Uniform Burnup Research Articles

Variational methods and speed-up of Monte Carlo perturbation computations for optimal design in nuclear systems

Seeking optimal material distribution in a nuclear system to maximize a response function of interest has been a subject of considerable interest in nuclear engineering. Examples are the optimal fuel distribution in a nuclear reactor core to achieve uniform burnup using minimum critical mass and the use of composite materials with an optimal mix of constituent elements in detection systems and radiation shielding. For such studies, variational methods have been found to be useful but, they have been used for standalone analyses often restricted to idealized models, while more elaborate design studies have required computationally expensive Monte Carlo simulations ill-suited to iterative schemes for optimization. Such an inherent disadvantage of Monte Carlo methods changed with the development of perturbation algorithms but, their efficiency is still dependent on the reference configuration for which a hit-and-trial approach is often used. In the first illustrative example, this paper explores the computational speedup for a bare cylindrical reactor core, achievable by using a variational result to enhance the computational efficiency of Monte Carlo design optimization simulation. In the second example, the effect of non-uniform material density in a fixed-source problem, applicable to optimal moderator and radiation shielding, is presented. While applications of this work are numerous, the objective of this paper is to present preliminary variational results as inputs to elaborate stochastic optimization by Monte Carlo simulation for large and realistic systems.

Total neutron emission generation and characterization for a Next Generation Safeguards Initiative spent fuel library

In March of 2009 the Next Generation Safeguards Initiative of the U.S. Department of Energy began a nominal five year spent fuel research effort with the goal of: (1) quantifying plutonium content in spent nuclear fuel through the use of non-destructive assay (NDA) techniques; (2) quantifying the capability of these NDA techniques to quantify burnup, cooling time, and initial enrichment, as well as detect pin diversions, for spent fuel assemblies. These NDA techniques were first scoped in computation space in order to understand detection limitations and down select for further testing. Multiple Monte Carlo based spent fuel libraries (SFL's) were developed to be used as source terms for assessing the detection limitations of each of these techniques. The characterization of the SFL's provide an analysis of different burn strategies used for spent fuel, functional fits for state point analysis, and the components of the neutron signal.This paper details both a methodology for generating the spatially dependent gross neutron emission (GNE) as well as a characterization of the GNE for a particular NGSI SFL for two different shuffle patterns. The GNE was characterized by examining three major components: (1) the total neutron emission; (2) the (α, n) and spontaneous fission emissions; and (3) the main contributors to the spontaneous fission and (α, n) emission signals. The total neutron emission was characterized by looking at two different shuffle patterns and using three different pin analysis techniques. The pin analysis techniques used were: (1) sum all the pins in the assembly; (2) create four pin zones to be summed; and (3) examine specific pins. These characterizations provide a large set of data with integral values (cooling time vs emission and burnup vs emission) and derivative values (isotopic percent composition) with visual depiction and detailed explanation. The characterizations also provide a comparison of the effects of shuffle schemes on the neutron emission signal for two different shuffle patterns. This data can be used in the development of time and burnup dependent normalized functions to describe the trends displayed in the data. This study also shows that the information which can be obtained from the GNE depends upon how the information is grouped and analyzed. Summing all the pins of the assembly or assuming a uniform burnup results in the loss of valuable neutron emission information. By splitting the pins into zones, trends such as the total emissions across the assembly become evident. By evaluating individual pins trends such as the effect of water rods on the neutron emission become evident. This type of analysis can be of great use to safeguards. Evaluating the GNE on a pin by pin basis leads to a much clearer picture of the constituents of the assembly.

Impact of Local Burnup on Prediction of Power Density in the NIST Research Reactor

This study addresses the overprediction of local power when the burnup distribution in each half-element of the National Institute of Standards and Technology (NIST) research reactor, known as the NBSR, is assumed to be uniform—a constraint in the full-core model used for neutronic analysis. A single-element model was utilized to quantify the impact of axial and platewise burnup on the power distribution within the NBSR fuel elements for both high-enriched uranium (HEU) and (proposed) low-enriched uranium (LEU) fuel. To validate this approach, key parameters in the single-element model were compared to parameters from an equilibrium core model, specifically, neutron energy spectrum, power distribution, and integral 235U vector. The power distribution changes significantly when incorporating local burnup effects and has lower power peaking relative to the uniform burnup case. In the uniform burnup case, the axial relative power peaking is overpredicted by as much as 59% in the HEU single element and 46% in the LEU single element. In the uniform burnup case, the platewise power peaking is overpredicted by as much as 23% in the HEU single element and 18% in the LEU single element. The degree of overprediction increases as a function of burnup cycle, with the greatest overprediction at the end of fuel element life. However, the overprediction in local power is always conservative in terms of the minimum critical heat flux ratio, a key safety parameter that depends on the local heat flux condition. The thermal flux peak is always in the midplane gap; this causes the local cumulative burnup near the midplane gap to be significantly higher than the fuel element average. Uniform burnup distribution throughout a half-element also causes a bias in fuel element reactivity worth particularly near end of life, primarily due to the importance of the fissile inventory in the midplane gap region. Despite this bias, comparisons of cycle length exhibit very good agreement between the core model with uniform burnup and the NBSR, which has many decades of operational experience with HEU fuel.

Using the Neutron Excess Concept to Determine Starting Fuel Requirements for Minimum Burnup Breed-and-Burn Reactors

In a breed-and-burn (B&B) reactor, the reactor is first started with enriched uranium or other fissile material but thereafter can be refueled with natural or depleted uranium. B&B reactors have the potential to achieve >10% uranium utilization in a once-through fuel cycle versus <1% for light water reactors. A newly developed method for analyzing B&B reactors—the “neutron excess” concept—is used to determine the minimum amount of startup fuel needed to establish a desired equilibrium cycle in a minimum burnup B&B reactor. Here, a minimum burnup B&B reactor is defined as one in which neutron leakage is minimized and feed fuel can be discharged at uniform burnup. The neutron excess concept reformulates the k-effective of a system in terms of material depletion quantities: the total number of neutrons absorbed and produced by a given volume of fuel, which are termed “neutron excess quantities.” This concept is useful because neutron excess quantities are straightforward to estimate using simple one-dimensional (1-D) and zero-dimensional (0-D) models. A set of equations is developed that allows the quantity of starter fuel needed to establish a given B&B equilibrium cycle to be expressed in terms of neutron excess quantities. A simple 1-D example of a sodium-cooled, metal fuel reactor with a startup enrichment of 15% is used to illustrate how the method is applied. An estimate for the required amount of starter fuel based on a 0-D depletion model is found to differ by only 3% from the actual amount computed using the 1-D example model.

Small PWR core design with coated particle based fuel with a novel composition

Abstract A conceptual design study for a small-pressurized water reactor (PWR) core has been presented in this research paper that utilizes carbon coated fuel particles with a novel composition. The utilization of novel tristructural-isotropic (TRISO) fuel particles can completely eliminate the use of soluble boron system and requirement of burnable poison if adequate number of control rods is used. Apart from the simplification of the design by excluding such as complex systems from the design, the utilization of TRISO fuel particles in PWR technology can enhance the design integrity by confining the radioactive fission products within fuel itself during reactor operation. Reactor simulation computer codes WIMS-D/4 and CITATION were used for neutronics design and RELAP5 was used for thermal hydraulic analysis. The designed core is small which has only 25 MWth power out put with more than 2 EFPY operating life. The values of Doppler, moderator and void reactivity coefficients were found negative. The Soluble Boron Free (SBF) and Burnable Poison Free (BPF) concepts are more viable in small and medium size reactors (SMRs) because it makes the plant simpler and uniform burnup can also be achieved throughout the core life.

06/02129 Recursively coupled thermal and mechanical FEM–analysis of lower plenum creep failure experiments: Willschütz, H. -G. et al. Annals of Nuclear Energy, 2006, 33, (2), 126–148.

A conceptual design study for a small-pressurized water reactor (PWR) core has been presented in this research paper that utilizes carbon coated fuel particles with a novel composition. The utilization of novel tristructural-isotropic (TRISO) fuel particles can completely eliminate the use of soluble boron system and requirement of burnable poison if adequate number of control rods is used. Apart from the simplification of the design by excluding such as complex systems from the design, the utilization of TRISO fuel particles in PWR technology can enhance the design integrity by confining the radioactive fission products within fuel itself during reactor operation. Reactor simulation computer codes WIMS-D/4 and CITATION were used for neutronics design and RELAP5 was used for thermal hydraulic analysis. The designed core is small which has only 25 MWth power out put with more than 2 EFPY operating life. The values of Doppler, moderator and void reactivity coefficients were found negative. The Soluble Boron Free (SBF) and Burnable Poison Free (BPF) concepts are more viable in small and medium size reactors (SMRs) because it makes the plant simpler and uniform burnup can also be achieved throughout the core life.

Transverse buckling effects on solitary burn-up waves

A three-dimensional one-group diffusion model with explicit effects of burnup and feedback is studied for a so-called “candle reactor”. By a perturbation method the problem is reduced to a one-dimensional one, for which a solitary wave solution was obtained by van Dam (2000) [Self-stabilizing criticality waves. Annals of Nuclear Energy 27, 1505]. Therefore, such a travelling burn-up wave exists as well in the multi-dimensional problem, which has a solitary wave profile in the long axial direction and a buckling shape, as in usual reactors, in the smaller transverse direction. Two solutions are obtained and discussed: (i) a uniform drift-speed solution in the case of a burn-up parameter, i.e. the value of neutron fluence associated with the maximum of k ∞, adapted to the transverse buckling; (ii) a nonuniform drift-speed solution for uniform burn-up parameters, where the contour lines of the wave pattern are distorted from symmetric ellipsis to sickle-shaped ones. The buckling effects on relationships of principal parameters are presented.

Assessment of a Nuclear Fuel Assembly Performance with Different Inner Water Channels

Advanced fuel designs have been developed with the aim of improving fuel cycle efficiency. With this idea, moderator distribution in boiling water reactor fuel assemblies has been improved. The current work analyzes two 9×9 fuel assemblies with different inner channel designs. The first design corresponds to an actual assembly, whereas the second is proposed with the aim of making comparisons between their performances. The former design is an internal parallelepipedal water channel, and the latter is an internal cylindrical water channel whose diameter is equal to one side of the first. It is observed that the former assembly has a better burnup. Reloads for Laguna Verde Nuclear Power Plant are simulated for each design. Better operational limits are obtained by using the latter assembly. The increase in the amount of water yields a more uniform burnup, although as shown in this study, this fact does not necessarily improve the plant operational limits.