Waclaw Gudowski

The burnup capabilities of the deep burn modular helium reactor analyzed by the Monte Carlo continuous energy code MCB

In the future development of nuclear energy, the graphite-moderated helium-cooled reactors may play an important role because of their valuable technical advantages: passive safety, low cost, flexi ...

Definition and application of proton source efficiency in accelerator-driven systems

Abstract In order to study the beam power amplification of an accelerator-driven system (ADS), a new parameter, the proton source efficiency ψ* is introduced. ψ* represents the average importance of the external proton source, relative to the average importance of the eigenmode production, and is closely related to the neutron source efficiency φ*, which is frequently used in the ADS field. φ* is commonly used in the physics of subcritical systems driven by any external source (spallation source, (d,d), (d,t), 252Cf spontaneous fissions, etc.). On the contrary, ψ* has been defined in this paper exclusively for ADS studies where the system is driven by a spallation source. The main advantage with using ψ* instead of φ* for ADS is that the way of defining the external source is unique and that it is proportional to the core power divided by the proton beam power, independent of the neutron source distribution. Numerical simulations have been performed with the Monte Carlo code MCNPX in order to study ψ* as a function of different design parameters. It was found that, in order to maximize ψ* and therefore minimize the proton current needs, a target radius as small as possible should be chosen. For target radii smaller than ~30 cm, lead-bismuth is a better choice of coolant material than sodium, regarding the proton source efficiency, while for larger target radii the two materials are equally good. The optimal axial proton beam impact was found to be located ~20 cm above the core center. Varying the proton energy, ψ*/Ep was found to have a maximum for proton energies between 1200 and 1400 MeV. Increasing the americium content in the fuel decreases ψ* considerably, in particular when the target radius is large.

Stochastic Approximation for Monte Carlo Calculation of Steady-State Conditions in Thermal Reactors

Abstract A new adaptive stochastic approximation method for an efficient Monte Carlo calculation of steady-state conditions in thermal reactor cores is described. The core conditions that we consider are spatial distributions of power, neutron flux, coolant density, and strongly absorbing fission products like 135Xe. These distributions relate to each other; thus, the steady-state conditions are described by a system of nonlinear equations. When a Monte Carlo method is used to evaluate the power or neutron flux, then the task turns to a nonlinear stochastic root-finding problem that is usually solved in the iterative manner by stochastic optimization methods. One of those methods is stochastic approximation where efficiency depends on a sequence of stepsize and sample size parameters. The stepsize generation is often based on the well-known Robbins-Monro algorithm; however, the efficient generation of the sample size (number of neutrons simulated at each iteration step) was not published yet. The proposed method controls both the stepsize and the sample size in an efficient way; according to the results, the method reaches the highest possible convergence rate.

Cross sections for nuclide production in a Fe-56 target irradiated by 300, 500, 750, 1000, 1500, and 2600 MeV protons compared with data on a hydrogen target irradiated by 300, 500, 750, 1000, and 1500 MeV/nucleon Fe-56 ions

This work presents the cross sections for radioactive nuclide production in Fe-56( p, x) reactions determined in six experiments using 300, 500, 750, 1000, 1500, and 2600 MeV protons of the externa ...

Feasibility study on transient identification in nuclear power plants using support vector machines

Support vector machines (SVMs), a relatively new paradigm in statistical learning theory, are studied for their potential to recognize transient behavior of detector signals corresponding to various accident events at nuclear power plants (NPPs). Transient classification is a major task for any computer-aided system for recognition of various malfunctions. The ability to identify the state of operation or events occurring at an NPP is crucial so that personnel can select adequate response actions. The Modular Accident Analysis Program, version 4 (MAAP4) is a program that can be used to model various normal and abnormal events in an NPP. This study uses MAAP signals describing various loss-of-coolant accidents in boiling water reactors. The simulated sensor readings corresponding to these events have been used to train and test SVM classifiers. SVM calculations have demonstrated that they can produce classifiers with good generalization ability for our data. This in turn indicates that SVMs show promise as classifiers for the learning problem of identifying transients.

Application of Burnable Absorbers in an Accelerator-Driven System

Abstract The application of burnable absorbers (BAs) to minimize power peaking, reactivity loss, and capture-to-fission probabilities in an accelerator-driven waste transmutation system has been investigated. Boron-10-enriched B4C absorber rods were introduced into a lead-bismuth-cooled core fueled with transuranic (TRU) discharges from light water reactors to achieve the smallest possible power peakings at beginning-of-life (BOL) subcriticality level of 0.97. Detailed Monte Carlo simulations show that a radial power peaking equal to 1.2 at BOL is attainable using a four-zone differentiation in BA content. Using a newly written Monte Carlo burnup code, reactivity losses were calculated to be 640 pcm per percent TRU burnup for unrecycled TRU discharges. Comparing to corresponding values in BA-free cores, BA introduction diminishes reactivity losses in TRU-fueled subcritical cores by ~20%. Radial power peaking after 300 days of operation at 1200-MW thermal power was <1.75 at a subcriticality level of ~0.92, which appears to be acceptable, with respect to limitations in cladding and fuel temperatures. In addition, the use of BAs yields significantly higher fission-to-capture probabilities in even-neutron-number nuclides. Fission-to-absorption probability ratio for 241Am equal to 0.33 was achieved in the configuration studied. Hence, production of the strong alpha-emitter 242Cm is reduced, leading to smaller fuel-swelling rates and pin pressurization. Disadvantages following BA introduction, such as increase of void worth and decrease of Doppler feedback in conjunction with small values of βeff, need to be addressed by detailed studies of subcritical core dynamics.

Measurement and simulation of the cross sections for nuclide production in natW and 181Ta targets irradiated with 0.04- to 2.6-GeV protons

The cross sections for nuclide production in thin natWand 181Ta targets irradiated by 0.04–2.6-GeV protons have been measured by direct γ spectrometry using two γ spectrometers with the resolutions of 1.8 and 1.7 keV in the 60Co 1332-keV γ line. As a result, 1895 yields of radioactive residual product nuclei have been obtained. The 27Al(p, x)22Na reaction has been used as a monitor reaction. The experimental data have been compared with the MCNPX (BERTINI, ISABEL), CEM03.02, INCL4.2, INCL4.5, PHITS, and CASCADE07 calculations.

Measurement and simulation of the cross sections for nuclide production in 93Nb and natNi targets irradiated with 0.04- to 2.6-GeV protons

The cross sections for nuclide production in thin 93Nb and natNi targets irradiated by 0.04- to 2.6-GeV protons have been measured by direct γ spectrometry using two γ spectrometers with the resolutions of 1.8 and 1.7 keV in the 60Co 1332-keV γ line. As a result, 1112 yields of radioactive residual nuclei have been obtained. The 27Al(p, x)22Na reaction has been used as a monitor reaction. The experimental data have been compared with the MCNPX (BERTINI, ISABEL), CEM03.02, INCL4.2, INCL4.5, PHITS, and CASCADE07 calculations.

Verification of high-energy transport codes on the basis of activation data

High-energy transport codes (HETCs) based on various versions of nuclear reaction models [generally, an intranuclear cascade model (INC) followed by different deexcitation models] are widely used in many nuclear centers for the analysis of experimental data, for calculations of nuclear accelerator shielding, and for design of new nuclear facilities. Each of the available codes was developed originally for well-specified tasks and the model parameters were estimated on the basis of fitting the corresponding data. With an expansion of tasks,

Comparative Studies of ENDF/B-6.8, JEF-2.2 and JENDL-3.2 Data Libraries by Monte Carlo Modeling of High Temperature Reactors on Plutonium Based Fuel Cycles

We performed a numerical comparative analysis of the burnup capability of the Gas Turbine-Modular Helium Reactor (GT-MHR) by the Monte Carlo Continuous Energy Burnup Code (MCB). The MCB code is an extension of MCNP that includes the burnup implementation; it adopts continuous energy cross sections and it evaluates the transmutation trajectories for over 2,400 decaying nuclides. We equipped the MCB code with three different nuclear data libraries: JENDL-3.2, JEF-2.2 and ENDF/B-6.8 processed for temperatures from 300 to 1,800K. The GT-MHR model studied in this paper is fueled by actinides coming from the Light Water Reactors waste, converted into two different types of fuel: Driver Fuel and Transmutation Fuel. The Driver Fuel supplies the fissile nuclides needed to maintain the criticality of the reactor, whereas the Transmutation Fuel depletes non-fissile isotopes and controls reactivity excess. We set the refueling and shuffling period to one year and the in-core fuel residency time to three years. The comparative analysis of the MCB code consists of accuracy and precision studies. In the accuracy studies, we performed the burnup calculation with different nuclear data libraries during the year at which the refueling and shuffling schedule set the equilibrium of the fuel composition. In the precision studies, we repeated the same simulations 20 times with a different pseudorandom number stride and the same nuclear data library.