Muitiobjective pressurized water reactor reload core design by nondominated genetic algorithm search

The design of pressurized water reactor reload cores is not only a formidable optimization problem but also, in many instances, a multiobjective problem. A genetic algorithm (GA) designed to perform true multiobjective optimization on such problems is described. Genetic algorithms simulate natural evolution. They differ from most optimization techniques by searching from one group of solutions to another, rather than from one solution to another. New solutions are generated by breeding from existing solutions. By selecting better (in a multiobjective sense) solutions as parents more often, the population can be evolved to reveal the trade-off surface between the competing objectives. An example illustrating the effectiveness of this novel method is presented and analyzed. It is found that in solving a reload design problem the algorithm evaluates a similar number of loading patterns to other state-of-the-art methods, but in the process reveals much more information about the nature of the problem being solved. The actual computational cost incurred depends: on the core simulator used; the GA itself is code independent.

Analysis of SEB and SEGR in super-junction MOSFETs

The electric field distribution in the super junction power MOSFET is analyzed using analytical modeling and numerical simulations in this paper. The single-event burn-out (SEB) and single-event gate rupture (SEGR) phenomena in this device are studied in detail. It is demonstrated that the super junction device is much less sensitive to SEB and SEGR compared to the standard power MOSFET. The physical mechanism is explained.

Detailed design of242mAm breeding in pressurized water reactors

There is growing interest in the use of 242mAm as a nuclear fuel. Because of its very high thermal fission cross section and its large number of neutrons released per fission, it can be used for various unique applications, such as space propulsion, medical applications, and compact energy sources. Since the thermal absorption cross section of 242mAm is very high, the best way to obtain 242mAm is by the capture of fast or epithermal neutrons in 241Am. However, fast spectrum reactors are not readily available. In this paper, we explore the possibility of producing 242mAm in existing pressurized water reactors (PWRs) with minimal interference in reactor performance. As suggested in previous studies on the subject, the 242mAm breeding targets are shielded with strong thermal absorbers in order to suppress the thermal neutron flux that causes 242mAm destruction. Since 242mAm enrichment within the Am target mainly depends on the neutron energy distribution, which in turn depends on the Am target thickness and on the neutron filter cutoff energy (thermal absorber type), this unique Am target design was developed. In our study, Cd, Sm, and Gd were considered as thermal neutron filters, as suggested by Cesana et al. The most favorable results were obtained by irradiating Am targets covered either with Gd or Cd. In these cases, up to 8.65% enrichment of 242mAm is obtained after 4.5 yr (three successive PWR fuel cycles) of irradiation. It was also found that significant quantities [up to 1.3 kg/GW (electric)-yr] of 242mAm can be obtained in PWR reactors without notable interference with reactor performance. However, in order to maintain the original fuel cycle length, the enrichment of the driver (UO2) fuel must be increased by ∼1%, raised from the conventional 4.5 to 5.5%, depending on the thermal neutron filter used. The most important reactivity feedback coefficients for fuel assemblies containing the 242mAm breeding targets were evaluated and found to be close to those of a standard PWR. Another product of neutron capture in the 241Am reaction is 238Pu. It was found that in a typical 1000 MW (electric) PWR core with one-third of the fuel assemblies containing 241Am targets, up to 15.1 kg of 238Pu enriched to 80% can be produced per year.

Effect of95 241Am(n, γ) reaction branching ratio on fuel cycle and reactor design characteristics

The growing interest in innovative reactors and advanced fuel cycle designs requires more accurate prediction of various transuranic actinide concentrations during irradiation or following discharge because of their effect on reactivity or spent-fuel emissions, such as gamma and neutron activity and decay heat. In this respect, many of the important actinides originate from the 241Am(n,γ) reaction, which leads to either the ground or the metastable state of 242Am. The branching ratio for this reaction depends on the incident neutron energy and has very large uncertainty in the current evaluated nuclear data files. This study examines the effect of accounting for the energy dependence of the 241Am(n,γ) reaction branching ratio calculated from different evaluated data files for different reactor and fuel types on the reactivity and concentrations of some important actinides. The results of the study confirm that the uncertainty in knowing the 241Am(n,γ) reaction branching ratio has a negligible effect on the characteristics of conventional light water reactor fuel. However, in advanced reactors with large loadings of actinides in general, and 241Am in particular, the branching ratio data calculated from the different data files may lead to significant differences in the prediction of the fuel criticality and isotopic composition. Moreover, it was found that neutron energy spectrum weighting of the branching ratio in each analyzed case is particularly important and may result in up to a factor of 2 difference in the branching ratio value. Currently, most of the neutronic codes have a single branching ratio value in their data libraries, which is sometimes difficult or impossible to update in accordance with the neutron spectrum shape for the analyzed system.

Efficient generation of one-group cross sections for coupled Monte Carlo depletion calculations

Coupled Monte Carlo depletion systems provide a versatile and an accurate tool for analyzing advanced thermal and fast reactor designs for a variety of fuel compositions and geometries. The main drawback of Monte Carlo-based systems is a long calculation time imposing significant restrictions on the complexity and amount of design-oriented calculations. This paper presents an alternative approach to interfacing the Monte Carlo and depletion modules aimed at addressing this problem. The main idea is to calculate the one-group cross sections for all relevant isotopes required by the depletion module in a separate module external to Monte Carlo calculations. Thus, the Monte Carlo module will produce the criticality and neutron spectrum only, without tallying of the individual isotope reaction rates. The onegroup cross section for all isotopes will be generated in a separate module by collapsing a universal multigroup (MG) cross-section library using the Monte Carlo calculated flux. Here, the term "universal" means that a single MG cross-section set will be applicable for all reactor systems and is independent of reactor characteristics such as a neutron spectrum; fuel composition; and fuel cell, assembly, and core geometries. This approach was originally proposed by Haeck et al. and implemented in the ALEPH code. Implementation of the proposed approach to Monte Carlo burnup interfacing was carried out through the BGCORE system. One-group cross sections generated by the BGCORE system were compared with those tallied directly by the MCNP code. Analysis of this comparison was carried out and led to the conclusion that in order to achieve the accuracy required for a reliable core and fuel cycle analysis, accounting for the background cross section (σ0) in the unresolved resonance energy region is essential. An extension of the one-group cross-section generation model was implemented and tested by tabulating and interpolating by a simplified σ0 model. A significant improvement of the one-group cross-section accuracy was demonstrated.

Use of axially graded burnable boron for hot-spot temperature reduction in a pressurized water reactor core

Shortly after the loading of a pressurized water reactor (PWR) core, the axial power distribution in fresh fuel has already attained the characteristic, almost flat shape, typical of burned fuel. At beginning of cycle (BOC), however, the axial distribution is centrally peaked. In assemblies hosting uniform burnable boron rods, this BOC peaking is even more pronounced. A reduction in the axial peaking is today often achieved by shortening the burnable boron rods by some 30 cm at each edge. It is shown that a two-zone grading of the boron rod leads, in a representative PWR cycle, to a reduction of the hot-spot temperature of approximately 70 °C, compared with the nongraded case. However, with a proper three-zone grading of the boron rod, an additional 20 °C may be cut off the hot-spot temperature. Further, with a slightly skewed application of this three-zone grading, an additional 50 °C may be cut off. The representative PWR cycle studied was cycle 11 of the Indian Point 2 station, with a simplification in the number of fuel types and in the burnup distribution. The analysis was based on a complete three-dimensional burnup calculation. The code system was ELCOS, with BOXER as an assembly code for the generation of burnup-dependent cross sections and SILWER as a three-dimensional core code with thermal-hydraulic feedback.

Sedimentation rates, nitrogen and phosphorus retentions in the largest urban Lake Donghu, China

Sediment core samples were collected in the largest urban Lake Donghu (Stations I and II) in China, and the activities of Pb-210, Ra-226 and Cs-137 were measured by gamma-ray spectrometry. The sedimentation rates, calculated by 210Pb constant rate of supply (CRS) model, ranged from 0.11 to 0.65 (average 0.39) cm(.)y(-1) at Station I, and from 0.21 to 0.78 (average 0.46) cm(.)y(-1) at Station II. Sedimentation rate calculated by Cs-137 as a time marker was 0.55 cm(.)y(-1) at Station II. Based on the average sedimentation rate, we obtained 769 and 147 t(.)y(-1) for nitrogen and phosphorus retentions in Lake Donghu sediments, respectively.

Design, simulation and testing of large area silicon drift detectors and detector array for X-ray spectroscopy

Large area (25 mm(2)) silicon drift detectors and detector arrays (5x5) have been designed, simulated, and fabricated for X-ray spectroscopy. On the anode side, the hexagonal drift detector was designed with self-biasing spiral cathode rings (p(+)) of fixed resistance between rings and with a grounded guard anode to separate surface current from the anode current. Two designs have been used for the P-side: symmetric self-biasing spiral cathode rings (p(+)) and a uniform backside p(+) implant. Only 3 to 5 electrodes are needed to bias the detector plus an anode for signal collection. With graded electrical potential, a sub-nanoamper anode current, and a very small anode capacitance, an initial FWHM of 1.3 keV, without optimization of all parameters, has been obtained for 5.9 keV Fe-55 X-ray at RT using a uniform backside detector.

Improved neutron radiation hardness for Si detectors: Application of low resistivity starting material and or manipulation of N-eff by selective filling of radiation-induced traps at low temperatures

Radiation-induced electrical changes in both space charge region (SCR) of Si detectors and bulk material (BM) have been studied for samples of diodes and resistors made on Si materials with different initial resistivities. The space charge sign inversion fluence (Phi(inv)) has been found to increase linearly with the initial doping concentration (the reciprocal of the resistivity), which gives improved radiation hardness to Si detectors fabricated from low resistivity material. The resistivity of the BM, on the other hand, has been observed to increase with the neutron fluence and approach a saturation value in the order of hundreds k Omega cm at high fluences, independent of the initial resistivity and material type. However, the fluence (Phi(s)), at which the resistivity saturation starts, increases with the initial doping concentrations and the value of Phi(s) is in the same order of that of Phi(inv) for all resistivity samples. Improved radiation hardness can also be achieved by the manipulation of the space charge concentration (N-eff) in SCR, by selective filling and/or freezing at cryogenic temperatures the charge state of radiation-induced traps, to values that will give a much smaller full depletion voltage. Models have been proposed to explain the experimental data.

Microscopic analysis of defects in a high resistivity silicon detector irradiated to 1.7x10(15)n/cm(2)

Current-based microscopic defect analysis methods with optical filling techniques, namely current deep level transient spectroscopy (I-DLTS) and thermally stimulated current (TSC), have been used to study defect levels in a high resistivity silicon detector (p(+)-n-n(+)) induced by very high fluence neutron (VHFN) irradiation (1.7x10(15) n/cm(2)). As many as fourteen deep levels have been detected by I-DLTS. Arrhenius plots of the I-DLTS data have shown defects with energy levels ranging from 0.03 eV to 0.5 eV in the energy band gap. Defect concentrations of relatively shallow levels (E(t) < 0.33 eV) are in the order of 10(13)cm(-3), while those for relatively deep levels (E(t) > 0.33 eV) are in the order of 10(14) cm(-3). TSC data have shown similar defect spectra. A full depletion voltage of about 27,000 volts has been estimated by C-V measurements for the as-irradiated detector, which corresponds to an effective space charge density (N-eff) in the order of 2x10(14) cm(-3). Both detector leakage current and full depletion voltage have been observed to increase with elevated temperature annealing (ETA). The increase of the full depletion voltage corresponds to the increase of some deep levels, especially the 0.39 eV level. Results of positron annihilation spectroscopy have shown a decrease of total concentration of vacancy related defects including vacancy clusters with ETA, suggesting the breaking up of vacancy clusters as possible source of vacancies for the formation of single defects during the reverse anneal.

STUDY OF THE LONG-TERM STABILITY OF THE EFFECTIVE CONCENTRATION OF IONIZED SPACE CHARGES (N-EFF) OF NEUTRON-IRRADIATED SILICON DETECTORS FABRICATED BY VARIOUS THERMAL-OXIDATION PROCESSES

Experimental study of the reverse annealing of the effective concentration of ionized space charges (N-eff, also called effective doping or impurity concentration) of neutron irradiated high resistivity silicon detectors fabricated on wafers with various thermal oxides has been conducted at room temperature (RT) and elevated temperature (ET). Various thermal oxidations with temperatures ranging from 975 degrees C to 1200 degrees C with and without trichlorethane (TCA), which result in different concentrations of oxygen and carbon impurities, have been used. It has been found that, the RT annealing of the N-eff is hindered initially (t < 42 days after the radiation) for detectors made on the oxides with high carbon concentrations, and there was no carbon effect on the long term (t > 42 days after the radiation) N-eff reverse annealing. No apparent effect of oxygen on the stability of N-eff has been observed at RT. At elevated temperature (80 degrees C), no significant difference in annealing behavior has been found for detectors fabricated on silicon wafers with various thermal oxides. It is apparent that for the initial stages (first and/or second) of N-eff reverse annealing, there may tie no dependence on the oxygen and carbon concentrations in the ranges studied.