Ehud Greenspan

The Encapsulated Nuclear Heat Source (ENHS) Reactor Core Design

Abstract A once-for-life, uniform composition, blanket-free and fuel-shuffling-free reference core has been designed for the Encapsulated Nuclear Heat Source (ENHS) to provide the design goals of a nearly zero burnup reactivity swing throughout ~20 yr of full-power operation up to the peak discharge burnup of more than 100 GWd/t HM. What limits the core life is radiation damage to the HT-9 structural material. The temperature coefficients of reactivity are all negative, except for that of the coolant expansion. However, the negative reactivity coefficient associated with the radial expansion of the core structure can compensate for the coolant thermal expansion. The void coefficient is positive but of no safety concern because the boiling temperature of lead or lead-bismuth is so high that there is no conceivable mechanism for the introduction of significant void fraction into the core. The core reactivity coefficients, reactivity worth, and power distributions are almost constant throughout the core life. It was found possible to design such once-for-life cores using different qualities of Pu and transuranics as long as U is used as the primary fertile material. It is also feasible to design ENHS cores using nitride rather than metallic fuel. Relative to the reference metallic fuel core, nitride fuel cores offer up to ~25% higher discharge burnup and longer life, up to ~38% more energy per core, a significantly more negative Doppler reactivity coefficient, and less positive coolant expansion and coolant void reactivity coefficient but a somewhat smaller negative fuel expansion reactivity coefficient. The pitch-to-diameter ratio (1.45 of the nitride fuel cores using enriched N) is larger than that (1.36) for the reference metallic fuel core, implying a reduction of the coolant friction loss, thus enabling an increase in the power level that can be removed from the core by natural circulation cooling. It is also possible to design Pu-U(10Zr) fueled ENHS-type cores using Na as the primary coolant with either Na or Pb-Bi secondary coolants. The Na-cooled cores feature a tighter lattice and are therefore more compact but have spikier power distribution, more positive coolant temperature reactivity coefficients, and smaller reactivity worth of the control elements.

Neutron Balance Analysis for Sustainability of Breed-and-Burn Reactors

Abstract One objective of the present work is to determine the minimum burnup (BU) required to sustain a breed-and-burn (B&B) mode of operation in a large 3000-MW(thermal) sodium-cooled fast reactor core fed with depleted uranium-based metallic fuel. Another objective is to assess the feasibility of using the fuel discharged at the minimum required BU for fabricating the starter of an additional B&B core without separation of actinides and most of the solid fission products. A melt-refining process is used to remove gaseous and volatile fission products and to replace the cladding when it reaches its 200 displacements per atom radiation damage limit. Additional objectives are to assess the validity of a simplified zero-dimensional (0-D) neutron balance analysis for determination of the minimum BU required and the maximum BU attainable in a B&B mode of operation and to apply this 0-D methodology to assess the feasibility of establishing a B&B mode of operation in fast reactor cores made of different combinations of fuels, coolants, and structural materials. It is found that the minimum BU required to sustain the B&B mode in the referenced depleted uranium-fueled B&B reactor is 19.4% FIMA. The number of excess neutrons that can be generated by the fuel discharged at 19.4% FIMA is found sufficient to establish the B&B mode in another B&B core. The net doubling time for starting new B&B reactors with fuel discharged from operating B&B reactors is 12.3 yr. The minimum BU required to sustain the B&B mode of operation in alternative core designs was found to be 29% FIMA when using Pb-Bi coolant with metallic uranium fuel and 40% FIMA when using nitride fuel with sodium coolant. The B&B mode of operation cannot be established using thorium fuel and liquid-metal coolant. The results derived from the neutron balance analysis strongly depend on the value of the estimated neutron leakage probability and the fraction of neutrons lost in the reactivity control systems. A neutron balance performed using a simplified 0-D core model, although not accurate due to, primarily, inaccurate spectra predictions, provides reasonable estimates of the minimum required and the maximum attainable BUs despite the fact that its k∞ evolution prediction is inaccurate. The 0-D approach can save much computational effort and time and is found to be useful for scoping analysis.

Considerations of Autocatalytic Criticality of Fissile Materials in Geologic Respositories

Potential routes to autocatalytic criticality in geologic repositories are systematically assessed. If highly enriched uranium (HEU) or {sup 239}Pu are transported and deposited in concentrations similar to natural uranium ore, in principle, criticality can occur. For some hypothesized critical configurations, removal of a small fraction of pore water provides a positive feedback mechanism that can lead to supercriticality. Rock heating and homogenization for these configurations can also significantly increase reactivity. At Yucca Mountain, it is highly unlikely that these configurations can occur; plutonium transport would occur primarily as colloids and deposit over short distances. HEU solute can move large distances in the Yucca Mountain setting; its ability to precipitate into critical configurations is unlikely because of a lack of active reducing agents. Appropriate engineering of the waste form and the repository can reduce any remaining probability of criticality.

Turbulence and the feasibility of self-cooled liquid metal blankets for fusion reactors

Magnetohydrodynamics (MHD) considerations are of paramount importance in the design and performance of self-cooled liquid-metal (LM) blankets; the interaction between the magnetic field and the flowing LM can have a significant effect on the pressure drop, the heat transfer rate, and the corrosion rate. The purpose of the present work is to assess the implications that the recent experimental findings might have on the performance of self-cooled LM blankets. The assessment is done by considering the poloidal blanket concept, which uses a vanadium alloy for the structure. Material strength and LM compatibility considerations limit the first-wall (FW) and FW/LM interface temperature to 750/sup 0/C. When the FW is subjected to a 0.5 MW/m/sup 2/ heat flux, a temperature drop of approx. 100/sup 0/C will be established across it, restricting the FW/LM interface temperature to T/sub int/ approx. 650/sup 0/C. It is concluded that the enhanced two-dimensional turbulence might significantly increase the attractiveness of self-cooled LM blankets by enabling (a) the design of simpler and, possibly, cheaper blankets, and (b) the attainment of lower pumping power requirements and higher energy conversion efficiency.

Possibilities for Improvements in Liquid-Metal Reactors Using Liquid-Metal Magnetohydrodynamic Energy Conversion

AbstractPossibilities for increasing efficiency, simplifying the design of the energy conversion system, and reducing the probability of sodium/water interaction in liquid-metal reactors (LMRs) using liquid-metal magnetohydrodynamic (LMMHD) energy conversion technology are investigated. Of the six different LMMHD power conversion systems considered, the LMMHD Rankine steam cycle offers the highest efficiency—up to 15% greater than a conventional LMR. The LMMHD Ericsson gas cycles, on the other hand, offer a significantly simplified and compact LMR plant design. All the LMMHD power conversion systems eliminate the sodium/water interaction problem. In addition to commercial applications, LMMHD energy conversion technology opens interesting new possibilities for special terrestrial as well as space applications of LMRs.

New Fuel Cycle and Fuel Management Options in Heavy Liquid Metal-Cooled Reactors

Abstract Fast reactors cooled by lead or lead-bismuth alloy offer new interesting fuel cycle and fuel management options by virtue of the superb neutronics and safety features of these heavy liquid metal (HLM) coolants. One option is once-for-life cores having relatively low power density. These cores are fueled in the factory; there is no refueling or fuel shuffling on site. A second option is very long-life cores being made of a fissioning zone and a natural uranium blanket zone. The fissioning zone very slowly drifts toward the blanket. A third option is multirecycling of light water reactor (LWR) discharged fuel without partitioning of transuranics (TRUs) in fuel-self-sustaining reactors. LWR spent fuel could provide the initial fuel loading after extracting fission products and ~90% of its uranium. The makeup fuel is natural or depleted uranium. A fourth option is the high-burnup once-through fuel cycle using natural or depleted uranium feed. The initial fuel loading of this reactor is a mixture of enriched and natural uranium. The natural uranium utilization is 10 to 20 times higher than that of a once-through LWR. A fifth option is transmutation of TRUs from LWRs using critical HLM-cooled reactors; such reactors could be designed to have the same high actinide burning capability of accelerator-driven systems and have comparable safety, but at a substantially lower cost. These novel reactor designs and fuel management options are hereby reviewed.

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.

IRIS: Proceeding Towards the Preliminary Design

The IRIS (International Reactor Innovative and Secure) project has completed the conceptual design phase and is moving towards completion of the preliminary design, scheduled for the end of 2002. Several other papers presented in this conference provide details on major aspects of the IRIS design. The three most innovative features which uniquely characterize IRIS are, in descending order of impact: 1. Safety-by-design, which takes maximum advantage of the integral configuration to eliminate from consideration some accidents, greatly lessen the consequence of other accident scenarios and decrease their probability of occurring; 2. Optimized maintenance, where the interval between maintenance shutdowns is extended to 48 months; and 3. Long core life, of at least four years without shuffling or partial refueling. Regarding feature 1, design and analyses will be supplemented by an extensive testing campaign to verify and demonstrate the performance of the integral components, individually as well as interactive systems. Test planning is being initiated. Test results will be factored into PRA analyses under an overall risk informed regulation approach, which is planned to be used in the IRIS licensing. Pre-application activities with NRC are also scheduled to start in mid 2002. Regarding feature 2, effort is being focused on advanced online diagnostics for the integral components, first of all the steam generators, which are the most critical component; several techniques are being investigated. Finally, a four year long life core design is well underway and some of the IRIS team members are examining higher enrichment, eight to ten year life cores which could be considered for reloads.Copyright

Performance of Large Breed-and-Burn Core

A sodium-cooled fast reactor breed-and-burn (B&B) core and fuel cycle concept are proposed to achieve uranium utilization in the vicinity of 50% without separation of most of the fission products from the actinides. This core is to be fueled with depleted uranium (DU) with the exception of the initial core loading that uses fissile fuel to achieve initial criticality. When the cladding reaches its radiation damage limit, the melt-refining process is used to recondition the fuel, and then the fuel is reloaded into the core. This fuel reconditioning continues until the fuel reaches the neutronically maximum attainable burnup. When a fuel assembly is discharged at its maximum attainable burnup, it is replaced with a fresh DU assembly. The maximum burnup attainable in a large 3000-MW(thermal) B&B core is found to be 57% fissions per initial metal atoms (FIMA). The discharged fuel characteristics such as the inventory of actinides, radiotoxicity, and decay heat are one order of magnitude smaller, per unit of energy generated, than those of a light water reactor operating with the once-through fuel cycle. It is also found that the minimum burnup required for sustaining the B&B mode of operation is 19.4% FIMA. The fuel discharged at this burnup has sufficient excess reactivity for establishing initial criticality in a new large B&B core. The theoretical minimum doubling time for new core spawning is estimated to be [approximately]10 effective full-power years; there is no need for any external fissile material supply beyond that required for the initial “mother” reactor. Successful development and deployment of the B&B core along with fuel reconditioning could possibly provide up to 3000 yr worth of the current global nuclear electricity generation by using the DU stockpiles already accumulated worldwide. However, a number of important feasibility issues are yet to be resolved.