Breeding K.S. Rajan Professor, School of Chemical & Biotechnology SASTRA University

Transcription

1 Breeding K.S. Rajan Professor, School of Chemical & Biotechnology SASTRA University Joint Initiative of IITs and IISc Funded by MHRD Page 1 of 7

2 Table of Contents 1 NEED FOR BREEDING COMPARISON OF THERMAL AND FAST SPECTRUM FOR BREEDING REFERENCES/ADDITIONAL READING... 7 Joint Initiative of IITs and IISc Funded by MHRD Page 2 of 7

3 In this lecture we shall discuss briefly about breeding in nuclear reactors At the end of this lecture, the learners will be able to (i) understand the need for breeding in nuclear reactors (ii) identify materials and conditions for breeding (iii) classify breeders 1 Need for Breeding Per-capita energy consumption of a nation is an indication of growth and prosperity of a nation. As all of us are aware that energy consumption throughout the world is increasing continuously due to several reasons that include population increase, better standard of living in terms of comfort and industrialization. Hence any mode of energy generation meant to serve industry and society needs to be sustainable. The problem of rapid depletion of oil and coal reserves is well known. One method of achieving sustainability is to improve the efficiency of energy generation/conversion such that the energy generated from unit mass of fuel is maximized. Another method is the re-use of material in one form or the other in order to minimize the amount of fresh fuel inventory required. Hence inline with these principles, power generation from nuclear fuel must also be made sustainable for long-term use. The natural uranium contain very little amount of fissile isotope (0.7 %) while the rest is U-238. Hence in typical pressurized heavy nuclear reactors, to utilize one kg of fissile isotope for power generation, nearly 100 kg of natural uranium is used. During fission reaction by neutron bombardment, the fissile material is destroyed or consumed. If fissile material could be produced from fertile material available abundantly in the core, inventory of fresh fuel required can be reduced. The conversion of fertile material to fissile material due to neutron irradiation in the core can be achieved by one of the nuclear reactions called nuclear transmutation. This is called breeding in which fissile isotopes are produced from fertile isotopes. Nuclear reactors that produce more fissile isotope than they consume are called breeders or breeder reactors. Depending upon the neutron energy utilized for breeding, breeders are classified as fast breeders or thermal breeders. 1.1 Comparison of thermal and fast spectrum for breeding Fast Breeders or Fast Breeder Reactors (FBR) are the common breeder reactors utilized in the world. For successful breeding, at least one neutron must be exclusively available for interaction with fertile isotope for every neutron absorbed by fuel, after taking into account of neutron absorption in core and structural materials. For Joint Initiative of IITs and IISc Funded by MHRD Page 3 of 7

4 example, for every fast neutron absorbed by Pu-239, 2.45 neutrons are produced. If one neutron is utilized for sustenance of chain reaction and on an average 0.45 neutron is lost, the remaining one neutron is available for conversion of fertile to fissile isotope. With U-235, an average of 2.1 neutrons are produced for every neutron absorbed. After accounting for neutrons required for sustenance of chain reaction and for losses, on an average less than one neutron is available which is insufficient for breeding. In other words, the number of neutrons produced per neutron absorbed must be greater than 2 by atleast a certain amount to facilitate breeding. Now let us analyze the scenario in thermal spectrum. The number of neutrons produced per neutron absorbed in Pu-239 and U-235 due to bombardment of thermal neutrons is 2.04 and 2.06 respectively, clearly indicating that thermal breeders with these fissile isotopes are impractical. Hence most of the breeder reactors are fast reactors utilizing Pu-239 as the fissile isotope. Hence such breeder reactors are called fast breeder reactors. Fast breeder technology for power generation is confined only to few countries like France, Russia, China, India, Japan and South Korea. It may be noted that thermal reactors are used in several countries for power generation while only the above countries have active fast breeder programme. One of the less-explored or lesser known fissile material is U-233. This isotope does not occur in nature and hence produced artificially in reactors using Th-232 as fertile isotope. In Indian fast reactor, Th-232 is used in the blanket, from which U-233 is produced. A fast reactor operating with U-233 as the fissile material will produce 2.31 neutrons per every neutron absorbed and hence with appropriate core design aimed at minimizing neutron loss, a fast breeder reactor operating with U-233 as fissile material can also be built. Breeding is possible even with thermal neutrons if U-233 is used as fissile material, as 2.26 neutrons are produced per neutron absorbed. A breeder reactor utilizing thermal neutrons is called thermal breeder. India is developing an Advanced Heavy Water Reactor that will act as thermal breeder utilizing U-233 as fissile isotope and Th-232 as fertile isotope. Availability of large reserves of thorium in India (world s 1/3 of total thorium reserves) is a prime reason for India s interest in breeders utilizing U Table 1. Number of neutrons produced per neutron absorbed for different isotopes in fast and thermal spectrum Neutron Energy Number of neutrons produced per neutron absorbed (η) U-235 Pu-239 U ev > 1 MeV Joint Initiative of IITs and IISc Funded by MHRD Page 4 of 7

5 It is to be noted that the number of neutrons produced per neutron absorbed (η), also called reproduction factor shown in Table 1 is not a measured quantity. It is determined from other measurable quantities as follows: η 0 =υσ f /σ a = υσ f /(σ c +σ a ) = υ/(1+α) (1) In the above equation, σ a, σ f and σ c refer to absorption, fission and capture cross sections. υ refers to the number of neutrons produced per fission. η 0 is the number of neutrons produced per neutron absorbed for a pure fissile material. When fuel is a mixture of fertile and fissile material, reproduction factor (η) is η=(ν 1υ 1σ f1 +N 2 υ 2σ f2 )/(Ν 1σ a1 + Ν 2σ a2 ) (2) In Equation (2), the suffices 1 and 2 represent the isotopes 1 and 2 (say U-235 and U- 238); N 1 and N 2 are the number fractions of isotopes 1 and 2. The following example shows calculation of reproduction factor for a typical scenario. Example 1: Determine the reproduction factor in a fast reactor operating with a pure fuel whose fission and absorption cross sections are 1.39 b and 1.64 b respectively. The average number of neutrons produced per fission is Recall Eq. (1) η 0 =υσ f /σ a = υσ f /(σ c +σ a ) = υ/(1+α) σ f = 1.39 b σ a = 1.64 b υ = Therefore, reproduction factor (η 0 ) = Example 2: The following data pertain to a nuclear reactor utilizing natural uranium (N U-235 /N U-238 =0.007) and operating in thermal spectrum: U-235 U-238 σ f = 586 b σ f = 0 b Joint Initiative of IITs and IISc Funded by MHRD Page 5 of 7

6 σ a = 681 b υ = 2.42 σ a = 2.7 b υ = 0 Determine the reproduction factor. Let 1 and 2 represent the isotopes U-235 and U-238 respectively. Using Eq. (2), η=(ν 1υ 1σ f1 +N 2 υ 2σ f2 )/(Ν 1σ a1 + Ν 2σ a2 ) Dividing the numerator and denominator of right side of the above Equation by N 2, we get η=(ν 1υ 1σ f1 / N 2 +υ 2σ f2 )/(Ν 1σ a1 /Ν 2+ σ a2 ) η=( )/( ) = 1.33 Therefore, the reproduction factor under the above circumstances is 1.33 Example 3: The following data correspond to U-233 in a thermal reactor. σ f = 531 b σ a = 576 b υ = 2.49 Determine the reproduction factor if the fuel is pure. Recall Eq. (1), η 0 =υσ f /σ a Substituting the data in Eq. (1), reproduction factor is obtained as 2.30 Example - 4: If a fast reactor is loaded with U-235 and U-238 such that N U-235 /N U- 238=0.25, comment on the possibility of breeding. Data on cross section and average number of neutrons produced per fission are given below: U-235 U-238 σ f = 1.4 b σ a = 1.65 b υ = 2.60 σ f = b σ a = b υ = 2.60 Joint Initiative of IITs and IISc Funded by MHRD Page 6 of 7

7 Let 1 and 2 represent the isotopes U-235 and U-238 respectively. Using Eq. (2), η=(ν 1υ 1σ f1 +N 2 υ 2σ f2 )/(Ν 1σ a1 + Ν 2σ a2 ) η=(ν1υ1σ f1 / N 2 +υ2σ f2 )/(Ν1σ a1 /Ν2+ σ a2 ) η=( )/( ) = 1.73 The reproduction factor is 1.73 and hence it is not possible to achieve breeding with this configuration. Hence one may understand from Examples 1 to 5, that the reproduction factor could be varied by using fuels with different enrichment levels or composition and by use of neutrons of appropriate energy (fast neutron or thermal neutron). In summary, fast breeder reactors are practical with Pu-239 as the fissile material, while thermal breeders can be built with U-233 as fuel. 2 References/Additional Reading 1. R.L. Murray, Nuclear Energy: An Introduction to the Concepts, Systems, and Applications of Nuclear Processes, 5/e, Butterworth Heinemann, 2000 (Chapters 11 and 13) 2. A.E. Waltar, D.R. Todd, P.V. Tsvetkov (Eds.), Fast Spectrum Reactors, Springer, 2012 (Chapter 1) Joint Initiative of IITs and IISc Funded by MHRD Page 7 of 7