Showing papers in "Nuclear Technology in 2004"

An Overview of Corrosion Issues for the Design and Operation of High-Temperature Lead- and Lead-Bismuth-Cooled Reactor Systems

Abstract: The viability of advanced Pb- or Pb-Bi–cooled fast reactor systems will depend on the development of classes of materials that can operate over the temperature range 650–1200°C. We briefly review t...

Use of Thorium for Transmutation of Plutonium and Minor Actinides in PWRs

Abstract: An assessment is made of the potential for Th-based fuel to minimize Pu and minor actinide (MA) production in pressurized water reactors (PWRs). Destruction rates and residual amounts of Pu and MA ...

Prognostics and Condition Based Maintenance: A New Approach to Precursive Metrics

Abstract: The assumptions used in the design basis of process equipment have always been as much art as science. The usually imprecise boundaries of the equipments' operational envelope provide opportunities for two major improvements in the operations and maintenance (O&M) of process machinery: (a) the actual versus intended machine environment can be understood and brought into much better alignment and (b) the end goal can define O&M strategies in terms of life cycle and economic management of plant assets. Scientists at the Pacific Northwest National Laboratory (PNNL) have performed experiments aimed at understanding and controlling aging of both safety-specific nuclear plant components and the infrastructure that supports essential plant processes. In this paper we examine the development of aging precursor metrics and their correlation with degradation rate and projected machinery failure. Degradation-specific correlations have been developed at PNNL that will allow accurate physics-based diagnostic and prognostic determinations to be derived from a new view of condition-based maintenance. This view, founded in root cause analysis, is focused on quantifying the primary stressor(s) responsible for degradation in the component of interest and formulating a deterministic relationship between the stressor intensity and the resulting degradation rate. This precursive relationship between the performance, degradation, and underlying stressor set is used to gain a first-principles approach to prognostic determinations. A holistic infrastructure approach, as applied through a conditions-based maintenance framework, will allow intelligent, automated diagnostic and prognostic programming to provide O&M practitioners with an understanding of the condition of their machinery today and an assurance of its operational state tomorrow.

Development of a Cobalt Dicarbollide/Polyethylene Glycol Solvent Extraction Process for Separation of Cesium and Strontium to Support Advanced Aqueous Reprocessing

Abstract: A chlorinated cobalt dicarbollide(CCD)/polyethylene glycol (PEG) based solvent extraction process is being developed for the separation of Cs and Sr from leached spent light water reactor (LWR) fue...

Biogeochemical Reactive Transport Model of the Redox Zone Experiment of the Äspö Hard Rock Laboratory in Sweden

Abstract: Underground facilities are being operated by several countries around the world for performing research and demonstration of the safety of deep radioactive waste repositories. The Aspo Hard Rock La...

Microheterogeneous thoria-urania fuels for pressurized water reactors

Abstract: A thorium-based fuel cycle for light water reactors will reduce the plutonium generation rate and enhance the proliferation resistance of the spent fuel. However, priming the thorium cycle with 235 U is necessary, and the 235 U fraction in the uranium must be limited to below 20% to minimize proliferation concerns. Thus, a once-through thorium-uranium dioxide (ThO 2 -UO 2 ) fuel cycle of no less than 25% uranium becomes necessary for normal pressurized water reactor (PWR) operating cycle lengths. Spatial separation of the uranium and thorium parts of the fuel can improve the achievable burnup of the thorium-uranium fuel designs through more effective breeding of 233 U from the 232 Th. Focus is on microheterogeneous fuel designs for PWRs, where the spatial separation of the uranium and thorium is on the order of a few millimetres to a few centimetres, including duplex pellet, axially microheterogeneous fuel, and a checkerboard of uranium and thorium pins. A special effort was made to understand the underlying reactor physics mechanisms responsible for enhancing the achievable burnup at spatial separation of the two fuels. The neutron spectral shift was identified as the primary reason for the enhancement of burnup capabilities. Mutual resonance shielding of uranium and thorium is also a factor; however, it is small in magnitude. It is shown that the microheterogeneous fuel can achieve higher burnups, by up to 15%, than the reference all-uranium fuel. However, denaturing of the 233 U in the thorium portion of the fuel with small amounts of uranium significantly impairs this enhancement. The denaturing is also necessary to meet conventional PWR thermal limits by improving the power share of the thorium region at the beginning of fuel irradiation. Meeting thermal-hydraulic design requirements by some of the microheterogeneous fuels while still meeting or exceeding the burnup of the all-uranium case is shown to be potentially feasible. However, the large power imbalance between the uranium and thorium regions creates several design challenges, such as higher fission gas release and cladding temperature gradients. A reduction of plutonium generation by a factor of 3 in comparison with all-uranium PWR fuel using the same initial 235 U content was estimated. In contrast to homogeneously mixed U-Th fuel, microheterogeneous fuel has a potential for economic performance comparable to the all-UO 2 fuel provided that the microheterogeneous fuel incremental manufacturing costs are negligibly small.

Studies of Polonium Removal from Molten Lead-Bismuth for Lead-Alloy-Cooled Reactor Applications

Abstract: The isotope 210Po is the main product of neutron activation in fast reactors cooled by molten lead-bismuth eutectic (LBE). The isotope 210Po is a pure alpha emitter with a half-life of 138.38 days....

Use of Thoria-Urania Fuels in PWRs: A General Review of a NERI Project to Assess Feasible Core Designs, Economics, Fabrication Methods, In-Pile Thermal/Mechanical Behavior, and Waste Form Characteristics

Abstract: This paper provides an introduction to and a summary of the remaining papers in this issue of Nuclear Technology. The papers in this issue present the important results from a U.S. Department of Energy-sponsored Nuclear Engineering Research Initiative (NERI) project to study the efficacy of the thorium-uranium dioxide (ThO2-UO2) once-through fuel cycle in current light water reactors. The project addressed fuel cycle neutronics and economics; ThO2-UO2 fuel manufacturing; the in-pile thermal/mechanical behavior of ThO2-UO2 fuel during normal, off-normal, and accident conditions; and the long-term stability of ThO2-UO2 waste. Results from this work show that a small-scale separation of the uranium and thorium will enhance the fuel reactivity and achievable burnup from uranium-thorium dioxide fuels. Under conditions that meet the thermal requirements in present pressurized water reactors (PWRs), a properly designed microheterogeneous fuel will have more reactivity than all-uranium fuel, and the overall production of plutonium is significantly reduced. The use of thorium as a host for actinide fuels when PWRs are used for actinide transmutation was also explored. It was also determined that there were no fundamental obstacles to converting the current plants that manufacture uranium oxide-only fuel to a mixed ThO2-UO2 fuel. Also, the in-service and transient thermal andmore » mechanical performance of homogeneous ThO2-UO2-based fuels with respect to safety is generally equal to or better than that of all-uranium fuel. Furthermore, a mixed thorium-uranium dioxide spent fuel appears to be a much more stable waste form than uranium oxide spent fuel.« less

Transient Fuel Behavior and Failure Condition in the CABRI-2 Experiments

Abstract: In the CABRI-2 program, 12 tests were performed under various transient conditions covering a wide range of accident scenarios using two types of preirradiated fast breeder reactor (FBR) fuel pins with different smear densities and burnups. For each fuel, a nonfailure-transient test was performed, and it provided basic information such as fuel thermal condition, fuel swelling, and gas release. From the failure tests, information on failure mode, failure time, and axial location was obtained. Based on this information, failure conditions such as fuel enthalpy and cladding temperature were evaluated. These failure conditions were compared with the CABRI-1 tests in which different fuels as well as different transient conditions were used. This comparison, together with supporting information available from existing in-pile and out-of-pile experiments, allowed an effective understanding on failure mechanisms depending on fuel and transient conditions. It is concluded that pellet-cladding mechanical interaction (PCMI) due to fuel thermal expansion and fission-gas-induced swelling is playing an important role on mechanical clad loading especially with high smear density and low fuel-heating-rate conditions. At very high heating-rate conditions, there is no sufficient time to allow significant fuel swelling, so that cavity pressurization with fuel melting becomes the likely failure mechanism. Fuel smear densitymore » and fission-gas retention have a strong impact both on PCMI and cavity pressurization. Furthermore, pin failure is strongly dependent on cladding temperature, which plays an important role in the axial failure location. With the low smear-density fuel, considerable PCMI mitigation is possible leading to a high failure threshold as well as in-pin molten-fuel relocation along the central hole. However, even with the low smear density fuel, PCMI failure could take place with an elevated cladding-temperature condition. On the other hand, in case of a sufficiently long transient timescale, such low smear density fuel has a potential to allow gas escape to plenum leading to a very effective mitigation of swelling-induced PCMI.In case of very high cladding temperature near its melting point, plenum-gas blowout at cladding rupture takes place before fuel disintegration. Fuel-disintegration behavior under this condition is dominated by fuel enthalpy, and no special effect of the high burnup can be identified through comparison with the CABRI-1 test results.« less

Design Strategy and Constraints for Medium-Power Lead-Alloy–Cooled Actinide Burners

Abstract: We outline the strategy and constraints adopted for the design of medium-power lead-alloy–cooled actinide-burning reactors that strive for a lower cost than accelerator-driven systems and for robus...

Fabrication and thermal conductivity of (Th, U)O2 pellets

Abstract: Techniques to fabricate thorium-uranium dioxide fuel [(Th,U)O2] have been developed, and the thermal conductivity of (Th,U)O2 pellets has been measured. Mixtures of thorium dioxide (ThO2) and urani...

Medium-Power Lead-Alloy Reactors: Missions for This Reactor Technology

Abstract: A multiyear project at the Idaho National Engineering and Environmental Laboratory and the Massachusetts Institute of Technology investigated the potential of medium-power lead-alloy-cooled technology to perform two missions: (1) the production of low-cost electricity and (2) the burning of actinides from light water reactor (LWR) spent fuel. The goal of achieving a high power level to enhance economic performance simultaneously with adoption of passive decay heat removal and modularity capabilities resulted in designs in the range of 600-800 MW(thermal), which we classify as a medium power level compared to the lower [{approx}100 MW(thermal)] and higher [2800 MW(thermal)] power ratings of other lead-alloy-cooled designs. The plant design that was developed shows promise of achieving all the Generation-IV goals for future nuclear energy systems: sustainable energy generation, low overnight capital cost, a very low likelihood and degree of core damage during any conceivable accident, and a proliferation-resistant fuel cycle. The reactor and fuel cycle designs that evolved to achieve these missions and goals resulted from study of the following key trade-offs: waste reduction versus reactor safety, waste reduction versus cost, and cost versus proliferation resistance. Secondary trade-offs that were also considered were monolithic versus modular design, active versus passive safety systems, forcedmore » versus natural circulation, alternative power conversion cycles, and lead versus lead-bismuth coolant.These studies led to a selection of a common modular design with forced convection cooling, passive decay heat removal, and a supercritical CO{sub 2} power cycle for all our reactor concepts. However, the concepts adopt different core designs to optimize the achievement of the two missions. For the low-cost electricity production mission, a design approach based on fueling with low enriched uranium operating without costly reprocessing in a once-through cycle was pursued to achieve a long operating cycle length by enhancing in-core breeding. For the actinide-burning mission three design variants were produced: (1) a fertile-free actinide burner, i.e., a single-tier strategy, (2) a minor actinide burner with plutonium burned in the LWR fleet, i.e., a two-tier strategy, and (3) an actinide burner with characteristics balanced to also favor economic electricity production.« less

Hydrogen Generation in Reflooding Experiments with LWR-Type Rod Bundles (QUENCH Program)

Abstract: The QUENCH bundle experiments together with pertinent separate-effects tests are run to investigate the hydrogen source term resulting from water injection into an uncovered core of a light water reactor for emergency cooling. The test bundle consists of 21 fuel rod simulators, 20 of which are heated electrically over a length of 1024 mm. The center rod is either an unheated fuel rod simulator or a control rod containing B{sub 4}C absorber material. The Zircaloy-4 rod cladding and the grid spacers are identical to those used in pressurized water reactors, whereas the fuel is represented by ZrO{sub 2} pellets. After transient heating to 2000 K and above, cooling of the test bundle is accomplished by injecting water or steam into the bottom of the test section. Hydrogen generation during cooling was found either to stop almost immediately or to increase for a certain time. Increased hydrogen generation was found in those tests in which local melting occurred, probably as a result of oxidation of the melt containing zirconium. Hydrogen release in the flooding/cooling phase of all QUENCH experiments performed so far seems to be insensitive to the coolant (water or steam) under similar test conditions.

Modeling of Colloid Transport Mechanisms Facilitating Migration of Radionuclides in Fractured Media

Abstract: Performance assessments of high-level radioactive waste disposal have emphasized the role of colloids in the migration of radionuclides in the geosphere. The transport of colloids often brings them in contact with fracture surfaces or porous rock matrix. Colloids that attach to these surfaces are treated as being immobile and are called filtered colloids. The filtered colloids could be released into the fracture again; that is, the attachment of colloids may be reversible. Also, the colloids in the fracture could diffuse into the porous matrix rock. A methodology is proposed to evaluate a predictive model to assess transport within the fractured rock as well as various phenomenological coefficients employed in the different mechanisms, such as filtration, remobilization, and matrix diffusion of colloids. The governing equations of colloids considering mechanisms of the colloidal transport in the fractured media, including filtration, remobilization, and matrix diffusion, have been modeled and solved analytically in previous studies. In the present study, transport equations of colloids and radionuclides that consider the combination of the aforementioned transport mechanisms have also been solved numerically and investigated. The total concentration of mobile radionuclides in the fracture becomes lower because the concentration of mobile colloids in the fracture decreases when themore » filtration coefficient for colloids increases. Additionally, the concentration of mobile radionuclides was increased at any given time step due to the higher sorption partition coefficient of radionuclides associated with colloids. The results also show that the concentration of radionuclides in the fracture zone decreases when the remobilization coefficient of colloids or the percentages of the matrix diffusion flux of colloids increase.« less

Aqueous dissolution of urania-thoria nuclear fuel

Abstract: The aqueous dissolution of irradiated and unirradiated uranium-thorium dioxide, (U,Th)O{sub 2}, fuel pellets in Yucca Mountain well water has been investigated. Whole and crushed pellets were reacted at 25 and 90 deg. C for periods of up to 195 days. The fuel dissolution was measured by analyzing the concentrations of soluble uranium, thorium, and important fission products ({sup 137}Cs, {sup 99}Tc, {sup 237}Np, {sup 239}Pu, {sup 240}Pu, and {sup 241}Am) in the well water. The surface-area-normalized fractional uranium release rates for unirradiated crushed uranium dioxide (UO{sub 2}) pellets were 10 to 40 times higher than the values for (U,Th)O{sub 2} fuel. Similarly, the dissolution rates of irradiated (U,Th)O{sub 2} pellets with compositions ranging from 2.0 to 5.2% UO{sub 2} were at least two orders of magnitude lower than reported literature values for pure UO{sub 2}. These results demonstrate an advantage of (U,Th)O{sub 2} over UO{sub 2} in terms of matrix dissolution in groundwater and suggest that (U,Th)O{sub 2} fuel is a more stable long-term waste form than conventional UO{sub 2} fuel.

Development and Validation of the 3-D Computational Fluid Dynamics Model for CANDU-6 Moderator Temperature Predictions

Abstract: A computational fluid dynamics (CFD) model for predicting the moderator circulation inside the Canada deuterium uranium (CANDU) reactor vessel has been developed to estimate the local subcooling of the moderator in the vicinity of the Calandria tubes. The buoyancy effect induced by internal heating is accounted for by Boussinesq approximation. The standard k-[curly epsilon] turbulence model associated with logarithmic wall treatment is applied to predict the turbulent jet flows from the inlet nozzles. The matrix of the Calandria tubes in the core region is simplified to porous media, in which anisotropic hydraulic impedance is modeled using an empirical correlation of the frictional pressure loss. The governing equations are solved by CFX-4.4, a commercial CFD code developed by AEA Technology. The CFD model has been successfully verified and validated against experimental data obtained at Stern Laboratories Inc. in Hamilton, Ontario, Canada.

Consistent Comparison of the Codes RELAP5/PARCS and TRAC-M/PARCS for the OECD MSLB Coupled Code Benchmark

Abstract: A generalized interface module was developed for coupling any thermal-hydraulic code to any spatial kinetic code. In the design used here the thermal-hydraulic and spatial kinetic codes function as independent processes and communicate using the Parallel Virtual Machine software. This approach helps maximize flexibility while minimizing modifications to the respective codes. Using this interface, the U.S. Nuclear Regulatory Commission (NRC) three-dimensional neutron kinetic code, Purdue Advanced Reactor Core Simulator (PARCS), has been coupled to the NRC system analysis codes RELAP5 and Modernized Transient Reactor Analysis Code (TRAC-M). Consistent comparison of code results for the Organization for Economic Cooperation and Development/Nuclear Energy Agency main steam line break benchmark problem using RELAP5/PARCS and TRAC-M/PARCS was made to assess code performance.

Advanced MOX Core Design Study of Sodium-Cooled Reactors in Current Feasibility Study on Commercialized Fast Reactor Cycle Systems in Japan

Abstract: Sodium-cooled mixed-oxide core design studies are performed with a target burnup of 150 GWd/t and possible measures against the recriticality issues in core disruptive accidents. Four types of core...

Fertile-Free Fast Lead-Cooled Incinerators for Efficient Actinide Burning

Abstract: Fertile-free fast lead-cooled modular reactors are proposed as efficient incinerators of plutonium and minor actinides (MAs) for application to advanced fuel cycles devoted to transmutation. Two concepts are presented: (1) an actinide burner reactor, designed to incinerate mostly plutonium and some MAs, and (2) a minor actinide burner reactor, devoted to burning mostly minor actinides and some plutonium. These transuranics are loaded in a fertile-free Zr-based metallic fuel to maximize the incineration rate. Both designs feature streaming fuel assemblies that enhance neutron leakage to achieve favorable neutronic feedback and a double-entry control rod system that reduces reactivity perturbations during seismic events and flattens the axial power profile. A detailed neutronic analysis shows that both designs have favorable neutronic characteristics and reactivity feedback mechanisms that yield passive safety features comparable to those of the Integral Fast Reactor. A safety analysis presents the response of the burners to anticipated transients without scram on the basis of (1) the integral parameter approach and (2) simulations of thermal-hydraulic accident scenario conditions. It is shown that both designs have large thermal margins that lead to safe shutdown without structural damage to the core components for a large spectrum of unprotected transients. Furthermore, the actinide destructionmore » rates are comparable to those of the accelerator transmutation of waste concept, and a fuel cycle cost analysis shows the potential for economical accomplishment of the transmutation mission compared to other proposed actinide-burning options.« less

Uranium Transport in a High-Throughput Electrorefiner for EBR-II Blanket Fuel

Abstract: A unique high-throughput Mk-V electrorefiner is being used in the electrometallurgical treatment of the metallic sodium-bonded blanket fuel from the Experimental Breeder Reactor II. Over many cycles, it transports uranium back and forth between the anodic fuel dissolution baskets and the cathode tubes until, because of imperfect adherence of the dendrites, it all ends up in the product collector at the bottom. The transport behavior of uranium in the high-throughput electrorefiner can be understood in terms of the sticking coefficients for uranium adherence to the cathode tubes in the forward direction and to the dissolution baskets in the reverse direction. The sticking coefficients are inferred from the experimental voltage and current traces and are correlated in terms of a single parameter representing the ratio of the cell current to the limiting current at the surface acting as the cathode. The correlations are incorporated into an engineering model that calculates the transport of uranium in the different modes of operation. The model also uses the experimentally derived electrorefiner operating maps that describe the relationship between the cell voltage and the cell current for the three principal transport modes. It is shown that the model correctly simulates the cycle-to-cycle variation of themore » voltage and current profiles. The model is used to conduct a parametric study of electrorefiner throughput rate as a function of the principal operating parameters. The throughput rate is found to improve with lowering of the basket rotation speed, reduction of UCl{sub 3} concentration in salt, and increasing the maximum cell current or cut-off voltage. Operating conditions are identified that can improve the throughput rate by 60 to 70% over that achieved at present.« less

Re-Evaluating Neptunium in Uranyl Phases Derived from Corroded Spent Fuel

Abstract: Interest in mechanisms that may control radioelement release from corroded commercial spent nuclear fuel (CSNF) has been heightened by the selection of the Yucca Mountain site in Nevada as the repository for high-level nuclear waste in the United States. Neptunium is an important radionuclide in repository models owing to its relatively long half-life and its high aqueous mobility as neptunyl [Np(V)O + 2 ]. The possibility of neptunium sequestration into uranyl alteration phases produced by corroding CSNF would suggest-a process for lowering neptunium concentration and subsequent migration from a geologic repository. However, there remains little experimental evidence that uranyl compounds will, in fact, serve as long-term host phases for the retention of neptunium under conditions expected in a deep geologic repository. To directly explore this possibility, we examined specimens of uranyl alteration phases derived from humid-air-corroded CSNF by X-ray absorption spectroscopy to better determine neptunium uptake in these phases. Although neptunium fluorescence was readily observed from as-received CSNF, it was not observed from the uranyl alteration rind. We establish upper limits for neptunium incorporation into CSNF alteration phases that are significantly below previously reported concentrations obtained by using electron energy loss spectroscopy (EELS). We attribute the discrepancy to a plural-scattering event that creates a spurious EELS peak at the neptunium-M V energy.

The Behavior of ThO2 Based Fuel Rods During Normal Operation and Transient Events in LWRs

Abstract: The thermal, mechanical, and chemical behavior of both thorium and uranium dioxide (ThO{sub 2}-UO{sub 2}) and thorium and plutonium dioxide (ThO{sub 2}-PuO{sub 2})-based fuels during in-service and hypothetical accident conditions in light water reactors (LWRs) is described These fuels offer the possibility for increased proliferation resistance and a reduction in the stockpile of weapons-grade and reactor-grade PuO{sub 2} as well as being a more stable waste form The behavior is described for three different designs of ThO{sub 2}-based fuels: a homogeneous mixture of ThO{sub 2}-UO{sub 2}, a microheterogeneous arrangement of the ThO{sub 2} and UO{sub 2}, and a homogeneous mixture of ThO{sub 2}-PuO{sub 2} The behavior was calculated with widely known LWR analysis tools extended for ThO{sub 2}-based fuels: (a) MATPRO for calculating material properties, (b) FRAPCON-3 for calculating in-service fuel temperature and fission-gas release, (c) VIPRE-01 for calculating the possibility for departure from nucleate boiling, (d) HEATING7 for calculating in-service two-dimensional temperature distributions in microheterogeneous fuel, (e) SCDAP/RELAP5-3D for calculating the transient reactor system behavior and fuel behavior during loss-of-coolant accidents, and (f) FRAP-T6 for calculating the vulnerability of the cladding to cracking due to swelling of the fuel during hypothetical reactivity-initiated accidentsThe analytical tools accounted for the followingmore » differences in ThO{sub 2}-based fuels relative to 100% UO{sub 2} fuel: (a) higher thermal conductivity, lower density and volumetric heat capacity, less thermal expansion, and higher melting point; (b) higher fission-gas production for {sup 233}U fission than {sup 235}U fission, but a lower gas diffusion coefficient in the ThO{sub 2} than in the UO{sub 2}; (c) less plutonium accumulation at the rim of the fuel pellets; (d) greater decay heat; (e) microheterogeneous arrangement of fuel; and (f) more-negative moderator temperature and Doppler coefficients and a smaller delayed-neutron fraction The newly developed models for ThO{sub 2} were checked against data from the light water breeder reactor program Calculations by these analytical tools indicate that the in-service and transient performance of homogeneous ThO{sub 2}-UO{sub 2}-based fuels with respect to safety is generally equal to or better than that of 100% UO{sub 2} fuel The in-service and transient temperatures in the most promising neutronic design of microheterogeneous ThO{sub 2}-UO{sub 2}-based fuel are greater than the temperatures in 100% UO{sub 2} fuel but are still within normal LWR safety limits The reactor kinetics parameters for ThO{sub 2}-PuO{sub 2}-based fuel cause a higher transient reactor power for some postulated accidents, but in general, the margin of safety for ThO{sub 2}-PuO{sub 2} fuels is equal to or greater than that in 100% UO{sub 2} fuels« less

Weapons-Grade Plutonium-Thorium PWR Assembly Design and Core Safety Analysis

Abstract: A light water reactor (LWR) fuel assembly design consisting of a blend of weapons-grade plutonium and natural thorium oxides was examined. The design meets current thermal-hydraulic and safety criteria. Such an assembly would have enough reactivity to achieve three cycles of operation. The pin power distribution indicates a fairly level distribution across the assembly, avoiding hot spots near guide tubes, corners, and other sections where excessive power would create significant loss to thermal-hydraulic margins.This work examined a number of physics and core safety analysis parameters that impact the operation and safety of power reactors. Such parameters as moderator coefficients of reactivity, Doppler coefficients, soluble boron worth, control rod worth, prompt neutron lifetime, and delayed-neutron fractions were considered. These in turn were used to examine reactor behavior during a number of operational conditions, transients, and accidents. Such conditions as shutdown from power with one rod stuck out, steam-line break accident, feedwater line break, loss of coolant flow, locked rotor accidents, control rod ejection accidents, and anticipated transients without scram (ATWSs) were examined.The analysis of selected reactor transients demonstrated that it is feasible to license and safely operate a reactor fueled with plutonium-thorium blended fuel. In most cases analyzed, the thorium mixturemore » had less-severe consequences than those for a core comprising low-enriched uranium fuel. In the analyzed cases where the consequences were more severe, they were still within acceptable limits. The ATWS accident condition requires more analysis.« less

TIMCOAT: An Integrated Fuel Performance Model for Coated Particle Fuel

Abstract: An integrated fuel performance model for coated particle fuel has been developed to comprehensively study the behavior of TRISO-coated fuel. Modeling of both pebble-bed and prismatic configurations is possible. In the case of the pebble-bed concept, refueling of pebbles is simulated to account for the nonuniform environment in the reactor core and history-dependent particle behavior. Monte Carlo sampling of particles is employed in fuel failure prediction to capture the statistical features of dimensions; material properties; and, in the case of the pebble-bed concept, the statistical nature of the refueling process. An advanced fuel failure model has been developed based on a probabilistic fracture mechanics approach. The mechanical analysis includes effects of anisotropic irradiation-induced dimensional changes and isotropic irradiation-induced creep, and the fluence-dependent Poisson ratio in irradiation creep. The stress analysis is benchmarked against the calculations of Japanese High Temperature Test Reactor (HTTR) first-loading fuel and finite element result on one case performed by the Idaho National Engineering and Environmental Laboratory. The failure model predictions are compared with NPR1, NPR2, and NPR1A capsule irradiation data. The model results compare very favorably with postirradiation examination results both in terms of failure probability, number of failed particles, and Kr{sup 85m} R/B evolution duringmore » irradiation.« less

The Use of Molybdenum-Based Ceramic-Metal (CerMet) Fuel for the Actinide Management in LWRs

Abstract: The technical and economic aspects of the use of molybdenum depleted in the isotope {sup 95}Mo (DepMo) for the transmutation of actinides in a light water reactor are discussed. DepMo has a low neutron absorption cross section and good physical and chemical properties. Therefore, DepMo is expected to be a good inert matrix in ceramic-metal fuel. The costs of the use of DepMo have been assessed, and it was concluded that these costs can be justified for the transmutation of the actinides neptunium, americium, and plutonium.

Alternative Applications of Homogeneous Thoria-Urania Fuel in Light Water Reactors to Enhance the Economics of the Thorium Fuel Cycle

Abstract: Since the thorium-based fuel has many incentives including the reduction of plutonium generation and long-lived radiotoxic isotope production, the research on the use of thorium as a nuclear fuel f...

Methods and Results for the MSLB NEA Benchmark Using SIMTRAN and RELAP-5

Abstract: The purpose of this paper is first to discuss the methods developed in our three-dimensional pressurized water reactor core dynamics code SIMTRAN and its coupling to the system code RELAP-5 for general transient and safety analysis. Then, we summarize its demonstration application to the Nuclear Energy Agency (NEA)/Organization for Economic Cooperation and Development (OECD) Benchmark on Main Steam Line Break (MSLB), co-sponsored by the U.S. Nuclear Regulatory Commission (NRC) and other regulatory institutions. In particular, our work has been supported by the Spanish 'Consejo de Seguridad Nuclear' (CSN) under a CSN research project.Our results for the steady states and the guided-core transients, proposed as exercise 2 of the MSLB benchmark, show small deviations from the mean results of all participants, especially in core average parameters. For the full-coupled core-plant transients, exercise 3, a detailed comparison with the University of Purdue-NRC results using PARCS/RELAP-5, shows quite good agreement in both integral and local parameters, especially for the more extreme return-to-power scenario.

Performance of the Lead-Alloy-Cooled Reactor Concept Balanced for Actinide Burning and Electricity Production

Abstract: A lead-bismuth-cooled fast reactor concept targeted for a balanced mission of actinide burning and low-cost electricity production is proposed and its performance analyzed. The design explores the potential benefits of thorium-based fuel in actinide-burning cores, in particular in terms of the reduction of the large reactivity swing and enhancement of the small Doppler coefficient typical of fertile-free actinide burners. Reduced electricity production cost is pursued through a longer cycle length than that used for fertile-free burners and thus a higher capacity factor. It is shown that the concept can achieve a high transuranics destruction rate, which is only 20% lower than that of an accelerator-driven system with fertile-free fuel. The small negative fuel temperature reactivity coefficient, small positive coolant temperature reactivity coefficient, and negative core radial expansion coefficient provide self-regulating characteristics so that the reactor is capable of inherent shutdown during major transients without scram, as in the Integral Fast Reactor. This is confirmed by thermal-hydraulic analysis of several transients without scram, including primary coolant pump trip, station blackout, and reactivity step insertion, which showed that the reactor was able to meet all identified thermal limits. However, the benefits of high actinide consumption and small reactivity swing can be attained more » only if the uranium from the discharged fuel is separated and not recycled. This additional uranium separation step and thorium reprocessing significantly increase the fuel cycle costs. Because the higher fuel cycle cost has a larger impact on the overall cost of electricity than the savings from the higher capacity factor afforded through use of thorium, this concept appears less promising than the fertile-free actinide burners. « less

A Simple and Efficient Control Rod Cusping Model for Three-Dimensional Pin-by-Pin Core Calculations

Abstract: A new solution for the control rod cusping problem in the three-dimensional pin-by-pin core calculation is proposed in this paper. The current advanced nodal code resolves this issue by estimating ...

Thorium-Based Transmuter Fuels for Light Water Reactors

Abstract: A light water reactor (LWR) fuel cycle is proposed where the reactor core mainly consists of standard uranium-dioxide (UO{sub 2}) fuel rods with typical {sup 235}U enrichment, along with thoria-urania (ThO{sub 2}-UO{sub 2}) or yttria-stablized zirconia fertile-free fuel rods containing the plutonium and minor actinides typical of 30-yr old UO{sub 2} fuel in 1/9 to 1/3 of the positions. The goals of this mono-recycling strategy or 'twice through fuel cycle' are to transmute the great majority of the long lived actinides in existing LWRs and to discharge a fuel form that is a very robust waste form and whose isotopic content is very proliferation resistant. The incorporation of plutonium into a ThO{sub 2} or yttria-stablized zirconia fertile-free matrix results in the consumption of already-separated plutonium without breeding significant additional {sup 239}Pu. The minor actinides (i.e., neptunium, americium, curium, berkelium, californium, etc.) are also included in the ThO{sub 2} or fertile-free transmuter fuel rods to further reduce the overall long-term radiotoxicity of the fuel cycle. Our analyses have shown that thorium-based or fertile-free fuels can reduce the amount of {sup 239}Pu needing further transmutation or going to a repository by {approx}90%. Also, thorium-based fuels produce a mixture of plutonium isotopes highmore » in {sup 238}Pu. Because of the high decay heat and spontaneous neutron generation of {sup 238}Pu, this isotope provides intrinsic proliferation resistance.« less