Sunday, September 30, 2007

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
CountryBillion KWh% of Total Generation
France428 78%
USA787 12.19%
Lithuania8.0 69%
Switzerland 26.4 37%
Japan291.5 30%
UK69.2 18%
India15.6 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
ParticularsAmountThermal EnergyElectricity
TWhGWyr.GWe-YrMWe
Uranium-Metal 61,000-T

-

-

-

-

in PHWR

-

7,99291333010,000
In FBR

-

1,027,616117,30842,2005,00,000
Thorium-Metal2,25,000-T

-

-

-

-

In Breeders

-

3,783,886 431,9501,50,000Very 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
NameYearType
ApsaraAugust 1956Swimming pool type
CIRUSJuly 1960 Canadian reactor
DhruvaAugust 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.




    Saturday, September 29, 2007

    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.


    Monday, September 17, 2007

    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.

    PotentialExisting 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.

    Sunday, September 16, 2007

    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.


    CountryReserves (tonnes)
    Australia300 000
    India290 000
    Norway170 000
    USA 160 000
    Canada100 000
    South Africa35 000
    Brazil 16 000
    Other countries95 000
    World total1 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


    Saturday, September 15, 2007

    Indian Nuclear Reactors

    The present status of The Nuclear Power plant in India

    Location

    Unit NameCapacity (net MWe)

    Utility

    Type Reactor SupplierPercent 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.

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