Fusion Breeder Research Articles

Application of discrete element method to study mechanical behaviors of ceramic breeder pebble beds

In this paper, the discrete element method (DEM) approach has been applied to study mechanical behaviors of ceramic breeder pebble beds. Directly simulating the contact state of each individual particle by the physically based interaction laws, the DEM numerical program is capable of predicting the mechanical behaviors of non-standard packing structures. The program can also provide the data to trace the evolution of contact characteristics and forces as deformation proceeds, as well as the particle movement when the pebble bed is subjected to external loadings. Our numerical simulations focus on predicting the mechanical behaviors of ceramic breeder pebble beds, which include typical fusion breeder materials in solid breeder blankets. Current numerical results clearly show that the packing density and the bed geometry can have an impact on the mechanical stiffness of the pebble beds. Statistical data show that the contact forces are highly related to the contact status of the pebbles.

06/00630 Radiation damage studies on the first wall of a HYLIFE-II type fusion breeder: Sahin, S. and Ubeyh, M. Energy Conversion and Management, 2005, 46, (20), 3185–3201.

06/00630 Radiation damage studies on the first wall of a HYLIFE-II type fusion breeder: Sahin, S. and Ubeyh, M. Energy Conversion and Management, 2005, 46, (20), 3185–3201.

Radiation damage studies on the first wall of a HYLIFE-II type fusion breeder

Abstract The radiation damage on the first wall [made of (1) a ferritic steel (9Cr–2WVTa), (2) a vanadium alloy (V–4Cr–4Ti) and (3) SiCf/SiC composite] of an inertial fusion energy (IFE) reactor of HYLIFE-II type is investigated. A protective liquid wall with variable thickness, containing Flibe + heavy metal salt (UF4 or ThF4) is used for first wall protection. The content of heavy metal salt is chosen as 4 and 12 mol%. Neutron transport calculations are performed with the aid of the SCALE4.3 System by solving the Boltzmann transport equation with the XSDRNPM code in 238 energy groups and S8–P3 approximation. A flowing wall with a thickness of ∼60 cm can extend the lifetime of the solid first wall structure to a plant lifetime of 30 years for 9Cr–2WVTa and V–4Cr–4Ti, whereas the SiCf/SiC composite as first wall needs a flowing wall with a thickness of ∼85 cm to maintain the radiation damage limit. Substantial extra revenue can be gained through the insertion of a heavy metal salt constituent into Flibe, which allows breeding fissile fuel for external reactors and increasing energy multiplication (233U with a value of up to 1,000,000,000 $/year or 239Pu for few 100,000,000 $/year).

04/02776 Modified APEX reactor as a fusion breeder: Sahin, S. and Übeyli, M. Energy Conversion and Management, 2004, 45, (9–10), 1497–1512

04/02776 Modified APEX reactor as a fusion breeder: Sahin, S. and Übeyli, M. Energy Conversion and Management, 2004, 45, (9–10), 1497–1512

Neutronic analysis of denaturing plutonium in a thorium fusion breeder and power flattening

The purpose of this study is to denature nuclear weapon grade quality plutonium in a thorium fusion breeder. Ten fuel rods containing the mixture of ThO 2 and PuO 2 are placed in a radial direction in the fissile zone where ThO 2 is mixed with variable amounts of PuO 2 to obtain a quasi-constant nuclear heat production density. The plutonium composition volume fractions in the fuel rods are gradually increased from 0.1% to 1% by 0.1% increments. The fissile fuel zone is cooled with four various coolants with a volume fraction ratio of 1 ( V coolant / V fuel = 1). These coolants are helium gas, flibe “Li 2 BeF 4 ”, natural lithium and eutectic lithium “Li 17 Pb 83 ”. Nuclear weapon grade quality 239 Pu in the fuel composition is denatured due to the accumulation of the 240 Pu isotope in the fissile zone after 18 months of plant operations. Under a first wall fusion neutron current load of 2.222 × 10 14 (14.1 MeV n/cm 2 s), which corresponds to 5 MW/m 2 , by a plant factor of 100%, at the end of the plant operation, the fissile fuel enrichment quality between 6.0% and 10% is obtained depending on the coolant types. During the plant operation, the tritium breeding ratio (TBR) should be at least 1.05. In the selected blanket, only the flibe coolant is already self sustaining at start up. The TBR increases steadily due to the higher neutron multiplication rate during the plant operation period. The highest TBR is obtained for the eutectic lithium coolant 1.4035, followed by the flibe coolant 1.3095, helium gas coolant 1.2172 and natural lithium coolant 1.0553 at the end of the operation period of 48 months. The energy multiplication factor M changed between 2.1731 and 6.6241 depending on coolant type during the operation period. The peak to average fission power density ratio Γ in the blanket decreases by ∼15%, which allows a more uniform power generation in the fissile zone. The isotopic percentage of 240 Pu reaches higher than 5% in all coolant types. This is very important for international safety.

Modified APEX reactor as a fusion breeder

An advanced fusion reactor project, called APEX, with improved effectiveness has been developed using a protective flowing liquid wall for tritium breeding and energy transfer. In the modified APEX concept, the flowing molten salt wall is composed of Flibe as the main constituent with increased mole fractions of heavy metal salt (ThF 4 or UF 4) for both fissile and fusile breeding purposes and to increase the energy multiplication. Neutron transport calculations are conducted with the help of the SCALE4.3 SYSTEM by solving the Boltzmann transport equation with the code XSDRNPM. By preserving a self sufficient tritium breeding ratio (TBR > 1.05) for a mole fraction up to 6% of ThF 4 or 12% of UF 4, the modified APEX reactor can produce up to ∼2800 kg of 233U/year or ∼4950 kg of 239Pu/year, assuming the same baseline fusion power production of 4000 MW th, as in the original APEX concept. With 6% ThF 4 or 12% UF 4 in the coolant, the total energy output will increase to 5560 MW th or 8440 MW th, respectively. For a plant operation period of 30 full power years, the atomic displacement and helium production rates remain well below the presumable limits. The additional benefits of fissionable metal salt in the flowing liquid in a fusion reactor can be summarized as breeding of high quality fissile fuel for external reactors and increase of total plant power output.

Heat resistant reduced activation 12% Cr steel of 16Cr12W2VTaB type-advanced structural material for fusion and fast breeder power reactors

Heat resistant 12% Cr steels of the 16Cr12W2VTaB type (12Cr-2W-V-Ta-B-0.16C) provide a reduced activation material that can be used as a structural material for fusion and fast breeder reactors. The composition under study meets scientific and engineering requirements and has an optimal base element composition to provide a δ-ferrite content of no more than 20%. It also has a minimum quantity of low melting impurity elements and non-metallic inclusions. Short-term tensile properties for the steel tested to 700 °C are provided after the standard heat treatment (normalization, temper). Rupture strength and creep properties for the steel depending on the initial heat treatment conditions are also given. The microstructural stability of the 16Cr12W2VTaB type steel at temperatures up to 650 °C is predicted to be good, and the properties of the steel after irradiation in BOR-60 are demonstrated.

Strategy of fusion reactor materials R&D in China

There are two phases of fusion reactor materials R&D in China. The first phase will focus on fusion materials R&D for fusion breeders (or breeders for short) and in the second phase fusion materials for commercial pure fusion reactors will be developed. China is now in the first phase. Now development of reduced activation oxide dispersion strengthened ferritic steels, low- and high- Z plasma facing materials, including carbon base materials, Be, advanced W-alloys and functionally graded material, will be emphasized. Limited research activities for the second phase, e.g. V–Cr–Ti alloys and SiC/SiC composite development, are also supported. Fusion reactor materials R&D status in China is introduced and some research results are also given in present paper.

New evaluation of displacement damage and gas production for breeder ceramics under IFMIF, fusion and fission neutron irradiation

Design and development of accelerator driven neutron sources, in particular the International Fusion Materials Irradiation Facility (IFMIF), demands neutron cross-sections with extended energy range up to 50 MeV, as standard data libraries are truncated at 20 MeV. Based on recently released new cross-section data for lithium isotopes, damage and gas production rates were re-evaluated for the lithium-based breeder ceramics Li 4SiO 4 and LiAlO 2 under several neutron irradiation conditions. Calculated irradiation responses for mixed spectrum neutrons (HFR-Petten), fusion neutrons (DEMO reactor) and the accelerator-based d-Li neutron source IFMIF are presented. This study indicates, that IFMIF provides not only favorable irradiation conditions for structural materials but also can be a suitable test bed for fusion breeder materials. In the latter case an appropriate neutron moderator/reflector should be used to adjust the low energy tail of the IFMIF spectrum to that one of a typical DEMO reactor blanket.

Neutronics Experiment of 6Li-enriched Breeding Blanket with Li2TiO3/Be/F82H Assembly Using D-T Neutrons

To improve the nuclear performance of fusion breeding blankets, the application of thermal type blankets with 6Li-enriched ceramics and low-activation ferritic steel is suggested. However, appropriate neutronics experiments to investigate the nuclear performance of such blankets have never been carried out. Therefore, we have done blanket neutronics experiments with stratified 95-% enriched 6Li2TiO3, stainless steel F82H and beryllium blocks using JAERI’s Fusion Neutronics Source (FNS) and verified the prediction accuracy of the tritium production rate and the production rate of selected isotopes in F82H in with the Monte Carlo code MCNP-4B and JENDL Fusion File, JENDL-3.2 and ENDF/B-VI data. Calculated tritium and 51Cr production rates were in good agreement with the measured values within the experimental error. However, discrepancies were found for 187W and 56Mn production rates.

Power flattening and minor actinide burning in a thorium fusion breeder

A neutronic analysis has been performed for a thorium fusion breeder with a special task of burning minor actinides 237 Np , 241 Am , 243 Am and 244 Cm and production of 233 U , 238 Pu , 242 m Am and 245 Cm for spacecraft application. 238 Pu is an important radioisotopic energy source for spacecraft generators. As potential nuclear fuels in the foreseeable future, 233 U , 242 m Am and 245 Cm would allow one to build extremely compact space reactors. Natural lithium has been selected as the coolant medium for the nuclear heat transfer out of the fuel zone. Minor actinides out of 5 and 10 units of LWRs per metre of blanket height have been mixed with ThO 2. Higher fission rates in minor actinides enables one to realise a power flattening in the fissile zone over three years of plant operation by a gradual increase in the radial direction at start-up. This has significant advantages with respect to plant operation over the long term and also with respect to a uniform utilisation of the nuclear fuel in the fissile zone. After three years of plant operation, the net 233 U production is ∼300 kg per metre of blanket height. The 238 Pu yield is 21 and 41 kg for a waste actinide charge out of 5 and 10 units of LWRs per metre of blanket height, respectively, and the 245 Cm yield is 1.1 and 2 kg, respectively. The net 242 m Am production is practically nil. With waste actinides out of 10 reactor units per metre of blanket height, the flattening of the nuclear heat production density in the fissile zone is almost perfect. Waste actinides out of five reactor units per metre of blanket height allow still an excellent power flattening. The quasi-constant power shape is saved over 36 months.

Investigation of the effects of the resonance absorption in a fusion breeder blanket

In generic neutronic studies of a fusion breeder, resonance absorption is mostly neglected. In this work, the effects of resonances on the neutronic parameters a fast and moderated fusion blanket are investigated. In the present case, a ( D, T) fusion reactor acts as an external high energetic (14.1 MeV) neutron source. The fissile fuel zone, containing 10 rows in radial direction, covers the cylindrical fusion plasma chamber. The fissile fuel is natural UO 2. Fissile zone is cooled (a) with pressurised helium gas for the fast blanket and (b) with light water for the moderated, each of them with a volume ratio of V coolant/ V fuel=2 in the fissile zone. The study has shown that careful resonance self-shielding calculations are indispensable for neutronic studies of a moderated fusion blanket. On the other hand the omission of the resonance self-shielding in generic studies of a fast fusion blanket can be tolerated to some degree. Furthermore, a correct description of the fusion neutron source spectrum has great importance on neutronic parameters, along with the resonance self-shielding calculations.

Cold trapping of traces of tritiated water from the helium loops of a fusion breeder blanket

The ITER Helium Cooled Pebble Bed (HCPB) Test Blanket Module (TBM) will comprise three helium loops designed for: tritium extraction from the breeder zone, heat removal, and purification of the coolant. The process step envisaged for tritium extraction as well as for coolant purification includes a cryogenic cold trap as main component for the removal of tritiated water vapour (mainly HTO, H 2O). The concentrations of water in the gas streams are expected to be extremely small, i.e. of the order of 10 ppm by volume. In this paper, we describe first runs with a cold trap using helium as the carrier gas at flow rates of 0.1 and 1.0 m 3/h. The range of water vapour concentration in the helium carrier gas was 0.5 to >200 ppm v. The experiments have demonstrated the ability of the cold trap to remove water vapour efficiently from the He stream down to concentrations of less than 0.02 ppm v when the inlet water concentration is in the range of 300–650 ppm v or higher.

Neutronic performance of proliferation hardened thorium fusion breeders

Production of denaturated fissile fuel in a thorium fusion breeder has been investigated by mixing the fertile fuel with natural-UO 2 and LWR (light water reactors) spent nuclear fuel. Four different coolants (pressurised helium, Flibe ‘Li 2BeF 4’, natural lithium and Li 17Pb 83 eutectic) are selected for the nuclear heat transfer. In order to obtain a power flattening in the fissile-fertile zone, the UO 2- or the spent fuel-fraction in the mixed oxide (MOX) fuel has been gradually increased in radial direction. Power plant operation periods between 13 and 29 months are evaluated to achieve a fissile fuel enrichment ∼4%, under a first-wall fusion neutron energy current of 5 MW m −2 (plant factor of 100%). For a plant operation over 4 years, enrichment grades between 6.0 and 11.5% are calculated for the investigated MOX fuel and coolant compositions. The 238U component of natural-UO 2 can provide a limited proliferation hardening only for the 233U component, whereas, the homogenous mixture of ThO 2 with a small quantity of (>4%) LWR spent nuclear fuel remains all-over non-prolific for both, uranium and plutonium components.

Self-stabilizing criticality waves

Production of denaturated fissile fuel in a thorium fusion breeder has been investigated by mixing the fertile fuel with natural-UO2 and LWR (light water reactors) spent nuclear fuel. Four different coolants (pressurised helium, Flibe ‘Li2BeF4’, natural lithium and Li17Pb83 eutectic) are selected for the nuclear heat transfer. In order to obtain a power flattening in the fissile-fertile zone, the UO2- or the spent fuel-fraction in the mixed oxide (MOX) fuel has been gradually increased in radial direction. Power plant operation periods between 13 and 29 months are evaluated to achieve a fissile fuel enrichment ∼4%, under a first-wall fusion neutron energy current of 5 MW m−2 (plant factor of 100%). For a plant operation over 4 years, enrichment grades between 6.0 and 11.5% are calculated for the investigated MOX fuel and coolant compositions. The 238U component of natural-UO2 can provide a limited proliferation hardening only for the 233U component, whereas, the homogenous mixture of ThO2 with a small quantity of (>4%) LWR spent nuclear fuel remains all-over non-prolific for both, uranium and plutonium components.

Proliferation hardening and power flattening of a thorium fusion breeder with triple mixed oxide fuel

The proliferation hardening of the 233U fuel in a thorium fusion breeder has been realised successfully with a homogenous mixture of ThO 2, natural-UO 2 and CANDU spent nuclear fuel in the form of a triple mixed oxide (TMOX) fuel. The new 233U component will be successfully hardened against proliferation with the help of the 238U component in the natural-UO 2 and spent fuel. The plutonium component remains non-prolific through the presence of the 240Pu isotope in the spent CANDU fuel due to its high spontaneous fission rate. A (D,T) fusion reactor acts as an external high energetic (14.1 MeV) neutron source. The fissile fuel zone, containing 10 fuel rod rows in the radial direction, covers the cylindrical fusion plasma chamber. A quasi-constant power density in the fissile zone has been achieved by reducing the ThO 2 component in the rods continuously in the radial direction (from 91 down to 64%). Three different coolants (pressurised helium, natural lithium and Li 17Pb 83 eutectic) are selected for the nuclear heat transfer out of the fissile fuel breeding zone with a volume ratio of V coolant V fuel =1 in the fissile zone. The fissile fuel breeding occurs through the neutron capture reaction in the 232Th (ThO 2), in the 238U (natural-UO 2 and CANDU spent fuel) isotopes. The fusion breeder increases the nuclear quality of the spent fuel, which can be defined with the help of the cumulative fissile fuel enrichment (CFFE) grade of the nuclear fuel calculated as the sum of the isotopic ratios of all fissile materials ( 233U+ 235U+ 239Pu+ 241Pu) in the TMOX fuel. Under a first-wall fusion neutron current load of 10 14 (14.1 MeV n/cm 2 s), corresponding to 2.25 MW/m 2 and by a plant factor of 100%, the TMOX fuel can achieve an enrichment degree of ∼1% after ∼12–15 months. A longer irradiation period (∼ 30 months) increases the fissile fuel enrichment levels of the TMOX towards much higher degrees (∼ 2%), opening new possibilities for utilisation in advanced CANDU thorium breeders. The selected TMOX fuel remains non-prolific over the entire period for both uranium and plutonium components. This is an important factor with regard to international safeguarding.

Comparison of nuclear irradiation parameters of fusion breeder materials in high flux fission test reactors and a fusion power demonstration reactor

Nuclear irradiation parameters relevant to displacement damage and burn-up of the breeder materials Li 2O, Li 4SiO 4 and Li 2TiO 3 have been evaluated and compared for a fusion power demonstration reactor and the high flux fission test reactor (HFR), Petten, the advanced test reactor (ATR, INEL) and the Japanese material test reactor (JMTR, JAERI). Based on detailed nuclear reactor calculations with the MCNP Monte Carlo code and binary collision approximation (BCA) computer simulations of the displacement damage in the polyatomic lattices with MARLOWE, it has been investigated how well the considered HFRs can meet the requirements for a fusion power reactor relevant irradiation. It is shown that a breeder material irradiation in these fission test reactors is well suited in this regard when the neutron spectrum is well tailored and the 6Li-enrichment is properly chosen. Requirements for the relevant nuclear irradiation parameters such as the displacement damage accumulation, the lithium burn-up and the damage production function W( T) can be met when taking into account these prerequisites. Irradiation times in the order of 2–3 full power years are necessary for the HFR to achieve the peak values of the considered fusion power Demo reactor blanket with regard to the burn-up and, at the same time, the dpa accumulation.

00/01176 Waste petroleum oils. Processing and legal and analytical problems

The proliferation hardening of the 233U fuel in a thorium fusion breeder has been realised successfully with a homogenous mixture of ThO2, natural-UO2 and CANDU spent nuclear fuel in the form of a triple mixed oxide (TMOX) fuel. The new 233U component will be successfully hardened against proliferation with the help of the 238U component in the natural-UO2 and spent fuel. The plutonium component remains non-prolific through the presence of the 240Pu isotope in the spent CANDU fuel due to its high spontaneous fission rate. A (D,T) fusion reactor acts as an external high energetic (14.1 MeV) neutron source. The fissile fuel zone, containing 10 fuel rod rows in the radial direction, covers the cylindrical fusion plasma chamber. A quasi-constant power density in the fissile zone has been achieved by reducing the ThO2 component in the rods continuously in the radial direction (from 91 down to 64%). Three different coolants (pressurised helium, natural lithium and Li17Pb83 eutectic) are selected for the nuclear heat transfer out of the fissile fuel breeding zone with a volume ratio of VcoolantVfuel=1 in the fissile zone. The fissile fuel breeding occurs through the neutron capture reaction in the 232Th (ThO2), in the 238U (natural-UO2 and CANDU spent fuel) isotopes. The fusion breeder increases the nuclear quality of the spent fuel, which can be defined with the help of the cumulative fissile fuel enrichment (CFFE) grade of the nuclear fuel calculated as the sum of the isotopic ratios of all fissile materials (233U+235U+239Pu+241Pu) in the TMOX fuel. Under a first-wall fusion neutron current load of 1014 (14.1 MeV n/cm2 s), corresponding to 2.25 MW/m2 and by a plant factor of 100%, the TMOX fuel can achieve an enrichment degree of ∼1% after ∼12–15 months. A longer irradiation period (∼ 30 months) increases the fissile fuel enrichment levels of the TMOX towards much higher degrees (∼ 2%), opening new possibilities for utilisation in advanced CANDU thorium breeders. The selected TMOX fuel remains non-prolific over the entire period for both uranium and plutonium components. This is an important factor with regard to international safeguarding.

Spent mixed oxide fuel rejuvenation in fusion breeders

A fusion breeder is presented for the rejuvenation of spent nuclear fuel. A (D, T) fusion reactor acts as an external high energetic (14.1 MeV) neutron source. The fissile fuel zone, containing ten rows in radial direction, covers the cylindrical fusion plasma chamber. The first three fuel rod rows contain Canadian deuterium uranium (CANDU) reactor spent nuclear fuel which was used down to a total enrichment grade of 0.418%. The following seven fuel rod rows contain light water reactor (LWR) spent nuclear fuel, which was used down to a total enrichment grade of 2.17%. This allows a certain degree of fission power flattening. Fissile zone is cooled with pressurised helium gas with volume ration of V coolant/ V fuel=2 in the fissile zone. Spent fuel rejuvenation occurs through the neutron capture reaction in 238U. The new fissile material increases the nuclear quality of the spent fuel which can be described as the cumulative fissile fuel enrichment (CFFE) grade of the nuclear fuel which is the sum of the isotopic ratios of all fissile material ( 235U+ 239Pu+ 241Pu) in the mixed oxide (MOX) fuel. Under a first-wall fusion neutron current load of 10 14 (14.1-MeV n/cm 2 s), corresponding to 2.25 MW/m 2 and by a plant factor of 100%, the CANDU spent fuel can achieve an enrichment degree of 1% after ∼7 months, suitable for reutilization in a CANDU reactor. LWR spent fuel requires >15 months to reach an enrichment grade ∼3.5%, suitable for reutilization in a LWR. A longer rejuvenation period (up to 48 months) increases the fissile fuel enrichment levels of the spent fuel reactor to much higher degrees (>3% for CANDU spent fuel and over 5% for LWR spent fuel), opening possibilities an increased burn-up in critical reactors and a re-utilization in multiple cycles.

Chemical interactions in the EXOTIC-7 experiment

Post-irradiation studies by X-ray microanalysis are reported on selected samples of the MANET steel lined capsule 28.2 of the EXOTIC-7 experiment within the fusion breeder blanket materials irradiation programme. A mixture of Be pebbles and SiO 2 doped Li 4SiO 4 pebbles with 52% 6Li enrichment has been irradiated in the HFR up to 18% burn-up. A Li/Si decrease in the breeder material is observed due to Li transmutation which results in higher amounts of Li 2SiO 3 in the Li 4SiO 4 pebbles. Furthermore, oxygen is released which oxidises the surface of the Be pebbles to BeO. Direct BeO–Li 2SiO 3 contact leads to the formation of Li 2BeSiO 4. The predominant Li 2O vapour species is transported to the colder, inner liner surface of the pebble bed and recondenses as stratified Li 2O and Li 8SiO 6 layers. The chemical reactions between the MANET steel liner and the breeder materials play a minor role.