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research-article  
Oriol Costa Garrido, Samir El Shawish and Leon Cizelj
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4036736
Large sets of fluid temperature histories and a recently proposed thermal fatigue assessment procedure are employed in this paper to deliver more accurate statistics of predicted lives of pipes and their uncertainties under turbulent fluid mixing circumstances. The wide variety of synthetic fluid temperatures, generated with an improved spectral method, results in a set of estimated distributions of fatigue lives through linear one-dimensional heat diffusion, thermal stress estimates and fatigue assessment codified rules. The results of the fatigue analysis indicate that, in order to avoid the inherent uncertainties due to comparatively short fluid temperature histories to the estimated fatigue lives, a conservative safe design against thermal fatigue could be attempted with the lower bounds of the predicted life distributions, such as the 5% failure probability life (5% of samples fail). The impact of the convection heat transfer coefficient on the predictions is also studied in a sensitivity analysis. This represents a detailed attempt to correlate the uncertainties in the physical fluid mixing conditions and heat transfer to the estimated fatigue life using spectral methods.
TOPICS: Fluids, Turbulence, Pipes, Fatigue life, Fatigue, Temperature, Uncertainty, Statistics as topic, Probability, Sensitivity analysis, Thermal diffusion, Heat transfer, Failure, Fatigue analysis, Thermal stresses, Convection, Design
research-article  
A. Gad-Briggs, P. Pilidis and T. Nikolaidis
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4036737
An important requirement for Generation IV Nuclear Power Plant (NPP) design is the control system, which enables part power operability. The choices of control system methods must ensure variation of load without severe drawbacks on cycle performance. The objective of this study is to assess the control of the NPP under part power operations. The cycles of interest are the Simple Cycle Recuperated (SCR) and the Intercooled Cycle Recuperated (ICR). Control strategies are proposed for NPPs but the focus is on the strategies that result in part power operation using the inventory control method. Firstly, results explaining the performance and load limiting factors of the inventory control method are documented; subsequently, the transient part power performances. The load versus efficiency curves were also derived from varying the load to understand the efficiency penalties. This is carried out using a modelling and performance simulation tool designed for this study. Results show that the ICR takes ~102% longer than the SCR to reduce the load to 50% in Design Point (DP) performance conditions for similar valve flows, which correlates to the volumetric increase for the ICR inventory tank. The efficiency penalties are comparable for both cycles at 50% part power, whereby a 22% drop in cycle efficiency was observed and indicates limiting time at very low part power. The analyses intend to aid the development of cycles for Generation IV NPPs specifically Gas Cooled Fast Reactors (GFRs) and Very High Temperature Reactors (VHTRs), where helium is the coolant.
TOPICS: Control systems, Gas turbines, Cycles, Helium, Nuclear power stations, Stress, Design, Very high temperature reactors, Fast neutron reactors, Flow (Dynamics), Modeling, Valves, Coolants, Transients (Dynamics), Simulation
research-article  
Aharon Ocherashvili, Tatjana Bogucarska, Arie Beck, Guy Heger, Marita Mosconi, Eric Roesgen, Jean-Michel Crochemore, Valeri Maiorov, Giovanni Varasano and Bent Pedersen
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4036698
We present a method with potential for the detection of special nuclear materials (SNM) in shielded containers which is both sensitive and applicable under field conditions. The method uses an external pulsed neutron source to induce fission in SNM and subsequent detection of the fast prompt fission neutrons. The detectors surrounding the container under investigation are liquid scintillation detectors able to distinguish gamma rays from fast neutrons by means of the pulse shape discrimination method (PSD). One advantage of these detectors, besides the ability for PSD analysis, is that the analogue signal from a detection event is of very short duration (typically few tens of nanoseconds). This allows the use of very short coincidence gates for the detection of the prompt fission neutrons in multiple detectors while benefiting from a low background coincidence rate yielding a low detection limit. Another principle advantage of this method derives from the fact that the external neutron source is pulsed. By proper time gating the interrogation can be conducted by epi-thermal source neutrons only. These source neutrons do not appear in the fast neutron signal following the PSD analysis thus providing a fundamental method for separating the interrogating source neutrons from the sample response in form of fast fission neutrons. The paper describes laboratory tests with a configuration of eight detectors in the Pulsed Neutron Interrogation Test Assembly (PUNITA). Both the photon and neutron signature for induced fission is observed, and the methods used to isolate these signatures is described and demonstrated.
TOPICS: Nuclear fission, Neutron sources, Neutrons, Sensors, Containers, Signals, Gamma rays, Photons, Scintillation counters, Manufacturing, Shapes
research-article  
Hiroo Kondo, Takuji Kanemura, Tomohiro Furukawa, Yasushi Hirakawa, Eiichi Wakai and Juan Knaster
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4036513
A liquid Li jet flowing at 15 m/s under a high vacuum of 10^-3 Pa is intended to serve as a beam target (Li target) in the planned International Fusion Materials Irradiation Facility (IFMIF). The engineering validation and engineering design activities (EVEDA) for the IFMIF are being implemented under the Broader Approach agreement. As a major activity of the Li target facility, the EVEDA Li test loop was constructed and a stable Li target was demonstrated. This study focuses on a cavitation-like acoustic noise detected in a downstream conduit where the Li target flowed under vacuum conditions. This noise was investigated using acoustic-emission (AE) sensors installed via acoustic wave guides. The sound intensity of the noise was examined against the cavitation number of the Li target. In addition, fast Fourier transform (FFT) and continuous wavelet transform (CWT) were performed to characterize the acoustic noise. Owing to the acoustic noise’s intermittency, high frequency, and the dependence on cavitation number, we conclude that this acoustic noise is generated when cavitation bubbles collapse. The location of the cavitation was fundamental for presuming the mechanism. In this study, the propagation of acoustic waves was used to localize the cavitation and a method to determine the location of cavitation was formulated. As a result, we found that cavitation occurred only in a narrow area where the Li target impinged on the downstream conduit; therefore, we concluded that this cavitation was induced by the impingement.
TOPICS: Irradiation (Radiation exposure), Cavitation, Lithium, Acoustics, Noise (Sound), Vacuum, Sensors, Waveguides, Wavelet transforms, Acoustic emissions, Acoustic intensity, Collapse, Fast Fourier transforms, Waves, Bubbles, Engineering design
Technical Brief  
A. Ben-Shlomo, G. Bartal, M. Mosseri and S. Shabat
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4036462
PURPOSE: The study aimed to determine how lumbar spine x-ray examinations effective dose changes at different patient positioning considering the x-ray tube heel effect. METHODS: The study used exposure of patient phantoms and Monte Carlo simulation of the effective doses. Results and conclusions: Using heel effect phenomenon the head to anode positioning of the patient reduces the effective dose by 5.0% when compared with the head to cathode direction and the same exposure.
TOPICS: X-rays, Diagnostic radiography, Lumbar spine, Phantoms, Anodes, Simulation
Technical Brief  
Jean Koch
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4036463
In Israel, a single regulatory body for radiation protection does not exist. Instead, its responsibilities and functions are shared between five government ministries and agencies. Accordingly, the existing legal framework for radiation safety is of a very heterogeneous nature. It is made of laws, acts, orders and regulations enacted during different periods, according to different principles. Moreover, some of the provisions of those legal instruments are obsolete or quote obsolete documents. The Standard for Radiation Protection (SRP) of the Israel Atomic Energy Commission (IAEC) was recently updated on the basis of the latest version of the IAEA International Basic Safety Standards (BSS). It is proposed that the SRP of the IAEC serve as a model for a comprehensive framework law that would be structured similarly, i.e. a division into three parts according to the three different types of exposure situation (planned, emergency, existing) defined by the ICRP and a subdivision of each part according to relevant exposure categories (occupational, public, medical). The adoption of such a structure would ensure that no aspect of radiation protection is left untreated. Furthermore, it would imply either creating a central regulatory body or strengthening coordination between the existing bodies.
TOPICS: Radiation (Physics), Safety, Instrumentation, Nuclear power, Regulations, Biomedicine, Emergencies, Governments
research-article  
Ilan Yaar, Rony Hakmon, Itzhak Halevy, Ronen Bar-Ziv, Noah Vainblat, Yacov Iflach, Maor Assulin, Tzipora Avraham, Michael D. Kaminski, Terry Stilman and Serre Shannon
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4036458
One of the preparation steps for a possible radiological attack, is the capability of fast and effective decontamination of critical infrastructures. This study describes the implementation of a test plan at an intermediate scale (between bench scale and large scale or wide area) to evaluate decontamination procedures, materials, technologies, and techniques for removal of radioactive material from various surfaces. Two radioisotopes were tested: cesium-137 (137Cs) and the short-lived simulant to 137Cs, rubidium-86 (86Rb). Two types of decontamination hydrogel products were evaluated: DeconGelTM and Argonne SuperGel. Tests were conducted at the assigned Chemical, Biological, Radiological, and Nuclear (CBRN) Israel Defense Forces (IDF) Home Front Command facility, and at the Nuclear Research Center Negev (NRCN), Israel. Results from these tests indicated similar removal and operational parameters for 86Rb and 137Cs, as expected from the chemical similarity of both elements. These results proved that the short-lived radioisotope 86Rb can be used in future experiments to simulate 137Cs. Results and conclusions from these experiments are presented and compared to results from laboratory-scale experiments performed on small coupons. In general, both hydrogel decontamination products may be used as a viable option to decontaminate large surfaces in a real-world event, removing between 30-90% of the contamination, depending on the surface type and porosity. However, both products may leave behind absorbed contamination, that will need to be addressed at a later stage. Yet, the likelihood of resuspension through use of these products is reduced.
TOPICS: Decontamination, Hydrogels, Building materials, Contamination, Radioisotopes, Materials science, Radioactive substances, Nuclear research, Defense industry, Porosity
Editorial  
Ilan Yaar and Jean Koch
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4036459
TOPICS: Nuclear engineering, Radiation (Physics)
research-article  
Aric Katz, Avraham Shtub and Ariel Roguin
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4036431
Advanced imaging systems, such as C-Arm machines, greatly improve physicians' diagnostic abilities and provide greater precision. Yet, these benefits come with a price of ionizing radiation exposure to medical teams and patients. Exposure to ionizing radiation can lead to physical maladies ranging from skin irritation to cancerous growths. Supplying proper training and skill improvement to operators on how to use this technology safely can help minimize risk of exposure. Previous studies on radiation knowledge among physicians and radiologists presented disturbing results of underestimated risk of exposure. This research is based on an innovation in Simulation-Based Training (SBT), a simulator using the WOZ (Wizard of Oz) concept that was used for training ER physicians and ultrasound technicians. This research integrated WOZ technology with a radiation exposure formula for training how to minimize unnecessary exposure to ionizing radiation. The simulator also incorporates 3D animation graphics, enabling trainees to simulate the control of different factors. Image quality and the operator's radiation exposure levels are also animated, assisting trainees to focus on their exposure based on their device settings. Contrary to previous studies we measured exposure doses to the operator and quantified image quality accordingly. Validation was done on different C-Arm machines. We also built a unique exposure formula and integrated it into our WOZ simulator enabling trainees to visualize their real-time and overall exposure based on their technique. Validation of learning outcomes was done using knowledge exams. Results from our validation exams presented significant improvement and high retention.
TOPICS: Ionizing radiation, Safety, Cardiology, Biomedicine, Radiation (Physics), Machinery, Risk, Innovation, Skin, Teams, Imaging, Simulation, Ultrasound, Performance
Technical Brief  
Magal Saphier, Oron Zamir, Polina Berzansky, Oshra Saphier and Dan Meyerstein
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4036432
The reaction of fluoride ions with alumina was found to strongly depend on the concentration of fluoride ions in the aqueous solution. At low concentrations ([F-]<0.1M in the case of KF), the aqueous concentration of aluminum ions is relatively high (ICP measurements), and the aluminum oxide dissolves as a fluoride complex. At high concentrations of fluoride ([F-]>0.5M in the case of KF), a new structure is formed on the alumina surface involving fluoride, aluminum, potassium ions (in the case of KF) and oxygen. The structure is characterized by XRD, SEM and EDS. The resulting structure improved the passivation of alumina, the solubility of aluminum ions decreasing compared to the “untreated” alumina. Aluminum surfaces “fluoride treated” showed a better resistance to dissolution in both acidic and basic media
TOPICS: Ions, Aluminum, Corrosion, Oxygen, Potassium
research-article  
Dimitry Ginzburg, Sophia Lantsberg and Sergio Faermann
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4036433
Hyperthyroid and Thyroid Carcinoma patients treated with radioactive 131I (RAI) are a potential source of external and internal exposure to members of the public, to medical staff and especially to family members, who are in close contact with these patients. The relationship between radiation dose rates and various clinical parameters, including gender, age, thyroid size and weight and iodine uptake, was assessed. Dose rates were measured on 8 patients with Thyroid Carcinoma after total or subtotal thyroidectomy and on 6 patients with Hyperthyroidism. All measurements were taken at 1 meter from the patient on two levels – anterior to the neck, and body center. Dose rates were measured at three or four times – at the time of RAI administration, and after 24 h, 48 h and 168 h. Based on these measurements, retention curves were obtained. The effective half-life was derived by fitting an exponential equation and was estimated to be 12.8 h.
TOPICS: Weight (Mass), Half-life, Radiation (Physics), Fittings, Biomedicine
research-article  
Aryeh M. Weiss, Itzhak Halevy, Naida Dzigal, Ernesto Chinea-Cano and Uri Admon
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4036434
Detection of microscopic fission track (FT) star-shaped clusters, developed in SSNTD by etching, created by fission fragments emitted from fissile particles irradiated by neutrons, is a key technique in nuclear forensics and safeguards investigation. It involves scanning and imaging a large area, typically 1-2 sq.cm, of a translucent SSNTD (e.g. polycarbonate sheet, mica, etc.) to identify the FT clusters, sparse as they may be, that must be distinguished from dirt and other artefacts present in the image. This task, if done manually, is time consuming, operator dependent, and prone to human errors. To solve the problem, an automated workflow have been developed for (a) scanning large area detectors, in order to acquire large images with adequate high resolution, and (b) an image processing scheme, implemented in ImageJ, to automatically detect the FT clusters. The scheme combines intensity-based segmentation approaches, a morphological algorithm capable of detecting and counting endpoints in putative FT clusters and thus enables rejection of non-FT artefacts. In this paper, the methodology is described and first very promising results shown.
TOPICS: Microscopy, Nuclear fission, Neutrons, Sensors, Particulate matter, Polycarbonate sheet, Resolution (Optics), Algorithms, Errors, Etching, Image processing, Image segmentation, Workflow, Imaging, Forensic sciences
research-article  
Yuval Ben Galim, Raymond Moreh and Itzhak Orion
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4036435
A multiline neutron source can be produced by using a metallic 232Th filter in conjunction with a white neutron source. The multiline spectrum consists of ~ 20 relatively strong intensity lines ranging from 10 to 4000 eV. It is shown that the width of each neutron line of the spectrum is strongly dependent on the absorber thickness. This neutron source is useful for accurate cross section measurements with precise neutron energies. The optimum thickness of the 232Th absorber which was found to yield a sharp multiline spectrum throughout the above energy range was found to be ~ 14 cm.
TOPICS: Optimization, Filters, Neutron sources, Neutrons
Technical Brief  
Jordan G. Gilbert, Scott B. Nokleby and Ed Waller
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4036354
Inspections of pressure tubes in CANDU reactors are a key part of maintaining safe operating conditions. The current inspection system, the Channel Inspection and Gauging Apparatus for Reactors (CIGAR), performs the job well but is limited by the fact that it can only inspect one channel at a time. A multidisciplinary team is currently developing a novel robotic inspection system. As part of this work, a Monte Carlo N-Particle (MCNP) model has been developed in order to predict the dose rates that the improved inspection system will be exposed to and, from this, predict the component lifetime. This MCNP model will be capable of predicting in-core dose rates at any location within the reactor, and as such could be used for other situations where the in-core dose rate needs to be know.Based on estimates from this model, it is expected that at 7 days after shutdown the improved inspection system could survive in core for approximately 7 hours, providing it uses a tungsten shield 2:5 cm in thickness around the integrated circuit components. This is expected to be sufficient to perform a single inspection. 1 Introduction and Background
TOPICS: Pressure, Inspection, Particulate matter, Simulation, Robotics, Integrated circuits, Teams, Tungsten
research-article  
Govert de with
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4036322
Fly ash is widely used as a supplementary cementitious material in the production of cement and concrete, and improves durability and strength of the concrete. However, as for all materials of mineral origin fly ash is a source for natural radioactivity; hence, its need for responsible use. The aim of this study is to investigate the radiation impact from fly ash as an additive to concrete compared against concrete without fly ash. For this purpose eight concrete mixtures are experimentally tested, followed by a computation of the radiation dose when used as bulk material in building constructions. The results demonstrate an increase in the total radiation dose from around 0.8 mSv with no fly ash up to 0.92 mSv when fly ash is used. The increase mostly comes from external radiation, while the radon exhalation factor reduces and sometimes even reduces the radon dose despite the higher radium content. The work has demonstrated that the impact from fly ash on the radiation exposure is limited when applied as a supplementary cementitious material. At the same time fly ash provides real benefits to the quality and durability of the concrete. For this reason exemption strategies for such applications should be developed.
TOPICS: Concretes, Radiation (Physics), Fly ash, Durability, Computation, Radioactivity, Radium, Cements (Adhesives), Bulk solids, Minerals
Technical Brief  
Eyal Peri, Adi Abraham, Tuvia Kravchik, Marcelo Weinstein, Dani Sattinger and Omer Pelled
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4036137
The ICRP in its statement on tissue reaction from April 2011 (?1) recommended to reduce the annual dose limit to the lens of the eye from 150 mSv to 20 mSv (averaged over five years), with no single year dose exceeding 50 mSv. IAEA TECDOC 1731 (?2) and ISO 15382 (?3) were published as guidelines to the implementation of the new limits for occupational radiation. The most accurate way to determine the dose to the lens of the eye is to use a dosimeter that is designed and calibrated for measuring Hp(3), but it will take some time (probably a couple of years) until such dosimetric system could be validated and implemented in routine monitoring. Meanwhile, in some cases, there is a need to estimate Hp(3), especially for retrospective reconstruction of dose to the lens of the eye. Therefore, in the following years, the measured values of Hp(10) and or Hp(0.07) can be used to calculate a conservative value for Hp(3). The present paper discusses a new, more accurate and less conservative way to estimate Hp(3) using Hp(0.07) and Hp(10) quantities. This new method could be used also in reporting historical personal lens of the eye doses, when Hp(3) dosimeters were not used, and in some cases it could reduce the need to use special Hp(3) dosimeters in the future.
TOPICS: Lenses (Optics), Dosimeters, Radiation (Physics), Biological tissues
research-article  
Abhishek Kumar Srivastava, Rakesh Chouhan, Ananta Borgohain, S. S. Jana, N. K. Maheshwari, D.S. Pilkhwal, A. Rama Rao, K. N. Hareendhran, S. Chowdhury, K. B. Modi, S. K. Raut and S. C. Parida
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4036027
Conceptual Molten Salt Breeder Reactor (MSBR) is under development in BARC with longterm objective of utilizing abundant thorium available in India. It is based on molten salts, which acts as fuel, blanket and coolant for the reactor. LiF-ThF4 (77.6-22.4 % mole) is proposed as a blanket salt for Indian MSBR. A laboratory scale molten salt natural circulation loop named, Molten Active Fluoride salt Loop (MAFL) has been setup for thermal-hydraulic, material compatibility and chemistry control studies. Various steady states and transient experiments have been performed in the operating temperature range of 600oC to 750oC. The loop operates in the power range of 250 W to 550 W. Steady state correlation given for natural circulation flow in a loop is compared with the steady state experimental data. The Reynolds number was found to in the range of 57 to 114. CFD simulation has also been performed for the same using OpenFOAM code and the results are compared with the experimental data generated in the loop. It has been found that the predictions of OpenFOAM are in good agreement with the experimental data. In this paper, features of the loop, its construction, the experimental and theoretical studies performed are discussed in detail.
TOPICS: Breeder reactors, Steady state, Operating temperature, Flow (Dynamics), Fuels, Reynolds number, Simulation, Construction, Coolants, Transients (Dynamics), Computational fluid dynamics, Chemistry
research-article  
Kurt E. Harris, Kevin J. Schillo, Yayu M. Hew, Akansha Kumar and Steven D. Howe
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4035974
In NASA's Design Reference Architecture 5.0 (DRA 5.0), fission surface power systems (FSPS) are described as “enabling for the human exploration of Mars”. This study investigates the design of a power conversion system (PCS) based on supercritical CO2 (S-CO2) Brayton configurations for a growing Martian colony. Various configurations utilizing regeneration, intercooling, and reheating are analyzed. A model to estimate the mass of the PCS is developed and used to obtain a realistic mass-optimized configuration. This mass model is conservative, being based on simple concentric tube counterflow heat exchangers and published data regarding turbomachinery masses. For load following and redundancy purposes, the FSPS consists of three 333 kWe reactors and PCS to provide a total of 1MWe for 15 years. The optimal configuration is a S-CO2 Brayton cycle with 60% regeneration and two stages of intercooling. Analyses are mostly performed in MATLAB, with certain data provided by a COMSOL model of part of a low-enriched uranium (LEU) ceramic metallic (CERMET) reactor core.
TOPICS: Nuclear fission, Optimization, Power conversion systems, Brayton cycle, Supercritical carbon dioxide, Design, Heat exchangers, Matlab, Turbomachinery, Uranium, Power systems (Machinery), Ceramics, Cermets, Stress, Redundancy (Engineering)
research-article  
Xiaoming Chai, Xiaolan Tu, Wei Lu, Zongjian Lu, Dong Yao, Qing Li and Wenbin Wu
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4035934
Due to powerful geometry treatment capability, Method Of Characteristics (MOC) currently becomes one of best method to solve neutron transport equation. In MOC method, boundary condition treatment, complex geometry representation, and advanced acceleration method are the key techniques to develop a powerful MOC code to solve complex problem. In this paper, we developed a powerful MOC module, which can treat different boundary conditions with two methods. For problems with special border shapes and boundary condition, such as rectangle, 1/8 of square, hexagon, 1/6 of hexagon problems with reflection, rotation, and translation boundary condition, the MOC module adopts periodic tracking method. For problems with general border shapes, the MOC module use ray prolongation method. Meanwhile, graphic user interface based on CAD software is developed to generate the geometry description file. The geometry and mesh can be described and modified correctly and fast. In order to accelerate the MOC transport calculation, the Generalized Coarse Mesh Finite Differential (GCMFD) is used, which can use irregular coarse mesh diffusion method to accelerate the transport equation. The MOC module was incorporated into advanced neutronics lattice code KYLIN-2, which developed by Nuclear Power Institute of China (NPIC) and used to simulate the assembly of current PWR reactor and advanced reactors. The numerical results show that the KYLIN-2 code can be used to calculate 2D neutron transport problems accurately and fast. In future, the KYLIN-2 code will be released and gradually become the main neutron transport lattice code in NPIC.
TOPICS: Rotation, Diffusion (Physics), Neutrons, Manufacturing, Reflection, Computer-aided design, Boundary-value problems, China, Computer software, Geometry, Nuclear power, Shapes, User interfaces, Pressurized water reactors
research-article  
Ella Israeli and Erez Gilad
ASME J of Nuclear Rad Sci   doi: 10.1115/1.4035883
Novel genetic algorithms are developed by using state of-the-art selection and crossover operators, e.g., rank selection or tournament selection instead of the traditional roulette (fitness proportionate) selection operator and novel crossover and mutation operators by considering the chromosomes as permutations (which is a specific feature of the loading pattern problem). The algorithm is implemented and applied to a representative model of a modern PWR core and for a single objective fitness function, i.e., k_eff. The results obtained for some reference cases using this setup are excellent and are obtained by utilizing a tournament selection operator with a linear ranking selection probability method, and a new geometric crossover operator that allows for geometrical swaps, rather than random, of genes segments between the chromosomes and control the sizes of the swapped segments. Finally, the effect of boundary conditions on the symmetry of the obtained best solutions is studied and the validity of the "symmetric loading patterns" assumption is tested.
TOPICS: Genetic algorithms, Fuel management, Probability, Pressurized water reactors, Algorithms, Boundary-value problems

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