Nuclear Engineering Institute Research Articles

Radiological protection of Young Scientist and Scientific Researcher trainees in controlled areas of the Argonaut reactor installation: analytical approach

Since the increasing use of ionizing radiations in medical and industrial proceedings, recommendations and actions have been necessaries to reduce the individual radiological risks. In the Nuclear Engineering Institute (IEN), there is the Argonauta research nuclear reactor (RARG) used for producing radionuclides, neutrongraphy studies and training young scientists (YST) and scientific researchers (SRT). During their trainings, these trainees are exposed to ionizing radiations meanly during the neutrongraphyc experiments when they need to adjust neutrongraphy plates in front the J9 channel with the RARG in critical condition. At this moment, they are exposed to neutron and gamma radiations in the RARG hall. This paper evaluated the gamma and neutron expositions that YST and SRT are submitted during the neutrongraphy experiments, by using area monitors to measure the ambient dose equivalent rates in two points near to J9 channel. Considering only this experiment, there was found that the annual individual effective dose could reach up to 2.5 mSv, which is more than 150% of the annual dose to member of the public (MP), but below the annual effective dose to occupationally exposed individual (OEI) as a worker. The question is if YST and SRT should be considered an OEI or a MP. This question is answered through an analytical approach of the regulations and recommendations. The conclusion is that YST and SRT shall be controlled as OEI, but must be submitted to MP’ restriction doses.

Design and beam dynamics simulation of an 8 MeV compact accelerator-driven neutron source

A compact accelerator-driven neutron source is proposed at Sino-French Institute of Nuclear Engineering and Technology, Sun Yat-Sen University, called Sun Yat-Sen University Proton Accelerator Facility (SYSU-PAFA). The proton accelerator is composed of a proton electron cyclotron resonance source, a four-vane radio frequency quadrupole (RFQ), and an alternative phase focusing drift tube linac (APF-DTL). It can accelerate 10 mA proton beam to 8 MeV. Due to the high current, beam matching is particularly important. In order to achieve beam matching between various components, beam transport sections are needed. The beam transport line is divided into three segments. The Low Energy Beam Transport (LEBT) ensures that the beam parameters are matched before entering the RFQ. The Medium Energy Beam Transport (MEBT) segment efficiently transfers the beam between the RFQ and DTL. The High Energy Beam Transport (HEBT) focuses on transporting the beam to the targets. The design goal of beam transport line is as short as possible while ensuring high efficiency of beam transportation. SYSU-PAFA has an overall transmission efficiency of 99%, with optimal transverse matching conditions between beam transport and RFQ or DTL accelerators. The efficient use of solenoids and magnets allows for a compact transmission section, resulting in a total length of 13.6 meters, shorter than most accelerators at the same beam energy. This paper will provide the detailed beam dynamics of the compact accelerator.

Comparison of the response of two scintillator detectors used in an environmental and occupational radiometric survey during the operation of the Argonauta reactor

The Argonauta research reactor at the Institute of Nuclear Engineering - IEN operates, under normal conditions, at powers ranging from 1.7 to 340 W, developing activities of a scientific-technological nature in the areas of neutron radiography, radionuclide production and conduction from experimental practices to graduate courses at CNEN research institutions and other Brazilian institutes of science and technology. During some security operations, the presence of investigators, students and/or operators inside and around the main hall of the Argonauta is required. Therefore, it is important to assess normal occupational exposure conditions when the reactor is in critical condition. This makes it possible to verify in greater detail the equivalent doses of ambient gamma radiation contained by the combustion of the fuel element inside the reactor room and in the surrounding areas, according to the Radiation Protection Plan (RPP) of this installation. In this work, a didactic activity was carried out to familiarize the students of the first period of the Master's Course in Nuclear Engineering, at the Military Institute of Engineering, with the procedures involved in the radiometric survey of the free, controlled and subsequent locations of this nuclear installation. This practice aims to identify the doses to which Occupationally Exposed Workers (OEW) and the general public are admitted, verifying compliance with regulatory requirements regarding the monitoring of environmental and local activities, as well as a comparison between the radiation monitors used.

Physical design and multi-physics analysis of a 200 MHz continuous wave radio frequency quadrupole accelerator for a proton accelerator facility

Based on the platform of a proton accelerator facility, a 200 MHz high intensity continuous wave (CW) radio frequency quadrupole (RFQ) accelerator has been designed at the Sino-French Institute of Nuclear Engineering and Technology of Sun Yat-Sen University, China. Employing the conventional four-vane structure, the RFQ can accelerate 20 mA proton beam from 20 keV to 2.5 MeV. In the beam dynamics design, the simulated transmission efficiency reaches 99.5% and the vane length is less than 4 m. In the electromagnetic design, for the long-term stable operation, Pi-mode Stabilizer Loops (PISLs) are utilized to achieve more than 10 MHz mode frequency separation and to decrease the effects of dipole modes. The 48 tuners are optimized to provide a range of ± 2 MHz frequency tuning. Additionally, the undercuts are optimized to ensure a good field flatness along the longitudinal direction. Based on the coolant channel design, the multi-physics analysis is performed to investigate the deformation and stress resulting from the dissipation of RF power within the cavity, as well as determine the temperature tuning coefficients of the coolants within the vane and wall. The entire design and analysis of PAFA-RFQ has been completed, and the scheme of the design can also be applied to the design of other RFQ cavities.

Decision Support Method in the Assessment of Nuclear Knowledge Management using Fuzzy Logic

The knowledge of workers constitute as valuable resources, as they enable organizations to perform their functions successfully. However, there are conditions that favor the loss of this knowledge in organizations, as for example, the natural aging of workers and consequently the retirement and staff turnover. Then, it becomes important for organization to seek the preservation these knowledge. For a successful implementation of Knowledge Management (KM), it is important to identify the barriers or critical factors that affect the success of the KM process. From the perspective of the nuclear organizations, no systematic framework exists on characterizing a set of critical success factors (CSFs) for implementing KM. Furthermore, the CSFs assessment deals with uncertainty and imprecision of human judgments. In this context, this paper presents a decision support method using CSFs and logic fuzzy to assess the KM in nuclear organizations. Fuzzy theory is essentially used in mapping quantitative models for decision support and representation methods in imprecise and uncertain environments. The decision support method was applied at the Laboratory of Human-Systems Interfaces of the Nuclear Engineering Institute. The results showed that the method is a good assessment tool for the elaboration of nuclear knowledge management strategies.

Active redundant hardware architecture for increased reliability in FPGA-based nuclear reactors critical systems

The hunt for increased reliability in systems for critical applications is a never-ending process and is a point of concern for designers in several different fields, such as nuclear reactors. This concern becomes more prominent when a new device technology is integrated into the options for the development of those systems, such as programmable logic devices like the FPGA . With the constant breakthroughs in this technology, there has been an increase in the capacity and the performance of FPGAs. Nevertheless, new methods to keep fault tolerance at an appropriate level for critical applications in hardware must be considered, particularly due to the transient nature of some radiation-induced faults. This work proposes a resilient and adaptable hardware architecture that increases the reliability and safety of FPGA-based systems, based on a classic active redundancy model, Triple Modular Redundancy with spares. Also, it brings forth a novel hardware architecture that can easily be ported to different FPGA models without compromising performance, reliability, and availability. We discuss and analyze these requirements for the proposed architecture and show that it is more reliable and keeps this reliability for longer periods of time than redundant solutions that use more area. Tests were carried out on a proton beam producing cyclotron at the Nuclear Engineering Institute. The results showed that the proposed architecture has a much lower number of failures than other redundancy solutions. • Increased reliability of circuits implemented in FPGAs to critical systems. • Resilient hardware architecture based on Triple Modular Redundancy with spares. • Hardware architecture reliable to critical system such as nuclear reactors. • Digital hardware model developed as recommended by nuclear reactors safety standards.

A Method for the evaluation of knowledge management systems

The explicit and tacit knowledge of workers constitute as valuable resources, as they enable organizations to perform their functions successfully. Such knowledge is derived from the qualification and experience of workers. On the other side, there are factors that favor the loss of this knowledge in organizations, as for example, the natural aging of workers and consequently the retirement and staff turnover. Because of these reasons, it becomes important for any organization to seek the preservation of its explicit and tacit knowledge. The Knowledge Management seeks to assist in the tasks of identification, organization, preservation and transmission of such knowledge. At the Nuclear Engineering Institute, three systems were developed and implemented to help in the Knowledge Management area: CarpeDIEN, a repository of research data based on the DSpace software; the “Nuclear Energy in Magazines”, a peer-reviewed academic journals management system based on the Open Journal Systems software; and WikiIEN, a wiki based on the MediaWiki software. Each of these systems presents a different proposal and workflow. The objective of this paper is to present the characteristics of each system, a method to compare them based on criteria presented in the literature and to point their weaknesses. The results showed that the method is a proactive tool to provide a basis for choosing the most appropriate system for a given need in an organization.

Short-Term Assessment of Radiological Impact and Potential Risk to Workers and Public from Argonaut Nuclear Reactor Accidental Release

In the early days of nuclear energy in Brazil, a reactor named Argonaut, designed at Argonne National Laboratory, reached criticality at the Institute of Nuclear Engineering (IEN). The presence of a nuclear research facility at the campus of Federal University of Rio de Janeiro is still a cause of concern with regard to the radiological safety of the surrounding community, even though this facility has been securely operating for more than 50 years. In addition, the risk premium paid to IEN workers has also been disputed by the National Office of Account Control. Thus, the aim of this study is to assess the radiological impact and potential risk from Argonaut reactor accidental releases. A recent accident scenario reassessment concluded that severe physical damage of the core after reactor shutdown should be the emergency situation with the greatest potential risk among feasible postulated accidents. The damage caused by failure of a handling crane dropping concrete shielding covers (each weighing 2.5 tons) on the core would lead to breaking of the aluminum coating and the nuclear fuel plates with their release to the reactor hall. This paper evaluates the short-term effective dose rates by inhalation and plume immersion for workers and members of the public, which would be induced by inventory partial release to the atmosphere. The conclusion is that potential risk remains above 1/10 of the limit of annual dose for workers while it stays below the transient levels for members of the public in unrestricted areas.

Analysis of the Production Routine of Iodine123 at IEN and an Adaptation Proposal for a New Radiopharmaceuticals Production Facility

The Brazilian Nuclear Energy Commission (CNEN) is the main manufacturer of radiopharmaceuticals in Brazil. The Nuclear Engineering Institute (IEN), located at Rio de Janeiro, is one of its main research and production centers, supplying public and private hospitals in the country. This radiopharmaceutical production is used in diagnostic and therapy procedures and allows one and a half million nuclear medicine procedures annually. However, IEN facilities were designed in the 60's and today its structure is inadequate in relation to the good manufacturing practices established by sanitary regulator (ANVISA) for the production of radiopharmaceuticals, leading to the need for a new project. In order to adapt and increase production in the country, a new plant will be built and integrated to the existing facilities with a new 30 MeV Cyclotron. Thus, it is proposed an analysis of the productive capacity of this plant in accordance with the principles of good manufacturing practices in order to define daily production and propose an optimized routine. As a result, the IEN / CNEN will in the future have a well-defined and standardized production process, with the flexibility to produce several radioisotopes required in nuclear medicine, obtaining a greater efficiency in its operation.

Proposal for a radiation shielding study aiming the implantation of neutrons beam shutter in the J-9 radiation channel of the Argonauta reactor of the Nuclear Engineering Institute.

Argonauta, the only nuclear research reactor situated in Rio de Janeiro, located at the Institute of Nuclear Engineering (IEN), regularly serves a network of users focused on research and development, and also provides its infrastructure for experimental classes and completion work course. Due to increasing demand for non-destructive thermal neutron assays and production of radioisotopes, there is a search for new procedures and/or devices that optimize users' exposure to neutrons. The implementation of mechanisms that allow access to the irradiation channels without the reactor being turned off and with a shielding configuration that limits the occupational doses at this location is very useful for the operation of the reactor. In order to achieve this, the present work proposes the establishment of a neutron beam shutter of the J-9 irradiation channel of the IEN’s Argonauta reactor. In a first step, experimental measurements were made in the irradiation channel of the reactor using a BF3 detector, which is coupled to a spectrometer. In this phase, the neutron beam was aligned to the spectrometer, and different materials were used as shields, aiming the attenuation of the beam. To validate and/or change the configuration of the barrier that best meets the material irradiation needs, a second planned phase is involving the neutron flux simulation of the reactor and the various shields with different boundary conditions using the particle transport code, Monte Carlo N-Particle Extended (MCNP- X).

Desenvolvimento de um vocabulário controlado para o repositório institucional CarpedIEN

Durante o ano de 2002, o Instituto de Engenharia Nuclear (IEN) apresentou a exposição itinerante “IEN – 40 anos” que, em 2003, chegou ao Museu de Astronomia e Ciências Afins (MAST). Parte da exposição trouxe um conjunto de fotografias que conta a história do IEN e, quando essa coleção retornou ao Instituto, decidiu-se que a guarda dela seria direcionada aos cuidados da biblioteca. Com isso, foi elaborado um projeto institucional que teve como estratégia principal digitalizá-las e disponibilizá-las no repositório institucional do IEN. Para que a recuperação dessas imagens pudesse ser mais consistente, identificou-se a necessidade da construção de um vocabulário controlado para indexá-las, e o projeto se estendeu a todos os itens do repositório. Essa proposta busca apresentar as etapas de construção desse vocabulário, desde a identificação da problemática, passando pelas estratégias de escolha das palavras-chave, a abordagem metodológica e o status atual do projeto, que ainda está em construção.

A Reliability Look at Energy Development

Professor Way Kuo, President of City University of Hong Kong, previously served on the senior management team at Oak Ridge National Laboratory and as Dean of Engineering at the University of Tennessee, and Head of Industrial Engineering at Texas A&M University. A pioneer in reliability, he is renowned for his work in design for reliability of electronics and energy systems. He is a member of the US National Academy of Engineering and Academia Sinica in Taiwan and a Foreign Member of the Chinese Academy of Engineering and the Russian Academy of Engineering. Dr. Chin Pan is Tsing Hua distinguished professor of the Institute of Nuclear Engineering and Science and the Department of Engineering and System Science of the National Tsing Hua University (NTHU), Taiwan. Dr. Pan previously served as Dean of the College of Nuclear Science of NTHU and Convener of the Energy Program of the Ministry of Science and Technology of Taiwan. Dr. Pan's research activities have been in the areas of two-phase flow, boiling heat transfer, nuclear thermal hydraulics, and energy engineering.

Use Dosimetry Virtual Tool for Security Studies Physics and Nuclear

Abstract The objective of this work is to develop a virtual detection tool of radioactive sources, such as to facilitate and assist the training and nuclear security planning. To accomplish this tool was created a virtual model of the Nuclear Engineering Institute (IEN) with virtual characters (avatars), able to move and interact with the environment, radiation detectors (fixed and portable) and sources radioactive. A tool developed enabled the simulation where individuals who were carrying radioactive sources were identified through the detectors installed at strategic points in the virtual environment. Furthermore, it was possible to detect and locate sources for handling portable detectors, operated by the characters within the virtual environment. The results showed a radiation detection system before the continuous profile radioactive sources, making it possible to evaluate the dose rate at any position of the virtual environment. So this work could help in both the security agents training and in the evaluation of the radiological safety of a nuclear site.

Improving the radiochemical purity determination of 123I-labeled metaiodobenzylguanidine

The HPLC method originally applied at the Nuclear Engineering Institute (IEN) for the radiochemical purity determination of 123iodine labeled m-iodobenzylguanidine (123I-mIBG) takes 18.5 min. The final product release also depends on this result, and to facilitate this stage, we aimed to decrease this analysis time. We also intended to use fewer toxic compounds, if feasible. The optimization approach used herein was a combination of factorial and mixture designs to study simultaneously the selected variables. Analysis time, resolution and chromatograms aspect were the measured responses. The qualitative analysis of these responses provided the best chromatographic separation conditions that were 52 mM KH2PO4 in a solution of ethanol and water (1:1), applying a flow rate of 0.50 mL min−1 and C18 column (4.6 × 250 mm, 5 μm). These optimum conditions not only decreased the analysis time in 61 %, but also allowed the reduction of mobile phase toxicity. To assure reliable data, method validation was performed for these conditions. The method has proved its specificity, the detection limit found was 3.70 × 10−4 MBq mL−1 and the quantification limit has corresponded to 1.11 × 10−3 MBq mL−1. Repeatability and intermediate precision has not exceeded 3 and 5 %, respectively, and the accuracy has matched the interval of 95–105 %. This new method has been routinely applied in the radiochemical purity determination of 123I-mIBG at IEN.

Studies of fuel loading pattern optimization for a typical pressurized water reactor (PWR) using improved pivot particle swarm method

Abstract An in-core fuel reload design tool using the improved pivot particle swarm method was developed for the loading pattern optimization problems in a typical PWR, such as Daya Bay Nuclear Power Plant. The discrete, multi-objective improved pivot particle swarm optimization, was integrated with the in-core physics calculation code ‘donjon’ based on finite element method, and assemblies’ group constant calculation code ‘dragon’, composing the optimization code for fuel arrangement. The codes of both ‘donjon’ and ‘dragon’ were programmed by Institute of Nuclear Engineering of Polytechnique Montreal, Canada. This optimization code was aiming to maximize the core effective multiplication factor (Keff), while keeping the local power peaking factor (Ppf) lower than a predetermined value to maintain fuel integrity. At last, the code was applied to the first cycle loading of Daya Bay Nuclear Power Plant. The result showed that, compared with the reference loading pattern design, the core effective multiplication factor increased by 9.6%, while the power peaking factor decreased by 0.6%, meeting the safety requirement.

An efficient Neuro-Fuzzy approach to nuclear power plant transient identification

► We investigate a Neuro-Fuzzy modeling tool use for able transient identification. ► The prelusive transient type identification is done by an artificial neural network. ► After, the fuzzy-logic system analyzes the results emitting reliability degree of it. ► The research support was made in a PWR simulator at the Brazilian Nuclear Engineering Institute. ► The results show the potential to help operators’ decisions in a nuclear power plant. Transient identification in nuclear power plants (NPP) is often a computational very hard task and may involve a great amount of human cognition. The early identification of unexpected departures from steady state behavior is an essential step for the operation, control and accident management in NPPs. The bases for the transient identification relay on the evidence that different system faults and anomalies lead to different pattern evolution in the involved process variables. During an abnormal event, the operator must monitor a great amount of information from the instruments that represents a specific type of event. Recently, several works have been developed for transient identification. These works frequently present a non reliable response, using the “don´t know” as the system output. In this work, we investigate the possibility of using a Neuro-Fuzzy modeling tool for efficient transient identification, aiming to helping the operator crew to take decisions relative to the procedure to be followed in situations of accidents/transients at NPPs. The proposed system uses artificial neural networks (ANN) as first level transient diagnostic. After the ANN has done the preliminary transient type identification, a fuzzy-logic system analyzes the results emitting reliability degree of it. A validation of this identification system was made at the three loops Pressurized Water Reactor (PWR) simulator of the Human-System Interface Laboratory (LABIHS) of the Nuclear Engineering Institute (IEN/CNEN/Brazil). The obtained results show the potential of this new transient identification system to be used in an operational NPP in order to assist the operators to take decisions during transients/accidents.

Influence of radiation on endotoxin test using the PTS TM for 18-FDG radiopharmaceutical

F-18 FDG (2-[18-F] fluoro-2-deoxy-D-glucose) is the most frequently used radiopharmaceutical for PET and PET CT imaging exams. The FDA recently approved the use of the PTS TM (Portable Test System) as an alternative to the standard test proposed by the United States Pharmacopeia using the LAL (Limulus Amebocyte Lysates), that takes longer to perform (about 1h) than the PTS TM (15 min). Recent studies have demonstrated that radiation could interfere with the PTS TM test. In order to study the effects of radiation on the PTS TM test and/or equipment, 27 batches of F-18 FDG produced in the Nuclear Engineering Institute were analyzed. The results showed that no direct correlation with radiation was found in any of the cases.

Comparison of Limulus Amebocyte Lysates with the United States Pharmacopeial Pyrogen Test and the Portable Test System for Radiopharmaceuticals

Radiopharmaceuticals play an important role in modern nuclear medicine, and 18F-FDG (2-[18F] fluoro-2-deoxy-D-glucose) is the most frequently used in imaging examinations today. The pyrogen test represents an important parameter for the quality of radiopharmaceuticals since most of them, 18F-FDG included, are injectable solutions. However, the standard test proposed by the U.S. Pharmacopeia using limulus amebocyte lysates takes too long (about 1 h). An alternative test is the Portable Test System, which takes no more than 15 min. In order to compare both tests, 10 batches of 18F-FDG produced in the Nuclear Engineering Institute, Rio de Janeiro, Brazil, were analyzed. The results showed that in 40% of the samples analyzed, different results were found. From the total analyzed, 20% showed totally diverging results, and 80% demonstrated partial similarity. These differences have a great impact in the choice of the test used for radiopharmaceutical injectable solutions and the importance of validation tests before the implementation in the daily routine.

TLD environmental monitoring at the Institute of Nuclear Engineering in Brazil

Since 2003 the Institute of Nuclear Engineering in Rio de Janeiro city, Brazil, operates a new cyclotron, RDS-111, to produce (18)F-Fluorodeoxyglucose to be used in nuclear medicine. Additionally, the IEN radioactive waste repository has been enlarged during the past last years, receiving a considerable amount of radioactive materials. Therefore, it became necessary to evaluate a possible increase of the environmental gamma exposure rates at the institute site due to the operation of the new accelerator and the enlargement of the institute waste repository as well. LiF:Mg,Cu,P, TLD-100H, and TL detectors were employed for environmental kerma rate evaluation and the results were compared with previous results obtained before the RDS-111 operation initialisation and the enlargement of IEN waste repository. No significant contribution for the enhancement of environmental gamma kerma rates was detected.

Particle swarm optimisation applied to nuclear engineering problems

Evolutionary computation (EC) techniques, and more specifically genetic algorithms (GA) and their variations, have been efficiently applied to many complex problems found in the nuclear engineering field. Such methods have been shown to be robust and efficient, but highly time consuming. Other population-based methods have been proposed as alternatives to these traditional EC techniques. The Particle Swarm Optimisation (PSO) technique has been shown to be faster and many times more efficient than GA. Motivated by that, investigations concerning applications of PSO to nuclear engineering problems have started in the Brazilian Nuclear Engineering Institute (IEN/CNEN) and Federal University of Rio de Janeiro (UFRJ). This paper describes applications of PSO to four classical nuclear engineering problems: (i) nuclear fuel reload, (ii) core design optimisation, (iii) surveillance tests planning and (iv) accident classification. Computational experiments demonstrate that PSO can be efficiently applied to the problems studied. Moreover, the results described are comparable with, or even better than, some good results (obtained by GA) found in the literature.