:The Magnox reactors had low maximum fuel temperatures and low coolant heat transfer capability thereby were several times larger than a LWR with the same power output.


3. Magnox reactors required construction of large, high purity graphite structures with tight tolerances and very large, high quality pressure vessels. Being very large to transport were built at site.


4. LWR imposed different constraints. The reactor internals were also carefully manufactured components with tight tolerances, but were small enough to be produced in a factory for later transport . The pressure vessel that enclosed the reactor internals was a challenging component and required a large investment in specialized manufacturing equipment, but the final product was small enough to be transported provided there were rail or water routes available. So the manufacturers were interested in a large no of deals to get back their investment.


5. The fuel used in the Magnox reactors was natural uranium metal clad with Magnox alloy. Initially maximum burn-up obtainable was about 3000 MWD/ Te of heavy metal, but it improved to about 6000 MWD/ Te ton . In 1960, the cost per kilogram of Natural uranium $18.00.


6. The fuel for the light water reactors was uranium oxide with a U-235 concentration of 3 percent clad with either stainless steel or zirconium alloy. At first, the maximum burn-up for this fuel was about 5000 MW days per ton, but it improved to about 25,000 MW days per ton within a few years. In 1962, the cost per kilogram of 3 percent enriched uranium hexafluoride (the direct product of the enrichment plants) was listed by the AEC as $254.00.


7. Disposal Costs: The natural uranium reactors produced a larger volume of high/medium level waste because of larger reactors with lower burn-up fuel . This tended to raise the cost estimates for decommissioning those reactors. This factor was countered by longer plant life estimates based on the lower stresses and lower neutron irradiation of the pressure vessel. The waste volume could also be reduced by fuel material and moderator recycling.


8. Large PWR have a significant cost disadvantage compared to gas cooled reactors as the pressure vessels are more highly contaminated and normally had to be cut up before disposal. The barges and rail lines that delivered the vessel were frequently at their capacity limits in moving an empty vessel, there is little space or weigh capacity left for adding the shielding.


9. Gas cooled vessels will also have to be dismantled, but it is far easier to cut a steel wall that is <> vessels.

Indian Renewable Power resources

Indian Government has accorded very high priority to develop and expand installed capacity base through non-conventional sources of electricity generation. There is a separate Ministry in the Government of India to exclusively focus on this important area of power generation.

	Potential	Existing capacity
Wind	45,000	1870 MW
Small Hydro (upto 25 MW)	15,000	1406MW
Solar (PV)	20 MW/Sq.Km	2490KW
Bio power (woody biomass/cogeneration)	57,000MW	542.8 MW
Urban/Industrial waste based plant	5000 MW	43.45MW

excerpts from the : Keynote Address in Global Energy Dialogue at Hanover (Germany) on April 25, 2006

The aggregated potential of the renewable energy resources is about 1,30, 000 MW.

Advanced Heavy water Reactor features

ADVANCED HEAVY WATER REACTOR (AHWR)

India is developing the Advanced Heavy Water reactor (AHWR) as the third stage in its plan to utilise thorium to fuel its overall nuclear power program. It is a 300MW vertical pressure tube type reactor using heavy water as moderator and boiling water as coolant in natural circulation mode at low pressure (~ 70 bar).

The calandria has 500 vertical pressure tubes and the coolant is boiling light water circulated by convection. Each fuel assembly has 30 Th-U-233 oxide pins and 24 Pu-Th oxide pins around a central rod with burnable absorber. Burn-up of 24 GWd/t is envisaged. It is designed to be self-sustaining in relation to U-233 bred from Th-232 and have a low Pu inventory and consumption, with slightly negative void coefficient of reactivity.

The major changes from the PHWR design is that AHWR is a thorium fuel based, Light water replaces the high pressure heavy water coolant circulated using a pump .
During its designed plant life of 100 years AHWR will generate 65% of the power from ThO2 based fuel. AHWR is the first of its kind in the world not only because of its most attractive feature of heat removal from the reactor core by natural circulation under all conditions but also due to the fact that it incorporates a host of other passive safety features that are in line with the approach being pursued world over for development of inherently safe reactor system by incorporating safety features that do not call for any human intervention or any active control devices for reactor safety.

The AHWR Fuel

The initial core will be made up of entirely (Th, Pu-239) MOX fuel assemblies.

• The U233 bred in the 54 pin (Th, Pu-239) MOX fuel pin will be progressively recovered and recycled as (Th, U233) MOX.
• At equilibrium, the core of AHWR will consist of composite fuel assemblies each having 24 nos. of (Th, Pu239) MOX pins and 30 nos. of (Th, U233) MOX pins arranged in three consecutive rings having fissile material compositions as shown hereunder:
  (i) 12 (Th-U233)O2 pins with U233 enrichment of 3.0% in inner most or 1st ring.
  (ii) 18 (Th-U233)O2 pins with U233 enrichment of 3.75% in intermediate or 2nd ring.
  (iii) 24 pins consists of (Th-Pu)O2 pellets with plutonium enrichment of 3.25% in outermost or 3rd ring.
  Fissile isotopes content of plutonium will go down from initial 75% to 25% level at equilibrium discharge burn-up level (which would not be possible to recycle in AHWR but can be recycled to FBR or ADSS with fast neutron spectrum).

To reduce the overall inventory of waste, it is envisaged that Th and U233 will be recycled in AHWR. Even though U234 produced (along with U235 and U236) by neutron capture in U233 has negative influence on reactivity, it might be possible to recycle U233 in AHWR with only a marginal penalty of less than 1000 MWd/Te on discharge burn-up for each recycling.
The initial core characteristics and equilibrium characteristics on AHWR are shown in Table 2 and Table 3 respectively. Plutonium in AHWR burns faster due to large absorption cross section

that leads to loss in reactivity. An option is kept available in AHWR to reconstitute the fuel cluster after an averaged discharge burn-up of 24,000 MWd/Te. In reconstitution, only plutonium pins in outer rings are replaced by fresh fuel. Rest of the fuel cluster remains as it is. It is possible to obtain an additional burn-up of upto 20,000 MWd/Te from the reconstituted cluster. The cluster reconstitution improves U233 production and reduces the reprocessing load due to increase in average cluster burn-up. Reconstitution of fuel cluster involves multiple enrichments for the (Th-Pu)O2 pins, which will affect the fuel fabrication. However, reconstitution improves fuel conversion and hence economics of fuel cycle.
The fuel cycle time of AHWR is 8 years : 4 years for residence in reactor residence and two years for cooling (to allow for >99.9% conversion of Pa-233 to U233), 1 year of reprocessing and 1 year for refabrication. For the initial few years, annual reload would consist of (Th-Pu)O2 clusters only.

Thorium Resources

The Thorium is a naturally-occurring,slightly radioactive metal discovered by Swedish chemist Jons Jakob Berzeliusin 1828.

Country	Reserves (tonnes)
Australia	300 000
India	290 000
Norway	170 000
USA	160 000
Canada	100 000
South Africa	35 000
Brazil	16 000
Other countries	95 000
World total	1 200 000

Thorium as a Nuclear fuel

Thorium, as well as uranium, can be used as a nuclear fuel. Although not fissile itself, thorium-232 (Th-232) will absorb slow neutrons to produce uranium-233 (U-233), which is fissile. Hence like uranium-238 (U-238) it is fertile.
In one significant respect U-233 is better than uranium-235 and plutonium-239, because of its higher neutron yield per neutron absorbed. Given a start with some other fissile material (U-235 or Pu-239), a breeding cycle similar to but more efficient than that with U-238 and plutonium (in slow-neutron reactors) can be set up. The Th-232 absorbs a neutron to become Th-233 which normally decays to protactinium-233 and then U-233. The irradiated fuel can then be unloaded from the reactor, the U-233 separated from the thorium, and fed back into another reactor as part of a closed fuel cycle.
Over the last 30 years there has been interest in utilising thorium as a nuclear fuel since it is more abundant in the Earth's crust than uranium. Also, all of the mined thorium is potentially useable in a reactor, compared with the 0.7% of natural uranium, so some 40 times the amount of energy per unit mass might theoretically be available (withouit recourse to fast breeder reactors

Indian Nuclear Reactors

The present status of The Nuclear Power plant in India

Location	Unit Name	Capacity (net MWe)	Utility	Type	Reactor Supplier	Percent Complete	Expected / Actual Date of operation
Kaiga Karnataka	Kaiga 1	202	NP	PHWR	NPCIL	75	11/1998	1999
Kaiga Karnataka	Kaiga 2	202	NP	PHWR	NPCIL	75	11/1998	2000
Kakrapar Gujarat	Kakrapar 1	202	NP	PHWR	DAE/NPCIL	100		11/1992
	Kakrapar 2	202	NP	PHWR	DAEC/NPCIL	100		03/1995
Kalpakkam, Tamil Nadu	Kalpakkam 1	155	NP	PHWR	DAE	100		07/1983
	Kalpakkam 2	155	NP	PHWR	DAE	100		09/1985
Kota, Rajasthan	Rajasthan 1	90	NP	PHWR	AECL	100		11/1972
	Rajasthan 2	187	NP	PHWR	AECL/DAE	100		11/1980
	Rajasthan 3	202	NP	PHWR	NPCIL	70	11/1998	2000
	Rajasthan 4	202	NP	PHWR	NPCIL	70	05/1999	2000
	Rajasthan 5	450	NP	PWHR	--	0	2007	2008
	Rajasthan 6	450	NP	PWHR	--	0	2008	2009
Kudankulam, Tamil Nadu	Kudankulam 1	1,000	NP	PWR	--	0	2006	2008
	Kadunkulam 2	1,000	NP	PWR	--	0	2008	2010
Narora, Uttar Pradesh	Narora 1	202	NP	PHWR	DAE/NPCIL	100		07/1989
	Narora 2	202	NP	PHWR	DAE/NPCIL	100		01/1992
Tarapur, Maharashtra	Tarapur 1	150	NP	BWR	GE	100		04/1969
	Tarapur 2	150	NP	BWR	GE	100		05/1969
	Tarapur 3	450	NP	PHWR	NPCIL	100	08/2003	8/2006
	Tarapur 4	450	NP	PHWR	NPCIL	100	05/2004	9/2005

Electricity demand in India has been increasing rapidly, and the 534 billion kilowatt hours produced in 2002 was almost double the 1990 output, though still representing only 505 kWh per capita for the year. This per capita figure is expected to almost triple by 2020, with 6.3% annual growth. Thermal power plants provides over half of the electricity at present, but reserves are expected to last about 80yrs.
Nuclear power supplied 15.6 billion kWh (2.6%) of India's electricity in 2006 from 3.5 GWe (of 110 GWe total) capacity and this will increase steadily as new plants come on line. India's shortage of fossil fuels, is driving the nuclear investment for electricity, and 25% nuclear contribution is foreseen by 2050, from one hundred times the 2002 capacity. Almost as much investment in the grid system as in power plants is necessary.
In 2006 almost US$ 9 billion was committed for power projects, including 9354 MWe of new generating capacity, taking forward projects to 43.6 GWe and US$ 51 billion.