Massimiliano Fratoni

CHAMBER DESIGN FOR THE LASER INERTIAL FUSION ENERGY (LIFE) ENGINE

Abstract The Laser Inertial Fusion Energy (LIFE) concept is being designed to operate as either a pure fusion or hybrid fusion-fission system. The present work focuses on the pure fusion option. A key component of a LIFE engine is the fusion chamber subsystem. It must absorb the fusion energy, produce fusion fuel to replace that burned in previous targets, and enable both target and laser beam transport to the ignition point. The chamber system also must mitigate target emissions, including ions, x-rays and neutrons and reset itself to enable operation at 10-15 Hz. Finally, the chamber must offer a high level of availability, which implies both a reasonable lifetime and the ability to rapidly replace damaged components. An integrated design that meets all of these requirements is described herein.

Neutronic Feasibility Assessment of Liquid Salt―Cooled Pebble Bed Reactors

Abstract This study investigates the neutronic characteristics of the Pebble Bed-Advanced High Temperature Reactor (PB-AHTR), which combines TRISO fuel technology and liquid salt [flibe (2LiF-Be2F)] cooling. Compared to equivalent helium-cooled cores, the flibe-cooled cores feature a significantly larger fraction of neutron loss to coolant absorption but also a reduced neutron loss to leakage. The flibe also significantly contributes to neutron slowing-down and allows an increase of the pebbles’ heavy metal-to-carbon volume ratio as compared to helium-cooled cores. In order to guarantee all negative reactivity coefficients, and in particular coolant void and temperature feedbacks, the carbon-to-heavy metal atom ratio must not exceed 300 to 400, depending on the fuel kernel diameter. The maximum burnup attainable from a PB-AHTR that is fueled with 10% enriched uranium and operated in continuous refueling is ˜130 GWd/t HM; this is comparable to the maximum burnup achieved in other high-temperature reactors, either liquid salt or gas cooled. Compared to helium-cooled pebble bed reactors, the PB-AHTR pebbles can be loaded with 2.5 times more fuel, resulting in a smaller number of pebbles to fabricate and a smaller spent-fuel volume to handle per energy generated. Relative to a light water reactor, the PB-AHTR offers improved natural uranium ore utilization and reduced enrichment capacity.

Design Summary of the Mark-I Pebble-Bed, Fluoride Salt–Cooled, High-Temperature Reactor Commercial Power Plant

Abstract The University of California, Berkeley (UCB), has developed a preconceptual design for a commercial pebble-bed (PB), fluoride salt–cooled, high-temperature reactor (FHR) (PB-FHR). The baseline design for this Mark-I PB-FHR (Mk1) plant is a 236-MW(thermal) reactor. The Mk1 uses a fluoride salt coolant with solid, coated-particle pebble fuel. The Mk1 design differs from earlier FHR designs because it uses a nuclear air-Brayton combined cycle designed to produce 100 MW(electric) of base-load electricity using a modified General Electric 7FB gas turbine. For peak electricity generation, the Mk1 has the ability to boost power output up to 242 MW(electric) using natural gas co-firing. The Mk1 uses direct heating of the power conversion fluid (air) with the primary coolant salt rather than using an intermediate coolant loop. By combining results from computational neutronics, thermal hydraulics, and pebble dynamics, UCB has developed a detailed design of the annular core and other key functional features. Both an active normal shutdown cooling system and a passive, natural-circulation-driven emergency decay heat removal system are included. Computational models of the FHR—validated using experimental data from the literature and from scaled thermal-hydraulic facilities—have led to a set of design criteria and system requirements for the Mk1 to operate safely and reliably. Three-dimensional, computer-aided-design models derived from the Mk1 design criteria are presented.

Generic repository design concepts and thermal analysis (FY11).

Reference concepts for geologic disposal of used nuclear fuel and high-level radioactive waste in the U.S. are developed, including geologic settings and engineered barriers. Repository thermal analysis is demonstrated for a range of waste types from projected future, advanced nuclear fuel cycles. The results show significant differences among geologic media considered (clay/shale, crystalline rock, salt), and also that waste package size and waste loading must be limited to meet targeted maximum temperature values. In this study, the UFD RD (2) waste generated from reprocessing of LWR UOX UNF to recover U and Pu, and subsequent direct disposal of used Pu-MOX fuel (also used in LWRs) in a modified-open cycle; and (3) waste generated by continuous recycling of metal fuel from fast reactors operating in a TRU burner configuration, with additional TRU material input supplied from reprocessing of LWR UOX fuel. The geologic setting provides the natural barriers, and establishes the boundary conditions for performance of engineered barriers. The composition and physical properties of the host medium dictate design and construction approaches, and determine hydrologic and thermal responses of the disposal system. Clay/shale, salt, and crystalline rock media are selected as the basis for reference mined geologic disposal concepts in this study, consistent with advanced international repository programs, and previous investigations in the U.S. The U.S. pursued deep geologic disposal programs in crystalline rock, shale, salt, and volcanic rock in the years leading up to the Nuclear Waste Policy Act, or NWPA (Rechard et al. 2011). The 1987 NWPA amendment act focused the U.S. program on unsaturated, volcanic rock at the Yucca Mountain site, culminating in the 2008 license application. Additional work on unsaturated, crystalline rock settings (e.g., volcanic tuff) is not required to support this generic study. Reference disposal concepts are selected for the media listed above and for deep borehole disposal, drawing from recent work in the U.S. and internationally. The main features of the repository concepts are discussed in Section 4.5 and summarized in Table ES-1. Temperature histories at the waste package surface and a specified distance into the host rock are calculated for combinations of waste types and reference disposal concepts, specifying waste package emplacement modes. Target maximum waste package surface temperatures are identified, enabling a sensitivity study to inform the tradeoff between the quantity of waste per disposal package, and decay storage duration, with respect to peak temperature at the waste package surface. For surface storage duration on the order of 100 years or less, waste package sizes for direct disposal of SNF are effectively limited to 4-PWR configurations (or equivalent size and output). Thermal results are summarized, along with recommendations for follow-on work including adding additional reference concepts, verification and uncertainty analysis for thermal calculations, developing descriptions of surface facilities and other system details, and cost estimation to support system-level evaluations.

Thorium fuel cycles with externally driven systems

Abstract Externally driven subcritical systems are closely associated with thorium, partially because thorium has no naturally occurring fissile isotopes. Both accelerator-driven systems (ADSs) and fusion-driven systems have been proposed. This paper highlights key literature related to the use of thorium in externally driven systems (EDSs) and builds upon this foundation to identify potential roles for EDSs in thorium fuel cycles. In fuel cycles with natural thorium feed and no enrichment, the potential roles are (1) a once-through breed-and-burn fuel cycle and (2) a fissile breeder (mainly 233U) to support a fleet of critical reactors. If enriched uranium is used in the fuel cycle in addition to thorium, EDSs may be used to burn transuranic material. These fuel cycles were evaluated in the recently completed U.S. Department of Energy Evaluation and Screening of nuclear fuel cycle options relative to the current once-through commercial nuclear fuel cycle in the United States. The evaluation was performed with respect to nine specified high-level criteria, such as waste management and resource utilization. Each of these fuel cycles presents significant potential benefits per unit energy generation compared to the present once-through uranium fuel cycle. A parametric study indicates that fusion-fission–hybrid systems perform better than ADSs in some missions due to a higher neutron source relative to the energy required to produce it. However, both potential externally driven technology choices face significant development and deployment challenges. In addition, there are significant challenges associated with the use of thorium fuel and with the transition from a uranium-based fuel cycle to a thorium-based fuel cycle.

Equilibrium Core Composition Search Methodologies for Pebble Bed Reactors

Abstract The capability to perform depletion analysis of pebble bed reactors has been traditionally limited to a few dedicated codes that are designed for helium-cooled reactors, rely on pregenerated problem-dependent group cross sections, and have limited flexibility in the materials and in the geometries they can model. This paper presents a newly developed tool to search for pebble bed reactor core equilibrium composition and calculate its neutronic characteristics. It uses MCNP for transport calculations and ORIGEN2 for depletion calculations and can generate effective one-group cross sections “on-the-fly” as pebbles move through the core using point-energy cross sections. This tool can be used for any coolant type including liquid salt, can model complex geometries, and can account for any level of heterogeneity. Also developed are two simplified methodologies that are based on unit-cell analysis and can considerably reduce the required computational time; they are useful for parametric studies.

Recycling-Independent Core Design for the ENHS Fuel Self-Sustaining Reactor

Abstract The feasibility of indefinite recycling in the Encapsulated Nuclear Heat Source (ENHS) core without changing the pitch-to-diameter (P/D) ratio, while maintaining a nearly zero burnup reactivity swing, is investigated. The P/D ratio required to achieve a nearly burnup-independent keff over the life of the ENHS core was found sensitive to the initial composition of the transuranium (TRU) loaded and to the number of recycles this fuel underwent. The longer the cooling time is of the TRU from light water reactor (LWR) spent fuel, the larger the optimal P/D ratio becomes. Whereas the optimal P/D ratio of the reference ENHS core that is fueled with TRU from LWR spent fuel discharged at 50 GWd/t heavy metal (HM) and cooled for 10 yr is 1.36, it is 1.54 for the equilibrium core that features a substantially smaller concentration of 241Pu as well as of 242Pu, a larger concentration of 239Pu, and a substantially larger concentration of minor actinides. It was found that by increasing the cooling period of the above LWR TRU to ~32 yr, the optimal first core P/D ratio is that of the equilibrium core. The burnup reactivity swing of the subsequent cores fueled with successive recycling of the ENHS discharged HM is satisfactory. There is no need to adjust the core P/D ratio from recycle to recycle. The power level that can be removed by natural circulation from the P/D = 1.54 core is ~36% higher than that of the reference ENHS core. The physical phenomena affecting the observed trends are discussed, and the neutronic characteristics of the equilibrium cores identified are summarized.

Advanced Burner Reactors with Breed-and-Burn Thorium Blankets for Improved Economics and Resource Utilization

Abstract This paper assesses the feasibility of designing seed-and-blanket (S&B) sodium-cooled fast reactor (SFR) cores to generate a significant fraction of the core power from radial thorium-fueled blankets that operate in the breed-and-burn (B&B) mode. The radiation damage on the cladding material in both seed and blanket does not exceed the presently acceptable constraint of 200 displacements per atom (dpa). The S&B core is designed to have an elongated seed (or driver) to maximize the fraction of neutrons that radially leak into the subcritical B&B blanket and reduce the neutron loss via axial leakage. A specific objective of this study is to maximize the fraction of core power generated by the B&B blanket that is proportional to the neutron leakage rate from the seed to the blanket. Since the blanket feed fuel is very inexpensive and requires no reprocessing and remote fuel fabrication, a larger fraction of power from the blanket will result in a lower fuel cycle cost per unit of electricity generated by the SFR core. It is found possible to design the seed of the S&B core to have a lower transuranics (TRU) conversion ratio (CR) than a conventional advanced burner reactor (ABR) core without deteriorating core safety. This is due to the unique synergism between a low CR seed and the B&B thorium blanket. The benefits of the synergism are maximized when using an annular seed surrounded by inner and outer thorium blankets. Two high-performance S&B cores are designed to benefit from the annular seed concept: (1) an ultra-long-cycle core having a CR = 0.5 seed and a cycle length of ~7 effective full-power years (EFPYs) and (2) a high-transmutation core having a TRU CR of 0.0. The TRU transmutation rate of the latter core is comparable to that of the reference ABR with a CR of 0.5, and the thorium blanket can generate close to 60% of the core power. Because of the high blanket power fraction along with the high discharge burnup of the CR = 0 seed, the reprocessing capacity per unit of core power required by this S&B core is only approximately 1/6th that of the reference ABR core with a TRU CR of 0.5. Although the seed fuel CR is nearly zero, the burnup reactivity swing is low enough to enable a cycle length of more than 4 EFPYs. This is attributed to a combination of reactivity gain in the thorium blankets over the cycle and the relatively high heavy metal inventory. Moreover, despite the very low leakage, the S&B cores feature a less positive coolant reactivity coefficient and large enough negative Doppler coefficient even when using nonfertile fuel for the seed, because of the unique physics properties of the 233U and Th in the thorium blankets. With the long cycles, the S&B SFR is expected to have a higher capacity factor, and therefore a lower cost of electricity, than conventional ABRs. The discharge burnup of the thorium blanket fuel is typically 70 MWd/kg such that the thorium fuel utilization is approximately 12 times that of natural uranium in light water reactors. A sensitivity study is subsequently undertaken to quantify the trade-off between the core performances and several design variables: amount of zirconium in the inert matrix seed fuel, active core height, coolant pressure drop, and radiation damage constraint. The effect of the criterion used for quantifying acceptable radiation damage is evaluated as well. It is concluded that a viable S&B core can be designed without significant deviation from typical SFR core design practices.

Core Design and Analysis of Axially Heterogeneous Boiling Water Reactor for Burning Transuranium Elements

Abstract The resource-renewable boiling water reactor (RBWR) is an innovative boiling water reactor that has the capability to breed or to burn transuranium elements (TRUs). Core characteristics of the RBWR of the TRU burner type were evaluated by two different core analysis methods. The RBWR core features an axially heterogeneous configuration, which consists of an internal blanket region between two seed regions, to achieve the TRU multi-recycling capability while maintaining a negative void reactivity coefficient. Axial power distribution of the TRU burner core tends to be more heterogeneous because the isotopic composition ratio of fertile TRUs to fissile TRUs becomes larger in the TRU burner–type core than in the breeder-type core and the seed regions need to be axially shorter than that of the breeder-type core. Thus core analysis of the TRU burner–type core is more challenging. A conventional diffusion calculation using nuclear constants prepared by two-dimensional lattice calculations was performed by Hitachi, while the calculation using nuclear constants prepared by three-dimensional calculations and axial discontinuity factors was performed by the University of Michigan to provide a more sophisticated treatment of the axial heterogeneity. Both calculations predicted similar axial power distributions except in the region near the boundary between fuel and plenum. Both calculations also predicted negative void reactivity coefficients throughout the operating cycle. Safety analysis was performed by Massachusetts Institute of Technology for the all-pump trip accident, which was identified as the limiting accident for the RBWR design. The analysis showed the peak cladding temperature remains below the safety limit. Detailed fuel cycle analysis by University of California, Berkeley, showed that per electrical power generated, the RBWR is capable of incinerating TRUs at about twice the rate at which they are produced in typical pressurized water reactors.

Assembly design of pressurized water reactors with fully ceramic microencapsulated fuel

Abstract Fully ceramic microencapsulated (FCM) fuel consists of TRISO (tristructural-isotropic) fuel particles embedded in a ceramic matrix (SiC) to form fuel pellets and rods and offers improved fission product retention and lower operating temperature with expected superior performance in normal and off-normal conditions compared to conventional fuel. When coupled with SiC cladding, FCM fuel eliminates zirconium altogether and is expected to drastically reduce hydrogen generation during a beyond-design-basis accident. In order to be deployed in current or future pressurized water reactors (PWRs), FCM fuel must meet or exceed the neutronic performance of conventional fuel. Limited by low heavy metal loading, an FCM fuel assembly requires increased enrichment and large fuel rods to match the cycle length of a conventional fuel assembly. This study investigated the core design, neutronics, and thermal hydraulics of a PWR loaded with FCM fuel and sought to optimize the assembly design to minimize the enrichment required to reach fuel performance similar to that of conventional fuel. It was found that the implementation of FCM fuel in a 17 × 17 assembly requires close to 20% enrichment and large fuel rods. Such design performs comparably to conventional fuel (4.5% enrichment) in terms of cycle length, reactivity coefficients, intra-assembly power peaking factor, burnable poison penalty, and control rod worth but requires an increase of pumping power. A parametric analysis spanned a large design space varying fuel outer diameter and pitch-to-diameter ratio (P/D) and downselected two alternate assembly designs: 11 × 11 (1.65-cm outer diameter and 1.18 P/D) and 9 × 9 (2.12-cm outer diameter and 1.12 P/D). These designs meet the cycle length requirement with 18.6% and 16.2% enrichments, respectively, but feature a smaller minimum departure from nucleate boiling ratio (MDNBR) compared to a reference assembly. It was estimated that a slight increase in rod outer diameter increases MDNBR to the desired level and implies a pressure drop increase of 10%.