Increasing Nuclear Power generation

[Arun_S]

This thread is for discussion about the current state of the art and what can be done to globally increase Nuclear power generation, with particular focus to India and emerging economies that need more energy as they catch-up to energy consumption pattern in Europe and USA. And for Europe and USA to decrease their carbon emission footprint to a level comparable to global average.

Clearly the traditional First generation nuclear reactors are too inefficient in use of the raw fuel, and its waste generating foot print has issues with quantity and quality of waste. One should take into consideration not just the reactor but the whole fuel cycle, that should be sustainable and fair.

In my very humble opinion, the Indian Second stage (FBR) and Third stage (Thorium) is an excellent template to solve not only Indian energy demands but also almost all other countries.

IMVHO global investment as well as cooperation in this field can be both profitable for companies and countries that develop key technologies as well as symbiotically it allows sustainable global growth.

A nearly closed fuel cycle is a very desirable goal for sustainability. And spent fuel handling and re-processing is not the devil that has been made by the myth making of 70's, 80's and 90's, it far more safe and benign than what has been painted in public psyche.

Greater automation in reprocessing is key for future use of Nuclear power.

Having said that significant correction are required in the Indo-US 123 agreement that precludes civil nuclear cooperation between India and the world. The way the 123 is drafted today Indian civilian power production from sustainable FBR and Thorium fuel cycle is in an isolated cocoon because the current 123 does not allow import of Plutonium nor enriched Uranium > 20%, nor allows locally enriching imported Uranium to greater than 20% enrichment. Thus limiting India to use imported fuel only for First stage reactors.

This is bad for India and this it is also bad for world. There has to be a new and more open civil nuclear cooperation with India. There is a lot that the world can gain from Indian experience and leadership in FBR and Thorium.

The current nuclear thread are to general and thus very suitable for this type of more focused discussion.

[Arun_S]

This document is a good primer to understand Thorium powered nuclear reactor and fuel cycle. Pls note that the AHWR type reactor can use both Plutonium or Enriched Uranium.
DAE: Shaping the 3rd Stage of Indian Nuclear Power Program

Nuclear Characteristics
The nuclear characteristics of thorium have an immense bearing on the selection and development of technologies associated with thorium fuel cycle. Let us have a look at some of these important characteristics. If one compares the absorption cross section1 for thermal neutrons, of 232Th vis a vis 238U, (7.4 barns vs. 2.7 barns), it becomes clear that 232Th offers greater competition to capture of the neutrons and as such a lower proportion of the neutrons will be lost to structural and other parasitic materials. Moreover this improves conversion of 232Th to 233U. 233U has an eta2 value greater than 2.0 and it remains constant over a wide energy range, in thermal as well as epithermal regions, unlike 235U and 239Pu. It makes thorium fuel cycle less sensitive to the type of thermal reactor.

The capture cross section of 233U is much smaller (46 barns) than the other two fissile isotopes (101 barns for 235U and 271 barns 239Pu) for thermal neutrons, while the fission cross section is of the same order (525 barns for 233U, 577 barns for 235U and 742 barns 239Pu). Thus nonfissile absorption leading to higher isotopes with higher absorption cross sections (234U/ 236U and 240Pu respectively) is much less probable. This makes recycling of 233U less of a problem from reactivity point of view compared to plutonium.

Thorium as Fissile Host
From fuel cycle analysis point of view one can compare the characteristic of the two fertile isotopes, viz., 238U and 232Th, as fissile hosts.
For simplicity we assume 235U as the fissile feed. The following conclusions are then arrived at:

! The amount of fissile enrichment required to achieve a given burn-up is higher for lower discharge burn-ups for thorium fuel. But at
higher discharge burn-ups3 , beyond about 50 GWD/t, the initial enrichment required is lower. This is due to the breeding of 233U in thorium system.

! In terms of energy extracted, i.e., fuel utilisation, thorium as fertile host overtakes 238U as host, at about 45 GWD/t. This is because the bred plutonium saturates at about 0.6% at high burn-ups, while 233U saturates to nearly 1.5%.

! In absolute terms, uranium (low enriched uranium, LEU) gives better utilisation up to a discharge burn-up of about 45 GWD/t, while higher burn-ups have to be achieved in thorium cycle to reap its benefits. These burn-up values, though considered high in early days, are now well within the current day water reactor fuel technology.

! There are other advantages of using 232Th that merit consideration. Variation of reactivity with burn-up is smaller in thorium based fuel due to a relatively higher value of fuel inventory ratio (FIR)4 leading to better operational characteristics such as flatter core power distribution that can help reduce the use of burnable or soluble poison. Finally, in thorium based fuel, there is effective utilisation of external fissile fuel whether it is 235U or Pu, added initially. It is for this interesting feature that thorium is getting increasing attention for plutonium dispositioning.

! In terms of operation, multiple recycling5 of plutonium has adverse effects on reactivity coefficients in thermal reactors, with even void reactivity becoming positive. For fuel dispositioning, employing inert matrix (IM, non-fertile metal alloys containing Pu) fuel also makes the reactivity coefficients so small that poses serious safety issues. As a result, only about one-third to one-fourth of the core can be loaded with such a fuel, bringing down the overall plutonium disposition rate via this route. Thorium as a plutonium carrier enables considerable improvement in both the cases.

! Use of (Th, Pu) MOX in Pressurised Water Reactors (PWRs), improves plutonium burn-up vis-a-vis (Th, U-233) MOX, but the reactivity coefficients turn highly negative which might lead to strong feed back effects. On the other hand, in a PHWR, full core can be loaded with (Th, Pu) MOX fuel without much deterioration in the safety characteristics of the core. Fraction of fissile plutonium burnt is also close to that in the case of IM.

The foregoing discussion brings out the major advantages of thorium utilisation in a Heavy Water Reactor (HWR). The design of the corresponding reactor system should meet the internationally stipulated criteria now being evolved for the next generation nuclear power plants. The Indian AHWR belongs to such class of next generation reactors.

Bottom line is that if India can import enriched U253 the AHWR can be fired up like crazy to meet the energy needs. Even without reprocessing the amount of feed fuel it needs for generating power will put less pressure on global uranium reserve and mining foot print. Thorium OTOH is more easily available and huge amount exists on surface itself (E.g. Monazite sand in the beaches of south India), thus Thorium mining footprint is negligible.

That does require removing the 20% enrichment limit on imported fuel from 123 agreement.