Plutonium and Minor Actinides Management in Thermal High-Temperature Reactors-The EU FP6 Project PUMA

ICENES 2007 - International Conference on Emerging Nuclear Enegy Systems Istanbul, Turkey, June 3-8, 2007 Plutonium and Minor Actinides Management in Thermal HighTemperature Reactors - The EU FP6 Project PUMA J.C. Kuijper1 NRG Westerduinweg 3, P.O.Box 25, NL-1755 ZG Petten, The Netherlands Tel: +31 224 564506, Fax:+31 224 568490 , Email: kuijper@nrg-nl.com Abstract – The PUMA project, a Specific Targeted Research Project (STREP) of the European Union EURATOM 6th Framework Program, is mainly aimed at providing additional key elements for the utilisation and transmutation of plutonium and minor actinides in contemporary and future (high temperature) gas-cooled reactor design, which are promising tools for improving the sustainability of the nuclear fuel cycle, hereby also contributing to the reduction of Pu and MA stockpiles, and to the development of safe and sustainable reactors for CO2-free energy generation. The project runs from September 1, 2006 until August 31, 2009. PUMA also contributes to technological goals of the Generation IV International Forum. It contributes to developing and maintaining the competence in reactor technology in the EU, and addresses European stakeholders on key issues for the future of nuclear energy in the EU. The paper presents an overview of planned activities and preliminary/expected results of the PUMA project. I. INTRODUCTION The sustainability of the nuclear fuel cycle and the reduction of plutonium (Pu) and Minor Actinides (MA) stockpiles are key issues in the definition of the future nuclear energy mix in Europe. The High Temperature gas-cooled Reactor (HTR) can fulfil a very useful niche for the purposes of Pu and MA incineration due to its unique and unsurpassed safety features, as well as to the attractive incentives offered by the nature of the coated particle fuel. Table I EU FP6 PUMA Consortium. 1 Partner No. 1 Partner acronym NRG 2 AGH 3 BN 4 CIRTEN 5 6 7 8 EDF GA USTUTT JRC-ITU 9 KTH 10 11 12 14 15 LISTO NEXIA NNC LGI DUT 16 17 FZJ AREVA Partner name Country Nuclear Research & consultancy Group University of Science and Technology of Cracow Belgonucleaire The Netherlands Poland Belgium Italy Electricité de France General Atomics University of Stuttgart Joint Research Centre Institute for TransUranics Royal Institute of Technology LISTO bvba NEXIA Solutions LaGrange Innovation Delft University of Technology Forschungszentrum Juelich AREVA On behalf of the EU FP6 PUMA consortium (see Table I) France USA Germany EU Germany Sweden Belgium UK UK France The Netherlands Germany France Main contact Person(s) J.C. Kuijper (coordinator) J. Cetnar S. Shihab G. Toury N. Cerullo G. Lomonaco E. Girardi F. Venneri W. Bernnat J. Somers J. Wallenius L. Van Den Durpel T. Abram D. Millington V. Chauvet J.L. Kloosterman H. Werner C. Trakas ICENES 2007 - International Conference on Emerging Nuclear Enegy Systems Istanbul, Turkey, June 3-8, 2007 The High Temperature gas-cooled Reactor (HTR) can fulfil a very useful niche for the purposes of Pu and Minor Actinide (MA) incineration due to its unique and unsurpassed safety features, as well as to the attractive incentives offered by the nature of the coated particle (CP) fuel. No European reactor of this type is currently available, but there has been, and still is, considerable interest internationally. Decisions to construct such a reactor in China and in South Africa have already been made or are about to be made. It is also developed in Europe, in particular within the RAPHAEL Integrated Project of the European Union 6th Framework Program (EU FP6) [1]. Apart from the unique and unsurpassed safety features offered by this reactor type, the nature of the CP fuel offers a number of attractive incentives. In particular, it can withstand burn-ups far beyond that in either LWR or FR systems. Demonstrations as high as 75% FIMA have been achieved. The coated particle itself offers significantly improved proliferation resistance, and finally with a correct choice of the kernel composition, it can be a very effective support for direct geological disposal of the fuel. The PUMA project is a Specific Targeted Research Project (STREP) within the EURATOM 6th Framework Programme (EU FP6) and runs from September 1, 2006, until August 31, 2009 [2]. The project consortium consists of 15 organisations from 8 European countries and 1 from the USA, as indicated in Table I. This paper presents an overview of planned activities and preliminary/expected results of the PUMA project. II. PUMA OBJECTIVES Complementary with other initiatives, the PUMA project aims at providing key results for the utilisation and transmutation of plutonium and minor actinides in HTRs. These results should further qualify the HTR design as a promising tool for the development of safe, sustainable and CO2-free energy generation. A number of important issues concerning the use of Pu and MA in gas-cooled reactors have already been dealt with in other projects, or are being treated in ongoing projects, e.g. as part of EU FP6. Therefore, the overall objective of the PUMA project is to provide additional key elements for the utilisation and transmutation of plutonium and minor actinides in contemporary and future (high temperature) gas-cooled reactor design, which are promising tools for improving the sustainability of the nuclear fuel cycle, hereby also contributing to the reduction of Pu and MA stockpiles, and to the development of safe and sustainable reactors for CO2-free energy generation. III. PROJECT STRUCTURE Earlier projects show favourable characteristics of HTRs with respect to Pu burning. However, further steps are required to demonstrate the potential of HTRs (and also of Gas-Cooled Fast Reactors - GCFRs) as Pu/MA transmuters based on realistic/feasible designs of CP Pu/MA fuel. So, core physics of Pu/MA fuel cycles for HTRs will be investigated to optimise the CP fuel and reactor characteristics and to assure nuclear stability of a Pu/MA HTR core. It is also envisaged to optimise present Pu CP design and to explore feasibility for MA fuel. New CP designs will be explored that can withstand very high burn-ups and are well adapted for disposal after irradiation. The project benefits greatly from access to past knowledge from Belgonucleaire’s Pu HTR fuel irradiation tests of the 1970’s, and also secures access to materials made at that time. (Very) High Temperature Reactor V/HTR Pu/MA transmuters are envisaged to operate in a global system of various reactor systems and fuel cycle facilities. Fuel cycle studies are envisaged to study the symbiosis between LWR, GCFR and ADS, and to quantify waste streams and radiotoxic inventories. The technical, economic, environmental and socio-political impact shall be assessed as well. The respective activities concerning these three subjects are being carried out in the three main Work Packages of the PUMA project, which will be described in the following sections. IV. CORE PHYSICS OF PU/MA LOADED V/HTR SYSTEMS The first PUMA Work Package is concerned with the core physics, including transient behaviour, of Pu/MA loaded HTRs. The main objectives of this Work Package are: · Demonstration of the full potential of contemporary and (near) future HTGR designs to utilise/transmute Pu and minor actinide fuel within the constraints of safe operation, and based upon realistic assumptions concerning the fuel composition. · To identify necessary additional qualification of the tools employed for the assessment of Pu/MA-loaded HTGR systems, and identify opportunities to obtain experimental data on which such additional qualification can be based. The additionally required qualification (validation) will be identified for the codes to be used in the assessment of envisaged (Pu/MA) HTR system/fuel/fuel cycle combinations. All contributing partners will initially provide information on the code system they intend to use for the analysis of the respective reactor/fuel/fuel cycle combinations. This includes information concerning the (lack of) specific qualification (validation) for the envisaged application and also concerning known deficiencies, as existing at the start of the PUMA project. ICENES 2007 - International Conference on Emerging Nuclear Enegy Systems Istanbul, Turkey, June 3-8, 2007 Table II Major Design and Operating Characteristics of the PBMR-400. PBMR Characteristic Installed thermal capacity Installed electric capacity Load following capability Availability Core configuration Fuel Primary coolant Primary coolant pressure Moderator Core outlet temperature Core inlet temperature Cycle type Coolant mass flow rate Cycle efficiency Effective annular core height Value 400 MW(t) 165 MW(e) 100-40-100% ≥ 95% Vertical with fixed centre graphite reflector TRISO ceramic coated U-235 in graphite spheres Helium 9 MPa Graphite 900°C. 500°C. Direct Brayton 192.7 kg/sec ≥ 41% 11 meters During the course of the project, while performing analyses, further information may arise concerning deficiencies, and lack of validation. Opportunities will be investigated and identified to obtain experimental data on which required/desired additional code qualification can be based. The possibilities will be investigated to perform further code qualification exercises, based on data from (irradiation) experiments on existing high quality Pu coated particles, presently stored at BN. Contacts and information exchange will be initiated/maintained with other EU FP6 projects, as well as with external organisations, such as the American GT-MHR programme (through partner GA), as well as British (through partners NEXIA and NNC) and Russian institutes (specifically the ASTRA critical facility at the Kurchatov Institute, Moscow, through ISTC and IAEA/INPRO), and the OECD/NEA. The latter is to insure that descriptions on which additional code qualification exercises are to be based, are welldefined and complete. For code qualification (benchmark) exercises adherence to OECD/NEA IRPhE requirements [3] is aimed for. However, the main task with this particular Work Package is concerned with the assessment of several HTR system/fuel/fuel cycle combinations. Based on results obtained in particular in EU FP5 projects HTR-N/-N1 [4,5], the core physics work in PUMA will establish optimised Pu/MA transmutation characteristics in HTRs. Transient analyses will seek to demonstrate the nuclear stability and safety of the optimised reactor/fuel designs. Results shall also comprise measures on other performance and safety-related parameters, such as the fast fluence in the CP coatings and the helium production in HTR coated particles, as well as an assessment of proliferation resistance. Generally more than a single, but not more than three, partner(s) will be assigned to the analysis of a specific HTR system/fuel/fuel cycle combination, in order to overcome/alleviate some of the uncertainties associated with the present, limited qualification of the code systems. Each contributing partner will at least analyse, as a base case, one out of two reference systems, loaded with reference (Pu-based) fuel, and operated according to the reference fuel cycle scenario. This will provide a set of reference values of the calculated parameters (obtained with the partner-specific code system), which facilitates the intercomparison of results for different reactor/fuel/fuel cycle combinations, obtained by the respective partners. Reference systems for these studies are contemporary representatives of the two main HTR designs, viz. the PBMR-400 [6,7] for the pebble-bed (continuous reload) type, and the GT-MHR [8] for the prismatic block type. As an example, the main characteristics of the PBMR-400 model are listed in Table II. Simplifications in the basic models have been applied, as the focus of these studies is not on detail of the reactor and fuel reload system, but on the influence of the usage of Pu/MA fuel on reactor behaviour, as well as on the attainable transmutation performance of such a system. Studies include: · Pu/MA deep burn in pebble-bed V/HTR. This also includes investigations on the utilisation IMF, such as (Zr,Y,Am)O2/(Zr,Y,Ce)O2-x or (CeAm)O2-x; · Th/Pu fuel cycle in pebble-bed V/HTR; · Pu/MA deep burn in prismatic V/HTR; · Th/Pu fuel cycle in prismatic V/HTR; ICENES 2007 - International Conference on Emerging Nuclear Enegy Systems Istanbul, Turkey, June 3-8, 2007 · · Reactivity transients of Pu/MA fuelled V/HTR; Integrated LWR-HTR-GCFR symbiotic fuel cycles. Some preliminary results have been obtained on the use of (U-free) Pu-containing coated particle fuel in a contemporary design of the PBMR-400 pebbled-bed HTR [9]. This has also been addressed by other recent work [6], although mainly focusing on Pu destruction characteristics. In addition to presenting information on Puburning capabilities, the present paper also treats the time-dependent behaviour of a PuO2-loaded PBMR-400 during a Loss of Forced Cooling and a Depressurised Loss of Forced Cooling incident, hereby further addressing the requirement of safe operation. The modelling of the PBMR-400 is based upon the description of the OECD/NEA PBMR-400 transient benchmark [7]. In the mid-1990s the Pebble Bed Modular Reactor (PBMR) came to the fore-front as a possible option for the installation of new generating capacity by ESKOM, the electric utility of South Africa. The PBMR is a pebble-type HTR. This PBMR power plant incorporates a closed cycle primary coolant system utilising helium to transport heat energy directly from the modular pebble bed reactor to a recuperative Brayton cycle power conversion unit with a single-shaft turbine/compressor/generator. This replacement of the steam cycle that is common in present nuclear power plants (NPP) with a direct gas cycle provides the benefits of simplification and a substantial increase in overall system efficiency and inherent safety with the attendant lowering of capital and operational costs. The PBMR functions under a direct Brayton cycle with primary coolant helium flowing downward through the core and exiting at 900°C. The helium then enters the turbine relinquishing energy to drive the electric generator and compressors. After leaving the turbine, the helium then passes consecutively through the LP primary side of the recuperator, then the pre-cooler, the low pressure compressor, intercooler, high pressure compressor and then on to the HP secondary side of the recuperator before re-entering the reactor vessel at 500°C. The following are the most important characteristics of the PBMR reactor: · Annular core with an outer diameter of 3.7 m and a ‘fixed central reflector’ with an outer diameter of 2 m; · Graphite side reflector of ~90 cm with the coolant inlet and risers; · Reactivity Control System (RCS) consisting of 24 partial length control rod positions in the side reflector, with 12 upper or control rods and 12 lower or shutdown rods, when fully inserted. During normal operation all 24 rods operate together. The rods have an effective length (B4C neutron absorbing material) of 6.5 m; · Reserve Shutdown System (RSS) consisting of eight Small Absorber Sphere (SAS) systems positioned in the fixed central reflector and filled with 1 cm diameter absorber spheres containing B4C when required; · Three fuel loading positions and three fuel unloading tubes, positioned equidistant in the centre of the fuel annulus. The core contains ~ 452,000 fuel spheres or “pebbles” with a packing fraction of 0.61. In the standard PBMR400 design, the fuel is uranium at 9 g per fuel sphere with the 235U enrichment at 9.6 wt%. The inner 5 cm of the fuel sphere contains the TRISO-coated UO2 kernels within a graphite matrix and surrounded by an outer graphite fuel free shell. Each coated particle acts as a fission product barrier. For the application as Pu-burner, a different fuel composition was assumed. Each fresh fuel pebble contains 2 g first generation Pu in coated particles with a kernel diameter of 240 mm (see Table III). No burnable poison is envisaged. The fuelling scheme employed is the Table III continuous on-line multi-pass method similar Characteristics of Pu-containing PBMR fuel pebbles. to the design used in the German MEDUL reactor. Fresh fuel elements are added to the Diameter 6.0 cm top of the reactor while used fuel pebbles are Diameter of fuelled zone 5.0 cm removed at the bottom to keep the reactor at full power. On average, each fuel pebble Kernel diameter/material 0.24 mm / PuO2 makes about six passes through the reactor Fuel loading 2.0 g Putout initial per pebble before being finally discharged to the spent Pu vector 2.59/53.85/23.66/13.13/6.7 fuel storage tanks with a target burn-up of Pu238/239/240/241/242/ 8 wt% about 760,000 MWd/t. This target value originates from experimental evidence (Peach Bottom) that PuOx coated particle fuel can withstand irradiation up to these levels [10]. The fuel handling system consists of a core-unloading device in each of the three de-fuelling chutes from where the fuel is moved to the burn-up assaying equipment. After the burn-up has been determined, the fuel is routed either to the spent fuel tanks or back to the core, depending on its burn-up. The fuel spheres are reloaded to the top of the core through three fuelling lines. Although the reference design for the PBMR-400 benchmark problem, and also for the analyses presented here, is derived from the design described above, several simplifications in the specification were made in order to ICENES 2007 - International Conference on Emerging Nuclear Enegy Systems Istanbul, Turkey, June 3-8, 2007 (a) (b) (c) (d) Fig. 1. Maps of the PBMR-400 (Pu) reactor with the distribution of the Power Density (a), Coolant Temperature (b), Solid or Structure Temperature (c) and the local mesh averaged Burn-up (d). The white lines indicate the core region and the dashed line gives the top of the pebble bed. In Fig. (b) the coolant entry can be seen at below-right and the coolant exit at belowmiddle. Note that the maps are cylinder-symmetric, with symmetry-axis at R = 0. limit the need for any further approximations to a minimum. During this process care has been taken to ensure that all the important characteristics of the reactor design were preserved. This ensures that the results from the benchmark will be representative of the actual design’s characteristics. The simplifications make the core design essentially two-dimensional (r,z). It includes flattening of the pebble bed’s upper surface and the removal of the bottom cone and defuelling chutes that results in a flat bottom reflector. Pebble flow channels within the pebble bed have been simplified to be parallel and at equal speed. Control rods in the side reflector are modelled as a cylindrical skirt (also referred to as a grey curtain) with a given B-10 concentration. Thermal-hydraulic simplifications include the specification of stagnant helium (no mass flow) between the side reflector and barrel as well as the barrel and RPV. Stagnant air is defined between the RPV and reactor cavity cooling system (RCCS, outer boundary). The coolant flow is simplified to the main engineered flow paths, i.e. upwards flow from the inlet below the core within a porous ring in the side reflector and downwards flow through the pebble bed to the outlet plenum. No reflector cooling or leakage paths were defined. Other engineered coolant flows excluded are flows like the control rod cooling flow and the core barrel leakage flow The effect of excluding specific coolant flows is to some extent balanced by the assumption that all heat sources (from fission) will be deposited locally, i.e. in the fuel and that no other heat sources exist outside the core (for example neutron absorption in the control rods and further down gamma heating). Simplifications are also made in the material thermal properties in as far as constant values are employed or specific correlations are employed. To arrive at the equilibrium steady state, 1500 fuel balls, with 3000 g initial Pu, were circulated per day in the simulation. To speed up the convergence to equilibrium at least 125 fresh fuel balls were added on top of the bed. After about 600 days the first pebbles in the unloaded fuel at the bottom had reached the nominal burn-up of ICENES 2007 - International Conference on Emerging Nuclear Enegy Systems Istanbul, Turkey, June 3-8, 2007 760 MWd/kg and were discarded. This amount was replenished at the top by fresh balls as well. After about 8000 days the amount of discarded fuel at the bottom and added fresh fuel at the top levelled equal, leading to a stable non-changing core inventory. At this equilibrium state about 250 fresh pebbles per day were needed. It should be noted that this situation can be considered as equilibrium in the strictest sense, as the spatial distribution of all fuel-related nuclides has become time-independent. The shortest time frame in which this situation can be established is governed by the time required for the fuel to reach the envisaged discharge burnup of 760 MWd/kg. At an average rating of about 0.442 MW/kg, this would take about 1720 days. The actual time depends on the particular run-in scenario, and this may be considerably longer, because of oscillatory behaviour, as is the case in the example shown. The resulting map of the power density, gas coolant temperature, solid temperature (temperature of the core, reflectors and vessels) and the (mesh) averaged burn-up in the core is shown in Figs. 1 a-d. Note the peaking of the power density adjacent to the inner and outer reflectors. Some additional tasks in this Work Package concern the calculation/prediction of the production of He in HTR fuel and an assessment of the proliferation risk of Pu and MA fuel in V/HTRs. This assessment evaluates the proliferation risks relative to those posed by the use of plutonium and minor actinide fuels in LWRs. It is expected that this assessment will be valuable in identifying if there are any areas where new technological requirements may be required to maintain adequate safeguards with plutonium and minor actinide fuels. The study will include an assessment of the proliferation risk for the reactor plant and its associated fuel cycle operations. The assessment will encompass al the fuel types of interest, low enriched uranium, plutoniumuranium, plutonium-thorium and minor actinide fuels. The report will also review safeguards procedures for prismatic and pebble bed cores; for pebble bed cores a different approach to safeguards will be needed because of the impossibility of individually verifying the individual fuel spheres. Possible safeguard schemes will be suggested which will highlight areas where future research and development may be needed. V. OPTIMISATION OF PU/MA LOADED V/HTR FUEL DESIGN For transmutation purposes, ultra-high burn up plutonium (Pu) and Minor Actinides (MA)-based coated particle (CP) fuels will be needed and will require the development of novel fuel kernels capable of either being reprocessed or suitable for direct geological disposal. The manufacture of HTR fuels incorporating plutonium and minor actinides will be more difficult than for the corresponding pellet type fuels. There are advantages to be gained by the use of inert matrix based fuel kernel to dilute the transuranic components [11,12]. The JRC-ITU has developed a method based on the infiltration of porous precursor kernels, which overrides some of the kernel production difficulties and cost issues. The fabrication process can be simplified and the extra volume generated in the buffer layer (for a given Pu mass and buffer layer thickness) can accommodate more fission gas, and higher burn-ups can be achieved. The traditional TRISO layers used in U based fuel are taken for PUMA fuels. A constant discussion point in the study of transmutation fuels is the level of separation of the actinides at the reprocessing plants. Today, U and Pu are separated individually, with the minor actinides going for disposal in the vitrified waste. Generation IV takes the extreme position that all actinides (U, Pu, MAs) should be subjected to group separation. For proliferation resistance, this is a perfectly laudable approach, but not necessarily the most practical solution. For transmutation fuels today, the general consensus is to avoid U as a matrix, as it generates further higher actinides on irradiation. Thus, its presence automatically leads to a decrease in the transmutation rate. Fuels for other transmutation reactor concepts, e.g. the accelerator driven system (ADS), rely on the inert matrix concept. Even here there is a debate on the choice of Pu and MA separation in the reprocessing step. From the fabrication standpoint, it is preferred to separate the Pu, so that the bulk of the fuel can be manufactured in a facility without excessive shielding, just as is done today in MELOX or SMP MOX fuel fabrication plants in France and the UK, respectively. In PUMA, the U/Pu separation philosophy will be followed and the fuels for PUMA should be considered as heterogeneous and will be of the form · PuO2 (probably with NpO2) · MAO2 · (Pu,MA)O2, whereby the Pu:MA ratio is significantly lower than in spent fuel. In principle Cm should be included in the fuel, but Cm management must be considered by alternative means, namely Cm storage to permit its natural decay to Pu, and then later transmutation of the resulting Pu. ICENES 2007 - International Conference on Emerging Nuclear Enegy Systems Istanbul, Turkey, June 3-8, 2007 Table V Novel kernel characteristics. Matrix Zr0.85Y0.15O2 TD (g.cm-3) Mass in 500 µm kernel (g) (95%TD) 5,96 3,47E-4 (Zr0.85Y0.15)1-y CeyO2-x * CeO2/Ce2O3 Zr2Ce2O7 Y2O3 C (graphite) 7,132/6,86 4,15E-4/ 3,99E-4 6.173 3,59E-4 5,01 2.92 E-4 2,25 1,31E-4 * Depends on the Ce content selected One real advantage of the HTR is the low power density of the core. This results in an abundance of space, so that fuel design concepts are more open than for other reactor systems. Although the reference PuO2 kernel has a diameter of 200 µm, this is a poor choice, and is based largely on historical reasons, even though it is still considered today for the GT-MHR in Russia. The early studies of Belgonucleaire clearly identified the Table IV Diluting inert matrices for HTR actinide (An = Pu, Am, or Np) fuel. YSZ YSZ-Ce Ceria PYR (pyrochlore) Zr doped Y2O3 Graphite Matrix Zr0.85Y0.15O2 (Zr0.85Y0.15)1-yCeyO2 CeO2 Zr2Ce2O7 (Y,Zr)2O3 C Fuel/target (Zr0.85Y0.15 )1-zAnzO2 ((Zr0.85Y0.15)1-yCey)1-zAnzO2 Ce1-zAnzO2 (Zr2Ce2-zAnz)O7 (Y,Zr,Anz)2O3 C + AnO2 advantages in diluting the Pu in an inert matrix and the fuel produced by them at that time was one of the very first inert matrix fuels (IMF). In that concept, carbon was chosen as the diluting matrix. Carbon is not ruled out today, but not given highest priority, as the then used fabrication process, based on powder metallurgy, is poorly compatible with handling of minor actinides. A much better option is based on actinide infiltration into a porous precursor kernel. This minimises wastes and reduces the handling steps involving MAs. Some of the most promising materials are given in Table IV. Carbon is still under consideration, and could be used in the infiltration process, once a suitable means to produce suitable precursors is developed. The premise in designing the diluted, coated particle with any of these inert matrices relies on taking a fixed actinide mass in each kernel. This mass corresponds to the volume occupied by the corresponding 200 µm actinide oxide kernel. In this way the power in each kernel is kept at or below that of a PuO2 kernel. Sticking with tradition, justified by the excellent results obtained with NUKEM produced UO2 fuel, the diluted kernel size is kept at 500 µm. At this size, one can expect the in pile fuel performance, i.e. that of the sealing coating layers, to be similar to that of a UO2 kernel, operated at the same power/kernel. The characteristics of these diluted kernels for Pu and MA fuels are given in Table V. Loading of fuel elements homogeneous in (Pu,MA)O2 fuel could be handled in the same way as in a conventional UO2 fuelled HTR (pebble bed or prismatic). Ideally the fuel should be homogeneous at the level of the kernel, i.e. a genuine, but diluted (see Table V), (PuMA)O2 kernel. Given the impracticalities of fabricating group reprocessed transuranium kernels and the pollution of a single large-scale facility with minor actinides, it is preferable to maintain compositional heterogeneity at the coated particle level, i.e. the spherical fuel elements or cylindrical compacts should contain both diluted PuO2 as well as diluted MAO2 coated particles in the required proportions. An additional quality control (QC) step to control the Pu and MA content would be required, as well as the need to develop fabrication procedures for two kernel types. In this way, a quasi homogenous compacts or SFE is maintained. In the case of prismatic cores, the ratio of Pu versus MA containing kernels in a compact can be selected to obtain an optimised power distribution and temperature gradient within the core. Yttria stabilised zirconia (YSZ) and ceria have been tested in previous irradiation tests using pellet fuel. Ceria has an added chemical advantage, as it can be incorporated in the fuel in its trivalent (III) form so that oxygen released through fission of the actinides can be gettered via the Ce(III/IV) buffer. Belgonucleaire used graphite ICENES 2007 - International Conference on Emerging Nuclear Enegy Systems Istanbul, Turkey, June 3-8, 2007 as a diluting matrix. The most readily reprocessible and proven matrix remains UO2, which depending on the strategy could be envisaged as a PUMA fuel, but is not a first choice. Further activities within the PUMA fuel Work Package include the further development of fuel performance models appropriate for HTR-type coated particle fuel. These models focus on the prediction of stresses in coated particles and prediction of release fraction of volatile species present in the kernel. The PUMA Fuel Work Package integrates previous knowledge gained in GA and Belgonucleaire programmes with new investigations on the optimal kernel composition. The programme makes use of Belgonucleaire archive PuO2 coated particles to investigate the behaviour of helium in coated particle fuel. In contrast to UO2 fuels, this is an issue of great importance for MA bearing fuels and must be incorporated in the fuel performance models being used to determine the optimal coated particle geometry. Finally, the design and safety study for a PUMA irradiation is being made. Independently of PUMA, the JRC-ITU is installing a Pu coated particle production facility, and it is planned that PUMA fuels will be irradiated in a subsequent EU FP7 programme. VI. ROLE FOR V/HTR SYSTEMS WITH A TRU MANAGEMENT FUNCTION IN EUROPE’S FUTURE NUCLEAR PARK HTR reactors have a potential in a mixed nuclear park, where fast reactors would be deployed to gain fuel sustainability, whereas HTRs could provide the means to transmute the MAs. A possible fuel cycle strategy is outlined in Fig. 2. Current day light water reactor (LWR) UO2 fuel is reprocessed with some of the Pu returning to the LWR in the form of MOX. Remaining Pu not used for LWR MOX can be fabricated into HTR fuel. In addition, the Pu extracted from LWR-MOX spent fuel can also be fabricated in the form of HTR fuel. Minor actinides from the LWR UO2 and MOX fuels will be separated and be processed for irradiation in the HTR. Irradiated HTR fuel can be considered for direct disposal or even reprocessing with the Pu and MA constituents Fabrication LWR – UO2 Fuel Fabrication Reactor LWR Disposal Partitioning Enhanced reprocessing UO2 Fuel MOX Fuel fabrication FP Pu, Np, Am, Cm Pu Pu, Np, Am, Cm HTR Fuel Fabrication PUMA HTR Reprocessing FP Geological Disposal Fig. 2. Transmutation Strategy with a PUMA HTR. being once again prepared for HTR irradiation. Other similar scenarios based on Gen IV fast reactors rather than LWRs can also be envisaged. In either case dedicated transmutation HTRs can be considered as a part of the nuclear park. The assessment of the role of V/HTRs delivering energy products (i.e. electricity, hydrogen and process heat in general), especially when fulfilling a TRU-management role, in a time-evolving European (nuclear) energy park is the prime topic addressed within the third Work Package of the PUMA-project. While the other PUMA Work Packages essentially address the scientific and technical aspects of reactors and fuels respectively, this WP aims at providing a holistic assessment of the potential role(s) of V/HTRs by focusing on four main facets, i.e.: · The technological impact of V/HTRs fulfilling a TRU-management function on the whole nuclear energy park and especially on the nuclear fuel cycle. Typical questions addressed are the technological feasibility and expected performance of the fuel cycle operations needed to accomplish the TRU-management mission ICENES 2007 - International Conference on Emerging Nuclear Enegy Systems Istanbul, Turkey, June 3-8, 2007 with V/HTRs as well as the synergy between V/HTRs with other reactors (LWRs, FRs) within Europe’s nuclear reactor park; · The economics and the potential market penetration for such V/HTRs and especially the economic impact on V/HTRs when performing a TRU-management role in future nuclear energy systems; · The environmental facet looks into a life cycle inventory analysis for (V)HTRs and especially to the fuel cycle aspects such as (secondary) waste arising, various waste management options, separated fissile material inventories and losses in the nuclear fuel cycle, as well as into non-nuclear material streams; · The socio-political facet focuses essentially on the proliferation risk of nuclear energy systems involving V/HTRs with TRU-management role. The outcome of this Work Package is a characterisation and uncertainty/sensitivity analysis of the technical feasibility, economic viability and environmental friendliness of various V/HTR-designs, as defined by the other PUMA Work Packages. The assessment of the potential of V/HTRs in their TRU-management function in future nuclear reactor parks will be based on internationally accepted assessment methodologies as has been recently proposed by the Generation-IV International Forum (GIF) activities [13] and IAEA’s INPRO’s Assessment Methodology [14] as well as practiced within the EU FP6 project RedImpact [15], among others. Specifically, a number of transmutation performance indicators will be established to enable comparison of different recycling scenarios and V/HTR variants as well as with other non-V/HTR scenarios based on literature review. The activities in this Work Package are essentially grouped into three major tasks with some specific sub-tasks being described in the following. Characterisation of (V)HTR and associated fuel cycle and waste management technologies with regard to the four facets described above. An assessment of the role that V/HTRs may play in a European nuclear reactor park needs to cover the economic, environmental and socio-political dimensions in addition to the pure technical feasibility of such reactors. This first task aims at specifically addressing these sustainability dimensions based on the technical information provided by the other PUMA Work Packages and collected from EC and other international projects on gas-cooled reactors with the main outcome being an as complete as possible description of the reactor and fuel cycle technologies as well as the technological challenges and limitations that may be faced in deploying such V/HTR-scenarios. Simulation of (European) nuclear energy systems providing a holistic assessment framework. Based on some (nuclear) energy demand scenarios from authoritative studies (IIASA, WEC, EC, ...), the deployment of various nuclear energy systems incorporating V/HTRs will be analysed. As both gas-cooled reactors may play a specific and important role in the sustainability character of nuclear energy, and especially also in the TRU-management for the whole nuclear energy system, a dynamic analysis of the evolution from today’s European reactor park to some future reactor parks will be analysed. Such dynamic analysis allows proper assessment of the mass-flows, waste arising, separated TRU inventories, the delay times in deployment, the economic impacts and so forth, e.g.: The timing of introduction of V/HTRs according to different fuel cycle options is a function of the evolution of the European reactor park and fuel cycle infrastructure and will therefore have to cope with different stocks of spent fuel and (separated) TRUs. The timing of introduction of V/HTRs shall therefore impact on the evolution of spent fuel, TRU-stock and other indicators as well as define the kind of (V)HTR core management needed in order to reduce such spent fuel and TRU-stock arising. The V/HTR fuel cycle infrastructure technologies (losses, transit times, …) will define the working inventories of TRUs in the fuel cycle and thus also the pace of V/HTR introduction in the European reactor park. Various dynamic simulation codes will be used, i.e. OSIRIS, ORION and DANESS and maybe others depending on availability. A benchmark exercise will also be undertaken to verify the mass-flow, economic and waste management simulations provided by these codes. An uncertainty/sensitivity analysis resulting in a mapping of the potential of V/HTR in future nuclear reactor parks and identification of the main drivers enabling such future role for V/HTRs with focus on: Identifying and analysing the prime important technological objectives (and associated uncertainties) to be achieved. This involves an iterative process of information exchange and scenario analysis based on new information gained from other Work Packages of PUMA The uncertainty/sensitivity analysis will be covering aspects of: · Fuel and core management in V/HTRs, i.e. burn-up, core-management, fuel type, cycle length, etc. · Fuel cycle: cooling times, reprocessing and fabrication parameters, waste management options, etc. · Economic parameters and especially the impact of technological choices induced by fuel/core options on expected capital, O&M and fuel cycle costs. The nuclear energy system scenarios investigating the potential future roles for (V)HTRs, i.e. the scenario families being considered are the following: · European LWR-park without (V)HTRs as reference scenario (first family of scenarios). Starting from the existing reactor park in Europe (i.e. taking into account the anticipated shutdown of reactors after ICENES 2007 - International Conference on Emerging Nuclear Enegy Systems Istanbul, Turkey, June 3-8, 2007 approximately 50 years lifetime), new Generation-III LWRs are introduced with a continuation of OTC and mono-Pu recycling as fuel cycle option; · European LWR-park without (V)HTRs with a TRU-management objective based on MOX-EU fuel cycle is studied in the second family of scenarios. This scenario family intends to show the ‘maximum’ achievable TRU-management capability of a LWR based reactor park allowing then to compare with a LWR+(V)HTR reactor park where the (V)HTRs serve a TRU-management role; · The third scenario family involves a mixed LWR and (V)HTR reactor park with the (V)HTRs using uranium-fuel in OTC-mode, i.e. no TRU-management role. Various reactor park variants with changing LWR and (V)HTR fractions shall be investigated based on energy market potential; · The fourth scenario family is comparable to the previous scenario except that (V)HTRs serves a TRUmanagement role. Different variants are considered, i.e.: 1. (V)HTRs in Pu deep-burn mode with the Pu coming from a. first or second generation UOX/MOX from the LWRs and/or; b. from recycled CP-fuels from (V)HTRs. 2. (V)HTRs in TRU deep-burn mode with the TRUs coming from a. First or second generation UOX/MOX from the LWRs and/or; b. From recycle CP-fuels from (V)HTRs. · The fifth family of scenarios involves mixed LWR+(V)HTR+(S)FR reactor parks intended to show the potential synergies between the reactor types and highlighting the potential role for (V)HTRs as well as comparison of (V)HTRs with (S)FRs with respect to TRU-management The Scenario Work Package of PUMA will also aim at integrating as much as possible the results and experience gained in previous or still ongoing EC projects such as RAPHAEL, RedImpact, GCFR as well as from other international projects especially with the objective to integrate best practices developed worldwide in performing this holistic assessment of V/HTRs futures. VII. CONCLUSION Summarizing, the PUMA project’s main goal is to provide additional key elements for the utilisation and transmutation of plutonium and minor actinides in contemporary and future (high temperature) gas-cooled reactor designs, which are promising tools for improving the sustainability of the nuclear fuel cycle, hereby also contributing to the reduction of Pu and MA stockpiles, and to the development of safe and sustainable reactors for CO2-free energy generation. The investigation on core physics aims at optimising the CP fuel and reactor characteristics, and assuring nuclear stability and safety of a Pu/MA V/HTR core. New CP designs will be explored in order to withstand very high burn-ups and obtain optimal adaptation for disposal after irradiation. In particular, helium production in Pu- and MA-based fuel will be assessed and supported by experiment. Fuel irradiation performance codes, developed and used by several organisations, will permit convergence on optimised design criteria. The impact of the introduction of Pu/MA-burning (V)HTRs on the fuel cycle and future nuclear energy mix will be assessed, with focus on the fuel cycle symbiosis with future nuclear energy systems in Europe (LWRs, Fast Reactors, ADS). This assessment involves the quantification of waste streams and radiotoxic inventories as well as the technical, economic, environmental and socio-political impacts of introducing (V)HTRs with a TRUmanagement mission in a future nuclear park. PUMA also contributes to technological goals of the Generation IV International Forum. It contributes to developing and maintaining the competence in reactor technology in the EU, and addresses European stakeholders on key issues for the future of nuclear energy in the EU. ACKNOWLEDGMENT The EU FP6 Project PUMA is partly financed by the Commission of the European Communities, EURATOM 6th Framework Program, contract no. 036457, signed October 3, 2006. REFERENCES 1. M.A. Fütterer, “RAPHAEL: The European Union’s (Very) High Temperature Reactor Technology Project”, Proc. ICAPP’06, Reno, NV, USA, June 4-8, 2006. 2. “PuMA - Plutonium and Minor Actinides Management by Gas-cooled Reactors”, European Union Sixth Framework Program contract no. 036475, signed October 3, 2006. 3. “Evaluation Guide for the International Reactor Physics Experiments Evaluation Project (IRPhEP)”, Document NEA/NSC/DOC(2006)2, Revision 8.9, OECD Nuclear Energy Agency, January 20, 2006. 4. “HTR-N - High-Temperature Reactor [Nuclear] Physics [, Waste] and Fuel Cycle Studies”, European Union Fifth Framework Program contracts no. 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