The Nuclear Fuel Cycle Lecture 4

The Nuclear Fuel Cycle Lecture 4 David J. Hamilton d.hamilton@physics.gla.ac.uk 31st January 2011

1. Overview The back-end of the cycle (continued): Once-through and geological disposal; Reprocessing-recycling strategies. Current MOX fuel utilisation Near-term advanced reactor and fuel technology

2. Recap (Lecture 3) UK gas-cooled reactor designs include Magnox and AGR. The Chernobyl accident is a good case-study in reactor kinetics. A prompt supercriticality occurred as a result of operating the reactor in an unstable configuration where the RBMK's positive void coefficient became dominant. Current fuel cycle strategies can be divided into two groups: Once-through cycle; Reprocessing/Recycling cycles; The front-end of the fuel cycle includes uranium mining and milling, enrichment and fuel manufacture. Spent reactor fuel is typically composed of 95% uranium, 1% plutonium, 4% FP and 0.1% MA. Radiotoxicity is an important measure of the ability of spent fuel to cause harm in the event of a release to the biosphere.

3. Spent Fuel Disposal Multiple Barrier System In the once-through cycle spent fuel is intended for disposal with no intention for retrieval. The primary goal in spent fuel disposal is isolation, containment and retardation of radiotoxicity over geological time-scales. To achieve this goal and prevent release into the biosphere a multiple barrier system is employed: The first barrier is the fuel matrix itself; Then comes the container system; Followed by the disposal vault; Finally, there is the geological barrier. Container system: spent fuel rods placed in steel or copper canisters (Swedish/Finnish design lifetime of 100,000 to 1 million years).

3. Spent Fuel Disposal Deep Geological Repository The spent fuel canisters are to be disposed of in stable deep geological repositories, such as: Yucca Mountain, USA (Tuff); Olkiluoto, Finland (Granite); Oskarshamn, Sweden (Granite); Gorleben, Germany (Salt). Mobility is an important characteristic associated with geological disposal. It related to the ease with which certain components of the spent fuel will be transported away in the presence of water. Pu and MA's have very low mobility. Oklo is an example of a natural analogue, highlighting the potential stability of radioactive disposal in the presence of groundwater over 1.8 billion years.

4. Problems with the Once-through Cycle RECYCLABLE MATERIALS URANIUM 95 to 96 % Geological disposal has associated with it issues with local acceptance, safety in the very long-term and uncertain costs. The once-through cycle is sometimes called by critics as the throw-away cycle, due to the inefficient use of resources. Recycling of the re-usable components of spent fuel not only presents a more sustainable use of resources but reduces significantly the requirements on repository design. For these reasons, some countries have adopted a reprocessing-recycling fuel cycle strategy. This is only a partial solution to the sustainability problem. RECYCLING PLUTONIUM 1% RECYCLING FISSION PRODUCTS 3 to 5 % WASTE DISPOSAL

4. Problems with the Once-through Cycle Can the Pu and MA components be removed and fully utilised?

5. Spent Fuel Reprocessing Separation of the U and Pu components of spent fuel is achieved through the use of aqueous reprocessing. The industry standard is the PUREX (Plutonium and Uranium by Extraction) process. The spent fuel is dissolved in nitric acid and the U and Pu are separated first from the FP's and then from each other by organic solvent extraction (TBP). This process depends on the different oxidation states of the various elements involved. The products of the PUREX process are: Uranium and plutonium for fuel use; A solution containing the FP's and MA's (HLW).

5. Spent Fuel Reprocessing Industrial Plants There are several industrial-scale reprocessing plants worldwide, the largest being the La Hague and Sellafield plants. The spent fuel is first cut into short lengths then heated to remove radioactive gases (such as Kr-85). It is then dissolved in nitric acid and filtered to remove undissolved remnants of the fuel assemblies. Following extraction, the uranium and plutonium nitrates are converted to oxides. The waste comprising FP's and MA's is vitrified (made into glass) for disposal. Reprocessing substantially reduces the volume of High Level Waste (HLW).

5. Spent Fuel Reprocessing Pu Stockpiles 2004 IAEA Declarations (INFCIRC/549) Unirradiated separated plutonium at reprocessing or fabricating plants Plutonium contained in unirradiated MOX at reactor sites Unirradiated separated plutonium held elsewhere Plutonium included above belonging to foreign countries Belgium France Germany UK 2.1 61.9 0 93.7 1.4 13.2 10.8 1.9 negligible 3.5 1.7 0.6 not sub. 30.5 not sub. 22.5 3.5 thm 48.1 thm 12.5 thm 73.7 thm Countries that have adopted civil reprocessing have had to contend with growing stockpiles of separated plutonium. These stockpiles are stored in oxide form at reprocessing sites. The question facing these nations is: How to best utilise the energetic value associated with these Pu stocks in the reactors of today and tomorrow. Optimisation of Pu consumption.

6. MOX Fuel The utilisation of plutonium today is only (partially) realised in the form of Mixed Oxide fuel (MOX). It is a blend of oxides of plutonium and uranium from reprocessing. The manufacturing process is similar to the UOX process, albeit with extra shielding needed because of the higher radioactivity associated with fresh MOX (due to the decay of Pu-241 to Am-241). The plutonium enrichment required to mimic low-enriched UOX fuel performance is typically between 5 and 10 %.

6. MOX Fuel Pu Composition Fissile Pu (%) Total Pu 100 Weapon-grade 90 Typical Plutonium isotopic composition for PWR spent fuel: plutonium 80 Reactor-grade 70 plutonium 60 50 0 10 1 20 30 2 40 3 4 Burn-up rate (GWd/tHM) Reactor years 52 % 24 % 15 % 6% 2% Pu-239 Pu-240 Pu-241 Pu-242 Pu-238 Only the odd-numbered isotopes of Pu are fissile, the rest are fertile. The required Pu enrichment therefore depends on the isotopic composition of the reprocessed spent fuel from which it came. This depends on the initial U enrichment, the burn-up and the irradiation time.

6. MOX Fuel In-core behaviour The in-core behaviour of plutonium-based fuels such as MOX differs from UOX in a number of significant ways. The delayed neutron fractions (β) for Pu-239 and Pu-241 are smaller than for U-235. This means that larger reactivity control systems (control rods and soluble poisons) are needed for MOX operation. The thermal fission and capture cross sections for Pu isotopes are larger than for U isotopes. This leads to a reduced thermal flux and an overall hardening of the neutron spectrum. A larger moderator-to-fuel ratio is therefore necessary with MOX. The lower thermal flux associated with MOX can lead to flux gradients in cores loaded with UOX and MOX, which can lead to local power peaking. To minimise this effect, several different Pu enrichments are employed in each MOX assembly. MOX has a lower thermal conductivity, meaning that it runs hotter than UOX fuel.

6. MOX Fuel Partial Core Loading Two MOX manufacturing plants are currently in operation: Areva's Melox plant (200 thm/yr); Sellafield MOX plant (120 thm/yr). There are currently 36 moxified reactors in operation in Europe. The MOX loading in these LWR systems is limited to 30 % of the core, the remainder being UOX. The reasons for this limit are due to the effects discussed in the previous slide: there is insufficient control and moderator in current LWR designs. E-ON, GKN, RWE, EnKK ELECTRABEL 2 11 EDF 3 20 NOK, KGD Pu consumption with 30% MOX core loading is: 0 kg/twhe Nonetheless, operational experience with compared to 100% UOX: MOX compares favourably with UOX, with -30 kg/twhe an increase of 12 % in the energy extracted from the original fuel.

7. Advanced Reactor Technology - Overview

8. Advanced LWR There are two main ALWR systems: Areva's 1650 MWe European Pressurised Water Reactor (EPR); Westinghouse's AP1000. The main improvements over previous PWR designs include: Increased plant safety through implementation of passive safety systems; Enhanced economic competitiveness, through larger plants and increased thermal efficiency; Longer plant lifetimes due to improvements in pressure vessel material technology,

8. Advanced LWR Full MOX Loading One of the other key improvements associated with ALWR systems is their ability to load 100% MOX cores. This is due to the fact that: There are more control rods and a larger soluble boron system compared to previous designs; The moderator-to-fuel volume ratio has been increased from 2:1 to 4:1. With only one ALWR burning 100% MOX, the Pu arisings from 7 UOX plants could be balanced. The Pu consumption for 100% MOX loading of an ALWR is: 60 kg/twhe MOX UOX Control rods PWR 900 EPR

8: Advanced LWR: IMF and Thorium Advanced thermal MOX programmes do not make full use of the energetic value of Pu and are not well-suited to multiple recycling. The use of Inert Matrix Fuels (IMF) and thorium-based fuels can lead to much higher Pu consumption rates and much smaller MA fractions in the spent fuel. Mixed ThO2 and PuO2 fuel can be easily manufactured for thermal LWR loading, leading to potential Pu consumption rates of up to 115 kg/twhe. Promising Pu-bearing IMF candidates for thermal LWR utilisation include yttria stabilised zirconia, molybdenum and ferritic steels. Much work is still needed to study in-core behaviour of these fuels for swelling and fission gas release.

9: Advanced HTR Technology Graphite moderated, helium cooled reactors. Outlet gas temperatures of 700 deg C. Small, modular designs (400 600 MWt). Capable of achieving very high burn-ups (750 GWd/tHM for MOX in Peach Bottom) Two different proliferation resistant fuel designs: pebble-bed or prismatic.

9. Advanced HTR Technology - Fuel Pu consumption = 100 kg/twhe

10. Goals of Advanced Fuel Cycles Sustainable use of resources. Reduction of radiotoxicity of spent fuel. Reduction of volume and lifetime of geological repositories. Extraction and re-use of not just plutonium but also the MA's.

11. Limitations of Pu Recycling in Thermal Reactors Single MOX recycling allows for a 12 % increase in energy extracted from the initial ore and a factor of 3 reduction in radiotoxicity c.f. Once-through. However, multi-recycling of MOX in thermal systems is not straightforward: The isotopic composition of the plutonium degrades at each step (meaning higher enrichments are necessary); The radiotoxicity of the resulting HLW actually becomes higher than in the once-through cycle because of MA build-up from radiative capture reactions on U-238 and Pu isotopes. These MA's (predominately americium, neptunium and curium isotopes) can not be utilised/disposed of in thermal critical reactors: They are strong thermal neutron poisons; Require fast neutrons to induce fission (fissionable but not fissile); Have very small delayed neutron fractions. Optimal Pu and MA incineration requires a fast reactor system.

12. Fast Critical Reactors Fast critical reactor systems utilise fast neutrons (> 1 kev) to induce fission and sustain a chain reaction. There is no moderator in a fast core. Extremely compact cores (high power densities) means large amounts of heat removal needed. These considerations led to the use of liquid metal coolant. Several candidates have been used: sodium, lead, lead-bismuth eutectic. Advantages include higher plant thermal efficiency, no boiling and atmospheric pressure operation. http://www.tpub.com/content/doe/h1019v1

12. Fast Critical Reactors Core Physics Fission cross sections are smaller for fast neutrons than for thermal (for U235 it is 500 b thermal and only 2 b fast). Higher fissile enrichment is needed (typical Pu enrichments of 15 35 %) due to the fact that: Prompt neutron lifetime in a fast system is 10-7 s (c.f. 10-4 s for thermal), though delayed neutron fractions (β) for fissile isotopes are around the same. Control rods are less effective in fast systems (B-10 absorption cross section is smaller, leading to a mean free path of 43 cm). Fast neutron fluxes are more homogeneous than thermal, due to the smaller cross sections. Typical burn-ups are much larger than even LWR systems, up to around 100 GWd/tHM.

12. Fast Critical Reactors Breeding The most important difference between fast and thermal systems for advanced fuel cycles is that there are more neutrons released per fission (larger η). In the 60's and 70's there were concerns about uranium price and resources, leading to development in several countries of a fast-breeder reactor programme. The principle involves utilising excess neutrons to breed more fuel (Pu-239) in the outer region of the core known as the blanket (U-238). This would lead to the much sought after closure of the fuel cycle. Several plants were built: SuperPhenix, Dounreay, BNR, Monju). http://www.whatisnuclear.com/articles/fast_reactor.html

12. Fast Critical Reactors Transmutation In recent years, focus has switched to utilising the excess neutrons in a fast reactor for MA incineration. This serves two purposes: Reduction of radiotoxicity, volume and heat of HLW. Use of energy from MA fission. Am-241 and Np-237 transmutation has been demonstrated. The major problem is that only small fractions of MA's can be incorporated into a fast critical core because of small delayed neutron fractions. http://www.whatisnuclear.com/articles/fast_reactor.html Also U-238 is still needed for reactivity control, leading to more MA production.