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Sample records for core research reactor

  1. Research on plasma core reactors

    Energy Technology Data Exchange (ETDEWEB)

    Jarvis, G.A.; Barton, D.M.; Helmick, H.H.; Bernard, W.; White, R.H.

    1977-01-01

    Experiments and theoretical studies are being conducted for NASA on critical assemblies with 1-m-diam by 1-m-long low-density cores surrounded by a thick beryllium reflector. These assemblies make extensive use of existing nuclear propulsion reactor components, facilities, and instrumentation. Due to excessive porosity in the reflector, the initial critical mass was 19 kg U(93.2). Addition of a 17-cm-thick by 89-cm-diam beryllium flux trap in the cavity reduced the critical mass to 7 kg when all the uranium was in the zone just outside the flux trap. A mockup aluminum UF/sub 6/ container was placed inside the flux trap and fueled with uranium-graphite elements. Fission distributions and reactivity worths of fuel and structural materials were measured. Finally, an 85,000-cm/sup 3/ aluminum canister in the central region was fueled with UF/sub 6/ gas and fission density distributions determined. These results will be used to guide the design of a prototype plasma core reactor which will test energy removal by optical radiation.

  2. Reactor pulse repeatability studies at the annular core research reactor

    Energy Technology Data Exchange (ETDEWEB)

    DePriest, K.R. [Applied Nuclear Technologies, Sandia National Laboratories, Mail Stop 1146, Post Office Box 5800, Albuquerque, NM 87185-1146 (United States); Trinh, T.Q. [Nuclear Facility Operations, Sandia National Laboratories, Mail Stop 0614, Post Office Box 5800, Albuquerque, NM 87185-1146 (United States); Luker, S. M. [Applied Nuclear Technologies, Sandia National Laboratories, Mail Stop 1146, Post Office Box 5800, Albuquerque, NM 87185-1146 (United States)

    2011-07-01

    The Annular Core Research Reactor (ACRR) at Sandia National Laboratories is a water-moderated pool-type reactor designed for testing many types of objects in the pulse and steady-state mode of operations. Personnel at Sandia began working to improve the repeatability of pulse operations for experimenters in the facility. The ACRR has a unique UO{sub 2}-BeO fuel that makes the task of producing repeatable pulses difficult with the current operating procedure. The ACRR produces a significant quantity of photoneutrons through the {sup 9}Be({gamma}, n){sup 8}Be reaction in the fuel elements. The photoneutrons are the result of the gammas produced during fission and in fission product decay, so their production is very much dependent on the reactor power history and changes throughout the day/week of experiments in the facility. Because the photoneutrons interfere with the delayed-critical measurements required for accurate pulse reactivity prediction, a new operating procedure was created. The photoneutron effects at delayed critical are minimized when using the modified procedure. In addition, the pulse element removal time is standardized for all pulse operations with the modified procedure, and this produces less variation in reactivity removal times. (authors)

  3. Neutron spectrometric methods for core inventory verification in research reactors

    CERN Document Server

    Ellinger, A; Hansen, W; Knorr, J; Schneider, R

    2002-01-01

    In consequence of the Non-Proliferation Treaty safeguards, inspections are periodically made in nuclear facilities by the IAEA and the EURATOM Safeguards Directorate. The inspection methods are permanently improved. Therefore, the Core Inventory Verification method is being developed as an indirect method for the verification of the core inventory and to check the declared operation of research reactors.

  4. Control Rod Reactivity Curves for the Annular Core Research Reactor

    Science.gov (United States)

    Depriest, K. Russell; Kajder, Karen C.; Frye, Jason N.; Denman, Matthew R.

    2009-08-01

    Experiments were conducted at the Annular Core Research Reactor (ACRR) to increase the fidelity of the control rod integral reactivity worth curve. This experiment series was designed to refine the integral reactivity curve used for pulse yield prediction and eliminate the need for operator compensation in the pulse setup. The experiment series consisted of delayed critical and positive period measurements with various ACRR cavity configurations. An improved integral reactivity worth curve for the ACRR control rods has been constructed using the positive period measurements, the delayed critical measurements, and radiation transport modeling of the reactor. A series of prompt period measurements is used to validate that the new control rod curve more accurately predicts the energy yield of the pulse operations. The new reactivity worth curve is compared with the current curve that was developed using traditional approaches.

  5. Researches on a reactor core in heavy ion inertial fusion

    CERN Document Server

    Kondo, S; Iinuma, T; Kubo, K; Kato, H; Kawata, S; Ogoyski, A I

    2016-01-01

    In this paper a study on a fusion reactor core is presented in heavy ion inertial fusion (HIF), including the heavy ion beam (HIB) transport in a fusion reactor, a HIB interaction with a background gas, reactor cavity gas dynamics, the reactor gas backflow to the beam lines, and a HIB fusion reactor design. The HIB has remarkable preferable features to release the fusion energy in inertial fusion: in particle accelerators HIBs are generated with a high driver efficiency of ~30-40%, and the HIB ions deposit their energy inside of materials. Therefore, a requirement for the fusion target energy gain is relatively low, that would be ~50 to operate a HIF fusion reactor with a standard energy output of 1GW of electricity. In a fusion reactor the HIB charge neutralization is needed for a ballistic HIB transport. Multiple mechanical shutters would be installed at each HIB port at the reactor wall to stop the blast waves and the chamber gas backflow, so that the accelerator final elements would be protected from the ...

  6. Characterization of Novel Calorimeters in the Annular Core Research Reactor *

    Directory of Open Access Journals (Sweden)

    Hehr Brian D.

    2016-01-01

    Full Text Available A series of pulsed irradiation experiments have been performed in the central cavity of Sandia National Laboratories' Annular Core Research Reactor (ACRR to characterize the responses of a set of elemental calorimeter materials including Si, Zr, Sn, Ta, W, and Bi. Of particular interest was the perturbing effect of the calorimeter itself on the ambient radiation field – a potential concern in dosimetry applications. By placing the calorimeter package into a neutron-thermalizing lead/polyethylene (LP bucket and irradiating both with and without a cadmium wrapper, it was demonstrated that prompt capture gammas generated inside the calorimeters can be a significant contributor to the measured dose in the active disc region. An MCNP model of the experimental setup was shown to replicate measured dose responses to within 10%. The internal (n,γ contribution was found to constitute as much as 50% of the response inside the LP bucket and up to 20% inside the nominal (unmodified cavity environment, with Ta and W exhibiting the largest enhancement due to their sizable (n,γ cross sections. Capture reactions in non-disc components of the calorimeter were estimated to be responsible for up to a few percent of the measured response.

  7. McCARD for Neutronics Design and Analysis of Research Reactor Cores

    Science.gov (United States)

    Shim, Hyung Jin; Park, Ho Jin; Kwon, Soonwoo; Seo, Geon Ho; Hyo Kim, Chang

    2014-06-01

    McCARD is a Monte Carlo (MC) neutron-photon transport simulation code developed exclusively for the neutronics design and analysis of nuclear reactor cores. McCARD is equipped with the hierarchical modeling and scripting functions, the CAD-based geometry processing module, the adjoint-weighted kinetics parameter and source multiplication factor estimation modules as well as the burnup analysis capability for the neutronics design and analysis of both research and power reactor cores. This paper highlights applicability of McCARD for the research reactor core neutronics analysis, as demonstrated for Kyoto University Critical Assembly, HANARO, and YALINA.

  8. Characterization of the fast neutron irradiation facility of the Portuguese Research Reactor after core conversion.

    Science.gov (United States)

    Marques, J G; Sousa, M; Santos, J P; Fernandes, A C

    2011-08-01

    The fast neutron irradiation facility of the Portuguese Research Reactor was characterized after the reduction in uranium enrichment and rearrangement of the core configuration. In this work we report on the determination of the hardness parameter and the 1MeV equivalent neutron flux along the facility, in the new irradiation conditions, following ASTM E722 standard.

  9. Characterization of the Annular Core Research Reactor (ACRR Neutron Radiography System Imaging Plane

    Directory of Open Access Journals (Sweden)

    Kaiser Krista

    2016-01-01

    Full Text Available The Annular Core Research Reactor (ACRR at Sandia National Laboratories (SNL is an epithermal pool-type research reactor licensed up to a thermal power of 2.4 MW. The ACRR facility has a neutron radiography facility that is used for imaging a wide range of items including reactor fuel and neutron generators. The ACRR neutron radiography system has four apertures (65:1, 125:1, 250:1, and 500:1 available to experimenters. The neutron flux and spectrum as well as the gamma dose rate were characterized at the imaging plane for the ACRR's neutron radiography system for the 65:1, 125:1 and 250:1 apertures.

  10. Characterization of the Annular Core Research Reactor (ACRR) Neutron Radiography System Imaging Plane

    Science.gov (United States)

    Kaiser, Krista; Chantel Nowlen, K.; DePriest, K. Russell

    2016-02-01

    The Annular Core Research Reactor (ACRR) at Sandia National Laboratories (SNL) is an epithermal pool-type research reactor licensed up to a thermal power of 2.4 MW. The ACRR facility has a neutron radiography facility that is used for imaging a wide range of items including reactor fuel and neutron generators. The ACRR neutron radiography system has four apertures (65:1, 125:1, 250:1, and 500:1) available to experimenters. The neutron flux and spectrum as well as the gamma dose rate were characterized at the imaging plane for the ACRR's neutron radiography system for the 65:1, 125:1 and 250:1 apertures.

  11. A study on improving the performance of a research reactor's equilibrium core

    Directory of Open Access Journals (Sweden)

    Muhammad Atta

    2013-01-01

    Full Text Available Utilizing low enriched uranium silicide fuel (U3Si2-Al of existing uranium density (3.285 g/cm3, different core configurations have been studied in search of an equilibrium core with an improved performance for the Pakistan Research Reactor-1. Furthermore, we have extended our analysis to the performance of higher density silicide fuels with a uranium density of 4.0 and 4.8 U g/cm3. The criterion used in selecting the best performing core was that of “unit flux time cycle length per 235U mass per cycle”. In order to analyze core performance by improving neutron moderation, utilizing higher-density fuel, the effect of the coolant channel width was also studied by reducing the number of plates in the standard/control fuel element. Calculations employing computer codes WIMSD/4 and CITATION were performed. A ten energy group structure for fission neutrons was used for the generation of microscopic cross-sections through WIMSD/4. To search the equilibrium core, two-dimensional core modelling was performed in CITATION. Performance indicators have shown that the higher-density uranium silicide-fuelled core (U density 4.8 g/cm3 without any changes in standard/control fuel elements, comprising of 15 standard and 4 control fuel elements, is the best performing of all analyzed cores.

  12. Technical basis in support of the conversion of the University of Missouri Research Reactor (MURR) core from highly-enriched to low-enriched uranium - core neutron physics

    Energy Technology Data Exchange (ETDEWEB)

    Stillman, J. [Argonne National Lab. (ANL), Argonne, IL (United States); Feldman, E. [Univ. of Missouri, Columbia, MO (United States). Columbia Research Reactor; Foyto, L [Univ. of Missouri, Columbia, MO (United States). Columbia Research Reactor; Kutikkad, K [Univ. of Missouri, Columbia, MO (United States). Columbia Research Reactor; McKibben, J C [Univ. of Missouri, Columbia, MO (United States). Columbia Research Reactor; Peters, N. [Univ. of Missouri, Columbia, MO (United States). Columbia Research Reactor; Stevens, J. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2012-09-01

    This report contains the results of reactor design and performance for conversion of the University of Missouri Research Reactor (MURR) from the use of highly-enriched uranium (HEU) fuel to the use of low-enriched uranium (LEU) fuel. The analyses were performed by staff members of the Global Threat Reduction Initiative (GTRI) Reactor Conversion Program at the Argonne National Laboratory (ANL) and the MURR Facility. The core conversion to LEU is being performed with financial support of the U. S. government.

  13. Safety analysis for operating the Annular Core Research Reactor with Cintichem-type targets installed in the central region of the core

    Energy Technology Data Exchange (ETDEWEB)

    PARMA JR.,EDWARD J.

    2000-01-01

    Production of the molybdenum-99 isotope at the Annular Core Research Reactor requires highly enriched, uranium oxide loaded targets to be irradiated for several days in the high neutron-flux region of the core. This report presents the safety analysis for the irradiation of up to seven Cintichem-type targets in the central region of the core and compares the results to the Annular Core Research Reactor Safety Analysis Report. A 19 target grid configuration is presented that allows one to seven targets to be irradiated, with the remainder of the grid locations filled with aluminum ''void'' targets. Analyses of reactor, neutronic, thermal hydraulics, and heat transfer calculations are presented. Steady-state operation and accident scenarios are analyzed with the conclusion that the reactor can be operated safely with seven targets in the grid, and no additional risk to the public.

  14. Georgia Institute of Technology research on the Gas Core Actinide Transmutation Reactor (GCATR)

    Science.gov (United States)

    Clement, J. D.; Rust, J. H.; Schneider, A.; Hohl, F.

    1976-01-01

    The program reviewed is a study of the feasibility, design, and optimization of the GCATR. The program is designed to take advantage of initial results and to continue work carried out on the Gas Core Breeder Reactor. The program complements NASA's program of developing UF6 fueled cavity reactors for power, nuclear pumped lasers, and other advanced technology applications. The program comprises: (1) General Studies--Parametric survey calculations performed to examine the effects of reactor spectrum and flux level on the actinide transmutation for GCATR conditions. The sensitivity of the results to neutron cross sections are to be assessed. Specifically, the parametric calculations of the actinide transmutation are to include the mass, isotope composition, fission and capture rates, reactivity effects, and neutron activity of recycled actinides. (2) GCATR Design Studies--This task is a major thrust of the proposed research program. Several subtasks are considered: optimization criteria studies of the blanket and fuel reprocessing, the actinide insertion and recirculation system, and the system integration. A brief review of the background of the GCATR and ongoing research is presented.

  15. Optimization of the core configuration design using a hybrid artificial intelligence algorithm for research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hedayat, Afshin, E-mail: ahedayat@aut.ac.i [Department of Nuclear Engineering and Physics, Amirkabir University of Technology (Tehran Polytechnic), 424 Hafez Avenue, P.O. Box 15875-4413, Tehran (Iran, Islamic Republic of); Reactor Research and Development School, Nuclear Science and Technology Research Institute (NSTRI), End of North Karegar Street, P.O. Box 14395-836, Tehran (Iran, Islamic Republic of); Davilu, Hadi [Department of Nuclear Engineering and Physics, Amirkabir University of Technology (Tehran Polytechnic), 424 Hafez Avenue, P.O. Box 15875-4413, Tehran (Iran, Islamic Republic of); Barfrosh, Ahmad Abdollahzadeh [Department of Computer Engineering, Amirkabir University of Technology (Tehran Polytechnic), 424 Hafez Avenue, P.O. Box 15875-4413, Tehran (Iran, Islamic Republic of); Sepanloo, Kamran [Reactor Research and Development School, Nuclear Science and Technology Research Institute (NSTRI), End of North Karegar Street, P.O. Box 14395-836, Tehran (Iran, Islamic Republic of)

    2009-12-15

    To successfully carry out material irradiation experiments and radioisotope productions, a high thermal neutron flux at irradiation box over a desired life time of a core configuration is needed. On the other hand, reactor safety and operational constraints must be preserved during core configuration selection. Two main objectives and two safety and operational constraints are suggested to optimize reactor core configuration design. Suggested parameters and conditions are considered as two separate fitness functions composed of two main objectives and two penalty functions. This is a constrained and combinatorial type of a multi-objective optimization problem. In this paper, a fast and effective hybrid artificial intelligence algorithm is introduced and developed to reach a Pareto optimal set. The hybrid algorithm is composed of a fast and elitist multi-objective genetic algorithm (GA) and a fast fitness function evaluating system based on the cascade feed forward artificial neural networks (ANNs). A specific GA representation of core configuration and also special GA operators are introduced and used to overcome the combinatorial constraints of this optimization problem. A software package (Core Pattern Calculator 1) is developed to prepare and reform required data for ANNs training and also to revise the optimization results. Some practical test parameters and conditions are suggested to adjust main parameters of the hybrid algorithm. Results show that introduced ANNs can be trained and estimate selected core parameters of a research reactor very quickly. It improves effectively optimization process. Final optimization results show that a uniform and dense diversity of Pareto fronts are gained over a wide range of fitness function values. To take a more careful selection of Pareto optimal solutions, a revision system is introduced and used. The revision of gained Pareto optimal set is performed by using developed software package. Also some secondary operational

  16. Neutronic and thermal-hydraulic analysis of new irradiation channels inside the Moroccan TRIGA Mark II research reactor core.

    Science.gov (United States)

    Chham, E; El Bardouni, T; Benaalilou, K; Boukhal, H; El Bakkari, B; Boulaich, Y; El Younoussi, C; Nacir, B

    2016-10-01

    This study was conducted to improve the capacity of radioisotope production in the Moroccan TRIGA Mark II research reactor, which is considered as one of the most important applications of research reactors. The aim of this study is to enhance the utilization of TRIGA core in the field of neutron activation and ensure an economic use of the fuel. The main idea was to create an additional irradiation channel (IC) inside the core. For this purpose, three new core configurations are proposed, which differ according to the IC position in the core. Thermal neutron flux distribution and other neutronic safety parameters such as power peaking factors, excess reactivity, and control rods worth reactivity were calculated using the Monte Carlo N-Particle Transport (MCNP) code and neutron cross-section library based on ENDF/B-VII evaluation. The calculated thermal flux in the central thimble (CT) and in the added IC for the reconfigured core is compared with the thermal flux in the CT of the existing core, which is taken as a reference. The results show that all the obtained fluxes in CTs are very close to the reference value, while a remarkable difference is observed between the fluxes in the new ICs and reference. This difference depends on the position of IC in the reactor core. To demonstrate that the Moroccan TRIGA reactor could safely operate at 2MW, with new configurations based on new ICs, different safety-related thermal-hydraulic parameters were investigated. The PARET model was used in this study to verify whether the safety margins are met despite the new modifications of the core. The results show that it is possible to introduce new ICs safely in the reactor core, because the obtained values of the parameters are largely far from compromising the safety of the reactor.

  17. INVAP's Research Reactor Designs

    Directory of Open Access Journals (Sweden)

    Eduardo Villarino

    2011-01-01

    Full Text Available INVAP, an Argentine company founded more than three decades ago, is today recognized as one of the leaders within the research reactor industry. INVAP has participated in several projects covering a wide range of facilities, designed in accordance with the requirements of our different clients. For complying with these requirements, INVAP developed special skills and capabilities to deal with different fuel assemblies, different core cooling systems, and different reactor layouts. This paper summarizes the general features and utilization of several INVAP research reactor designs, from subcritical and critical assemblies to high-power reactors.

  18. Numerical Research on Hybrid Fuel Locking Device for Upward Flow Core-Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Huh, Hyung; Cho, Yeong-Garp; Yoo, Yeon-Sik; Ryu, Jeong-Soo [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    The assembly must be held firmly against these forces, but cannot be permanently attached to the support stand because periodic refueling of the reactor requires removal or relocation of each assembly. There are so many kinds of fuel locking device, but they are operated manually. As a part of a new project, we have investigated a hybrid fuel locking device (HFLD) for research reactor which is operated automatically. Prior method of holding down the fuel assembly includes a hybrid zero electromagnet consisting of an electromagnet and a permanent magnet. The role of an electromagnet is converged to zero power for overcoming the lifting power of a permanent magnet by controlling the coil current. At this time, a HFLD is an unlocking state. On the contrary, it is locking state that only a permanent magnet works when the power of an electromagnet is off. The results of a FEM in this work lead to the following conclusions: (1) It is possible that an electromagnet is converged to zero power for overcoming the lifting power of a permanent magnet by remote controlling the coil current. (2) At this time, it is able to detect remotely using proximity sensor whether a HFLD is latched or not.

  19. Review of Transient Fuel Test Results at Sandia National Laboratories and the Potential for Future Fast Reactor Fuel Transient Testing in the Annular Core Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Wright, Steven A.; Pickard, Paul S.; Parma, Edward J.; Vernon, Milton E.; Kelly, John; Tikare, Veena [Sandia National Laboratories, Org 6872 MS-1146, PO Box 5800 Albuquerque, New Mexico 87185 (United States)

    2009-06-15

    Reactor driven transient tests of fast reactor fuels may be required to support the development and certification of new fuels for Fast Reactors. The results of the transient fuel tests will likely be needed to support licensing and to provide validation data to support the safety case for a variety of proposed fast fuel types and reactors. In general reactor driven transient tests are used to identify basic phenomenology during reactor transients and to determine the fuel performance limits and margins to failure during design basis accidents such as loss of flow, loss of heat sink, and reactivity insertion accidents. This paper provides a summary description of the previous Sandia Fuel Disruption and Transient Axial Relocation tests that were performed in the Annular Core Research Reactor (ACRR) for the U.S. Nuclear Regulatory Commission almost 25 years ago. These tests consisted of a number of capsule tests and flowing gas tests that used fission heating to disrupt fresh and irradiated MOX fuel. The behavior of the fuel disruption, the generation of aerosols and the melting and relocation of fuel and cladding was recorded on high speed cinematography. This paper will present videos of the fuel disruption that was observed in these tests which reveal stark differences in fuel behavior between fresh and irradiated fuel. Even though these tests were performed over 25 years ago, their results are still relevant to today's reactor designs. These types of transient tests are again being considered by the Advanced Fuel Cycle Initiative to support the Global Nuclear Energy Partnership because of the need to perform tests on metal fuels and transuranic fuels. Because the Annular Core Research Reactor is the only transient test facility available within the US, a brief summary of Sandia's continued capability to perform these tests in the ACRR will also be provided. (authors)

  20. Core calculations for the upgrading of the IEA-R1 research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Santos, Adimir dos; Perrotta, Jose A.; Bastos, Jose Luis F.; Yamaguchi, Mitsuo; Umbehaun, Pedro E. [Instituto de Pesquisas Energeticas e Nucleares (IPEN), Sao Paulo, SP (Brazil)]. E-mail: asantos@net.ipen.br; perrotta@net.ipen.br; mitsuo@net.ipen.br

    1998-07-01

    The IEA-R1 Research Reactor is a multipurpose reactor. It has been used for basic and applied research in the nuclear area, training and radioisotopes production since 1957. In 1995, the Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP) took the decision to modernize and upgrade the power from 2 to 5 MW and increase the operational cycle. This work presents the design requirements and the calculations effectuated to reach this goal. (author)

  1. Kinetic parameters study based on burn-up for improving the performance of research reactor equilibrium core

    Directory of Open Access Journals (Sweden)

    Muhammad Atta

    2014-01-01

    Full Text Available In this study kinetic parameters, effective delayed neutron fraction and prompt neutron generation time have been investigated at different burn-up stages for research reactor's equilibrium core utilizing low enriched uranium high density fuel (U3Si2-Al fuel with 4.8 g/cm3 of uranium. Results have been compared with reference operating core of Pakistan research Reactor-1. It was observed that by increasing fuel burn-up, effective delayed neutron fraction is decreased while prompt neutron generation time is increased. However, over all ratio beff/L is decreased with increasing burn-up. Prompt neutron generation time L in the understudy core is lower than reference operating core of reactor at all burn-up steps due to hard spectrum. It is observed that beff is larger in the understudy core than reference operating core of due to smaller size. Calculations were performed with the help of computer codes WIMSD/4 and CITATION.

  2. Neutron Environment Characterization of the Central Cavity in the Annular Core Research Reactor *

    Directory of Open Access Journals (Sweden)

    Parma Edward J.

    2016-01-01

    Full Text Available Characterization of the neutron environment in the central cavity of the Sandia National Laboratories' Annular Core Research Reactor (ACRR is important in order to provide experimenters with the most accurate spectral information and maintain a high degree of fidelity in performing reactor experiments. Characterization includes both modeling and experimental efforts. Building accurate neutronic models of the ACRR and the central cavity “bucket” environments that can be used by experimenters is important in planning and designing experiments, as well as assessing the experimental results and quantifying uncertainties. Neutron fluence characterizations of two bucket environments, LB44 and PLG, are presented. These two environments are used frequently and represent two extremes in the neutron spectrum. The LB44 bucket is designed to remove the thermal component of the neutron spectrum and significantly attenuate the gamma-ray fluence. The PLG bucket is designed to enhance the thermal component of the neutron spectrum and attenuate the gamma-ray fluence. The neutron characterization for each bucket was performed by irradiating 20 different activation foil types, some of which were cadmium covered, resulting in 37 different reactions at the peak axial flux location in each bucket. The dosimetry results were used in the LSL-M2 spectrum adjustment code with a 640-energy group MCNP-generated trial spectrum, self-shielding correction factors, the SNLRML or IRDFF dosimetry cross-section library, trial spectrum uncertainty, and trial covariance matrix, to generate a least-squares adjusted neutron spectrum, spectrum uncertainty, and covariance matrix. Both environment character-izations are well documented and the environments are available for use by experimenters.

  3. Neutron Environment Characterization of the Central Cavity in the Annular Core Research Reactor

    Science.gov (United States)

    Parma, Edward J.; Naranjo, Gerald E.; Lippert, Lance L.; Vehar, David W.

    2016-02-01

    Characterization of the neutron environment in the central cavity of the Sandia National Laboratories' Annular Core Research Reactor (ACRR) is important in order to provide experimenters with the most accurate spectral information and maintain a high degree of fidelity in performing reactor experiments. Characterization includes both modeling and experimental efforts. Building accurate neutronic models of the ACRR and the central cavity "bucket" environments that can be used by experimenters is important in planning and designing experiments, as well as assessing the experimental results and quantifying uncertainties. Neutron fluence characterizations of two bucket environments, LB44 and PLG, are presented. These two environments are used frequently and represent two extremes in the neutron spectrum. The LB44 bucket is designed to remove the thermal component of the neutron spectrum and significantly attenuate the gamma-ray fluence. The PLG bucket is designed to enhance the thermal component of the neutron spectrum and attenuate the gamma-ray fluence. The neutron characterization for each bucket was performed by irradiating 20 different activation foil types, some of which were cadmium covered, resulting in 37 different reactions at the peak axial flux location in each bucket. The dosimetry results were used in the LSL-M2 spectrum adjustment code with a 640-energy group MCNP-generated trial spectrum, self-shielding correction factors, the SNLRML or IRDFF dosimetry cross-section library, trial spectrum uncertainty, and trial covariance matrix, to generate a least-squares adjusted neutron spectrum, spectrum uncertainty, and covariance matrix. Both environment character-izations are well documented and the environments are available for use by experimenters. Work supported by the United States Department of Energy at Sandia National Laboratories. Sandia National Laboratories is a multi-program laboratory managed and operated by Sandia Corporation, a wholly owned

  4. Preliminary study of degradation from neutron effects of core-structural materials of Thai Research Reactor TRR-1/M1

    Science.gov (United States)

    Ampornrat, P.; Boonsuwan, P.; Sangkaew, S.; Angwongtrakool, T.

    2017-06-01

    Thai research reactor went first critical in 1962. The reactor was converted in 1977 from an MTR-type with high-enriched uranium fuel to a TRIGA-MARK III type using low-enriched uranium fuel, called TRR-1/M1. Since the TRR-1/M1 has been operated for almost 40 years, degradation of reactor structural materials is expected. In this preliminary study, the potential degradation from neutron effects of core-structural materials, e.g., fuel clad (SS304) and core components (Al6061) were studied. Assessment included calculation of neutron energy, flux and fluence in the reactor core to evaluate displacement rate (dpa) and irradiation effects on the material properties. Results showed maximum displacement rates on SS304 was 5.24×10-8 per cm3·sec and on Al6061 was 1.14×10-8 per cm3·sec. The corresponding maximum displacement levels were ∼17 dpa for SS304, and ∼4 dpa for Al6061. At these levels of displacement, it is possible for the materials to result in tensile strength increasing and ductility reduction. Further inspection on the core-structural materials needs to be conducted to validate the assessment results from this study.

  5. INVAP's Research Reactor Designs

    OpenAIRE

    Eduardo Villarino; Alicia Doval

    2011-01-01

    INVAP, an Argentine company founded more than three decades ago, is today recognized as one of the leaders within the research reactor industry. INVAP has participated in several projects covering a wide range of facilities, designed in accordance with the requirements of our different clients. For complying with these requirements, INVAP developed special skills and capabilities to deal with different fuel assemblies, different core cooling systems, and different reactor layouts. This paper ...

  6. Fabrication of uranium dioxide fuel pellets in support of a SLOWPOKE-2 research reactor HEU to LEU core conversion

    Energy Technology Data Exchange (ETDEWEB)

    Bergeron, A. [Aomic Energy of Canada Limited, Chalk River, Ontario (Canada)

    2014-07-01

    The International Centre for Environmental and Nuclear Sciences (ICENS) at the University of the West Indies in Jamaica operates a SLOWPOKE-2 research reactor that is currently fuelled with highly-enriched uranium (HEU). As part of the Global Threat Reduction Initiative, Atomic Energy of Canada Ltd. has been subcontracted to fabricate low-enriched uranium (LEU) fuel for the ICENS SLOWPOKE-2. The low enriched uranium core consists of a fuel cage containing uranium dioxide fuelled elements. This paper describes the fabrication of the low-enriched uranium dioxide fuel pellets for the SLOWPOKE-2 core conversion. (author)

  7. Analysis of partial and total flow blockage of a single fuel assembly of an MTR research reactor core

    Energy Technology Data Exchange (ETDEWEB)

    Adorni, Martina [Dipartimento di Ingegneria Meccanica, Nucleare e della Produzione, Facolta di Ingegneria, Universita di Pisa, Via Diotisalvi, 2-56100 Pisa (Italy)]. E-mail: martina_adorni@tin.it; Bousbia-Salah, Anis [Dipartimento di Ingegneria Meccanica, Nucleare e della Produzione, Facolta di Ingegneria, Universita di Pisa, Via Diotisalvi, 2-56100 Pisa (Italy); Hamidouche, Tewfik [Commissariat a l' Energie Atomique, Centre de Recherche Nucleaire d' Alger-Algeria, 02 Boulevard Frantz fanon, BP 399 Alger-gare (Algeria); Maro, Beniamino Di [Dipartimento di Ingegneria Meccanica, Nucleare e della Produzione, Facolta di Ingegneria, Universita di Pisa, Via Diotisalvi, 2-56100 Pisa (Italy); Pierro, Franco [Dipartimento di Ingegneria Meccanica, Nucleare e della Produzione, Facolta di Ingegneria, Universita di Pisa, Via Diotisalvi, 2-56100 Pisa (Italy); D' Auria, Francesco [Dipartimento di Ingegneria Meccanica, Nucleare e della Produzione, Facolta di Ingegneria, Universita di Pisa, Via Diotisalvi, 2-56100 Pisa (Italy)

    2005-10-15

    The main aim of the following study is to perform a safety analysis of the IAEA 10 MW MTR Pool type Research Reactor [IAEA-TECDOC-233, 1980. IAEA Research Reactor Core Conversion from the use of high-enriched uranium to the use of low enriched uranium fuels Guidebook] under flow blockage of a single Fuel Assembly (FA) conditions. Such event was rarely investigated in the open literature notwithstanding the fat that it constitutes a severe accident that may lead to local dryout and eventually to loss of the FA integrity. The transients herein considered are related to partial and total obstruction of the cooling channel of a single Fuel Assembly of the reactor core. This study constitutes the first step of a larger work, which consists in performing a 3D simulation using the Best Estimate coupled code technique. However, as a first approach the instantaneous reactor power is derived through the point kinetic approach of the used thermal-hydraulic system code.

  8. Neutron flux parameters for k{sub 0}-NAA method at the Malaysian nuclear agency research reactor after core reconfiguration

    Energy Technology Data Exchange (ETDEWEB)

    Yavar, A.R. [School of Applied Physics, Faculty of Science and Technology, University Kebangsaan Malaysia (UKM), Bangi, Selangor 43600 (Malaysia); Sarmani, S. [School of Chemical Sciences and Food Technology, Faculty of Science and Technology, University Kebangsaan Malaysia (UKM), Bangi, Selangor 43600 (Malaysia); Wood, A.K. [Analytical Chemistry Application Group, Industrial Technology Division, Malaysian Nuclear Agency (MNA), Bangi, Kajang, Selangor 43000 (Malaysia); Fadzil, S.M. [School of Applied Physics, Faculty of Science and Technology, University Kebangsaan Malaysia (UKM), Bangi, Selangor 43600 (Malaysia); Masood, Z. [Analytical Chemistry Application Group, Industrial Technology Division, Malaysian Nuclear Agency (MNA), Bangi, Kajang, Selangor 43000 (Malaysia); Khoo, K.S., E-mail: khoo@ukm.m [School of Applied Physics, Faculty of Science and Technology, University Kebangsaan Malaysia (UKM), Bangi, Selangor 43600 (Malaysia)

    2011-02-15

    The Malaysian Nuclear Agency (MNA) research reactor, commissioned in 1982, is a TRIGA Mark II swimming pool type reactor. When the core configuration changed in June 2009, it became essential to re-determine such neutron flux parameters as thermal to epithermal neutron flux ratio (f), epithermal neutron flux shape factor ({alpha}), thermal neutron flux ({phi}{sub th}) and epithermal neutron flux ({phi}{sub epi}) in the irradiation positions of MNA research reactor in order to guarantee accuracy in the application of k{sub 0}-neutron activation analysis (k{sub 0}-NAA).The f and {alpha} were determined using the bare bi-isotopic monitor and bare triple monitor methods, respectively; Au and Zr monitors were utilized in present study. The results for four irradiation positions are presented and discussed in the present work. The calculated values of f and {alpha} ranged from 33.49 to 47.33 and -0.07 to -0.14, respectively. The {phi}{sub th} and the {phi}{sub epi} were measured as 2.03 x 10{sup 12} (cm{sup -2} s{sup -1}) and 6.05 x 10{sup 10} (cm{sup -2} s{sup -1}) respectively. These results were compared to those of previous studies at this reactor as well as to those of reactors in other countries. The results indicate a good conformity with other findings.

  9. Study on Reduced-Moderation Water Reactor (RMWR) core design. Joint research report (FY1998-1999)

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2000-09-01

    The Reduce-Moderation Water Reactor (RMWR) is a next generation water-cooled reactor aiming at effective utilization of uranium resource, high burn-up and long operation cycle, and plutonium multi-recycle. Japan Atomic Energy Research Institute (JAERI) started a joint research program for conceptual design of RMWR core in collaboration with the Japan Atomic Power Company (JAPC) since 1998. The research area includes the RMWR core conceptual designs, development of analysis methods for rector physics and thermal-hydraulics to design the RMWR cores with higher accuracy and preparation of MOX critical experiment to confirm the feasibility from the reactor physics point of view. The present report describes the results of joint research program 'RMWR core design Phase 1' performed by JAERI and JAPC in FY 1998 and 1999. The results obtained from the joint research program are as follows: Conceptual design study on the RMWR core has been performed. A core concept with a conversion ratio more than about 1 is basically feasible to multiple recycling of plutonium. Investigating core characteristics at the equilibrium, some promising core concepts to satisfy above aims have been established. As for BWR-type concepts with negative void reactivity coefficients, three types of design have been obtained as follows; (1) one feasible to attain high conversion ratio about 1.1, (2) one feasible to attain operation cycle of about 2 years and burn-up of about 60 GWd/t with conversion ratio more than 1 or (3) one in simple design based on the ABWR assembly and without blanket attaining conversion ratio more than 1. And as for PWR-type concepts with negative void reactivity coefficients, two types of design have been obtained as follows; (1) one feasible to attain high conversion ratio about 1.05 by using heavy water as a coolant and (2) one feasible to attain conversion ratio about l by using light water. In the study of nuclear calculation method, a reactor analysis code

  10. Monte Carlo Calculation of Core Reactivity and Fluxes for the Development of the BNCT Neutron Source at the Kyiv Research Reactor

    Science.gov (United States)

    Gritzay, Olena; Kalchenko, Oleksandr; Klimova, Nataliya; Razbudey, Volodymyr; Sanzhur, Andriy; Binney, Stephen

    2005-05-01

    The presented results show our consecutive steps in developing a neutron source with parameters required by Boron Neutron Capture Therapy (BNCT) at the Kyiv Research Reactor (KRR). The main goal of this work was to analyze the influence of installation of different types of uranium converters close to the reactor core on neutron beam characteristics and on level of reactor safety. The general Monte Carlo radiation transport code MCNP, version 4B, has been used for these calculations.

  11. Neutronic analysis for core conversion (HEU–LEU) of the low power research reactor using the MCNP4C code

    OpenAIRE

    Aldawahra Saadou; Khattab Kassem; Saba Gorge

    2015-01-01

    Comparative studies for conversion of the fuel from HEU to LEU in the miniature neutron source reactor (MNSR) have been performed using the MCNP4C code. The HEU fuel (UAl4-Al, 90% enriched with Al clad) and LEU (UO2 12.6% enriched with zircaloy-4 alloy clad) cores have been analyzed in this study. The existing HEU core of MNSR was analyzed to validate the neutronic model of reactor, while the LEU core was studied to prove the possibility of fuel conversion of the existing HEU core. The propos...

  12. Neutronic analysis for core conversion (HEU–LEU of the low power research reactor using the MCNP4C code

    Directory of Open Access Journals (Sweden)

    Aldawahra Saadou

    2015-06-01

    Full Text Available Comparative studies for conversion of the fuel from HEU to LEU in the miniature neutron source reactor (MNSR have been performed using the MCNP4C code. The HEU fuel (UAl4-Al, 90% enriched with Al clad and LEU (UO2 12.6% enriched with zircaloy-4 alloy clad cores have been analyzed in this study. The existing HEU core of MNSR was analyzed to validate the neutronic model of reactor, while the LEU core was studied to prove the possibility of fuel conversion of the existing HEU core. The proposed LEU core contained the same number of fuel pins as the HEU core. All other structure materials and dimensions of HEU and LEU cores were the same except the increase in the radius of control rod material from 0.195 to 0.205 cm and keeping the outer diameter of the control rod unchanged in the LEU core. The effective multiplication factor (keff, excess reactivity (ρex, control rod worth (CRW, shutdown margin (SDM, safety reactivity factor (SRF, delayed neutron fraction (βeff and the neutron fluxes in the irradiation tubes for the existing and the potential LEU fuel were investigated. The results showed that the safety parameters and the neutron fluxes in the irradiation tubes of the LEU fuels were in good agreements with the HEU results. Therefore, the LEU fuel was validated to be a suitable choice for fuel conversion of the MNSR in the future.

  13. Multi purpose research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Raina, V.K. [Research Reactor Design and Projects Division, Bhabha Atomic Research Centre, Mumbai 400085 (India)]. E-mail: vkrain@magnum.barc.ernet.in; Sasidharan, K. [Research Reactor Design and Projects Division, Bhabha Atomic Research Centre, Mumbai 400085 (India); Sengupta, Samiran [Research Reactor Design and Projects Division, Bhabha Atomic Research Centre, Mumbai 400085 (India); Singh, Tej [Research Reactor Services Division, Bhabha Atomic Research Centre, Mumbai 400085 (India)

    2006-04-15

    At present Dhruva and Cirus reactors provide the majority of research reactor based facilities to cater to the various needs of a vast pool of researchers in the field of material sciences, physics, chemistry, bio sciences, research and development work for nuclear power plants and production of radio isotopes. With a view to further consolidate and expand the scope of research and development in nuclear and allied sciences, a new 20 MWt multi purpose research reactor is being designed. This paper describes some of the design features and safety aspects of this reactor.

  14. Neutronic simulation of a research reactor core of (232Th, 235U)O2 fuel using MCNPX2.6 code

    Indian Academy of Sciences (India)

    Seyed Amir Hossein Feghhi; Marzieh Rezazadeh; Yachine Kadi; Claudio Tenreiro; Morteza Aref; Zohreh Gholamzadeh

    2013-01-01

    The small reactor design for the remote and less developed areas of the user countries should have simple features in view of the lack of infra-structure and resources. Many researchers consider long core life with no on-site refuelling activity as a primary feature for the small reactor design. Long core life can be achieved by enhancing internal conversion rate of fertile to fissile materials. For that purpose, thorium cycle can be adopted because a high fissile production rate of 233U converted from 232Th can be expected in the thermal energy region. A simple nuclear reactor core arranged 19 assemblies in hexagonal structure, using thorium-based fuel and heavy water as coolant and moderator was simulated using MCNPX2.6 code, aiming an optimized critical assembly. Optimized reflector thickness and gap between assemblies were determined to achieve minimum neutron leakage and void reactivity. The result was a more compact core, where assemblies were designed having 19-fuel pins in 1.25 pitch-to-diameter ratio. Optimum reflector thickness of 15 cm resulted in minimal neutron leakage in view of economic limitations. A 0.5 cm gap between assembles achieved more safety and 2.2% enrichment requirements. The present feasibility study suggests a thermal core of acceptable neutronic parameters to achieve a simple and safe core.

  15. Effect of the New WIMSD Libraries on the Neutronic Parameters of the First and Equilibrium Cores of the Indonesian Multipurpose Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Sembiring, T.M.; Suparlina, L. [Center for Reactor Technology and Nuclear Safety, National Nuclear Energy Agency (BATAN), Kawasan PUSPIPTEK Gd. No. 80, Serpong, Tangerang Selatan 15310 (Indonesia); Hong, Liem Peng [Research Laboratory for Nuclear Reactors, Tokyo Institute of Technology O-okayama, Meguro, Tokyo 152-8550 (Japan)

    2011-07-01

    Accurate neutronic parameters are very important in the design and safety analysis of a research reactor. The calculation method and neutron cross-section data play an important role to obtain those neutronic parameters. In this work, we reported our new results on the effects of the recent neutron cross-section data of the WIMSD libraries, JEFF-3.1 and ENDF/B-VII.0, on the criticality parameters of the first and equilibrium cores of the Indonesian multipurpose reactor, RSG-GAS reactor. WIMSD lattice code together with WIMSD libraries, coupled with the in- core fuel management code BATAN-FUEL, has been used routinely in the in-core management analysis since 1996. The objective of this work is to assess the accuracy of the recent WIMSD libraries before being adopted for routine in-core fuel management. As the first step, a series of core calculations were carried out for determining the effective multiplication factor, burn-up swing reactivity and Xenon equilibrium reactivity for the first and equilibrium cores. The results show that WIMSD library based on the evaluated nuclear data ENDF/B-VII.0 gives better agreement with the experimental results for criticality evaluation (underestimation of less than 0.4%). For burn-up swing evaluation, we found relatively large differences among the libraries (max. 5%). However, there is no significant difference among three WIMSD libraries in determining the Xenon equilibrium reactivity. (author)

  16. An Innovative Passive Residual Heat Removal System of an Open-Pool Type Research Reactor with Pump Flywheel and Gravity Core Cooling Tank

    Directory of Open Access Journals (Sweden)

    Kwon-Yeong Lee

    2015-01-01

    Full Text Available In an open-pool type research reactor, the primary cooling system can be designed to have a downward flow inside the core during normal operation because of the plate type fuel geometry. There is a flow inversion inside the core from the downward flow by the inertia force of the primary coolant to the upward flow by the natural circulation when the pump is turned off. To delay the flow inversion time, an innovative passive system with pump flywheel and GCCT is developed to remove the residual heat. Before the primary cooling pump starts up, the water level of the GCCT is the same as that of the reactor pool. During the primary cooling pump operation, the water in the GCCT is moved into the reactor pool because of the pump suction head. After the pump stops, the potential head generates a downward flow inside the core by moving the water from the reactor pool to the GCCT and removes the residual heat. When the water levels of the two pools are the same again, the core flow has an inversion of the flow direction, and natural circulation is developed through the flap valves.

  17. Reactor Materials Research

    Energy Technology Data Exchange (ETDEWEB)

    Van Walle, E

    2001-04-01

    The activities of the Reactor Materials Research Department of the Belgian Nuclear Research Centre SCK-CEN in 2000 are summarised. The programmes within the department are focussed on studies concerning (1) fusion, in particular mechanical testing; (2) Irradiation Assisted Stress Corrosion Cracking (IASCC); (3) nuclear fuel; and (4) Reactor Pressure Vessel Steel (RPVS)

  18. Development of inherent core technologies for advanced reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Keung Koo; Noh, J.M.; Hwang, D.H. [and others

    1999-03-01

    Recently, the developed countries made their effort on developing the advanced reactor which will result in significantly enhanced safety and economy. However, they will protect the advanced reactor and its design technology with patent and proprietary right. Therefore, it is very important to develop our own key core concepts and inherent core design technologies which can form a foundation of indigenous technologies for development of the domestic advanced reactor in order to keep the superiority in the nuclear plant building market among the developing countries. In order to provide the basic technology for the core design of advanced reactor, this project is for developing the inherent core design concepts with enhanced safety and economy, and associated methodologies and technologies for core analyses. The feasibility study of constructing domestic critical facilities are performed by surveying the status and utilization of foreign facilities and by investigating the demand for domestic facilities. The research results developed in this project, such as core analysis methodologies for hexagonal core, conceptual core design based on hexagonal fuel assemblies and soluble boron core design and control strategies, will provide a technical foundation in developing core design of domestic advanced reactor. Furthermore, they will strengthen the competitiveness of Korean nuclear technology. We also expect that some of the design concepts developed in this project to improve the reactor safety and economy can be applicable to the design of advanced reactor. This will significantly reduce the public anxiety on the nuclear power plant, and will contribute to the economy of construction and operation for the future domestic reactors. Even though the critical facility will not be constructed right now, the investigation of the status and utilization of foreign critical facility will contribute to the future critical facility construction. (author). 150 refs., 34 tabs., 103

  19. Wire core reactor for nuclear thermal propulsion

    Science.gov (United States)

    Harty, Richard B.; Brengle, Robert G.

    1993-01-01

    Studies have been performed of a compact high-performance nuclear rocket reactor that incorporates a tungsten alloy wire fuel element. This reactor, termed the wire core reactor, can deliver a specific impulse of 1,000 s using an expander cycle and a nozzle expansion ratio of 500 to 1. The core is constructed of layers of 0.8-mm-dia fueled tungsten wires wound over alternate layers of spacer wires, which forms a rugged annular lattice. Hydrogen flow in the core is annular, flowing from inside to outside. In addition to the concepts compact size and good heat transfer, the core has excellent power-flow matching features and can resist vibration and thermal stresses during star-up and shutdown.

  20. Advanced reactor physics methods for heterogeneous reactor cores

    Science.gov (United States)

    Thompson, Steven A.

    To maintain the economic viability of nuclear power the industry has begun to emphasize maximizing the efficiency and output of existing nuclear power plants by using longer fuel cycles, stretch power uprates, shorter outage lengths, mixed-oxide (MOX) fuel and more aggressive operating strategies. In order to accommodate these changes, while still satisfying the peaking factor and power envelope requirements necessary to maintain safe operation, more complexity in commercial core designs have been implemented, such as an increase in the number of sub-batches and an increase in the use of both discrete and integral burnable poisons. A consequence of the increased complexity of core designs, as well as the use of MOX fuel, is an increase in the neutronic heterogeneity of the core. Such heterogeneous cores introduce challenges for the current methods that are used for reactor analysis. New methods must be developed to address these deficiencies while still maintaining the computational efficiency of existing reactor analysis methods. In this thesis, advanced core design methodologies are developed to be able to adequately analyze the highly heterogeneous core designs which are currently in use in commercial power reactors. These methodological improvements are being pursued with the goal of not sacrificing the computational efficiency which core designers require. More specifically, the PSU nodal code NEM is being updated to include an SP3 solution option, an advanced transverse leakage option, and a semi-analytical NEM solution option.

  1. Usage of burnable poison on research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Villarino, Eduardo Anibal [INVAP S.E., San Carlos de Bariloche (Argentina)

    2002-07-01

    The fuel assemblies with burnable poison are widely used on power reactors, but there are not commonly used on research reactors. This paper shows a neutronic analysis of the advantages and disadvantages of the burnable poison usage on research reactors. This paper analyses both burnable poison design used on research reactors: Boron on the lateral wall and Cadmium wires. Both designs include a parametric study on the design parameters like the amount and geometry of the burnable poison. This paper presents the design flexibility using burnable poisons, it does not find an optimal or final design, which it will strongly depend on the core characteristics and fuel management strategy. (author)

  2. Reactor Materials Research

    Energy Technology Data Exchange (ETDEWEB)

    Van Walle, E

    2002-04-01

    The activities of SCK-CEN's Reactor Materials Research Department for 2001 are summarised. The objectives of the department are: (1) to evaluate the integrity and behaviour of structural materials used in nuclear power industry; (2) to conduct research to unravel and understand the parameters that determine the material behaviour under or after irradiation; (3) to contribute to the interpretation, the modelling of the material behaviour and to develop and assess strategies for optimum life management of nuclear power plant components. The programmes within the department are focussed on studies concerning (1) Irradiation Assisted Stress Corrosion Cracking (IASCC); (2) nuclear fuel; and (3) Reactor Pressure Vessel Steel.

  3. Gas core reactors for coal gasification

    Science.gov (United States)

    Weinstein, H.

    1976-01-01

    The concept of using a gas core reactor to produce hydrogen directly from coal and water is presented. It is shown that the chemical equilibrium of the process is strongly in favor of the production of H2 and CO in the reactor cavity, indicating a 98% conversion of water and coal at only 1500 K. At lower temperatures in the moderator-reflector cooling channels the equilibrium strongly favors the conversion of CO and additional H2O to CO2 and H2. Furthermore, it is shown the H2 obtained per pound of carbon has 23% greater heating value than the carbon so that some nuclear energy is also fixed. Finally, a gas core reactor plant floating in the ocean is conceptualized which produces H2, fresh water and sea salts from coal.

  4. Gas core reactors for coal gasification

    Science.gov (United States)

    Weinstein, H.

    1976-01-01

    The concept of using a gas core reactor to produce hydrogen directly from coal and water is presented. It is shown that the chemical equilibrium of the process is strongly in favor of the production of H2 and CO in the reactor cavity, indicating a 98% conversion of water and coal at only 1500 K. At lower temperatures in the moderator-reflector cooling channels the equilibrium strongly favors the conversion of CO and additional H2O to CO2 and H2. Furthermore, it is shown the H2 obtained per pound of carbon has 23% greater heating value than the carbon so that some nuclear energy is also fixed. Finally, a gas core reactor plant floating in the ocean is conceptualized which produces H2, fresh water and sea salts from coal.

  5. Site Investigation for Detection of KIJANG Reactor Core Center

    Energy Technology Data Exchange (ETDEWEB)

    Kwon, Tae-Hyun; Kim, Jun Yeon; Kim, Jeeyoung [KAERI, Daejeon (Korea, Republic of)

    2015-05-15

    It was planned for the end of March 2017 and extended to April 2018 according to the government budget adjustment. The KJRR project is intended for filling the self-sufficiency of RI demand including Mo-99, increasing the NTD capacity and developing technologies related to the research reactor. In project, site investigation is the first activity that defines seismologic and related geologic aspects of the site. Site investigation was carried out from Oct. 2012 to Jan. 2014 and this study is intended to describe detail procedures in locating the reactor core center. The location of the reactor core center was determined by collectively reviewing not only geological information but also information from architects engineering. EL 50m was selected as ground level by levering construction cost. Four recommended locations (R-1a - R-1d) are displayed for the reactor core center. R-1a was found optimal in consideration of medium rock contour, portion of medium rock covering reactor buildings, construction cost, physical protection and electrical resistivity. It is noted that engineering properties of the medium rock is TCR/RQD 100/53, elastic modulus 7,710 - 8,720MPa, permeability coefficient 2.92E-06cm/s, and S-wave velocity 1,380m/s, sound for foundations of reactor buildings.

  6. Nuclear reactor core modelling in multifunctional simulators

    Energy Technology Data Exchange (ETDEWEB)

    Puska, E.K. [VTT Energy, Nuclear Energy, Espoo (Finland)

    1999-06-01

    The thesis concentrates on the development of nuclear reactor core models for the APROS multifunctional simulation environment and the use of the core models in various kinds of applications. The work was started in 1986 as a part of the development of the entire APROS simulation system. The aim was to create core models that would serve in a reliable manner in an interactive, modular and multifunctional simulator/plant analyser environment. One-dimensional and three-dimensional core neutronics models have been developed. Both models have two energy groups and six delayed neutron groups. The three-dimensional finite difference type core model is able to describe both BWR- and PWR-type cores with quadratic fuel assemblies and VVER-type cores with hexagonal fuel assemblies. The one- and three-dimensional core neutronics models can be connected with the homogeneous, the five-equation or the six-equation thermal hydraulic models of APROS. The key feature of APROS is that the same physical models can be used in various applications. The nuclear reactor core models of APROS have been built in such a manner that the same models can be used in simulator and plant analyser applications, as well as in safety analysis. In the APROS environment the user can select the number of flow channels in the three-dimensional reactor core and either the homogeneous, the five- or the six-equation thermal hydraulic model for these channels. The thermal hydraulic model and the number of flow channels have a decisive effect on the calculation time of the three-dimensional core model and thus, at present, these particular selections make the major difference between a safety analysis core model and a training simulator core model. The emphasis on this thesis is on the three-dimensional core model and its capability to analyse symmetric and asymmetric events in the core. The factors affecting the calculation times of various three-dimensional BWR, PWR and WWER-type APROS core models have been

  7. NEUTRON RADIOGRAPHY (NRAD) REACTOR 64-ELEMENT CORE UPGRADE

    Energy Technology Data Exchange (ETDEWEB)

    John D. Bess

    2014-03-01

    The neutron radiography (NRAD) reactor is a 250 kW TRIGA (registered) (Training, Research, Isotopes, General Atomics) Mark II , tank-type research reactor currently located in the basement, below the main hot cell, of the Hot Fuel Examination Facility (HFEF) at the Idaho National Laboratory (INL). It is equipped with two beam tubes with separate radiography stations for the performance of neutron radiography irradiation on small test components. The interim critical configuration developed during the core upgrade, which contains only 62 fuel elements, has been evaluated as an acceptable benchmark experiment. The final 64-fuel-element operational core configuration of the NRAD LEU TRIGA reactor has also been evaluated as an acceptable benchmark experiment. Calculated eigenvalues differ significantly (approximately +/-1%) from the benchmark eigenvalue and have demonstrated sensitivity to the thermal scattering treatment of hydrogen in the U-Er-Zr-H fuel.

  8. Thermal Hydraulic Tests for Reactor Core Safety

    Energy Technology Data Exchange (ETDEWEB)

    Moon, S. K.; Baek, W. P.; Chun, S. Y. (and others)

    2007-06-15

    The main objectives of the present project are to resolve the current issues of reactor core thermal hydraulics, to develop an advanced measurement and analytical techniques, and to perform reactor core safety verification tests. 6x6 reflood experiments, various heat transfer experiments using Freon, and experiments on the spacer grids effects on the post-dryout are carried out using spacer grids developed in Korea in order to resolve the current issues of the reactor core thermal hydraulics. In order to develop a reflood heat transfer model, the detailed reflood phenomena are visualized and measured using round tube and 2x2 rod bundle. A detailed turbulent mixing phenomenon for subchannels is measured using advanced measurement techniques such as LDV and PIV. MARS and MATRA codes developed in Korea are assessed, verified and improved using the obtained experimental data. Finally, a systematic quality assurance program and experimental data generation system has been constructed in order to increase the reliability of the experimental data.

  9. Benchmark Evaluation of the NRAD Reactor LEU Core Startup Measurements

    Energy Technology Data Exchange (ETDEWEB)

    J. D. Bess; T. L. Maddock; M. A. Marshall

    2011-09-01

    The Neutron Radiography (NRAD) reactor is a 250-kW TRIGA-(Training, Research, Isotope Production, General Atomics)-conversion-type reactor at the Idaho National Laboratory; it is primarily used for neutron radiography analysis of irradiated and unirradiated fuels and materials. The NRAD reactor was converted from HEU to LEU fuel with 60 fuel elements and brought critical on March 31, 2010. This configuration of the NRAD reactor has been evaluated as an acceptable benchmark experiment and is available in the 2011 editions of the International Handbook of Evaluated Criticality Safety Benchmark Experiments (ICSBEP Handbook) and the International Handbook of Evaluated Reactor Physics Benchmark Experiments (IRPhEP Handbook). Significant effort went into precisely characterizing all aspects of the reactor core dimensions and material properties; detailed analyses of reactor parameters minimized experimental uncertainties. The largest contributors to the total benchmark uncertainty were the 234U, 236U, Er, and Hf content in the fuel; the manganese content in the stainless steel cladding; and the unknown level of water saturation in the graphite reflector blocks. A simplified benchmark model of the NRAD reactor was prepared with a keff of 1.0012 {+-} 0.0029 (1s). Monte Carlo calculations with MCNP5 and KENO-VI and various neutron cross section libraries were performed and compared with the benchmark eigenvalue for the 60-fuel-element core configuration; all calculated eigenvalues are between 0.3 and 0.8% greater than the benchmark value. Benchmark evaluations of the NRAD reactor are beneficial in understanding biases and uncertainties affecting criticality safety analyses of storage, handling, or transportation applications with LEU-Er-Zr-H fuel.

  10. The reprocessing of reactor core materials

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Jing, E-mail: wang-jing@nuaa.edu.cn [State Key Laboratory of Mechanics and Control of Mechanial Structures, Nanjing University of Aeronautics and Astronautics, Nanjing 210016 (China); Liu, Bing; Shao, Youlin; Lu, Zhenming; Liu, Malin [Institute of Nuclear and New Energy Technology, Tsinghua University, Beijing 100084 (China)

    2014-05-01

    Generation IV high temperature gas-cooled reactors (HTGR) are preferentially fueled by spherical fuel elements, which are composed of a fuel zone of triso-coated uranium oxide (UO{sub 2}) particles and a matrix graphite layer. Unqualified coated particles and spherical fuel elements unavoidablely occur during the processing of coating UO{sub 2} kernels and embedding the coated particles into the graphite matrix. So it is necessary to reprocess the UO{sub 2} in the unqualified coated particles and spherical fuel elements to maximize the use of the reactor core materials. In this work, we present several methods to: (1) separate the coated particles from the graphite matrix and, (2) expose and recover the UO{sub 2} kernels from the coated particles. The comparison of different methods shows that the thermal oxidation of graphite by a fixed bed burner and the jet grinding of the unqualified coated particles are prosing in practice for the separation of coated particles from the graphite matrix and recovering the uranium dioxide kernels, respectively. Some other methods, such as etching the SiC layer with the active fluorine species in plasma generated by the dielectric barrier discharge (DBD) under the atmosphere also show their great potential values in the reprocessing of reactor core materials, especially for the activated and contaminated fuels.

  11. Nuclear research reactors in Brazil

    Energy Technology Data Exchange (ETDEWEB)

    Cota, Anna Paula Leite; Mesquita, Amir Zacarias, E-mail: aplc@cdtn.b, E-mail: amir@cdtn.b [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2011-07-01

    The rising concerns about global warming and energy security have spurred a revival of interest in nuclear energy, giving birth to a 'nuclear power renaissance' in several countries in the world. Particularly in Brazil, in the recent years, the nuclear power renaissance can be seen in the actions that comprise its nuclear program, summarily the increase of the investments in nuclear research institutes and the government target to design and build the Brazilian Multipurpose research Reactor (BMR). In the last 50 years, Brazilian research reactors have been used for training, for producing radioisotopes to meet demands in industry and nuclear medicine, for miscellaneous irradiation services and for academic research. Moreover, the research reactors are used as laboratories to develop technologies in power reactors, which are evaluated today at around 450 worldwide. In this application, those reactors become more viable in relation to power reactors by the lowest cost, by the operation at low temperatures and, furthermore, by lower demand for nuclear fuel. In Brazil, four research reactors were installed: the IEA-R1 and the MB-01 reactors, both at the Instituto de Pesquisas Energeticas Nucleares (IPEN, Sao Paulo); the Argonauta, at the Instituto de Engenharia Nuclear (IEN, Rio de Janeiro) and the IPR-R1 TRIGA reactor, at the Centro de Desenvolvimento da Tecnologia Nuclear (CDTN, Belo Horizonte). The present paper intends to enumerate the characteristics of these reactors, their utilization and current academic research. Therefore, through this paper, we intend to collaborate on the BMR project. (author)

  12. Nuclear research reactors in Brazil

    Energy Technology Data Exchange (ETDEWEB)

    Cota, Anna Paula Leite; Mesquita, Amir Zacarias, E-mail: aplc@cdtn.b, E-mail: amir@cdtn.b [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2011-07-01

    The rising concerns about global warming and energy security have spurred a revival of interest in nuclear energy, giving birth to a 'nuclear power renaissance' in several countries in the world. Particularly in Brazil, in the recent years, the nuclear power renaissance can be seen in the actions that comprise its nuclear program, summarily the increase of the investments in nuclear research institutes and the government target to design and build the Brazilian Multipurpose research Reactor (BMR). In the last 50 years, Brazilian research reactors have been used for training, for producing radioisotopes to meet demands in industry and nuclear medicine, for miscellaneous irradiation services and for academic research. Moreover, the research reactors are used as laboratories to develop technologies in power reactors, which are evaluated today at around 450 worldwide. In this application, those reactors become more viable in relation to power reactors by the lowest cost, by the operation at low temperatures and, furthermore, by lower demand for nuclear fuel. In Brazil, four research reactors were installed: the IEA-R1 and the MB-01 reactors, both at the Instituto de Pesquisas Energeticas Nucleares (IPEN, Sao Paulo); the Argonauta, at the Instituto de Engenharia Nuclear (IEN, Rio de Janeiro) and the IPR-R1 TRIGA reactor, at the Centro de Desenvolvimento da Tecnologia Nuclear (CDTN, Belo Horizonte). The present paper intends to enumerate the characteristics of these reactors, their utilization and current academic research. Therefore, through this paper, we intend to collaborate on the BMR project. (author)

  13. Sodium fast reactor evaluation: Core materials

    Science.gov (United States)

    Cheon, Jin Sik; Lee, Chan Bock; Lee, Byoung Oon; Raison, J. P.; Mizuno, T.; Delage, F.; Carmack, J.

    2009-07-01

    In the framework of the Generation IV Sodium Fast Reactor (SFR) Program the Advanced Fuel Project has conducted an evaluation of the available fuel systems supporting future sodium cooled fast reactors. In this paper the status of available and developmental materials for SFR core cladding and duct applications is reviewed. To satisfy the Generation IV SFR fuel requirements, an advanced cladding needs to be developed. The candidate cladding materials are austenitic steels, ferritic/martensitic (F/M) steels, and oxide dispersion strengthened (ODS) steels. A large amount of irradiation testing is required, and the compatibility of cladding with TRU-loaded fuel at high temperatures and high burnup must be investigated. The more promising F/M steels (compared to HT9) might be able to meet the dose requirements of over 200 dpa for ducts in the GEN-IV SFR systems.

  14. Multilevel transport solution of LWR reactor cores

    Energy Technology Data Exchange (ETDEWEB)

    Jose Ignacio Marquez Damian; Cassiano R.E. de Oliveira; HyeonKae Park

    2008-09-01

    This work presents a multilevel approach for the solution of the transport equation in typical LWR assemblies and core configurations. It is based on the second-order, even-parity formulation of the transport equation, which is solved within the framework provided by the finite element-spherical harmonics code EVENT. The performance of the new solver has been compared with that of the standard conjugate gradient solver for diffusion and transport problems on structured and unstruc-tured grids. Numerical results demonstrate the potential of the multilevel scheme for realistic reactor calculations.

  15. 77 FR 30435 - In-core Thermocouples at Different Elevations and Radial Positions in Reactor Core

    Science.gov (United States)

    2012-05-23

    ... COMMISSION 10 CFR Part 50 In-core Thermocouples at Different Elevations and Radial Positions in Reactor Core... ``require all holders of operating licenses for nuclear power plants (``NPP'') to operate NPPs with in-core thermocouples at different elevations and radial positions throughout the reactor core.'' DATES: Submit...

  16. Analysis of High Temperature Reactor Control Rod Worth for the Initial and Full Core

    Science.gov (United States)

    Oktajianto, Hammam; Setiawati, Evi; Anam, Khoirul; Sugito, Heri

    2017-01-01

    Control rod is one important component in a nuclear reactor. In nuclear reactor operations the control rod functions to shut down the reactor. This research analyses ten control rods worth of HTR (High Temperature Reactor) at initial and full core. The HTR in this research adopts HTR-10 China and HTR- of pebble bed. Core calculations are performed by using MCNPX code after modelling the entire parts of core in condition of ten control rods fully withdrawn, all control rods in with 20 cm ranges of depth and the use of one control rod. Pebble bed and moderator balls are distributed in the core zone using a Body Centred Cubic (BCC) lattice by ratio of 57:43. The research results are obtained that the use of one control rod will decrease the reactor criticality of 2.04±0.12 %Δk/k at initial core and 1.57±0.10 %Δk/k at full core. The deeper control rods are in, the lesser criticality of reactor is with reactivity of ten control rods of 16.41±0.11 %Δk/k at initial core and 15.43±0.11 %Δk/k at full core. The results show that the use of ten control rods at full core will keep achieving subcritical condition even though the reactivity is smaller than reactivity at initial core.

  17. Strengthening IAEA Safeguards for Research Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Reid, Bruce D. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Anzelon, George A. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Budlong-Sylvester, Kory [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2016-09-01

    During their December 10-11, 2013, workshop in Grenoble France, which focused on the history and future of safeguarding research reactors, the United States, France and the United Kingdom (UK) agreed to conduct a joint study exploring ways to strengthen the IAEA’s safeguards approach for declared research reactors. This decision was prompted by concerns about: 1) historical cases of non-compliance involving misuse (including the use of non-nuclear materials for production of neutron generators for weapons) and diversion that were discovered, in many cases, long after the violations took place and as part of broader pattern of undeclared activities in half a dozen countries; 2) the fact that, under the Safeguards Criteria, the IAEA inspects some reactors (e.g., those with power levels under 25 MWt) less than once per year; 3) the long-standing precedent of States using heavy water research reactors (HWRR) to produce plutonium for weapons programs; 4) the use of HEU fuel in some research reactors; and 5) various technical characteristics common to some types of research reactors that could provide an opportunity for potential proliferators to misuse the facility or divert material with low probability of detection by the IAEA. In some research reactors, for example, such characteristics include rapid on-line refueling, and a core design with room for such a large number of assemblies or targets that it is difficult to detect diversion or undeclared irradiation. In addition, infrastructure associated with research reactors, such as hot cells, where plutonium could be separated, could pose a safeguards challenge because, in some cases, they are not declared (because they are not located in the facility or because nuclear materials are not foreseen to be processed inside) and may not be accessible to inspectors in States without an Additional Protocol in force.

  18. Development of an automated core model for nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Mosteller, R.D.

    1998-12-31

    This is the final report of a three-year, Laboratory Directed Research and Development (LDRD) project at the Los Alamos National Laboratory (LANL). The objective of this project was to develop an automated package of computer codes that can model the steady-state behavior of nuclear-reactor cores of various designs. As an added benefit, data produced for steady-state analysis also can be used as input to the TRAC transient-analysis code for subsequent safety analysis of the reactor at any point in its operating lifetime. The basic capability to perform steady-state reactor-core analysis already existed in the combination of the HELIOS lattice-physics code and the NESTLE advanced nodal code. In this project, the automated package was completed by (1) obtaining cross-section libraries for HELIOS, (2) validating HELIOS by comparing its predictions to results from critical experiments and from the MCNP Monte Carlo code, (3) validating NESTLE by comparing its predictions to results from numerical benchmarks and to measured data from operating reactors, and (4) developing a linkage code to transform HELIOS output into NESTLE input.

  19. Review on Application of Control Algorithms to Power Regulations of Reactor Cores

    OpenAIRE

    2016-01-01

    This research is to solve the stability analysis issue of nonlinear pressurized water reactor cores. On the basis of modeling a nonlinear pressurized water reactor core using the lumped parameter method, its linearized model is achieved via the small perturbation linearization way. Linearized models of the nonlinear core at six power levels are selected as local models of this core. The T-S fuzzy idea for the core is exploited to construct the T-S fuzzy model of the nonlinear core based on th...

  20. Control Rod Ejection Accident while Using 6- and 8-Tube IRT-4M Fuel Assemblies in WWR-SM Research Reactor Core

    Energy Technology Data Exchange (ETDEWEB)

    Baytelesov, S.; Kungurov, F.; Safarov, A.; Salikhbaev, U.

    2011-07-01

    The WWR-SM reactor at the Institute of Nuclear Physics of the Academy of Sciences (INP AS) in Uzbekistan was converted to 6-tube IRT-4M LEU (19.7%) fuel in 2009. Presently, INP intends to also use IRT-4M 8-tube FA, and a safety analysis for these 'mixed' (8-tube and 6-tube FA) cores is required by the regulatory authorities. This paper presents results of control rod ejection transient analysis for these mixed cores

  1. Need for space-time analyses of research reactor transients

    Energy Technology Data Exchange (ETDEWEB)

    Jatuff, F.E. [INVAP S.E., de Bariloche (Argentina)

    1997-12-01

    The success of the point-kinetics approximation to represent the time behavior of research reactors relies on the fact that research reactor cores are small enough to be neutronically tightly coupled; the core is small when measured in diffusion lengths. This fact implies that a certain change in a part of the core is immediately observed by the whole system. The propagation of changes is so fast that the core exhibits a shape function that is practically unchanged during the transient; the amplitude function, the only unknown of the problem, represents the full knowledge of the core response. One is immediately warned to look for the truth of this assumption. How small should a research reactor core be to be sure that point kinetics is a valid assumption? This question is becoming increasingly important because the tendency is to increase the size of research reactor cores to make them capable of various simultaneous uses (multipurpose characteristics), with powers in the range of tens of thermal megawatts. One of the lines of investigation at the Department of Reactor Physics is related to scenarios of Materials Test Reactor (MTR)-type research reactor transients for which space-time kinetics would bring a more profound insight than point kinetics.

  2. Investigative studies on the effects of cadmium rabbits on high enriched uranium-fueled and low enriched uranium-fueled cores of Ghana Research Reactor-1 using MCNP5 code

    Energy Technology Data Exchange (ETDEWEB)

    Boffie, J., E-mail: jboffie@yahoo.com [Department of Nuclear Engineering and Material Science, School of Nuclear and Allied Sciences (SNAS), University of Ghana, P.O. Box AE 1, Atomic Energy, Accra (Ghana); National Nuclear Research Institute, Ghana Atomic Energy Commission, P.O. Box LG 80, Legon, Accra (Ghana); Akaho, E.H.K. [Department of Nuclear Engineering and Material Science, School of Nuclear and Allied Sciences (SNAS), University of Ghana, P.O. Box AE 1, Atomic Energy, Accra (Ghana); Nyarko, B.J.B.; Odoi, H.C.; Tuffour-Achampong, K.; Abrefah, R.G. [Department of Nuclear Engineering and Material Science, School of Nuclear and Allied Sciences (SNAS), University of Ghana, P.O. Box AE 1, Atomic Energy, Accra (Ghana); National Nuclear Research Institute, Ghana Atomic Energy Commission, P.O. Box LG 80, Legon, Accra (Ghana)

    2013-12-15

    Highlights: • The operating parameters for both the HEU core and proposed LEU core were similar. • The length of the Cd in the capsules must be increased for its use in the LEU core. • Cd rabbits can emergently be used to shut down MNSRs. - Abstract: Miniature Neutron Source Reactors (MNSRs) are noted to be among highly safe research reactors. However, because of its use of one control rod for reactivity control and shutdown purposes, alternative methods of shutting it down are important. The Ghana MNSR uses four cadmium rabbits of approximate dimensions 6.5 cm × 5.0 cm × 0.1 cm and mass of 9.48 g each to emergently shut down the reactor. The Monte Carlo N-Particle code; version 5, (MCNP5) was used to design the high enriched uranium (HEU) and low enriched uranium (LEU) cores of the MNSR with four cadmium rabbits inserted in four inner irradiation sites of each core. The operating parameters and shutdown parameters for both cores with the central control rod (CCR) either fully withdrawn or fully inserted had similar results with the HEU core having slightly better results in terms of safety. However, the results show that the four inserted cadmium rabbits make the HEU core subcritical whiles in the LEU core, it still remains critical (k{sub eff} = 1.00005 ± 0.00007). The length of the cadmium material in each cadmium rabbit must therefore be increased by at least 0.5 cm in order to attain subcriticality (k{sub eff} = 0.99989 ± 0.00006) and shutdown margin of 0.11 mk when inserted in the LEU core.

  3. On the oxidation of uraninite from natural reactor cores

    Energy Technology Data Exchange (ETDEWEB)

    Cui, D.; Eriksen, T.; Eklund, U.B.

    1999-07-01

    Natural nuclear reactors provide unique evidence in helping to understand the processes that might occur over long timescales in radioactive waste disposal sites. In the presented work, the extent and kinetics of oxidation of core material from the Oklo-Bangombe natural reactors are investigated. The X-ray powder diffraction analysis shows that the uraninites core samples from the Bangombe Reactor and Oklo Reactor 2, and Oklo Reactor 13 have the same unit-cell parameters as synthetic UO{sub 2.25}. A significant amount of fourmarierite, Pb(UO{sub 2}){sub 4}O{sub 3}(OH){sub 4}.4H{sub 2}O, was identified in the core samples from two shallow reactors Bangombe and Oklo 2, but not in the deeper reactor Oklo 13. The results of U(IV)/U(IV) measurements indicate that the extent of oxidative weathering of shallow reactors (Bangombe and Oklo 2) is greater than for the deeper reactor Oklo 13. Evaporable organic compounds found in the uraninite inclusion containing bitumen at the edge of Okelobondo Reactor (400 C) and in the black shale immediately above the Bangombe Reactor (260 C) may work as a reducing buffer or/and a hydrophobic water shield to depress the oxidative dissolution of the uraninite cores.

  4. State space modeling of reactor core in a pressurized water reactor

    Science.gov (United States)

    Ashaari, A.; Ahmad, T.; Shamsuddin, Mustaffa; M, Wan Munirah W.; Abdullah, M. Adib

    2014-07-01

    The power control system of a nuclear reactor is the key system that ensures a safe operation for a nuclear power plant. However, a mathematical model of a nuclear power plant is in the form of nonlinear process and time dependent that give very hard to be described. One of the important components of a Pressurized Water Reactor is the Reactor core. The aim of this study is to analyze the performance of power produced from a reactor core using temperature of the moderator as an input. Mathematical representation of the state space model of the reactor core control system is presented and analyzed in this paper. The data and parameters are taken from a real time VVER-type Pressurized Water Reactor and will be verified using Matlab and Simulink. Based on the simulation conducted, the results show that the temperature of the moderator plays an important role in determining the power of reactor core.

  5. State space modeling of reactor core in a pressurized water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ashaari, A.; Ahmad, T.; M, Wan Munirah W. [Department of Mathematical Science, Faculty of Science, Universiti Teknologi Malaysia, 81310 Skudai, Johor (Malaysia); Shamsuddin, Mustaffa [Institute of Ibnu Sina, Universiti Teknologi Malaysia, 81310 Skudai, Johor (Malaysia); Abdullah, M. Adib [Swinburne University of Technology, Faculty of Engineering, Computing and Science, Jalan Simpang Tiga, 93350 Kuching, Sarawak (Malaysia)

    2014-07-10

    The power control system of a nuclear reactor is the key system that ensures a safe operation for a nuclear power plant. However, a mathematical model of a nuclear power plant is in the form of nonlinear process and time dependent that give very hard to be described. One of the important components of a Pressurized Water Reactor is the Reactor core. The aim of this study is to analyze the performance of power produced from a reactor core using temperature of the moderator as an input. Mathematical representation of the state space model of the reactor core control system is presented and analyzed in this paper. The data and parameters are taken from a real time VVER-type Pressurized Water Reactor and will be verified using Matlab and Simulink. Based on the simulation conducted, the results show that the temperature of the moderator plays an important role in determining the power of reactor core.

  6. Hanging core support system for a nuclear reactor. [LMFBR

    Science.gov (United States)

    Burelbach, J.P.; Kann, W.J.; Pan, Y.C.; Saiveau, J.G.; Seidensticker, R.W.

    1984-04-26

    For holding the reactor core in the confining reactor vessel, a support is disclosed that is structurally independent of the vessel, that is dimensionally accurate and stable, and that comprises tandem tension linkages that act redundantly of one another to maintain stabilized core support even in the unlikely event of the complete failure of one of the linkages. The core support has a mounting platform for the reactor core, and unitary structure including a flange overlying the top edge of the reactor vessels, and a skirt and box beams between the flange and platform for establishing one of the linkages. A plurality of tension rods connect between the deck closing the reactor vessel and the platform for establishing the redundant linkage. Loaded Belleville springs flexibly hold the tension rods at the deck and separable bayonet-type connections hold the tension rods at the platform.

  7. Gaseous fuel nuclear reactor research

    Science.gov (United States)

    Schwenk, F. C.; Thom, K.

    1975-01-01

    Gaseous-fuel nuclear reactors are described; their distinguishing feature is the use of fissile fuels in a gaseous or plasma state, thereby breaking the barrier of temperature imposed by solid-fuel elements. This property creates a reactor heat source that may be able to heat the propellant of a rocket engine to 10,000 or 20,000 K. At this temperature level, gas-core reactors would provide the breakthrough in propulsion needed to open the entire solar system to manned and unmanned spacecraft. The possibility of fuel recycling makes possible efficiencies of up to 65% and nuclear safety at reduced cost, as well as high-thrust propulsion capabilities with specific impulse up to 5000 sec.

  8. VIPRE modeling of VVER-1000 reactor core for DNB analyses

    Energy Technology Data Exchange (ETDEWEB)

    Sung, Y.; Nguyen, Q. [Westinghouse Electric Corporation, Pittsburgh, PA (United States); Cizek, J. [Nuclear Research Institute, Prague, (Czech Republic)

    1995-09-01

    Based on the one-pass modeling approach, the hot channels and the VVER-1000 reactor core can be modeled in 30 channels for DNB analyses using the VIPRE-01/MOD02 (VIPRE) code (VIPRE is owned by Electric Power Research Institute, Palo Alto, California). The VIPRE one-pass model does not compromise any accuracy in the hot channel local fluid conditions. Extensive qualifications include sensitivity studies of radial noding and crossflow parameters and comparisons with the results from THINC and CALOPEA subchannel codes. The qualifications confirm that the VIPRE code with the Westinghouse modeling method provides good computational performance and accuracy for VVER-1000 DNB analyses.

  9. Reactor core design and characteristics of the Fugen

    Energy Technology Data Exchange (ETDEWEB)

    Matsumoto, Mitsuo; Kowata, Yasuki; Sugawara, Satoru; Deshimaru, Takehide

    1988-03-01

    The heavy water moderated, boiling light water cooled pressure tube type reactor Fugen uses plutonium-uranium mixed oxide as a fuel. Heavy water as the moderator and the light water of coolant are separated by the pressure tubes and calandria tubes. Thereby, the reactor core is heterogenes compared with that of LWRs. This paper describes the development of reactor core design procedure based on the feature of the Fugen type reactor, the feasibility test and the validity of nuclear and thermalhydraulic design based on the operating experience.

  10. Core management of the prototype heavy water reactor FUGEN

    Energy Technology Data Exchange (ETDEWEB)

    Deshimaru, Takehide; Furubayashi, Toshiyuki; Matsumoto, Mitsuo (Power Reactor and Nuclear Fuel Development Corp., Tokyo (Japan))

    1982-12-01

    In this paper, the core management which has been implemented so far for the prototype heavy water reactor FUGEN is described. First, the outline of the core is introduced. The core management is generally the repetition of planning, practice and evaluation, but the evaluation is specifically important in FUGEN because FUGEN is a prototype reactor. In the reactor FUGEN, the fuel replacement plan which determines the number and position of fuels to be replaced, and fuel procurement plan based on the replacement plan are prepared. The control rod pattern is determined so that the thermal limit for the fuel assembly is secured throughout the fuel cycle, but the output flattening by control rods is scarcely necessary by adopting a distributed replacement method. After a replaced core has been composed, the maximum excess reactivity and reactivity shut-down margin are mainly measured at the start-up of the reactor to confirm the predetermined characteristics of the replaced core. The core life can be simply and accurately estimated by the measurement of /sup 10/B concentration in heavy water. The output distribution in the core is an important parameter for calculating the performance of the FUGEN reactor core. The output increasing procedure is also controlled in accordance with that of light water reactors.

  11. Burnup concept for a long-life fast reactor core using MCNPX.

    Energy Technology Data Exchange (ETDEWEB)

    Holschuh, Thomas Vernon,; Lewis, Tom Goslee,; Parma, Edward J.,

    2013-02-01

    This report describes a reactor design with a burnup concept for a long-life fast reactor core that was evaluated using Monte Carlo N-Particle eXtended (MCNPX). The current trend in advanced reactor design is the concept of a small modular reactor (SMR). However, very few of the SMR designs attempt to substantially increase the lifetime of a reactor core, especially without zone loading, fuel reshuffling, or other artificial mechanisms in the core that %E2%80%9Cflatten%E2%80%9D the power profile, including non-uniform cooling, non-uniform moderation, or strategic poison placement. Historically, the limitations of computing capabilities have prevented acceptable margins in the temporal component of the spatial excess reactivity in a reactor design, due primarily to the error in burnup calculations. This research was performed as an initial scoping analysis into the concept of a long-life fast reactor. It can be shown that a long-life fast reactor concept can be modeled using MCNPX to predict burnup and neutronics behavior. The inherent characteristic of this conceptual design is to minimize the change in reactivity over the lifetime of the reactor. This allows the reactor to operate substantially longer at full power than traditional Light Water Reactors (LWRs) or other SMR designs. For the purpose of this study, a single core design was investigated: a relatively small reactor core, yielding a medium amount of power (~200 to 400 MWth). The results of this scoping analysis were successful in providing a preliminary reactor design involving metal U-235/U-238 fuel with HT-9 fuel cladding and sodium coolant at a 20% volume fraction.

  12. Analysis of the Gas Core Actinide Transmutation Reactor (GCATR)

    Science.gov (United States)

    Clement, J. D.; Rust, J. H.

    1977-01-01

    Design power plant studies were carried out for two applications of the plasma core reactor: (1) As a breeder reactor, (2) As a reactor able to transmute actinides effectively. In addition to the above applications the reactor produced electrical power with a high efficiency. A reactor subsystem was designed for each of the two applications. For the breeder reactor, neutronics calculations were carried out for a U-233 plasma core with a molten salt breeding blanket. A reactor was designed with a low critical mass (less than a few hundred kilograms U-233) and a breeding ratio of 1.01. The plasma core actinide transmutation reactor was designed to transmute the nuclear waste from conventional LWR's. The spent fuel is reprocessed during which 100% of Np, Am, Cm, and higher actinides are separated from the other components. These actinides are then manufactured as oxides into zirconium clad fuel rods and charged as fuel assemblies in the reflector region of the plasma core actinide transmutation reactor. In the equilibrium cycle, about 7% of the actinides are directly fissioned away, while about 31% are removed by reprocessing.

  13. A Preliminary Analysis of Reactor Performance Test (LOEP) for a Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hyeonil; Park, Su-Ki [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    The final phase of commissioning is reactor performance test, which is to prove the integrated performance and safety of the research reactor at full power with fuel loaded such as neutron power calibration, Control Absorber Rod/Second Shutdown Rod drop time, InC function test, Criticality, Rod worth, Core heat removal with natural mechanism, and so forth. The last test will be safety-related one to assure the result of the safety analysis of the research reactor is marginal enough to be sure about the nuclear safety by showing the reactor satisfies the acceptance criteria of the safety functions such as for reactivity control, maintenance of auxiliaries, reactor pool water inventory control, core heat removal, and confinement isolation. After all, the fuel integrity will be ensured by verifying there is no meaningful change in the radiation levels. To confirm the performance of safety equipment, loss of normal electric power (LOEP), possibly categorized as Anticipated Operational Occurrence (AOO), is selected as a key experiment to figure out how safe the research reactor is before turning over the research reactor to the owner. This paper presents a preliminary analysis of the reactor performance test (LOEP) for a research reactor. The results showed how different the transient between conservative estimate and best estimate will look. Preliminary analyses have shown all probable thermal-hydraulic transient behavior of importance as to opening of flap valve, minimum critical heat flux ratio, the change of flow direction, and important values of thermal-hydraulic parameters.

  14. Solid0Core Heat-Pipe Nuclear Batterly Type Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ehud Greenspan

    2008-09-30

    This project was devoted to a preliminary assessment of the feasibility of designing an Encapsulated Nuclear Heat Source (ENHS) reactor to have a solid core from which heat is removed by liquid-metal heat pipes (HP).

  15. Nuclear waste disposal utilizing a gaseous core reactor

    Science.gov (United States)

    Paternoster, R. R.

    1975-01-01

    The feasibility of a gaseous core nuclear reactor designed to produce power to also reduce the national inventories of long-lived reactor waste products through nuclear transmutation was examined. Neutron-induced transmutation of radioactive wastes is shown to be an effective means of shortening the apparent half life.

  16. Core Physics of Pebble Bed High Temperature Nuclear Reactors

    NARCIS (Netherlands)

    Auwerda, G.J.

    2014-01-01

    To more accurately predict the temperature distribution inside the reactor core of pebble bed type high temperature reactors, in this thesis we investigated the stochastic properties of randomly stacked beds and the effects of the non-homogeneity of these beds on the neutronics and thermal-hydraulic

  17. Development of core design and analyses technology for integral reactor

    Energy Technology Data Exchange (ETDEWEB)

    Zee, Sung Quun; Lee, C. C.; Song, J. S. and others

    1999-03-01

    Integral reactors are developed for the applications such as sea water desalination, heat energy for various industries, and power sources for large container ships. In order to enhance the inherent and passive safety features, low power density concept is chosen for the integral reactor SMART. Moreover, ultra-longer cycle and boron-free operation concepts are reviewed for better plant economy and simple design of reactor system. Especially, boron-free operation concept brings about large difference in core configurations and reactivity controls from those of the existing large size commercial nuclear power plants and also causes many differences in the safety aspects. The ultimate objectives of this study include detailed core design of a integral reactor, development of the core design system and technology, and finally acquisition of the system design certificate. The goal of the first stage is the conceptual core design, that is, to establish the design bases and requirements suitable for the boron-free concept, to develop a core loading pattern, to analyze the nuclear, thermal and hydraulic characteristics of the core and to perform the core shielding design. Interface data for safety and performance analyses including fuel design data are produced for the relevant design analysis groups. Nuclear, thermal and hydraulic, shielding design and analysis code systems necessary for the core conceptual design are established through modification of the existing design tools and newly developed methodology and code modules. Core safety and performance can be improved by the technology development such as boron-free core optimization, advaned core monitoring and operational aid system. Feasiblity study on the improvement of the core protection and monitoring system will also contribute toward core safety and performance. Both the conceptual core design study and the related technology will provide concrete basis for the next design phase. This study will also

  18. Power distributions in fresh and depleted LEU and HEU cores of the MITR reactor.

    Energy Technology Data Exchange (ETDEWEB)

    Wilson, E.H.; Horelik, N.E.; Dunn, F.E.; Newton, T.H., Jr.; Hu, L.; Stevens, J.G. (Nuclear Engineering Division); (2MIT Nuclear Reactor Laboratory and Nuclear Science and Engineering Department)

    2012-04-04

    The Massachusetts Institute of Technology Reactor (MITR-II) is a research reactor in Cambridge, Massachusetts designed primarily for experiments using neutron beam and in-core irradiation facilities. It delivers a neutron flux comparable to current LWR power reactors in a compact 6 MW core using Highly Enriched Uranium (HEU) fuel. In the framework of its non-proliferation policies, the international community presently aims to minimize the amount of nuclear material available that could be used for nuclear weapons. In this geopolitical context, most research and test reactors both domestic and international have started a program of conversion to the use of Low Enriched Uranium (LEU) fuel. A new type of LEU fuel based on an alloy of uranium and molybdenum (UMo) is expected to allow the conversion of U.S. domestic high performance reactors like the MITR-II reactor. Toward this goal, core geometry and power distributions are presented. Distributions of power are calculated for LEU cores depleted with MCODE using an MCNP5 Monte Carlo model. The MCNP5 HEU and LEU MITR models were previously compared to experimental benchmark data for the MITR-II. This same model was used with a finer spatial depletion in order to generate power distributions for the LEU cores. The objective of this work is to generate and characterize a series of fresh and depleted core peak power distributions, and provide a thermal hydraulic evaluation of the geometry which should be considered for subsequent thermal hydraulic safety analyses.

  19. Cooling Performance of Natural Circulation for a Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Park, Suki; Chun, J. H.; Yum, S. B. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    This paper deals with the core cooling performance by natural circulation during normal operation and a flow channel blockage event in an open tank-in-pool type research reactor. The cooling performance is predicted by using the RELAP5/ MOD3.3 code. The core decay heat is usually removed by natural circulation to the reactor pool water in open tank-in-pool type research reactors with the thermal power less than several megawatts. Therefore, these reactors have generally no active core cooling system against a loss of normal forced flow. In reactors with the thermal power less than around one megawatt, the reactor core can be cooled down by natural circulation even during normal full power operation. The cooling performance of natural circulation in an open tank-in-pool type research reactor has been investigated during the normal natural circulation and a flow channel blockage event. It is found that the maximum powers without void generation at the hot channel are around 1.16 MW and 820 kW, respectively, for the normal natural circulation and the flow channel blockage event.

  20. Nuclear gas core propulsion research program

    Science.gov (United States)

    Diaz, Nils J.; Dugan, Edward T.; Anghaie, Samim

    1993-01-01

    Viewgraphs on the nuclear gas core propulsion research program are presented. The objectives of this research are to develop models and experiments, systems, and fuel elements for advanced nuclear thermal propulsion rockets. The fuel elements under investigation are suitable for gas/vapor and multiphase fuel reactors. Topics covered include advanced nuclear propulsion studies, nuclear vapor thermal rocket (NVTR) studies, and ultrahigh temperature nuclear fuels and materials studies.

  1. COMSORS: A light water reactor chemical core catcher

    Energy Technology Data Exchange (ETDEWEB)

    Forsberg, C.W.; Parker, G.W.; Rudolph, J.C.; Osborne-Lee, I.W. [Oak Ridge National Lab., TN (United States). Chemical Technology Div.; Kenton, M.A. [Creare Inc., Hanover, NH (United States)

    1997-02-24

    The Core-Melt Source Reduction System (COMSORS) is a new approach to terminate lightwater reactor (LWR) core-melt accidents and ensure containment integrity. A special dissolution glass made of lead oxide (PbO) and boron oxide (B{sub 2}O{sub 3}) is placed under the reactor vessel. If molten core debris is released onto the glass, the following sequence happens: (1) the glass absorbs decay heat as its temperature increases and the glass softens; (2) the core debris dissolves into the molten glass; (3) molten glass convective currents create a homogeneous high-level waste (HLW) glass; (4) the molten glass spreads into a wider pool, distributing the heat for removal by radiation to the reactor cavity above or transfer to water on top of the molten glass; and (5) the glass solidifies as increased surface cooling area and decreasing radioactive decay heat generation allows heat removal to exceed heat generation.

  2. A study on the recriticality possibilities of fast reactor cores after a hypothetical core meltdown accident

    Energy Technology Data Exchange (ETDEWEB)

    Na, Byung Chan; Han, Do Hee; Kim, Young Cheol

    1997-04-01

    The preliminary and parametric sensitivity study on recriticality risk of fast reactor cores after a hypothetical total core meltdown accident was performed. Only the neutronic aspects of the accident was considered for this study, independent of the accident scenario. Estimation was made for the quantities of molten fuel which must be ejected out of the core in order to assure a sub-critical state. Diverse parameters were examined: molten pool type (homogenized or stratified), fuel temperature, conditions of the reactor core, core size (small or large), and fuel type (oxide, nitride, metal) (author). 7 refs.

  3. Optimization of a Potential New Core of the TRIGA Mark II Reactor Vienna

    Energy Technology Data Exchange (ETDEWEB)

    Khan, R.; Villa, M.; Bock, H.; Abele, H.; Steinhauser, G. [Vienna University of Technology-Atominstitut, Vienna (Austria)

    2011-07-01

    The TRIGA Mark II Vienna is one of the last TRIGA reactors utilizing a mixed core with High Enrich Uranium (HEU) fuel. Due to the US Fuel Return Program, the Vienna University of Technology/Atominstitut (ATI) is obliged to return its HEU fuel by 2019. There is no final decision on any further utilization of the Vienna research reactor beyond that point. However, of all possible scenarios of the future, the conversion of the current core into Low Enriched Uranium (LEU) fuel and the complete replacement of all existing 83 burned FE(s) by new fresh FE(s) are investigated herein. This paper presents detailed reactor design calculations for three different reactor cores. The core 1 employs 104-type, core 2 uses 108-type and core 3 is loaded with mixed TRIGA fuels (i.e. 104 and 108). The combination of the Monte Carlo based neutronics code MCNP5, Oak Ridge Isotope Generation and depletion code ORIGEN2 and diffusion theory based reactor physics program TRIGLAV is used for this study. On the basis of this neutronics study, the amount of fuel required for a possible future reactor operation and its cost minimization is presented in this paper. The criticality, core excess reactivity, length of initial life cycle and thermal flux density distribution is simulated for three different cores. Keeping the utilization of existing fourteen 104-type FE(s) (i.e. six burned and eight fresh FE(s)) in view, the core 3 is found the most economical, enduring and safe option for future of the TRIGA Mark II reactor in Vienna. (author)

  4. Fast reactor core concepts to improve transmutation efficiency

    Energy Technology Data Exchange (ETDEWEB)

    Fujimura, Koji; Kawashima, Katsuyuki [Hitachi Research Laboratory, Hitachi, Ltd., 7-1-1, Omika-cho, Hitachi-shi, Ibaraki, 319-1221 Japan (Japan); Itooka, Satoshi [Hitachi-GE Nuclear Energy, Ltd., 3-1-1, Saiwai-cho, Hitachi-shi, Ibaraki, 317-0073 Japan (Japan)

    2015-12-31

    Fast Reactor (FR) core concepts to improve transmutation efficiency were conducted. A heterogeneous MA loaded core was designed based on the 1000MWe-ABR breakeven core. The heterogeneous MA loaded core with Zr-H loaded moderated targets had a better transmutation performance than the MA homogeneous loaded core. The annular pellet rod design was proposed as one of the possible design options for the MA target. It was shown that using annular pellet MA rods mitigates the self-shielding effect in the moderated target so as to enhance the transmutation rate.

  5. Monte Carlo modelling of TRIGA research reactor

    Science.gov (United States)

    El Bakkari, B.; Nacir, B.; El Bardouni, T.; El Younoussi, C.; Merroun, O.; Htet, A.; Boulaich, Y.; Zoubair, M.; Boukhal, H.; Chakir, M.

    2010-10-01

    The Moroccan 2 MW TRIGA MARK II research reactor at Centre des Etudes Nucléaires de la Maâmora (CENM) achieved initial criticality on May 2, 2007. The reactor is designed to effectively implement the various fields of basic nuclear research, manpower training, and production of radioisotopes for their use in agriculture, industry, and medicine. This study deals with the neutronic analysis of the 2-MW TRIGA MARK II research reactor at CENM and validation of the results by comparisons with the experimental, operational, and available final safety analysis report (FSAR) values. The study was prepared in collaboration between the Laboratory of Radiation and Nuclear Systems (ERSN-LMR) from Faculty of Sciences of Tetuan (Morocco) and CENM. The 3-D continuous energy Monte Carlo code MCNP (version 5) was used to develop a versatile and accurate full model of the TRIGA core. The model represents in detailed all components of the core with literally no physical approximation. Continuous energy cross-section data from the more recent nuclear data evaluations (ENDF/B-VI.8, ENDF/B-VII.0, JEFF-3.1, and JENDL-3.3) as well as S( α, β) thermal neutron scattering functions distributed with the MCNP code were used. The cross-section libraries were generated by using the NJOY99 system updated to its more recent patch file "up259". The consistency and accuracy of both the Monte Carlo simulation and neutron transport physics were established by benchmarking the TRIGA experiments. Core excess reactivity, total and integral control rods worth as well as power peaking factors were used in the validation process. Results of calculations are analysed and discussed.

  6. 3D computer visualization and animation of CANDU reactor core

    Energy Technology Data Exchange (ETDEWEB)

    Qian, T.; Echlin, M.; Tonner, P.; Sur, B. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)

    1999-07-01

    Three-dimensional (3D) computer visualization and animation models of typical CANDU reactor cores (Darlington, Point Lepreau) have been developed using world-wide-web (WWW) browser based tools: JavaScript, hyper-text-markup language (HTML) and virtual reality modeling language (VRML). The 3D models provide three-dimensional views of internal control and monitoring structures in the reactor core, such as fuel channels, flux detectors, liquid zone controllers, zone boundaries, shutoff rods, poison injection tubes, ion chambers. Animations have been developed based on real in-core flux detector responses and rod position data from reactor shutdown. The animations show flux changing inside the reactor core with the drop of shutoff rods and/or the injection of liquid poison. The 3D models also provide hypertext links to documents giving specifications and historical data for particular components. Data in HTML format (or other format such as PDF, etc.) can be shown in text, tables, plots, drawings, etc., and further links to other sources of data can also be embedded. This paper summarizes the use of these WWW browser based tools, and describes the resulting 3D reactor core static and dynamic models. Potential applications of the models are discussed. (author)

  7. An approach to model reactor core nodalization for deterministic safety analysis

    Science.gov (United States)

    Salim, Mohd Faiz; Samsudin, Mohd Rafie; Mamat @ Ibrahim, Mohd Rizal; Roslan, Ridha; Sadri, Abd Aziz; Farid, Mohd Fairus Abd

    2016-01-01

    Adopting good nodalization strategy is essential to produce an accurate and high quality input model for Deterministic Safety Analysis (DSA) using System Thermal-Hydraulic (SYS-TH) computer code. The purpose of such analysis is to demonstrate the compliance against regulatory requirements and to verify the behavior of the reactor during normal and accident conditions as it was originally designed. Numerous studies in the past have been devoted to the development of the nodalization strategy for small research reactor (e.g. 250kW) up to the bigger research reactor (e.g. 30MW). As such, this paper aims to discuss the state-of-arts thermal hydraulics channel to be employed in the nodalization for RTP-TRIGA Research Reactor specifically for the reactor core. At present, the required thermal-hydraulic parameters for reactor core, such as core geometrical data (length, coolant flow area, hydraulic diameters, and axial power profile) and material properties (including the UZrH1.6, stainless steel clad, graphite reflector) have been collected, analyzed and consolidated in the Reference Database of RTP using standardized methodology, mainly derived from the available technical documentations. Based on the available information in the database, assumptions made on the nodalization approach and calculations performed will be discussed and presented. The development and identification of the thermal hydraulics channel for the reactor core will be implemented during the SYS-TH calculation using RELAP5-3D® computer code. This activity presented in this paper is part of the development of overall nodalization description for RTP-TRIGA Research Reactor under the IAEA Norwegian Extra-Budgetary Programme (NOKEBP) mentoring project on Expertise Development through the Analysis of Reactor Thermal-Hydraulics for Malaysia, denoted as EARTH-M.

  8. An approach to model reactor core nodalization for deterministic safety analysis

    Energy Technology Data Exchange (ETDEWEB)

    Salim, Mohd Faiz, E-mail: mohdfaizs@tnb.com.my; Samsudin, Mohd Rafie, E-mail: rafies@tnb.com.my [Nuclear Energy Department, Regulatory Economics & Planning Division, Tenaga Nasional Berhad (Malaysia); Mamat Ibrahim, Mohd Rizal, E-mail: m-rizal@nuclearmalaysia.gov.my [Prototypes & Plant Development Center, Malaysian Nuclear Agency (Malaysia); Roslan, Ridha, E-mail: ridha@aelb.gov.my; Sadri, Abd Aziz [Nuclear Installation Divisions, Atomic Energy Licensing Board (Malaysia); Farid, Mohd Fairus Abd [Reactor Technology Center, Malaysian Nuclear Agency (Malaysia)

    2016-01-22

    Adopting good nodalization strategy is essential to produce an accurate and high quality input model for Deterministic Safety Analysis (DSA) using System Thermal-Hydraulic (SYS-TH) computer code. The purpose of such analysis is to demonstrate the compliance against regulatory requirements and to verify the behavior of the reactor during normal and accident conditions as it was originally designed. Numerous studies in the past have been devoted to the development of the nodalization strategy for small research reactor (e.g. 250kW) up to the bigger research reactor (e.g. 30MW). As such, this paper aims to discuss the state-of-arts thermal hydraulics channel to be employed in the nodalization for RTP-TRIGA Research Reactor specifically for the reactor core. At present, the required thermal-hydraulic parameters for reactor core, such as core geometrical data (length, coolant flow area, hydraulic diameters, and axial power profile) and material properties (including the UZrH{sub 1.6}, stainless steel clad, graphite reflector) have been collected, analyzed and consolidated in the Reference Database of RTP using standardized methodology, mainly derived from the available technical documentations. Based on the available information in the database, assumptions made on the nodalization approach and calculations performed will be discussed and presented. The development and identification of the thermal hydraulics channel for the reactor core will be implemented during the SYS-TH calculation using RELAP5-3D{sup ®} computer code. This activity presented in this paper is part of the development of overall nodalization description for RTP-TRIGA Research Reactor under the IAEA Norwegian Extra-Budgetary Programme (NOKEBP) mentoring project on Expertise Development through the Analysis of Reactor Thermal-Hydraulics for Malaysia, denoted as EARTH-M.

  9. A design study of reactor core optimization for direct nuclear heat-to-electricity conversion in a space power reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yoshikawa, Hidekazu; Takahashi, Makoto; Shimoda, Hiroshi; Takeoka, Satoshi [Kyoto Univ. (Japan); Nakagawa, Masayuki; Kugo, Teruhiko

    1998-01-01

    To propose a new design concept of a nuclear reactor used in the space, research has been conducted on the conceptual design of a new nuclear reactor on the basis of the following three main concepts: (1) Thermionic generation by thermionic fuel elements (TFE), (2) reactivity control by rotary reflector, and (3) reactor cooling by liquid metal. The outcomes of the research are: (1) A calculation algorithm was derived for obtaining convergent conditions by repeating nuclear characteristic calculation and thermal flow characteristic calculation for the space nuclear reactor. (2) Use of this algorithm and the parametric study established that a space nuclear reactor using 97% enriched uranium nitride as the fuel and lithium as the coolant and having a core with a radius of about 25 cm, a height of about 50 cm and a generation efficiency of about 7% can probably be operated continuously for at least more than ten years at 100 kW only by reactivity control by rotary reflector. (3) A new CAD/CAE system was developed to assist design work to optimize the core characteristics of the space nuclear reactor comprehensively. It is composed of the integrated design support system VINDS using virtual reality and the distributed system WINDS to collaboratively support design work using Internet. (N.H.)

  10. Safe operation and maintenance of research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Munsorn, S. [Reactor Operation Division, Office of Atomic Energy for Peace, Chatuchak, Bangkok (Thailand)

    1999-10-01

    The first Thai Research Reactor (TRR-1) was established in 1961 at the Office of Atomic Energy for Peace (OAEP), Bangkok. The reactor was light water moderated and cooled, using HEU plate-type with U{sub 3}O{sub 8}- Al fuel meat and swimming pool type. The reactor went first critical on October 27, 1962 and had been licensed to operate at 1 MW (thermal). On June 30, 1975 the reactor was shutdown for modification and the core and control system was disassemble and replaced by that of TRIGA Mark III type while the pool cooling system, irradiation facilities and other were kept. Thus the name TRR-1/M1' has been designed due to this modification the fuel has been changed from HEU plate type to Uranium Zirconium Hydride (UZrH) Low Enrichment Uranium (LEU) which include 4 Fuel Follower Control Rods and 1 Air Follower Control Rod. The TRR-1/M1 went critical on November 7, 1977 and the purpose of the operation are training, isotope production and research. Nowadays the TRR-1/M1 has been operated with core loading No.12 which released power of 1,056 MWD. (as of October 1998). The TRR-1/M1 has been operated at the power of 1.2 MW, three days a week with 34 hours per week, Shut-down on Monday for weekly maintenance and Tuesday for special experiment. The everage energy released is about 40.8 MW-hour per week. Every year, the TRR-1/M1 is shut-down about 2 months between February to March for yearly maintenance. (author)

  11. Nuclear research reactors activities in INVAP

    Energy Technology Data Exchange (ETDEWEB)

    Ordonez, Juan Pablo [INVAP, Bariloche (Argentina)

    2013-07-01

    This presentation describes the different activities in the research reactor field that are being carried out by INVAP. INVAP is presently involved in the design of three new research reactors in three different countries. The RA-10 is a multipurpose reactor, in Argentina, planned as a replacement for the RA-3 reactor. INVAP was contracted by CNEA for carrying out the preliminary engineering for this reactor, and has recently been contracted by CNEA for the detailed engineering. CNEA groups are strongly involved in the design of this reactor. The RMB is a multipurpose reactor, planned by CNEN from Brazil. CNEN, through REDETEC, has contracted INVAP to carry out the preliminary engineering for this reactor. As the user requirements for RA-10 and RMB are very similar, an agreement was signed between Argentina and Brasil governments to cooperate in these two projects. The agreement included that both reactors would use the OPAL reactor in Australia, design and built by INVAP, as a reference reactor. INVAP has also designed the LPRR reactor for KACST in Saudi Arabia. The LPRR is a 30 kw reactor for educational purposes. KACST initially contracted INVAP for the engineering for this reactor and has recently signed the contract with INVAP for building the reactor. General details of these three reactors will be presented.

  12. Study on core design for reduced-moderation water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Okubo, Tsutomu [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2002-12-01

    The Reduced-Moderation Water Reactor (RMWR) is a water-cooled reactor with the harder neutron spectrum comparing with the LWR, resulting from low neutron moderation due to reduced water volume fraction. Based on the difference from the spectrum from the LWR, the conversion from U-238 to Pu-239 is promoted and the new cores preferable to effective utilization of uranium resource can be possible Design study of the RMWR core started in 1997 and new four core concepts (three BWR cores and one PWR core) are recently evaluated in terms of control rod worths, plutonium multiple recycle, high burnup and void coefficient. Comparative evaluations show needed incorporation of control rod programming and simplified PUREX process as well as development of new fuel cans for high burnup of 100 GW-d/t. Final choice of design specifications will be made at the next step aiming at realization of the RMWR. (T. Tanaka)

  13. Study on secondary shutdown systems in Tehran research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jalali, H.R.; Fadaei, A.H., E-mail: Fadaei_amir@aut.ac.ir; Gharib, M.

    2015-09-15

    Highlights: • A study was undertaken to summarize the techniques for secondary shutdown systems (SSS). • Neutronic calculation performed for proposed systems as SSS. • Dumping the heavy water stored in the reflector vessel is capable to shut down reactor. • Neutronic and transient calculation was done for validating the selected SSS. • All calculation shown that this system has advantages in safety and neutron economy. - Abstract: One important safety aspect of any research reactor is the ability to shut down the reactor. Usually, research reactors, currently in operation, have a single shutdown system based on the simultaneous insertion of the all control rods into the reactor core through gravity. Nevertheless, the International Atomic Energy Agency currently recommends use of two shutdown systems which are fully independent from each other to guarantee secure shutdown when one of them fails. This work presents an investigative study into secondary shutdown systems, which will be an important safety component in the research reactor and will provide another alternative way to shut down the reactor emergently. As part of this project, a study was undertaken to summarize the techniques that are currently used at world-wide research reactors for recognizing available techniques to consider in research reactors. Removal of the reflector, removal of the fuels, change in critical shape of reactor core and insertion of neutron absorber between the core and reflector are selected as possible techniques in mentioned function. In the next step, a comparison is performed for these methods from neutronic aspects. Then, chosen method is studied from the transient behavior point of view. Tehran research reactor which is a 5 MW open-pool reactor selected as a case study and all calculations are carried out for it. It has 5 control rods which serve the purpose of both reactivity control and shutdown of reactor under abnormal condition. Results indicated that heavy

  14. Studying the effects of dynamical parameters on reactor core temperature

    Directory of Open Access Journals (Sweden)

    R Khodabakhsh

    2015-01-01

    Full Text Available In order to increase productivity, reduce depreciation, and avoid possible accidents in a system such as fuel rods' melting and overpressure, control of temperature changes in the reactor core is an important factor. There are several methods for solving and analysing the stability of point kinetics equations. In most previous analyses, the effects of various factors on the temperature of the reactor core have been ignored. In this work, the effects of various dynamical parameters on the temperature of the reactor core and stability of the system in the presence of temperature feedback reactivity with external reactivity step, ramp and sinusoidal for six groups of delayed neutrons were studied using the method of Lyapunov exponent. The results proved to be in good agreement with other works

  15. In-core fuel management for pebble-bed reactors

    Energy Technology Data Exchange (ETDEWEB)

    Milian Perez, Daniel; Rodriguez Garcia, Lorena; Garcia Hernandez, Carlos; Milian Lorenzo, Daniel, E-mail: dperez@instec.cu, E-mail: cgh@instec.cu, E-mail: dmilian@instec.cu [Higher Institute of Technologies and Applied Sciences, Havana (Cuba); Velasco, Abanades, E-mail: abanades@etsii.upm.es [Department of Simulation of Thermo Energy Systems, Polytechnic University of Madrid (Spain)

    2013-07-01

    In this paper a calculation procedure to reduce the power peak in the core of a Very High Temperature pebble bed Reactor is presented. This procedure combines the fuel depletion and the neutronic behavior of the fuel in the reactor core, modeling once-through-then-out cycles as well as cycles in which pebbles are recirculated through the core an arbitrary number of times, obtaining the asymptotic fuel-loading pattern. The procedure consists in several coupled computational codes, which are used iteratively until convergence is reached. The utilization of the MCNPX 2.6e, as one of these computational codes, is validated through the calculation of benchmarks announced by IAEA (IAEA-TECDOC-1249, 2001). To complete the verification of the calculation procedure a base case described in (Annals of Nuclear Energy 29 (2002) 1345-1364), was performed. The procedure has been applied to a model of Pebble Bed Modular Reactor (200 MW) design. (author)

  16. Assessment of HCDA energetics in the CRBRP heterogeneous reactor core

    Energy Technology Data Exchange (ETDEWEB)

    Rhow, S K; Switick, D M; McElroy, J L; Joe, B W; Elawar, Z J

    1981-03-27

    The results of hypothetical core disruptive event analyses for the CRBRP heterogeneous reactor core are reported. The analytical results cover a large number of parametric cases including variations in design parameters and phenomenological assumptions. Reactor core configurations at the beginning of cycle one and end of cycle four are evaluated. The energetic consequences are evaluated based upon both fuel expansion thermodynamic work potential and a relative probability assignment. It is concluded that the structural loads, which result from 101 megajoules of available expansion work at sodium slug impact on the reactor closure head (equivalent to 661 megajoules of fuel expansion work to one atmosphere), is an adequate energetic consequence envelope for use in specifying the Structural Margins Beyond the Design Basis.

  17. Estimative of core damage frequency in IPEN IEA-R1 research reactor due to the initiating events of loss of flow caused by channel blockage and loss of coolant caused by a large rupture in the pipe of the primary circuit - PSA level 1

    Energy Technology Data Exchange (ETDEWEB)

    Hirata, Daniel Massami [Centro Tecnologico da Marinha em Sao Paulo (CTMSP), Sao Paulo, SP (Brazil); Sabundjian, Gaiane, E-mail: gdjian@ipen.b [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP) Sao Paulo, SP (Brazil)

    2011-07-01

    This work applies the methodology of Probabilistic Safety Assessment Level 1 to the research reactor IEA-R1 IPEN-CNEN/SP. Two categories of identified initiating events of accidents in the reactor are studied: loss of flow and loss of primary coolant. Among the initiating events, blockage of flow channel and loss of cooling fluid caused by large pipe rupture in the primary circuit are chosen for a detailed analysis. The event tree technique is used to analyze the evolution of the accident, including the actuation or the fail of actuation of the safety systems and the reactor damages. Using the fault tree the reliability of the following reactor safety systems is evaluated: reactor shutdown system, isolation of the reactor pool, Emergency Core Cooling System (ECCS) and the electric system. Estimative for the frequency of damage to the reactor core and the probability of failure of the analyzed systems are calculated. The estimated values for the frequencies of core damage are within the expected margins and are of the same order of magnitude as those found for similar reactors. The reliability of the reactor shutdown system, isolation of the reactor pool and ECCS are satisfactory for the conditions in which these systems are required. However, for the electric system it is suggested an upgrade to increase its reliability. (author)

  18. Design and development of small and medium integral reactor core

    Energy Technology Data Exchange (ETDEWEB)

    Zee, Sung Quun; Chang, M. H.; Lee, C. C.; Song, J. S.; Cho, B. O.; Kim, K. Y.; Kim, S. J.; Park, S. Y.; Lee, K. B.; Lee, C. H.; Chun, T. H.; Oh, D. S.; In, W. K.; Kim, H. K.; Lee, C. B.; Kang, H. S.; Song, K. N.

    1997-07-01

    Recently, the role of small and medium size integral reactors is remarkable in the heat applications rather than the electrical generations. Such a range of possible applications requires extensive used of inherent safety features and passive safety systems. It also requires ultra-longer cycle operations for better plant economy. Innovative and evolutionary designs such as boron-free operations and related reactor control methods that are necessary for simple reactor system design are demanded for the small and medium reactor (SMR) design, which are harder for engineers to implement in the current large size nuclear power plants. The goals of this study are to establish preliminary design criteria, to perform the preliminary conceptual design and to develop core specific technology for the core design and analysis for System-integrated Modular Advanced ReacTor (SMART) of 330 MWt power. Based on the design criteria of the commercial PWR`s, preliminary design criteria will be set up. Preliminary core design concept is going to be developed for the ultra-longer cycle and boron-free operation and core analysis code system is constructed for SMART. (author). 100 refs., 40 tabs., 92 figs.

  19. Heat Pipe Reactor Dynamic Response Tests: SAFE-100 Reactor Core Prototype

    Science.gov (United States)

    Bragg-Sitton, Shannon M.

    2005-01-01

    The SAFE-I00a test article at the NASA Marshall Space Flight Center was used to simulate a variety of potential reactor transients; the SAFEl00a is a resistively heated, stainless-steel heat-pipe (HP)-reactor core segment, coupled to a gas-flow heat exchanger (HX). For these transients the core power was controlled by a point kinetics model with reactivity feedback based on core average temperature; the neutron generation time and the temperature feedback coefficient are provided as model inputs. This type of non-nuclear test is expected to provide reasonable approximation of reactor transient behavior because reactivity feedback is very simple in a compact fast reactor (simple, negative, and relatively monotonic temperature feedback, caused mostly by thermal expansion) and calculations show there are no significant reactivity effects associated with fluid in the HP (the worth of the entire inventory of Na in the core is .core thermal expansion via deflection measurements. It was found that core deflection was a strung function of how the SAFE-100 modules were fabricated and assembled (in terms of straightness, gaps, and other tolerances). To remove the added variable of how this particular core expands as compared to a different concept, it was decided to use a temperature based feedback model (based on several thermocouples placed throughout the core).

  20. Uranium droplet nuclear reactor core with MHD generator

    Science.gov (United States)

    Anghaie, Samim; Kumar, Ratan

    An innovative concept employing liquid uranium droplets as fuel in an ultrahigh-temperature vapor core reactor (UTVR) magnetohydrodynamic (MHD) generator power system for space power generation has been studied. Metallic vapor in superheated form acts as a working fluid for a closed-Rankine-type thermodynamic cycle. Usage of fuel and working fluid in this form assures certain advantages. The major technical issues emerging as a result involve a method for droplet generation, droplet transport in the reactor core, heat generation in the fuel and transport to the metallic vapor, and materials compatibility. A qualitative and quantitative attempt to resolve these issues has indicated the promise and tentative feasibility of the system.

  1. Core Flow Distribution from Coupled Supercritical Water Reactor Analysis

    Directory of Open Access Journals (Sweden)

    Po Hu

    2014-01-01

    Full Text Available This paper introduces an extended code package PARCS/RELAP5 to analyze steady state of SCWR US reference design. An 8 × 8 quarter core model in PARCS and a reactor core model in RELAP5 are used to study the core flow distribution under various steady state conditions. The possibility of moderator flow reversal is found in some hot moderator channels. Different moderator flow orifice strategies, both uniform across the core and nonuniform based on the power distribution, are explored with the goal of preventing the reversal.

  2. Thermal-hydraulic behavior of physical quantities at critical velocities in a nuclear research reactor core channel using plate type fuel

    Directory of Open Access Journals (Sweden)

    Sidi Ali Kamel

    2012-01-01

    Full Text Available The thermal-hydraulic study presented here relates to a channel of a nuclear reactor core. This channel is defined as being the space between two fuel plates where a coolant fluid flows. The flow velocity of this coolant should not generate vibrations in fuel plates. The aim of this study is to know the distribution of the temperature in the fuel plates, in the cladding and in the coolant fluid at the critical velocities of Miller, of Wambsganss, and of Cekirge and Ural. The velocity expressions given by these authors are function of the geometry of the fuel plate, the mechanical characteristics of the fuel plate’s material and the thermal characteristics of the coolant fluid. The thermal-hydraulic study is made under steady-state; the equation set-up of the thermal problem is made according to El Wakil and to Delhaye. Once the equation set-up is validated, the three critical velocities are calculated and then used in the calculations of the different temperature profiles. The average heat flux and the critical heat flux are evaluated for each critical velocity and their ratio reported. The recommended critical velocity to be used in nuclear channel calculations is that of Wambsganss. The mathematical model used is more precise and all the physical quantities, when using this critical velocity, stay in safe margins.

  3. Interim irradiated fuel storage facility for research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Lolich, Jose [INVAP SE, Bariloche (Argentina)

    2002-07-01

    In most research reactors irradiated fuel discharged from the reactor is initially stored underwater inside the reactor building for along period of time. This allows for heat dissipation and fission product decay. In most cases this initial storage is done in a irradiated fuel storage facility pool located closed to the reactor core. After a certain cooling time, the fuel discharged should be relocated for long-term interim storage in a Irradiated Fuel Storage (IFS) Facility. IFS facilities are required for the safe storage of irradiated nuclear fuel before it is reprocessed or conditioned for disposal as radioactive waste. The IFS Facility described in this report is not an integral part of an operating nuclear reactor. This facility many be either co-located with nuclear facilities (such as a nuclear reactor or reprocessing plant) or sited independently of other nuclear facilities. (author)

  4. Refurbishment of existing research reactors for BNCT

    Energy Technology Data Exchange (ETDEWEB)

    Jatuff, F.E.; Gessaghi, V. [INVAP S.E., de Bariloche (Argentina)

    1997-12-01

    Some research reactors have been selected for the development of boron neutron capture therapy (BNCT) in the United States like the Massachusetts Institute of Technology research reactor, the University of Missouri research reactor 2 or the Brookhaven Medical Research Reactor, among others. These reactors have received excellent analyses and designs to accommodate extremely optimized beam shaping assemblies (BSAs) for the proper tuning of neutron spectra and absorption of undesired particles such as photons and fast neutrons. Due to the importance of BNCT in these facilities, the physicists and engineers have used many degrees of freedom for the optimization process.

  5. Related activities on management of ageing of Dalat Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Pham Van Lam [Reactor Dept., Nuclear Research Institute, Dalat (Viet Nam)

    1998-10-01

    The Dalat Nuclear Research Reactor (DNRR) is a pool type research reactor which was reconstructed in 1982 from the previous 250 kW TRIGA-MARK II reactor. The reactor core, the control and instrumentation system, the primary and secondary cooling systems as well as other associated systems were newly designed and installed. The renovated reactor reached its initial criticality in November 1983 and attained its nominal power of 500 kW in February 1984. Since then DNRR has been operated safely. Retained structures of the former reactor such as the reactor aluminum tank, the graphite reflector, the thermal column, the horizontal beam tubes and the radiation concrete shielding are 35 years old. During the recent years, in-service inspection has been carried out, the reactor control and instrumentation system were renovated due to ageing and obsolescence of its components, reactor general inspection and refurbishment were performed. Efforts are being made to cope with ageing of old reactor components to maintain safe operation of the DNRR. (author)

  6. Gamma thermometer based reactor core liquid level detector

    Science.gov (United States)

    Burns, Thomas J.

    1983-01-01

    A system is provided which employs a modified gamma thermometer for determining the liquid coolant level within a nuclear reactor core. The gamma thermometer which normally is employed to monitor local core heat generation rate (reactor power), is modified by thermocouple junctions and leads to obtain an unambiguous indication of the presence or absence of coolant liquid at the gamma thermometer location. A signal processor generates a signal based on the thermometer surface heat transfer coefficient by comparing the signals from the thermocouples at the thermometer location. The generated signal is a direct indication of loss of coolant due to the change in surface heat transfer when coolant liquid drops below the thermometer location. The loss of coolant indication is independent of reactor power at the thermometer location. Further, the same thermometer may still be used for the normal power monitoring function.

  7. Wide-range structurally optimized channel for monitoring the certified power of small-core reactors

    Energy Technology Data Exchange (ETDEWEB)

    Koshelev, A. S., E-mail: alexsander.coshelev@yandex.ru; Kovshov, K. N.; Ovchinnikov, M. A.; Pikulina, G. N.; Sokolov, A. B. [Russian Federal Nuclear Center—All-Russian Scientific Research Institute of Experimental Physics (Russian Federation)

    2016-12-15

    The results of tests of a prototype version of a channel for monitoring the certified power of small-core reactors performed at the BR-K1 reactor at the All-Russian Scientific Research Institute of Experimental Physics are reported. An SNM-11 counter and commercial KNK-4 and KNK-3 compensated ion chambers were used as neutron detectors in the tested channel, and certified NCMM and CCMM measurement modules controlled by a PC with specialized software were used as measuring instruments. The specifics of metrological assurance of calibration of the channel in the framework of reactor power monitoring are discussed.

  8. Core Physics and Kinetics Calculations for the Fissioning Plasma Core Reactor

    Science.gov (United States)

    Butler, C.; Albright, D.

    2007-01-01

    Highly efficient, compact nuclear reactors would provide high specific impulse spacecraft propulsion. This analysis and numerical simulation effort has focused on the technical feasibility issues related to the nuclear design characteristics of a novel reactor design. The Fissioning Plasma Core Reactor (FPCR) is a shockwave-driven gaseous-core nuclear reactor, which uses Magneto Hydrodynamic effects to generate electric power to be used for propulsion. The nuclear design of the system depends on two major calculations: core physics calculations and kinetics calculations. Presently, core physics calculations have concentrated on the use of the MCNP4C code. However, initial results from other codes such as COMBINE/VENTURE and SCALE4a. are also shown. Several significant modifications were made to the ISR-developed QCALC1 kinetics analysis code. These modifications include testing the state of the core materials, an improvement to the calculation of the material properties of the core, the addition of an adiabatic core temperature model and improvement of the first order reactivity correction model. The accuracy of these modifications has been verified, and the accuracy of the point-core kinetics model used by the QCALC1 code has also been validated. Previously calculated kinetics results for the FPCR were described in the ISR report, "QCALC1: A code for FPCR Kinetics Model Feasibility Analysis" dated June 1, 2002.

  9. Status of Research on Online Fuel Damage Detection and Core Damage Assessment System

    Institute of Scientific and Technical Information of China (English)

    XU; Xi-an; JI; Song-tao; GAO; Yong-guang; SHI; Xiao-lei

    2012-01-01

    <正>The technique research on the online fuel element damage detection and reactor core damage assessment is one project in the research program of the technical research for reactor key equipment maintenance and detection. The main research objective is to develop an online fuel damage detection system (FDDS), a core damage assessment system (CDAS) and make the integration of the two systems.

  10. Core Optimization of a Deep-Burn Pebble Bed Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Brian Boer; Abderrafi M. Ougouag

    2010-06-01

    Achieving a high fuel burnup in the Deep-Burn (DB) pebble bed reactor design, while remaining within the limits for fuel temperature, power peaking and temperature reactivity feedback, is challenging. The high content of Pu and Minor Actinides in the Deep-Burn fuel significantly impacts the thermal neutron energy spectrum as compared to a ’standard’ UO2 fueled core. This can result in power and temperature peaking in the pebble bed core in locally thermalized regions near the graphite reflectors. Furthermore, the interplay of the Pu resonances of the neutron absorption cross sections at low-lying energies can lead to a positive temperature reactivity coefficient for the graphite moderator at certain operating conditions. The DB concept focuses on the destruction of spent fuel transuranics in TRISO coated particle fueled gas-cooled reactors with the aim of a fractional fuel burnup of 60-70% in fissions per initial metal atom (FIMA), using a single-pass, multi in-core fuel (re)cycling scheme. In principle, the DB pebble bed concept employs the same reactor designs as the present low enriched uranium core designs, i.e. the 400 MWth Pebble Bed Modular Reactor (PBMR-400). A Pu and Minor Actinide fueled PBMR-400 design serves as the starting point for a core optimization study. The fuel temperature, power peak, temperature reactivity coefficients, and burnup capabilities of the modified designs are analyzed with the PEBBED code. A code-to-code coupling with the PASTA code allows for the analysis of the TRISO fuel performance for both normal and Loss Of Forced Cooling conditions. An improved core design is sought, maximizing the fuel discharge burnup, while retaining negative temperature reactivity feedback coefficients for the entire temperature range and avoiding high fuel temperatures (fuel failure probabilities).

  11. Preliminary Core Analysis of a Micro Modular Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jo, Chang Keun; Chang, Jongwa [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Venneri, Francesco [Ultra Safe Nuclear Corporation, Los Alamos (United States); Hawari, Ayman [NC State Univ., Raleigh (United States)

    2014-05-15

    The Micro Modular Reactor (MMR) will be 'melt-down proof'(MDP) under all circumstances, including the complete loss of coolant, and will be easily transportable and retrievable, and suitable for use with very little site preparation and Balance of Plant (BOP) requirements for a variety of applications, from power generation and process heat applications in remote areas to grid-unattached locations, including ship propulsion. The Micro Modular Reactor design proposed in this paper has 3 meter diameter core (2 meter active core) which is suitable for 'factory manufactured' and has few tens year of service life for remote deployment. We confirmed the feasibility of long term service life by a preliminary neutronic analysis in terms of the excess reactivity, the temperature feedback coefficient, and the control margins. We are able to achieve a reasonably long core life time of 5 ∼ 10 years under typical thermal hydraulic condition of a helium cooled reactor. However, on a situation where longer service period and safety is important, we can reduce the power density to the level of typical pebble bed reactor. In this case we can design 10 MWt MMR with core diameter for 10 ∼ 40 years core life time without much loss in the economics. Several burnable poisons are studied and it is found that erbia mixed in the compact matrix seems reasonably good poison. The temperature feedback coefficients were remaining negative during lifetime. Drum type control rods at reflector region and few control rods inside core region are sufficient to control the reactivity during operation and to achieve safe cold shutdown state.

  12. Convective cooling in a pool-type research reactor

    Science.gov (United States)

    Sipaun, Susan; Usman, Shoaib

    2016-01-01

    A reactor produces heat arising from fission reactions in the nuclear core. In the Missouri University of Science and Technology research reactor (MSTR), this heat is removed by natural convection where the coolant/moderator is demineralised water. Heat energy is transferred from the core into the coolant, and the heated water eventually evaporates from the open pool surface. A secondary cooling system was installed to actively remove excess heat arising from prolonged reactor operations. The nuclear core consists of uranium silicide aluminium dispersion fuel (U3Si2Al) in the form of rectangular plates. Gaps between the plates allow coolant to pass through and carry away heat. A study was carried out to map out heat flow as well as to predict the system's performance via STAR-CCM+ simulation. The core was approximated as porous media with porosity of 0.7027. The reactor is rated 200kW and total heat density is approximately 1.07+E7 Wm-3. An MSTR model consisting of 20% of MSTR's nuclear core in a third of the reactor pool was developed. At 35% pump capacity, the simulation results for the MSTR model showed that water is drawn out of the pool at a rate 1.28 kg s-1 from the 4" pipe, and predicted pool surface temperature not exceeding 30°C.

  13. Convective cooling in a pool-type research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Sipaun, Susan, E-mail: susan@nm.gov.my [Malaysian Nuclear Agency, Industrial Technology Division, Blok 29T, Bangi 43200, Selangor (Malaysia); Usman, Shoaib, E-mail: usmans@mst.edu [Missouri University of Science and Technology, Nuclear Engineering, 222 Fulton Hall 301 W.14th St., Rolla 64509 MO (United States)

    2016-01-22

    A reactor produces heat arising from fission reactions in the nuclear core. In the Missouri University of Science and Technology research reactor (MSTR), this heat is removed by natural convection where the coolant/moderator is demineralised water. Heat energy is transferred from the core into the coolant, and the heated water eventually evaporates from the open pool surface. A secondary cooling system was installed to actively remove excess heat arising from prolonged reactor operations. The nuclear core consists of uranium silicide aluminium dispersion fuel (U{sub 3}Si{sub 2}Al) in the form of rectangular plates. Gaps between the plates allow coolant to pass through and carry away heat. A study was carried out to map out heat flow as well as to predict the system’s performance via STAR-CCM+ simulation. The core was approximated as porous media with porosity of 0.7027. The reactor is rated 200kW and total heat density is approximately 1.07+E7 Wm{sup −3}. An MSTR model consisting of 20% of MSTR’s nuclear core in a third of the reactor pool was developed. At 35% pump capacity, the simulation results for the MSTR model showed that water is drawn out of the pool at a rate 1.28 kg s{sup −1} from the 4” pipe, and predicted pool surface temperature not exceeding 30°C.

  14. Optimized core design and fuel management of a pebble-bed type nuclear reactor

    NARCIS (Netherlands)

    Boer, B.

    2009-01-01

    The core design of a pebble-bed type Very High Temperature Reactor (VHTR) is optimized, aiming for an increase of the coolant outlet temperature to 1000 C, while retaining its inherent safety features. The VHTR has been selected by the international Generation IV research initiative as one of the si

  15. Development and Assessment of Advanced Reactor Core Protection System

    Science.gov (United States)

    in, Wang-Kee; Park, Young-Ho; Baeg, Seung-Yeob

    An advanced core protection system for a pressurized water reactor, Reactor Core Protection System(RCOPS), was developed by adopting a high performance hardware platform and optimal system configuration. The functional algorithms of the core protection system were also improved to enhance the plant availability by reducing unnecessary reactor trips and increasing operational margin. The RCOPS consists of four independent safety channels providing a two-out-of-four trip logic. The reliability analysis using the reliability block diagram method showed the unavailability of the RCOPS to be lower than the conventional system. The failure mode and effects analysis demonstrated that the RCOPS does not lose its intended safety functions for most failures. New algorithms for the RCOPS functional design were implemented in order to avoid unnecessary reactor trips by providing auxiliary pre-trip alarms and signal validation logic for the control rod position. The new algorithms in the RCOPS were verified by comparing the RCOPS calculations with reference results. The new thermal margin algorithm for the RCOPS was expected to increase the operational margin to the limit for Departure from Nucleate Boiling Ratio (DNBR) by approximately 1%.

  16. Documented Safety Analysis Addendum for the Neutron Radiography Reactor Facility Core Conversion

    Energy Technology Data Exchange (ETDEWEB)

    Boyd D. Christensen

    2009-05-01

    The Neutron Radiography Reactor Facility (NRAD) is a Training, Research, Isotope Production, General Atomics (TRIGA) reactor which was installed in the Idaho National Laboratory (INL) Hot Fuels Examination Facility (HFEF) at the Materials and Fuels Complex (MFC) in the mid 1970s. The facility provides researchers the capability to examine both irradiated and non-irradiated materials in support of reactor fuel and components programs through non-destructive neutron radiography examination. The facility has been used in the past as one facet of a suite of reactor fuels and component examination facilities available to researchers at the INL and throughout the DOE complex. The facility has also served various commercial research activities in addition to the DOE research and development support. The reactor was initially constructed using Fuel Lifetime Improvement Program (FLIP)- type highly enriched uranium (HEU) fuel obtained from the dismantled Puerto Rico Nuclear Center (PRNC) reactor. In accordance with international non-proliferation agreements, the NRAD core will be converted to a low enriched uranium (LEU) fuel and will continue to utilize the PRNC control rods, control rod drives, startup source, and instrument console as was previously used with the HEU core. The existing NRAD Safety Analysis Report (SAR) was created and maintained in the preferred format of the day, combining sections of both DOE-STD-3009 and Nuclear Regulatory Commission Regulatory Guide 1.70. An addendum was developed to cover the refueling and reactor operation with the LEU core. This addendum follows the existing SAR format combining required formats from both the DOE and NRC. This paper discusses the project to successfully write a compliant and approved addendum to the existing safety basis documents.

  17. System Study: Reactor Core Isolation Cooling 1998–2012

    Energy Technology Data Exchange (ETDEWEB)

    T. E. Wierman

    2013-10-01

    This report presents an unreliability evaluation of the reactor core isolation cooling (RCIC) system at 31 U.S. commercial boiling water reactors. Demand, run hours, and failure data from fiscal year 1998 through 2012 for selected components were obtained from the Equipment Performance and Information Exchange (EPIX). The unreliability results are trended for the most recent 10 year period while yearly estimates for system unreliability are provided for the entire active period. No statistically significant increasing trend was identified in the HPCI results. Statistically significant decreasing trends were identified for RCIC start-only and 8-hour trends.

  18. System Study: Reactor Core Isolation Cooling 1998-2014

    Energy Technology Data Exchange (ETDEWEB)

    Schroeder, John Alton [Idaho National Lab. (INL), Idaho Falls, ID (United States). Risk Assessment and Management Services Dept.

    2015-12-01

    This report presents an unreliability evaluation of the reactor core isolation cooling (RCIC) system at 31 U.S. commercial boiling water reactors. Demand, run hours, and failure data from fiscal year 1998 through 2014 for selected components were obtained from the Institute of Nuclear Power Operations (INPO) Consolidated Events Database (ICES). The unreliability results are trended for the most recent 10 year period, while yearly estimates for system unreliability are provided for the entire active period. No statistically significant trends were identified in the RCIC results.

  19. System Study: Reactor Core Isolation Cooling 1998–2013

    Energy Technology Data Exchange (ETDEWEB)

    Schroeder, John Alton [Idaho National Lab. (INL), Idaho Falls, ID (United States). Risk Assessment and Management Services Dept.

    2015-01-31

    This report presents an unreliability evaluation of the reactor core isolation cooling (RCIC) system at 31 U.S. commercial boiling water reactors. Demand, run hours, and failure data from fiscal year 1998 through 2013 for selected components were obtained from the Institute of Nuclear Power Operations (INPO) Consolidated Events Database (ICES). The unreliability results are trended for the most recent 10-year period, while yearly estimates for system unreliability are provided for the entire active period. No statistically significant trends were identified in the RCIC results.

  20. Gas core reactor power plants designed for low proliferation potential

    Energy Technology Data Exchange (ETDEWEB)

    Lowry, L.L. (comp.)

    1977-09-01

    The feasibility of gas core nuclear power plants to provide adequate power while maintaining a low inventory and low divertability of fissile material is studied. Four concepts were examined. Two used a mixture of UF/sub 6/ and helium in the reactor cavities, and two used a uranium-argon plasma, held away from the walls by vortex buffer confinement. Power levels varied from 200 to 2500 MWth. Power plant subsystems were sized to determine their fissile material inventories. All reactors ran, with a breeding ratio of unity, on /sup 233/U born from thorium. Fission product removal was continuous. Newly born /sup 233/U was removed continuously from the breeding blanket and returned to the reactor cavities. The 2500-MWth power plant contained a total of 191 kg of /sup 233/U. Less than 4 kg could be diverted before the reactor shut down. The plasma reactor power plants had smaller inventories. In general, inventories were about a factor of 10 less than those in current U.S. power reactors.

  1. 77 FR 36014 - Initial Test Program of Emergency Core Cooling Systems for Boiling-Water Reactors

    Science.gov (United States)

    2012-06-15

    ... COMMISSION Initial Test Program of Emergency Core Cooling Systems for Boiling-Water Reactors AGENCY: Nuclear...-1277, ``Initial Test Program of Emergency Core Cooling Systems for Boiling- Water Reactors.'' This... testing features of emergency core cooling systems (ECCSs) for boiling-water reactors (BWRs). DATES...

  2. Antineutrino emission and gamma background characteristics from a thermal research reactor

    CERN Document Server

    Bui, V M; Fallot, M; Communeau, V; Cormon, S; Estienne, M; Lenoir, M; Peuvrel, N; Shiba, T; Cucoanes, A S; Elnimr, M; Martino, J; Onillon, A; Porta, A; Pronost, G; Remoto, A; Thiolliere, N; Yermia, F; Zakari-Issoufou, A -A

    2016-01-01

    The detailed understanding of the antineutrino emission from research reactors is mandatory for any high sensitivity experiments either for fundamental or applied neutrino physics, as well as a good control of the gamma and neutron backgrounds induced by the reactor operation. In this article, the antineutrino emission associated to a thermal research reactor: the OSIRIS reactor located in Saclay, France, is computed in a first part. The calculation is performed with the summation method, which sums all the contributions of the beta decay branches of the fission products, coupled for the first time with a complete core model of the OSIRIS reactor core. The MCNP Utility for Reactor Evolution code was used, allowing to take into account the contributions of all beta decayers in-core. This calculation is representative of the isotopic contributions to the antineutrino flux which can be found at research reactors with a standard 19.75\\% enrichment in $^{235}$U. In addition, the required off-equilibrium correction...

  3. Role of research reactors for nuclear power program in Indonesia

    Energy Technology Data Exchange (ETDEWEB)

    Soentono, S.; Arbie, B. [National Atomic Energy Agency, Batan (Indonesia)

    1994-12-31

    The main objectives of nuclear development program in Indonesia are to master nuclear science and technology, as well as to utilise peaceful uses of nuclear know-how, aiming at stepwisely socioeconomic development. A Triga Mark II, previously of 250 kW, reactor in Bandung has been in operation since 1965 and its design power has been increased to 1000 kW in 1972. Using core grid of the Triga 250 kW, BATAN designed and constructed the Kartini Reactor in Yogyakarta which started its operation in 1979. Both of these Triga reactors have served a wide spectrum of utilisation, such as training of manpower in nuclear engineering as well as radiochemistry, isotope production and beam research activities in solid state physics. In order to support the nuclear power development program in general and to suffice the reactor experiments further, simultaneously meeting the ever increasing demand for radioisotope, the third reactor, a multipurpose reactor of 30 MW called GA. Siwabessy (RSG-GAS) has been in operation since 1987 at Serpong near Jakarta. Each of these reactors has strong cooperation with Universities, namely the Bandung Institute of Technology at Bandung, the Gadjah Mada University at Yogyakarta, and the Indonesia University at Jakarta and has facilitated the man power development required. The role of these reactors, especially the multipurpose GA. Siwabessy reactor, as essential tools in nuclear power program are described including the experience gained during preproject, construction and commissioning, as well as through their operation, maintenance and utilisation.

  4. Piezoelectric material for use in a nuclear reactor core

    Science.gov (United States)

    Parks, D. A.; Reinhardt, Brian; Tittmann, B. R.

    2012-05-01

    In radiation environments ultrasonic nondestructive evaluation has great potential for improving reactor safety and furthering the understanding of radiation effects and materials. In both nuclear power plants and materials test reactors, elevated temperatures and high levels of radiation present challenges to ultrasonic NDE methodologies. The challenges are primarily due to the degradation of the ultrasonic sensors utilized. We present results from the operation of a ultrasonic piezoelectric transducer, composed of bulk single crystal AlN, in a nuclear reactor core for over 120 MWHrs. The transducer was coupled to an aluminum cylinder and operated in pulse echo mode throughout the irradiation. In addition to the pulse echo testing impedance data were obtained. Further, the piezoelectric coefficient d33 was measured prior to irradiation and found to be 5.5 pC/N which is unchanged from as-grown samples, and in fact higher than the measured d33 for many as-grown samples.

  5. Superconducting shielded core reactor with reduced AC losses

    Energy Technology Data Exchange (ETDEWEB)

    Cha, Yung S.; Hull, John R.

    2006-04-04

    A superconducting shielded core reactor (SSCR) operates as a passive device for limiting excessive AC current in a circuit operating at a high power level under a fault condition such as shorting. The SSCR includes a ferromagnetic core which may be either closed or open (with an air gap) and extends into and through a superconducting tube or superconducting rings arranged in a stacked array. First and second series connected copper coils each disposed about a portion of the iron core are connected to the circuit to be protected and are respectively wound inside and outside of the superconducting tube or rings. A large impedance is inserted into the circuit by the core when the shielding capability of the superconducting arrangement is exceeded by the applied magnetic field generated by the two coils under a fault condition to limit the AC current in the circuit. The proposed SSCR also affords reduced AC loss compared to conventional SSCRs under continuous normal operation.

  6. Study for improvement of performance of the test and research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Sakurai, Fumio [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1999-03-01

    Current utilization needs for the test and research reactors become more advanced and diversified along with the advance of nuclear science and technology. Besides, the requested safety for the research and test reactors grows strictly every year as well as a case of the power reactors. Under this circumstance, every effort to improve reactor performance including its safety is necessary to be sustained for allowing more effective utilization of the test and research reactors as experimental apparatus for advanced researches. In this study, the following three themes i.e., JMTR high-performance fuel element, evaluation method of fast neutron irradiation dose in the JMTR, evaluation method of performance of siphon break valve as core covering system for water-cooled test and research reactors, were investigated respectively from the views of improvement of core performance as a neutron source, utilization performance as an experimental apparatus, and safety as a reactor plant. (author)

  7. Sequential reactions directed by core/shell catalytic reactors.

    Science.gov (United States)

    Wei, Yanhu; Soh, Siowling; Apodaca, Mario M; Kim, Jiwon; Grzybowski, Bartosz A

    2010-04-09

    Millimeter-sized reactor particles made of permeable polymer doped with catalysts arranged in a core/shell fashion direct sequences of chemical reactions (e.g., alkyne coupling followed by hydrogenation or hydrosilylation followed by hydrogenation). Spatial compartmentalization of catalysts coupled with the diffusion of substrates controls reaction order and avoids formation of byproducts. The experimentally observed yields of reaction sequences are reproduced by a theoretical model, which accounts for the reaction kinetics and the diffusion of the species involved.

  8. Performance of a multipurpose research electrochemical reactor

    Energy Technology Data Exchange (ETDEWEB)

    Henquin, E.R. [Programa de Electroquimica Aplicada e Ingenieria Electroquimica (PRELINE), Facultad de Ingenieria Quimica, Universidad Nacional del Litoral, Santiago del Estero 2829, S3000AOM Santa Fe (Argentina); Bisang, J.M., E-mail: jbisang@fiq.unl.edu.ar [Programa de Electroquimica Aplicada e Ingenieria Electroquimica (PRELINE), Facultad de Ingenieria Quimica, Universidad Nacional del Litoral, Santiago del Estero 2829, S3000AOM Santa Fe (Argentina)

    2011-07-01

    Highlights: > For this reactor configuration the current distribution is uniform. > For this reactor configuration with bipolar connection the leakage current is small. > The mass-transfer conditions are closely uniform along the electrode. > The fluidodynamic behaviour can be represented by the dispersion model. > This reactor represents a suitable device for laboratory trials. - Abstract: This paper reports on a multipurpose research electrochemical reactor with an innovative design feature, which is based on a filter press arrangement with inclined segmented electrodes and under a modular assembly. Under bipolar connection, the fraction of leakage current is lower than 4%, depending on the bipolar Wagner number, and the current distribution is closely uniform. When a turbulence promoter is used, the local mass-transfer coefficient shows a variation of {+-}10% with respect to its mean value. The fluidodynamics of the reactor responds to the dispersion model with a Peclet number higher than 10. It is concluded that this reactor is convenient for laboratory research.

  9. Testing of an Integrated Reactor Core Simulator and Power Conversion System with Simulated Reactivity Feedback

    Science.gov (United States)

    Bragg-Sitton, Shannon M.; Hervol, David S.; Godfroy, Thomas J.

    2010-01-01

    A Direct Drive Gas-Cooled (DDG) reactor core simulator has been coupled to a Brayton Power Conversion Unit (BPCU) for integrated system testing at NASA Glenn Research Center (GRC) in Cleveland, Ohio. This is a closed-cycle system that incorporates an electrically heated reactor core module, turboalternator, recuperator, and gas cooler. Nuclear fuel elements in the gas-cooled reactor design are replaced with electric resistance heaters to simulate the heat from nuclear fuel in the corresponding fast spectrum nuclear reactor. The thermodynamic transient behavior of the integrated system was the focus of this test series. In order to better mimic the integrated response of the nuclear-fueled system, a simulated reactivity feedback control loop was implemented. Core power was controlled by a point kinetics model in which the reactivity feedback was based on core temperature measurements; the neutron generation time and the temperature feedback coefficient are provided as model inputs. These dynamic system response tests demonstrate the overall capability of a non-nuclear test facility in assessing system integration issues and characterizing integrated system response times and response characteristics.

  10. Consequence analysis of core meltdown accidents in liquid metal fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Suk, S.D.; Hahn, D. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    2001-07-01

    Core disruptive accidents have been investigated at Korea Atomic Energy Research Institute(KAERI) as part of work to demonstrate the inherent and ultimate safety of the conceptual design of the Korea Advanced Liquid Metal Reactor(KALIMER), a 150 Mw pool-type sodium cooled prototype fast reactor that uses U-Pu-Zr metallic fuel. In this study, a simple method was developed using a modified Bethe-Tait method to simulate the kinetics and hydraulic behavior of a homogeneous spherical core over the period of the super-prompt critical power excursion induced by the ramp reactivity insertion. Calculations of energy release during excursions in the sodium-voided core of the KALIMER were subsequently performed using the method for various reactivity insertion rates up to 100 $/s, which has been widely considered to be the upper limit of ramp rates due to fuel compaction. Benchmark calculations were made to compare with the results of more detailed analysis for core meltdown energetics of the oxide fuelled fast reactor. A set of parametric studies was also performed to investigate the sensitivity of the results on the various thermodynamics and reactor parameters. (author)

  11. Advanced research reactor fuel development

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Chang Kyu; Pak, H. D.; Kim, K. H. [and others

    2000-05-01

    The fabrication technology of the U{sub 3}Si fuel dispersed in aluminum for the localization of HANARO driver fuel has been launches. The increase of production yield of LEU metal, the establishment of measurement method of homogeneity, and electron beam welding process were performed. Irradiation test under normal operation condition, had been carried out and any clues of the fuel assembly breakdown was not detected. The 2nd test fuel assembly has been irradiated at HANARO reactor since 17th June 1999. The quality assurance system has been re-established and the eddy current test technique has been developed. The irradiation test for U{sub 3}Si{sub 2} dispersed fuels at HANARO reactor has been carried out in order to compare the in-pile performance of between the two types of U{sub 3}Si{sub 2} fuels, prepared by both the atomization and comminution processes. KAERI has also conducted all safety-related works such as the design and the fabrication of irradiation rig, the analysis of irradiation behavior, thermal hydraulic characteristics, stress analysis for irradiation rig, and thermal analysis fuel plate, for the mini-plate prepared by international research cooperation being irradiated safely at HANARO. Pressure drop test, vibration test and endurance test were performed. The characterization on powders of U-(5.4 {approx} 10 wt%) Mo alloy depending on Mo content prepared by rotating disk centrifugal atomization process was carried out in order to investigate the phase stability of the atomized U-Mo alloy system. The {gamma}-U phase stability and the thermal compatibility of atomized U-16at.%Mo and U-14at.%Mo-2at.%X(: Ru, Os) dispersion fuel meats at an elevated temperature have been investigated. The volume increases of U-Mo compatibility specimens were almost the same as or smaller than those of U{sub 3}Si{sub 2}. However the atomized alloy fuel exhibited a better irradiation performance than the comminuted alloy. The RERTR-3 irradiation test of nano

  12. Reactor containment research and development

    Energy Technology Data Exchange (ETDEWEB)

    Weil, N. A.

    1963-06-15

    An outline is given of containment concepts, sources and release rates of energy, responses of containment structures, effects of projectiles, and leakage rates of radioisotopes, with particular regard to major reactor accidents. (T.F.H.)

  13. Simulation on reactor TRIGA Puspati core kinetics fueled with thorium (Th) based fuel element

    Science.gov (United States)

    Mohammed, Abdul Aziz; Pauzi, Anas Muhamad; Rahman, Shaik Mohmmed Haikhal Abdul; Zin, Muhamad Rawi Muhammad; Jamro, Rafhayudi; Idris, Faridah Mohamad

    2016-01-01

    In confronting global energy requirement and the search for better technologies, there is a real case for widening the range of potential variations in the design of nuclear power plants. Smaller and simpler reactors are attractive, provided they can meet safety and security standards and non-proliferation issues. On fuel cycle aspect, thorium fuel cycles produce much less plutonium and other radioactive transuranic elements than uranium fuel cycles. Although not fissile itself, Th-232 will absorb slow neutrons to produce uranium-233 (233U), which is fissile. By introducing Thorium, the numbers of highly enriched uranium fuel element can be reduced while maintaining the core neutronic performance. This paper describes the core kinetic of a small research reactor core like TRIGA fueled with a Th filled fuel element matrix using a general purpose Monte Carlo N-Particle (MCNP) code.

  14. Simulation on reactor TRIGA Puspati core kinetics fueled with thorium (Th) based fuel element

    Energy Technology Data Exchange (ETDEWEB)

    Mohammed, Abdul Aziz, E-mail: azizM@uniten.edu.my; Rahman, Shaik Mohmmed Haikhal Abdul [Universiti Tenaga Nasional. Jalan Ikram-UNITEN, 43000 Kajang, Selangor (Malaysia); Pauzi, Anas Muhamad, E-mail: anas@uniten.edu.my; Zin, Muhamad Rawi Muhammad; Jamro, Rafhayudi; Idris, Faridah Mohamad [Malaysian Nuclear Agency, Bangi, 43000 Kajang, Selangor (Malaysia)

    2016-01-22

    In confronting global energy requirement and the search for better technologies, there is a real case for widening the range of potential variations in the design of nuclear power plants. Smaller and simpler reactors are attractive, provided they can meet safety and security standards and non-proliferation issues. On fuel cycle aspect, thorium fuel cycles produce much less plutonium and other radioactive transuranic elements than uranium fuel cycles. Although not fissile itself, Th-232 will absorb slow neutrons to produce uranium-233 ({sup 233}U), which is fissile. By introducing Thorium, the numbers of highly enriched uranium fuel element can be reduced while maintaining the core neutronic performance. This paper describes the core kinetic of a small research reactor core like TRIGA fueled with a Th filled fuel element matrix using a general purpose Monte Carlo N-Particle (MCNP) code.

  15. Irradiation-Accelerated Corrosion of Reactor Core Materials. Final Report

    Energy Technology Data Exchange (ETDEWEB)

    Jiao, Zhujie [Univ. of Michigan, Ann Arbor, MI (United States); Was, Gary [Univ. of Michigan, Ann Arbor, MI (United States); Bartels, David [Univ. of Notre Dame, IN (United States)

    2015-04-02

    This project aims to understand how radiation accelerates corrosion of reactor core materials. The combination of high temperature, chemically aggressive coolants, a high radiation flux and mechanical stress poses a major challenge for the life extension of current light water reactors, as well as the success of most all GenIV concepts. Of these four drivers, the combination of radiation and corrosion places the most severe demands on materials, for which an understanding of the fundamental science is simply absent. Only a few experiments have been conducted to understand how corrosion occurs under irradiation, yet the limited data indicates that the effect is large; irradiation causes order of magnitude increases in corrosion rates. Without a firm understanding of the mechanisms by which radiation and corrosion interact in film formation, growth, breakdown and repair, the extension of the current LWR fleet beyond 60 years and the success of advanced nuclear energy systems are questionable. The proposed work will address the process of irradiation-accelerated corrosion that is important to all current and advanced reactor designs, but remains very poorly understood. An improved understanding of the role of irradiation in the corrosion process will provide the community with the tools to develop predictive models for in-reactor corrosion, and to address specific, important forms of corrosion such as irradiation assisted stress corrosion cracking.

  16. Power Control Method for Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Baang, Dane; Suh, Yongsuk; Park, Cheol [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    Considering safety-oriented design concept and other control environment, we developed a simple controller that provides limiting function of power change- rate as well as fine tracking performance. The design result has been well-proven via simulation and actual application to a TRIGA-II type research reactor. The proposed controller is designed to track the PDM(Power Demand) from operator input as long as maintaining the power change rate lower than a certain value for stable reactor operation. A power control method for a TRIGA-II type research reactor has been designed, simulated, and applied to actual reactor. The control performance during commissioning test shows that the proposed controller provides fine control performance for various changes in reference values (PDM), even though there is large measurement noise from neutron detectors. The overshoot at low power level is acceptable in a sense of reactor operation.

  17. Neutron flux optimization in irradiation channels at NUR research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Meftah, B. [Division Reacteur, Centre de Recherche Nucleaire Draria (CRND), BP 43 Sebala DRARIA, Alger (Algeria)]. E-mail: b_meftah@yahoo.com; Zidi, T. [Division Reacteur, Centre de Recherche Nucleaire Draria (CRND), BP 43 Sebala DRARIA, Alger (Algeria); Bousbia-Salah, A. [Dipartimento di Ingegneria Meccanica, Nucleari e della Produzione, Facolta di Ingegneria, Universita di Pisa, Via Diotisalvi, 2 - 56126 Pisa (Italy)

    2006-09-15

    Optimization of neutron fluxes in experimental channels is of great concern in research reactor utilization. The general approach used at the NUR research reactor for neutron flux optimization in irradiation channels is presented. The approach is essentially based upon a judicious optimization of the core configuration combined with the improvement of reflector characteristics. The method allowed to increase the thermal neutron flux for radioisotope production purposes by more than 800%. Increases of up to 60% are also observed in levels of useful fluxes available for neutron diffraction experiments (small angle neutron scattering (SANS), neutron reflectometry, etc.). Such improvements in the neutronic characteristics of the NUR reactor opened new perspectives in terms of its utilization. More particularly, it is now possible to produce at industrial scales major radio-isotopes for medicine and industry and to perform, for the first time, material testing experiments. The cost of the irradiations in the optimized configuration is generally small when compared to those performed in the old configuration and an average reduction factor of about of 10 is expected in the case of production of Molybdenum-99 (isotope required for the manufacturing of Technetium-99 medical kits). In addition to these important results, safety analysis studies showed that the more symmetrical nature of the core geometry leads to a more adequately balanced reactivity control system and contributes quite efficiently to the operational safety of the NUR reactor. Results of comparisons between calculations and measurements for a series of parameters of importance in reactor operation and safety showed good agreement.

  18. Reactivity feedback coefficients Pakistan research reactor-1 using PRIDE code

    Energy Technology Data Exchange (ETDEWEB)

    Mansoor, Ali; Ahmed, Siraj-ul-Islam; Khan, Rustam [Pakistan Institute of Engineering and Applied Sciences, Islamabad (Pakistan). Dept. of Nuclear Engineering; Inam-ul-Haq [Comsats Institute of Information Technology, Islamabad (Pakistan). Dept. of Physics

    2017-05-15

    Results of the analyses performed for fuel, moderator and void's temperature feedback reactivity coefficients for the first high power core configuration of Pakistan Research Reactor - 1 (PARR-1) are summarized. For this purpose, a validated three dimensional model of PARR-1 core was developed and confirmed against the reference results for reactivity calculations. The ''Program for Reactor In-Core Analysis using Diffusion Equation'' (PRIDE) code was used for development of global (3-dimensional) model in conjunction with WIMSD4 for lattice cell modeling. Values for isothermal fuel, moderator and void's temperature feedback reactivity coefficients have been calculated. Additionally, flux profiles for the five energy groups were also generated.

  19. Computation system for nuclear reactor core analysis. [LMFBR

    Energy Technology Data Exchange (ETDEWEB)

    Vondy, D.R.; Fowler, T.B.; Cunningham, G.W.; Petrie, L.M.

    1977-04-01

    This report documents a system which contains computer codes as modules developed to evaluate nuclear reactor core performance. The diffusion theory approximation to neutron transport may be applied with the VENTURE code treating up to three dimensions. The effect of exposure may be determined with the BURNER code, allowing depletion calculations to be made. The features and requirements of the system are discussed and aspects common to the computational modules, but the latter are documented elsewhere. User input data requirements, data file management, control, and the modules which perform general functions are described. Continuing development and implementation effort is enhancing the analysis capability available locally and to other installations from remote terminals.

  20. Dismantling decontamination of research reactor equipment

    Energy Technology Data Exchange (ETDEWEB)

    Voronik, N. I.; Davydov, Yu. P.; Shatilo, N. N. [Institute of Radioecological Problems Belarus Ac. Sci., Minsk-Sosny (Belarus)

    1999-07-01

    The purpose of the work was to check applicability of the existing and new compositions for decontamination and their adjustment to the specific conditions dealing with operation of the research reactor. (author)

  1. Development of a research nuclear reactor simulator using LABVIEW®

    Energy Technology Data Exchange (ETDEWEB)

    Lage, Aldo Marcio Fonseca; Mesquita, Amir Zacarias; Pinto, Antonio Juscelino; Souza, Luiz Claudio Andrade [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2015-07-01

    The International Atomic Energy Agency recommends the use of safety and friendly interfaces for monitoring and controlling the operational parameters of the nuclear reactors. The most important variable in the nuclear reactors control is the power released by fission of the fuel in the core which is directly proportional to neutron flux. It was developed a digital system to simulate the neutron evolution flux and monitoring their interaction on the other operational parameters. The control objective is to bring the reactor power from its source level (mW) to a few W. It is intended for education of basic reactor neutronic principles such as the multiplication factor, criticality, reactivity, period, delayed neutron and control by rods. The 250 kW IPR-R1 TRIGA research reactor at Nuclear Technology Development Center - CDTN (Belo Horizonte/Brazil) was used as reference. TRIGA reactors, developed by General Atomics (GA), are the most widely used research reactor in the world. They are cooled by light water under natural convection and are characterized by being inherently safety. The simulation system was developed using the LabVIEW® (Laboratory Virtual Instruments Engineering Workbench) software, considering the modern concept of virtual instruments (VI's). The main purpose of the system is to provide to analyze the behavior, and the tendency of some processes that occur in the reactor using a user-friendly operator interface. The TRIGA simulator system will allow the study of parameters, which affect the reactor operation, without the necessity of using the facility.(author)

  2. Determination of the kinetic parameters of the CALIBAN metallic core reactor from stochastic neutron measurements

    Energy Technology Data Exchange (ETDEWEB)

    Casoli, P.; Authier, N.; Chapelle, A. [Commissariat a l' Energie Atomique et Aux Energies Alternatives, CEA, DAM, F-21120 Is sur Tille (France)

    2012-07-01

    Several experimental devices are operated by the Criticality and Neutron Science Research Dept. of the CEA Valduc Laboratory. One of these is the Caliban metallic core reactor. The purpose of this study is to develop and perform experiments allowing to determinate some of fundamental kinetic parameters of the reactor. The prompt neutron decay constant and particularly its value at criticality can be measured with reactor noise techniques such as Rossi-{alpha} and Feynman variance-to-mean methods. Subcritical, critical, and even supercritical experiments were performed. Fission chambers detectors were put nearby the core and measurements were analyzed with the Rossi-{alpha} technique. A new value of the prompt neutron decay constant at criticality was determined, which allows, using the Nelson number method, new evaluations of the effective delayed neutron fraction and the in core neutron lifetime. As an introduction of this paper, some motivations of this work are given in part 1. In part 2, principles of the noise measurements experiments performed at the CEA Valduc Laboratory are reminded. The Caliban reactor is described in part 3. Stochastic neutron measurements analysis techniques used in this study are then presented in part 4. Results of fission chamber experiments are summarized in part 5. Part 6 is devoted to the current work, improvement of the experimental device using He 3 neutron detectors and first results obtained with it. Finally, conclusions and perspectives are given in part 7. (authors)

  3. Supply of enriched uranium for research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Mueller, H. [NUKEM GmbH, Alzenau (Germany)

    1997-08-01

    Since the RERTR-meeting In Newport/USA in 1990 the author delivered a series of papers in connection with the fuel cycle for research reactors dealing with its front-end. In these papers the author underlined the need for unified specifications for enriched uranium metal suitable for the production of fuel elements and made proposals with regard to the re-use of in Europe reprocessed highly enriched uranium. With regard to the fuel cycle of research reactors the research reactor community was since 1989 more concentrating on the problems of its back-end since the USA stopped the acceptance of spent research reactor fuel on December 31, 1988. Now, since it is apparent that these back-end problem have been solved by AEA`s ability to reprocess and the preparedness of the USA to again accept physically spent research reactor fuel the author is focusing with this paper again on the front-end of the fuel cycle on the question whether there is at all a safe supply of low and high enriched uranium for research reactors in the future.

  4. BENCHMARK EVALUATION OF THE START-UP CORE REACTOR PHYSICS MEASUREMENTS OF THE HIGH TEMPERATURE ENGINEERING TEST REACTOR

    Energy Technology Data Exchange (ETDEWEB)

    John Darrell Bess

    2010-05-01

    The benchmark evaluation of the start-up core reactor physics measurements performed with Japan’s High Temperature Engineering Test Reactor, in support of the Next Generation Nuclear Plant Project and Very High Temperature Reactor Program activities at the Idaho National Laboratory, has been completed. The evaluation was performed using MCNP5 with ENDF/B-VII.0 nuclear data libraries and according to guidelines provided for inclusion in the International Reactor Physics Experiment Evaluation Project Handbook. Results provided include updated evaluation of the initial six critical core configurations (five annular and one fully-loaded). The calculated keff eigenvalues agree within 1s of the benchmark values. Reactor physics measurements that were evaluated include reactivity effects measurements such as excess reactivity during the core loading process and shutdown margins for the fully-loaded core, four isothermal temperature reactivity coefficient measurements for the fully-loaded core, and axial reaction rate measurements in the instrumentation columns of three core configurations. The calculated values agree well with the benchmark experiment measurements. Fully subcritical and warm critical configurations of the fully-loaded core were also assessed. The calculated keff eigenvalues for these two configurations also agree within 1s of the benchmark values. The reactor physics measurement data can be used in the validation and design development of future High Temperature Gas-cooled Reactor systems.

  5. Degraded core analysis for the pressurized-water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Gittus, J.H.

    1987-02-09

    An analysis of the likelihood and the consequences of 'degraded-core accidents' has been undertaken for the proposed Sizewell B PWR. In such accidents, degradation of the core geometry occurs as a result of overheating. Radionuclides are released and may enter the environment, causing harmful effects. The analysis concludes that degraded-core accidents are highly improbable, the plant having been designed to reduce the frequency of such accidents to a value of order 10/sup -6/ per year. Tbe building containing the reactor would only fail in a small proportion of degraded-core accidents. In the great majority of cases the containment would remain intact and the release of radioactivity to the environment would be small. The risk to individuals have been calculated for both immediate and long term effects. Although the estimates of risk are approximate, studies to investigate the uncertainties, and sensitivities to different assumptions, show that potential errors are small compared with the very large 'margin of safety' between the risks estimated for Sizewell B and those that already exist in society.

  6. Facility for a Low Power Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Chalker, R. G.

    1949-09-14

    Preliminary investigation indicates that a reactor facility with ample research provisions for use by University or other interested groups, featuring safety in design, can be economically constructed in the Los Angeles area. The complete installation, including an underground gas-tight reactor building, with associated storage and experiment assembly building, administration offices, two general laboratory buildings, hot latoratory and lodge, can be constructed for approxinately $1,500,000. This does not include the cost of the reactor itself or of its auxiliary equipment,

  7. Design Features of a Core Protection System for an Integral Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Koo, Bon Seung; In, Wang Kee; Kim, Keung Koo; Lee, Chung Chan; Zee, Sung Quun [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    2005-07-01

    A system-integrated modular advanced research reactor is under development in the KAERI. Therefore, it is required to design an advanced core protection system for an integral reactor and an online digital core protection system, SCOPS is being developed as a part of plant protection system. SCOPS calculates the minimum CHFR(Critical Heat Flux Ratio) and maximum LPD(Local Power Density) based on the several online measured system parameters, such as the excore detector signal, CEA positions, MCP pump speed, pressure and temperature. Calculated values are compared with predetermined limiting values and the trip signal is generated if necessary. This paper describes the basic design features of SCOPS and several output parameters for a simple test case are presented.

  8. Strengthening IAEA Safeguards for Research Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Reid, Bruce D. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Anzelon, George A. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Budlong-Sylvester, Kory [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2016-09-01

    During their December 10-11, 2013, workshop in Grenoble France, which focused on the history and future of safeguarding research reactors, the United States, France and the United Kingdom (UK) agreed to conduct a joint study exploring ways to strengthen the IAEA’s safeguards approach for declared research reactors. This decision was prompted by concerns about: 1) historical cases of non-compliance involving misuse (including the use of non-nuclear materials for production of neutron generators for weapons) and diversion that were discovered, in many cases, long after the violations took place and as part of broader pattern of undeclared activities in half a dozen countries; 2) the fact that, under the Safeguards Criteria, the IAEA inspects some reactors (e.g., those with power levels under 25 MWt) less than once per year; 3) the long-standing precedent of States using heavy water research reactors (HWRR) to produce plutonium for weapons programs; 4) the use of HEU fuel in some research reactors; and 5) various technical characteristics common to some types of research reactors that could provide an opportunity for potential proliferators to misuse the facility or divert material with low probability of detection by the IAEA. In some research reactors it is difficult to detect diversion or undeclared irradiation. In addition, infrastructure associated with research reactors could pose a safeguards challenge. To strengthen the effectiveness of safeguards at the State level, this paper advocates that the IAEA consider ways to focus additional attention and broaden its safeguards toolbox for research reactors. This increase in focus on the research reactors could begin with the recognition that the research reactor (of any size) could be a common path element on a large number of technically plausible pathways that must be considered when performing acquisition pathway analysis (APA) for developing a State Level Approach (SLA) and Annual Implementation Plan (AIP). To

  9. Detailed Burnup Calculations for Research Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Leszczynski, F. [Centro Atomico Bariloche (CNEA), 8400 S. C. de Bariloche (Argentina)

    2011-07-01

    A general method (RRMCQ) has been developed by introducing a microscopic burn up scheme which uses the Monte Carlo calculated spatial power distribution of a research reactor core and a depletion code for burn up calculations, as a basis for solving nuclide material balance equations for each spatial region in which the system is divided. Continuous energy dependent cross-section libraries and full 3D geometry of the system is input for the calculations. The resulting predictions for the system at successive burn up time steps are thus based on a calculation route where both geometry and cross-sections are accurately represented, without geometry simplifications and with continuous energy data. The main advantage of this method over the classical deterministic methods currently used is that RRMCQ System is a direct 3D method without the limitations and errors introduced on the homogenization of geometry and condensation of energy of deterministic methods. The Monte Carlo and burn up codes adopted until now are the widely used MCNP5 and ORIGEN2 codes, but other codes can be used also. For using this method, there is a need of a well-known set of nuclear data for isotopes involved in burn up chains, including burnable poisons, fission products and actinides. For fixing the data to be included on this set, a study of the present status of nuclear data is performed, as part of the development of RRMCQ method. This study begins with a review of the available cross-section data of isotopes involved in burn up chains for research nuclear reactors. The main data needs for burn up calculations are neutron cross-sections, decay constants, branching ratios, fission energy and yields. The present work includes results of selected experimental benchmarks and conclusions about the sensitivity of different sets of cross-section data for burn up calculations, using some of the main available evaluated nuclear data files. Basically, the RRMCQ detailed burn up method includes four

  10. USGS Core Research Center (CRC) Collection of Core

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The Core Research Center (CRC) was established in 1974 by the U.S. Geological Survey (USGS) to preserve valuable rock cores for use by scientists and educators from...

  11. PRIZMA predictions of in-core detection indications in the VVER-1000 reactor

    Science.gov (United States)

    Kandiev, Yadgar Z.; Kashayeva, Elena A.; Malyshin, Gennady N.; Modestov, Dmitry G.; Khatuntsev, Kirill E.

    2014-06-01

    The paper describes calculations which were done by the PRIZMA code(1) to predict indications of in-core rhodium detectors in the VVER-1000 reactor for some core fragments with allowance for fuel and rhodium burnout.

  12. Conversion Preliminary Safety Analysis Report for the NIST Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Diamond, D. J. [Brookhaven National Lab. (BNL), Upton, NY (United States); Baek, J. S. [Brookhaven National Lab. (BNL), Upton, NY (United States); Hanson, A. L. [Brookhaven National Lab. (BNL), Upton, NY (United States); Cheng, L-Y [Brookhaven National Lab. (BNL), Upton, NY (United States); Brown, N. [Brookhaven National Lab. (BNL), Upton, NY (United States); Cuadra, A. [Brookhaven National Lab. (BNL), Upton, NY (United States)

    2015-01-30

    The NIST Center for Neutron Research (NCNR) is a reactor-laboratory complex providing the National Institute of Standards and Technology (NIST) and the nation with a world-class facility for the performance of neutron-based research. The heart of this facility is the NIST research reactor (aka NBSR); a heavy water moderated and cooled reactor operating at 20 MW. It is fueled with high-enriched uranium (HEU) fuel elements. A Global Threat Reduction Initiative (GTRI) program is underway to convert the reactor to low-enriched uranium (LEU) fuel. This program includes the qualification of the proposed fuel, uranium and molybdenum alloy foil clad in an aluminum alloy, and the development of the fabrication techniques. This report is a preliminary version of the Safety Analysis Report (SAR) that would be submitted to the U.S. Nuclear Regulatory Commission (NRC) for approval prior to conversion. The report follows the recommended format and content from the NRC codified in NUREG-1537, “Guidelines for Preparing and Reviewing Applications for the Licensing of Non-power Reactors,” Chapter 18, “Highly Enriched to Low-Enriched Uranium Conversions.” The emphasis in any conversion SAR is to explain the differences between the LEU and HEU cores and to show the acceptability of the new design; there is no need to repeat information regarding the current reactor that will not change upon conversion. Hence, as seen in the report, the bulk of the SAR is devoted to Chapter 4, Reactor Description, and Chapter 13, Safety Analysis.

  13. Corrosion Minimization for Research Reactor Fuel

    Energy Technology Data Exchange (ETDEWEB)

    Eric Shaber; Gerard Hofman

    2005-06-01

    Existing university research reactors are being converted to use low-enriched uranium fue to eliminate the use of highly-enriched uranium. These conversions require increases in fuel loading that will result in the use of elements with more fuel plates, resulting in a net decrease in the water annulus between fuel plates. The proposed decrease in the water annulus raises questions about the requirements and stability of the surface hydroxide on the aluminum fuel cladding and the potential for runaway corrosion resulting in fuel over-temperature incidents. The Nuclear Regulatory Commission (NRC), as regulator for these university reactors, must ensure that proposed fuel modifications will not result in any increased risk or hazard to the reactor operators or the public. This document reviews the characteristics and behavior of aluminum hydroxides, analyzes the drivers for fuel plate corrosion, reviews relevant historical incidents, and provides recommendations on fuel design, surface treatment, and reactor operational practices to avoid corrosion issues.

  14. Decommissioning of the Salaspils Research Reactor

    Directory of Open Access Journals (Sweden)

    Abramenkovs Andris

    2011-01-01

    Full Text Available In May 1995, the Latvian government decided to shut down the Salaspils Research Reactor and to dispense with nuclear energy in the future. The reactor has been out of operation since July 1998. A conceptual study on the decommissioning of the Salaspils Research Reactor was drawn up by Noell-KRC-Energie- und Umwelttechnik GmbH in 1998-1999. On October 26th, 1999, the Latvian government decided to start the direct dismantling to “green-field” in 2001. The upgrading of the decommissioning and dismantling plan was carried out from 2003-2004, resulting in a change of the primary goal of decommissioning. Collecting and conditioning of “historical” radioactive wastes from different storages outside and inside the reactor hall became the primary goal. All radioactive materials (more than 96 tons were conditioned for disposal in concrete containers at the radioactive wastes depository “Radons” at the Baldone site. Protective and radiation measurement equipment of the personnel was upgraded significantly. All non-radioactive equipment and materials outside the reactor buildings were released for clearance and dismantled for reuse or conventional disposal. Contaminated materials from the reactor hall were collected and removed for clearance measurements on a weekly basis.

  15. The study of core melting phenomena in reactor severe accident of PWR

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jae Hong; Jeun, Gyoo Dong; Park, Seh In; Lim, Jae Hyuck; Park, Seong Yong [Hanyang Univ., Seoul (Korea, Republic of); Bang, Kwang Hyun; Kim, Ki Yong [Korea Maritime Univ., Busan (Korea, Republic of)

    1999-03-15

    After TMI-2 accident, it has been paid much attention to severe accidents beyond the design basis accidents and the research on the progress of severe accidents and mitigation and the closure of severe accidents has been actively performed. In particular, a great deal of uncertainties yet exist in the phase of late core melt progression and thus the research on this phase of severe accident progress has a key role in obtaining confidence in severe accident mitigation and nuclear reactor safety. In the present study, physics of late core melt progression, experimental data and the major phenomenological models of computer codes are reviewed and a direction of reducing the uncertainties in the late core melt progression is proposed.

  16. The study of core melting phenomena in reactor severe accident of PWR

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jae Hong; Jeun, Gyoo Dong; Park, Seh In; Lim, Jae Hyuck; Park, Seong Yong [Hanyang Univ., Seoul (Korea, Republic of); Bang, Kwang Hyun; Kim, Ki Yong [Korea Maritime Univ., Busan (Korea, Republic of)

    1999-03-15

    After TMI-2 accident, it has been paid much attention to severe accidents beyond the design basis accidents and the research on the progress of severe accidents and mitigation and the closure of severe accidents has been actively performed. In particular, a great deal of uncertainties yet exist in the phase of late core melt progression and thus the research on this phase of severe accident progress has a key role in obtaining confidence in severe accident mitigation and nuclear reactor safety. In the present study, physics of late core melt progression, experimental data and the major phenomenological models of computer codes are reviewed and a direction of reducing the uncertainties in the late core melt progression is proposed.

  17. A study on the late core melt progression in pressurized water reactor severe accidents

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jae Hong; Jeun Gyoo Dong; Bang, Kwang Hyun; Park, Seh In; Lim, Jae Hyuck; Park, Seong Yong [Hanyang Univ., Seoul (Korea, Republic of); Back, Hyung Hmm [Korea Maritime Univ., Busan (Korea, Republic of)

    1998-03-15

    After TMI-2 accidents, it has been paid much attention to severe accidents beyond the design basis accidents and the research on the progress of severe accidents and mitigation and the closure of severe accidents has been actively performed. In particular, a great deal of uncertainties yet exist in the phase of late core melt progression and thus the research on this phase of severe accident progress has a key role in obtaining in severe accident mitigation and nuclear reactor safety. In the present study, physics of late core melt progression, experimental data and the major phenomenological models of computer codes are reviewed and a direction of reducing the uncertainties in the late core melt progression os proposed.

  18. Design requirement for electrical system of an advanced research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Hoan Sung; Kim, H. K.; Kim, Y. K.; Wu, J. S.; Ryu, J. S

    2004-12-01

    An advanced research reactor is being designed since 2002 and the conceptual design has been completed this year for the several types of core. Also the fuel was designed for the potential cores. But the process system, the I and C system, and the electrical system design are under pre-conceptual stage. The conceptual design for those systems will be developed in the next year. Design requirements for the electrical system set up to develop conceptual design. The same goals as reactor design - enhance safety, reliability, economy, were applied for the development of the requirements. Also the experience of HANARO design and operation was based on. The design requirements for the power distribution, standby power supply, and raceway system will be used for the conceptual design of electrical system.

  19. Machine Learning of the Reactor Core Loading Pattern Critical Parameters

    Directory of Open Access Journals (Sweden)

    Krešimir Trontl

    2008-01-01

    Full Text Available The usual approach to loading pattern optimization involves high degree of engineering judgment, a set of heuristic rules, an optimization algorithm, and a computer code used for evaluating proposed loading patterns. The speed of the optimization process is highly dependent on the computer code used for the evaluation. In this paper, we investigate the applicability of a machine learning model which could be used for fast loading pattern evaluation. We employ a recently introduced machine learning technique, support vector regression (SVR, which is a data driven, kernel based, nonlinear modeling paradigm, in which model parameters are automatically determined by solving a quadratic optimization problem. The main objective of the work reported in this paper was to evaluate the possibility of applying SVR method for reactor core loading pattern modeling. We illustrate the performance of the solution and discuss its applicability, that is, complexity, speed, and accuracy.

  20. LMFBR type reactor core and its fuel exchange method

    Energy Technology Data Exchange (ETDEWEB)

    Ishibashi, Yoko; Koyama, Jun-ichi; Aoyama, Motoo; Haikawa, Katsumasa; Yamanaka, Akihiro

    1996-08-20

    Upon initial loading, two kinds of fuel assemblies including first fuel assemblies having a highest enrichment degree and second fuel assemblies having a lowest enrichment degree are loaded. The average fuel enrichment degree of an upper region of the first fuel assembly is made greater than that of the lower region. The reactivity of the lower region of the first fuel assembly is made lower than that of the upper portion to reduce power peak. Upon transfer from a first cycle to a second cycle, at least one of the second fuel assemblies is exchanged by the same number of the third fuel assemblies. In this case, an average fuel enrichment degree of the upper region of the third fuel assembly is made greater than that of the lower region to suppress the reactivity in the lower region of the third fuel assembly lower than the reactivity in the upper region thereby reducing the power peak. Thus, the upper power peak over the entire reactor core is moderated thereby capable of ensuring the reactor shut down margin without deteriorating the same. (N.H.)

  1. Localization of Vibrating Noise Sources in Nuclear Reactor Cores

    Energy Technology Data Exchange (ETDEWEB)

    Hultqvist, Pontus

    2004-09-01

    In this thesis the possibility of locating vibrating noise sources in a nuclear reactor core from the neutron noise has been investigated using different localization methods. The influence of the vibrating noise source has been considered to be a small perturbation of the neutron flux inside the reactor. Linear perturbation theory has been used to construct the theoretical framework upon which the localization methods are based. Two different cases have been considered: one where a one-dimensional one-group model has been used and another where a two-dimensional two-energy group noise simulator has been used. In the first case only one localization method is able to determine the position with good accuracy. This localization method is based on finding roots of an equation and is sensitive to other perturbations of the neutron flux. It will therefore work better with the assistance of approximative methods that reconstruct the noise source to determine if the results are reliable or not. In the two-dimensional case the results are more promising. There are several different localization techniques that reproduce both the vibrating noise source position and the direction of vibration with enough precision. The approximate methods that reconstruct the noise source are substantially better and are able to support the root finding method in a more constructive way. By combining the methods, the results will be more reliable.

  2. Construction of linear empirical core models for pressurized water reactor in-core fuel management

    Energy Technology Data Exchange (ETDEWEB)

    Okafor, K.C.; Aldemir, T. (The Ohio State Univ., Dept. of Mechanical Engineering, Nuclear Engineering Program, 206 West 18th Ave., Columbus, OH (US))

    1988-06-01

    An empirical core model construction procedure for pressurized water reactor (PWR) in-core fuel management problems is presented that (a) incorporates the effect of composition changes in all the control zones in the core of a given fuel assembly, (b) is valid at all times during the cycle for a given range of control variables, (c) allows determining the optimal beginning of cycle (BOC) kappainfinity distribution as a single linear programming problem,and (d) provides flexibility in the choice of the material zones to describe core composition. Although the modeling procedure assumes zero BOC burnup, the predicted optimal kappainfinity profiles are also applicable to reload cores. In model construction, assembly power fractions and burnup increments during the cycle are regarded as the state (i.e., dependent) variables. Zone enrichments are the control (i.e., independent) variables. The model construction procedure is validated and implemented for the initial core of a PWR to determine the optimal BOC kappainfinity profiles for two three-zone scatter loading schemes. The predicted BOC kappainfinity profiles agree with the results of other investigators obtained by different modeling techniques.

  3. Emergency reactor core cooling water injection device for light water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Oda, Junro.

    1994-05-13

    A reactor pressure vessel is immersed in pool water of a reactor container. A control valve is interposed to a water supplying pipelines connecting pool water and a pressure vessel. A valve actuation means for opening/closing the control valve comprises a lifting tank. The inner side of the lifting tank and the inner side of the pressure vessel are connected by a communication pipeline (a syphon pipe) at upper and lower two portions. The lifting tank and the control valve are connected by a link mechanism. When a water level in the pressure vessel is lowered, the water level in the lifting tank is lowered to the same level as that in the pressure vessel. This reduces the weight of the lifting tank, the lifting tank is raised, to open the control valve by way of a link mechanism. As a result, liquid phase in the pressure vessel is in communication with the pool water, and the pool water flows down into the pressure vessel to maintain the reactor core in a flooded state. (I.N.).

  4. Annular core liquid-salt cooled reactor with multiple fuel and blanket zones

    Science.gov (United States)

    Peterson, Per F.

    2013-05-14

    A liquid fluoride salt cooled, high temperature reactor having a reactor vessel with a pebble-bed reactor core. The reactor core comprises a pebble injection inlet located at a bottom end of the reactor core and a pebble defueling outlet located at a top end of the reactor core, an inner reflector, outer reflector, and an annular pebble-bed region disposed in between the inner reflector and outer reflector. The annular pebble-bed region comprises an annular channel configured for receiving pebble fuel at the pebble injection inlet, the pebble fuel comprising a combination of seed and blanket pebbles having a density lower than the coolant such that the pebbles have positive buoyancy and migrate upward in said annular pebble-bed region toward the defueling outlet. The annular pebble-bed region comprises alternating radial layers of seed pebbles and blanket pebbles.

  5. The influence of reactor core parameters on effective breeding coefficient Keff

    Institute of Scientific and Technical Information of China (English)

    Liu Li-Po; Liu Yi-Bao; Wang Juan; Yang Bo; Zhang Tao

    2008-01-01

    The values of effective breeding coefficient Keff in a reactor core of nuclear power plant are calculated for different values of parameters (core structure, fuel assembly component) by using the Monte Carlo method. The obtained values of Keff are compared and analysed, which can provide theoretical basis for reactor design.

  6. Discussion about modeling the effects of neutron flux exposure for nuclear reactor core analysis

    Energy Technology Data Exchange (ETDEWEB)

    Vondy, D.R.

    1986-04-01

    Methods used to calculate the effects of exposure to a neutron flux are described. The modeling of the nuclear-reactor core history presents an analysis challenge. The nuclide chain equations must be solved, and some of the methods in use for this are described. Techniques for treating reactor-core histories are discussed and evaluated.

  7. Modular Pebble-Bed Reactor Project: Laboratory-Directed Research and Development Program FY 2002 Annual Report

    Energy Technology Data Exchange (ETDEWEB)

    Petti, David Andrew; Dolan, Thomas James; Miller, Gregory Kent; Moore, Richard Leroy; Terry, William Knox; Ougouag, Abderrafi Mohammed-El-Ami; Oh, Chang H; Gougar, Hans D

    2002-11-01

    This report documents the results of our research in FY-02 on pebble-bed reactor technology under our Laboratory Directed Research and Development (LDRD) project entitled the Modular Pebble-Bed Reactor. The MPBR is an advanced reactor concept that can meet the energy and environmental needs of future generations under DOE’s Generation IV initiative. Our work is focused in three areas: neutronics, core design and fuel cycle; reactor safety and thermal hydraulics; and fuel performance.

  8. Modular Pebble-Bed Reactor Project: Laboratory-Directed Research and Development Program FY 2002 Annual Report

    Energy Technology Data Exchange (ETDEWEB)

    Petti, David Andrew; Dolan, Thomas James; Miller, Gregory Kent; Moore, Richard Leroy; Terry, William Knox; Ougouag, Abderrafi Mohammed-El-Ami; Oh, Chang H; Gougar, Hans D

    2002-11-01

    This report documents the results of our research in FY-02 on pebble-bed reactor technology under our Laboratory Directed Research and Development (LDRD) project entitled the Modular Pebble-Bed Reactor. The MPBR is an advanced reactor concept that can meet the energy and environmental needs of future generations under DOE’s Generation IV initiative. Our work is focused in three areas: neutronics, core design and fuel cycle; reactor safety and thermal hydraulics; and fuel performance.

  9. In-Core-Instrumentation Methods for 3-Dimensional Distribution Information of Reactor Core Temperatures and Melt-down

    Energy Technology Data Exchange (ETDEWEB)

    Shin, Yeong Cheol [KHNP, Daejeon (Korea, Republic of); Eun, Myoung; Kim, Sung Jun [Woojin Inc., Hwaseong (Korea, Republic of)

    2014-08-15

    The tsunami-induced nuclear accident at Japanese Fukushima power plants in March 2011 has revealed some weaknesses in the severe accident monitoring system. The plant instrumentation did not provide utility, safety experts, and government officials with adequate and reliable information. The information on the reactor core damage and coolability is critical for making decisions correctly as well as in a timely manner during the course of the mitigation of severe accidents. Current Pressurized Water Reactor (PWR)s have an In-Core-Instrumentation (ICI) system that measures the temperature distribution of the top surface (i.e. Core Exit Temperatures) of the reactor core mainly to indicate when to begin Severe Accident Mitigation Guidelines (SAMG). This design concept giving only the core exit temperature has limitations in terms of sufficiency as well as availability of the information necessary for diagnosis on the status of the degraded core and the effectiveness of the measures taken as mitigation strategies. The reactor core exit temperatures are not sufficient to support the assessment of the degree of the core damage and the location of the molten core debris and recognition whether the core damage progresses on or it is mitigated. The ICI location being at the top of the reactor core also makes the ICI thermocouples vulnerable to melt-down because the upper part of the reactor core uncovers first, thereby melt down at the early stage of the accident. This means that direct indication of reactor core temperature will be lost and unavailable during the later stages of severe accident. To address the aforementioned weaknesses of the current ICIs, it is necessary to develop a new ICI system that provides information that is more expanded and more reliable for accident mitigation. With the enhanced information available, the SAMG can be prepared in more refined and effective way based on the direct and suitable indication of status of damages and the 3-dimensional

  10. Characteristic Research and Modeling of the Three-Core Magnetically Controlled Shunt Reactor%三铁柱磁饱和电抗器的建模与特性研究

    Institute of Scientific and Technical Information of China (English)

    赫志远; 李光友; 赵庆春

    2014-01-01

    针对三铁柱磁饱和电抗器的特点,导出了该电抗器的数学模型。在此基础上,利用磁路分段法建立了电抗器的仿真模型。在Matlab软件环境下对样机进行了仿真,得出了该样机的谐波特性、控制特性和响应时间,仿真结果与理论分析一致。%ABSTRACT:According to the characteristics of the three-core magnetically controlled shunt reactor,this paper deduces the mathematical model of the magnetically controlled shunt reactor. On this basis,the simulation model of the three-core magnetically controlled shunt reactor is established by using magnetic circuit decomposition approach. In Matlab software environment and by the simulation of the prototype it is concluded that the harmonic characteristics, the control characteristics and response time of the prototype simulation results are consistent with theoretical analysis.

  11. An experimental substantiation of the design functions imposed on the additional system for passively flooding the core of a VVER reactor

    Science.gov (United States)

    Morozov, A. V.; Remizov, O. V.

    2012-05-01

    Results obtained from a research work on experimentally substantiating the serviceability of the additional system for passively flooding the core of a VVER reactor from the second-stage hydro accumulators are presented.

  12. A Small Modular Reactor Core Design using FCM Fuel and BISO BP particles

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Jae Yeon; Hwang, Dae Hee; Yoo, Ho Seong; Hong, Ser Gi [Kyung Hee University, Yongin (Korea, Republic of)

    2016-10-15

    The objective of this work is to design a PWR small modular reactor which employs the advanced fuel technology of FCM particle fuels including BISO burnable poisons and advanced cladding of SiC in order to improve the fuel economy and safety by increasing fuel burnup and temperature, and by reducing hydrogen generation under accidents. Recently, many countries including USA have launched projects to develop the accident tolerant fuels (ATF) which can cope with the accidents such as LOCA (Loss of Coolant Accident). In general, the ATF fuels are required to meet the PWR operational, safety, and fuel cycle constraints which include enhanced burnup, lower or no generation of hydrogen, lower operating temperatures, and enhanced retention of fission products. Another stream of research and development in nuclear society is to develop advanced small modular reactors in order to improve inherent passive safety and to reduce the risk of large capital investment. In this work, a small PWR modular reactor core was neutronically designed and analyzed. The SMR core employs new 13x13 fuel assemblies which are loaded with thick FCM fuel rods in which TRISO fuel particles AO and also the first cycle has the AOs which are within the typical design limit. Also, this figure shows that the evolutions of AO for the cycles 6 and 7 are nearly the same. we considered the SiC cladding for reduction of hydrogen generation under accidents. From the results of core design and analysis, it is shown that the core has long cycle length of 732 -1191 EFPDs, high discharge burnup of 101-105 MWD/kg, low power peaking factors, low axial offsets, negative MTCs, and large shutdown margins except for BOC of the first cycle. So, it can be concluded that the new SMR core is neutronically feasible.

  13. ITHNA.SYS: An Integrated Thermal Hydraulic and Neutronic Analyzer SYStem for NUR research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Mazidi, S., E-mail: samirmazidi@gmail.com [Division Physique et Applications Nucléaires, Centre de Recherche Nucléaire de Draria (CRND), BP 43 Sebala, Draria, Alger (Algeria); Meftah, B., E-mail: b_meftah@yahoo.com [Division Physique et Applications Nucléaires, Centre de Recherche Nucléaire de Draria (CRND), BP 43 Sebala, Draria, Alger (Algeria); Belgaid, M., E-mail: belgaidm@yahoo.com [Faculté de Physique, Université Houari Boumediene, USTHB, BP 31, Bab Ezzouar, Alger (Algeria); Letaim, F., E-mail: fletaim@yahoo.fr [Faculté des Sciences et Technologies, Université d’El-oued, PO Box 789, El-oued (Algeria); Halilou, A., E-mail: hal_rane@yahoo.fr [Division Réacteur NUR, Centre de Recherche Nucléaire de Draria, BP 43 Sebala, Draria, Alger (Algeria)

    2015-08-15

    Highlights: • We develop a neutronic and thermal hydraulic MTR reactor analyzer. • The analyzer allows a rapid determination of the reactor core parameters. • Some NUR reactor parameters have been analyzed. - Abstract: This paper introduces the Integrated Thermal Hydraulic and Neutronic Analyzer SYStem (ITHNA.SYS) that has been developed for the Algerian research reactor NUR. It is used both as an operating aid tool and as a core physics engineering analysis tool. The system embeds three modules of the MTR-PC software package developed by INVAP SE: the cell calculation code WIMSD, the core calculation code CITVAP and the program TERMIC for thermal hydraulic analysis of a material testing reactor (MTR) core in forced convection. ITHNA.SYS operates both in on-line and off-line modes. In the on-line mode, the system is linked, via the computer parallel port, to the data acquisition console of the reactor control room and allows a real time monitoring of major physical and safety parameters of the NUR core. PC-based ITHNA.SYS provides a viable and convenient way of using an accumulated and often complex reactor physics stock of knowledge and frees the user from the intricacy of adequate reactor core modeling. This guaranties an accurate, though rapid, determination of a variety of neutronic and thermal hydraulic parameters of importance for the operation and safety analysis of the NUR research reactor. Instead of the several hours usually required, the processing time for the determination of such parameters is now reduced to few seconds. Validation of the system was performed with respect to experimental measurements and to calculations using reference codes. ITHNA.SYS can be easily adapted to accommodate other kinds of MTR reactors.

  14. Complete dismantling of the research reactor DIORIT

    Energy Technology Data Exchange (ETDEWEB)

    Beer, Hans-Frieder [Paul Scherrer Inst. (PSI), Villigen (Switzerland)

    2013-08-01

    The research reactor DIORIT at the Paul Scherrer Institute was a natural uranium reactor moderated by D{sub 2}O. It was put into operation in August 1960 and finally shut down in August 1977. The original dismantling plan, developed in 1980, comprised 3 phases and 13 steps. The dismantling started in 1982. It was interrupted for several times due to financial restrictions and during the last dismantling step due to the unexpected occurrence of asbestos. The dismantling could be successfully finished on September 11{sup th}, 2012. (orig.)

  15. Ultrahigh temperature vapor core reactor-MHD system for space nuclear electric power

    Science.gov (United States)

    Maya, Isaac; Anghaie, Samim; Diaz, Nils J.; Dugan, Edward T.

    1991-01-01

    The conceptual design of a nuclear space power system based on the ultrahigh temperature vapor core reactor with MHD energy conversion is presented. This UF4 fueled gas core cavity reactor operates at 4000 K maximum core temperature and 40 atm. Materials experiments, conducted with UF4 up to 2200 K, demonstrate acceptable compatibility with tungsten-molybdenum-, and carbon-based materials. The supporting nuclear, heat transfer, fluid flow and MHD analysis, and fissioning plasma physics experiments are also discussed.

  16. Fresh-Core Reload of the Neutron Radiography (NRAD) Reactor with Uranium(20)-Erbium-Zirconium-Hydride Fuel

    Energy Technology Data Exchange (ETDEWEB)

    John D. Bess; Thomas L. Maddock; Margaret A. Marshall; Leland M. Montierth

    2014-03-01

    The neutron radiography (NRAD) reactor is a 250 kW TRIGA® (Training, Research, Isotopes, General Atomics) Mark II , tank-type research reactor currently located in the basement, below the main hot cell, of the Hot Fuel Examination Facility (HFEF) at the Idaho National Laboratory (INL). It is equipped with two beam tubes with separate radiography stations for the performance of neutron radiography irradiation on small test components. The 60-fuel-element operational core configuration of the NRAD LEU TRIGA reactor has been evaluated as an acceptable benchmark experiment. The initial critical configuration developed during the fuel loading process, which contains only 56 fuel elements, has not been evaluated as it is very similar to the evaluated core configuration. The benchmark eigenvalue is 1.0012 ± 0.0029. Calculated eigenvalues differ significantly (~±1%) from the benchmark eigenvalue and have demonstrated sensitivity to the thermal scattering treatment of hydrogen in the U-Er-Zr-H fuel.

  17. Fresh-Core Reload of the Neutron Radiography (NRAD) Reactor with Uranium(20)-Erbium-Zirconium-Hydride Fuel

    Energy Technology Data Exchange (ETDEWEB)

    John D. Bess; Thomas L. Maddock; Margaret A. Marshall; Leland M. Montierth

    2011-03-01

    The neutron radiography (NRAD) reactor is a 250 kW TRIGA® (Training, Research, Isotopes, General Atomics) Mark II , tank-type research reactor currently located in the basement, below the main hot cell, of the Hot Fuel Examination Facility (HFEF) at the Idaho National Laboratory (INL). It is equipped with two beam tubes with separate radiography stations for the performance of neutron radiography irradiation on small test components. The 60-fuel-element operational core configuration of the NRAD LEU TRIGA reactor has been evaluated as an acceptable benchmark experiment. The initial critical configuration developed during the fuel loading process, which contains only 56 fuel elements, has not been evaluated as it is very similar to the evaluated core configuration. The benchmark eigenvalue is 1.0012 ± 0.0029. Calculated eigenvalues differ significantly (~±1%) from the benchmark eigenvalue and have demonstrated sensitivity to the thermal scattering treatment of hydrogen in the U-Er-Zr-H fuel.

  18. Investigation on the Core Bypass Flow in a Very High Temperature Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hassan, Yassin

    2013-10-22

    Uncertainties associated with the core bypass flow are some of the key issues that directly influence the coolant mass flow distribution and magnitude, and thus the operational core temperature profiles, in the very high-temperature reactor (VHTR). Designers will attempt to configure the core geometry so the core cooling flow rate magnitude and distribution conform to the design values. The objective of this project is to study the bypass flow both experimentally and computationally. Researchers will develop experimental data using state-of-the-art particle image velocimetry in a small test facility. The team will attempt to obtain full field temperature distribution using racks of thermocouples. The experimental data are intended to benchmark computational fluid dynamics (CFD) codes by providing detailed information. These experimental data are urgently needed for validation of the CFD codes. The following are the project tasks: • Construct a small-scale bench-top experiment to resemble the bypass flow between the graphite blocks, varying parameters to address their impact on bypass flow. Wall roughness of the graphite block walls, spacing between the blocks, and temperature of the blocks are some of the parameters to be tested. • Perform CFD to evaluate pre- and post-test calculations and turbulence models, including sensitivity studies to achieve high accuracy. • Develop the state-of-the art large eddy simulation (LES) using appropriate subgrid modeling. • Develop models to be used in systems thermal hydraulics codes to account and estimate the bypass flows. These computer programs include, among others, RELAP3D, MELCOR, GAMMA, and GAS-NET. Actual core bypass flow rate may vary considerably from the design value. Although the uncertainty of the bypass flow rate is not known, some sources have stated that the bypass flow rates in the Fort St. Vrain reactor were between 8 and 25 percent of the total reactor mass flow rate. If bypass flow rates are on the

  19. Mitigating IASCC of Reactor Core Internals by Post-Irradiation Annealing

    Energy Technology Data Exchange (ETDEWEB)

    Was, Gary [Univ. of Michigan, Ann Arbor, MI (United States)

    2017-06-02

    This final report summarizes research performed during the period between September 2012 and December 2016, with the objective of establishing the effectiveness of post-irradiation annealing (PIA) as an advanced mitigation strategy for irradiation-assisted stress corrosion cracking (IASCC). This was completed by using irradiated 304SS control blade material to conduct crack initiation and crack growth rate (CGR) experiments in simulated BWR environment. The mechanism by which PIA affects IASCC susceptibility will also be verified. The success of this project will provide a foundation for the use of PIA as a mitigation strategy for core internal components in commercial reactors.

  20. Ageing investigation and upgrading of components/systems of Kartini research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Syarip; Widi Setiawan [Yogyakarta Nuclear Research Centre, Yogyakarta (Indonesia)

    1998-10-01

    Kartini research reactor has been operated in good condition and has demonstrated successful operation for the past 18 years, utilized for: reactor kinetic and control studies, instrumentation tests, neutronic and thermohydraulic studies, routine neutron activation analysis, reactor safety studies, training for research reactor operators and supervisors, and reactor physics experiments. Several components of Kartini reactor use components from the abandoned IRT-2000 Project at Serpong and from Bandung Reactor Centre such as: reactor tank, reactor core, heat exchanger, motor blower for ventilation system, fuel elements, etc. To maintain a good operating performance and also for aging investigation purposes, the component failure data collection has been done. The method used is based on the Manual on Reliability Data Collection For Research Reactor PSAs, IAEA TECDOC 636, and analyzed by using Data Entry System (DES) computer code. Analysis result shows that the components/systems failure rate of Kartini reactor is around 1,5.10{sup -4} up to 2,8.10{sup -4} per hour, these values are within the ranges of the values indicated in IAEA TECDOC 478. Whereas from the analysis of irradiation history shows that the neutron fluence of fuel element with highest burn-up (2,05 gram U-235 in average) is around 1.04.10{sup 16} n Cm{sup -2} and this value is still far below its limiting value. Some reactor components/systems have been replaced and upgraded such as heat exchanger, instrumentation and control system (ICS), etc. The new reactor ICS was installed in 1994 which is designed as a distributed structure by using microprocessor based systems and bus system technology. The characteristic and operating performance of the new reactor ICS, as well as the operation history and improvement of the Kartini research reactor is presented. (J.P.N.)

  1. Neutron scattering at Australia's replacement research reactor

    Science.gov (United States)

    Robinson, R. A.; Kennedy, S. J.

    2002-01-01

    On August 25 1999, the Australian government gave final approval to build a research reactor to replace the existing HIFAR reactor at Lucas Heights. The replacement reactor, which will commence operation in 2005, will be multipurpose in function, with capabilities for both neutron-beam research and radioisotope production. Regarding beams, cold and thermal neutron sources are to be installed and the intent is to use supermirror guides, with coatings with critical angles up to 3 times that of natural Ni, to transport cold and thermal neutron beams into a large modern guide hall. The reactor and all the associated infrastructure, with the exception of the neutron beam instruments, is to be built by INVAP, SE and subcontractors in a turnkey contract. The goal is to have at least eight leading-edge neutron-beam instruments ready in 2005, and they will be developed by ANSTO and other contracted organisations, in consultation with the Australian user community and interested overseas parties. A review of the planned scientific capabilities, a description of the facility and a status report on the activities so far is given.

  2. The current status of Kartini research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Tri Wulan Tjiptono; Syarip

    1998-10-01

    The Kartini reactor reached the first criticality on January 25, 1979. In the first three years, the reactor power is limited up to 50 kW thermal power and on July 1, 1982 has been increased to 100 kW. It has been used as experiments facility by researcher of Atomic Energy National Agency and students of the Universities. Three beam tubes used as experiments facilities, the first, is used as a neutron source for H{sub 2}O-Natural Uranium Subcritical Assembly, the second, is developed for neutron radiography facility and the third, is used for gamma radiography facility. The other facilities are rotary rack and two pneumatic transfer systems, one for delayed neutron counting system and the other for the new Neutron Activation Analysis (NAA) facility. The rotary rack used for isotope production for NAA purpose (for long time irradiation), the delayed neutron counting system used for analysis the Uranium contents of the ores and the new NAA is provided for short live elements analysis. In the last three years the Reactor Division has a joint use program with the Nuclear Component and Engineering Center in research reactor instrumentation and control development. (author)

  3. Shape optimization of a Sodium Fast Reactor core

    Directory of Open Access Journals (Sweden)

    Dombre Emmanuel

    2013-01-01

    Full Text Available We apply in this paper a geometrical shape optimization method for the design of the core of a SFR (Sodium-cooled Fast Reactor in order to minimize a thermal counter-reaction known as the sodium void effect. In this kind of reactors, by increasing the temperature, the core may become liable to a strong increase of reactivity, a key-parameter governing the chain-reaction at quasi-static states. We first use the one group energy diffusion model and give the generalization to the two groups energy equation. We then give some numerical results in the case of the one group energy equation. Note that the application of our method leads to some designs whose interfaces can be parametrized by very smooth curves which can stand very far from realistic designs. We don’t explain here the method that it would be possible to use for recovering an operational design but there exists several penalization methods (see [2] that could be employed to this end. On applique dans cet article une méthode d’optimisation géométrique dans le cadre de la conception d’un cœur de réacteur SFR (Sodium-cooled Fast Reactor, i.e. réacteur à neutron rapide refroidi au sodium dans le but de minimiser une contre réaction thermique connue sous le nom d’effet de vidange sodium. Lorsqu’une augmentation de température survient, ce type de réacteur peut être sujet à une forte augmentation de réactivité, un paramètre clé dans le contrôle de la réaction en chaîne en régime quasi-statique. On a recours à l’équation de diffusion à un groupe puis on donne la généralisation du modèle d’optimisation pour l’équation de la diffusion à deux groupes d’énergie. On présente ensuite quelques résultats numériques obtenus dans le cas de l’équation à un groupe d’énergie. On note que l’application de cette méthode conduit à des designs de cœur présentant des interfaces très régulières qui sont loin d’un design de cœur faisable sur le

  4. A computer program to determine the specific power of prismatic-core reactors

    Energy Technology Data Exchange (ETDEWEB)

    Dobranich, D.

    1987-05-01

    A computer program has been developed to determine the maximum specific power for prismatic-core reactors as a function of maximum allowable fuel temperature, core pressure drop, and coolant velocity. The prismatic-core reactors consist of hexagonally shaped fuel elements grouped together to form a cylindrically shaped core. A gas coolant flows axially through circular channels within the elements, and the fuel is dispersed within the solid element material either as a composite or in the form of coated pellets. Different coolant, fuel, coating, and element materials can be selected to represent different prismatic-core concepts. The computer program allows the user to divide the core into any arbitrary number of axial levels to account for different axial power shapes. An option in the program allows the automatic determination of the core height that results in the maximum specific power. The results of parametric specific power calculations using this program are presented for various reactor concepts.

  5. Control rod reactivity worth determination of a typical MTR research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Rizwan, M.; Raza, S.S.; Khan, R. [Pakistan Institute of Engineering and Applied Sciences (PIEAS), Islamabad (Pakistan). Dept. of Nuclear Engineering

    2015-10-15

    The safe and reliable utilization of research reactor demands the possible accurate information of control rod (CR) worths. The criticality positions of the control rods changes with time due to build up fission products. It is therefore important to determine the reactivity worth of control rods. The aim of this article is to estimate the reactivity worth of controls rods in the equilibrium core of a Materials Testing Reactor (MTR). A deterministic model of the reactor core was developed and confirmed against the reference results of excess reactivity, shutdown margin and combined control rod reactivity worth using the combination of WIMS/D4 and CITATION computer codes.

  6. Heat transfer and core neutronics considerations of the heat pipe cooled thermionic reactor

    Science.gov (United States)

    Determan, W. R.; Lewis, Brian

    1991-01-01

    The authors summarize the results of detailed neutronic and thermal-hydraulic evaluations of the heat pipe cooled thermionic (HPTI) reactor design, identify its key design attributes, and quantify its performance characteristics. The HPTI core uses modular, liquid-metal core heat transfer assemblies to replace the liquid-metal heat transport loop employed by in-core thermionic reactor designs of the past. The nuclear fuel, power conversion, heat transport, and heat rejection functions are all combined into a single modular unit. The reactor/converter assembly uses UN fuel pins to obtain a critical core configuration with in-core safety rods and reflector controls added to complete the subassembly. By thermally bonding the core heat transfer assemblies during the reactor core is coupled neutronically, thermally, and electrically into a modular assembly of individual power sources with cross-tied architecture. A forward-facing heat pipe radiator assembly extends from the reactor head in the shape of a frustum of a cone on the opposite side of the power system from the payload. Important virtues of the concept are the absence of any single-point failures and the ability of the core to effectively transfer the TFE waste heat load laterally to other in-core heat transfer assemblies in the event of multiple failures in either in-core and radiator heat pipes.

  7. Heat transfer and core neutronics considerations of the heat pipe cooled thermionic reactor

    Science.gov (United States)

    Determan, W. R.; Lewis, Brian

    The authors summarize the results of detailed neutronic and thermal-hydraulic evaluations of the heat pipe cooled thermionic (HPTI) reactor design, identify its key design attributes, and quantify its performance characteristics. The HPTI core uses modular, liquid-metal core heat transfer assemblies to replace the liquid-metal heat transport loop employed by in-core thermionic reactor designs of the past. The nuclear fuel, power conversion, heat transport, and heat rejection functions are all combined into a single modular unit. The reactor/converter assembly uses UN fuel pins to obtain a critical core configuration with in-core safety rods and reflector controls added to complete the subassembly. By thermally bonding the core heat transfer assemblies during the reactor core is coupled neutronically, thermally, and electrically into a modular assembly of individual power sources with cross-tied architecture. A forward-facing heat pipe radiator assembly extends from the reactor head in the shape of a frustum of a cone on the opposite side of the power system from the payload. Important virtues of the concept are the absence of any single-point failures and the ability of the core to effectively transfer the TFE waste heat load laterally to other in-core heat transfer assemblies in the event of multiple failures in either in-core and radiator heat pipes.

  8. Numerical Simulation of Flow Field in Flow-guide Tank of China Advanced Research Reactor

    Institute of Scientific and Technical Information of China (English)

    2001-01-01

    The flow-guide tank of China advanced research reactor (CARR) is located at the top of the reactor vessel and connected with the inlet coolant pipe. It acts as a reactor inlet coolant distributor and plays an important role in reducing the flow-induced vibration of the internal components of the reactor core. Several designs of the flow-guide tank have been proposed, however, the final design option has to be made after detailed investigation of the velocity profile within the flow-guide tank for each configuration.

  9. Turbulence coefficients and stability studies for the coaxial flow or dissimiliar fluids. [gaseous core nuclear reactors

    Science.gov (United States)

    Weinstein, H.; Lavan, Z.

    1975-01-01

    Analytical investigations of fluid dynamics problems of relevance to the gaseous core nuclear reactor program are presented. The vortex type flow which appears in the nuclear light bulb concept is analyzed along with the fluid flow in the fuel inlet region for the coaxial flow gaseous core nuclear reactor concept. The development of numerical methods for the solution of the Navier-Stokes equations for appropriate geometries is extended to the case of rotating flows and almost completes the gas core program requirements in this area. The investigations demonstrate that the conceptual design of the coaxial flow reactor needs further development.

  10. Steady Thermal Field Simulation of Forced Air-cooled Column-type Air-core Reactor

    Institute of Scientific and Technical Information of China (English)

    DENG Qiu; LI Zhenbiao; YIN Xiaogen; YUAN Zhao

    2013-01-01

    Modeling the steady thermal field of the column-type air-core reactor,and further analyzing its distribution regularity,will help optimizing reactor design as well as improving its quality.The operation mechanism and inner insulation structure of a novel current limiting column-type air-core reactor is introduced in this paper.The finite element model of five encapsulation forced air-cooled column type air-core reactor is constructed using Fluent.Most importantly,this paper present a new method that,the steady thermal field of reactor working under forced air-cooled condition is simulated without arbitrarily defining the convection heat transfer coefficient for the initial condition; The result of the thermal field distribution shows that,the maximum steady temperature rise of forced air-cooled columntype air-core reactor happens approximately 5% to its top.The law of temperature distribution indicates:In the 1/3part of the reactor to its bottom,the temperature will rise rapidly to the increasing of height,yet the gradient rate is gradually decreasing; In the 5 % part of the reactor to its top,the temperature will drop rapidly to the increasing of height; In the part between,the temperature will rise slowly to the increasing of height.The conclusion draws that more thermal withstand capacity should be considered at the 5 % part of the reactor to its top to achieve optimal design solution.

  11. Intelligent uranium fission converter for neutron production on the periphery of the nuclear reactor core (MARIA reactor in Swierk - Poland)

    Energy Technology Data Exchange (ETDEWEB)

    Gryzinski, M.A.; Wielgosz, M. [National Centre for Nuclear Research, Andrzeja Soltana 7, 05-400 Otwock-Swierk (Poland)

    2015-07-01

    The multipurpose, high flux research reactor MARIA in Otwock - Swierk is an open-pool type, water and beryllium moderated and graphite reflected. There are two not occupied experimental H1 and H2 horizontal channels with complex of empty rooms beside them. Making use of these two channels is not in conflict with other research or commercial employing channels. They can work simultaneously, moreover commercial channels covers the cost of reactor working. Such conditions give beneficial possibility of creating epithermal neutron stand for researches in various field at the horizontal channel H2 of MARIA reactor (co-organization of research at H1 channel is additionally planned). At the front of experimental channels the neutron flux is strongly thermalized - neutrons with energies above 0.625 eV constitute only ∼2% of the total flux. This thermalized neutron flux will be used to achieve high flux of epithermal neutrons at the level of 2x10{sup 9} n cm{sup -2}s{sup -1} by uranium neutron converter (fast neutron production - conversion of reactor core thermal neutrons to fast neutrons - and then filtering, moderating and finally cutting of unwanted gamma radiation). The intelligent converter will be placed in the reactor pool, near the front of the H2 channel. It will replace one graphite block at the periphery of MARIA graphite reflector. The converter will consist of 20 fuel elements - low enriched uranium plates. A fuel plate will be a part which will measure 110 mm wide by 380 mm long and will consist of a thin layer of uranium sealed between two aluminium plates. These plates, once assembled, form the fuel element used in converter. The plates will be positioned vertically. There are several important requirements which should be taken into account at the converter design stage: -maximum efficiency of the converter for neutrons conversion, -cooling of the converter need to be integrated with the cooling circuit of the reactor pool and if needed equipped with

  12. Irradiation capabilities of LR-0 reactor with VVER-1000 Mock-Up core.

    Science.gov (United States)

    Košťál, Michal; Rypar, Vojtěch; Svadlenková, Marie; Cvachovec, František; Jánský, Bohumil; Milčák, Ján

    2013-12-01

    Even low power reactors, such as zero power reactors, are sufficient for semiconductor radiation hardness effect investigation. This reflects the fact that fluxes necessary for affecting semiconductor electrical resistance are much lower than fluxes necessary to affect material parameters. The paper aims to describe the irradiation possibilities of the LR-0 reactor with a special core arrangement corresponding to VVER-1000 dosimetry Mock-Up.

  13. Fault current limiter-predominantly resistive behavior of a BSCCO shielded-core reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ennis, M. G.; Tobin, T. J.; Cha, Y. S.; Hull, J. R.

    2000-06-30

    Tests were conducted to determine the electrical and magnetic characteristics of a superconductor shielded core reactor (SSCR). The results show that a closed-core SSCR is predominantly a resistive device and an open-core SSCR is a hybrid resistive/inductive device. The open-core SSCR appears to dissipate less than the closed-core SSCR. However, the impedance of the open-core SSCR is less than that of the closed-core SSCR. Magnetic and thermal diffusion are believed to be the mechanism that facilitates the penetration of the superconductor tube under fault conditions.

  14. Sensitivity Analysis of Core Neutronic Parameters in Electron Accelerator-driven Subcritical Advanced Liquid Metal Reactor

    Directory of Open Access Journals (Sweden)

    Marziye Ebrahimkhani

    2016-02-01

    Full Text Available Calculation of the core neutronic parameters is one of the key components in all nuclear reactors. In this research, the energy spectrum and spatial distribution of the neutron flux in a uranium target have been calculated. In addition, sensitivity of the core neutronic parameters in accelerator-driven subcritical advanced liquid metal reactors, such as electron beam energy (Ee and source multiplication coefficient (ks, has been investigated. A Monte Carlo code (MCNPX_2.6 has been used to calculate neutronic parameters such as effective multiplication coefficient (keff, net neutron multiplication (M, neutron yield (Yn/e, energy constant gain (G0, energy gain (G, importance of neutron source (φ∗, axial and radial distributions of neutron flux, and power peaking factor (Pmax/Pave in two axial and radial directions of the reactor core for four fuel loading patterns. According to the results, safety margin and accelerator current (Ie have been decreased in the highest case of ks, but G and φ∗ have increased by 88.9% and 21.6%, respectively. In addition, for LP1 loading pattern, with increasing Ee from 100 MeV up to 1 GeV, Yn/e and G improved by 91.09% and 10.21%, and Ie and Pacc decreased by 91.05% and 10.57%, respectively. The results indicate that placement of the Np–Pu assemblies on the periphery allows for a consistent keff because the Np–Pu assemblies experience less burn-up.

  15. Safety and core design of large liquid-metal cooled fast breeder reactors

    Science.gov (United States)

    Qvist, Staffan Alexander

    In light of the scientific evidence for changes in the climate caused by greenhouse-gas emissions from human activities, the world is in ever more desperate need of new, inexhaustible, safe and clean primary energy sources. A viable solution to this problem is the widespread adoption of nuclear breeder reactor technology. Innovative breeder reactor concepts using liquid-metal coolants such as sodium or lead will be able to utilize the waste produced by the current light water reactor fuel cycle to power the entire world for several centuries to come. Breed & burn (B&B) type fast reactor cores can unlock the energy potential of readily available fertile material such as depleted uranium without the need for chemical reprocessing. Using B&B technology, nuclear waste generation, uranium mining needs and proliferation concerns can be greatly reduced, and after a transitional period, enrichment facilities may no longer be needed. In this dissertation, new passively operating safety systems for fast reactors cores are presented. New analysis and optimization methods for B&B core design have been developed, along with a comprehensive computer code that couples neutronics, thermal-hydraulics and structural mechanics and enables a completely automated and optimized fast reactor core design process. In addition, an experiment that expands the knowledge-base of corrosion issues of lead-based coolants in nuclear reactors was designed and built. The motivation behind the work presented in this thesis is to help facilitate the widespread adoption of safe and efficient fast reactor technology.

  16. Numerical Analysis of Magnetic Force of Dry-Type Air-Core Reactor

    Institute of Scientific and Technical Information of China (English)

    LIUZhi-gang; GENGYing-san; WANGJian-hua

    2004-01-01

    This paper presents a coupled magnetic-circuit method for computing the magnetic force of air-core reactor under short-time current. The current and the magnetic flux density are computed first and then the magnetic force is obtained. Thus, the dynamic stability performance of air-core reactor can be analyzed at the design stage to reduce experimental cost and shorten the lead-time of product development.

  17. A new high performance research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Abbate, Pablo M. [INVAP S.E., Bariloche, Rio Negro (Argentina)

    2002-07-01

    A contract for the design, construction and commissioning of the Replacement Research Reactor was signed in July 2000 between Australia authorities and INVAP from Argentina. Since then the detailed design has been completed, an application for a construction license was made in May 2001 and the construction authorisation was issued on 4{sup th} April 2002. This paper explains the safety philosophy embedded into the design together with the approach taken for main elements of the design and their relation to the proposed applications of the reactor. Also information is provided on the suit of neutron beam facilities and irradiation facilities being constructed. Finally it is presented an outline of the project management organisation, project planing and schedule. (author)

  18. On use of ZPR research reactors and associated instrumentation and measurement methods for reactor physics studies

    Energy Technology Data Exchange (ETDEWEB)

    Chauvin, J.P. [CEA,DEN, DER, SPEX, Experimental Physics Service, Cadarache, F-13108 St-Paul-Lez-Durance (France); Blaise, P. [CEA, DEN, DER, SPEX Experimental Programs Laboratory, Cadarache, F-13108 St-Paul-Lez-Durance (France); Lyoussi, A. [CEA, DEN, DER, Instrumentation Sensors and Dosimetry Laboratory, Cadarache, F-13108 St-Paul-Lez-Durance (France)

    2015-07-01

    The French atomic and alternative energies -CEA- is strongly involved in research and development programs concerning the use of nuclear energy as a clean and reliable source of energy and consequently is working on the present and future generation of reactors on various topics such as ageing plant management, optimization of the plutonium stockpile, waste management and innovative systems exploration. Core physics studies are an essential part of this comprehensive R and D effort. In particular, the Zero Power Reactor (ZPR) of CEA: EOLE, MINERVE and MASURCA play an important role in the validation of neutron (as well photon) physics calculation tools (codes and nuclear data). The experimental programs defined in the CEA's ZPR facilities aim at improving the calculation routes by reducing the uncertainties of the experimental databases. They also provide accurate data on innovative systems in terms of new materials (moderating and decoupling materials) and new concepts (ADS, ABWR, new MTR (e.g. JHR), GENIV) involving new fuels, absorbers and coolant materials. Conducting such interesting experimental R and D programs is based on determining and measuring main parameters of phenomena of interest to qualify calculation tools and nuclear data 'libraries'. Determining these parameters relies on the use of numerous and different experimental techniques using specific and appropriate instrumentation and detection tools. Main ZPR experimental programs at CEA, their objectives and challenges will be presented and discussed. Future development and perspectives regarding ZPR reactors and associated programs will be also presented. (authors)

  19. The AFR. An approved network of research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hampel, Gabriele [Mainz Univ. (Germany). Arbeitsgemeinschaft fuer Betriebs- und Sicherheitsfragen an Forschungsreaktoren (AFR)

    2012-10-15

    AFR (Arbeitsgemeinschaft fuer Betriebs- und Sicherheitsfragen an Forschungsreaktoren) is the German acronym for 'Association for Research Reactor Operation and Safety Issues' which was founded in 1959. Reactor managers of European research reactors mainly from the German linguistic area meet regularly for their mutual benefit to exchange experience and knowledge in all areas of operating, managing and utilization of research reactors. In the last 2 years joint meetings were held together with the French association of research reactors CER (Club d'Exploitants des Reacteurs). In this contribution the AFR, its members, work and aims as well as the French partner CER are presented. (orig.)

  20. Calculation of kinetic parameters of Caliban metallic core experimental reactor from stochastic neutron measurements

    Energy Technology Data Exchange (ETDEWEB)

    Casoli, P.; Authier, N.; Baud, J. [Commissariat a l' energie Atomique, Centre de Valduc, 21120 Is-sur-Tille (France)

    2009-07-01

    Several experimental devices are operated by the Criticality and Neutron Science Research Department of the CEA Valduc Laboratory. One of these is the metallic core reactor Caliban. The knowledge of the fundamental kinetic parameters of the reactor is very useful, indeed necessary, to the operator. The purpose of this study was to develop and perform experiments allowing to determinate some of these parameters. The prompt neutron decay constant and particularly its value at criticality can be measured with reactor noise techniques such as the interval-distribution, the Feynman variance-to-mean, and the Rossi-{alpha} methods. By introducing the Nelson number, the effective delayed neutron fraction and the average neutron lifetime can also be calculated with the Rossi-{alpha} method. Subcritical, critical, and even supercritical experiments were performed. With the Rossi-{alpha} technique, it was found that the prompt neutron decay constant at criticality was (6.02*10{sup 5} {+-} 9%). Experiments also brought out the limitations of the used experimental parameters. (authors)

  1. Research reactor de-fueling and fuel shipment

    Energy Technology Data Exchange (ETDEWEB)

    Ice, R.D.; Jawdeh, E.; Strydom, J.

    1998-08-01

    Planning for the Georgia Institute of Technology Research Reactor operations during the 1996 Summer Olympic Games began in early 1995. Before any details could be outlined, several preliminary administrative decisions had to be agreed upon by state, city, and university officials. The two major administrative decisions involving the reactor were (1) the security level and requirements and (2) the fuel status of the reactor. The Georgia Tech Research Reactor (GTRR) was a heavy-water moderated and cooled reactor, fueled with high-enriched uranium. The reactor was first licensed in 1964 with an engineered lifetime of thirty years. The reactor was intended for use in research applications and as a teaching facility for nuclear engineering students and reactor operators. Approximately one year prior to the olympics, the Georgia Tech administration decided that the GTRR fuel would be removed. In addition, a heightened, beyond regulatory requirements, security system was to be implemented. This report describes the scheduling, operations, and procedures.

  2. The reactor core TRIGA Mark-III with fuels type 30/20; El nucleo del reactor TRIGA Mark-III con combustible tipo 30/20

    Energy Technology Data Exchange (ETDEWEB)

    Aguilar H, F., E-mail: fortunato.aguilar@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2012-10-15

    This work describes the calculation series carried out with the program MCNP5 in order to define the configuration of the reactor core with fuels 30/20 (fuels with 30% of uranium content in the Or-Zr-H mixture and a nominal enrichment of 20%). To select the configuration of the reactor core more appropriate to the necessities and future uses of the reactor, the following criterions were taken into account: a) the excess in the reactor reactivity, b) the switch out margin and c) to have new irradiation facilities inside the reactor core. Taking into account these criterions is proceeded to know the characteristics of the components that form the reactor core (dimensions, geometry, materials, densities and positions), was elaborated a base model of the reactor core, for the MCNP5 code, with a configuration composed by 85 fuel elements, 4 control bars and the corresponding structural elements. The high reactivity excess obtained with this model, gave the rule to realize other models of the reactor core in which the reactivity excess and the switch out margin were approximate to the values established in the technical specifications of the reactor operation. Several models were realized until finding the satisfactory model; this is composite for 74 fuels, 4 control bars and 6 additional experimental positions inside the reactor core. (Author)

  3. Reactor Safety Research: Semiannual report, January-June 1986: Reactor Safety Research Program

    Energy Technology Data Exchange (ETDEWEB)

    1987-05-01

    Sandia National Laboratories is conducting, under USNRC sponsorship, phenomenological research related to the safety of commercial nuclear power reactors. The research includes experiments to simulate the phenomenology of accident conditions and the development of analytical models, verified by experiment, which can be used to predict reactor and safety systems performance behavior under abnormal conditions. The objective of this work is to provide NRC requisite data bases and analytical methods to (1) identify and define safety issues, (2) understand the progression of risk-significant accident sequences, and (3) conduct safety assessments. The collective NRC-sponsored effort at Sandia National Laboratories is directed at enhancing the technology base supporting licensing decisions.

  4. Determination of the Sensitivity of the Antineutrino Probe for Reactor Core Monitoring

    Science.gov (United States)

    Cormon, S.; Fallot, M.; Bui, V.-M.; Cucoanes, A.; Estienne, M.; Lenoir, M.; Onillon, A.; Shiba, T.; Yermia, F.; Zakari-Issoufou, A.-A.

    2014-06-01

    This paper presents a feasibility study of the use of the detection of reactor-antineutrinos (νbare) for non proliferation purpose. To proceed, we have started to study different reactor designs with our simulation tools. We use a package called MCNP Utility for Reactor Evolution (MURE), initially developed by CNRS/IN2P3 labs to study Generation IV reactors. The MURE package has been coupled to fission product beta decay nuclear databases for studying reactor antineutrino emission. This method is the only one able to predict the antineutrino emission from future reactor cores, which don't use the thermal fission of 235U, 239Pu and 241Pu. It is also the only way to include off-equilibrium effects, due to neutron captures and time evolution of the fission product concentrations during a reactor cycle. We will present here the first predictions of antineutrino energy spectra from innovative reactor designs (Generation IV reactors). We will then discuss a summary of our results of non-proliferation scenarios involving the latter reactor designs, taking into account reactor physics constraints.

  5. Determination of the Sensitivity of the Antineutrino Probe for Reactor Core Monitoring

    Energy Technology Data Exchange (ETDEWEB)

    Cormon, S.; Fallot, M., E-mail: fallot@subatech.in2p3.fr; Bui, V.-M.; Cucoanes, A.; Estienne, M.; Lenoir, M.; Onillon, A.; Shiba, T.; Yermia, F.; Zakari-Issoufou, A.-A.

    2014-06-15

    This paper presents a feasibility study of the use of the detection of reactor-antineutrinos (ν{sup ¯}{sub e}) for non proliferation purpose. To proceed, we have started to study different reactor designs with our simulation tools. We use a package called MCNP Utility for Reactor Evolution (MURE), initially developed by CNRS/IN2P3 labs to study Generation IV reactors. The MURE package has been coupled to fission product beta decay nuclear databases for studying reactor antineutrino emission. This method is the only one able to predict the antineutrino emission from future reactor cores, which don't use the thermal fission of {sup 235}U, {sup 239}Pu and {sup 241}Pu. It is also the only way to include off-equilibrium effects, due to neutron captures and time evolution of the fission product concentrations during a reactor cycle. We will present here the first predictions of antineutrino energy spectra from innovative reactor designs (Generation IV reactors). We will then discuss a summary of our results of non-proliferation scenarios involving the latter reactor designs, taking into account reactor physics constraints.

  6. Termination of light-water reactor core-melt accidents with a chemical core catcher: the core-melt source reduction system (COMSORS)

    Energy Technology Data Exchange (ETDEWEB)

    Forsberg, C.W.; Parker, G.W.; Rudolph, J.C.; Osborne-Lee, I.W. [Oak Ridge National Lab., TN (United States); Kenton, M.A. [Dames and Moore, Westmont, IL (United States)

    1996-09-01

    The Core-Melt Source Reduction System (COMSORS) is a new approach to terminate light-water reactor core melt accidents and ensure containment integrity. A special dissolution glass is placed under the reactor vessel. If core debris is released onto the glass, the glass melts and the debris dissolves into the molten glass, thus creating a homogeneous molten glass. The molten glass, with dissolved core debris, spreads into a wide pool, distributing the heat for removal by radiation to the reactor cavity above or by transfer to water on top of the molten glass. Expected equilibrium glass temperatures are approximately 600 degrees C. The creation of a low-temperature, homogeneous molten glass with known geometry permits cooling of the glass without threatening containment integrity. This report describes the technology, initial experiments to measure key glass properties, and modeling of COMSORS operations.

  7. 100-KE REACTOR CORE REMOVAL PROJECT ALTERNATIVE ANALYSIS WORKSHOP REPORT

    Energy Technology Data Exchange (ETDEWEB)

    HARRINGTON RA

    2010-01-15

    On December 15-16, 2009, a 100-KE Reactor Core Removal Project Alternative Analysis Workshop was conducted at the Washington State University Consolidated Information Center, Room 214. Colburn Kennedy, Project Director, CH2M HILL Plateau Remediation Company (CHPRC) requested the workshop and Richard Harrington provided facilitation. The purpose of the session was to select the preferred Bio Shield Alternative, for integration with the Thermal Shield and Core Removal and develop the path forward to proceed with project delivery. Prior to this workshop, the S.A. Robotics (SAR) Obstruction Removal Alternatives Analysis (565-DLV-062) report was issued, for use prior to and throughout the session, to all the team members. The multidisciplinary team consisted ofrepresentatives from 100-KE Project Management, Engineering, Radcon, Nuclear Safety, Fire Protection, Crane/Rigging, SAR Project Engineering, the Department of Energy Richland Field Office, Environmental Protection Agency, Washington State Department of Ecology, Defense Nuclear Facility Safety Board, and Deactivation and Decommission subject matter experts from corporate CH2M HILL and Lucas. Appendix D contains the workshop agenda, guidelines and expectations, opening remarks, and attendance roster going into followed throughout the workshop. The team was successful in selecting the preferred alternative and developing an eight-point path forward action plan to proceed with conceptual design. Conventional Demolition was selected as the preferred alternative over two other alternatives: Diamond Wire with Options, and Harmonic Delamination with Conventional Demolition. The teams preferred alternative aligned with the SAR Obstruction Removal Alternative Analysis report conclusion. However, the team identified several Path Forward actions, in Appendix A, which upon completion will solidify and potentially enhance the Conventional Demolition alternative with multiple options and approaches to achieve project delivery

  8. Architecture dependent availability analysis of RPS for Research Reactor Applications

    Energy Technology Data Exchange (ETDEWEB)

    Rahman, Khalilur; Heo, Gyunyoung [Kyung Hee Univ., Yongin (Korea, Republic of); Son, Hanseong [Joongbu Univ., Geumsan (Korea, Republic of); Kim, Youngki; Park, Jaekwan [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-05-15

    The research reactors are categorized into two broad categories, Low power research reactors and medium to high power research reactors. According to IAEA TECDOC-1234, Research reactors with 0.250- 2.0 MW power rating or 2.5-10 Χ 10{sup 11} η/cm{sup 2}. s flux are termed low power reactor whereas research reactors ranging from 2-10 MW power rating or 0.1-10 Χ 10{sup 13} η/cm{sup 2}. s are considered as Medium to High power research reactors. Some other standards (IAEA NP-T-5.1) define multipurpose research reactor ranging from power few hundred KW to 10 MW as low power research reactor. The aim of this research, in this article, was to identify a configuration of architecture which gives highest availability with maintaining low cost of manufacturing. In this regard, two configurations of a single channel of RPS are formulated in the current article and their fault trees were developed using AIMS PSA software to get the unavailability. This is a starting point of attempt towards the standardization of I and C architecture for low and medium power research reactors.

  9. Fresh-Core Reload of the Neutron Radiography (NRAD) Reactor with Uranium(20)-Erbium-Zirconium-Hydride Fuel

    Energy Technology Data Exchange (ETDEWEB)

    John D. Bess; Thomas L. Maddock; Margaret A. Marshall; Leland M. Montierth

    2013-03-01

    The neutron radiography (NRAD) reactor is a 250 kW TRIGA® (Training, Research, Isotopes, General Atomics) Mark II , tank-type research reactor currently located in the basement, below the main hot cell, of the Hot Fuel Examination Facility (HFEF) at the Idaho National Laboratory (INL). It is equipped with two beam tubes with separate radiography stations for the performance of neutron radiography irradiation on small test components. The initial critical configuration developed during the fuel loading process, which contains only 56 fuel elements, has been evaluated as an acceptable benchmark experiment. The 60-fuel-element operational core configuration of the NRAD LEU TRIGA reactor has also been evaluated as an acceptable benchmark experiment. Calculated eigenvalues differ significantly (~±1%) from the benchmark eigenvalue and have demonstrated sensitivity to the thermal scattering treatment of hydrogen in the U-Er-Zr-H fuel.

  10. Modernization of the VVR-TS reactor core for the increasing of the radionuclides production

    Science.gov (United States)

    Fomin, R. V.; Kolesov, V. V.; Kochnov, O. Yu

    2017-01-01

    We have studied the possibility of increasing the 99Mo and 131I production through modernization of the VVR-TS reactor core. It has been found in particular that the introducing of beryllium reflector on the core periphery can significantly increase the starting reactivity margin. This, in turn, give us possibility to create an additional channel for increasing their production.

  11. Engineering review of the core support structure of the Gas Cooled Fast Breeder Reactor

    Energy Technology Data Exchange (ETDEWEB)

    None

    1978-09-01

    The review of the core support structure of the gas cooled fast breeder reactor (GCFR) covered such areas as the design criteria, the design and analysis of the concepts, the development plan, and the projected manufacturing costs. Recommendations are provided to establish a basis for future work on the GCFR core support structure.

  12. Neutronic study of SCR core for under-sea scientific research vessel

    Energy Technology Data Exchange (ETDEWEB)

    Odano, Naoteru; Ishida, Toshihisa [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Wada, Koji; Imai, Hiroshi [Hitachi Engineering Co., Ltd., Hitachi, Ibaraki (Japan)

    2001-07-01

    A very small reactor, SCR (Submersible Compact Reactor), whose thermal output is 1250 kW, is an integral-pressurized type reactor to be used as a power source for a scientific research vessel in medium region to approximately 300-m in depth of the Arctic Ocean. Neutronic study has been carried out for design of the SCR core of which could achieve continuous long-term operation without refueling for 10 years considering 50% of load factor of the core. In the present study, arrangement of fuel rods, {sup 235}U enrichment of UO{sub 2} fuel rods and reflector materials were surveyed. The {sup 235}U enrichment has been determined to be 9.5 wt% to satisfy design criteria. Design of the reflector as well as that of the fuel enrichment is one of important factors for determination of the core life because leakage of neutrons from a very small reactor is large. In the present study Be metal was adopted as a reflector material. The square and triangle arrangements of fuel rods were studied and the later one was adopted for the present design to make possible to reduce size of the core. Reactor physics parameters including reactivity coefficients and power distributions were evaluated for the determined core specifications. Reactor physics parameters related to core safety were also confirmed and the evaluated parameters indicated that the determined core specifications in this study satisfied design conditions. (author)

  13. Development and application of neutron transport methods and uncertainty analyses for reactor core calculations. Technical report; Entwicklung und Einsatz von Neutronentransportmethoden und Unsicherheitsanalysen fuer Reaktorkernberechnungen. Technischer Bericht

    Energy Technology Data Exchange (ETDEWEB)

    Zwermann, W.; Aures, A.; Bernnat, W.; and others

    2013-06-15

    This report documents the status of the research and development goals reached within the reactor safety research project RS1503 ''Development and Application of Neutron Transport Methods and Uncertainty Analyses for Reactor Core Calculations'' as of the 1{sup st} quarter of 2013. The superordinate goal of the project is the development, validation, and application of neutron transport methods and uncertainty analyses for reactor core calculations. These calculation methods will mainly be applied to problems related to the core behaviour of light water reactors and innovative reactor concepts. The contributions of this project towards achieving this goal are the further development, validation, and application of deterministic and stochastic calculation programmes and of methods for uncertainty and sensitivity analyses, as well as the assessment of artificial neutral networks, for providing a complete nuclear calculation chain. This comprises processing nuclear basis data, creating multi-group data for diffusion and transport codes, obtaining reference solutions for stationary states with Monte Carlo codes, performing coupled 3D full core analyses in diffusion approximation and with other deterministic and also Monte Carlo transport codes, and implementing uncertainty and sensitivity analyses with the aim of propagating uncertainties through the whole calculation chain from fuel assembly, spectral and depletion calculations to coupled transient analyses. This calculation chain shall be applicable to light water reactors and also to innovative reactor concepts, and therefore has to be extensively validated with the help of benchmarks and critical experiments.

  14. Generation of XS library for the reflector of VVER reactor core using Monte Carlo code Serpent

    Science.gov (United States)

    Usheva, K. I.; Kuten, S. A.; Khruschinsky, A. A.; Babichev, L. F.

    2017-01-01

    A physical model of the radial and axial reflector of VVER-1200-like reactor core has been developed. Five types of radial reflector with different material composition exist for the VVER reactor core and 1D and 2D models were developed for all of them. Axial top and bottom reflectors are described by the 1D model. A two-group XS library for diffusion code DYN3D has been generated for all types of reflectors by using Serpent 2 Monte Carlo code. Power distribution in the reactor core calculated in DYN3D is flattened in the core central region to more extent in the 2D model of the radial reflector than in its 1D model.

  15. Research and development on next generation reactor (phase I)

    Energy Technology Data Exchange (ETDEWEB)

    Park, Jong Kyoon; Chang, Moon Heuy; Hwang, Yung Dong [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)] [and others

    1994-10-01

    The objective of the study is to improve the volume of nuclear power plant which adopts passive safety system concept. The passive safety system reactor is characterized by excellent safety and reliability. But the volume of NSSS (Nuclear Steam Supply System) of the passive safety system reactor is so small that it should be upgraded for commercial operation. For volume upgrade, detailed analyses are performed as follows; core design, hydraulics, design and mechnical structures, and safety analysis. In addition to above analysis, some investigations must be supplied as follows: power density vs. DNB margin decrease, outlet temperature vs. EPRI-URD, additional tests for upgraded reactor, dynamic analysis of mechanical vibration according to expanded reactor vessel and expanded in-core structures, and Merit loss of passive safety system reactor according to design margin decrease. (Author).

  16. Very High Temperature Reactor (VHTR) Deep Burn Core and Fuel Analysis -- Complete Design Selection for the Pebble Bed Reactor

    Energy Technology Data Exchange (ETDEWEB)

    B. Boer; A. M. Ougouag

    2010-09-01

    The Deep-Burn (DB) concept focuses on the destruction of transuranic nuclides from used light water reactor fuel. These transuranic nuclides are incorporated into TRISO coated fuel particles and used in gas-cooled reactors with the aim of a fractional fuel burnup of 60 to 70% in fissions per initial metal atom (FIMA). This high performance is expected through the use of multiple recirculation passes of the fuel in pebble form without any physical or chemical changes between passes. In particular, the concept does not call for reprocessing of the fuel between passes. In principle, the DB pebble bed concept employs the same reactor designs as the presently envisioned low-enriched uranium core designs, such as the 400 MWth Pebble Bed Modular Reactor (PBMR-400). Although it has been shown in the previous Fiscal Year (2009) that a PuO2 fueled pebble bed reactor concept is viable, achieving a high fuel burnup, while remaining within safety-imposed prescribed operational limits for fuel temperature, power peaking and temperature reactivity feedback coefficients for the entire temperature range, is challenging. The presence of the isotopes 239-Pu, 240-Pu and 241-Pu that have resonances in the thermal energy range significantly modifies the neutron thermal energy spectrum as compared to a ”standard,” UO2-fueled core. Therefore, the DB pebble bed core exhibits a relatively hard neutron energy spectrum. However, regions within the pebble bed that are near the graphite reflectors experience a locally softer spectrum. This can lead to power and temperature peaking in these regions. Furthermore, a shift of the thermal energy spectrum with increasing temperature can lead to increased absorption in the resonances of the fissile Pu isotopes. This can lead to a positive temperature reactivity coefficient for the graphite moderator under certain operating conditions. The effort of this task in FY 2010 has focused on the optimization of the core to maximize the pebble discharge

  17. Preparation macroconstants to simulate the core of VVER-1000 reactor

    Science.gov (United States)

    Seleznev, V. Y.

    2017-01-01

    Dynamic model is used in simulators of VVER-1000 reactor for training of operating staff and students. As a code for the simulation of neutron-physical characteristics is used DYNCO code that allows you to perform calculations of stationary, transient and emergency processes in real time to a different geometry of the reactor lattices [1]. To perform calculations using this code, you need to prepare macroconstants for each FA. One way of getting macroconstants is to use the WIMS code, which is based on the use of its own 69-group macroconstants library. This paper presents the results of calculations of FA obtained by the WIMS code for VVER-1000 reactor with different parameters of fuel and coolant, as well as the method of selection of energy groups for further calculation macroconstants.

  18. A New In-core Production Method of Co-60 in CANDU Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Lyu, Jinqi; Kim, Woosong; Kim, Yonghee [KAIST, Daejeon (Korea, Republic of); Park, Younwon [BEES Inc, Daejeon (Korea, Republic of)

    2016-05-15

    This study introduces an innovative method for Co-60 production in the CANDU6 core. In this new scheme, the central fuel element is replaced by a Co-59 target and Co-60 is obtained after the fuel bundle is discharged. It has been shown that the new method can produce significantly higher amount of Co-60 than the conventional Co production method in CANDU6 reactors without compromising the fuel burnup by removing some (<50%) of the adjuster rods in the whole core. The coolant void reactivity is noticeably reduced when a Co-59 target is loaded into the central pin of the fuel bundle. Meanwhile, the peak power in a fuel bundle is just a little higher due to the central Co-59 target than in conventional CANDU6 fuel design. The basic technology for Co-60 producing was developed by MDS Nordion and Atomic Energy of Canada Limited (AECL) in 1946 and the same technology was adapted and applied in CANDU6 power reactors. The standard CANDU6 reactor has 21 adjuster rods which are fully inserted into the core during normal operation. The stainless steel adjuster rods are replaced with neutronically-equivalent Co-59 adjusters to produce Co-60. Nowadays, the roles of the adjuster rods are rather vague since nuclear reactors cannot be quickly restarted after a sudden reactor trip due to more stringent regulations. In some Canadian CANDU6 reactors, some or all the adjuster rods are removed from the core to maximize the uranium utilization.

  19. Jordan's First Research Reactor Project: Driving Forces, Present Status and the Way Ahead

    Energy Technology Data Exchange (ETDEWEB)

    Xoubi, Ned, E-mail: Ned@Xoubi.co [Jordan Atomic Energy Commission (JAEC), P.O.Box 70, Shafa Badran, 11934 Amman (Jordan)

    2011-07-01

    In a gigantic step towards establishing Jordan's nuclear power program, Jordan Atomic Energy Commission (JAEC) is building the first nuclear research and test reactor in the Kingdom. The new reactor will serve as the focal point for Jordan Center for Nuclear Research (JCNR), a comprehensive state of the art nuclear center not only for Jordan but for the whole region, the center will include in addition to the reactor a radioisotopes production plant, a nuclear fuel fabrication plant, a cold neutron source (CNS), a radioactive waste treatment facility, and education and training center. The JRTR reactor is the only research reactor new build worldwide in 2010, it is a 5 MW light water open pool multipurpose reactor, The reactor core is composed of 18 fuel assemblies, MTR plate type, with 19.75% enriched uranium silicide (U{sub 3}Si{sub 2}) in an aluminum matrix. It is reflected on all sides by beryllium and graphite blocks. Reactor power is upgradable to 10 MW with a maximum thermal flux of 1.45x10{sup 14} cm{sup -2}s{sup -1}. The reactor reactivity is controlled by four Hafnium Control Absorber Rods (CAR). Jordan Center for Nuclear Research is located in Ramtha city, it is owned by Jordan Atomic Energy Commission (JAEC), and is contracted to Korea Atomic Energy Research Institute (KAERI) and Daewoo E and C. The JCNR project is a 56 months EPC fixed price contract for the design engineering, construction, and commissioning the JCNR reactor, and other nuclear facilities. The project presents many challenges for both the owner and the contractor, being the first nuclear reactor for Jordan, and the first nuclear export for Korea. The driving forces, present status and the way ahead will be presented in this paper. (author)

  20. Development of Monju easy-to-introduce system for total evaluation of reactor core

    Energy Technology Data Exchange (ETDEWEB)

    Kitano, Akihiro; Teruyama, Hidehiko; Nishi, Hiroshi [Japan Nuclear Cycle Development Inst., Tsuruga Head Office, Tsuruga, Fukui (Japan); Yamaoka, Mitsuaki; Moriki, Yasuyuki [Toshiba Corp., Tokyo (Japan); Nakagawa, Masatoshi [AITEL Corp., Tokyo (Japan)

    2002-06-01

    An interactive core analysis tool for use on a PC MEISTER (Monju Easy-to-Introduce System for Total Evaluation of Reactor Core) has been developed, enabling evaluation of the overall core characteristics, including core physics, thermal hydraulics and structural integrity, in an easy but accurate manner. Data processing, such as the preparation of input data, data transfer from one code to another or display of output data, can be achieved in a systematic manner by graphical user interfaces. MEISTER has been verified by the analysis of core criticality, control rod worth and the maximum linear heat rate measured in the Monju start-up tests. (author)

  1. Proposed design for the PGAA facility at the TRIGA IPR-R1 research reactor

    OpenAIRE

    Guerra, Bruno T.; Jacimovic, Radojko; Menezes, Maria Angela BC; Leal,Alexandre S.

    2013-01-01

    Background This work presents an initial proposed design of a Prompt Gamma Activation Analysis (PGAA) facility to be installed at the TRIGA IPR-R1, a 60 years old research reactor of the Centre of Development of Nuclear Technology (CDTN) in Brazil. The basic characteristics of the facility and the results of the neutron flux are presented and discussed. Findings The proposed design is based on a quasi vertical tube as a neutron guide from the reactor core, inside the reactor pool, 6 m below t...

  2. Fuel-coolant interaction (FCI) phenomena in reactor safety. Current understanding and future research needs

    Energy Technology Data Exchange (ETDEWEB)

    Speis, T.P. [Maryland Univ., College Park, MD (United States); Basu, S.

    1998-01-01

    This paper gives an account of the current understanding of fuel-coolant interaction (FCI) phenomena in the context of reactor safety. With increased emphasis on accident management and with emerging in-vessel core melt retention strategies for advanced light water reactor (ALWR) designs, recent interest in FCI has broadened to include an evaluation of potential threats to the integrity of reactor vessel lower head and ex-vessel structural support, as well as the role of FCI in debris quenching and coolability. The current understanding of FCI with regard to these issues is discussed, and future research needs to address the issues from a risk perspective are identified. (author)

  3. 小型自然循环钠冷堆堆芯初步设计研究%Preliminary Design Research on Core of Small Natural-circulation Sodium-cooled Reactor

    Institute of Scientific and Technical Information of China (English)

    陆道纲; 张勋

    2014-01-01

    A small natural-circulation sodium-cooled reactor-AM TEC system was pres-ented as a distributed generation system . According to the criticality calculation and thermal-hydraulic analysis of the core when the total fuel mass was constant ,the influ-ence of pellet radius ,fuel pin length and the number of rings of fuel pins upon the effec-tive multiplication factor kef , pressure drop across the core and heat transfer was analyzed .Additionally ,the B4 C absorber thickness and beginning-of-life excess reactivi-ty under different numbers of rings of the fuel pins were studied when the assumed addi-tional shutdown margin was different .The results show that decreasing the number of rings and active fuel length would increase kef and decrease pressure drop across the core w hen the core equivalent diameter and total cross-sectional area of the coolant channel are unchanged .In order to increase additional shutdow n margin ,the absorber thickness should be increased while the beginning-of-life excess reactivity and power plant econo-my would decrease .%提出了一种适用于分布式发电系统的小型自然循环钠冷堆-AMTEC系统。通过对堆芯的临界计算和热工水力分析,研究了堆芯燃料装载量不变情况下,芯块半径、燃料棒长度和圈数对堆芯有效增殖因数 kef 、堆芯压降和传热的影响。同时分析了不同额外停堆裕量下,B4 C吸收层厚度和堆芯初始剩余反应性随燃料棒圈数的变化关系。计算结果表明:保持堆芯当量直径和冷却剂通道总截面积不变的情况下,减少燃料棒圈数和活性区长度不仅可增加 kef ,且能降低堆芯压降;为提高额外停堆裕量需增加吸收层厚度,但降低了堆芯初始剩余反应性,不利于电厂的经济性。

  4. Design of a New Research Reactor: Preliminary Conceptual Design (3rd Year)

    Energy Technology Data Exchange (ETDEWEB)

    Park, Cheol; Lee, B. C.; Chae, H. T. and others

    2006-01-15

    A research reactor design is a kind of integral engineering project and a process to obtain a concrete shape through several years of concept development, conceptual design, basic design and detail design. So it requires close cooperation in various areas as well as lots of manpower and cost. The overall process at each stage may be said to be similar except for some stage-specific works. In 2005 as last year of a concept development stage, investigations on the various concepts of the fuel, reactor structure and systems which can meet the requirements established. The requirements for the process systems and I and C systems have also been embodied. The major tasks planned at the early of 2005 have been performed for each area of reactor design as follows: Establishment of the fuel and reactor core concept, and the core analysis, Preliminary thermal-hydraulic and safety analyses for the conceptual cores, Establishment and improvement of analysis system, Concept developments of the reactor structures and major systems, Test and test plan to verify the developed concepts, International cooperation to establish the foundations for exporting a research reactor.

  5. Advanced reactor material research requirements; Necesidades de investigacion en materiales para reactores avanzados

    Energy Technology Data Exchange (ETDEWEB)

    Greene, C. A.; Muscara, J.; Srinivasan, M.

    2003-07-01

    The metal and graphite components used in high temperature gas-cooled reactors (HTGR) may suffer physical-chemical alterations, irradiation damage and mechanical alterations. Their failure may call the security of these reactors into question by affecting the integrity of the pressure control system, core geometry or its cooling, among other aspects. This article analyses the work currently being done in the matter by the US Nuclear Regulatory Commission. (Author)

  6. The current status of nuclear research reactor in Thailand

    Energy Technology Data Exchange (ETDEWEB)

    Sittichai, C.; Kanyukt, R.; Pongpat, P. [Office of Atomic Energy for Peace, Bangkok (Thailand)

    1998-10-01

    Since 1962, the Thai Research Reactor has been serving for various kinds of activities i.e. the production of radioisotopes for medical uses and research and development on nuclear science and technology, for more than three decades. The existing reactor site should be abandoned and relocated to the new suitable site, according to Thai cabinet`s resolution on the 27 December 1989. The decommissioning project for the present reactor as well as the establishment of new nuclear research center were planned. This paper discussed the OAEP concept for the decommissioning programme and the general description of the new research reactor and some related information were also reported. (author)

  7. Analysis on Electromagnetic Characteristics of Research Reactor Control Rod Drive Mechanism for Thrust Force Improvement

    Energy Technology Data Exchange (ETDEWEB)

    Huh, Hyung; Choi, Myoung Hwan; Yu, Je Yong; Cho, Yeong Garp; Kim, Jong In [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2010-10-15

    The control rod drive mechanism (CRDM) is the part of reactor regulating system (RRS), which is located in the reactor pool top or the room below the reactor pool. The function of the CRDM is to insert, withdraw or maintain neutron absorbing material (control rod) at any required position within the reactor core, in order to the reactivity of the core. There are so many kinds of CRDM, such as magnetic-jack type, hydraulic type, rack and pinion type, chain type and linear or rotary step motor and so on. As a part of a new project, we are investigating the movable coil electromagnetic drive mechanism (MCEDM) which is new scheme for the reactor control rod adopted by China Advanced Research Reactor (CARR). To have a better knowledge of the electromagnetic and magnetic characteristics, numerical models of MCEDM are proposed. Especially in order to achieve improved thrust force, numerical magnetic field calculations for various kinds of magnetic and electromagnetic configuration have been performed. As a result, we present the improved design of MCEDM for research reactor

  8. Meteodiffusive Characterization of Algiers' Nuclear Research Reactor

    Directory of Open Access Journals (Sweden)

    Mourad Messaci

    2007-01-01

    Full Text Available In the framework of the environmental impact studies of the nuclear research reactor of Algiers, we will present the work related to the atmospheric dispersion of releases due to the installation in normal operation, which dealt with the assessment of spatial distribution of yearly average values of atmospheric dilution factor. The aim of this work is a characterization of the site in terms of diffusivity, which is basic for the radiological impact evaluation of the reactor. The meteorological statistics result from the National Office of Meteorology and concern 15 years of hourly records. According to the nature and features of these data, a Gaussian-type model with wind direction sectors was used. Values of wind speed at release height were estimated from measurement values at 10 m from ground. For the assessment of vertical dispersion coefficient, we used Briggs' formulas related to a sampling time of one hour. Areas of maximum impact were delimited and points of highest concentration within these zones were identified.

  9. The research reactors their contribution to the reactors physics; Les reacteurs de recherche leur apport sur la physique des reacteurs

    Energy Technology Data Exchange (ETDEWEB)

    Barral, J.C. [Electricite de France (EDF), 75 - Paris (France); Zaetta, A. [CEA/Cadarache, Direction des Reacteurs Nucleaires, DRN, 13 - Saint-Paul-lez-Durance (France); Johner, J. [CEA/Cadarache, Dept. de Recherches sur la Fusion Controlee (DRFC), 13 - Saint Paul lez Durance (France); Mathoniere, G. [CEA/Saclay, DEN, 91 - Gif sur Yvette (France)] [and others

    2000-07-01

    The 19 october 2000, the french society of nuclear energy organized a day on the research reactors. This associated report of the technical session, reactors physics, is presented in two parts. The first part deals with the annual meeting and groups general papers on the pressurized water reactors, the fast neutrons reactors and the fusion reactors industry. The second part presents more technical papers about the research programs, critical models, irradiation reactors (OSIRIS and Jules Horowitz) and computing tools. (A.L.B.)

  10. Fuel burnup analysis for Thai research reactor by using MCNPX computer code

    Science.gov (United States)

    Sangkaew, S.; Angwongtrakool, T.; Srimok, B.

    2017-06-01

    This paper presents the fuel burnup analysis of the Thai research reactor (TRR-1/M1), TRIGA Mark-III, operated by Thailand Institute of Nuclear Technology (TINT) in Bangkok, Thailand. The modelling software used in this analysis is MCNPX (MCNP eXtended) version 2.6.0, a Fortran90 Monte Carlo radiation transport computer code. The analysis results will cover the core excess reactivity, neutron fluxes at the irradiation positions and neutron detector tubes, power distribution, fuel burnup, and fission products based on fuel cycle of first reactor core arrangement.

  11. Sodium Based Heat Pipe Modules for Space Reactor Concepts: Stainless Steel SAFE-100 Core

    Science.gov (United States)

    Martin, James J.; Reid, Robert S.

    2004-01-01

    A heat pipe cooled reactor is one of several candidate reactor cores being considered for advanced space power and propulsion systems to support future space exploration applications. Long life heat pipe modules, with designs verified through a combination of theoretical analysis and experimental lifetime evaluations, would be necessary to establish the viability of any of these candidates, including the heat pipe reactor option. A hardware-based program was initiated to establish the infrastructure necessary to build heat pipe modules. This effort, initiated by Los Alamos National Laboratory and referred to as the Safe Affordable Fission Engine (SAFE) project, set out to fabricate and perform non-nuclear testing on a modular heat pipe reactor prototype that can provide 100 kilowatt from the core to an energy conversion system at 700 C. Prototypic heat pipe hardware was designed, fabricated, filled, closed-out and acceptance tested.

  12. Annular core for Modular High-Temperature Gas-Cooled Reactor (MHTGR)

    Energy Technology Data Exchange (ETDEWEB)

    Turner, R.F.; Baxter, A.M.; Stansfield, O.M.; Vollman, R.E.

    1987-08-01

    The active core of the 350 MW(t) MHTGR is annular in configuration, shaped to provide a large external surface-to-volume ratio for the transport of heat radially to the reactor vessel in case of a loss of coolant flow. For a given fuel temperature limit, the annular core provides approximately 40% greater power output over a typical cylindrical configuration. The reactor core is made up of columns of hexagonal blocks, each 793-mm high and 360-mm wide. The active core is 3.5 m in o.d., 1.65 m in i.d., and 7.93-m tall. Fuel elements contain TRISO-coated microspheres of 19.8% enriched uranium oxycarbide and of fertile thorium oxide. The core is controlled by 30 control rods which enter the inner and outer side reflectors from above.

  13. Annular core for the Modular High-Temperature Gas-cooled Reactor (MHTGR)

    Energy Technology Data Exchange (ETDEWEB)

    Turner, R.F.; Baxter, A.M.; Stansfield, O.M.; Vollman, R.E.

    The active core of the 350 MW(t) MHTGR is annular in configuration, shaped to provide a large external surface-to-volume ratio for the transport of heat radially to the reactor vessel in case of a loss of coolant flow. For a given fuel temperature limit, the annular core provides approximately 40% greater power output over a typical cylindrical configuration. The reactor core is made up to columns of hexagonal blocks, each 793 mm high and 360 mm wide. The active core is 3.5 m in outside diameter, 1.65 m in inside diameter, and 7.93 m tall. Fuel elements contain TRISO-coated microspheres of 19.8% enriched uranium oxycarbide and of fertile thorium oxide. The core is controlled by 30 control rods which enter the inner and outer side reflectors from above.

  14. Research Reactor Design for Export to Myanmar

    Energy Technology Data Exchange (ETDEWEB)

    Win Naing, Lay Lay Myint and Myung-Hyun Kim [Kyunghee Univ. Yongin (Korea, Republic of)

    2006-07-01

    Myanmar is striving to acquire the innovative technology in all field areas including maritime, aerospace and nuclear engineering. There is a high intention to construct a new research reactor for peaceful purposes. The Ministry of Science and Technology (MOST) and Ministry of Education (MOE) are the important government organizations for Myanmar's education and they control most of institutes, universities and colleges. The Department of Atomic Energy (DAE), one of the departments under MOST, leads research projects such as for radiation protection as well as radiation application and coordinates government departments and institutions regarding nuclear energy and its applications. Myanmar's Scientific and Technological Research Department (MSTRD) under MOST guides researches in metallurgy, polymer, pharmacy and biotechnology and so on, and acts as an official body for Myanmar industrial standard. The Department of Higher Education (DHE) under MOE controls art and science universities and colleges including research centers such as Asia Research Center (ARC), Universities Research Center (URC), Microbiology Research Center and so on and does to expand research areas and to utilize advanced technology in science. The wide use of radiation and radioisotopes is developed in Myanmar especially for the field areas such as Medical Science and Agricultural Science. Co{sup 60}, I{sup 131} and Tc{sup 99} are the major use of radioisotopes in diagnosis and therapy. In Agricultural Science, H{sup 3}, C{sup 14}, C{sup 60} etc are used to provide biological effects of radiations on plants, radio-isotopic study of soil physics and tracer studies.

  15. Evaluation of the Shielding Characteristics Test around the Reactor Core in the Prototype Fbr Monju

    Science.gov (United States)

    Usami, Shin; Suzuoki, Zenro; Deshimaru, Takehide; Nakashima, Fumiaki; Hikichi, Takuo

    2003-06-01

    In Monju, shielding measurements were made around the reactor core as a part of the system start-up tests in order to evaluate the design margins of the shielding performance, to demonstrate the validity of the shielding analysis method, and to acquire basic data for use in future FBR design. The measured reaction rates have been obtained radially from the core to the in-vessel storage rack and axially to the reactor vessel upper plenum. The measured values (E) were compared with the calculated values (C) obtained with the FBR shielding analysis system on the basis of the nuclear data library JENDL-3.2. Based upon these results, the design margins around the reactor core have been examined.

  16. Reactors

    CERN Document Server

    International Electrotechnical Commission. Geneva

    1988-01-01

    This standard applies to the following types of reactors: shunt reactors, current-limiting reactors including neutral-earthing reactors, damping reactors, tuning (filter) reactors, earthing transformers (neutral couplers), arc-suppression reactors, smoothing reactors, with the exception of the following reactors: small reactors with a rating generally less than 2 kvar single-phase and 10 kvar three-phase, reactors for special purposes such as high-frequency line traps or reactors mounted on rolling stock.

  17. Exposure calculation code module for reactor core analysis: BURNER

    Energy Technology Data Exchange (ETDEWEB)

    Vondy, D.R.; Cunningham, G.W.

    1979-02-01

    The code module BURNER for nuclear reactor exposure calculations is presented. The computer requirements are shown, as are the reference data and interface data file requirements, and the programmed equations and procedure of calculation are described. The operating history of a reactor is followed over the period between solutions of the space, energy neutronics problem. The end-of-period nuclide concentrations are determined given the necessary information. A steady state, continuous fueling model is treated in addition to the usual fixed fuel model. The control options provide flexibility to select among an unusually wide variety of programmed procedures. The code also provides user option to make a number of auxiliary calculations and print such information as the local gamma source, cumulative exposure, and a fine scale power density distribution in a selected zone. The code is used locally in a system for computation which contains the VENTURE diffusion theory neutronics code and other modules.

  18. Exposure calculation code module for reactor core analysis: BURNER

    Energy Technology Data Exchange (ETDEWEB)

    Vondy, D.R.; Cunningham, G.W.

    1979-02-01

    The code module BURNER for nuclear reactor exposure calculations is presented. The computer requirements are shown, as are the reference data and interface data file requirements, and the programmed equations and procedure of calculation are described. The operating history of a reactor is followed over the period between solutions of the space, energy neutronics problem. The end-of-period nuclide concentrations are determined given the necessary information. A steady state, continuous fueling model is treated in addition to the usual fixed fuel model. The control options provide flexibility to select among an unusually wide variety of programmed procedures. The code also provides user option to make a number of auxiliary calculations and print such information as the local gamma source, cumulative exposure, and a fine scale power density distribution in a selected zone. The code is used locally in a system for computation which contains the VENTURE diffusion theory neutronics code and other modules.

  19. Technical Research for Dedicated Isotope Production Reactor of South Africa

    Institute of Scientific and Technical Information of China (English)

    ZOU; Yao; LIU; Xing-min; CHEN; Hui-qiang; SUN; Zhen; WU; Yuan-yuan

    2012-01-01

    <正>Research reactor plays an important part in nuclear science and technology, application and power development. Currently, many countries in Middle East and Africa are ready to develop their own nuclear industry. South Africa sent its User Requirements Specification (URS) for a dedicated isotope production reactor to several institutes or companies, among of which Department of Reactor Engineering Research and Design (DRERD) in China Institute of Atomic Energy (CIAE) is a competitive candidate.

  20. Initial decommissioning planning for the Budapest research reactor

    OpenAIRE

    Toth Gabor

    2011-01-01

    The Budapest Research Reactor is the first nuclear research facility in Hungary. The reactor is to remain in operation for at least another 13 years. At the same time, the development of a decommissioning plan is a mandatory requirement under national legislation. The present paper describes the current status of decommissioning planning which is aimed at a timely preparation for the forthcoming decommissioning of the reactor.

  1. Initial decommissioning planning for the Budapest research reactor

    Directory of Open Access Journals (Sweden)

    Toth Gabor

    2011-01-01

    Full Text Available The Budapest Research Reactor is the first nuclear research facility in Hungary. The reactor is to remain in operation for at least another 13 years. At the same time, the development of a decommissioning plan is a mandatory requirement under national legislation. The present paper describes the current status of decommissioning planning which is aimed at a timely preparation for the forthcoming decommissioning of the reactor.

  2. Materials science research for sodium cooled fast reactors

    Indian Academy of Sciences (India)

    Baldev Raj

    2009-06-01

    The paper gives an insight into basic as well as applied research being carried out at the Indira Gandhi Centre for Atomic Research for the development of advanced materials for sodium cooled fast reactors towards extending the life of reactors to nearly 100 years and the burnup of fuel to 2,00,000 MWd/t with an objective of providing fast reactor electricity at an affordable and competitive price.

  3. Core and Refueling Design Studies for the Advanced High Temperature Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Holcomb, David Eugene [ORNL; Ilas, Dan [ORNL; Varma, Venugopal Koikal [ORNL; Cisneros, Anselmo T [ORNL; Kelly, Ryan P [ORNL; Gehin, Jess C [ORNL

    2011-09-01

    The Advanced High Temperature Reactor (AHTR) is a design concept for a central generating station type [3400 MW(t)] fluoride-salt-cooled high-temperature reactor (FHR). The overall goal of the AHTR development program is to demonstrate the technical feasibility of FHRs as low-cost, large-size power producers while maintaining full passive safety. This report presents the current status of ongoing design studies of the core, in-vessel structures, and refueling options for the AHTR. The AHTR design remains at the notional level of maturity as important material, structural, neutronic, and hydraulic issues remain to be addressed. The present design space exploration, however, indicates that reasonable options exist for the AHTR core, primary heat transport path, and fuel cycle provided that materials and systems technologies develop as anticipated. An illustration of the current AHTR core, reactor vessel, and nearby structures is shown in Fig. ES1. The AHTR core design concept is based upon 252 hexagonal, plate fuel assemblies configured to form a roughly cylindrical core. The core has a fueled height of 5.5 m with 25 cm of reflector above and below the core. The fuel assembly hexagons are {approx}45 cm across the flats. Each fuel assembly contains 18 plates that are 23.9 cm wide and 2.55 cm thick. The reactor vessel has an exterior diameter of 10.48 m and a height of 17.7 m. A row of replaceable graphite reflector prismatic blocks surrounds the core radially. A more complete reactor configuration description is provided in Section 2 of this report. The AHTR core design space exploration was performed under a set of constraints. Only low enrichment (<20%) uranium fuel was considered. The coated particle fuel and matrix materials were derived from those being developed and demonstrated under the Department of Energy Office of Nuclear Energy (DOE-NE) advanced gas reactor program. The coated particle volumetric packing fraction was restricted to at most 40%. The pressure

  4. In-core materials testing under LWR conditions in the Halden reactor

    Energy Technology Data Exchange (ETDEWEB)

    Bennett, P.J.; Hauso, E.; Hoegberg, N.W.; Karlsen, T.M.; McGrath, M.A. [OECD Halden Reactor Project (Norway)

    2002-07-01

    The Halden boiling water reactor (HBWR) has been in operation since 1958. It is a test reactor with a maximum power of 18 MW and is cooled and moderated by boiling heavy water, with a normal operating temperature of 230 C and a pressure of 34 bar. In the past 15 years increasing emphasis has been placed on materials testing, both of in-core structural materials and fuel claddings. These tests require representative light water reactor (LWR) conditions, which are achieved by housing the test rigs in pressure flasks that are positioned in fuel channels in the reactor and connected to dedicated water loops, in which boiling water reactor (BWR) or pressurised water reactor (PWR) conditions are simulated. Understanding of the in-core behaviour of fuel or reactor materials can be greatly improved by on-line measurements during power operation. The Halden Project has performed in-pile measurements for a period of over 35 years, beginning with fuel temperature measurements using thermocouples and use of differential transformers for measurement of fuel pellet or cladding dimensional changes and internal rod pressure. Experience gained over this period has been applied to on-line instrumentation for use in materials tests. This paper gives details of the systems used at Halden for materials testing under LWR conditions. The techniques used to provide on-line data are described and illustrative results are presented. (authors)

  5. Thermal limits validation of gamma thermometer power adaption in CFE Laguna Verde 2 reactor core

    Energy Technology Data Exchange (ETDEWEB)

    Cuevas V, G.; Banfield, J. [GE-Hitachi Nuclear Energy Americas LLC, Global Nuclear Fuel, Americas LLC, 3901 Castle Hayne Road, Wilmingtonm, North Carolina (United States); Avila N, A., E-mail: Gabriel.Cuevas-Vivas@ge.com [Comision Federal de Electricidad, Central Nucleoelectrica Laguna Verde, Carretera Cardel-Nautla Km 42.5, Alto Lucero, Veracruz (Mexico)

    2016-09-15

    This paper presents the status of GEH work on Gamma Thermometer (GT) validation using the signals of the instruments installed in the Laguna Verde Unit 2 reactor core. The long-standing technical collaboration between Comision Federal de Electricidad (CFE), Global Nuclear Fuel - Americas LLC (GNF) and GE-Hitachi Nuclear Energy Americas LLC (GEH) is moving forward with solid steps to a final implementation of GTs in a nuclear reactor core. Each GT is integrated into a slightly modified Local Power Range Monitor (LPRM) assembly. Six instrumentation strings are equipped with two gamma field detectors for a total of twenty-four bundles whose calculated powers are adapted to the instrumentation readings in addition to their use as calibration instruments for LPRMs. Since November 2007, the six GT instrumentation strings have been operable with almost no degradation by the strong neutron and gamma fluxes in the Laguna Verde Unit 2 reactor core. In this paper, the thermal limits, Critical Power Ratio (CPR) and maximum Linear Heat Generation Rate (LHGR), of bundles directly monitored by either Traverse In-core Probes (TIPs) or GTs are used to establish validation results that confirm the viability of TIP system replacement with automatic fixed in-core probes (AFIPs, GTs, in a Boiling Water Reactor. The new GNF steady-state reactor core simulator AETNA02 is used to obtain power and exposure distribution. Using this code with an updated methodology for GT power adaption, a reduced value of the GT interpolation uncertainty is obtained that is fed into the LHGR calculation. This new method achieves margin recovery for the adapted thermal limits for use in the Economic Simplified Boiling Water Reactor (ESBWR) or any other BWR in the future that employs a GT based AFIP system for local power measurements. (Author)

  6. Nodal Diffusion Burnable Poison Treatment for Prismatic Reactor Cores

    Energy Technology Data Exchange (ETDEWEB)

    A. M. Ougouag; R. M. Ferrer

    2010-10-01

    The prismatic block version of the High Temperature Reactor (HTR) considered as a candidate Very High Temperature Reactor (VHTR)design may use burnable poison pins in locations at some corners of the fuel blocks (i.e., assembly equivalent structures). The presence of any highly absorbing materials, such as these burnable poisons, within fuel blocks for hexagonal geometry, graphite-moderated High Temperature Reactors (HTRs) causes a local inter-block flux depression that most nodal diffusion-based method have failed to properly model or otherwise represent. The location of these burnable poisons near vertices results in an asymmetry in the morphology of the assemblies (or blocks). Hence the resulting inadequacy of traditional homogenization methods, as these “spread” the actually local effect of the burnable poisons throughout the assembly. Furthermore, the actual effect of the burnable poison is primarily local with influence in its immediate vicinity, which happens to include a small region within the same assembly as well as similar regions in the adjacent assemblies. Traditional homogenization methods miss this artifact entirely. This paper presents a novel method for treating the local effect of the burnable poison explicitly in the context of a modern nodal method.

  7. Research on Precaution and Detection Technology for Flow Blockage of Plate-type Fuel Element in Research Reactors

    Institute of Scientific and Technical Information of China (English)

    DING; Li; QIAO; Ya-xin; ZHANG; Nian-peng; LUO; Bei-bei; HUA; Xiao; JIA; Shu-jie; YAN; Hui-yang

    2013-01-01

    The main aim of this study is to offer the technical support for safety operation and management of research reactors using plate-type fuel assemblies in China,which is performed from analysis of precaution measures for flow blockage and detection methods of accidents.Study shows that most accidents were induced by in-core foreign objects and the swelling of fuel

  8. Neutron flux mapping of Argonauta reactor in the new configuration of its reactor core; Mapeamento do fluxo de neutrons do reator Argonauta na nova configuracao do seu nucleo

    Energy Technology Data Exchange (ETDEWEB)

    Souza, Maria Ines Silvani; Furieri, Rosanne Cefaly de Aranda Amado [Instituto de Engenharia Nuclear (IEN), Rio de Janeiro, RJ (Brazil)

    2000-07-01

    Whenever tasks involving the use of a nuclear reactor are carried out, e.g., radioisotope production, activation analysis, neutrongraphy, etc., it is necessary to know the magnitude of the associate neutron flux. The Argonauta reactor operating in Rio de Janeiro, at Instituto de Engenharia Nuclear - IEN/CNEN, was submitted to some modifications in its core, which made necessary to measure again its new neutronic characteristics, not only in the core itself, but also at the irradiation pads. In this type of research reactor, the neutrons are energetically distributed from values below 1 eV, to values reaching the magnitude of MeV. Therefore, depending on the kind of experiment to be conducted, it may become necessary to know the integrated neutron flux within certain energy ranges. In this work, the neutron flux for thermal and epithermal regions were determined by using the foil activation method. To accomplish this goal, two different techniques were applied. In the first technique {beta}-{gamma} gamma coincidence measurements were performed using a proportional 4{pi}{beta} gaseous detector and a NaI(Tl) scintillation detector, while in the second one, gamma spectroscopy was carried out using Hp-Ge and NaI(Tl) detectors. In both cases, the flux was computed using the FLUXO software, specially developed for this purpose. (author)

  9. Status of neutron beam utilization at the Dalat nuclear research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Dien, Nguyen Nhi; Hai, Nguyen Canh [Nuclear Research Institute, Dalat (Viet Nam)

    2003-03-01

    The 500-kW Dalat nuclear research reactor was reconstructed from the USA-made 250-kW TRIGA Mark II reactor. After completion of renovation and upgrading, the reactor has been operating at its nominal power since 1984. The reactor is used mainly for radioisotope production, neutron activation analysis, neutron beam researches and reactor physics study. In the framework of the reconstruction and renovation project of the 1982-1984 period, the reactor core, the control and instrumentation system, the primary and secondary cooling systems, as well as other associated systems were newly designed and installed by the former Soviet Union. Some structures of the reactor, such as the reactor aluminum tank, the graphite reflector, the thermal column, horizontal beam tubes and the radiation concrete shielding have been remained from the previous TRIGA reactor. As a typical configuration of the TRIGA reactor, there are four neutron beam ports, including three radial and one tangential. Besides, there is a large thermal column. Until now only two-neutron beam ports and the thermal column have been utilized. Effective utilization of horizontal experimental channels is one of the important research objectives at the Dalat reactor. The research program on effective utilization of these experimental channels was conducted from 1984. For this purpose, investigations on physical characteristics of the reactor, neutron spectra and fluxes at these channels, safety conditions in their exploitation, etc. have been carried out. The neutron beams, however, have been used only since 1988. The filtered thermal neutron beams at the tangential channel have been extracted using a single crystal silicon filter and mainly used for prompt gamma neutron activation analysis (PGNAA), neutron radiography (NR) and transmission experiments (TE). The filtered quasi-monoenergetic keV neutron beams using neutron filters at the piercing channel have been used for nuclear data measurements, study on

  10. Radiation dosimetry for NCT facilities at the Brookhaven Medical Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Holden, N.E.; Hu, J.P.; Greenberg, D.D.; Reciniello, R.N.

    1998-12-31

    Brookhaven Medical Research Reactor (BMRR) is a 3 mega-watt (MW) heterogeneous, tank-type, light water cooled and moderated, graphite reflected reactor, which was designed for medical and biological studies and became operational in 1959. Over time, the BMRR was modified to provide thermal and epithermal neutron beams suitable for research studies. NCT studies have been performed at both the epithermal neutron irradiation facility (ENIF) on the east side of the BMRR reactor core and the thermal neutron irradiation facility (TNIF) on the west side of the core. Neutron and gamma-ray dosimetry performed from 1994 to the present in both facilities are described and the results are presented and discussed.

  11. Comparison between triangular and hexagonal modeling of a hexagonal-structured reactor core using box method

    Energy Technology Data Exchange (ETDEWEB)

    Malmir, Hessam, E-mail: malmir@energy.sharif.edu [Department of Energy Engineering, Sharif University of Technology, Azadi Street, Tehran (Iran, Islamic Republic of); Moghaddam, Nader Maleki [Department of Nuclear Engineering and Physics, Amir Kabir University of Technology (Tehran Polytechnique), Hafez Street, Tehran (Iran, Islamic Republic of); Zahedinejad, Ehsan [Department of Energy Engineering, Sharif University of Technology, Azadi Street, Tehran (Iran, Islamic Republic of)

    2011-02-15

    A hexagonal-structured reactor core (e.g. VVER-type) is mostly modeled by structured triangular and hexagonal mesh zones. Although both the triangular and hexagonal models give good approximations over the neutronic calculation of the core, there are some differences between them that seem necessary to be clarified. For this purpose, the neutronic calculations of a hexagonal-structured reactor core have to be performed using the structured triangular and hexagonal meshes based on box method of discretisation and then the results of two models should be benchmarked in different cases. In this paper, the box method of discretisation is derived for triangular and hexagonal meshes. Then, two 2-D 2-group static simulators for triangular and hexagonal geometries (called TRIDIF-2 and HEXDIF-2, respectively) are developed using the box method. The results are benchmarked against the well-known CITATION computer code in case of a VVER-1000 reactor core. Furthermore, the relative powers calculated by the TRIDIF-2 and HEXDIF-2 along with the ones obtained by the CITATION code are compared with the verified results which have been presented in the Final Safety Analysis Report (FSAR) of the aforementioned reactor. Different benchmark cases revealed the reliability of the box method in contrast with the CITATION code. Furthermore, it is shown that the triangular modeling of the core is more acceptable compared with the hexagonal one.

  12. Photon spectrum behind biological shielding of the LVR-15 research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Klupak, V.; Viererbl, L.; Lahodova, Z.; Marek, M.; Vins, M. [Research Centre Rez Ltd., Husinec-Rez 130 (Czech Republic)

    2011-07-01

    The LVR-15 reactor is a light water research reactor situated at the Research Centre Rez, near Prague. It operates as a multipurpose facility with a maximum thermal power of 10 MW. The reactor core usually contains from 28 to 32 fuel assemblies with a total mass of {sup 235}U of about 5 kg. Emitted radiation from the fuel caused by fission is shielded by moderating water, a steel reactor vessel, and heavy concrete. This paper deals with measurement and analysis of the gamma spectrum near the outer surface of the concrete wall, behind biological shielding, mainly in the 3- to 10-MeV energy range. A portable HPGe detector with a portable multichannel analyzer was used to measure gamma spectra. The origin of energy lines in gamma detector spectra was identified. (authors)

  13. Conceptual Nuclear Design of a 20 MW Multipurpose Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Seo, Chul Gyo; Kim, Hak Sung; Park, Cheol [KAERI, Daejeon (Korea, Republic of); Nghiem, Huynh Ton; Vinh, Le Vinh; Dang, Vo Doan Hai [Dalat Nuclear Research Reactor, Hanoi (Viet Nam)

    2007-08-15

    A conceptual nuclear design of a 20 MW multi-purpose research reactor for Vietnam has been jointly done by the KAERI and the DNRI (VAEC). The AHR reference core in this report is a right water cooled and a heavy water reflected open-tank-in-pool type multipurpose research reactor with 20 MW. The rod type fuel of a dispersed U{sub 3}Si{sub 2}-Al with a density of 4.0 gU/cc is used as a fuel. The core consists of fourteen 36-element assemblies, four 18-element assemblies and has three in-core irradiation sites. The reflector tank filled with heavy water surrounds the core and provides rooms for various irradiation holes. Major analyses have been done for the relevant nuclear design parameters such as the neutron flux and power distributions, reactivity coefficients, control rod worths, etc. For the analysis, the MCNP, MVP, and HELIOS codes were used by KAERI and DNRI (VAEC). The results by MCNP (KAERI) and MVP (DNRI) showed good agreements and can be summarized as followings. For a clean, unperturbed core condition such that the fuels are all fresh and there are no irradiation holes in the reflector region, the fast neutron flux (E{sub n}{>=}1.0 MeV) reaches 1.47x10{sup 14} n/cm{sup 2}s and the maximum thermal neutron flux (E{sub n}{<=}0.625 eV) reaches 4.43x10{sup 14} n/cm{sup 2}s in the core region. In the reflector region, the thermal neutron peak occurs about 28 cm far from the core center and the maximum thermal neutron flux is estimated to be 4.09x10{sup 14} n/cm{sup 2}s. For the analysis of the equilibrium cycle core, the irradiation facilities in the reflector region were considered. The cycle length was estimated as 38 days long with a refueling scheme of replacing three 36-element fuel assemblies or replacing two 36-element and one 18-element fuel assemblies. The excess reactivity at a BOC was 103.4 mk, and 24.6 mk at a minimum was reserved at an EOC. The assembly average discharge burnup was 54.6% of initial U-235 loading. For the proposed fuel management

  14. Validation of Reactor Physics-Thermal hydraulics Calculations for Research Reactors Cooled by the Laminar Flow of Water

    Energy Technology Data Exchange (ETDEWEB)

    Jordan, K. A.; Schubring, D. [Univ. of Florida, Florida (United States); Girardin, G.; Pautz, A. [Swiss Federal Institute of Technology, Zuerich (Switzerland)

    2013-07-01

    A collaboration between the University of Florida and the Swiss Federal Institute of Technology, Lausanne (EPFL) has been formed to develop and validate detailed coupled multiphysics models of the zero-power (100 W) CROCUS reactor at EPFL and the 100 kW University of Florida Training Reactor, for the comprehensive analysis of the reactor behavior under transient (neutronic or thermal-hydraulic induced) conditions. These two reactors differ significantly in the core design and thermal power output, but share unique heat transfer and flow characteristics. They are characterized by single-phase laminar water flow at near-atmospheric pressures in complex geometries with the possibility of mechanically entrained air bubbles. Validation experiments will be designed to expand the validation domain of these existing models, computational codes and techniques. In this process, emphasis will be placed on validation of the coupled models developed to gain confidence in their applicability for safety analysis. EPFL is responsible for the design and implementation of transient experiments to generate a database of reactor parameters (flow distribution, power profile, and power evolution) to be used to validate against code predictions. The transient experiments performed at EPFL will be simulated on the basis of developed models for these tasks. Comparative analysis will be performed with SERPENT and MCNPX reference core models. UF focuses on the generation of the coupled neutron kinetics and thermal-hydraulic models, including implementation of a TRACE/PARCS reactor simulator model, a PARET model, and development of full-field computational fluid dynamics models (using OpenFOAM) for refined thermal-hydraulics physics treatments. In this subtask of the project, the aim is to verify by means of CFD the validity of TRACE predictions for near-atmospheric pressure water flow in the presence of mechanically entrained air bubbles. The scientific understanding of these multiphysics

  15. Development of a decommissioning strategy for the MR research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Bylkin, Boris; Gorlinsky, Yury; Kolyadin, Vyacheslav; Pavlenko, Vitaly [RRC Kurchatov Institute, Moscow (Russian Federation); Craig, David; Fecitt, Lorna [NUKEM Limited, Dounreay (United Kingdom); Harman, Neil; Jackson, Roger [Serco Technical and Assurance Services, Warrington (United Kingdom); Lobach, Yury [Inst. for Nuclear Research of NASU, Kiev (Ukraine)

    2010-03-15

    A description of the selected decommissioning strategy for the research reactor MR at the site of the Kurchatov Institute in Moscow is presented. The MR reactor hall is planned to be used as a temporary fuel store for the other research reactors on the site. On the basis of the site-specific conditions and over-all decommissioning goals, it was identified that phased immediate decommissioning is the preferable option. The current status of the reactor, expected final conditions and the sequence of decommissioning works are shown. (orig.)

  16. Proceedings of the sixth Asian symposium on research reactors

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1999-08-01

    The symposium consisted of 16 sessions with 58 submitted papers. Major fields were: (1) status and future plan of research and testing reactors, (2) operating experiences, (3) design and modification of the facility, and reactor fuels, (4) irradiation studies, (5) irradiation facilities, (6) reactor characteristics and instrumentation, and (7) neutron beam utilization. Panel discussion on the 'New Trends on Application of Research and Test Reactors' was also held at the last of the symposium. About 180 people participated from China, Korea, Indonesia, Thailand, Bangladesh, Vietnam, Chinese Taipei, Belgium, France, USA, Japan and IAEA. The 58 of the presented papers are indexed individually. (J.P.N.)

  17. Monte Carlo simulation of a research reactor with nominal power of 7 MW to design new control safety rods

    Science.gov (United States)

    Shoushtari, M. K.; Kakavand, T.; Sadat Kiai, S. M.; Ghaforian, H.

    2010-03-01

    The Monte Carlo simulation has been established for a research reactor with nominal power of 7 MW. A detailed model of the reactor core was employed including standard and control fuel elements, reflectors, irradiation channels, control rods, reactor pool and thermal column. The following physical parameters of reactor core were calculated for the present LEU core: core reactivity ( ρ), control rod (CR) worth, thermal and epithermal neutron flux distributions, shutdown margin and delayed neutron fraction. Reduction of unfavorable effects of blockage probability of control safety rod (CSR)s in their interiors because of not enough space in their sites, and lack of suitable capabilities to fabricate very thin plates for CSR cladding, is the main aim of the present study. Making the absorber rod thinner and CSR cladding thicker by introducing a better blackness absorbing material and a new stainless steel alloy, respectively, are two studied ways to reduce the effects of mentioned problems.

  18. Experimental research in neutron physic and thermal-hydraulic at the CDTN Triga reactor

    Energy Technology Data Exchange (ETDEWEB)

    Mesquita, Amir Z.; Souza, Rose Mary G.P.; Ferreira, Andrea V.; Pinto, Antonio J.; Costa, Antonio C.L.; Rezende, Hugo C., E-mail: amir@cdtn.b, E-mail: souzarm@cdtn.b, E-mail: avf@cdtn.b, E-mail: ajp@cdtn.b, E-mail: aclc@cdtn.b, E-mail: hcr@cdtn.b [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

    2011-07-01

    The IPR-R1 TRIGA (Training, Research, Isotopes production, General Atomics) at Nuclear Technology Development Center (CDTN) is a pool type reactor cooled by natural circulation of light water and an open surface. TRIGA reactors, developed by General Atomics (GA), are the most widely used research reactor in the world and characterized by inherent safety. The IPR-R1 is the only Brazilian nuclear research reactor available and able to perform experiments in which interaction between neutronic and thermal-hydraulic areas occurs. The IPR-R1 has started up on November 11th, 1960. At that time the maximum thermal power was 30 kW. The present forced cooling system was built in the 70th and the power was upgraded to 100 kW. Recently the core configuration and instrumentation was upgraded again to 250 kW at steady state, and is awaiting the license of CNEN to operate definitely at this new power. This paper describes the experimental research project carried out in the IPR-R1 reactor that has as objective evaluate the behaviour of the reactor operational parameters, and mainly to investigate the influence of temperature on the neutronic variables. The research was supported by Research Support Foundation of the State of Minas Gerais (FAPEMIG) and Brazilian Council for Scientific and Technological Development (CNPq). The research project meets the recommendations of the IAEA, for safety, modernization and development of strategic plan for research reactors utilization. This work is in line with the strategic objectives of Brazil, which aims to design and construct the Brazilian Multipurpose research Reactor (RMB). (author)

  19. Application of gaseous core reactors for transmutation of nuclear waste

    Science.gov (United States)

    Schnitzler, B. G.; Paternoster, R. R.; Schneider, R. T.

    1976-01-01

    An acceptable management scheme for high-level radioactive waste is vital to the nuclear industry. The hazard potential of the trans-uranic actinides and of key fission products is high due to their nuclear activity and/or chemical toxicity. Of particular concern are the very long-lived nuclides whose hazard potential remains high for hundreds of thousands of years. Neutron induced transmutation offers a promising technique for the treatment of problem wastes. Transmutation is unique as a waste management scheme in that it offers the potential for "destruction" of the hazardous nuclides by conversion to non-hazardous or more manageable nuclides. The transmutation potential of a thermal spectrum uranium hexafluoride fueled cavity reactor was examined. Initial studies focused on a heavy water moderated cavity reactor fueled with 5% enriched U-235-F6 and operating with an average thermal flux of 6 times 10 to the 14th power neutrons/sq cm-sec. The isotopes considered for transmutation were I-129, Am-241, Am-242m, Am-243, Cm-243, Cm-244, Cm-245, and Cm-246.

  20. Application of gaseous core reactors for transmutation of nuclear waste

    Science.gov (United States)

    Schnitzler, B. G.; Paternoster, R. R.; Schneider, R. T.

    1976-01-01

    An acceptable management scheme for high-level radioactive waste is vital to the nuclear industry. The hazard potential of the trans-uranic actinides and of key fission products is high due to their nuclear activity and/or chemical toxicity. Of particular concern are the very long-lived nuclides whose hazard potential remains high for hundreds of thousands of years. Neutron induced transmutation offers a promising technique for the treatment of problem wastes. Transmutation is unique as a waste management scheme in that it offers the potential for "destruction" of the hazardous nuclides by conversion to non-hazardous or more manageable nuclides. The transmutation potential of a thermal spectrum uranium hexafluoride fueled cavity reactor was examined. Initial studies focused on a heavy water moderated cavity reactor fueled with 5% enriched U-235-F6 and operating with an average thermal flux of 6 times 10 to the 14th power neutrons/sq cm-sec. The isotopes considered for transmutation were I-129, Am-241, Am-242m, Am-243, Cm-243, Cm-244, Cm-245, and Cm-246.

  1. Full core analysis of IRIS reactor by using MCNPX.

    Science.gov (United States)

    Amin, E A; Bashter, I I; Hassan, Nabil M; Mustafa, S S

    2016-07-01

    This paper describes neutronic analysis for fresh fuelled IRIS (International Reactor Innovative and Secure) reactor by MCNPX code. The analysis included criticality calculations, radial power and axial power distribution, nuclear peaking factor and axial offset percent at the beginning of fuel cycle. The effective multiplication factor obtained by MCNPX code is compared with previous calculations by HELIOS/NESTLE, CASMO/SIMULATE, modified CORD-2 nodal calculations and SAS2H/KENO-V code systems. It is found that k-eff value obtained by MCNPX is closer to CORD-2 value. The radial and axial powers are compared with other published results carried out using SAS2H/KENO-V code. Moreover, the WIMS-D5 code is used for studying the effect of enriched boron in form of ZrB2 on the effective multiplication factor (K-eff) of the fuel pin. In this part of calculation, K-eff is calculated at different concentrations of Boron-10 in mg/cm at different stages of burnup of unit cell. The results of this part are compared with published results performed by HELIOS code.

  2. Core Plasma Characteristics of a Spherical Tokamak D-3He Fusion Reactor

    Institute of Scientific and Technical Information of China (English)

    Shi Bingren

    2005-01-01

    The magnetic fusion reactor using the advanced D-3He fuels has the advantage of much less-neutron productions so that the consequent damages to the first wall are less serious. If the establishment of this kind of reactor becomes realistic, the exploration of 3He on the moon will be largely motivated. Based on recent progresses in the spherical torus (ST) research, we have physically designed a D-3He fusion reactor using the extrapolated results from the ST experiments and also the present-day tokamak scaling. It is found that the reactor size significantly depends on the wall reflection coefficient of the synchrotron radiation and of the impurity contaminations.The secondary reaction between D-D that promptly leads to the D-T reaction producing 14 MeV neutrons is also estimated. Comparison of this D-3He ST reactor with the D-T reactor is made.

  3. Transient analyses for a molten salt fast reactor with optimized core geometry

    Energy Technology Data Exchange (ETDEWEB)

    Li, R., E-mail: rui.li@kit.edu [Institute for Nuclear and Energy Technologies (IKET), Karlsruhe Institute of Technology (KIT), Hermann-von-Helmholtz-Platz 1, D-76344 Eggenstein-Leopoldshafen (Germany); Wang, S.; Rineiski, A.; Zhang, D. [Institute for Nuclear and Energy Technologies (IKET), Karlsruhe Institute of Technology (KIT), Hermann-von-Helmholtz-Platz 1, D-76344 Eggenstein-Leopoldshafen (Germany); Merle-Lucotte, E. [Laboratoire de Physique Subatomique et de Cosmologie – IN2P3 – CNRS/Grenoble INP/UJF, 53, rue des Martyrs, 38026 Grenoble (France)

    2015-10-15

    Highlights: • MSFR core is analyzed by fully coupling neutronics and thermal-hydraulics codes. • We investigated four types of transients intensively with the optimized core geometry. • It demonstrates MSFR has a high safety potential. - Abstract: Molten salt reactors (MSRs) have encountered a marked resurgence of interest over the past decades, highlighted by their inclusion as one of the six candidate reactors of the Generation IV advanced nuclear power systems. The present work is carried out in the framework of the European FP-7 project EVOL (Evaluation and Viability Of Liquid fuel fast reactor system). One of the project tasks is to report on safety analyses: calculations of reactor transients using various numerical codes for the molten salt fast reactor (MSFR) under different boundary conditions, assumptions, and for different selected scenarios. Based on the original reference core geometry, an optimized geometry was proposed by Rouch et al. (2014. Ann. Nucl. Energy 64, 449) on thermal-hydraulic design aspects to avoid a recirculation zone near the blanket which accumulates heat and very high temperature exceeding the salt boiling point. Using both fully neutronics thermal-hydraulic coupled codes (SIMMER and COUPLE), we also re-confirm the efforts step by step toward a core geometry without the recirculation zone in particular as concerns the modifications of the core geometrical shape. Different transients namely Unprotected Loss of Heat Sink (ULOHS), Unprotected Loss of Flow (ULOF), Unprotected Transient Over Power (UTOP), Fuel Salt Over Cooling (FSOC) are intensively investigated and discussed with the optimized core geometry. It is demonstrated that due to inherent negative feedbacks, an MSFR plant has a high safety potential.

  4. Reducing numerical costs for core wide nuclear reactor CFD simulations by the Coarse-Grid-CFD

    Science.gov (United States)

    Viellieber, Mathias; Class, Andreas G.

    2013-11-01

    Traditionally complete nuclear reactor core simulations are performed with subchannel analysis codes, that rely on experimental and empirical input. The Coarse-Grid-CFD (CGCFD) intends to replace the experimental or empirical input with CFD data. The reactor core consists of repetitive flow patterns, allowing the general approach of creating a parametrized model for one segment and composing many of those to obtain the entire reactor simulation. The method is based on a detailed and well-resolved CFD simulation of one representative segment. From this simulation we extract so-called parametrized volumetric forces which close, an otherwise strongly under resolved, coarsely-meshed model of a complete reactor setup. While the formulation so far accounts for forces created internally in the fluid others e.g. obstruction and flow deviation through spacers and wire wraps, still need to be accounted for if the geometric details are not represented in the coarse mesh. These are modelled with an Anisotropic Porosity Formulation (APF). This work focuses on the application of the CGCFD to a complete reactor core setup and the accomplishment of the parametrization of the volumetric forces.

  5. Partial Safety Analysis for a Reduced Uranium Enrichment Core for the High Flux Isotope Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Primm, Trent [ORNL; Gehin, Jess C [ORNL

    2009-04-01

    A computational model of the reactor core of the High Flux Isotope Rector (HFIR) was developed in order to analyze non-destructive accidents caused by transients during reactor operation. The reactor model was built for the latest version of the nuclear analysis software package called Program for the Analysis of Reactor Transients (PARET). Analyses performed with the model constructed were compared with previous data obtained with other tools in order to benchmark the code. Finally, the model was used to analyze the behavior of the reactor under transients using a different nuclear fuel with lower enrichment of uranium (LEU) than the fuel currently used, which has a high enrichment of uranium (HEU). The study shows that the presence of fertile isotopes in LEU fuel, which increases the neutron resonance absorption, reduces the impact of transients on the fuel and enhances the negative reactivity feedback, thus, within the limitations of this study, making LEU fuel appear to be a safe alternative fuel for the reactor core.

  6. Reactor Core Coolability Analysis during Hypothesized Severe Accidents of OPR1000

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Yongjae; Seo, Seungwon; Kim, Sung Joong [Hanyang University, Seoul (Korea, Republic of); Ha, Kwang Soon; Kim, Hwan-Yeol [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-10-15

    Assessment of the safety features over the hypothesized severe accidents may be performed experimentally or numerically. Due to the considerable time and expenditures, experimental assessment is implemented only to the limited cases. Therefore numerical assessment has played a major role in revisiting severe accident analysis of the existing or newly designed power plants. Computer codes for the numerical analysis of severe accidents are categorized as the fast running integral code and detailed code. Fast running integral codes are characterized by a well-balanced combination of detailed and simplified models for the simulation of the relevant phenomena within an NPP in the case of a severe accident. MAAP, MELCOR and ASTEC belong to the examples of fast running integral codes. Detailed code is to model as far as possible all relevant phenomena in detail by mechanistic models. The examples of detailed code is SCDAP/RELAP5. Using the MELCOR, Carbajo. investigated sensitivity studies of Station Black Out (SBO) using the MELCOR for Peach Bottom BWR. Park et al. conduct regulatory research of the PWR severe accident. Ahn et al. research sensitivity analysis of the severe accident for APR1400 with MELCOR 1.8.4. Lee et al. investigated RCS depressurization strategy and developed a core coolability map for independent scenarios of Small Break Loss-of-Coolant Accident (SBLOCA), SBO, and Total Loss of Feed Water (TLOFW). In this study, three initiating cases were selected, which are SBLOCA without SI, SBO, and TLOFW. The initiating cases exhibit the highest probability of transitioning into core damage according to PSA 1 of OPR 1000. The objective of this study is to investigate the reactor core coolability during hypothesized severe accidents of OPR1000. As a representative indicator, we have employed Jakob number and developed JaCET and JaMCT using the MELCOR simulation. Although the RCS pressures for the respective accident scenarios were different, the JaMCT and Ja

  7. 78 FR 63516 - Initial Test Program of Emergency Core Cooling Systems for New Boiling-Water Reactors

    Science.gov (United States)

    2013-10-24

    ... COMMISSION Initial Test Program of Emergency Core Cooling Systems for New Boiling-Water Reactors AGENCY... Cooling Systems for New Boiling-Water Reactors.'' This RG describes testing methods the NRC staff considers acceptable for demonstrating the operability of emergency core cooling systems (ECCSs) for boiling...

  8. Status and some safety philosophies of the China advanced research reactor CARR

    Energy Technology Data Exchange (ETDEWEB)

    Luzheng Yuan [China Inst. of Atomic Energy, Beijing, BJ (China). Reactor Engineering Research and Design Dept.

    2001-07-01

    The existing two research reactors, HWRR (heavy water research reactor) and SPR (swimming pool reactor), have been operated by China Institute of Atomic Energy (CIAE) since, respectively, 1958 and 1964, and are both in extending service and facing the aging problem. It is expected that they will be out of service successively in the beginning decade of the 21{sup st} century. A new, high performance and multipurpose research reactor called China advanced research reactor (CARR) will replace these two reactors. This new reactor adopts the concept of inverse neutron trap compact core structure with light water as coolant and heavy water as the outer reflector. Its design goal is as follows: under the nuclear power of 60MW, the maximum unperturbed thermal neutron flux in peripheral D{sub 2}O reflector not less than 8 x 10{sup 14} n/cm{sup 2}. s while in central experimental channel, if the central cell to be replaced by an experimental channel, the corresponding value not less than 1 x 10{sup 15} n/cm{sup 2}. s. The main applications for this research reactor will cover RI production, neutron scattering experiments, NAA and its applications, neutron photography, NTD for monocrystaline silicon and applications on reactor engineering technology. By the end of 1999, the preliminary design of CARR was completed, then the draft of preliminary safety analysis report (PSAR) was submitted to the relevant authority at the end of 2000 for being reviewed. Now, the CARR project has entered the detail design phase and safety reviewing procedure for obtaining the construction permit from the relevant licensing authority. This paper will only briefly introduce some aspects of safety philosophy of CARR design and PSAR. (orig.)

  9. 10 MW research reactor simulation using fuel plate type

    Energy Technology Data Exchange (ETDEWEB)

    Mustafa, M. El Sayed, E-mail: memmm67@yahoo.com [Reactors Department, Nuclear Researches Center, Inshas (Egypt); Shaat, M. [Reactors Department, Nuclear Researches Center, Inshas (Egypt); Kady, M. El [Mechanical Power Engineering Department, Faculty of Engineering, Al Azhar University, Cairo (Egypt)

    2016-04-15

    A computer code was established named ET-RR-1-10 to investigate the thermal hydraulic behavior of the ETRR1 (first Egyptian research reactor) research reactor when its power upgraded to 10 MW using the new fuel plate elements type. The work done include both normal and flow reduction conditions. The code modeled primary loop, secondary lop, and reactor kinetics. All code models used finite difference technique. The code results were tested against the available corresponding experimental data taken from a similar research reactor MITR (Massachusetts Institute of Technology research reactor) for the sake of code validation. The results showed good agreement, and the code can be used for thermal hydraulic calculations.

  10. Core questions in domestication research.

    Science.gov (United States)

    Zeder, Melinda A

    2015-03-17

    The domestication of plants and animals is a key transition in human history, and its profound and continuing impacts are the focus of a broad range of transdisciplinary research spanning the physical, biological, and social sciences. Three central aspects of domestication that cut across and unify this diverse array of research perspectives are addressed here. Domestication is defined as a distinctive coevolutionary, mutualistic relationship between domesticator and domesticate and distinguished from related but ultimately different processes of resource management and agriculture. The relative utility of genetic, phenotypic, plastic, and contextual markers of evolving domesticatory relationships is discussed. Causal factors are considered, and two leading explanatory frameworks for initial domestication of plants and animals, one grounded in optimal foraging theory and the other in niche-construction theory, are compared.

  11. Deterministic Analysis of a Beyond Design Basis Accident in a Low Power, Pin-Type Fuel Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Nagah Abdou, Hesham Mohammed [INVAP S. E., Bariloche (Argentina)

    2013-07-01

    A Beyond Design Basis Accident has been analyzed for a pool type research reactor with pin-type, Zry4 clad fuel. This is a low power research reactor (maximum power: 100kW) with neutron beam facilities. Two scenarios are considered: a neutron beam rapture that results in a fraction of the core submerged in water and a catastrophic failure that results in a fully uncovered core. The paper discusses the different cooling mechanisms for these two BDBAs and compares results for both scenarios, with predictions of no core damage in any situation. Core damage is defined as CHFR↔1.5 and/or Tclad→T start of breakaway oxidation temperature. In addition, the paper compares calculations with a thermalhydraulic code and an analytical model. This paper allows to analyze the applicability of regular thermalhydraulic codes to BDBA accident scenarios in low power research reactors.

  12. The effects of stainless steel radial reflector on core reactivity for small modular reactor

    Science.gov (United States)

    Kang, Jung Kil; Hah, Chang Joo; Cho, Sung Ju; Seong, Ki Bong

    2016-01-01

    Commercial PWR core is surrounded by a radial reflector, which consists of a baffle and water. Radial reflector is designed to reflect neutron back into the core region to improve the neutron efficiency of the reactor and to protect the reactor vessels from the embrittling effects caused by irradiation during power operation. Reflector also helps to flatten the neutron flux and power distributions in the reactor core. The conceptual nuclear design for boron-free small modular reactor (SMR) under development in Korea requires to have the cycle length of 4˜5 years, rated power of 180 MWth and enrichment less than 5 w/o. The aim of this paper is to analyze the effects of stainless steel radial reflector on the performance of the SMR using UO2 fuels. Three types of reflectors such as water, water/stainless steel 304 mixture and stainless steel 304 are selected to investigate the effect on core reactivity. Additionally, the thickness of stainless steel and double layer reflector type are also investigated. CASMO-4/SIMULATE-3 code system is used for this analysis. The results of analysis show that single layer stainless steel reflector is the most efficient reflector.

  13. The effects of stainless steel radial reflector on core reactivity for small modular reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Jung Kil, E-mail: jkkang@email.kings.ac.kr; Hah, Chang Joo, E-mail: changhah@kings.ac.kr [KINGS, 658-91, Haemaji-ro, Seosaeng-myeon, Ulju-gun, Ulsan, 689-882 (Korea, Republic of); Cho, Sung Ju, E-mail: sungju@knfc.co.kr; Seong, Ki Bong, E-mail: kbseong@knfc.co.kr [KNFC, Daedeok-daero 989beon-gil, Yuseong-gu, Daejeon, 305-353 (Korea, Republic of)

    2016-01-22

    Commercial PWR core is surrounded by a radial reflector, which consists of a baffle and water. Radial reflector is designed to reflect neutron back into the core region to improve the neutron efficiency of the reactor and to protect the reactor vessels from the embrittling effects caused by irradiation during power operation. Reflector also helps to flatten the neutron flux and power distributions in the reactor core. The conceptual nuclear design for boron-free small modular reactor (SMR) under development in Korea requires to have the cycle length of 4∼5 years, rated power of 180 MWth and enrichment less than 5 w/o. The aim of this paper is to analyze the effects of stainless steel radial reflector on the performance of the SMR using UO{sub 2} fuels. Three types of reflectors such as water, water/stainless steel 304 mixture and stainless steel 304 are selected to investigate the effect on core reactivity. Additionally, the thickness of stainless steel and double layer reflector type are also investigated. CASMO-4/SIMULATE-3 code system is used for this analysis. The results of analysis show that single layer stainless steel reflector is the most efficient reflector.

  14. A Burst Mode, Ultrahigh Temperature UF4 Vapor Core Reactor Rankine Cycle Space Power System Concept

    Science.gov (United States)

    Dugan, E. T.; Kahook, S. D.; Diaz, N. J.

    1996-01-01

    Static and dynamic neutronic analyses have been performed on an innovative burst mode (100's of MW output for a few thousand seconds) Ulvahigh Temperature Vapor Core Reactor (UTVR) space nuclear power system. The NVTR employs multiple, neutronically-coupled fissioning cores and operates on a direct, closed Rankine cycle using a disk Magnetohydrodynamic (MHD) generater for energy conversion. The UTVR includes two types of fissioning core regions: (1) the central Ultrahigh Temperature Vapor Core (UTVC) which contains a vapor mixture of highly enriched UF4 fuel and a metal fluoride working fluid and (2) the UF4 boiler column cores located in the BeO moderator/reflector region. The gaseous nature of the fuel the fact that the fuel is circulating, the multiple coupled fissioning cores, and the use of a two phase fissioning fuel lead to unique static and dynamic neutronic characteristics. Static neutronic analysis was conducted using two-dimensional S sub n, transport theory calculations and three-dimensional Monte Carlo transport theory calculations. Circulating-fuel, coupled-core point reactor kinetics equations were used for analyzing the dynamic behavior of the UTVR. In addition to including reactivity feedback phenomena associated with the individual fissioning cores, the effects of core-to-core neutronic and mass flow coupling between the UTVC and the surrounding boiler cores were also included in the dynamic model The dynamic analysis of the UTVR reveals the existence of some very effectlve inherent reactivity feedback effects that are capable of quickly stabilizing this system, within a few seconds, even when large positive reactivity insertions are imposed. If the UTVC vapor fuel density feedback is suppressed, the UTVR is still inherently stable because of the boiler core liquid-fuel volume feedback; in contrast, suppression of the vapor fuel density feedback in 'conventional" gas core cavity reactors causes them to become inherently unstable. Due to the

  15. Core Design and Deployment Strategy of Heavy Water Cooled Sustainable Thorium Reactor

    Directory of Open Access Journals (Sweden)

    Naoyuki Takaki

    2012-08-01

    Full Text Available Our previous studies on water cooled thorium breeder reactor based on matured pressurized water reactor (PWR plant technology concluded that reduced moderated core by arranging fuel pins in a triangular tight lattice array and using heavy water as coolant is appropriate for achieving better breeding performance and higher burn-up simultaneously [1–6]. One optimum core that produces 3.5 GW thermal energy using Th-233U oxide fuel shows a breeding ratio of 1.07 and averaged burn-up of about 80 GWd/t with long cycle length of 1300 days. The moderator to fuel volume ratio is 0.6 and required enrichment of 233U for the fresh fuel is about 7%. The coolant reactivity coefficient is negative during all cycles despite it being a large scale breeder reactor. In order to introduce this sustainable thorium reactor, three-step deployment scenario, with intermediate transition phase between current light water reactor (LWR phase and future sustainer phase, is proposed. Both in transition phase and sustainer phase, almost the same core design can be applicable only by changing fissile materials mixed with thorium from plutonium to 233U with slight modification in the fuel assembly design. Assuming total capacity of 60 GWe in current LWR phase and reprocessing capacity of 800 ton/y with further extensions to 1600 ton/y, all LWRs will be replaced by heavy water cooled thorium reactors within about one century then thorium reactors will be kept operational owing to its potential to sustain fissile fuels while reprocessing all spent fuels until exhaustion of massive thorium resource.

  16. Feasibility study on nuclear core design for soluble boron free small modular reactor

    Science.gov (United States)

    Rabir, Mohamad Hairie; Hah, Chang Joo; Ju, Cho Sung

    2015-04-01

    A feasibility study on nuclear core design of soluble boron free (SBF) core for small size (150MWth) small modular reactor (SMR) was investigated. The purpose of this study was to design a once through cycle SMR core, where it can be used to supply electricity to a remote isolated area. PWR fuel assembly design with 17×17 arrangement, with 264 fuel rods per assembly was adopted as the basis design. The computer code CASMO-3/MASTER was used for the search of SBF core and fuel assembly analysis for SMR design. A low critical boron concentration (CBC) below 200 ppm core with 4.7 years once through cycle length was achieved using 57 fuel assemblies having 170 cm of active height. Core reactivity controlled using mainly 512 number of 4 wt% and 960 12 wt% Gd rods.

  17. Feasibility study on nuclear core design for soluble boron free small modular reactor

    Energy Technology Data Exchange (ETDEWEB)

    Rabir, Mohamad Hairie, E-mail: m-hairie@nuclearmalaysia.gov.my; Hah, Chang Joo; Ju, Cho Sung [Department of NPP Engineering, KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of)

    2015-04-29

    A feasibility study on nuclear core design of soluble boron free (SBF) core for small size (150MWth) small modular reactor (SMR) was investigated. The purpose of this study was to design a once through cycle SMR core, where it can be used to supply electricity to a remote isolated area. PWR fuel assembly design with 17×17 arrangement, with 264 fuel rods per assembly was adopted as the basis design. The computer code CASMO-3/MASTER was used for the search of SBF core and fuel assembly analysis for SMR design. A low critical boron concentration (CBC) below 200 ppm core with 4.7 years once through cycle length was achieved using 57 fuel assemblies having 170 cm of active height. Core reactivity controlled using mainly 512 number of 4 wt% and 960 12 wt% Gd rods.

  18. Modular Pebble Bed Reactor Project, University Research Consortium Annual Report

    Energy Technology Data Exchange (ETDEWEB)

    Petti, David Andrew

    2000-07-01

    This project is developing a fundamental conceptual design for a gas-cooled, modular, pebble bed reactor. Key technology areas associated with this design are being investigated which intend to address issues concerning fuel performance, safety, core neutronics and proliferation resistance, economics and waste disposal. Research has been initiated in the following areas: · Improved fuel particle performance · Reactor physics · Economics · Proliferation resistance · Power conversion system modeling · Safety analysis · Regulatory and licensing strategy Recent accomplishments include: · Developed four conceptual models for fuel particle failures that are currently being evaluated by a series of ABAQUS analyses. Analytical fits to the results are being performed over a range of important parameters using statistical/factorial tools. The fits will be used in a Monte Carlo fuel performance code, which is under development. · A fracture mechanics approach has been used to develop a failure probability model for the fuel particle, which has resulted in significant improvement over earlier models. · Investigation of fuel particle physio-chemical behavior has been initiated which includes the development of a fission gas release model, particle temperature distributions, internal particle pressure, migration of fission products, and chemical attack of fuel particle layers. · A balance of plant, steady-state thermal hydraulics model has been developed to represent all major components of a MPBR. Component models are being refined to accurately reflect transient performance. · A comparison between air and helium for use in the energy-conversion cycle of the MPBR has been completed and formed the basis of a master’s degree thesis. · Safety issues associated with air ingress are being evaluated. · Post shutdown, reactor heat removal characteristics are being evaluated by the Heating-7 code. · PEBBED, a fast deterministic neutronic code package suitable for

  19. Sodium fast reactor safety and licensing research plan. Volume II.

    Energy Technology Data Exchange (ETDEWEB)

    Ludewig, H. (Brokhaven National Laboratory, Upton, NY); Powers, D. A.; Hewson, John C.; LaChance, Jeffrey L.; Wright, A. (Argonne National Laboratory, Argonne, IL); Phillips, J.; Zeyen, R. (Institute for Energy Petten, Saint-Paul-lez-Durance, France); Clement, B. (IRSN/DPAM.SEMIC Bt 702, Saint-Paul-lez-Durance, France); Garner, Frank (Radiation Effects Consulting, Richland, WA); Walters, Leon (Advanced Reactor Concepts, Los Alamos, NM); Wright, Steve; Ott, Larry J. (Oak Ridge National Laboratory, Oak Ridge, TN); Suo-Anttila, Ahti Jorma; Denning, Richard (Ohio State University, Columbus, OH); Ohshima, Hiroyuki (Japan Atomic Energy Agency, Ibaraki, Japan); Ohno, S. (Japan Atomic Energy Agency, Ibaraki, Japan); Miyhara, S. (Japan Atomic Energy Agency, Ibaraki, Japan); Yacout, Abdellatif (Argonne National Laboratory, Argonne, IL); Farmer, M. (Argonne National Laboratory, Argonne, IL); Wade, D. (Argonne National Laboratory, Argonne, IL); Grandy, C. (Argonne National Laboratory, Argonne, IL); Schmidt, R.; Cahalen, J. (Argonne National Laboratory, Argonne, IL); Olivier, Tara Jean; Budnitz, R. (Lawrence Berkeley National Laboratory, Berkeley, CA); Tobita, Yoshiharu (Japan Atomic Energy Agency, Ibaraki, Japan); Serre, Frederic (Centre d' %C3%94etudes nucl%C3%94eaires de Cadarache, Cea, France); Natesan, Ken (Argonne National Laboratory, Argonne, IL); Carbajo, Juan J. (Oak Ridge National Laboratory, Oak Ridge, TN); Jeong, Hae-Yong (Korea Atomic Energy Research Institute, Daejeon, Korea); Wigeland, Roald (Idaho National Laboratory, Idaho Falls, ID); Corradini, Michael (University of Wisconsin-Madison, Madison, WI); Thomas, Justin (Argonne National Laboratory, Argonne, IL); Wei, Tom (Argonne National Laboratory, Argonne, IL); Sofu, Tanju (Argonne National Laboratory, Argonne, IL); Flanagan, George F. (Oak Ridge National Laboratory, Oak Ridge, TN); Bari, R. (Brokhaven National Laboratory, Upton, NY); Porter D. (Idaho National Laboratory, Idaho Falls, ID); Lambert, J. (Argonne National Laboratory, Argonne, IL); Hayes, S. (Idaho National Laboratory, Idaho Falls, ID); Sackett, J. (Idaho National Laboratory, Idaho Falls, ID); Denman, Matthew R.

    2012-05-01

    Expert panels comprised of subject matter experts identified at the U.S. National Laboratories (SNL, ANL, INL, ORNL, LBL, and BNL), universities (University of Wisconsin and Ohio State University), international agencies (IRSN, CEA, JAEA, KAERI, and JRC-IE) and private consultation companies (Radiation Effects Consulting) were assembled to perform a gap analysis for sodium fast reactor licensing. Expert-opinion elicitation was performed to qualitatively assess the current state of sodium fast reactor technologies. Five independent gap analyses were performed resulting in the following topical reports: (1) Accident Initiators and Sequences (i.e., Initiators/Sequences Technology Gap Analysis), (2) Sodium Technology Phenomena (i.e., Advanced Burner Reactor Sodium Technology Gap Analysis), (3) Fuels and Materials (i.e., Sodium Fast Reactor Fuels and Materials: Research Needs), (4) Source Term Characterization (i.e., Advanced Sodium Fast Reactor Accident Source Terms: Research Needs), and (5) Computer Codes and Models (i.e., Sodium Fast Reactor Gaps Analysis of Computer Codes and Models for Accident Analysis and Reactor Safety). Volume II of the Sodium Research Plan consolidates the five gap analysis reports produced by each expert panel, wherein the importance of the identified phenomena and necessities of further experimental research and code development were addressed. The findings from these five reports comprised the basis for the analysis in Sodium Fast Reactor Research Plan Volume I.

  20. PHYSICS AND SAFETY ANALYSIS FOR THE NIST RESEARCH REACTOR.

    Energy Technology Data Exchange (ETDEWEB)

    Cheng, L.; Diamond, D.; Xu, J.; Carew, J.; Rorer, D.

    2004-03-31

    Detailed reactor physics and safety analyses have been performed for the 20 MW D{sub 2}O moderated research reactor (NBSR) at the National Institute of Standards and Technology (NIST). The analyses provide an update to the Final Safety Analysis Report (FSAR) and employ state-of-the-art calculational methods. Three-dimensional Monte Carlo neutron and photon transport calculations were performed with the MCNP code to determine the safety parameters for the NBSR. The core depletion and determination of the fuel compositions were performed with MONTEBURNS. MCNP calculations were performed to determine the beginning, middle, and end-of-cycle power distributions, moderator temperature coefficient, and shim safety arm, beam tube and void reactivity worths. The calculational model included a plate-by-plate description of each fuel assembly, axial mid-plane water gap, beam tubes and the tubular geometry of the shim safety arms. The time-dependent analysis of the primary loop was determined with a RELAP5 transient analysis model that includes the pump, heat exchanger, fuel element geometry, and flow channels for both the six inner and twenty-four outer fuel elements. The statistical analysis used to assure protection from critical heat flux (CHF) was performed using a Monte Carlo simulation of the uncertainties contributing to the CHF calculation. The power distributions used to determine the local fuel conditions and margin to CHF were determined with MCNP. Evaluations were performed for the following accidents: (1) the control rod withdrawal startup accident, (2) the maximum reactivity insertion accident, (3) loss-of-flow resulting from loss of electrical power, (4) loss-of-flow resulting from a primary pump seizure, (5) loss-of-flow resulting from inadvertent throttling of a flow control valve, (6) loss-of-flow resulting from failure of both shutdown cooling pumps and (7) misloading of a fuel element. In both the startup and maximum reactivity insertion accidents, the

  1. High Temperature Reactor (HTR) Deep Burn Core and Fuel Analysis: Design Selection for the Prismatic Block Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Francesco Venneri; Chang-Keun Jo; Jae-Man Noh; Yonghee Kim; Claudio Filippone; Jonghwa Chang; Chris Hamilton; Young-Min Kim; Ji-Su Jun; Moon-Sung Cho; Hong-Sik Lim; MIchael A. Pope; Abderrafi M. Ougouag; Vincent Descotes; Brian Boer

    2010-09-01

    The Deep Burn (DB) Project is a U.S. Department of Energy sponsored feasibility study of Transuranic Management using high burnup fuel in the high temperature helium cooled reactor (HTR). The DB Project consists of seven tasks: project management, core and fuel analysis, spent fuel management, fuel cycle integration, TRU fuel modeling, TRU fuel qualification, and HTR fuel recycle. In the Phase II of the Project, we conducted nuclear analysis of TRU destruction/utilization in the HTR prismatic block design (Task 2.1), deep burn fuel/TRISO microanalysis (Task 2.3), and synergy with fast reactors (Task 4.2). The Task 2.1 covers the core physics design, thermo-hydraulic CFD analysis, and the thermofluid and safety analysis (low pressure conduction cooling, LPCC) of the HTR prismatic block design. The Task 2.3 covers the analysis of the structural behavior of TRISO fuel containing TRU at very high burnup level, i.e. exceeding 50% of FIMA. The Task 4.2 includes the self-cleaning HTR based on recycle of HTR-generated TRU in the same HTR. Chapter IV contains the design and analysis results of the 600MWth DB-HTR core physics with the cycle length, the average discharged burnup, heavy metal and plutonium consumptions, radial and axial power distributions, temperature reactivity coefficients. Also, it contains the analysis results of the 450MWth DB-HTR core physics and the analysis of the decay heat of a TRU loaded DB-HTR core. The evaluation of the hot spot fuel temperature of the fuel block in the DB-HTR (Deep-Burn High Temperature Reactor) core under full operating power conditions are described in Chapter V. The investigated designs are the 600MWth and 460MWth DB-HTRs. In Chapter VI, the thermo-fluid and safety of the 600MWth DB-HTRs has been analyzed to investigate a thermal-fluid design performance at the steady state and a passive safety performance during an LPCC event. Chapter VII describes the analysis results of the TRISO fuel microanalysis of the 600MWth and 450

  2. Self powered neutron detectors as in-core detectors for Sodium-cooled Fast Reactors

    Science.gov (United States)

    Verma, V.; Barbot, L.; Filliatre, P.; Hellesen, C.; Jammes, C.; Svärd, S. Jacobsson

    2017-07-01

    Neutron flux monitoring system forms an integral part of the design of a Generation IV sodium cooled fast reactor. Diverse possibilities of detector system installation must be studied for various locations in the reactor vessel in order to detect any perturbations in the core. Results from a previous paper indicated that it is possible to detect changes in neutron source distribution initiated by an inadvertent withdrawal of outer control rod with in-vessel fission chambers located azimuthally around the core. It is, however, not possible to follow inner control rod withdrawal and precisely know the location of the perturbation in the core. Hence the use of complimentary in-core detectors coupled with the peripheral fission chambers is proposed to enable robust core monitoring across the radial direction. In this paper, we assess the feasibility of using self-powered neutron detectors (SPNDs) as in-core detectors in fast reactors for detecting local changes in the power distribution when the reactor is operated at nominal power. We study the neutron and gamma contributions to the total output current of the detector modelled with Platinum as the emitter material. It is shown that this SPND placed in an SFR-like environment would give a sufficiently measurable prompt neutron induced current of the order of 600 nA/m. The corresponding induced current in the connecting cable is two orders of magnitude lower and can be neglected. This means that the SPND can follow in-core power fluctuations. This validates the operability of an SPND in an SFR-like environment.

  3. Licensing of the Australian replacement research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Garea, Veronica B. [INVAP S.E., San Carlos de Bariloche (Argentina)

    2002-07-01

    The Australian Nuclear Science and Technology Organisation (ANSTO)'s Replacement Research Reactor has been submitted to a comprehensive licensing process of which peer review has been a fundamental part. Following Australian Regulation, an application for a site licence was the first step supported by an Environmental Impact Statement approved by The Minister for the Environment and Heritage, and a Reference Accident Analysis. After the site licence had been granted and the contract awarded to the Designer and Constructor, INVAP S.E:, a 2500 page Preliminary Safety Analysis Report was submitted by ANSTO to the Australian Radiation Protection and Nuclear Safety Agency (ARPANSA), which conducted its review. ARPANSA requested that the PSAR be also reviewed by an experts mission from the International Atomic Energy Agency. The PSAR was also reviewed by the Argentine Regulatory Body, it was submitted to public examination in Australia and it was reviewed by international experts hired as consultants by several Australian organisations. A public forum was also held in Sydney. The Regulator, the applicant and the Designer-Constructor maintained constant interaction during the whole process, so that questions, comments and observations that arose from the review of the PSAR were fed back to the designers. This process allowed for a robust, safe design enriched by the results of the safety analysis and review process. (author)

  4. Evaluation of the Start-Up Core Physics Tests at Japan's High Temperature Engineering Test Reactor (Annular Core Loadings)

    Energy Technology Data Exchange (ETDEWEB)

    John D. Bess; Nozomu Fujimoto; James W. Sterbentz; Luka Snoj; Atsushi Zukeran

    2010-03-01

    The High Temperature Engineering Test Reactor (HTTR) of the Japan Atomic Energy Agency (JAEA) is a 30 MWth, graphite-moderated, helium-cooled reactor that was constructed with the objectives to establish and upgrade the technological basis for advanced high-temperature gas-cooled reactors (HTGRs) as well as to conduct various irradiation tests for innovative high-temperature research. The core size of the HTTR represents about one-half of that of future HTGRs, and the high excess reactivity of the HTTR, necessary for compensation of temperature, xenon, and burnup effects during power operations, is similar to that of future HTGRs. During the start-up core physics tests of the HTTR, various annular cores were formed to provide experimental data for verification of design codes for future HTGRs. The Japanese government approved construction of the HTTR in the 1989 fiscal year budget; construction began at the Oarai Research and Development Center in March 1991 and was completed May 1996. Fuel loading began July 1, 1998, from the core periphery. The first criticality was attained with an annular core on November 10, 1998 at 14:18, followed by a series of start-up core physics tests until a fully-loaded core was developed on December 16, 1998. Criticality tests were carried out into January 1999. The first full power operation with an average core outlet temperature of 850ºC was completed on December 7, 2001, and operational licensing of the HTTR was approved on March 6, 2002. The HTTR attained high temperature operation at 950 ºC in April 19, 2004. After a series of safety demonstration tests, it will be used as the heat source in a hydrogen production system by 2015. Hot zero-power critical, rise-to-power, irradiation, and safety demonstration testing , have also been performed with the HTTR, representing additional means for computational validation efforts. Power tests were performed in steps from 0 to 30 MW, with various tests performed at each step to confirm

  5. An overview of the FZJ-tools for HTR core design and reactor dynamics, the past, present and future

    Energy Technology Data Exchange (ETDEWEB)

    Reitsma, F. [Nuclear Engineering Analysis (NEA), PBMR (Pty Ltd), cennturion (South Africa); Rutten, H.J.; Scherer, W. [Forschungzentrum Julich GmbH, Institute for Safety Research and Reactor Technology, Julich (Germany)

    2005-07-01

    The development of the pebble-bed type high-temperature reactor in Germany was actively supported by the research centre Juelich (FZJ, former KFA) i.e. with the development of theoretical methods and computational tools to perform core neutronics design, reactor operation simulation and transient analysis. The tools, developed as the outcome of research activities, made huge contributions not only to the understanding of the technology and its physical behaviour but were also used in support of licensing of German HTR projects. Today these codes are used in the design and licensing of current commercial projects such as the Pebble Bed Modular Reactor (PBMR) but is also used to design the upcoming Gen IV reactors of HTR pebble-bed type. The renewed interest in these codes especially with respect to the pebble-bed designs is due to their unique features such as the fuel management algorithms, the simultaneous treatment of nuclear, thermal-hydraulic and fluid-dynamic problems and the description of fast and long-term transients. The paper provides an overview of the codes VSOP and TINTE, provides an update of recent developments in the codes and gives specific examples of applications such as the PBMR 400 MW running-in phase, a comparison with the SANA pebble bed effective thermal conductivity experiment and very recent results obtained simulating the corrosion due to an air ingress event simulated in the NACOK facility. Finally some ideas on the future development of these codes are discussed. (authors)

  6. Cosmic Ray Radiography of the Damaged Cores of the Fukushima Reactors

    CERN Document Server

    Borozdin, Konstantin; Lukić, Zarija; Milner, Edward Cas; Miyadera, Haruo; Morris, Christopher; Perry, John

    2012-01-01

    The passage of muons through matter is dominated by the Coulomb interaction with electrons and nuclei. The interaction with the electrons leads to continuous energy loss and stopping of the muons. The interaction with nuclei leads to angle diffusion. Two muon imaging methods that use flux attenuation and multiple Coulomb scattering of cosmic-ray muons are being studied as tools for diagnosing the damaged cores of the Fukushima reactors. Here we compare these two methods. We conclude that the scattering method can provide detailed information about the core. Attenuation has low contrast and little sensitivity to the core.

  7. Investigation of Abnormal Heat Transfer and Flow in a VHTR Reactor Core

    Energy Technology Data Exchange (ETDEWEB)

    Kawaji, Masahiro [City College of New York, NY (United States); Valentin, Francisco I. [City College of New York, NY (United States); Artoun, Narbeh [City College of New York, NY (United States); Banerjee, Sanjoy [City College of New York, NY (United States); Sohal, Manohar [Idaho National Lab. (INL), Idaho Falls, ID (United States); Schultz, Richard [Idaho National Lab. (INL), Idaho Falls, ID (United States); McEligot, Donald M. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-12-21

    The main objective of this project was to identify and characterize the conditions under which abnormal heat transfer phenomena would occur in a Very High Temperature Reactor (VHTR) with a prismatic core. High pressure/high temperature experiments have been conducted to obtain data that could be used for validation of VHTR design and safety analysis codes. The focus of these experiments was on the generation of benchmark data for design and off-design heat transfer for forced, mixed and natural circulation in a VHTR core. In particular, a flow laminarization phenomenon was intensely investigated since it could give rise to hot spots in the VHTR core.

  8. Multipurpose epithermal neutron beam on new research station at MARIA research reactor in Swierk-Poland

    Energy Technology Data Exchange (ETDEWEB)

    Gryzinski, M.A.; Maciak, M. [National Centre for Nuclear Research, Andrzeja Soltana 7, 05-400 Otwock-Swierk (Poland)

    2015-07-01

    MARIA reactor is an open-pool research reactor what gives the chance to install uranium fission converter on the periphery of the core. It could be installed far enough not to induce reactivity of the core but close enough to produce high flux of fast neutrons. Special design of the converter is now under construction. It is planned to set the research stand based on such uranium converter in the near future: in 2015 MARIA reactor infrastructure should be ready (preparation started in 2013), in 2016 the neutron beam starts and in 2017 opening the stand for material and biological research or for medical training concerning BNCT. Unused for many years, horizontal channel number H2 at MARIA research rector in Poland, is going to be prepared as a part of unique stand. The characteristics of the neutron beam will be significant advantage of the facility. High flux of neutrons at the level of 2x10{sup 9} cm{sup -2}s{sup -1} will be obtainable by uranium neutron converter located 90 cm far from the reactor core fuel elements (still inside reactor core basket between so called core reflectors). Due to reaction of core neutrons with converter U{sub 3}Si{sub 2} material it will produce high flux of fast neutrons. After conversion neutrons will be collimated and moderated in the channel by special set of filters and moderators. At the end of H2 channel i.e. at the entrance to the research room neutron energy will be in the epithermal energy range with neutron intensity at least at the level required for BNCT (2x10{sup 9} cm{sup -2}s{sup -1}). For other purposes density of the neutron flux could be smaller. The possibility to change type and amount of installed filters/moderators which enables getting different properties of the beam (neutron energy spectrum, neutron-gamma ratio and beam profile and shape) is taken into account. H2 channel is located in separate room which is adjacent to two other empty rooms under the preparation for research laboratories (200 m2). It is

  9. Core Research Program, Year 5

    Science.gov (United States)

    2002-01-01

    Dramatic losses of bone mineral density (BMD) and muscle strength are two of the best documented changes observed in humans after prolonged exposure to microgravity. Recovery of muscle upon return to a 1-G environment is well studied, however, far less is known about the rate and completeness of BMD recovery to pre-flight values. Using the mature tail-suspended adult rat model, this proposal will focus on the temporal course of recovery in tibial bone following a 28-d period of skeletal unloading. Through the study of bone density and muscle strength in the same animal, time-points during recovery from simulated microgravity will be identified when bone is at an elevated risk for fracture. These will occur due to the rapid recovery of muscle strength coupled with a slower recovery of bone, producing a significant mismatch in functional strength of these two tissues. Once the time-point of maximal mismatch is defined, various mechanical and pharmacological interventions will be tested at and around this time-point in attempt to minimize the functional difference of bone and muscle. The outcomes of this research will have high relevance for optimizing the rehabilitation of astronauts upon return to Earth, as well as upon landing on the Martian surface before assuming arduous physical tasks. Further. it will impact significantly on rehabilitation issues common to patients experiencing long periods of limb immobilization or bed rest.

  10. Evaluation of thermal-hydraulic parameter uncertainties in a TRIGA research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Mesquita, Amir Z.; Costa, Antonio C.L.; Ladeira, Luiz C.D.; Rezende, Hugo C., E-mail: amir@cdtn.br, E-mail: aclc@cdtn.br, E-mail: lcdl@cdtn.br, E-mail: hcr@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil); Palma, Daniel A.P., E-mail: dapalma@cnen.gov.br [Comissao Nacional de Energia Nuclear (CNEN), Rio de Janeiro, RJ (Brazil)

    2015-07-01

    Experimental studies had been performed in the TRIGA Research Nuclear Reactor of CDTN/CNEN to find out the its thermal hydraulic parameters. Fuel to coolant heat transfer patterns must be evaluated as function of the reactor power in order to assess the thermal hydraulic performance of the core. The heat generated by nuclear fission in the reactor core is transferred from fuel elements to the cooling system through the fuel-cladding (gap) and the cladding to coolant interfaces. As the reactor core power increases the heat transfer regime from the fuel cladding to the coolant changes from single-phase natural convection to subcooled nucleate boiling. This paper presents the uncertainty analysis in the results of the thermal hydraulics experiments performed. The methodology used to evaluate the propagation of uncertainty in the results was done based on the pioneering article of Kline and McClintock, with the propagation of uncertainties based on the specification of uncertainties in various primary measurements. The uncertainty analysis on thermal hydraulics parameters of the CDTN TRIGA fuel element is determined, basically, by the uncertainty of the reactor's thermal power. (author)

  11. Shippingport operations with the Light Water Breeder Reactor core. (LWBR Development Program)

    Energy Technology Data Exchange (ETDEWEB)

    Budd, W.A. (ed.)

    1986-03-01

    This report describes the operation of the Shippingport Atomic Power Station during the LWBR (Light Water Breeder Reactor) Core lifetime. It also summarizes the plant-oriented operations during the period preceding LWBR startup, which include the defueling of The Pressurized Water Reactor Core 2 (PWR-2) and the installation of the LWBR Core, and the operations associated with the defueling of LWBR. The intent of this report is to examine LWBR experience in retrospect and present pertinent and significant aspects of LWBR operations that relate primarily to the nuclear portion of the Station. The nonnuclear portion of the Station is discussed only as it relates to overall plant operation or to unusual problems which result from the use of conventional equipment in radioactive environments. 30 refs., 69 figs., 27 tabs.

  12. Thermal Hydraulic Analysis of 3 MW TRIGA Research Reactor of Bangladesh Considering Different Cycles of Burnup

    Directory of Open Access Journals (Sweden)

    M.H. Altaf

    2014-12-01

    Full Text Available Burnup dependent steady state thermal hydraulic analysis of TRIGA Mark-II research reactor has been carried out utilizing coupled point kinetics, neutronics and thermal hydraulics code EUREKA-2/RR. From the previous calculations of neutronics parameters including percentage burnup of individual fuel elements performed so far for 700 MWD burnt core of TRIGA reactor showed that the fuel rod predicted as hottest at the beginning of cycle (fresh core was found to remain as the hottest until 200 MWD of burn, but, with the progress of core burn, the hottest rod was found to be shifted and another rod in the core became the hottest. The present study intends to evaluate the thermal hydraulic parameters of these hottest fuel rods at different cycles of burnup, from beginning to 700 MWD core burnt considering reactor operates under steady state condition. Peak fuel centerline temperature, maximum cladding and coolant temperatures of the hottest channels were calculated. It revealed that maximum temperature reported for fuel clad and fuel centerline found to lie below their melting points which indicate that there is no chance of burnout on the fuel cladding surface and no blister in the fuel meat throughout the considered cycles of core burnt.

  13. Current Status of the Transmutation Reactor Technology and Preliminary Evaluation of Transmutation Performance of the KALIMER Core

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Ser Gi; Sim, Yoon Sub; Kim, Yeong Il; Kim, Young Gyum; Lee, Byung Woon; Song, Hoon; Lee, Ki Bog; Jang, Jin Wook; Lee, Dong Uk

    2005-08-15

    Recently the most countries using the nuclear power plants for electricity generation have been faced with the problem of the preparation of the repository for the disposition of the nuclear waste generated from LWR. It was well-known that the issues related with long term risk of the radioactive wastes for the future generations are due only to 1% of the total waste. This small fraction of 1% consists of transuranic (TRU) nuclides such as Pu, Np, Am, Cm and the long lived fission products such as Tc and I. For the transuranic (TRU) nuclides, their half lives range from several years to several hundred thousands years and hence their radioactive toxicity can be lasted over very long time period. This has made the change of the rule of the fast spectrum reactor from the economical use of uranium resource through breeding to the reduction of the nuclear waste through the transmutation. The purpose of this study is to obtain the basic knowledge on the nuclear transmutation technology and to suggest the technical solution ways for the future technology development and enhancement through a survey of the state-of-art of the international research on the nuclear transmutation. The increase of the transmutation rate requires the reduction of the breeding ratio. In fact, the transmutation rate is determined by the breeding ratio. The reduction of the breeding ratio can be achieved by reducing the U-238 content in fuel or increasing the neutron leakage through core boundary or absorbing the neutrons by using some absorbers. However, the reduction of the U-238 content results in the degradation of the fuel Doppler coefficient that is one of the most important safety-related parameters and the reduction of the effective delayed neutron fraction that is related with the controllability of the reactor core. Also, the increase of the transmutation rate can lead to the increase of the coolant void reactivity worth unless some ways to reduce the coolant void reactivity are not

  14. Evaluation of isotopic composition of fast reactor core in closed nuclear fuel cycle

    Science.gov (United States)

    Tikhomirov, Georgy; Ternovykh, Mikhail; Saldikov, Ivan; Fomichenko, Peter; Gerasimov, Alexander

    2017-09-01

    The strategy of the development of nuclear power in Russia provides for use of fast power reactors in closed nuclear fuel cycle. The PRORYV (i.e. «Breakthrough» in Russian) project is currently under development. Within the framework of this project, fast reactors BN-1200 and BREST-OD-300 should be built to, inter alia, demonstrate possibility of the closed nuclear fuel cycle technologies with plutonium as a main source of energy. Russia has a large inventory of plutonium which was accumulated in the result of reprocessing of spent fuel of thermal power reactors and conversion of nuclear weapons. This kind of plutonium will be used for development of initial fuel assemblies for fast reactors. The closed nuclear fuel cycle concept of the PRORYV assumes self-supplied mode of operation with fuel regeneration by neutron capture reaction in non-enriched uranium, which is used as a raw material. Operating modes of reactors and its characteristics should be chosen so as to provide the self-sufficient mode by using of fissile isotopes while refueling by depleted uranium and to support this state during the entire period of reactor operation. Thus, the actual issue is modeling fuel handling processes. To solve these problems, the code REPRORYV (Recycle for PRORYV) has been developed. It simulates nuclide streams in non-reactor stages of the closed fuel cycle. At the same time various verified codes can be used to evaluate in-core characteristics of a reactor. By using this approach various options for nuclide streams and assess the impact of different plutonium content in the fuel, fuel processing conditions, losses during fuel processing, as well as the impact of initial uncertainties on neutron-physical characteristics of reactor are considered in this study.

  15. Studies for a multipurpose research reactor for the CRCN/CNEN-PE

    Energy Technology Data Exchange (ETDEWEB)

    Barroso, Antonio C.O. [Comissao Nacional de Energia Nuclear (CNEN), Rio de Janeiro, RJ (Brazil); Maiorino, Jose R.; Bastos, Jose L.; Silva, Jose E.R. da; Yamaguchi, Mitsuo; Umbehaum, Pedro E.; Ferreira, Carlos R.; Maprelian, E.; Silva, Graciete S. de A.; Yoryiaz, H.; Terremoto, Luiz A.A. [Instituto de Pesquisas Energeticas e Nucleares (IPEN), Sao Paulo, SP (Brazil)]. E-mail: maiorino@net.ipen.br; mitsuo@net.ipen.br; emaprel@net.ipen.br; gsasilva@net.ipen.br; Moreira, Joao M.L. [Centro Tecnologico da Marinha (CTMSP), Sao Paulo, SP (Brazil); Lima, Fernando R. de A.; Lyra, Carlos A.B.O. [Pernambuco Univ., Recife, PE (Brazil). Dept. de Energia Nuclear; Azevedo, Carlos V.G. de; Filho, Jose A.B. [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN), Belo Horizonte, MG (Brazil); Santos, Rubens S. dos [Instituto de Engenharia Nuclear (IEN), Rio de Janeiro, RJ (Brazil)

    1998-07-01

    This paper presents a conceptual proposal for the design and construction of an irradiation research facility at the CRCN (Regional Center of Nuclear Sciences) site, in cooperation with an international partner. The planned irradiation facility is based on a multipurpose research reactor with an innovative design feature, which is a core with two sub-critical parts coupled by a heavy water tank for enhancing and flatten the thermal fluxes, improving safety, and improving beam applications. (author)

  16. Coupled neutronic core and subchannel analysis of nanofluids in VVER-1000 type reactor

    Energy Technology Data Exchange (ETDEWEB)

    Zarifi, Ehsan; Sepanloo, Kamran [Nuclear Science and Technology Research Institute (NSTRI), Tehran (Iran, Islamic Republic of). Reactor and Nuclear Safety School; Jahanfarnia, Golamreza [Islamic Azad Univ., Tehran (Iran, Islamic Republic of). Dept. of Nuclear Engineering, Science and Research Branch

    2017-05-15

    This study is aimed to perform the coupled thermal-hydraulic/neutronic analysis of nanofluids as the coolant in the hot fuel assembly of VVER-1000 reactor core. Water-based nanofluid containing various volume fractions of Al{sub 2}O{sub 3} nanoparticle is analyzed. WIMS and CITATION codes are used for neutronic simulation of the reactor core, calculating neutron flux and thermal power distribution. In the thermal-hydraulic modeling, the porous media approach is used to analyze the thermal behavior of the reactor core and the subchannel analysis is used to calculate the hottest fuel assembly thermal-hydraulic parameters. The derived conservation equations for coolant and conduction heat transfer equation for fuel and clad are discretized by Finite volume method and solved numerically using visual FORTRAN program. Finally the analysis results for nanofluids and pure water are compared together. The achieved results show that at low concentration (0.1 percent volume fraction) alumina is the optimum nanoparticles for normal reactor operation.

  17. Analysis on approach of safeguards implementation at research reactor handling item count and bulk material

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hyun Jo; Lee, Sung Ho; Lee, Byung Doo; Jung, Juang [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    KiJang research reactor (KJRR) will be constructed to produce the radioisotope such as Mo-99 etc., provide the neutron transmutation doping (NTD) service of silicon, and develop the core technologies of research reactor. In this paper, the features of the process and nuclear material flow are reviewed and the material balance area (MBA) and key measurement point (KMP) are established based on the nuclear material flow. Also, this paper reviews the approach on safeguards and nuclear material accountancy at the facility level for Safeguards-by-Design at research reactor handling item count and bulk material. In this paper, MBA and KMPs are established through the analysis on facility features and major process at KJRR handling item count and bulk material. Also, this paper reviews the IAEA safeguards implementation and nuclear material accountancy at KJRR. It is necessary to discuss the safeguards approach on the fresh FM target assemblies and remaining uranium in the intermediate level liquid wastes.

  18. Current status, research progress and future plan of Kartini research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Sardjono, Y.; Syarip; Tjiptono, T.W. [Yogyakarta Nuclear Research Center, Batan (Indonesia)

    1999-10-01

    The current status, research progress and future plan of the Kartini Research Reactor (KRR) is presented. The measurements of axial burn-up distributions for each fuel element by gamma scanning techniques, core axial power distribution display, fuel management for safeguards purpose as well as some research progress activities i.e.; utilization of beamport for: neutron radiography, application neutron activation analysis and history record of KRR power operations is also presented. The KRR is 100 kW pool water reactor type which uses natural circulation and provided by: five beamports in which one of them already coupled with natural uranium subcritical assembly, two thermalizing columns in which one of them is prepared for developing Boron Neutron Capture Therapy (BNCT), two rabbit systems utilized for special analysis uranium ore by delayed neutron counting techniques, one center timbre and 40 irradiation rack (lazy susan) for neutron activation analysis. The KRR was constructed as a second research reactor in Indonesia with special purpose for training and education, high safety margin with involve in high negative temperature coefficient which achieved its first criticality on January 25, 1979. The maximum power level on first criticality is 50 kW and since August 1981 up to now is operating 100 kW. Base on the KRR design limit, it is planned to increase the power level up to 250 kW in the future plan. The preliminary activities such as Non Destructive Testing (NDT) for some reactor components especially water tank and thermal column should be done before decided to increase power level. (author)

  19. Status and future of the WWR-M research reactor in Kiev

    Energy Technology Data Exchange (ETDEWEB)

    Bazavov, D.A.; Gavrilyuk, V.I.; Kirischuk, V.I.; Kochetkov, V.V.; Lysenko, M.V.; Makarovskiy, V.N.; Scherbachenko, A.M.; Shevel, V.N.; Slisenko, V.I. [Institute for Nuclear Research, Kiev (Ukraine)

    2001-07-01

    Kiev WWR-M Research Reactor, operated at maximum power of 10 MW, was put into operation in 1960 and during its 40-years history has been used to perform numerous studies in different areas of science and technology. Due to a number of technical problems the Research Reactor, the only one in Ukraine, was shut down in 1993 and then put into operation in 1999 again. Now there is an intention to reconstruct Kiev Research Reactor. The upgraded Research Reactor would allow solving such problems as the safe operation of Ukrainian NPPs, radioisotope production and, naturally, fundamental and applied research. The main problem for the successful operation of Kiev Research Reactor is the management and storage of spent fuel at the site, since after core unloading the spent fuel storage appears to be practically completed. So it is absolutely necessary to ship the most part of the spent fuel for reprocessing and as soon as possible. Besides, there is a need to build up the new spent fuel storage, because the tank of available storage requires careful inspection for corrosion. (author)

  20. A Metropolis algorithm combined with Nelder-Mead Simplex applied to nuclear reactor core design

    Energy Technology Data Exchange (ETDEWEB)

    Sacco, Wagner F. [Depto. de Modelagem Computacional, Instituto Politecnico, Universidade do Estado do Rio de Janeiro, R. Alberto Rangel, s/n, P.O. Box 972285, Nova Friburgo, RJ 28601-970 (Brazil)], E-mail: wfsacco@iprj.uerj.br; Filho, Hermes Alves; Henderson, Nelio [Depto. de Modelagem Computacional, Instituto Politecnico, Universidade do Estado do Rio de Janeiro, R. Alberto Rangel, s/n, P.O. Box 972285, Nova Friburgo, RJ 28601-970 (Brazil); Oliveira, Cassiano R.E. de [Nuclear and Radiological Engineering Program, George W. Woodruff School of Mechanical Engineering, Georgia Institute of Technology, Atlanta, GA 30332-0405 (United States)

    2008-05-15

    A hybridization of the recently introduced Particle Collision Algorithm (PCA) and the Nelder-Mead Simplex algorithm is introduced and applied to a core design optimization problem which was previously attacked by other metaheuristics. The optimization problem consists in adjusting several reactor cell parameters, such as dimensions, enrichment and materials, in order to minimize the average peak-factor in a three-enrichment-zone reactor, considering restrictions on the average thermal flux, criticality and sub-moderation. The new metaheuristic performs better than the genetic algorithm, particle swarm optimization, and the Metropolis algorithms PCA and the Great Deluge Algorithm, thus demonstrating its potential for other applications.

  1. Fission product release phenomena during core melt accidents in metal fueled heavy water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Ellison, P G; Hyder, M L; Monson, P R; Randolph, H W [Westinghouse Savannah River Co., Aiken, SC (USA); Hagrman, D L [EG and G Idaho, Inc., Idaho Falls, ID (USA); McClure, P R; Leonard, M T [Science Applications International Corp., Albuquerque, NM (USA)

    1990-01-01

    The phenomena that determine fission product release rates from a core melting accident in a metal-fueled, heavy water reactor are described in this paper. This information is obtained from the analysis of the current metal fuel experimental data base and from the results of analytical calculations. Experimental programs in place at the Savannah River Site are described that will provide information to resolve uncertainties in the data base. The results of the experiments will be incorporated into new severe accident computer codes recently developed for this reactor design. 47 refs., 4 figs.

  2. Development of a steady thermal-hydraulic analysis code for the China Advanced Research Reactor

    Institute of Scientific and Technical Information of China (English)

    TIAN Wenxi; QIU Suizheng; GUO Yun; SU Guanghui; JIA Dounan; LIU Tiancai; ZHANG Jianwei

    2007-01-01

    A multi-channel model steady-state thermalhydraulic analysis code was developed for the China Advanced Research Reactor (CARR). By simulating the whole reactor core, the detailed mass flow distribution in the core was obtained. The result shows that structure size plays the most important role in mass flow distribution, and the influence of core power could be neglected under singlephase flow. The temperature field of the fuel element under unsymmetrical cooling condition was also obtained, which is necessary for further study such as stress analysis, etc. Of the fuel element. At the same time, considering the hot channel effect including engineering factor and nuclear factor, calculation of the mean and hot channel was carried out and it is proved that all thermal-hydraulic parameters satisfy the "Safety design regulation of CARR".

  3. Research reactor power controller design using an output feedback nonlinear receding horizon control method

    Energy Technology Data Exchange (ETDEWEB)

    Etchepareborda, Andres [Department of Nuclear Engineering, Argentine National Atomic Energy Commission, Centro Atomico Bariloche, Av. E. Bustillo 9500, Bariloche 8400 (Argentina)]. E-mail: etche@cab.cnea.gov.ar; Lolich, Jose [INVAP S.E., Moreno 1089, Bariloche 8400 (Argentina)

    2007-02-15

    A constrained, output feedback nonlinear receding horizon control (NRHC) method is applied to design a research reactor power controller. The method uses a nonlinear plant model subject to state, control and terminal set constraints; a nonlinear cost function; and a high gain observer. The controller regulates reactor power from 1% to 100% of full power; considers known disturbances, such as reactivity insertions and changes in core inlet flow and temperature; and includes upper limits constraints on neutron flux, neutron flux rate, core outlet temperature and core inlet-outlet temperature difference. Simulation results show an excellent performance for power regulation and known disturbances rejection: all process variables are kept within the admissible limits avoiding the actuation of the safety systems.

  4. 78 FR 56174 - In-Core Thermocouples at Different Elevations and Radial Positions in Reactor Core

    Science.gov (United States)

    2013-09-12

    ... coolant Temperature--Enthalpy (T-H) properties. The coolant steady-state properties (i.e., temperature) do... the rapid zirconium-steam reaction, core exit temperatures were measured at around 800 F. (Leyse-4... some liquefaction of core components because of eutectic reactions (i.e., the eutectic reaction...

  5. Nuclear design of the burst power ultrahigh temperature UF4 vapor core reactor system

    Science.gov (United States)

    Kahook, Samer D.; Dugan, Edward T.

    1991-01-01

    Static and dynamic neutronic analyses are being performed, as part of an integrated series of studies, on an innovative burst power UF4 Ultrahigh Temperature Vapor Core Reactor (UTVR)/Disk Magnetohydrodynamic (MHD) generator for space nuclear power applications. This novel reactor concept operates on a direct, closed Rankine cycle in the burst power mode (hundreds of MWe for thousands of seconds). The fuel/working fluid is a mixture of UF4 and metal fluoride. Preliminary calculations indicate high overall system efficiencies (≊20%), small radiator size (≊5 m2/MWe), and high specific power (≊5 kWe/kg). Neutronic analysis has revealed a number of attractive features for this novel reactor concept. These include some unique and very effective inherent negative reactivity control mechanisms such as the vapor-fuel density power coefficient of reactivity, the direct neutronic coupling among the multiple fissioning core regions (the central vapor core and the surrounding boiler columns), and the mass flow coupling feedback between the fissioning cores.

  6. Advanced Core Design And Fuel Management For Pebble-Bed Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hans D. Gougar; Abderrafi M. Ougouag; William K. Terry

    2004-10-01

    A method for designing and optimizing recirculating pebble-bed reactor cores is presented. At the heart of the method is a new reactor physics computer code, PEBBED, which accurately and efficiently computes the neutronic and material properties of the asymptotic (equilibrium) fuel cycle. This core state is shown to be unique for a given core geometry, power level, discharge burnup, and fuel circulation policy. Fuel circulation in the pebble-bed can be described in terms of a few well?defined parameters and expressed as a recirculation matrix. The implementation of a few heat?transfer relations suitable for high-temperature gas-cooled reactors allows for the rapid estimation of thermal properties critical for safe operation. Thus, modeling and design optimization of a given pebble-bed core can be performed quickly and efficiently via the manipulation of a limited number key parameters. Automation of the optimization process is achieved by manipulation of these parameters using a genetic algorithm. The end result is an economical, passively safe, proliferation-resistant nuclear power plant.

  7. Recent advances on thermohydraulic simulation of HTR-10 nuclear reactor core using realistic CFD approach

    Energy Technology Data Exchange (ETDEWEB)

    Silva, Alexandro S., E-mail: alexandrossilva@ifba.edu.br [Instituto Federal de Educacao, Ciencia e Tecnologia da Bahia (IFBA), Vitoria da Conquista, BA (Brazil); Mazaira, Leorlen Y.R., E-mail: leored1984@gmail.com, E-mail: cgh@instec.cu [Instituto Superior de Tecnologias y Ciencias Aplicadas (INSTEC), La Habana (Cuba); Dominguez, Dany S.; Hernandez, Carlos R.G., E-mail: alexandrossilva@gmail.com, E-mail: dsdominguez@gmail.com [Universidade Estadual de Santa Cruz (UESC), Ilheus, BA (Brazil). Programa de Pos-Graduacao em Modelagem Computacional; Lira, Carlos A.B.O., E-mail: cabol@ufpe.br [Universidade Federal de Pernambuco (UFPE), Recife, PE (Brazil)

    2015-07-01

    High-temperature gas-cooled reactors (HTGRs) have the potential to be used as possible energy generation sources in the near future, owing to their inherently safe performance by using a large amount of graphite, low power density design, and high conversion efficiency. However, safety is the most important issue for its commercialization in nuclear energy industry. It is very important for safety design and operation of an HTGR to investigate its thermal-hydraulic characteristics. In this article, it was performed the thermal-hydraulic simulation of compressible flow inside the core of the pebble bed reactor HTR (High Temperature Reactor)-10 using Computational Fluid Dynamics (CFD). The realistic approach was used, where every closely packed pebble is realistically modelled considering a graphite layer and sphere of fuel. Due to the high computational cost is impossible simulate the full core; therefore, the geometry used is a FCC (Face Centered Cubic) cell with the half height of the core, with 21 layers and 95 pebbles. The input data used were taken from the thermal-hydraulic IAEA Bechmark. The results show the profiles of velocity and temperature of the coolant in the core, and the temperature distribution inside the pebbles. The maximum temperatures in the pebbles do not exceed the allowable limit for this type of nuclear fuel. (author)

  8. Development of Liquid-Vapor Core Reactors with MHD Generator for Space Power and Propulsion Applications

    Energy Technology Data Exchange (ETDEWEB)

    Samim Anghaie

    2002-08-13

    Any reactor that utilizes fuel consisting of a fissile material in a gaseous state may be referred to as a gaseous core reactor (GCR). Studies on GCRs have primarily been limited to the conceptual phase, mostly due to budget cuts and program cancellations in the early 1970's. A few scientific experiments have been conducted on candidate concepts, primarily of static pressure fissile gas filling a cylindrical or spherical cavity surrounded by a moderating shell, such as beryllium, heavy water, or graphite. The main interest in this area of nuclear power generation is for space applications. The interest in space applications has developed due to the promise of significant enhancement in fuel utilization, safety, plant efficiency, special high-performance features, load-following capabilities, power conversion optimization, and other key aspects of nuclear power generation. The design of a successful GCR adapted for use in space is complicated. The fissile material studied in the pa st has been in a fluorine compound, either a tetrafluoride or a hexafluoride. Both of these molecules have an impact on the structural material used in the making of a GCR. Uranium hexafluoride as a fuel allows for a lower operating temperature, but at temperatures greater than 900K becomes essentially impossible to contain. This difficulty with the use of UF6 has caused engineers and scientists to use uranium tetrafluoride, which is a more stable molecule but has the disadvantage of requiring significantly higher operating temperatures. Gas core reactors have traditionally been studied in a steady state configuration. In this manner a fissile gas and working fluid are introduced into the core, called a cavity, that is surrounded by a reflector constructed of materials such as Be or BeO. These reactors have often been described as cavity reactors because the density of the fissile gas is low and criticality is achieved only by means of the reflector to reduce neutron leakage from the

  9. China Advanced Research Reactor Project Progress in 2011

    Institute of Scientific and Technical Information of China (English)

    2011-01-01

    2011, China Advanced Research Reactor (CARR) Project finished the B stage commissioning and resolved the relative technical problems. Meanwhile, the acceptance items and the cold neutron source were carrying out.

  10. Remediation of Site of Decommissioning Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Danilovich, A.S.; Ivanov, O.P.; Lemus, A.V.; Pavlenko, V.I.; Potapov, V.N.; Semenov, S.G.; Shisha, A.D.; Chesnokov, A.V. [National Research Center ' Kurchatov Institute' , 123182, Moscow (Russian Federation)

    2014-07-01

    In the world the most widespread method of soil decontamination consists of removing the contaminated upper layer and sending it for long-term controlled storage. However, implementation of this soil cleanup method for remediation of large contaminated areas would involve high material and financial expenditures, because it produces large amounts of radioactive waste demanding removal to special storage sites. Contaminated soil extraction and cleanup performed right on the spot of remediation activities represents a more advanced and economically acceptable method. Radiological separation of the radioactive soil allows reducing of amount of radwaste. Studies performed during the liquidation of the Chernobyl accident consequences revealed that a considerable fraction of radioactivity is accumulated in minute soil grains. So, the separation of contaminated soil by size fractions makes it possible to extract and concentrate the major share of radioactivity in the fine fraction. Based on these researches water gravity separation technology was proposed by Bochvar Institute. The method extracts the fine fraction from contaminated soil. Studies carried out by Bochvar Institute experts showed that, together with the fine fraction (amounting to 10-20% of the initial soil), this technology can remove up to 85-90% of contaminating radionuclides. The resulting 'dirty' soil fraction could be packaged into containers and removed as radwaste, and decontaminated fractions returned back to their extraction site. Use of radiological and water gravity separations consequently increases the productivity of decontamination facility. Efficiency of this technology applied for contaminated soil cleanup was confirmed in the course of remediation of the contaminated territories near decommissioning research reactor in the Kurchatov Institute. For soil cleaning purposes, a special facility implementing the technology of water gravity separation and radiometric monitoring of soil

  11. Conceptual Study for development of a low power research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Park, C.; Kim, H. S.; Park, J. H.; Chae, H. T.; Lee, B. C. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-07-01

    Even though the nuclear society is again facing with difficult situations after Fukusima accident, some countries still continues to consider nuclear power as one option of national energy sources and to introduce nuclear energy. As a research reactor has been regarded as a step-stone to establish infrastructures for the nuclear power development program, some countries that have plan to introduce the nuclear power energy are considering to construct a research reactor. Particularly, a low power research reactor whose main purpose is basic researches on the nuclear technology and education/training would be of interest to developing countries when taking the economy and level of science and technology into consideration. And many low power research reactors at operation are obsolescent and their numbers are decreasing. Hence, some concepts on a low power research reactor are being studied for the future needs. This paper presents the conceptual study on the basic requirements and the preliminary design features of a low power research reactor.

  12. Renewing Liquid Fueled Molten Salt Reactor Research and Development

    Science.gov (United States)

    Towell, Rusty; NEXT Lab Team

    2016-09-01

    Globally there is a desperate need for affordable, safe, and clean energy on demand. More than anything else, this would raise the living conditions of those in poverty around the world. An advanced reactor that utilizes liquid fuel and molten salts is capable of meeting these needs. Although, this technology was demonstrated in the Molten Salt Reactor Experiment (MSRE) at ORNL in the 60's, little progress has been made since the program was cancelled over 40 years ago. A new research effort has been initiated to advance the technical readiness level of key reactor components. This presentation will explain the motivation and initial steps for this new research initiative.

  13. Thermal hydraulic parametric investigation of decay heat removal from degraded core of a sodium cooled fast Breeder reactor

    Energy Technology Data Exchange (ETDEWEB)

    Verma, Lokesh [Department of Physics and Astrophysics, University of Delhi, Delhi 110007 (India); Kumar Sharma, Anil, E-mail: aksharma@igcar.gov.in [Reactor Design Group, Indira Gandhi Centre for Atomic Research, HBNI, Kalpakkam (India); Velusamy, K. [Reactor Design Group, Indira Gandhi Centre for Atomic Research, HBNI, Kalpakkam (India)

    2017-03-15

    Highlights: • Decay heat removal from degraded core of a typical SFR is highlighted. • Influence of number of DHXs in operation on PAHR is analyzed. • Investigations on structural integrity of the inner vessel and core catcher. • Feasibility study for retention of a part of debris in upper pool of SFR. - Abstract: Ensuring post accident decay heat removal with high degree of reliability following a Core Disruptive Accident (CDA) is very important in the design of sodium cooled fast reactors (SFR). In the recent past, a lot of research has been done towards the design of an in-vessel core catcher below the grid plate to prevent the core debris reaching the main vessel in a pool type SFR. However, during an energetic CDA, the entire core debris is unlikely to reach the core catcher. A significant part of the debris is likely to settle in core periphery between radial shielding subassemblies and the inner vessel. Failure of inner vessel due to the decay heat can lead to core debris reaching the main vessel and threatening its integrity. On the other hand, retention of a part of debris in core periphery can reduce the load on main core catcher. Towards achieving an optimum design of SFR and safety evaluation, it is essential to quantify the amount of heat generating core debris that can be retained safely within the primary vessel. This has been performed by a mathematical simulation comprising solution of 2-D transient form of the governing equations of turbulent sodium flow and heat transfer with Boussinesq approximations. The conjugate conduction-convection model adopted for this purpose is validated against in-house experimental data. Transient evolutions of natural convection in the pools and structural temperatures in critical components have been predicted. It is found that 50% of the core debris can be safely accommodated in the gap between radial shielding subassemblies and inner vessel without exceeding structural temperature limit. It is also

  14. Design and implementation progress of multi-purpose simulator for nuclear research reactor using LabVIEW

    Energy Technology Data Exchange (ETDEWEB)

    Arafa, Amany Abdel Aziz; Saleh, Hassan Ibrahim [Atomic Energy Authority, Cairo (Egypt). Radiation Engineering Dept.; Ashoub, Nagieb [Atomic Energy Authority, Cairo (Egypt). Nuclear Research Center

    2015-11-15

    This paper illustrates the neutronic and thermal hydraulic models that were implemented in the nuclear research reactor simulator based on LabVIEW. It also describes the system and transient analysis of the simulator that takes into consideration the temperature effects and poisoning. This simulator is designed to be a multi-purpose in which the operator could understand the effects of the input parameters on the reactor. A designer can study different solutions for virtual reactor accident scenarios. The main features of the simulator are the flexibility to design and maintain the interface and the ability to redesign and remodel the reactor core engine. The developed reactor simulator permits to acquire hands-on the experience of the physics and technology of nuclear reactors including reactivity control, thermodynamics, technology design and safety system design. This simulator can be easily customizable and upgradable and new opportunities for collaboration between academic groups could be conducted.

  15. Test problem for thermal-hydraulics and neutronic coupled calculation fore ALFREAD reactor core

    Science.gov (United States)

    Filip, A.; Darie, G.; Saldikov, I. S.; Smirnov, A. D.; Tikhomirov, G. V.

    2017-01-01

    The beginning of a new era of nuclear reactor requires technological advances and also multiples studies. The European Liquid metal cooled Fast breeder Reactor is one of the designs for the generation IV nuclear reactor, selected by ENEA. A pioneer of its time, ELFR needs a demonstrator in order to prove the feasibility of this project and to acquire more data and experience in operating a LFR. For this reason the ALFRED project was started and it is expected to be under operation by the year 2030. This paper has the objective of analyzing the neutronic and thermohydraulics of the ALFRED core by the means of a coupled scheme. The selected code for neutronic simulation is MCNP and the selected code for thermohydraulics is ANSYS.

  16. Dosimetry assessments for the reactor pressure vessel and core barrel in UK PWR plant

    Energy Technology Data Exchange (ETDEWEB)

    Thornton, D.A.; Allen, D.A.; Huggon, A.P.; Picton, D.J.; Robinson, A.T.; Steadman, R.J. [Serco, Rutherford House, Quedgeley, Gloucester, Gl2 4NF (United Kingdom); Seren, T.; Lipponen, M.; Kekki, T. [VTT, Technical Research Centre of Finland, Otakaari 3 K, P.O. BOX 1000, Espoo, FI-02044 (Finland)

    2011-07-01

    Specimens for the Sizewell B reactor pressure vessel (RPV) inservice steels surveillance program are irradiated inside eight capsules located within the reactor pressure vessel and loaded prior to commissioning. The periodic removal of these capsules and testing of their contents provides material properties data at intervals during the lifetime of the plant. Neutron activation measurements and radiation transport calculations play an essential role in assessing the neutron exposure of the specimens and RPV. Following the most recent withdrawal, seven capsules have now been removed covering nine cycles of reactor operation. This paper summarizes the dosimetry results of the Sizewell B surveillance program obtained to date. In addition to an overview of the calculational methodology it includes a review of the measurements. Finally, it describes an extension of the methodology to provide dosimetry recommendations for the core barrel and briefly discusses the results that were obtained. (authors)

  17. Development of an educational nuclear research reactor simulator

    Energy Technology Data Exchange (ETDEWEB)

    Arafa, Amany Abdel Aziz; Saleh, Hassan Ibrahim [Atomic Energy Authority, Cairo (Egypt). Radiation Engineering Dept.; Ashoub, Nagieb [Atomic Energy Authority, Cairo (Egypt). Reactor Physics Dept.

    2014-12-15

    This paper introduces the development of a research reactor educational simulator based on LabVIEW that allows the training of operators and studying different accident scenarios and the effects of operational parameters on the reactor behavior. Using this simulator, the trainee can test the interaction between the input parameters and the reactor activities. The LabVIEW acts as an engine implements the reactor mathematical models. In addition, it is used as a tool for implementing the animated graphical user interface. This simulator provides the training requirements for both of the reactor staff and the nuclear engineering students. Therefore, it uses dynamic animation to enhance learning and interest for a trainee on real system problems and provides better visual effects, improved communications, and higher interest levels. The benefits of conducting such projects are to develop the expertise in this field and save costs of both operators training and simulation courses.

  18. CFD analysis of PWR core top and reactor vessel upper plenum internal subdomain models

    Energy Technology Data Exchange (ETDEWEB)

    Kao, Min-Tsung; Wu, Chung-Yun [National Tsing Hua University, Hsinchu 30043, Taiwan (China); Chieng, Ching-Chang, E-mail: cchieng@ess.nthu.edu.tw [National Tsing Hua University, Hsinchu 30043, Taiwan (China); Xu Yiban; Yuan Kun; Dzodzo, Milorad; Conner, Michael; Beltz, Steven; Ray, Sumit; Bissett, Teresa [Westinghouse Electric Company, Cranberry Township, PA 16066 (United States)

    2011-10-15

    Highlights: > The paper develops a CFD flow model for upper portion of AP1000 and determines how lateral flow in the top core and upper plenum. > Mesh sensitivities and geometrical modification strategies give the guidelines to reduce the size of overall computation mesh. > Pressure drop measurement data act as a guideline for the mesh selection. > Lateral flows are mainly exiting through upper and lower windows of guide tubes ({approx}81%) and 18% flow through small side gaps. > The interactions between guide tubes and neighboring support column as well as flow characteristic are revealed. - Abstract: One aspect of the Westinghouse AP1000{sup TM} reactor design is the reduction in the number of major components and simplification in manufacturing. One design change relative to current Westinghouse reactors of similar size is that AP1000 reactor vessel has two nozzles/hot legs instead of three. With regard to fuel performance, this design difference creates a different flow field in the reactor vessel upper plenum. The flow exiting from the core and entering the upper plenum must turn toward one of the two outlet nozzles and flow laterally around numerous control rod guide tubes and support columns. Also, below the upper plenum are the upper core plate and the top core region of the 157 fuel assemblies and 69 guidetube assemblies. To determine how the lateral flow in the top of the core and upper plenum compares to the current reactors a CFD model of the flow in the upper portion of the AP1000 reactor vessel was created. Before detailed CFD simulations of the flow in the entire upper plenum and top core regions were performed, conducting local simulations for smaller sections of the domain provided crucial and detailed physical aspects of the flow. These sub-domain models were used to perform mesh sensitivities and to assess what geometrical details may be eliminated from the larger model in order to reduce mesh size and computational requirements. In this paper

  19. Status of reduced enrichment programs for research reactors in Japan

    Energy Technology Data Exchange (ETDEWEB)

    Kanda, Keiji; Nishihara, Hedeaki [Kyoto Univ., Osaka (Japan); Shirai, Eiji; Oyamada, Rokuro; Sanokawa, Konomo [Japan Atomic Energy Research Institute, Tokyo (Japan)

    1997-08-01

    The reduced enrichment programs for the JRR-2, JRR-3, JRR-4 and JMTR of Japan Atomic Energy Research Institute (JAERI), and the KUR of Kyoto University Research Reactor Institute (KURRI) have been partially completed and are mostly still in progress under the Joint Study Programs with Argonne National Laboratory (ANL). The JMTR and JRR-2 have been already converted to use MEU aluminide fuels in 1986 and 1987, respectively. The operation of the upgraded JRR-3(JRR-3M) has started in March 1990 with the LEU aluminide fuels. Since May 1992, the two elements have been inserted in the KUR. The safety review application for the full core conversion to use LEU silicide in the JMTR was approved in February 1992 and the conversion has been done in January 1994. The Japanese Government approved a cancellation of the KUHFR Project in February 1991, and in April 1994 the U.S. Government gave an approval to utilize HEU in the KUR instead of the KUHFR. Therefore, the KUR will be operated with HEU fuel until 2001. Since March 1994, Kyoto University is continuing negotiation with UKAEA Dounreay on spent fuel reprocessing and blending down of recovered uranium, in addition to that with USDOE.

  20. Initial verification and validation of RAZORBACK - A research reactor transient analysis code

    Energy Technology Data Exchange (ETDEWEB)

    Talley, Darren G. [Sandia National Laboratories (SNL-NM), Albuquerque, NM (United States)

    2015-09-01

    This report describes the work and results of the initial verification and validation (V&V) of the beta release of the Razorback code. Razorback is a computer code designed to simulate the operation of a research reactor (such as the Annular Core Research Reactor (ACRR)) by a coupled numerical solution of the point reactor kinetics equations, the energy conservation equation for fuel element heat transfer, and the mass, momentum, and energy conservation equations for the water cooling of the fuel elements. This initial V&V effort was intended to confirm that the code work to-date shows good agreement between simulation and actual ACRR operations, indicating that the subsequent V&V effort for the official release of the code will be successful.

  1. Use of research and test reactors for SPD development and calibration

    Energy Technology Data Exchange (ETDEWEB)

    LaFontaine, M.W.R. [Physics Solutions Inc., Kitchener, ON (Canada)

    2011-07-01

    Prior to using a research or test reactor for performance studies or calibration of self powered detectors, it is first necessary to fully characterize the reactor environment in the region to be utilized. This presentation details Characterization Experiments performed to quantify research/test reactor core/site parameters as they would apply for use with SPD applications. Methods will be described to: Determine the Westcott parameter, r (T {sub n}/T {sub o}) , for the region of interest; Characterize the neutron energy spectrum in terms of the cadmium absorption cut-off, i.e., consider neutrons of energy 5kT < 0.13 eV to be thermal neutrons, and neutrons of energy 5kT > 0.13 eV to be epithermal neutrons; Determine T {sub n}, the effective neutron temperature, in the region of interest; Determine the gamma flux in the region of interest; and, Establish SPD calibration standard detectors.

  2. Impact of beryllium reflector ageing on Safari–1 reactor core parameters / L.E. Moloko

    OpenAIRE

    Moloko, Lesego Ernest

    2011-01-01

    The build–up of 6Li and 3He, that is, the strong thermal neutron absorbers or the so called "neutron poisons", in the beryllium reflector changes the physical characteristics of the reactor, such as reactivity, neutron spectra, neutron flux level, power distribution, etc.; furthermore,gaseous isotopes such as 3H and 4He induce swelling and embrittlement of the reflector. The SAFARI–1 research reactor, operated by Necsa at Pelindaba in South Africa, uses a beryllium reflector on...

  3. Determination of the optimal positions for installing gamma ray detection systems at Tehran Research Reactor

    Science.gov (United States)

    Sayyah, A.; Rahmani, F.; Khalafi, H.

    2015-09-01

    Dosimetric instruments must constantly monitor radiation dose levels in different areas of nuclear reactor. Tehran Research Reactor (TRR) has seven beam tubes for different research purposes. All the beam tubes extend from the reactor core to Beam Port Floor (BPF) of the reactor facility. During the reactor operation, the gamma rays exiting from each beam tube outlet produce a specific gamma dose rate field in the space of the BPF. To effectively monitor the gamma dose rates on the BPF, gamma ray detection systems must be installed in optimal positions. The selection of optimal positions is a compromise between two requirements. First, the installation positions must possess largest gamma dose rates and second, gamma ray detectors must not be saturated in these positions. In this study, calculations and experimental measurements have been carried out to identify the optimal positions of the gamma ray detection systems. Eight three dimensional models of the reactor core and related facilities corresponding to eight scenarios have been simulated using MCNPX Monte Carlo code to calculate the gamma dose equivalent rate field in the space of the BPF. These facilities are beam tubes, thermal column, pool, BPF space filled with air, facilities such as neutron radiography facility, neutron powder diffraction facility embedded in the beam tubes as well as biological shields inserted into the unused beam tubes. According to the analysis results of the combined gamma dose rate field, three positions on the north side and two positions on the south side of the BPF have been recognized as optimal positions for installing the gamma ray detection systems. To ensure the consistency of the simulation data, experimental measurements were conducted using TLDs (600 and 700) pairs during the reactor operation at 4.5 MW.

  4. Examination of offsite radiological emergency measures for nuclear reactor accidents involving core melt. [PWR

    Energy Technology Data Exchange (ETDEWEB)

    Aldrich, D.C.; McGrath, P.E.; Rasmussen, N.C.

    1978-06-01

    Evacuation, sheltering followed by population relocation, and iodine prophylaxis are evaluated as offsite public protective measures in response to nuclear reactor accidents involving core-melt. Evaluations were conducted using a modified version of the Reactor Safety Study consequence model. Models representing each measure were developed and are discussed. Potential PWR core-melt radioactive material releases are separated into two categories, ''Melt-through'' and ''Atmospheric,'' based upon the mode of containment failure. Protective measures are examined and compared for each category in terms of projected doses to the whole body and thyroid. Measures for ''Atmospheric'' accidents are also examined in terms of their influence on the occurrence of public health effects.

  5. Review of core disruptive accident analysis for liquid-metal cooled fast reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Y. C.; Na, B. C.; Hahn, D. H

    1997-04-01

    Analysis methodologies of core disruptive accidents (CDAs) are reviewed. The role of CDAS in the overall safety evaluation of fast reactors has not always been well defined nor universally agreed upon. However, they have become a traditional issue in LMR safety, design, and licensing. The study is for the understanding of fast reactor behavior under CDA conditions to establish the consequences of such conditions and to provide a basis for evaluating consequence limiting design features for the KALIMER developments. The methods used to analyze CDAs from initiating event to complete core disruption are described. Two examples of CDA analyses for CRBRP and ALMR are given and R and D needed for better understanding of CDA phenomena are proposed. (author). 10 refs., 2 tabs., 3 figs

  6. Thermal-hydraulic analysis techniques for axisymmetric pebble bed nuclear reactor cores. [PEBBLE code

    Energy Technology Data Exchange (ETDEWEB)

    Stroh, K.R.

    1979-03-01

    The pebble bed reactor's cylindrical core volume contains a random bed of small, spherical fuel-moderator elements. These graphite spheres, containing a central region of dispersed coated-particle fissile and fertile material, are cooled by high pressure helium flowing through the connected interstitial voids. A mathematical model and numerical solution technique have been developed which allow calculation of macroscopic values of thermal-hydraulic variables in an axisymmetric pebble bed nuclear reactor core. The computer program PEBBLE is based on a mathematical model which treats the bed macroscopically as a generating, conducting porous medium. The steady-state model uses a nonlinear Forchheimer-type relation between the coolant pressure gradient and mass flux, with newly derived coefficients for the linear and quadratic resistance terms. The remaining equations in the model make use of mass continuity, and thermal energy balances for the solid and fluid phases.

  7. Experimental study of flow inversion in MTR upward flow research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Abdel-Hadi, Ead A. [Benha Univ., Cairo (Egypt). Shobra Faculty of Engineering; Khedr, Ahmed; Talha, Kamal Eldin Aly; Abdel-Latif, Salwa Helmy

    2014-06-15

    The core cooling of upward flow MTR pool type Research Reactor (RR) at the later stage of pump coast down is experimentally handled to clarify the effect of some operating parameters on RR core cooling. Therefore, a test rig is designed and built to simulate the core cooling loop at this stage. The core is simulated as two vertical channels, electrically heated, and extended between upper and lower plenums. Two elevated tanks filled with water are connected to the two plenums. The first one constitutes a left branch, connected to the lower plenum, and is electrically heated to simulate the core return pipe. The second one constitutes the right branch, connected to the upper plenum, and is cooled by refrigerant circuit to simulate the reactor pool. Channel coolant and wall temperatures at different power and branch temperatures are measured, registered and analyzed. The results show that at this stage of core cooling two cooling loops are established; an internal circulation loop between the channels dominated by the difference in channel's power and an external circulation loop between the branches dominated by the temperature difference between branches. Also, there is a double inversion in core flow, upward-downward-upward flow. This double inversion increases largely the channel's wall temperature. Complementary safety analysis to evaluate this phenomenon must be performed. (orig.)

  8. Oak Ridge National Laboratory Research Reactor Experimenters' Guide

    Energy Technology Data Exchange (ETDEWEB)

    Cagle, C.D. (comp.)

    1982-04-01

    The Oak Ridge National Laboratory (ORNL) operates six research reactors dedicated to research and development work as well as radioisotope production. These reactors are used by ORNL and qualified non-ORNL research and development groups. The purpose of this report is to provide information to research personnel concerning the facilities and the ORNL research and services groups available to assist in the design, fabrication, operation, and post-operation examination of irradiation assemblies. Safety and operability reviews and quality assurance requirements are also described.

  9. Computer simulation of Angra-2 PWR nuclear reactor core using MCNPX code

    Energy Technology Data Exchange (ETDEWEB)

    Medeiros, Marcos P.C. de; Rebello, Wilson F., E-mail: eng.cavaliere@ime.eb.br, E-mail: rebello@ime.eb.br [Instituto Militar de Engenharia - Secao de Engenharia Nuclear, Rio de Janeiro, RJ (Brazil); Oliveira, Claudio L. [Universidade Gama Filho, Departamento de Matematica, Rio de Janeiro, RJ (Brazil); Vellozo, Sergio O., E-mail: vellozo@cbpf.br [Centro Tecnologico do Exercito. Divisao de Defesa Quimica, Biologica e Nuclear, Rio de Janeiro, RJ (Brazil); Silva, Ademir X. da, E-mail: ademir@nuclear.ufrj.br [Coordenacao dos Programas de Pos Gaduacao de Engenharia (COPPE/UFRJ), Rio de Janeiro, RJ (Brazil)

    2011-07-01

    In this work the MCNPX (Monte Carlo N-Particle Transport Code) code was used to develop a computerized model of the core of Angra 2 PWR (Pressurized Water Reactor) nuclear reactor. The model was created without any kind of homogenization, but using real geometric information and material composition of that reactor, obtained from the FSAR (Final Safety Analysis Report). The model is still being improved and the version presented in this work is validated by comparing values calculated by MCNPX with results calculated by others means and presented on FSAR. This paper shows the results already obtained to K{sub eff} and K{infinity}, general parameters of the core, considering the reactor operating under stationary conditions of initial testing and operation. Other stationary operation conditions have been simulated and, in all tested cases, there was a close agreement between values calculated computationally through this model and data presented on the FSAR, which were obtained by other codes. This model is expected to become a valuable tool for many future applications. (author)

  10. Analyzing the thermionic reactor critical experiments. [thermal spectrum of uranium 235 core

    Science.gov (United States)

    Niederauer, G. F.

    1973-01-01

    The Thermionic Reactor Critical Experiments (TRCE) consisted of fast spectrum highly enriched U-235 cores reflected by different thicknesses of beryllium or beryllium oxide with a transition zone of stainless steel between the core and reflector. The mixed fast-thermal spectrum at the core reflector interface region poses a difficult neutron transport calculation. Calculations of TRCE using ENDF/B fast spectrum data and GATHER library thermal spectrum data agreed within about 1 percent for the multiplication factor and within 6 to 8 percent for the power peaks. Use of GAM library fast spectrum data yielded larger deviations. The results were obtained from DOT R Theta calculations with leakage cross sections, by region and by group, extracted from DOT RZ calculations. Delineation of the power peaks required extraordinarily fine mesh size at the core reflector interface.

  11. A Development of Technical Specification of a Research Reactor with Plate Fuels Cooled by Upward Flow

    Energy Technology Data Exchange (ETDEWEB)

    Park, Sujin; Kim, Jeongeun; Kim, Hyeonil [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    The contents of the TS(Technical Specifications) are definitions, safety limits, limiting safety system settings, limiting conditions for operation, surveillance requirements, design features, and administrative controls. TS for Nuclear Power Plants (NPPs) have been developed since many years until now. On the other hands, there are no applicable modernized references of TS for research reactors with many differences from NPPs in purpose and characteristics. Fuel temperature and Departure from Nuclear Boiling Ratio (DNBR) are being used as references from the thermal-hydraulic analysis point of view for determining whether the design of research reactors satisfies acceptance criteria for the nuclear safety or not. Especially for research reactors using plate-type fuels, fuel temperature and critical heat flux, however, are very difficult to measure during the reactor operation. This paper described the outline of main contents of a TS for open-pool research reactor with plate-type fuels using core cooling through passive systems, where acceptance criteria for nuclear safety such as CHF and fuel temperature cannot be directly measured, different from circumstances in NPPs. Thus, three independent variables instead of non-measurable acceptance criteria: fuel temperature and CHF are considered as safety limits, i.e., power, flow, and flow temperature.

  12. Comments on ``large enhancement of TLD-100 sensitivity by irradiation in a reactor core''

    Science.gov (United States)

    Lakshmanan, A. R.; Chandra, Bhuwan; Bhatt, R. C.

    1987-06-01

    The large enhancement of TLD-100 sensitivity on irradiation in a reactor core reported by Lau et al. [Nucl. Instr. and Meth. B17 (1986) 170] is false and is in complete contradiction with the results reported earlier in the literature and with our recent findings. Lau et al. have misinterpreted the TL signal from thermal neutron induced 3H betas in LiF as due to enhanced TL sensitivity because of neutron induced traps/luminescent centres.

  13. Fusion-power-core design of a Compact Reversed-Field Pinch Reactor (CRFPR)

    Science.gov (United States)

    Copenhaver, C.; Schnurr, N. M.; Krakowski, R. A.; Hagenson, R. L.; Mynard, R. C.; Cappiello, C.; Lujan, R. E.; Davidson, J. W.; Chaffee, A. D.; Battat, M. E.

    A conceptual design of a fusion power core (FPC, i.e., plasma chamber, first wall, blanket, shield, coils) based on a Reversed-Field Pinch (RFP) has been completed. After a brief statement of rationale and description of the reactor configuraton, the FPC integration is described in terms of power balance, thermal-hydraulics, and mechanical design. The engineering versatility, promise, and problems of this high-power-density approach to fusion are addressed.

  14. An algorithm for multi-group two-dimensional neutron diffusion kinetics in nuclear reactor cores

    OpenAIRE

    Marcelo Schramm

    2016-01-01

    The objective of this thesis is to introduce a new methodology for two{dimensional multi{ group neutron diffusion kinetics in a reactor core. The presented methodology uses a polyno- mial approximation in a rectangular homogeneous domain with non{homogeneous boundary conditions. As it consists on a truncated Taylor series, its error estimates varies with the size of the rectangle. The coefficients are obtained mainly by their relations with the independent term, which is determined by the dif...

  15. Effects of neutronics characteristics for a generic gas core reactor when selected parameters are changed

    Energy Technology Data Exchange (ETDEWEB)

    Oliveira, Claudio Luiz de [Instituto Militar de Engenharia (IME), Rio de Janeiro, RJ (Brazil). Dept. de Engenharia Nuclear

    2000-07-01

    The Battelle Revised Thermos code, that solves the integral neutron transport equation is used to perform the analysis of a chosen system, which includes materials and number densities that are typically encountered in gaseous core reactors, where selected parameters as pressure (or number density, if the gas temperature is kept constant) of the hydrogen; and temperature and material of the external moderator are changed. (author)

  16. Feasibility study of boiling water reactor core based on thorium-uranium fuel concept

    Energy Technology Data Exchange (ETDEWEB)

    Nunez-Carrera, Alejandro [Comision Nacional de Seguridad Nuclear y Salvaguardias, Dr. Barragan 779, Col Narvarte, 03020 Mexico D.F. (Mexico); Francois Lacouture, Juan Luis; Martin del Campo, Cecilia [Universidad Nacional Autonoma de Mexico, Facultad de Ingenieria, Paseo Cuauhnahuac 8532, Jiutepec, Mor. (Mexico); Espinosa-Paredes, Gilberto [Area de Ingenieria en Recursos Energeticos, Universidad Autonoma Metropolitana Iztapalapa, Apartado Postal 55-534, Mexico D.F. 09340 (Mexico)], E-mail: gepe@xanum.uam.mx

    2008-01-15

    The design of a boiling water reactor (BWR) equilibrium core using the thorium-uranium (blanket-seed) concept in the same integrated fuel assembly is presented in this paper. The lattice design uses the thorium conversion capability to {sup 233}U in a BWR spectrum. A core design was developed to achieve an equilibrium cycle of one effective full power year in a standard BWR with a reload of 104 fuel assemblies designed with an average {sup 235}U enrichment of 7.5 w/o in the seed sub-lattice. The main core operating parameters were obtained. It was observed that the analyzed parameters behave like those obtained in a standard BWR. The economic analysis shows that the fuel cycle cost of the proposed core design can be competitive with a standard uranium core design. Finally, a comparison of the toxicity of the spent fuel showed that the toxicity is lower in the thorium cycle than in other fuel cycles (UO{sub 2} and MOX uranium and plutonium) in the case of the once through cycle for light water reactors (LWR)

  17. A nuclear reactor core fuel reload optimization using artificial ant colony connective networks

    Energy Technology Data Exchange (ETDEWEB)

    Lima, Alan M.M. de [Universidade Federal do Rio de Janeiro, PEN/COPPE - UFRJ, Ilha do Fundao s/n, CEP 21945-970 Rio de Janeiro (Brazil)], E-mail: alanmmlima@yahoo.com.br; Schirru, Roberto [Universidade Federal do Rio de Janeiro, PEN/COPPE - UFRJ, Ilha do Fundao s/n, CEP 21945-970 Rio de Janeiro (Brazil)], E-mail: schirru@lmp.ufrj.br; Carvalho da Silva, Fernando [Universidade Federal do Rio de Janeiro, PEN/COPPE - UFRJ, Ilha do Fundao s/n, CEP 21945-970 Rio de Janeiro (Brazil)], E-mail: fernando@con.ufrj.br; Medeiros, Jose Antonio Carlos Canedo [Universidade Federal do Rio de Janeiro, PEN/COPPE - UFRJ, Ilha do Fundao s/n, CEP 21945-970 Rio de Janeiro (Brazil)], E-mail: canedo@lmp.ufrj.br

    2008-09-15

    The core of a nuclear Pressurized Water Reactor (PWR) may be reloaded every time the fuel burn-up is such that it is not more possible to maintain the reactor operating at nominal power. The nuclear core fuel reload optimization problem consists in finding a pattern of burned-up and fresh-fuel assemblies that maximize the number of full operational days. This is an NP-Hard problem, meaning that complexity grows exponentially with the number of fuel assemblies in the core. Moreover, the problem is non-linear and its search space is highly discontinuous and multi-modal. Ant Colony System (ACS) is an optimization algorithm based on artificial ants that uses the reinforcement learning technique. The ACS was originally developed to solve the Traveling Salesman Problem (TSP), which is conceptually similar to the nuclear core fuel reload problem. In this work a parallel computational system based on the ACS, called Artificial Ant Colony Networks is introduced to solve the core fuel reload optimization problem.

  18. Improvement of Core Performance by Introduction of Moderators in a Blanket Region of Fast Reactors

    Directory of Open Access Journals (Sweden)

    Toshio Wakabayashi

    2013-01-01

    Full Text Available An application of deuteride moderator for fast reactor cores is proposed for power flattening that can mitigate thermal spikes and alleviate the decrease in breeding ratio, which sometimes occurs when hydrogen moderator is applied as a moderator. Zirconium deuteride is employed in a form of pin arrays at the inner most rows of radial blanket fuel assemblies, which works as a reflector in order to flatten the radial power distribution in the outer core region of MONJU. The power flattening can be utilized to increase core average burn-up by increasing operational time. The core characteristics have been evaluated with a continuous-energy model Monte Carlo code MVP and the JENDL-3.3 cross-section library. The result indicates that the discharged fuel burn-up can be increased by about 7% relative to that of no moderator in the blanket region due to the power flattening when the number of deuteride moderator pins is 61. The core characteristics and core safety such as void reactivity, Doppler coefficient, and reactivity insertion that occurred at dissolution of deuteron were evaluated. It was clear that the serious drawback did not appear from the viewpoints of the core characteristics and core safety.

  19. Assessment of CANDU reactor physics effects using a simplified whole-core MCNP model

    Energy Technology Data Exchange (ETDEWEB)

    Kozier, K.S

    2002-07-01

    A whole-core Monte Carlo n-particle (MCNP) model of a simplified CANDU reactor was developed and used to study core configurations and reactor physics phenomena of interest in CANDU safety analysis. The resulting reactivity data were compared with values derived from corresponding WIMS-AECL/RFSP, two-neutron-energy-group diffusion theory core simulations, thereby extending the range of CANDU-related code-to-code benchmark comparisons to include whole-core representations. These comparisons show a systematic discrepancy of about 6 mk between the respective absolute k{sub eff} values, but very good agreement to within about -0.15 {+-} 0.06 mk for the reactivity perturbation induced by G-core checkerboard coolant voiding. These findings are generally consistent with the results of much simpler uniform-lattice comparisons involving only WIMS-AECL and MCNP. In addition, MCNP fission-energy tallies were used to evaluate other core-wide properties, such as fuel bundle and total-channel power distributions, as well as intra-bundle details, such as outer-fuel-ring relative power densities and outer-ring fuel element azimuthal power variations, which cannot be determined directly from WIMS-AECL/RFSP core calculations. The average MCNP values for the ratio of outer fuel element to average fuel element power density agreed well with corresponding values derived from WIMS-AECL lattice-cell cases, showing a small systematic discrepancy of about 0.5 %, independent of fuel bum-up. For fuel bundles containing the highest-power fuel elements, the maximum peak-to-average outer-element azimuthal power variation was about 2.5% for cases where a statistically significant trend was observed, while much larger peak-to-average outer-element azimuthal power variations of up to around 42% were observed in low-power fuel bundles at the core/radial-neutron-reflector interface. (author)

  20. Analysis of three-dimensional thermo-hydraulic phenomena in the reactor core of LMFBR

    Energy Technology Data Exchange (ETDEWEB)

    Hu, S.; Lee, Y. B.; Jang, W. P.; Ha, K. S.; Jung, H. Y. [KAERI, Taejon (Korea, Republic of)

    2004-07-01

    The mismatch between power and flow under the transient condition of LMFBR (Liquid Metal cooled Fast Breeder Reactor) core results in thermal stratification in hot pool. Since the fluid of hot pool enters IHXs, the temperature distribution of hot pool can alter the overall system response, therefore three-dimensional analysis of thermo-hydraulic phenomena is necessary. In this study, the thermo-hydraulic phenomena under normal operating condition and unprotected transient condition of LMFBR is investigated using which is the three-dimensional analysis code, COMMIX-1AR/P. The basic input data is based on the design data of KALIMER-600, which is sodium-cooled fast breeder reactor developed by KAERI. COMMIX-1AR/P code has not a reactivity model and the power and core flowrate must be supplied in the input data. In this study, results of SSC-K calculation is used. The temperature and velocity distributions are calculated and compared with those of SSC-K calculation results. The UTOF(Unprotected Loss Of Flow) accident is calculated using COMMIX-1AR/P and the temperature and velocity distributions in the total reactor core are calculated and the natural circulation mode under this transient condition is investigated.

  1. Optimization of a heterogeneous fast breeder reactor core with improved behavior during unprotected transients

    Energy Technology Data Exchange (ETDEWEB)

    Poumerouly, S.; Schmitt, D.; Massara, S.; Maliverney, B. [EDF R and D, 1 avenue du general de Gaulle, 92140 Clamart (France)

    2012-07-01

    Innovative Sodium-cooled Fast Reactors (SFRs) are currently being investigated by CEA, AREVA and EDF in the framework of a joint French collaboration, and the construction of a GEN IV prototype, ASTRID (Advanced Sodium Technical Reactor for Industrial Demonstration), is scheduled in the years 2020. Significant improvements are expected so as to improve the reactor safety: the goal is to achieve a robust safety demonstration of the mastering of the consequences of a Core Disruptive Accident (CDA), whether by means of prevention or mitigation features. In this framework, an innovative design was proposed by CEA in 2010. It aims at strongly reducing the sodium void effect, thereby improving the core behavior during unprotected loss of coolant transients. This design is strongly heterogeneous and includes, amongst others, a fertile plate, a sodium plenum associated with a B{sub 4}C upper blanket and a stepwise modulation of the fissile height of the core (onwards referred to as the 'diabolo shape'). In this paper, studies which were entirely carried out at EDF are presented: the full potential of this heterogeneous concept is thoroughly investigated using the SDDS methodology. (authors)

  2. Activities for extending the lifetime of MINT research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Bokhari, Adnan; Kassim, Mohammad Suhaimi [Malaysian Inst. for Nuclear Technology Research (MINT), Bangi, Kajang (Malaysia)

    1998-10-01

    MINT TRIGA Reactor is a 1-MW swimming pool nuclear reactor commissioned in June 1982. Since then, it has been used for research, isotope production, neutron activation, neutron radiography and manpower training. The total operating time till the end on September 1997 is 16968 hours with cumulative total energy release of 11188 MW-hours. After more than fifteen years of successful operation, some deterioration in components and associated systems has been observed. This paper describes some of the activities carried out to increase the lifetime and to reduce the shutdown time of the reactor. (author)

  3. Neutron beam facilities at the Australian Replacement Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kennedy, Shane; Robinson, Robert; Hunter, Brett [Physics Division, ANSTO, Lucas Heights (Australia)

    2001-03-01

    Australia is building a research reactor to replace the HIFAR reactor at Lucas Heights by the end of 2005. Like HIFAR, the Replacement Research Reactor will be multipurpose with capabilities for both neutron beam research and radioisotope production. It will be a pool-type reactor with thermal neutron flux (unperturbed) of 4 x 10{sup 14} n/cm{sup 2}/sec and a liquid D{sub 2} cold neutron source. Cold and thermal neutron beams for neutron beam research will be provided at the reactor face and in a large neutron guide hall. Supermirror neutron guides will transport cold and thermal neutrons to the guide hall. The reactor and the associated infrastructure, with the exception of the neutron beam instruments, is to be built by INVAP S.E. under contract. The neutron beam instruments will be developed by ANSTO, in consultation with the Australian user community. This status report includes a review the planned scientific capabilities, a description of the facility and a summary of progress to date. (author)

  4. Background Radiation Measurements at High Power Research Reactors

    CERN Document Server

    Ashenfelter, J; Baldenegro, C X; Band, H R; Barclay, G; Bass, C D; Berish, D; Bowden, N S; Bryan, C D; Cherwinka, J J; Chu, R; Classen, T; Davee, D; Dean, D; Deichert, G; Dolinski, M J; Dolph, J; Dwyer, D A; Fan, S; Gaison, J K; Galindo-Uribarri, A; Gilje, K; Glenn, A; Green, M; Han, K; Hans, S; Heeger, K M; Heffron, B; Jaffe, D E; Kettell, S; Langford, T J; Littlejohn, B R; Martinez, D; McKeown, R D; Morrell, S; Mueller, P E; Mumm, H P; Napolitano, J; Norcini, D; Pushin, D; Romero, E; Rosero, R; Saldana, L; Seilhan, B S; Sharma, R; Stemen, N T; Surukuchi, P T; Thompson, S J; Varner, R L; Wang, W; Watson, S M; White, B; White, C; Wilhelmi, J; Williams, C; Wise, T; Yao, H; Yeh, M; Yen, Y -R; Zhang, C; Zhang, X

    2016-01-01

    Research reactors host a wide range of activities that make use of the intense neutron fluxes generated at these facilities. Recent interest in performing measurements with relatively low event rates, e.g. reactor antineutrino detection, at these facilities necessitates a detailed understanding of background radiation fields. Both reactor-correlated and naturally occurring background sources are potentially important, even at levels well below those of importance for typical activities. Here we describe a comprehensive series of background assessments at three high-power research reactors, including $\\gamma$-ray, neutron, and muon measurements. For each facility we describe the characteristics and identify the sources of the background fields encountered. The general understanding gained of background production mechanisms and their relationship to facility features will prove valuable for the planning of any sensitive measurement conducted therein.

  5. Background radiation measurements at high power research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Ashenfelter, J. [Wright Laboratory, Department of Physics, Yale University, New Haven, CT 06520 (United States); Balantekin, B. [Department of Physics, University of Wisconsin, Madison, WI 53706 (United States); Baldenegro, C.X. [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Band, H.R. [Wright Laboratory, Department of Physics, Yale University, New Haven, CT 06520 (United States); Barclay, G. [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Bass, C.D. [Department of Chemistry and Physics, Le Moyne College, Syracuse, NY 13214 (United States); Berish, D. [Department of Physics, Temple University, Philadelphia, PA 19122 (United States); Bowden, N.S., E-mail: nbowden@llnl.gov [Nuclear and Chemical Sciences Division, Lawrence Livermore National Laboratory, Livermore, CA 94550 (United States); Bryan, C.D. [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Cherwinka, J.J. [Physical Sciences Laboratory, University of Wisconsin, Madison, WI 53706 (United States); Chu, R. [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Department of Physics, University of Tennessee, Knoxville, TN 37996 (United States); Classen, T. [Nuclear and Chemical Sciences Division, Lawrence Livermore National Laboratory, Livermore, CA 94550 (United States); Davee, D. [Department of Physics, College of William and Mary, Williamsburg, VA 23187 (United States); Dean, D.; Deichert, G. [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Dolinski, M.J. [Department of Physics, Drexel University, Philadelphia, PA 19104 (United States); Dolph, J. [Physics Department, Brookhaven National Laboratory, Upton, NY 11973 (United States); Dwyer, D.A. [Physics Division, Lawrence Berkeley National Laboratory, Berkeley, CA 94720 (United States); Fan, S. [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Department of Physics, University of Tennessee, Knoxville, TN 37996 (United States); and others

    2016-01-11

    Research reactors host a wide range of activities that make use of the intense neutron fluxes generated at these facilities. Recent interest in performing measurements with relatively low event rates, e.g. reactor antineutrino detection, at these facilities necessitates a detailed understanding of background radiation fields. Both reactor-correlated and naturally occurring background sources are potentially important, even at levels well below those of importance for typical activities. Here we describe a comprehensive series of background assessments at three high-power research reactors, including γ-ray, neutron, and muon measurements. For each facility we describe the characteristics and identify the sources of the background fields encountered. The general understanding gained of background production mechanisms and their relationship to facility features will prove valuable for the planning of any sensitive measurement conducted therein.

  6. Model Based Cyber Security Analysis for Research Reactor Protection System

    Energy Technology Data Exchange (ETDEWEB)

    Sho, Jinsoo; Rahman, Khalil Ur; Heo, Gyunyoung [Kyung Hee Univ., Yongin (Korea, Republic of); Son, Hanseong [Joongbu Univ., Geumsan (Korea, Republic of)

    2013-07-01

    The study on the qualitative risk due to cyber-attacks into research reactors was performed using bayesian Network (BN). This was motivated to solve the issues of cyber security raised due to digitalization of instrumentation and control (I and C) system. As a demonstrative example, we chose the reactor protection system (RPS) of research reactors. Two scenarios of cyber-attacks on RPS were analyzed to develop mitigation measures against vulnerabilities. The one is the 'insertion of reactor trip' and the other is the 'scram halt'. The six mitigation measures are developed for five vulnerability for these scenarios by getting the risk information from BN.

  7. Neutron scattering at Australia's replacement research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kennedy, S.J.; Robinson, R.A.; Hunter, B.A. [Physics Division, ANSTO, Lucas Heights, NSW (Australia)

    2001-03-01

    On August 25{sup th} 1999, the Australian government gave final approval to build a research reactor to replace the existing HIFAR reactor at Lucas Heights. The replacement reactor, which will commence operation in 2005, will be multipurpose with capabilities for both neutron beam research and radioisotope production. Cold, and thermal neutron sources are to be installed and supermirror guides will transport cold and thermal neutron beams into a large modern guide hall. The reactor and associated infrastructure is to be built by INVAP, SE and subcontractors under contract. The neutron beam instruments will be developed by ANSTO in consultation with the Australian user community and interested overseas parties. We review the planned scientific capabilities, give a description of the facility and a status report on the activities so far. (author)

  8. Experimental and Analytic Study on the Core Bypass Flow in a Very High Temperature Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Richard Schultz

    2012-04-01

    Core bypass flow has been one of key issues in the very high temperature reactor (VHTR) design for securing core thermal margins and achieving target temperatures at the core exit. The bypass flow in a prismatic VHTR core occurs through the control element holes and the radial and axial gaps between the graphite blocks for manufacturing and refueling tolerances. These gaps vary with the core life cycles because of the irradiation swelling/shrinkage characteristic of the graphite blocks such as fuel and reflector blocks, which are main components of a core's structure. Thus, the core bypass flow occurs in a complicated multidimensional way. The accurate prediction of this bypass flow and counter-measures to minimize it are thus of major importance in assuring core thermal margins and securing higher core efficiency. Even with this importance, there has not been much effort in quantifying and accurately modeling the effect of the core bypass flow. The main objectives of this project were to generate experimental data for validating the software to be used to calculate the bypass flow in a prismatic VHTR core, validate thermofluid analysis tools and their model improvements, and identify and assess measures for reducing the bypass flow. To achieve these objectives, tasks were defined to (1) design and construct experiments to generate validation data for software analysis tools, (2) determine the experimental conditions and define the measurement requirements and techniques, (3) generate and analyze the experimental data, (4) validate and improve the thermofluid analysis tools, and (5) identify measures to control the bypass flow and assess its performance in the experiment.

  9. Subchannel analysis of a small ultra-long cycle fast reactor core

    Energy Technology Data Exchange (ETDEWEB)

    Seo, Han; Kim, Ji Hyun; Bang, In Cheol, E-mail: icbang@unist.ac.kr

    2014-04-01

    Highlights: • The UCFR-100 is small-sized one of 60 years long-life nuclear reactors without refueling. • The design safety limits of the UCFR-100 are evaluated using MATRA-LMR. • The subchannel results are below the safety limits of general SFR design criteria. - Abstract: Thermal-hydraulic evaluation of a small ultra-long cycle fast reactor (UCFR) core is performed based on existing safety regulations. The UCFR is an innovative reactor newly designed with long-life core based on the breed-and-burn strategy and has a target electric power of 100 MWe (UCFR-100). Low enriched uranium (LEU) located at the bottom region of the core play the role of igniter to operate the UCFR for 60 years without refueling. A metallic form is selected as a burning fuel region material after the LEU location. HT-9 and sodium are used as cladding and coolant materials, respectively. In the present study, MATRA-LMR, subchannel analysis code, is used for evaluating the safety design limit of the UCFR-100 in terms of fuel, cladding, and coolant temperature distributions in the core as design criteria of a general fast reactor. The start-up period (0 year of operation), the middle of operating period (30 years of operation), and the end of operating cycle (60 years of operation) are analyzed and evaluated. The maximum cladding surface temperature (MCST) at the BOC (beginning of core life) is 498 °C on average and 551 °C when considering peaking factor, while the MCST at the MOC (middle of core life) is 498 °C on average and 548 °C in the hot channel, respectively, and the MCST at the EOC (end of core life) is 499 °C on average and 538 °C in the hot channel, respectively. The maximum cladding surface temperature over the long cycle is found at the BOC due to its high peaking factor. It is found that all results including fuel rods, cladding, and coolant exit temperature are below the safety limit of general SFR design criteria.

  10. Nuclear Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hogerton, John

    1964-01-01

    This pamphlet describes how reactors work; discusses reactor design; describes research, teaching, and materials testing reactors; production reactors; reactors for electric power generation; reactors for supply heat; reactors for propulsion; reactors for space; reactor safety; and reactors of tomorrow. The appendix discusses characteristics of U.S. civilian power reactor concepts and lists some of the U.S. reactor power projects, with location, type, capacity, owner, and startup date.

  11. Advanced computational methods for the assessment of reactor core behaviour during reactivity initiated accidents. Final report; Fortschrittliche Rechenmethoden zum Kernverhalten bei Reaktivitaetsstoerfaellen. Abschlussbericht

    Energy Technology Data Exchange (ETDEWEB)

    Pautz, A.; Perin, Y.; Pasichnyk, I.; Velkov, K.; Zwermann, W.; Seubert, A.; Klein, M.; Gallner, L.; Krzycacz-Hausmann, B.

    2012-05-15

    The document at hand serves as the final report for the reactor safety research project RS1183 ''Advanced Computational Methods for the Assessment of Reactor Core Behavior During Reactivity-Initiated Accidents''. The work performed in the framework of this project was dedicated to the development, validation and application of advanced computational methods for the simulation of transients and accidents of nuclear installations. These simulation tools describe in particular the behavior of the reactor core (with respect to neutronics, thermal-hydraulics and thermal mechanics) at a very high level of detail. The overall goal of this project was the deployment of a modern nuclear computational chain which provides, besides advanced 3D tools for coupled neutronics/ thermal-hydraulics full core calculations, also appropriate tools for the generation of multi-group cross sections and Monte Carlo models for the verification of the individual calculational steps. This computational chain shall primarily be deployed for light water reactors (LWR), but should beyond that also be applicable for innovative reactor concepts. Thus, validation on computational benchmarks and critical experiments was of paramount importance. Finally, appropriate methods for uncertainty and sensitivity analysis were to be integrated into the computational framework, in order to assess and quantify the uncertainties due to insufficient knowledge of data, as well as due to methodological aspects.

  12. Assessment of mass fraction and melting temperature for the application of limestone concrete and siliceous concrete to nuclear reactor basemat considering molten core-concrete interaction

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Ho Jae; Kim, Do Gyeum [Korea Institute of Civil Engineering and Building Technology, Goyang (Korea, Republic of); Cho, Jae Leon [Korea Hydro and Nuclear Power Co., Ulsan (Korea, Republic of); Yoon, Eui Sik [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of); Cho, Myung Suk [Korea Hydro and Nuclear Power Co., Central Research Institute, Daejeon (Korea, Republic of)

    2016-04-15

    Severe accident scenarios in nuclear reactors, such as nuclear meltdown, reveal that an extremely hot molten core may fall into the nuclear reactor cavity and seriously affect the safety of the nuclear containment vessel due to the chain reaction caused by the reaction between the molten core and concrete. This paper reports on research focused on the type and amount of vapor produced during the reaction between a high-temperature molten core and concrete, as well as on the erosion rate of concrete and the heat transfer characteristics at its vicinity. This study identifies the mass fraction and melting temperature as the most influential properties of concrete necessary for a safety analysis conducted in relation to the thermal interaction between the molten core and the basemat concrete. The types of concrete that are actually used in nuclear reactor cavities were investigated. The H2O content in concrete required for the computation of the relative amount of gases generated by the chemical reaction of the vapor, the quantity of CO2 necessary for computing the cooling speed of the molten core, and the melting temperature of concrete are evaluated experimentally for the molten core-concrete interaction analysis.

  13. Experimental distribution of coolant in the IPR-R1 Triga nuclear reactor core

    Energy Technology Data Exchange (ETDEWEB)

    Mesquita, Amir Z., E-mail: amir@cdtn.b [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil). Servico de Tecnologia de Reatores; Palma, Daniel A.P., E-mail: dapalma@cnen.gov.b [Comissao Nacional de Energia Nuclear (CNEN/RJ), Rio de Janeiro, RJ (Brazil); Costa, Antonella L.; Pereira, Claubia; Veloso, Maria A.F.; Reis, Patricia A.L., E-mail: claubia@nuclear.ufmg.b, E-mail: dora@nuclear.ufmg.b [Universidade Federal de Minas Gerais (DEN/UFMG), Belo Horizonte, MG (Brazil). Dept. de Engenharia Nuclear

    2011-07-01

    The IPR-R1 is a typical TRIGA Mark I light-water and open pool type reactor. The core has an annular configuration of six rings and is cooled by natural circulation. The core coolant channels extend from the bottom grid plate to the top grid plate. The cooling water flows through the holes in the bottom grid plate, passes through the lower unheated region of the element, flows upwards through the active region, passes through the upper unheated region, and finally leaves the channel through the differential area between a triangular spacer block on the top of the fuel element and a round hole in the grid. Direct measurement of the flow rate in a coolant channel is difficult because of the bulky size and low accuracy of flow meters. The flow rate through the channel may be determined indirectly from the heat balance across the channel using measurements of the water inlet and outlet temperatures. This paper presents the experiments performed in the IPR-R1 reactor to monitoring some thermo-hydraulic parameters in the core coolant channels, such as: the radial and axial temperature profile, temperature, velocity, mass flow rate, mass flux and Reynolds's number. Some results were compared with theoretical predictions, as it was expected the variables follow the power distribution (or neutron flux) in the core. (author)

  14. The rehabilitation/upgrading of Philippine Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Renato, T. Banaga [Philippines Nuclear Research Inst., Quezon (Philippines)

    1998-10-01

    The Philippine Research Reactor (PRR-1) is the only research reactor in the Philippines. It was acquired through the Bilateral Agreement with the United States of America. The General Electric (G.E.) supplied PRR-1 first become operational in 1963 and used MTR plate type fuel. The original one-megawatt G.E. reactor was shutdown and converted into a 3 MW TRIGA PULSING REACTOR in 1984. The conversion includes the upgrading of the cooling system, replacement of new reactor coolant pumps, heat exchanger, cooling tower, replacement of new nuclear instrumentation and standard TRIGA console, TRIGA fuel supplied by General Atomic (G.A.). Philippine Nuclear Research Institute (PNRI) provided the old reactor, did the detailed design of the new cooling system, provided the new non-nuclear instrumentation and electrical power supply system and performed all construction, installation and modification work on site. The TRIGA conversion fuel is contained in a shrouded 4-rod cluster which fit into the original grid plate. The new fuel is a E{sub 1}-U-Z{sub 1}-H{sub 1.6} TRIGA fuel, has a 20% wt Uranium loading with 19.7% U-235 enrichment and about 0.5 wt % Erbium. The Start-up, calibration and Demonstration of Pulsing and Full Power Operation were completed during a three week start-up phase which were performed last March 1968. A few days after, a leak in the pool liner was discovered. The reactor was shutdown again for repair and up to present the reactor is still in the process of rehabilitation. This paper will describe the rehabilitation/upgrading done on the PRR-1 since 1988 up to present. (author)

  15. Status of research and development on reduced-moderation water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Iwamura, Takamichi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2002-12-01

    To improve uranium utilization, a design study of the Reduced-Moderation Water Reactor (RMWR) has been carried out intensively since 1998 at the Japan Atomic Energy Research Institute (JAERI). In this reactor, the nuclear fission reaction is designed to be realized mainly by high energy neutrons. To achieve this, the volume of water used to cool the fuel rods is decreased by reducing the gap width between the fuel rods. Conversion ratio greater than 1.0 is expected whether the core i-s cooled by boiling water or pressurized water and whether the core size is small or large. Status of the RMWR design is reviewed and planning of R and D for future deployment of this reactor after 20-20 is presented. To improve economics of this reactor, development of fuel cans for high burnup and low-cost reprocessing technology of mixed oxide spect fuels are highly needed. R and D has been conducted under the cooperation with utilities, industry, research organization and academia. (T. Tanaka)

  16. Electrically Heated Testing of the Kilowatt Reactor Using Stirling Technology (KRUSTY) Experiment Using a Depleted Uranium Core

    Science.gov (United States)

    Briggs, Maxwell H.; Gibson, Marc A.; Sanzi, James

    2017-01-01

    The Kilopower project aims to develop and demonstrate scalable fission-based power technology for systems capable of delivering 110 kW of electric power with a specific power ranging from 2.5 - 6.5 Wkg. This technology could enable high power science missions or could be used to provide surface power for manned missions to the Moon or Mars. NASA has partnered with the Department of Energys National Nuclear Security Administration, Los Alamos National Labs, and Y-12 National Security Complex to develop and test a prototypic reactor and power system using existing facilities and infrastructure. This technology demonstration, referred to as the Kilowatt Reactor Using Stirling TechnologY (KRUSTY), will undergo nuclear ground testing in the summer of 2017 at the Nevada Test Site. The 1 kWe variation of the Kilopower system was chosen for the KRUSTY demonstration. The concept for the 1 kWe flight system consist of a 4 kWt highly enriched Uranium-Molybdenum reactor operating at 800 degrees Celsius coupled to sodium heat pipes. The heat pipes deliver heat to the hot ends of eight 125 W Stirling convertors producing a net electrical output of 1 kW. Waste heat is rejected using titanium-water heat pipes coupled to carbon composite radiator panels. The KRUSTY test, based on this design, uses a prototypic highly enriched uranium-molybdenum core coupled to prototypic sodium heat pipes. The heat pipes transfer heat to two Advanced Stirling Convertors (ASC-E2s) and six thermal simulators, which simulate the thermal draw of full scale power conversion units. Thermal simulators and Stirling engines are gas cooled. The most recent project milestone was the completion of non-nuclear system level testing using an electrically heated depleted uranium (non-fissioning) reactor core simulator. System level testing at the Glenn Research Center (GRC) has validated performance predictions and has demonstrated system level operation and control in a test configuration that replicates the one

  17. Human reliability analysis of the Tehran research reactor using the SPAR-H method

    Directory of Open Access Journals (Sweden)

    Barati Ramin

    2012-01-01

    Full Text Available The purpose of this paper is to cover human reliability analysis of the Tehran research reactor using an appropriate method for the representation of human failure probabilities. In the present work, the technique for human error rate prediction and standardized plant analysis risk-human reliability methods have been utilized to quantify different categories of human errors, applied extensively to nuclear power plants. Human reliability analysis is, indeed, an integral and significant part of probabilistic safety analysis studies, without it probabilistic safety analysis would not be a systematic and complete representation of actual plant risks. In addition, possible human errors in research reactors constitute a significant part of the associated risk of such installations and including them in a probabilistic safety analysis for such facilities is a complicated issue. Standardized plant analysis risk-human can be used to address these concerns; it is a well-documented and systematic human reliability analysis system with tables for human performance choices prepared in consultation with experts in the domain. In this method, performance shaping factors are selected via tables, human action dependencies are accounted for, and the method is well designed for the intended use. In this study, in consultations with reactor operators, human errors are identified and adequate performance shaping factors are assigned to produce proper human failure probabilities. Our importance analysis has revealed that human action contained in the possibility of an external object falling on the reactor core are the most significant human errors concerning the Tehran research reactor to be considered in reactor emergency operating procedures and operator training programs aimed at improving reactor safety.

  18. Operating manual for the Health Physics Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    1985-11-01

    This manual is intended to serve as a guide in the operation and maintenance of the Health Physics Researh Reactor (HPRR) of the Health Physics Dosimetry Applications Research (DOSAR) Facility. It includes descriptions of the HPRR and of associated equipment such as the reactor positioning devises and the derrick. Procedures for routine operation of the HPRR are given in detail, and checklists for the various steps are provided where applicable. Emergency procedures are similarly covered, and maintenance schedules are outlined. Also, a bibliography of references giving more detailed information on the DOSAR Facility is included. Changes to this manual will be approved by at least two of the following senior staff members: (1) the Operations Division Director, (2) the Reactor Operations Department Head, (3) the Supervisor of Reactor Operations TSF-HPRR Areas. The master copy and the copy of the manual issued to the HPRR Operations Supervisor will always reflect the latest revision. 22 figs.

  19. Use of Distribution Devices for Hydraulic Profiling of Coolant Flow in Core Gas-cooled Reactors

    Directory of Open Access Journals (Sweden)

    A. A. Satin

    2014-01-01

    Full Text Available In setting up a reactor plant for the transportation-power module of the megawatt class an important task is to optimize the path of flow, i.e. providing moderate hydraulic resistance, uniform distribution of the coolant. Significant contribution to the hydraulic losses makes one selected design of the coolant supplies. It is, in particular, hemispherical or semi-elliptical shape of the supply reservoir, which is selected to reduce its mass, resulting in the formation of torusshaped vortex in the inlet manifold, that leads to uneven coolant velocity at the inlet into the core, the flow pulsations, hydraulic losses.To control the flow redistribution in the core according to the level of energy are used the switchgear - deflectors installed in a hemispherical reservoir supplying coolant to the fuel elements (FE of the core of gas-cooled reactor. This design solution has an effect on the structure of the flow, rate in the cooling duct, and the flow resistance of the collector.In this paper we present the results of experiments carried out on the gas dynamic model of coolant paths, deflectors, and core, comprising 55 fuel rod simulators. Numerical simulation of flow in two-parameter model, using the k-ε turbulence model, and the software package ANSYS CFX v14.0 is performed. The paper demonstrates that experimental results are in compliance with calculated ones.The results obtained suggest that the use of switchgear ensures a coolant flow balance directly at the core inlet, thereby providing temperature reduction of fuel rods with a uniform power release in the cross-section. Considered options to find constructive solutions for deflectors give an idea to solve the problem of reducing hydraulic losses in the coolant paths, to decrease pulsation components of flow in the core and length of initial section of flow stabilization.

  20. CORE ANALYSIS, DESIGN AND OPTIMIZATION OF A DEEP-BURN PEBBLE BED REACTOR

    Energy Technology Data Exchange (ETDEWEB)

    B. Boer; A. M. Ougouag

    2010-05-01

    Achieving a high burnup in the Deep-Burn pebble bed reactor design, while remaining within the limits for fuel temperature, power peaking and temperature reactivity feedback, is challenging. The high content of Pu and Minor Actinides in the Deep-Burn fuel significantly impacts the thermal neutron energy spectrum. This can result in power and temperature peaking in the pebble bed core in locally thermalized regions near the graphite reflectors. Furthermore, the interplay of the Pu resonances of the neutron absorption cross sections at low-lying energies can lead to a positive temperature reactivity coefficient for the graphite moderator at certain operating conditions. To investigate the aforementioned effects a code system using existing codes has been developed for neutronic, thermal-hydraulic and fuel depletion analysis of Deep-Burn pebble bed reactors. A core analysis of a Deep-Burn Pebble Bed Modular Reactor (400 MWth) design has been performed for two Deep-Burn fuel types and possible improvements of the design with regard to power peaking and temperature reactivity feedback are identified.

  1. Measurements of actinide-fission product yields in Caliban and Prospero metallic core reactor fission neutron fields

    Energy Technology Data Exchange (ETDEWEB)

    Casoli, P.; Authier, N. [CEA, Centre de Valduc, 21120 Is-sur-Tille (France); Laurec, J.; Bauge, E.; Granier, T. [CEA, Centre DIF, 91297 Arpajon (France)

    2011-07-01

    In the 1970's and early 1980's, an experimental program was performed on the facilities of the CEA Valduc Research Center to measure several actinide-fission product yields. Experiments were, in particular, completed on the Caliban and Prospero metallic core reactors to study fission-neutron-induced reactions on {sup 233}U, {sup 235}U, and {sup 239}Pu. Thick actinide samples were irradiated and the number of nuclei of each fission product was determined by gamma spectrometry. Fission chambers were irradiated simultaneously to measure the numbers of fissions in thin deposits of the same actinides. The masses of the thick samples and the thin deposits were determined by mass spectrometry and alpha spectrometry. The results of these experiments will be fully presented in this paper for the first time. A description of the Caliban and Prospero reactors, their characteristics and performances, and explanations about the experimental approach will also be given in the article. A recent work has been completed to analyze and reinterpret these measurements and particularly to evaluate the associated uncertainties. In this context, calculations have also been carried out with the Monte Carlo transport code Tripoli-4, using the published benchmarked Caliban description and a three-dimensional model of Prospero, to determine the average neutron energy causing fission. Simulation results will be discussed in this paper. Finally, new fission yield measurements will be proposed on Caliban and Prospero reactors to strengthen the results of the first experiments. (authors)

  2. BURNUR.SYS: A 2-D code system for NUR research reactor burn up analysis

    Energy Technology Data Exchange (ETDEWEB)

    Meftah, B. [Division Reacteur NUR, Centre de Recherche Nucleaire de Draria, BP 43 Sebala, Alger (Algeria)], E-mail: b_meftah@yahoo.com; Halilou, A. [Division Reacteur NUR, Centre de Recherche Nucleaire de Draria, BP 43 Sebala, Alger (Algeria); Letaim, F.; Mazidi, S. [Faculte de Physique, Universite Haouri Boumediene, USTHB, BP 31 Bab Ezzouar, Alger (Algeria); Mokeddem, M.Y. [Division Physique et Applications Nucleaires, Centre de Recherche Nucleaire de Draria, BP 43 Sebala, Alger (Algeria); Zeggar, F. [Division Surete Nucleaire et Radioprotection, Centre de Recherche Nucleaire de Draria, BP 43 Sebala, Alger (Algeria)

    2008-04-15

    Adequate knowledge of burn up levels of fuel elements within a research reactor is of great importance for its optimum operation. Such knowledge is required for the monitoring of reactivity parameters and flux and power distributions throughout the reactor core, the estimation of the radioactive source term needed in accidental situations analysis, the evaluation of the amount of fissile materials present at any moment within the fuel for safeguards purposes and the estimation of cooling and shielding requirements for interim storage or transport of spent fuel elements. This paper presents the approach of fuel burn up evaluation used at the NUR research reactor. The approach is essentially based upon the utilization of BURNUR.SYS code, an in-house developed software. BURNUR.SYS is an object oriented program under DELPHI 7 that integrates the cell calculation code WIMSD-4 and the core calculation code CITVAP. BURNUR.SYS calculates the evolution in time of pertinent quantities such as: the concentrations of U235 and others actinides, the concentrations of major poisons (Xe135 and Sm149), the distributions of power densities and burn up levels within fuel elements, the effective multiplication factor and core reactivity. The results are displayed in user friendly graphical and numerical formats.

  3. Health services research doctoral core competencies

    Science.gov (United States)

    Forrest, Christopher B; Martin, Diane P; Holve, Erin; Millman, Anne

    2009-01-01

    This manuscript presents an initial description of doctoral level core competencies for health services research (HSR). The competencies were developed by a review of the literature, text analysis of institutional accreditation self-studies submitted to the Council on Education for Public Health, and a consensus conference of HSR educators from US educational institutions. The competencies are described in broad terms which reflect the unique expertise, interests, and preferred learning methods of academic HSR programs. This initial set of core competencies is published to generate further dialogue within and outside of the US about the most important learning objectives and methods for HSR training and to clarify the unique skills of HSR training program graduates. PMID:19555485

  4. Health services research doctoral core competencies

    Directory of Open Access Journals (Sweden)

    Holve Erin

    2009-06-01

    Full Text Available Abstract This manuscript presents an initial description of doctoral level core competencies for health services research (HSR. The competencies were developed by a review of the literature, text analysis of institutional accreditation self-studies submitted to the Council on Education for Public Health, and a consensus conference of HSR educators from US educational institutions. The competencies are described in broad terms which reflect the unique expertise, interests, and preferred learning methods of academic HSR programs. This initial set of core competencies is published to generate further dialogue within and outside of the US about the most important learning objectives and methods for HSR training and to clarify the unique skills of HSR training program graduates.

  5. Porosity Effect in the Core Thermal Hydraulics for Ultra High Temperature Gas-cooled Reactor

    Directory of Open Access Journals (Sweden)

    Motoo Fumizawa

    2008-12-01

    Full Text Available This study presents an experimental method of porosity evaluation and a predictive thermal-hydraulic analysis with packed spheres in a nuclear reactor core. The porosity experiments were carried out in both a fully shaken state with the closest possible packing and in a state of non-vibration. The predictive analysis considering the fixed porosity value was applied as a design condition for an Ultra High Temperature Reactor Experiment (UHTREX. The thermal-hydraulic computer code was developed and identified as PEBTEMP. The highest outlet coolant temperature of 1316 oC was achieved in the case of an UHTREX at Los Alamos Scientific Laboratory, which was a small scale UHTR. In the present study, the fuel was changed to a pebble type, a porous media. In order to compare the present pebble bed reactor and UHTREX, a calculation based on HTGR-GT300 was carried out in similar conditions with UHTREX; in other words, with an inlet coolant temperature of 871oC, system pressure of 3.45 MPa and power density of 1.3 w/cm3. As a result, the fuel temperature in the present pebble bed reactor showed an extremely lower value compared to that of UHTREX.

  6. A simulation of a pebble bed reactor core by the MCNP-4C computer code

    Directory of Open Access Journals (Sweden)

    Bakhshayesh Moshkbar Khalil

    2009-01-01

    Full Text Available Lack of energy is a major crisis of our century; the irregular increase of fossil fuel costs has forced us to search for novel, cheaper, and safer sources of energy. Pebble bed reactors - an advanced new generation of reactors with specific advantages in safety and cost - might turn out to be the desired candidate for the role. The calculation of the critical height of a pebble bed reactor at room temperature, while using the MCNP-4C computer code, is the main goal of this paper. In order to reduce the MCNP computing time compared to the previously proposed schemes, we have devised a new simulation scheme. Different arrangements of kernels in fuel pebble simulations were investigated and the best arrangement to decrease the MCNP execution time (while keeping the accuracy of the results, chosen. The neutron flux distribution and control rods worth, as well as their shadowing effects, have also been considered in this paper. All calculations done for the HTR-10 reactor core are in good agreement with experimental results.

  7. Technological Transfer from Research Nuclear Reactors to New Generation Nuclear Power Reactors

    Science.gov (United States)

    Radulescu, Laura; Pavelescu, Margarit

    2010-01-01

    The goal of this paper is the analysis of the technological transfer role in the nuclear field, with particular emphasis on nuclear reactors domain. The presentation is sustained by historical arguments. In this frame, it is very important to start with the achievements of the first nuclear systems, for instant those with natural uranium as fuel and heavy water as moderator, following in time through the history until the New Generation Nuclear Power Reactors. Starting with 1940, the accelerated development of the industry has implied the increase of the global demand for energy. In this respect, the nuclear energy could play an important role, being essentially an unlimited source of energy. However, the nuclear option faces the challenges of increasingly demanding safety requirements, economic competitiveness and public acceptance. Worldwide, a significant amount of experience has been accumulated during development, licensing, construction, and operation of nuclear power reactors. The experience gained is a strong basis for further improvements. Actually, the nuclear programs of many countries are addressing the development of advanced reactors, which are intended to have better economics, higher reliability, improved safety, and proliferation-resistant characteristics in order to overcome the current concerns about nuclear power. Advanced reactors, now under development, may help to meet the demand for energy power of both developed and developing countries as well as for district heating, desalination and for process heat. The paper gives historical examples that illustrate the steps pursued from first research nuclear reactors to present advanced power reactors. Emphasis was laid upon the fact that the progress is due to the great discoveries of the nuclear scientists using the technological transfer.

  8. Reduced enrichment for research and test reactors: Proceedings

    Energy Technology Data Exchange (ETDEWEB)

    1988-05-01

    The international effort to develop new research reactor fuel materials and designs based on the use of low-enriched uranium, instead of highly-enriched uranium, has made much progress during the eight years since its inception. To foster direct communication and exchange of ideas among the specialist in this area, the Reduced Enrichment Research and Test Reactor (RERTR) Program, at the Argonne National Laboratory, sponsored this meeting as the ninth of a series which began in 1978. All previous meetings of this series are listed on the facing page. The focus of this meeting was on the LEU fuel demonstration which was in progress at the Oak Ridge Research (ORR) reactor, not far from where the meeting was held. The visit to the ORR, where a silicide LEU fuel with 4.8 g A/cm/sup 3/ was by then in routine use, illustrated how far work has progressed.

  9. A probabilistic safety analysis of incidents in nuclear research reactors.

    Science.gov (United States)

    Lopes, Valdir Maciel; Agostinho Angelo Sordi, Gian Maria; Moralles, Mauricio; Filho, Tufic Madi

    2012-06-01

    This work aims to evaluate the potential risks of incidents in nuclear research reactors. For its development, two databases of the International Atomic Energy Agency (IAEA) were used: the Research Reactor Data Base (RRDB) and the Incident Report System for Research Reactor (IRSRR). For this study, the probabilistic safety analysis (PSA) was used. To obtain the result of the probability calculations for PSA, the theory and equations in the paper IAEA TECDOC-636 were used. A specific program to analyse the probabilities was developed within the main program, Scilab 5.1.1. for two distributions, Fischer and chi-square, both with the confidence level of 90 %. Using Sordi equations, the maximum admissible doses to compare with the risk limits established by the International Commission on Radiological Protection (ICRP) were obtained. All results achieved with this probability analysis led to the conclusion that the incidents which occurred had radiation doses within the stochastic effects reference interval established by the ICRP-64.

  10. Prevention of ageing of research reactors by design

    Energy Technology Data Exchange (ETDEWEB)

    Boado, J. [INVAP, Bariloche (Argentina); Lolich, J. [INVAP, Bariloche (Argentina)

    1995-12-31

    However, it is our experience as designers and builders of research reactors, that the most important cause of ageing of any experimental installation, is the loss of motivation of the personnel involved in the operation and maintenance, when the objectives for the utilisation of the facility change or research programs are abandoned for whatever reason. We therefore have endeavoured to design and construct research reactors with several engineering features and with an untraditional approach to the training of the future operator of the facility. During all phases of the design and construction of the reactor, we develop in the future operator of the facility the capacity, not only to operate it properly, but to innovate and to adapt the installation to the daily operating problems due to new requirements and options that might not have been foreseen when the facility was ordered. The versatility of the operator is thus a further guarantee against ageing by abandonment. (orig./HP)

  11. Radiation protection personnel training in Research Reactors; Capacitacion en proteccion radiologica para reactores de investigacion

    Energy Technology Data Exchange (ETDEWEB)

    Fernandez, Carlos Dario; Lorenzo, Nestor Pedro de [Comision Nacional de Energia Atomica, Rio Negro (Argentina). Centro Atomico Bariloche. Instituto Balseiro

    1996-07-01

    The RA-6 research reactor is considering the main laboratory in the training of different groups related with radiological protection. The methodology applied to several courses over 15 years of experience is shown in this work. The reactor is also involved in the construction, design, start-up and sell of different installation outside Argentina for this reason several theoretical and practical courses had been developed. The acquired experience obtained is shown in this paper and the main purpose is to show the requirements to be taken into account for every group (subjects, goals, on-job training, etc) (author)

  12. RADIATION DOSIMETRY AT THE BNL HIGH FLUX BEAM REACTOR AND MEDICAL RESEARCH REACTOR.

    Energy Technology Data Exchange (ETDEWEB)

    HOLDEN,N.E.

    1999-09-10

    RADIATION DOSIMETRY MEASUREMENTS HAVE BEEN PERFORMED OVER A PERIOD OF MANY YEARS AT THE HIGH FLUX BEAM REACTOR (HFBR) AND THE MEDICAL RESEARCH REACTOR (BMRR) AT BROOKHAVEN NATIONAL LABORATORY TO PROVIDE INFORMATION ON THE ENERGY DISTRIBUTION OF THE NEUTRON FLUX, NEUTRON DOSE RATES, GAMMA-RAY FLUXES AND GAMMA-RAY DOSE RATES. THE MCNP PARTICLE TRANSPORT CODE PROVIDED MONTE CARLO RESULTS TO COMPARE WITH VARIOUS DOSIMETRY MEASUREMENTS PERFORMED AT THE EXPERIMENTAL PORTS, AT THE TREATMENT ROOMS AND IN THE THIMBLES AT BOTH HFBR AND BMRR.

  13. Decommissioning Project for the Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Chung, U. S.; Park, J. H.; Paik, S. T. (and others)

    2009-02-15

    In 2008, tried to complete the whole decommissioning project of KRR-1 and KRR-2 and preparing work for memorial museum of KRR-1 reactor. Now the project is delayed for 3 months because of finding unexpected soil contamination around facility and treatment of. To do final residual radioactivity assessment applied by MARSSIM procedure. Accumulated decommissioning experiences and technologies will be very usefully to do decommissioning other nuclear related facility. At the decommissioning site of the uranium conversion plant, the decontamination of the dismantled carbon steel waste are being performed and the lagoon 1 sludge waste is being treated this year. The technologies and experiences obtained from the UCP dismantling works are expected to apply to other fuel cycle facilities decommissioning. The lagoon sludge treatment technology is the first applied technology in the actual field and it is expected that this technology could be applied to other country.

  14. A simple reactivity feedback model accounting for radial core expansion effects in the liquid metal fast reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kwon, Young Min; Lee, Yong Bum; Chang, Won Pyo; Haha, Do Hee [KAERI, Taejon (Korea, Republic of)

    2002-10-01

    The radial core expansion due to the structure temperature rise is one of major negative reactivity insertion mechanisms in metallic fueled reactor. Thermal expansion is a result of both the laws of nature and the particular core design and it causes negative reactivity feedback by the combination of increased core volume captures and increased core surface leakage. The simple radial core expansion reactivity feedback model developed for the SSC-K code was evaluated by the code-to-code comparison analysis. From the comparison results, it can be stated that the radial core expansion reactivity feedback model employed into the SSC-K code may be reasonably accurate in the UTOP analysis.

  15. ASRL core research program 2010 - 2011

    Energy Technology Data Exchange (ETDEWEB)

    Anon.

    2010-01-15

    This article summarized the core research program of Alberta Sulphur Research Ltd. The high-priority projects are improved liquid sulfur degassing technologies, improved tail gas treatment processes, oxygen consumption in amine systems, formation of hydrogen sulphide (H{sub 2}S) in shale gas reservoirs and during the steam assisted gravity drainage (SAGD) recovery of heavy oil and bitumen, designer hydrocarbon sulfur solvents for sour gas production, determination of the kinetics of H{sub 2}S oxidation in compression systems, H{sub 2}S and sulphur dioxide (SO{sub 2}) solubility in liquid sulfur updating and checking FTIR calibrations, low tonnage sulfur recovery, the effect of BTX on reduction catalysts used in Claus tail gas processing, measurement of acid gas properties at high pressure, catalytic tail gas incineration, and sulfur dust properties. The projects identified as important areas of research are acid gas injection water holding capacity for acid gas mixtures; rate of decomposition of polymeric sulfur; ammonium salt plugging in the Claus Converter Train; re-examination of catalytic partial oxidation for sulfur recovery from low H{sub 2}S content hydrocarbon contaminated acid gas; primary upgrading of oil sands bitumen; prediction of sulfur deposition in sour gas reservoirs; and new extended uses of elemental sulfur. There are two fundamental research programs, which include ongoing research and partial external funding: production of C{sub 3} - C{sub 6} olefins, high octane alkylate, and valuable petrochemicals and computational modeling of catalytic systems. The commercial and specific objectives of each project were described. Two special projects, which aim to take Alberta Sulphur Research Ltd. (ASRL) core research to the commercial demonstration phase, involve injection of SO{sub 2} into disposal reservoirs and above-ground sulfur storage. 1 tab., 22 figs.

  16. Gas Cooled Fast Reactor Research and Development in the European Union

    Directory of Open Access Journals (Sweden)

    Richard Stainsby

    2009-01-01

    Full Text Available Gas-cooled fast reactor (GFR research is directed towards fulfilling the ambitious goals of Generation IV (Gen IV, that is, to develop a safe, sustainable, reliable, proliferation-resistant and economic nuclear energy system. The research is directed towards developing the GFR as an economic electricity generator, with good safety and sustainability characteristics. Fast reactors maximise the usefulness of uranium resources by breeding plutonium and can contribute to minimising both the quantity and radiotoxicity nuclear waste by actinide transmutation in a closed fuel cycle. Transmutation is particularly effective in the GFR core owing to its inherently hard neutron spectrum. Further, GFR is suitable for hydrogen production and process heat applications through its high core outlet temperature. As such GFR can inherit the non-electricity applications that will be developed for thermal high temperature reactors in a sustainable manner. The Euratom organisation provides a route by which researchers in all European states, and other non-European affiliates, can contribute to the Gen IV GFR system. This paper summarises the achievements of Euratom's research into the GFR system, starting with the 5th Framework programme (FP5 GCFR project in 2000, through FP6 (2005 to 2009 and looking ahead to the proposed activities within the 7th Framework Programme (FP7.

  17. Analysis of cracked core spray piping from the Quad Cities Unit 2 boiling water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Diercks, D.R.; Gaitonde, S.M.

    1982-09-01

    The results of a metallurgical analysis of leaking cracks detected in the core spray injection piping of Commonwealth Edison Company's Quad Cities Unit 2 Boiling Water Reactor are described. The cracks were present in a welded 105/sup 0/ elbow assembly in the line, and were found to be caused by intergranular stress corrosion cracking associated with the probable presence of dissolved oxygen in the reactor cooling water and the presence of grain boundary sensitization and local residual stresses induced by welding. The failure is unusual in several respects, including the very large number of cracks (approximately 40) present in the failed component, the axial orientation of the cracks, and the fact that at least one crack completely penetrated a circumferential weld. Virtually all of the cracking occurred in forged material, and the microstructural evidence presented suggests that the orientation of the cracks was influenced by the presence of axially banded delta ferrite in the microstructure of the forged components.

  18. Analysis of Postulated Core Meltdown of an SRP Reactor - Final Report

    Energy Technology Data Exchange (ETDEWEB)

    Durant, W.S.; Brown, R.J.

    1970-10-01

    An analysis was made to determine the consequences of a postulated accident in which the core of a Savannah River Plant reactor melts down following the loss of coolant. The study was made to determine (1) the potential damage to the reactor building that could impair its integrity for confining activity and (2) the need for additional facilities to prevent the activity confinement system from being overheated by the decay heat in the debris. A preliminary report on this analysis was issued previously. The sequence of events during and following the loss of coolant has now been studied in more detail, and a computer program has been written and used to investigate transient heating effects. This is the final report of the analysis and presents the conclusions.

  19. Cost-based optimization of a nuclear reactor core design: a preliminary model

    Energy Technology Data Exchange (ETDEWEB)

    Sacco, Wagner F.; Alves Filho, Hermes [Universidade do Estado do Rio de Janeiro (UERJ), Nova Friburgo, RJ (Brazil). Inst. Politecnico. Dept. de Modelagem Computacional]. E-mails: wfsacco@iprj.uerj.br; halves@iprj.uerj.br; Pereira, Claudio M.N.A. [Instituto de Engenharia Nuclear (IEN), Rio de Janeiro, RJ (Brazil). Div. de Reatores]. E-mail: cmnap@ien.gov.br

    2007-07-01

    A new formulation of a nuclear core design optimization problem is introduced in this article. Originally, the optimization problem consisted in adjusting several reactor cell parameters, such as dimensions, enrichment and materials, in order to minimize the radial power peaking factor in a three-enrichment zone reactor, considering restrictions on the average thermal flux, criticality and sub-moderation. Here, we address the same problem using the minimization of the fuel and cladding materials costs as the objective function, and the radial power peaking factor as an operational constraint. This cost-based optimization problem is attacked by two metaheuristics, the standard genetic algorithm (SGA), and a recently introduced Metropolis algorithm called the Particle Collision Algorithm (PCA). The two algorithms are submitted to the same computational effort and their results are compared. As the formulation presented is preliminary, more elaborate models are also discussed (author)

  20. Study on core radius minimization for long life Pb-Bi cooled CANDLE burnup scheme based fast reactor

    Science.gov (United States)

    Afifah, Maryam; Miura, Ryosuke; Su'ud, Zaki; Takaki, Naoyuki; Sekimoto, H.

    2015-09-01

    Fast Breeder Reactor had been interested to be developed over the world because it inexhaustible source energy, one of those is CANDLE reactor which is have strategy in burn-up scheme, need not control roads for control burn-up, have a constant core characteristics during energy production and don't need fuel shuffling. The calculation was made by basic reactor analysis which use Sodium coolant geometry core parameter as a reference core to study on minimum core reactor radius of CANDLE for long life Pb-Bi cooled, also want to perform pure coolant effect comparison between LBE and sodium in a same geometry design. The result show that the minimum core radius of Lead Bismuth cooled CANDLE is 100 cm and 500 MWth thermal output. Lead-Bismuth coolant for CANDLE reactor enable to reduce much reactor size and have a better void coefficient than Sodium cooled as the most coolant for FBR, then we will have a good point in safety analysis.

  1. A reload and startup plan for conversion of the NIST research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Diamond, D. J. [Brookhaven National Lab. (BNL), Upton, NY (United States)

    2016-03-31

    The National Institute of Standards and Technology operates a 20 MW research reactor for neutron-based research. The heavy-water moderated and cooled reactor is fueled with high-enriched uranium (HEU) but a program to convert the reactor to low-enriched uranium (LEU) fuel is underway. Among other requirements, a reload and startup test plan must be submitted to the U.S. Nuclear Regulatory Commission (NRC) for their approval. The NRC provides guidance for what should be in the plan to ensure that the licensee has sufficient information to operate the reactor safely. Hence, a plan has been generated consisting of two parts. The reload portion of the plan specifies the fuel management whereby initially only two LEU fuel elements are in the core for eight fuel cycles. This is repeated until a point when the optimum approach is to place four fresh LEU elements into the reactor each cycle. This final transition is repeated and after eight cycles the reactor is completely fueled with LEU. By only adding two LEU fuel elements initially, the plan allows for the consumption of HEU fuel elements that are expected to be in storage at the time of conversion and provides additional qualification of production LEU fuel under actual operating conditions. Because the reload is to take place over many fuel cycles, startup tests will be done at different stages of the conversion. The tests, to be compared with calculations to show that the reactor will operate as planned, are the measurement of critical shim arm position and shim arm and regulating rod reactivity worths. An acceptance criterion for each test is specified based on technical specifications that relate to safe operation. Additional tests are being considered that have less safety significance but may be of interest to bolster the validation of analysis tools.

  2. Reactor Safety Research: Semiannual report, July-December 1986

    Energy Technology Data Exchange (ETDEWEB)

    1987-11-01

    Sandia National Laboratories is conducting, under USNRC sponsorship, phenomenological research related to the safety of commercial nuclear power reactors. The research includes experiments to simulate the phenomenology of the accident conditions and the development of analytical models, verified by experiment, which can be used to predict reactor and safety systems performance and behavior under abnormal conditions. The objective of this work is to provide NRC requisite data bases and analytical methods to (1) identify and define safety issues, (2) understand the progression of risk-significant accident sequences, and (3) conduct safety assessments. The collective NRC-sponsored effort at Sandia National Laboratories is directed at enhancing the tehcnology base supporting licensing decisions.

  3. A Study on Comparison of HANARO and KIJANG Research Reactor in Nuclear Safeguards

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Juang; Lee, Sung Ho; Kim, Hyun-Jo [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    As one of major national projects for nuclear science and engineering in Korea, the KIJANG Research Reactor(KJRR) project was commenced in order to develop the core research reactor(RR) technologies for strengthening the competitiveness of the RR export and also to stabilize the supply of key radioisotopes for medical and industrial applications. This paper is about applying IAEA safeguards at new nuclear facility (KJRR). The beginning of this project is comparing of HANARO and KIJANG research reactor in nuclear safeguards for nuclear material accountancy method. As mentioned before, research reactor is basically item counting facility. In Fig 1, first two processes are belonging to item counting. But last two processes are for bulk handling. So KIJANG RR would be treated item counting facility as well as bulk handling facility by fission moly production facility. For this reason, nuclear material accountancy method for KJRR is not easy compared to existing one. This paper accounted for solution of KJRR nuclear material accountancy briefly. Future study on the suitable nuclear material accountancy method for mixed facility between item counting facility and bulk handling facility will be conducted more specifically.

  4. Advanced Test Reactor Core Modeling Update Project Annual Report for Fiscal Year 2010

    Energy Technology Data Exchange (ETDEWEB)

    Rahmat Aryaeinejad; Douglas S. Crawford; Mark D. DeHart; George W. Griffith; D. Scott Lucas; Joseph W. Nielsen; David W. Nigg; James R. Parry; Jorge Navarro

    2010-09-01

    Legacy computational reactor physics software tools and protocols currently used for support of Advanced Test Reactor (ATR) core fuel management and safety assurance and, to some extent, experiment management are obsolete, inconsistent with the state of modern nuclear engineering practice, and are becoming increasingly difficult to properly verify and validate (V&V). Furthermore, the legacy staff knowledge required for application of these tools and protocols from the 1960s and 1970s is rapidly being lost due to staff turnover and retirements. In 2009 the Idaho National Laboratory (INL) initiated a focused effort to address this situation through the introduction of modern high-fidelity computational software and protocols, with appropriate V&V, within the next 3-4 years via the ATR Core Modeling and Simulation and V&V Update (or “Core Modeling Update”) Project. This aggressive computational and experimental campaign will have a broad strategic impact on the operation of the ATR, both in terms of improved computational efficiency and accuracy for support of ongoing DOE programs as well as in terms of national and international recognition of the ATR National Scientific User Facility (NSUF).

  5. Multimegawatt nuclear electric propulsion with gaseous and vapor core reactors with MHD

    Science.gov (United States)

    Knight, Travis; Anghaie, Samim; Smith, Blair; Houts, Michael

    2001-02-01

    This study investigated the development of a system concept for space power generation and nuclear electric propulsion based on a fissioning plasma core reactor (FPCR) with magnetohydrodynamic (MHD) power conversion system, coupled to a magnetoplasmadynamic (MPD) thruster. The FPCR is a liquid-vapor core reactor concept operating with metallic uranium or uranium tetrafluoride (UF4) vapor as the fissioning fuel and alkali metals or their fluorides as working fluid in a closed Rankine cycle with MHD energy conversion. Candidate working fluids include K, Li, Na, KF, LiF, NaF, etc. The system features core outlet temperatures of 3000 to 4000 K at pressures of about 1 to 10 MPa, MHD temperatures of 2000 to 3000 K, and radiator temperatures of 1200 to 2000 K. This combination of parameters offers the potential for low total system specific mass in the range of 0.4 to 0.6 kg/kWe. The MHD output could be coupled with minimal power conditioning to the variable specific impulse magnetoplasma rocket (VASIMR), MPD thrusters or other types of thruster for producing thrust at very high specific impulse (Isp=1500 to 10,000 s). .

  6. Fast reactors fuel cycle core physics results from the CAPRA-CADRA programme

    Energy Technology Data Exchange (ETDEWEB)

    Vasile, A.; Rimpault, G.; Tommasi, J.; Saint Jean, C. de; Delpech, M. [CEA Cadarache, 13 - Saint Paul lez Durance (France); Hesketh, K. [BNFL, Inc., Denver, CO (United States); Beaumont, H.M.; Sunderland, R.E. [NNC Ltd. (United Kingdom); Newton, T.; Smith, P. [AEA Technology (United Kingdom); Raedt, Ch. de [SCK.CEN, Mol (Belgium); Vambenepe, G. [Electricite de France (EDF), 75 - Paris (France); Lefevre, J.C. [FRAMATOME, 92 - Paris-La-Defence (France); Maschek, W.; Haas, D

    2001-07-01

    This paper presents an overview of fast reactor core physics results obtained in the context of the CAPRA-CADRA European collaborative programme, whose aim is to investigate a broad range of possible options for plutonium and radioactive waste management. Different types of fast reactors have been studied to evaluate their potential capabilities with respect to the long term management of plutonium, minor actinides (MAs) and long- lived fission products (LLFPs). Among the several options aiming at reducing waste and consequently radio toxicity are: homogeneous recycling of Minor Actinides, heterogeneous recycling of Minor Actinides either without or with moderation, dedicated critical cores (fuelled mainly with Minor Actinides) and Accelerator Driven System (ADS) variants. In order to achieve a detailed understanding of the potential of the various options, advanced core physics methods have been implemented and tested and applied, for example, to improving control rod modeling and to studying safety aspects. There has also been code development and experimental work carried out to improve the understanding of fuel performance behaviors. (author)

  7. THR-TH: a high-temperature gas-cooled nuclear reactor core thermal hydraulics code

    Energy Technology Data Exchange (ETDEWEB)

    Vondy, D.R.

    1984-07-01

    The ORNL version of PEBBLE, the (RZ) pebble bed thermal hydraulics code, has been extended for application to a prismatic gas cooled reactor core. The supplemental treatment is of one-dimensional coolant flow in up to a three-dimensional core description. Power density data from a neutronics and exposure calculation are used as the basic information for the thermal hydraulics calculation of heat removal. Two-dimensional neutronics results may be expanded for a three-dimensional hydraulics calculation. The geometric description for the hydraulics problem is the same as used by the neutronics code. A two-dimensional thermal cell model is used to predict temperatures in the fuel channel. The capability is available in the local BOLD VENTURE computation system for reactor core analysis with capability to account for the effect of temperature feedback by nuclear cross section correlation. Some enhancements have also been added to the original code to add pebble bed modeling flexibility and to generate useful auxiliary results. For example, an estimate is made of the distribution of fuel temperatures based on average and extreme conditions regularly calculated at a number of locations.

  8. Experimental simulation of fragmentation and stratification of core debris on the core catcher of a fast breeder reactor

    Energy Technology Data Exchange (ETDEWEB)

    Pillai, Dipin S.; Vignesh, R. [Indian Institute of Technology, Chennai, Tamil Nadu (India); Sudha, A. Jasmin, E-mail: jasmin@igcar.gov.in [Safety Engineering Division, Reactor Design Group, Indira Gandhi Centre for Atomic Research, Kalpakkam 603 102, Tamil Nadu (India); Pushpavanam, S.; Sundararajan, T. [Indian Institute of Technology, Chennai, Tamil Nadu (India); Nashine, B.K.; Selvaraj, P. [Safety Engineering Division, Reactor Design Group, Indira Gandhi Centre for Atomic Research, Kalpakkam 603 102, Tamil Nadu (India)

    2016-05-15

    Highlights: • Fragmentation of two simultaneous metals jets in a bulk coolant analysed. • Particle size from experiments compared with theoretical analysis. • Jet breakup modes explained using dimensionless numbers. • Settling aspects of aluminium and lead debris on collector plate studied. • Results analysed in light of core debris settling on core catcher in a FBR. - Abstract: The complex and coupled phenomena of two simultaneous molten metal jets fragmenting inside a quiescent liquid pool and settling on a collector plate are experimentally analysed in the context of safety analysis of a fast breeder reactor (FBR) in the post accident heat removal phase. Following a hypothetical core melt down accident in a FBR, a major portion of molten nuclear fuel and clad/structural material which are collectively termed as ‘corium’ undergoes fragmentation in the bulk coolant sodium in the lower plenum of the reactor main vessel and settles on the core catcher plate. The coolability of this decay heat generating debris bed is dependent on the particle size distribution and its layering i.e., stratification. Experiments have been conducted with two immiscible molten metals of different densities poured inside a coolant medium to understand their fragmentation behaviour and to assess the possibility of formation of a stratified debris bed. Molten aluminium and lead have been used as simulants in place of molten stainless steel and nuclear fuel to facilitate easy handling. This paper summarizes the major findings from these experiments. The fragmentation of the two molten metals are explained in the light of relevant dimensionless numbers such as Reynolds number and Weber Number. The mass median diameter of the fragmented debris is predicted from nonlinear stability analysis of slender jets for lead jet and using Rayleigh's classical theory of jet breakup for aluminium jet. The agreement of the predicted values with the experimental results is good. These

  9. Core/Combustor Noise - Research Overview

    Science.gov (United States)

    Hultgren, Lennart S.

    2017-01-01

    Contributions from the combustor to the overall propulsion noise of civilian transport aircraft are starting to become important due to turbofan design trends and advances in mitigation of other noise sources. Future propulsion systems for ultra-efficient commercial air vehicles are projected to be of increasingly higher bypass ratio from larger fans combined with much smaller cores, with ultra-clean burning fuel-flexible combustors. Unless effective noise-reduction strategies are developed, combustor noise is likely to become a prominent contributor to overall airport community noise in the future. This presentation gives a brief overview of the NASA outlook on pertinent issues and far-term research needs as well as current and planned research in the core/combustor-noise area. The research described herein is aligned with the NASA Ultra-Efficient Commercial Transport strategic thrust and is supported by the NASA Advanced Air Vehicle Program, Advanced Air Transport Technology Project, under the Aircraft Noise Reduction Subproject. The overarching goal of the Advanced Air Transport Technology (AATT) Project is to explore and develop technologies and concepts to revolutionize the energy efficiency and environmental compatibility of fixed wing transport aircrafts. These technological solutions are critical in reducing the impact of aviation on the environment even as this industry and the corresponding global transportation system continue to grow.

  10. Neutronic and thermal-hydraulic coupling for 3D reactor core modeling combining MCB and fluent

    Directory of Open Access Journals (Sweden)

    Królikowski Igor P.

    2015-09-01

    Full Text Available Three-dimensional simulations of neutronics and thermal hydraulics of nuclear reactors are a tool used to design nuclear reactors. The coupling of MCB and FLUENT is presented, MCB allows to simulate neutronics, whereas FLUENT is computational fluid dynamics (CFD code. The main purpose of the coupling is to exchange data such as temperature and power profile between both codes. Temperature required as an input parameter for neutronics is significant since cross sections of nuclear reactions depend on temperature. Temperature may be calculated in thermal hydraulics, but this analysis needs as an input the power profile, which is a result from neutronic simulations. Exchange of data between both analyses is required to solve this problem. The coupling is a better solution compared to the assumption of estimated values of the temperatures or the power profiles; therefore the coupled analysis was created. This analysis includes single transient neutronic simulation and several steady-state thermal simulations. The power profile is generated in defined points in time during the neutronic simulation for the thermal analysis to calculate temperature. The coupled simulation gives information about thermal behavior of the reactor, nuclear reactions in the core, and the fuel evolution in time. Results show that there is strong influence of neutronics on thermal hydraulics. This impact is stronger than the impact of thermal hydraulics on neutronics. Influence of the coupling on temperature and neutron multiplication factor is presented. The analysis has been performed for the ELECTRA reactor, which is lead-cooled fast reactor concept, where the coolant fl ow is generated only by natural convection

  11. Hafnium as a neutron absorber in research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Jatuff, F.E.; Villarino, E.A. [INVAP S.E., S.C. de Bariloche (Argentina)

    1995-12-31

    Research reactors usually require the periodic replacement of control rods because of the burnup of the neutron absorber material, even if devices have been mainly withdrawn. The purpose of this paper is the study of the reactivity worth of hafnium as a neutron absorber in the multi-purpose-type research reactor (MPTRR). The MPTRR design develops 20 MW, and it is configured by five rows of six low-enriched uranium (200% enrichment) materials test reactor fuels inside a Zircaloy chimney. The standard control plates are given by two rows of three plates of silver-indium-cadmium alloy (85% silver, 10% indium, and 5% cadmium, in weight percents; 0.26 cm of maximum absorber width), separating the central three rows of fuel elements from the peripheral first and fifth rows. The burnup of control plates decreases the reactivity worth below the licensing criteria in {approximately}6 months.

  12. Axial power distribution calculation using a neural network in the nuclear reactor core

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Y. H.; Cha, K. H.; Lee, S. H. [Korea Electric Power Research Institute, Taejon (Korea, Republic of)

    1997-12-31

    This paper is concerned with an algorithm based on neural networks to calculate the axial power distribution using excore detector signals in the nuclear reactor core. The fundamental basis of the algorithm is that the detector response can be fairly accurately estimated using computational codes. In other words, the training set, which represents relationship between detector signals and axial power distributions, for the neural network can be obtained through calculations instead of measurements. Application of the new method to the Yonggwang nuclear power plant unit 3 (YGN-3) shows that it is superior to the current algorithm in place. 7 refs., 4 figs. (Author)

  13. BOLD VENTURE COMPUTATION SYSTEM for nuclear reactor core analysis, Version III

    Energy Technology Data Exchange (ETDEWEB)

    Vondy, D.R.; Fowler, T.B.; Cunningham, G.W. III.

    1981-06-01

    This report is a condensed documentation for VERSION III of the BOLD VENTURE COMPUTATION SYSTEM for nuclear reactor core analysis. An experienced analyst should be able to use this system routinely for solving problems by referring to this document. Individual reports must be referenced for details. This report covers basic input instructions and describes recent extensions to the modules as well as to the interface data file specifications. Some application considerations are discussed and an elaborate sample problem is used as an instruction aid. Instructions for creating the system on IBM computers are also given.

  14. Development of core fuel management code system for WWER-type reactors

    Institute of Scientific and Technical Information of China (English)

    2006-01-01

    In this article, a core fuel management program for hexagonal pressurized water type WWER reactors (CFMHEX) has been developed, which is based on advanced three-dimensional nodal method and integrated with thermal hydraulic code to realize the coupling of neutronics and thermal-hydraulics. In CFMHEX, all these feedback effects such as burnup, power distribution, moderator density, and control rod insertion are considered. The verification and validation of the code system have been examined through the IAEA WWER-1000-type Kalinin NPP benchmark problem. The numerical results are in good agreement with measurements and are close to those of other international institutes.

  15. Reactor pressure vessel structural integrity research

    Energy Technology Data Exchange (ETDEWEB)

    Pennell, W.E.; Corwin, W.R. [Oak Ridge National Lab., TN (United States)

    1995-04-01

    Development continues on the technology used to assess the safety of irradiation-embrittled nuclear reactor pressure vessels (RPVs) containing flaws. Fracture mechanics tests on RPV steel, coupled with detailed elastic-plastic finite-element analyses of the crack-tip stress fields, have shown that (1) constraint relaxation at the crack tip of shallows surface flaws results in increased data scatter but no increase in the lower-bound fracture toughness, (2) the nil ductility temperature (NDT) performs better than the reference temperature for nil ductility transition (RT{sub NDT}) as a normalizing parameter for shallow-flaw fracture toughness data, (3) biaxial loading can reduce the shallow-flaw fracture toughness, (4) stress-based dual-parameter fracture toughness correlations cannot predict the effect of biaxial loading on a shallow-flaw fracture toughness because in-plane stresses at the crack tip are not influenced by biaxial loading, and (5) an implicit strain-based dual-parameter fracture toughness correlation can predict the effect of biaxial loading on shallow-flaw fracture toughness. Experimental irradiation investigations have shown that (1) the irradiation-induced shift in Charpy V-notch vs temperature behavior may not be adequate to conservatively assess fracture toughness shifts due to embrittlement, and (2) the wide global variations of initial chemistry and fracture properties of a nominally uniform material within a pressure vessel may confound accurate integrity assessments that require baseline properties.

  16. Verification of HELIOS/MASTER Nuclear Analysis System for SMART Research Reactor, Rev. 1.0

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Kyung Hoon; Kim, Kang Seog; Cho, Jin Young; Lee, Chung Chan; Zee, Sung Quun

    2005-12-15

    Nuclear design for the SMART reactor is performed by using the transport lattice code HELIOS and the core analysis code MASTER. HELIOS code developed by Studsvik Scandpower in Norway is a transport lattice code for the neutron and gamma behavior, and is used to generate few group constants. MASTER code is a nodal diffusion code developed by KAERI, and is used to analyze reactor physics. This nuclear design code package requires verification. Since the SMART reactor is unique, it is impossible to verify this code system through the comparison of the calculation results with the measured ones. Therefore, the uncertainties for the nuclear physics parameters calculated by HELIOS/MASTER have been evaluated indirectly. Since Monte Carlo calculation includes least approximations an assumptions to simulate a neutron behavior, HELIOS/MASTER has been verified by this one. Monte Carlo code has been verified by the Kurchatov critical experiments similar to SMART reactor, and HELIOS/MASTER code package has been verified by Monte Carlo calculations for the SMART research reactor.

  17. Verification of HELIOS/MASTER Nuclear Analysis System for SMART Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Kang Seog; Cho, Jin Young; Lee, Chung Chan; Zee, Sung Quun

    2005-07-15

    Nuclear design for the SMART reactor is performed by using the transport lattice code HELIOS and the core analysis code MASTER. HELIOS code developed by Studsvik Scandpower in Norway is a transport lattice code for the neutron and gamma behavior, and is used to generate few group constants. MASTER code is a nodal diffusion code developed by KAERI, and is used to analyze reactor physics. This nuclear design code package requires verification. Since the SMART reactor is unique, it is impossible to verify this code system through the comparison of the calculation results with the measured ones. Therefore, the uncertainties for the nuclear physics parameters calculated by HELIOS/MASTER have been evaluated indirectly. Since Monte Carlo calculation includes least approximations an assumptions to simulate a neutron behavior, HELIOS/MASTER has been verified by this one. Monte Carlo code has been verified by the Kurchatov critical experiments similar to SMART reactor, and HELIOS/MASTER code package has been verified by Monte Carlo calculations for the SMART research reactor.

  18. Preliminary conceptual design for electrical and I and C system of a new research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Hoan Sung; Kim, Y. K.; Kim, M. J.; Kim, H. K.; Ryu, J. S

    2004-01-01

    The core type and the process system design will be varied according to the reactor's application and capacity. A New research reactor is being designed by KAERI since 2002 and the process systems are not fixed yet. But control and instrument systems are similar to each other even though the application and the size are not same. So the C and I system that encompasses reactor protection system, reactor control system, and computer system was designed conceptually according to the requirements based on new digital technology and HANARO's proven design. The plant electrical system consists of off-site system that delivers bulk electrical power to the reactor site and on-site system that distributes and controls electrical power at the facility. The electrical system includes building service system that consist of lighting, communication, fire detection, grounding, cathodic protection, etc. also. This report describes the design requirements of on-site and off-site electric power system that set up from the codes and standards and the conceptual design based on the design requirements.

  19. A novel concept for CRIEC-driven subcritical research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Nieto, M.; Miley, G.H. [Illinois Univ., Fusion Studies Lab., Dept. of Nuclear, Plasma, and Radiological Engineering, Urbana, IL (United States)

    2001-07-01

    A novel scheme is proposed to drive a low-power subcritical fuel assembly by means of a long Cylindrical Radially-convergent Inertial Electrostatic Confinement (CRIEC) used as a neutron source. The concept is inherently safe in the sense that the fuel assembly remains subcritical at all times. Previous work has been done for the possible implementation of CRIEC as a subcritical assembly driver for power reactors. However, it has been found that the present technology and stage of development of IEC-based neutron sources can not meet the neutron flux requirements to drive a system as big as a power reactor. Nevertheless, smaller systems, such as research and training reactors, could be successfully driven with levels of neutron flux that seem more reasonable to be achieved in the near future by IEC devices. The need for custom-made expensive nuclear fission fuel, as in the case of the TRIGA reactors, is eliminated, and the CRIEC presents substantial advantages with respect to the accelerator-driven subcritical reactors in terms of simplicity and cost. In the present paper, a conceptual design for a research/training CRIEC-driven subcritical assembly is presented, emphasizing the description, principle of operation and performance of the CRIEC neutron source, highlighting its advantages and discussing some key issues that require study for the implementation of this concept. (author)

  20. Generic experiments at the sump model 'Zittauer Stroemungswanne' (ZSW) for the behaviour of mineral wool in the sump and the reactor core

    Energy Technology Data Exchange (ETDEWEB)

    Alt, Soeren; Hampel, R.; Kaestner, Wolfgang [Hochschule Zittau/Goerlitz (DE)] (and others)

    2011-03-15

    The investigation of insulation debris transport, sedimentation, penetration into the reactor core and head loss build up becomes important to reactor safety research for PWR and BWR, when considering the long-term behaviour of emergency core cooling systems during loss of coolant accidents. Research projects are being performed in cooperation between the University of Applied Sciences Zittau/Goerlitz and the Helmholtz-Zentrum Dresden-Rossendorf. The projects include experimental investigations of different processes and phenomena of insulation debris in coolant flow and the development of CFD models. Generic complex experiments serve for building up a data base for the validation of models for single effects and their coupling in CFD codes. This paper includes the description of the experimental facility for complex generic experiments (ZSW), an overview about experimental boundary conditions and results for upstream and down-stream phenomena as well as for the long-time behaviour due to corrosive processes. (orig.)

  1. Power up-grading study for the first Egyptian research reactor

    Energy Technology Data Exchange (ETDEWEB)

    El-Sawy Temraz, H.; Ashoub, N. E-mail: nageeb@pcn.aea.sci.eg; Fathallah, A

    2001-09-01

    In the present work, power up-grading study is performed, for the first Egyptian Research Reactor (ET-RR-1), using the present fuel basket with 4x4 fuel rods, (17.5 mm pitch), and a proposed fuel basket with 5x5 fuel rods, (14.0 mm pitch), without violating the thermal hydraulic safety criteria. These safety criteria are; fuel centerline temperature (fuel melting), clad surface temperature (surface boiling), outlet coolant temperature, and maximum heat flux (critical heat flux ratio). Different thermal reactor powers (2-10 MW) and different core coolant flow rates (450, 900, 1350 m{sup 3} h{sup -1}) are considered. The thermal hydraulic analysis was performed using the subchannel code COBRA-IIIC for the estimation of temperatures, coolant velocities and critical heat flux. The neutronic calculations were performed using WIMS-D4 code with 5-group neutron cross section library. These cross sections were adapted to use in the two-dimensional (2-D) diffusion code DIXY for core calculations. The study concluded that ET-RR-1 power can be upgraded safely up to 4 MW with the present 4x4-fuel basket and with the proposed 5x5-fuel basket up to 5 MW with the present coolant flow rate (900 m{sup 3} h{sup -1}). With the two fuel arrays, the reactor power can be upgraded to 6 MW with coolant flow rate of 1350 m{sup 3} h{sup -1} without violating the safety criterion. It is also concluded that, loading the ET-RR-1 core with the proposed fuel basket (5x5) increases the excess reactivity of the reactor core than the present 4x4 fuel matrix with equal U-235 mass load and gave better fuel economy of fuel utilization.

  2. Reactor Physics Methods and Preconceptual Core Design Analyses for Conversion of the Advanced Test Reactor to Low-Enriched Uranium Fuel Annual Report for Fiscal Year 2012

    Energy Technology Data Exchange (ETDEWEB)

    David W. Nigg; Sean R. Morrell

    2012-09-01

    Under the current long-term DOE policy and planning scenario, both the ATR and the ATRC will be reconfigured at an appropriate time within the next several years to operate with low-enriched uranium (LEU) fuel. This will be accomplished under the auspices of the Reduced Enrichment Research and Test Reactor (RERTR) Program, administered by the DOE National Nuclear Security Administration (NNSA). At a minimum, the internal design and composition of the fuel element plates and support structure will change, to accommodate the need for low enrichment in a manner that maintains total core excess reactivity at a suitable level for anticipated operational needs throughout each cycle while respecting all control and shutdown margin requirements and power distribution limits. The complete engineering design and optimization of LEU cores for the ATR and the ATRC will require significant multi-year efforts in the areas of fuel design, development and testing, as well as a complete re-analysis of the relevant reactor physics parameters for a core composed of LEU fuel, with possible control system modifications. Ultimately, revalidation of the computational physics parameters per applicable national and international standards against data from experimental measurements for prototypes of the new ATR and ATRC core designs will also be required for Safety Analysis Report (SAR) changes to support routine operations with LEU. This report is focused on reactor physics analyses conducted during Fiscal Year (FY) 2012 to support the initial development of several potential preconceptual fuel element designs that are suitable candidates for further study and refinement during FY-2013 and beyond. In a separate, but related, effort in the general area of computational support for ATR operations, the Idaho National Laboratory (INL) is conducting a focused multiyear effort to introduce modern high-fidelity computational reactor physics software and associated validation protocols to replace

  3. RETU. The Finnish research programme on reactor safety. Interim report 1995 - May 1997

    Energy Technology Data Exchange (ETDEWEB)

    Vanttola, T.; Puska, E.K. [VTT Energy, Espoo (Finland). Nuclear Energy] [eds.

    1997-08-01

    The Finnish national research programme on Reactor Safety (RETU, 1995-1998) concentrates on the search of safe limits of nuclear fuel and the reactor core, accident management methods and risk management of the operation of nuclear power plants. The annual volume of the programme has been about 26 person years and the annual funding FIM 15 million. This report summarises the structure and objectives of the programme, research fields included and the main results obtained during the period 1995 - May 1997. In the field of operational margins of a nuclear reactor, the behaviour of high burnup nuclear fuel is studied both in normal operation and during power transients. The static and dynamic reactor analysis codes are developed and validated to cope with new fuel designs and complicated three-dimensional reactivity transients and accidents. Research on accident management aims at development and validation of calculation methods needed to plan preventive measures and to train the personnel to severe accident mitigation. Other goals are to reduce uncertainties in phenomena important in severe accidents and to study actions planned for accident management. In the field of risk management probabilistic methods are developed for safety related decision making and for complex phenomena and event sequences. Effects of maintenance on nuclear power plant safety are studied and more effective methods for the assessment of human reliability and safety critical organisations are searched. 135 refs.

  4. An assessment of coupling algorithms for nuclear reactor core physics simulations

    Science.gov (United States)

    Hamilton, Steven; Berrill, Mark; Clarno, Kevin; Pawlowski, Roger; Toth, Alex; Kelley, C. T.; Evans, Thomas; Philip, Bobby

    2016-04-01

    This paper evaluates the performance of multiphysics coupling algorithms applied to a light water nuclear reactor core simulation. The simulation couples the k-eigenvalue form of the neutron transport equation with heat conduction and subchannel flow equations. We compare Picard iteration (block Gauss-Seidel) to Anderson acceleration and multiple variants of preconditioned Jacobian-free Newton-Krylov (JFNK). The performance of the methods are evaluated over a range of energy group structures and core power levels. A novel physics-based approximation to a Jacobian-vector product has been developed to mitigate the impact of expensive on-line cross section processing steps. Numerical simulations demonstrating the efficiency of JFNK and Anderson acceleration relative to standard Picard iteration are performed on a 3D model of a nuclear fuel assembly. Both criticality (k-eigenvalue) and critical boron search problems are considered.

  5. Determination and variation of core bacterial community in a two-stage full-scale anaerobic reactor treating high-strength pharmaceutical wastewater.

    Science.gov (United States)

    Ma, Haijun; Ye, Lin; Hu, Haidong; Zhang, Lulu; Ding, Lili; Ren, Hongqiang

    2017-08-25

    The functional characterization and temporal variation of anaerobic bacterial population is important to better understanding of microbial process of two-stage anaerobic reactor. However, due to the high diversity of anaerobic bacteria, close attention should be prioritized to be paid to the frequently abundant bacteria that were defined as core bacteria and putatively functionally important. Here in this study, using Miseq sequencing technology, the core bacterial community of 98 operational taxonomic units (OTUs) was determined in a two-stage upflow blanket filter reactors treating pharmaceutical wastewater. The core bacterial community accounted for 61.66% of the total sequences and accurately predicted the sample location in the principal coordinates analysis (PCoA) scatter plot as the total bacterial OTUs did. The core bacterial community in the first-stage (FS) and second-stage (SS) reactors were generally distinct that FS core bacterial community was indicated to be more related to higher-level fermentation process and SS core bacterial community contained more microbes in syntrophic cooperation with methanogens. Moreover, the different responses of FS and SS core bacterial community to the temperature shock and influent disturbance caused by solid contamination were fully investigated. Co-occurring analysis at the order level implied that Bacteroidales, Selenomonadales, Anaerolineales, Syneristales and Thermotogales might play keystone roles in anaerobic digestion due to their high abundance and tight correlation with other microbes. These findings advanced our knowledge about the core bacteria community and its temporal variability for future comparative research and the improvement of the two-stage anaerobic system operation.

  6. China Advanced Research Reactor Project Progress in 2012

    Institute of Scientific and Technical Information of China (English)

    ZHAO; Tie-jun

    2012-01-01

    <正>In 2012, all the commissioning for the China Advanced Research Reactor (CARR) had been finished and the diffraction pattern had been successfully obtained on the neutron scattering spectrometer. Meanwhile, the cold neutron source project and the acceptance items of CARR project had been carrying out.

  7. Kartini Research Reactor prospective studies for neutron scattering application

    Energy Technology Data Exchange (ETDEWEB)

    Widarto [Yogyakarta Nuclear Research Center, BATAN (Indonesia)

    1999-10-01

    The Kartini Research Reactor (KRR) is located in Yogyakarta Nuclear Research Center, Yogyakarta - Indonesia. The reactor is operated for 100 kW thermal power used for research, experiments and training of nuclear technology. There are 4 beam ports and 1 column thermal are available at the reactor. Those beam ports have thermal neutron flux around 10{sup 7} n/cm{sup 2}s each other and used for sub critical assembly, neutron radiography studies and Neutron Activation Analysis (NAA). Design of neutron collimator has been done for piercing radial beam port and the calculation result of collimated neutron flux is around 10{sup 9} n/cm{sup 2}s. This paper describes experiment facilities and parameters of the Kartini research reactor, and further more the prospective studies for neutron scattering application. The purpose of this paper is to optimize in utilization of the beam ports facilities and enhance the manpower specialty. The special characteristic of the beam ports and preliminary studies, pre activities regarding with neutron scattering studies for KKR is presented. (author)

  8. Reduced enrichment for research and test reactors: Proceedings

    Energy Technology Data Exchange (ETDEWEB)

    1993-07-01

    The 15th annual Reduced Enrichment for Research and Test Reactors (RERTR) international meeting was organized by Ris{o} National Laboratory in cooperation with the International Atomic Energy Agency and Argonne National Laboratory. The topics of the meeting were the following: National Programs, Fuel Fabrication, Licensing Aspects, States of Conversion, Fuel Testing, and Fuel Cycle. Individual papers have been cataloged separately.

  9. Numerical simulation of a Hypothetical Core Disruptive Accident in a small-scale model of a nuclear reactor

    Energy Technology Data Exchange (ETDEWEB)

    Robbe, M.F. E-mail: robbe@aquilon.cea.frmfrobbe@cea.fr; Lepareux, M.; Treille, E.; Cariou, Y

    2003-08-01

    In the case of a Hypothetical Core Disruptive Accident (HCDA) in a Liquid Metal Fast Breeder Reactor, it is assumed that the core of the nuclear reactor has melted partially and that the chemical interaction between molten fuel and liquid sodium has created a high-pressure gas bubble in the core. The violent expansion of this bubble loads and deforms the reactor vessel and the internal structures, thus endangering the safety of the nuclear plant. The MARA 10 experimental test simulates a HCDA in a 1/30-scale mock-up schematising a reactor block. In the mock-up, the liquid sodium cooling the reactor core is replaced by water and the argon blanket laying below the reactor roof is simulated by an air blanket. The explosion is triggered by an explosive charge. This paper presents a numerical simulation of the test with the EUROPLEXUS code and an analysis of the computed results. In particular, the evolution of the fluid flows and the deformations of the internal and external structures are analysed in detail. Finally, the current computed results are compared with the experimental ones and with previous numerical results computed with the SIRIUS and CASTEM-PLEXUS codes.

  10. Comparison between different flux traps assembled in the core of the nuclear reactor IPEN/MB-01 by measuring of the thermal and epithermal neutron fluxes using activation foils

    Energy Technology Data Exchange (ETDEWEB)

    Mura, Luiz Ernesto Credidio; Bitelli, Ulysses d' Utra; Mura, Luis Felipe Liambos; Carluccio, Thiago; Andrade, Graciete Simoes de, E-mail: ubitelli@ipen.b, E-mail: gsasilva@ipen.b [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2011-07-01

    The production of radioisotopes is one of the most important applications of nuclear research reactors. This study investigated a method called Flux Trap, which is used to increase the yield of production of radioisotopes in nuclear reactors. The method consists in the rearrangement of the fuel rods to allow the increase of the thermal neutron flux in the irradiation region inside the reactor core, without changing the standard reactor power level. Various configurations were assembled with the objective of finding the configuration with the highest thermal neutron flux in the region of irradiation. The method of activation analysis was used to measure the thermal neutron flux and determine the most efficient reactor core configuration . It was found that there was an increase in the thermal neutron flux of 337% in the most efficient configuration, which demonstrates the effectiveness of the method. (author)

  11. COREMAP: Graphical user interface for displaying reactor core data in an interactive hexagon map

    Energy Technology Data Exchange (ETDEWEB)

    Muscat, F.L.; Derstine, K.L.

    1995-06-01

    COREMAP is a Graphical User Interface (GUI) designed to assist users read and check reactor core data from multidimensional neutronic simulation models in color and/or as text in an interactive 2D planar grid of hexagonal subassemblies. COREMAP is a complete GEODST/RUNDESC viewing tool which enables the user to access multi data set files (e.g. planes, moments, energy groups ,... ) and display up to two data sets simultaneously, one as color and the other as text. The user (1) controls color scale characteristics such as type (linear or logarithmic) and range limits, (2) controls the text display based upon conditional statements on data spelling, and value. (3) chooses zoom features such as core map size, number of rings and surrounding subassemblies, and (4) specifies the data selection for supplied popup subwindows which display a selection of data currently off-screen for a selected cell, as a list of data and/or as a graph. COREMAP includes a RUNDESC file editing tool which creates ``proposed`` Run-description files by point and click revisions to subassembly assignments in an existing EBRII Run-description file. COREMAP includes a fully automated printing option which creates high quality PostScript color or greyscale images of the core map independent of the monitor used, e.g. color prints can be generated with a session from a color or monochrome monitor. The automated PostScript output is an alternative to the xgrabsc based printing option. COREMAP includes a plotting option which creates graphs related to a selected cell. The user specifies the X and Y coordinates types (planes, moment, group, flux ,... ) and a parameter, P, when displaying several curves for the specified (X, Y) pair COREMAP supports hexagonal geometry reactor core configurations specified by: the GEODST file and binary Standard Interface Files and the RUNDESC ordering.

  12. Thermal hydraulics characterization of the core and the reactor vessel type BWR; Caracterizacion termohidraulica del nucleo y de la vasija de un reactor tipo BWR

    Energy Technology Data Exchange (ETDEWEB)

    Zapata Y, M.; Lopez H, L.E. [CFE, Carretera Cardel-Nautla Km. 42.5, Municipio Alto Lucero, Veracruz (Mexico)]. e-mail: marxlenin.zapata@cfe.gob.mx

    2008-07-01

    The thermal hydraulics design of a reactor type BWR 5 as the employees in the nuclear power plant of Laguna Verde involves the coupling of at least six control volumes: Pumps jet region, Stratification region, Core region, Vapor dryer region, Humidity separator region and Reactor region. Except by the regions of the core and reactor, these control volumes only are used for design considerations and their importance as operative data source is limited. It is for that is fundamental to complement the thermal hydraulics relations to obtain major data that allow to determine the efficiency of internal components, such as pumps jet, humidity separator and vapor dryer. Like example of the previous thing, calculations are realized on the humidity of the principal vapor during starting, comparing it with the values at the moment incorporated in the data banks of the computers of process of both units. (Author)

  13. Research and development studies on the seismic behaviour of the PEC fast reactor (safety analysis detailed report no. 8)

    Energy Technology Data Exchange (ETDEWEB)

    Martelli, A.; Forni, M.; Masoni, P.; Maresca, G.; Castoldi, A.; Muzzi, F. (ENEA, Rome (Italy); Ansaldo Spa, Genoa (Italy); ISMES Spa, Bergamo (Italy))

    1988-01-15

    This paper presents the main features and results of the numerical and experimental studies that were carried out by ENEA (Italian Commission for Alternative Energy Sources) for the seismic verification of the Italian PEC fast reactor test facility. More precisely, the paper focuses on the wide-ranging research and development programme that has been performed (and recently completed) on the reactor building, the reactor-block, the main vessel, the core and the shutdown system. The needs of these detailed studies are stressed and the feed-backs on the design, necessary safisfy the seismic safety requirements, are recalled. The general validity of the analyses in the framework of the research and development activities for nuclear reactor is also pointed out.

  14. Fuel shuffling optimization for the Delft research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Geemert, R. van; Hoogenboom, J.E.; Gibcus, H.P.M. [Delft Univ. of Technology, Interfaculty Reactor Inst., Delft (Netherlands); Quist, A.J. [Delft Univ., Fac. of Applied Mathematics and Informatics, Delft (Netherlands)

    1997-07-01

    A fuel shuffling optimization procedure is proposed for the Hoger Onderwijs Reactor (HOR) in Delft, the Netherlands, a 2 MWth swimming-pool type research reactor. In order to cope with the fluctuatory behaviour of objective functions in loading pattern optimization, the proposed cyclic permutation optimization procedure features a gradual transition from global to local search behaviour via the introduction of stochastic tests for the number of fuel assemblies involved in a cyclic permutation. The possible objectives and the safety and operation constraints, as well as the optimization procedure, are discussed, followed by some optimization results for the HOR. (author)

  15. Nonlinear control for core power of pressurized water nuclear reactors using constant axial offset strategy

    Directory of Open Access Journals (Sweden)

    Gholam Reza Ansarifar

    2015-12-01

    Full Text Available One of the most important operations in nuclear power plants is load following, in which an imbalance of axial power distribution induces xenon oscillations. These oscillations must be maintained within acceptable limits otherwise the nuclear power plant could become unstable. Therefore, bounded xenon oscillation is considered to be a constraint for the load following operation. In this paper, the design of a sliding mode control (SMC, which is a robust nonlinear controller, is presented. SMC is a means to control pressurized water nuclear reactor (PWR power for the load following operation problem in a way that ensures xenon oscillations are kept bounded within acceptable limits. The proposed controller uses constant axial offset (AO strategy to ensure xenon oscillations remain bounded. The constant AO is a robust state constraint for the load following problem. The reactor core is simulated based on the two-point nuclear reactor model with a three delayed neutron groups. The stability analysis is given by means of the Lyapunov approach, thus the control system is guaranteed to be stable within a large range. The employed method is easy to implement in practical applications and moreover, the SMC exhibits the desired dynamic properties during the entire output-tracking process independent of perturbations. Simulation results are presented to demonstrate the effectiveness of the proposed controller in terms of performance, robustness, and stability. Results show that the proposed controller for the load following operation is so effective that the xenon oscillations are kept bounded in the given region.

  16. Nonlinear control for core power of pressurized water nuclear reactors using constant axial offset strategy

    Energy Technology Data Exchange (ETDEWEB)

    Ansarifar, Gholam Reza; Saadatzi, Saeed [Dept. of Nuclear Engineering, Faculty of Advanced Sciences and Technology, University of Isfahan, Isfahan (Iran, Islamic Republic of)

    2015-12-15

    One of the most important operations in nuclear power plants is load following, in which an imbalance of axial power distribution induces xenon oscillations. These oscillations must be maintained within acceptable limits otherwise the nuclear power plant could become unstable. Therefore, bounded xenon oscillation is considered to be a constraint for the load following operation. In this paper, the design of a sliding mode control (SMC), which is a robust nonlinear controller, is presented. SMC is a means to control pressurized water nuclear reactor (PWR) power for the load following operation problem in a way that ensures xenon oscillations are kept bounded within acceptable limits. The proposed controller uses constant axial offset (AO) strategy to ensure xenon oscillations remain bounded. The constant AO is a robust state constraint for the load following problem. The reactor core is simulated based on the two-point nuclear reactor model with a three delayed neutron groups. The stability analysis is given by means of the Lyapunov approach, thus the control system is guaranteed to be stable within a large range. The employed method is easy to implement in practical applications and moreover, the SMC exhibits the desired dynamic properties during the entire output-tracking process independent of perturbations. Simulation results are presented to demonstrate the effectiveness of the proposed controller in terms of performance, robustness, and stability. Results show that the proposed controller for the load following operation is so effective that the xenon oscillations are kept bounded in the given region.

  17. High temperature UF6 RF plasma experiments applicable to uranium plasma core reactors

    Science.gov (United States)

    Roman, W. C.

    1979-01-01

    An investigation was conducted using a 1.2 MW RF induction heater facility to aid in developing the technology necessary for designing a self critical fissioning uranium plasma core reactor. Pure, high temperature uranium hexafluoride (UF6) was injected into an argon fluid mechanically confined, steady state, RF heated plasma while employing different exhaust systems and diagnostic techniques to simulate and investigate some potential characteristics of uranium plasma core nuclear reactors. The development of techniques and equipment for fluid mechanical confinement of RF heated uranium plasmas with a high density of uranium vapor within the plasma, while simultaneously minimizing deposition of uranium and uranium compounds on the test chamber peripheral wall, endwall surfaces, and primary exhaust ducts, is discussed. The material tests and handling techniques suitable for use with high temperature, high pressure, gaseous UF6 are described and the development of complementary diagnostic instrumentation and measurement techniques to characterize the uranium plasma, effluent exhaust gases, and residue deposited on the test chamber and exhaust system components is reported.

  18. Approaches to optimization of core reactivity coefficirnts for the “MASTER” heat supply reactor

    Directory of Open Access Journals (Sweden)

    D.M. Titov

    2015-09-01

    Full Text Available After increasing the power output of heat supply reactor «MASTER» by insertion of the annular channel with coolant, feedback coefficients are deteriorated. Thereby, there was need to find ways for changing reactivity coefficients in new reactor design and at the same time to save natural circulation, low core pressure and outlet core temperature of coolant. Reactivity coefficients have been calculated depending on width and locations radius of annular coolant channel at once to fuel enrichment. Neutron-physical code WIMS-D4 was used as calculation tool. The results showed that the feedback coefficients optimum can be achieved by reducing of annular channel width and increasing of fuel enrichment. At the same time reactivity coefficients are insensitive to location of annular coolant channel radius changes. Restrictions for fuel enrichment (IAEA requirements coupled with geometry restrictions of annular channel listed above (impossible to remove the thermal power or significant increasing of heat exchangers height have shown that prospect of feedbacks improving via width and location of annular channel is used up. Possible improvements can be achieved by changing type of burnable poison and neutron spectrum.

  19. Hybrid parallel code acceleration methods in full-core reactor physics calculations

    Energy Technology Data Exchange (ETDEWEB)

    Courau, T.; Plagne, L.; Ponicot, A. [EDF R and D, 1, Avenue du General de Gaulle, 92141 Clamart Cedex (France); Sjoden, G. [Nuclear and Radiological Engineering, Georgia Inst. of Technology, Atlanta, GA 30332 (United States)

    2012-07-01

    When dealing with nuclear reactor calculation schemes, the need for three dimensional (3D) transport-based reference solutions is essential for both validation and optimization purposes. Considering a benchmark problem, this work investigates the potential of discrete ordinates (Sn) transport methods applied to 3D pressurized water reactor (PWR) full-core calculations. First, the benchmark problem is described. It involves a pin-by-pin description of a 3D PWR first core, and uses a 8-group cross-section library prepared with the DRAGON cell code. Then, a convergence analysis is performed using the PENTRAN parallel Sn Cartesian code. It discusses the spatial refinement and the associated angular quadrature required to properly describe the problem physics. It also shows that initializing the Sn solution with the EDF SPN solver COCAGNE reduces the number of iterations required to converge by nearly a factor of 6. Using a best estimate model, PENTRAN results are then compared to multigroup Monte Carlo results obtained with the MCNP5 code. Good consistency is observed between the two methods (Sn and Monte Carlo), with discrepancies that are less than 25 pcm for the k{sub eff}, and less than 2.1% and 1.6% for the flux at the pin-cell level and for the pin-power distribution, respectively. (authors)

  20. Design of a Resistively Heated Thermal Hydraulic Simulator for Nuclear Rocket Reactor Cores

    Science.gov (United States)

    Litchford, Ron J.; Foote, John P.; Ramachandran, Narayanan; Wang, Ten-See; Anghaie, Samim

    2007-01-01

    A preliminary design study is presented for a non-nuclear test facility which uses ohmic heating to replicate the thermal hydraulic characteristics of solid core nuclear reactor fuel element passages. The basis for this testing capability is a recently commissioned nuclear thermal rocket environments simulator, which uses a high-power, multi-gas, wall-stabilized constricted arc-heater to produce high-temperature pressurized hydrogen flows representative of reactor core environments, excepting radiation effects. Initially, the baseline test fixture for this non-nuclear environments simulator was configured for long duration hot hydrogen exposure of small cylindrical material specimens as a low cost means of evaluating material compatibility. It became evident, however, that additional functionality enhancements were needed to permit a critical examination of thermal hydraulic effects in fuel element passages. Thus, a design configuration was conceived whereby a short tubular material specimen, representing a fuel element passage segment, is surrounded by a backside resistive tungsten heater element and mounted within a self-contained module that inserts directly into the baseline test fixture assembly. With this configuration, it becomes possible to create an inward directed radial thermal gradient within the tubular material specimen such that the wall-to-gas heat flux characteristics of a typical fuel element passage are effectively simulated. The results of a preliminary engineering study for this innovative concept are fully summarized, including high-fidelity multi-physics thermal hydraulic simulations and detailed design features.

  1. Advanced Test Reactor Core Modeling Update Project Annual Report for Fiscal Year 2013

    Energy Technology Data Exchange (ETDEWEB)

    David W. Nigg

    2013-09-01

    Legacy computational reactor physics software tools and protocols currently used for support of Advanced Test Reactor (ATR) core fuel management and safety assurance, and to some extent, experiment management, are inconsistent with the state of modern nuclear engineering practice, and are difficult, if not impossible, to verify and validate (V&V) according to modern standards. Furthermore, the legacy staff knowledge required for effective application of these tools and protocols from the 1960s and 1970s is rapidly being lost due to staff turnover and retirements. In late 2009, the Idaho National Laboratory (INL) initiated a focused effort, the ATR Core Modeling Update Project, to address this situation through the introduction of modern high-fidelity computational software and protocols. This aggressive computational and experimental campaign will have a broad strategic impact on the operation of the ATR, both in terms of improved computational efficiency and accuracy for support of ongoing DOE programs as well as in terms of national and international recognition of the ATR National Scientific User Facility (NSUF).

  2. The scalability of OTR (out-of-core thermionic reactor) space nuclear power systems

    Energy Technology Data Exchange (ETDEWEB)

    Gallup, D.R.

    1990-03-01

    In this document, masses of the STAR-C power system and an optimized out-of-core thermionic reactor (OTR) power system versus power level are investigated. The impacts of key system parameters on system performance are also addressed. The STAR-C is mass competitive below about 15 kWe, but at higher power levels the scalability is relatively poor. An optimized OR is the least massive space nuclear power system below 25 kWe, and scales well to 50 kWe. The system parameters that have a significant impact on the scalability of the STAR-C are core thermal flux, thermionic converter efficiency, and core length to diameter ratio. The emissivity of the core surface is shown to be a relatively unimportant parameter. For an optimized OR power system, the most significant system parameter is the maximum allowable fuel temperature. It is also shown that if advanced radiation-hardened electronics are used in the satellite payload, a very large mass savings is realized. 10 refs., 23 figs., 7 tabs.

  3. Iaea Activities Supporting the Applications of Research Reactors in 2013

    Science.gov (United States)

    Peld, Nathan D.; Ridikas, Danas

    2014-02-01

    As the underutilization of research reactors around the world persists as a primary topic of concern among facility owners and operators, the IAEA responded in 2013 with a broad range of activities to address the planning, execution and improvement of many experimental techniques. The revision of two critical documents for planning and diversifying a facility's portfolio of applications, TECDOC 1234 “The Applications of Research Reactors” and TECDOC 1212 “Strategic Planning for Research Reactors”, is in progress in order to keep this information relevant, corresponding to the dynamism of experimental techniques and research capabilities. Related to the latter TECDOC, the IAEA convened a meeting in 2013 for the expert review of a number of strategic plans submitted by research reactor operators in developing countries. A number of activities focusing on specific applications are either continuing or beginning as well. In neutron activation analysis, a joint round of inter-comparison proficiency testing sponsored by the IAEA Technical Cooperation Department will be completed, and facility progress in measurement accuracy is described. Also, a training workshop in neutron imaging and Coordinated Research Projects in reactor benchmarks, automation of neutron activation analysis and neutron beam techniques for material testing intend to advance these activities as more beneficial services to researchers and other users.

  4. Yale High Energy Physics Research: Precision Studies of Reactor Antineutrinos

    Energy Technology Data Exchange (ETDEWEB)

    Heeger, Karsten M. [Yale Univ., New Haven, CT (United States)

    2014-09-13

    This report presents experimental research at the intensity frontier of particle physics with particular focus on the study of reactor antineutrinos and the precision measurement of neutrino oscillations. The experimental neutrino physics group of Professor Heeger and Senior Scientist Band at Yale University has had leading responsibilities in the construction and operation of the Daya Bay Reactor Antineutrino Experiment and made critical contributions to the discovery of non-zero$\\theta_{13}$. Heeger and Band led the Daya Bay detector management team and are now overseeing the operations of the antineutrino detectors. Postdoctoral researchers and students in this group have made leading contributions to the Daya Bay analysis including the prediction of the reactor antineutrino flux and spectrum, the analysis of the oscillation signal, and the precision determination of the target mass yielding unprecedented precision in the relative detector uncertainty. Heeger's group is now leading an R\\&D effort towards a short-baseline oscillation experiment, called PROSPECT, at a US research reactor and the development of antineutrino detectors with advanced background discrimination.

  5. Neutron transport with the method of characteristics for 3-D full core boiling water reactor applications

    Science.gov (United States)

    Thomas, Justin W.

    2006-12-01

    The Numerical Nuclear Reactor (NNR) is a code suite that is being developed to provide high-fidelity multi-physics capability for the analysis of light water nuclear reactors. The focus of the work here is to extend the capability of the NNR by incorporation of the neutronics module, DeCART, for Boiling Water Reactor (BWR) applications. The DeCART code has been coupled to the NNR fluid mechanics and heat transfer module STAR-CD for light water reactor applications. The coupling has been accomplished via an interface program, which is responsible for mapping the STAR-CD and DeCART meshes, managing communication, and monitoring convergence. DeCART obtains the solution of the 3-D Boltzmann transport equation by performing a series of 2-D modular ray tracing-based method of characteristics problems that are coupled within the framework of 3-D coarse-mesh finite difference. The relatively complex geometry and increased axial heterogeneity found in BWRs are beyond the modeling capability of the original version of DeCART. In this work, DeCART is extended in three primary areas. First, the geometric capability is generalized by extending the modular ray tracing scheme and permitting an unstructured mesh in the global finite difference kernel. Second, numerical instabilities, which arose as a result of the severe axial heterogeneity found in BWR cores, have been resolved. Third, an advanced nodal method has been implemented to improve the accuracy of the axial flux distribution. In this semi-analytic nodal method, the analytic solution to the transverse-integrated neutron diffusion equation is obtained, where the nonhomogeneous neutron source was first approximated by a quartic polynomial. The successful completion of these three tasks has allowed the application of the coupled DeCART/STAR-CD code to practical BWR problems.

  6. Nodalization effects on RELAP5 results related to MTR research reactor transient scenarios

    Directory of Open Access Journals (Sweden)

    Khedr Ahmed

    2005-01-01

    Full Text Available The present work deals with the anal y sis of RELAP5 results obtained from the evaluation study of the total loss of flow transient with the deficiency of the heat removal system in a research reactor using two different nodalizations. It focuses on the effect of nodalization on the thermal-hydraulic evaluation of the re search reactor. The analysis of RELAP5 results has shown that nodalization has a big effect on the predicted scenario of the postulated transient. There fore, great care should be taken during the nodalization of the reactor, especially when the avail able experimental or measured data are insufficient for making a complete qualification of the nodalization. Our analysis also shows that the research reactor pool simulation has a great effect on the evaluation of natural circulation flow and on other thermal-hydraulic parameters during the loss of flow transient. For example, the on set time of core boiling changes from less than 2000 s to 15000 s, starting from the beginning of the transient. This occurs if the pool is simulated by two vertical volumes in stead of one vertical volume.

  7. Measurements and analysis of critical assemblies for research reactors with mixed enrichments

    Energy Technology Data Exchange (ETDEWEB)

    Deen, J.R.; Snelgrove, J.L.; Hobbs, R.W.

    1982-01-01

    As part of the RERTR Program whole-core demonstration in the Ford Nuclear Reactor (FNR) at the University of Michigan, data have been obtained which will allow more extensive validation of neutronics methods for whole-core calculations of an equilibrium high-enriched-uranium (HEU) core and a fresh low-enriched-uranium (LEU) core. A series of experiments designed to provide the data needed for these validations has been performed in the Pool Critical Assembly (PCA) at the Oak Ridge National laboratory (ORNL). This paper reports the results of the measurements and of the subsequent validation calculations performed at ANL. Measurements were made on approximately 20 different critical configurations in the PCA during the period June 15 to 26, 1981. The normal PCA fuel elements contained high-enriched uranium (HEU, 93 wt% /sup 235/U) while the reduced-enrichment fuel elements, obtained for irradiation testing in the Oak Ridge Research Reactor (ORR) under the fuel demonstration activity of the RERTR Program, contained either medium-enriched uranium (MEU, 45 wt% /sup 235/U) or low-enriched uranium (LEU, 19.8 wt% /sup 235/U).

  8. A neutronic feasibility study for LEU conversion of the WWR-SM research reactor in Uzbekistan.

    Energy Technology Data Exchange (ETDEWEB)

    Rakhmanov, A.

    1998-10-19

    The WWR-SM research reactor in Uzbekistan has operated at 10 MW since 1979, using Russian-supplied IRT-3M fuel assemblies containing 90% enriched uranium. Burnup tests of three full-sized IRT-3M FA with 36% enrichment were successfully completed to a burn up of about {approximately}50% in 1987-1989. In August 1998, four IRT-3M FA with 36% enriched uranium were loaded into the core to initiate conversion of the entire core to 36% enriched fuel. This paper presents the results of equilibrium fuel cycle comparisons of the reactor using HEU (90%) and HEU (36%) IRT-3M fuel and compares results with the performance of IRT-4M FA containing LEU (19.75%). The results show that an LEU (19.75%) density of 3.8 g/cm{sup 3} is required to match the cycle length of the HEU (90%) core and an LEU density 3.9 g/cm{sup 3} is needed to match the cycle length of the HEU (36%) core.

  9. Comparative Study on Various Geometrical Core Design of 300 MWth Gas Cooled Fast Reactor with UN-PuN Fuel Longlife without Refuelling

    Science.gov (United States)

    Dewi Syarifah, Ratna; Su'ud, Zaki; Basar, Khairul; Irwanto, Dwi

    2017-07-01

    Nuclear power has progressive improvement in the operating performance of exiting reactors and ensuring economic competitiveness of nuclear electricity around the world. The GFR use gas coolant and fast neutron spectrum. This research use helium coolant which has low neutron moderation, chemical inert and single phase. Comparative study on various geometrical core design for modular GFR with UN-PuN fuel long life without refuelling has been done. The calculation use SRAC2006 code both PIJ calculation and CITATION calculation. The data libraries use JENDL 4.0. The variation of fuel fraction is 40% until 65%. In this research, we varied the geometry of core reactor to find the optimum geometry design. The variation of the geometry design is balance cylinder; it means that the diameter active core (D) same with height active core (H). Second, pancake cylinder (D>H) and third, tall cylinder (Dcore, when we use the balance geometry, the k-eff value flattest and more stable than the others.

  10. Dynamic neutronic and stability analysis of a burst mode, single cavity gas core reactor Brayton cycle space power system

    Science.gov (United States)

    Dugan, Edward T.; Kutikkad, Kiratadas

    The conceptual, burst-mode gaseous-core reactor (GCR) space nuclear power system presently subjected to reactor-dynamics and system stability studies operates on a closed Brayton cycle, via disk MHD generator for energy conversion. While the gaseous fuel density power coefficient of reactivity is found to be capable of rapidly stabilizing the GCR system, the power of this feedback renders standard external reactivity insertions inadequate for significant power-level changes during normal operation.

  11. Temperature oscillations near natural nuclear reactor cores and the potential for prebiotic oligomer synthesis.

    Science.gov (United States)

    Adam, Zachary R

    2016-06-01

    Geologic settings capable of driving prebiotic oligomer synthesis reactions remain a relatively unexplored aspect of origins of life research. Natural nuclear reactors are an example of Precambrian energy sources that produced unique temperature fluctuations. Heat transfer models indicate that water-moderated, convectively-cooled natural fission reactors in porous host rocks create temperature oscillations that resemble those employed in polymerase chain reaction (PCR) devices to artificially amplify oligonucleotides. This temperature profile is characterized by short-duration pulses up to 70-100 °C, followed by a sustained period of temperatures in the range of 30-70 °C, and finally a period of relaxation to ambient temperatures until the cycle is restarted by a fresh influx of pore water. For a given reactor configuration, temperature maxima and the time required to relax to ambient temperatures depend most strongly on the aggregate effect of host rock permeability in decreasing the thermal expansion and increasing the viscosity and evaporation temperature of the pore fluids. Once formed, fission-fueled reactors can sustain multi-kilowatt-level power production for 10(5)-10(6) years, ensuring microenvironmental longevity and chemical output. The model outputs indicate that organic synthesis on young planetary bodies with a sizeable reservoir of fissile material can involve more sophisticated energy dissipation pathways than modern terrestrial analog settings alone would suggest.

  12. Temperature oscillations near natural nuclear reactor cores and the potential for prebiotic oligomer synthesis

    Science.gov (United States)

    Adam, Zachary R.

    2016-06-01

    Geologic settings capable of driving prebiotic oligomer synthesis reactions remain a relatively unexplored aspect of origins of life research. Natural nuclear reactors are an example of Precambrian energy sources that produced unique temperature fluctuations. Heat transfer models indicate that water-moderated, convectively-cooled natural fission reactors in porous host rocks create temperature oscillations that resemble those employed in polymerase chain reaction (PCR) devices to artificially amplify oligonucleotides. This temperature profile is characterized by short-duration pulses up to 70-100 °C, followed by a sustained period of temperatures in the range of 30-70 °C, and finally a period of relaxation to ambient temperatures until the cycle is restarted by a fresh influx of pore water. For a given reactor configuration, temperature maxima and the time required to relax to ambient temperatures depend most strongly on the aggregate effect of host rock permeability in decreasing the thermal expansion and increasing the viscosity and evaporation temperature of the pore fluids. Once formed, fission-fueled reactors can sustain multi-kilowatt-level power production for 105-106 years, ensuring microenvironmental longevity and chemical output. The model outputs indicate that organic synthesis on young planetary bodies with a sizeable reservoir of fissile material can involve more sophisticated energy dissipation pathways than modern terrestrial analog settings alone would suggest.

  13. Needs and Requirements for Future Research Reactors (ORNL Perspectives)

    Energy Technology Data Exchange (ETDEWEB)

    Ilas, Germina [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Bryan, Chris [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Gehin, Jess C. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2016-02-10

    The High Flux Isotope Reactor (HFIR) is a vital national and international resource for neutron science research, production of radioisotopes, and materials irradiation. While HFIR is expected to continue operation for the foreseeable future, interest is growing in understanding future research reactors features, needs, and requirements. To clarify, discuss, and compile these needs from the perspective of Oak Ridge National Laboratory (ORNL) research and development (R&D) missions, a workshop, titled “Needs and Requirements for Future Research Reactors”, was held at ORNL on May 12, 2015. The workshop engaged ORNL staff that is directly involved in research using HFIR to collect valuable input on the reactor’s current and future missions. The workshop provided an interactive forum for a fruitful exchange of opinions, and included a mix of short presentations and open discussions. ORNL staff members made 15 technical presentations based on their experience and areas of expertise, and discussed those capabilities of the HFIR and future research reactors that are essential for their current and future R&D needs. The workshop was attended by approximately 60 participants from three ORNL directorates. The agenda is included in Appendix A. This document summarizes the feedback provided by workshop contributors and participants. It also includes information and insights addressing key points that originated from the dialogue started at the workshop. A general overview is provided on the design features and capabilities of high performance research reactors currently in use or under construction worldwide. Recent and ongoing design efforts in the US and internationally are briefly summarized, followed by conclusions and recommendations.

  14. USGS Core Research Center (CRC) Collection of Cuttings

    Data.gov (United States)

    U.S. Geological Survey, Department of the Interior — The Core Research Center (CRC) was established in 1974 by the U.S. Geological Survey (USGS) to preserve valuable rock cores for use by scientists and educators from...

  15. Developing Fully Coupled Dynamical Reactor Core Isolation System Models in RELAP-7 for Extended Station Black-Out Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Haihua Zhao; Ling Zou; Hongbin Zhang; David Andrs; Richard Martineau

    2014-04-01

    The reactor core isolation cooling (RCIC) system in a boiling water reactor (BWR) provides makeup water to the reactor vessel for core cooling when the main steam lines are isolated and the normal supply of water to the reactor vessel is lost. It was one of the very few safety systems still available during the Fukushima Daiichi accidents after the tsunamis hit the plants and the system successfully delayed the core meltdown for a few days for unit 2 & 3. Therefore, detailed models for RCIC system components are indispensable to understand extended station black-out accidents (SBO) for BWRs. As part of the effort to develop the new generation reactor system safety analysis code RELAP-7, major components to simulate the RCIC system have been developed. This paper describes the models for those components such as turbine, pump, and wet well. Selected individual component test simulations and a simplified SBO simulation up to but before core damage is presented. The successful implementation of the simplified RCIC and wet well models paves the way to further improve the models for safety analysis by including more detailed physical processes in the near future.

  16. Source term derivation and radiological safety analysis for the TRICO II research reactor in Kinshasa

    Energy Technology Data Exchange (ETDEWEB)

    Muswema, J.L., E-mail: jeremie.muswem@unikin.ac.cd [Faculty of Science, University of Kinshasa, P.O. Box 190, KIN XI (Congo, The Democratic Republic of the); Ekoko, G.B. [Faculty of Science, University of Kinshasa, P.O. Box 190, KIN XI (Congo, The Democratic Republic of the); Lukanda, V.M. [Faculty of Science, University of Kinshasa, P.O. Box 190, KIN XI (Congo, The Democratic Republic of the); Democratic Republic of the Congo' s General Atomic Energy Commission, P.O. Box AE1 (Congo, The Democratic Republic of the); Lobo, J.K.-K. [Faculty of Science, University of Kinshasa, P.O. Box 190, KIN XI (Congo, The Democratic Republic of the); Darko, E.O. [Radiation Protection Institute, Ghana Atomic Energy Commission, P.O. Box LG 80, Legon, Accra (Ghana); Boafo, E.K. [University of Ontario Institute of Technology, 2000 Simcoe St. North, Oshawa, ONL1 H7K4 (Canada)

    2015-01-15

    Highlights: • Atmospheric dispersion modeling for two credible accidents of the TRIGA Mark II research reactor in Kinshasa (TRICO II) was performed. • Radiological safety analysis after the postulated initiating events (PIE) was also carried out. • The Karlsruhe KORIGEN and the HotSpot Health Physics codes were used to achieve the objectives of this study. • All the values of effective dose obtained following the accident scenarios were below the regulatory limits for reactor staff members and the public, respectively. - Abstract: The source term from the 1 MW TRIGA Mark II research reactor core of the Democratic Republic of the Congo was derived in this study. An atmospheric dispersion modeling followed by radiation dose calculation were performed based on two possible postulated accident scenarios. This derivation was made from an inventory of peak radioisotope activities released in the core by using the Karlsruhe version of isotope generation code KORIGEN. The atmospheric dispersion modeling was performed with HotSpot code, and its application yielded to radiation dose profile around the site using meteorological parameters specific to the area under study. The two accident scenarios were picked from possible accident analyses for TRIGA and TRIGA-fueled reactors, involving the case of destruction of the fuel element with highest activity release and a plane crash on the reactor building as the worst case scenario. Deterministic effects of these scenarios are used to update the Safety Analysis Report (SAR) of the reactor, and for its current version, these scenarios are not yet incorporated. Site-specific meteorological conditions were collected from two meteorological stations: one installed within the Atomic Energy Commission and another at the National Meteorological Agency (METTELSAT), which is not far from the site. Results show that in both accident scenarios, radiation doses remain within the limits, far below the recommended maximum effective

  17. Development of Reactor Core for Nuclear Thermal Propulsion%核热推进堆芯方案的发展

    Institute of Scientific and Technical Information of China (English)

    解家春; 赵守智

    2012-01-01

    Nuclear thermal propulsion heats propellant with fission energy. It's specific impulse is double of chemical rockets. It could play an important role in space mission. During the research process about nuclear thermal propulsion in USA and Russia, many reactors were well developed. The details of the reactors core were described, the characteristics of design were indicated, and the trend of development was summarized.%核热推进利用核裂变能加热工质,比冲可达化学火箭的2倍多,在空间活动中有广阔的应用前景.在美国和俄罗斯的研究过程中,对多个核热推进堆芯方案进行了较深入的研究.本工作介绍了这些堆芯方案的情况,详细说明了其设计特点,并总结了堆芯方案的发展趋势.

  18. Fundamental approaches for analysis thermal hydraulic parameter for Puspati Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hashim, Zaredah, E-mail: zaredah@nm.gov.my; Lanyau, Tonny Anak, E-mail: tonny@nm.gov.my; Farid, Mohamad Fairus Abdul; Kassim, Mohammad Suhaimi [Reactor Technology Centre, Technical Support Division, Malaysia Nuclear Agency, Ministry of Science, Technology and Innovation, Bangi, 43000, Kajang, Selangor Darul Ehsan (Malaysia); Azhar, Noraishah Syahirah [Universiti Teknologi Malaysia, 80350, Johor Bahru, Johor Darul Takzim (Malaysia)

    2016-01-22

    The 1-MW PUSPATI Research Reactor (RTP) is the one and only nuclear pool type research reactor developed by General Atomic (GA) in Malaysia. It was installed at Malaysian Nuclear Agency and has reached the first criticality on 8 June 1982. Based on the initial core which comprised of 80 standard TRIGA fuel elements, the very fundamental thermal hydraulic model was investigated during steady state operation using the PARET-code. The main objective of this paper is to determine the variation of temperature profiles and Departure of Nucleate Boiling Ratio (DNBR) of RTP at full power operation. The second objective is to confirm that the values obtained from PARET-code are in agreement with Safety Analysis Report (SAR) for RTP. The code was employed for the hot and average channels in the core in order to calculate of fuel’s center and surface, cladding, coolant temperatures as well as DNBR’s values. In this study, it was found that the results obtained from the PARET-code showed that the thermal hydraulic parameters related to safety for initial core which was cooled by natural convection was in agreement with the designed values and safety limit in SAR.

  19. Major update of Safety Analysis Report for Thai Research Reactor-1/Modification 1

    Energy Technology Data Exchange (ETDEWEB)

    Tippayakul, Chanatip [Thailand Institute of Nuclear Technology, Bangkok (Thailand)

    2013-07-01

    Thai Research Reactor-1/Modification 1 (TRR-1/M1) was converted from a Material Testing Reactor in 1975 and it had been operated by Office of Atom for Peace (OAP) since 1977 until 2007. During the period, Office of Atom for Peace had two duties for the reactor, that is, to operate and to regulate the reactor. However, in 2007, there was governmental office reformation which resulted in the separation of the reactor operating organization from the regulatory body in order to comply with international standard. The new organization is called Thailand Institute of Nuclear Technology (TINT) which has the mission to promote peaceful utilization of nuclear technology while OAP remains essentially the regulatory body. After the separation, a new ministerial regulation was enforced reflecting a new licensing scheme in which TINT has to apply for a license to operate the reactor. The safety analysis report (SAR) shall be submitted as part of the license application. The ministerial regulation stipulates the outlines of the SAR almost equivalent to IAEA standard 35-G1. Comparing to the IAEA 35-G1 standard, there were several incomplete and missing chapters in the original SAR of TRR1/M1. The major update of the SAR was therefore conducted and took approximately one year. The update work included detail safety evaluation of core configuration which used two fuel element types, the classification of systems, structures and components (SSC), the compilation of detail descriptions of all SSCs and the review and evaluation of radiation protection program, emergency plan and emergency procedure. Additionally, the code of conduct and operating limits and conditions were revised and finalized in this work. A lot of new information was added to the SAR as well, for example, the description of commissioning program, information on environmental impact assessment, decommissioning program, quality assurance program and etc. Due to the complexity of this work, extensive knowledge was

  20. Determination of the level of water in the core of reactors PWR using neutron detectors signal ex core; Determinacion del nivel del agua del nucleo de reactores PWR usando la senal de detectores neutronicos excore

    Energy Technology Data Exchange (ETDEWEB)

    Bernal, A.; Abarca, A.; Miro, R.; Verdu, G.

    2014-07-01

    The level of water from the core provides relevant information of the neutronic and thermal hydraulic of the reactor as the power, k EFF and cooling capacity. In fact, this level monitoring can be used for prediction of LOCA and reduction of cooling that can cause damage to the core. There are several teams that measure a variety of parameters of the reactor, as opposed to the level of the water of the core. However, the detectors 'excore' measure fast neutrons which escape from the core and there are studies that demonstrate the existence of a relationship between them and the water level of the kernel due to the water shield. Therefore, a methodology has been developed to determine this relationship, using the Monte Carlo method using the MCNP code and apply variance reduction techniques based on the attached flow that is obtained using the method of discrete ordinates using code TORT. (Author)

  1. A STRONGLY COUPLED REACTOR CORE ISOLATION COOLING SYSTEM MODEL FOR EXTENDED STATION BLACK-OUT ANALYSES

    Energy Technology Data Exchange (ETDEWEB)

    Zhao, Haihua [Idaho National Laboratory; Zhang, Hongbin [Idaho National Laboratory; Zou, Ling [Idaho National Laboratory; Martineau, Richard Charles [Idaho National Laboratory

    2015-03-01

    The reactor core isolation cooling (RCIC) system in a boiling water reactor (BWR) provides makeup cooling water to the reactor pressure vessel (RPV) when the main steam lines are isolated and the normal supply of water to the reactor vessel is lost. The RCIC system operates independently of AC power, service air, or external cooling water systems. The only required external energy source is from the battery to maintain the logic circuits to control the opening and/or closure of valves in the RCIC systems in order to control the RPV water level by shutting down the RCIC pump to avoid overfilling the RPV and flooding the steam line to the RCIC turbine. It is generally considered in almost all the existing station black-out accidents (SBO) analyses that loss of the DC power would result in overfilling the steam line and allowing liquid water to flow into the RCIC turbine, where it is assumed that the turbine would then be disabled. This behavior, however, was not observed in the Fukushima Daiichi accidents, where the Unit 2 RCIC functioned without DC power for nearly three days. Therefore, more detailed mechanistic models for RCIC system components are needed to understand the extended SBO for BWRs. As part of the effort to develop the next generation reactor system safety analysis code RELAP-7, we have developed a strongly coupled RCIC system model, which consists of a turbine model, a pump model, a check valve model, a wet well model, and their coupling models. Unlike the traditional SBO simulations where mass flow rates are typically given in the input file through time dependent functions, the real mass flow rates through the turbine and the pump loops in our model are dynamically calculated according to conservation laws and turbine/pump operation curves. A simplified SBO demonstration RELAP-7 model with this RCIC model has been successfully developed. The demonstration model includes the major components for the primary system of a BWR, as well as the safety

  2. Characterization of a Neutron Beam Following Reconfiguration of the Neutron Radiography Reactor (NRAD Core and Addition of New Fuel Elements

    Directory of Open Access Journals (Sweden)

    Aaron E. Craft

    2016-02-01

    Full Text Available The neutron radiography reactor (NRAD is a 250 kW Mark-II Training, Research, Isotopes, General Atomics (TRIGA reactor at Idaho National Laboratory, Idaho Falls, ID, USA. The East Radiography Station (ERS is one of two neutron beams at the NRAD used for neutron radiography, which sits beneath a large hot cell and is primarily used for neutron radiography of highly radioactive objects. Additional fuel elements were added to the NRAD core in 2013 to increase the excess reactivity of the reactor, and may have changed some characteristics of the neutron beamline. This report discusses characterization of the neutron beamline following the addition of fuel to the NRAD. This work includes determination of the facility category according to the American Society for Testing and Materials (ASTM standards, and also uses an array of gold foils to determine the neutron beam flux and evaluate the neutron beam profile. The NRAD ERS neutron beam is a Category I neutron radiography facility, the highest possible quality level according to the ASTM. Gold foil activation experiments show that the average neutron flux with length-to-diameter ratio (L/D = 125 is 5.96 × 106 n/cm2/s with a 2σ standard error of 2.90 × 105 n/cm2/s. The neutron beam profile can be considered flat for qualitative neutron radiographic evaluation purposes. However, the neutron beam profile should be taken into account for quantitative evaluation.

  3. Reactor-produced radionuclides at the University of Missouri Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Ketring, A.R.; Evans-Blumer, M.S.; Ehrhardt, G.J. [University of Missouri Research Reactor, Colombia (United States). Departments of Radiology, Chemistry and Nuclear Engineering

    1997-10-01

    Nuclear medicine has primarily been a diagnostic science for many years, but today is facing considerable challenges from other modalities in this area. However, these competing techniques (magnetic resonance imaging, ultrasound, and computer-assisted tomography) in general are not therapeutic. Although early nuclear medicine therapy was of limited efficacy, in recent years a revolution in radiotherapy has been developing base don more sophisticated targeting methods, including radioactive intra-arterial microspheres, chemically-guided bone agents, labelled monoclonal antibodies, and isotopically-tagged polypeptide receptor-binding agents. Although primarily used for malignancies, therapeutic nuclear medicine is also applicable to the treatment of rheumatoid arthritis and possibly coronary artery re closure following angioplasty. The isotopes of choice for these applications are reactor-produced beta emitters such as Sm-153, Re-186, Re-188, Ho-166, Lu-177, and Rh-105. Although alpha emitters possess greater cell toxicity due to their high LET, the greater range of beta emitters and the typically inhomogeneous deposition of radiotherapy agents in lesions leads to greater beta `crossfire` and better overall results. The University of Missouri Research Reactor (MURR) has been in the forefront of research into means of preparing, handling and supplying these high-specific-activity isotopes in quantities appropriate not only for research, but also for patient trials in the US and around the world. Researchers at MURR in collaboration with others at the University of Missouri (MU) developed Sm-153 Quadramet{sup TM}, a drug recently approved in the US for palliation of bone tumor pain. In conjunction with researchers at the University of Missouri-Rolla, MURR also developed Y-90 TheraSphere{sup TM}, an agent for the treatment of liver cancer now approved in Canada. Considerable effort has been expended to develop techniques for irradiation, handling, and shipping isotopes

  4. 75 FR 27368 - Aerotest Operations, Inc., Aerotest Radiography and Research Reactor; Notice of Consideration of...

    Science.gov (United States)

    2010-05-14

    ... COMMISSION Aerotest Operations, Inc., Aerotest Radiography and Research Reactor; Notice of Consideration of... INFORMATION CONTACT: Cindy Montgomery, Project Manager, Research and Test Reactors Licensing Branch, Division... Operating License No. R-98 for the Aerotest Radiography and Research Reactor (ARRR), currently held by...

  5. Two Step Procedure Using a 1-D Slab Spectral Geometry in a Pebble Bed Reactor Core Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Hyun Chul; Kim, Kang Seog; Noh, Jae Man; Joo, Hyung Kook [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    2005-07-01

    A strong spectral interaction between the core and the reflector has been one of the main concerns in the analysis of pebble bed reactor cores. To resolve this problem, VSOP adopted iteration between the spectrum calculation in a spectral zone and the global core calculation. In VSOP, the whole problem domain is divided into many spectral zones in which the fine group spectrum is calculated using bucklings for fast groups and albedos for thermal groups from the global core calculation. The resulting spectrum in each spectral zone is used to generate broad group cross sections of the spectral zone for the global core calculation. In this paper, we demonstrate a two step procedure in a pebble bed reactor core analysis. In the first step, we generate equivalent cross sections from a 1-D slab spectral geometry model with the help of the equivalence theory. The equivalent cross sections generated in this way include the effect of the spectral interaction between the core and the reflector. In the second step, we perform a diffusion calculation using the equivalent cross sections generated in the first step. A simple benchmark problem derived from the PMBR-400 Reactor was introduced to verify this approach. We compared the two step solutions with the Monte Carlo (MC) solutions for the problem.

  6. Dosimetry at the Portuguese research reactor using thermoluminescence measurements and Monte Carlo calculations.

    Science.gov (United States)

    Fernandes, A C; Gonçalves, I C; Santos, J; Cardoso, J; Santos, L; Ferro Carvalho, A; Marques, J G; Kling, A; Ramalho, A J G; Osvay, M

    2006-01-01

    This work presents an extensive study on Monte Carlo radiation transport simulation and thermoluminescent (TL) dosimetry for characterising mixed radiation fields (neutrons and photons) occurring in nuclear reactors. The feasibility of these methods is investigated for radiation fields at various locations of the Portuguese Research Reactor (RPI). The performance of the approaches developed in this work is compared with dosimetric techniques already existing at RPI. The Monte Carlo MCNP-4C code was used for a detailed modelling of the reactor core, the fast neutron beam and the thermal column of RPI. Simulations using these models allow to reproduce the energy and spatial distributions of the neutron field very well (agreement better than 80%). In the case of the photon field, the agreement improves with decreasing intensity of the component related to fission and activation products. (7)LiF:Mg,Ti, (7)LiF:Mg,Cu,P and Al(2)O(3):Mg,Y TL detectors (TLDs) with low neutron sensitivity are able to determine photon dose and dose profiles with high spatial resolution. On the other hand, (nat)LiF:Mg,Ti TLDs with increased neutron sensitivity show a remarkable loss of sensitivity and a high supralinearity in high-intensity fields hampering their application at nuclear reactors.

  7. Research and Development Roadmaps for Liquid Metal Cooled Fast Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kim, T. K. [Argonne National Lab. (ANL), Argonne, IL (United States); Grandy, C. [Argonne National Lab. (ANL), Argonne, IL (United States); Natesan, K. [Argonne National Lab. (ANL), Argonne, IL (United States); Sienicki, J. [Argonne National Lab. (ANL), Argonne, IL (United States); Hill, R. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2017-04-20

    The United States Department of Energy (DOE) commissioned the development of technology roadmaps for advanced (non-light water reactor) reactor concepts to help focus research and development funding over the next five years. The roadmaps show the research and development needed to support demonstration of an advanced (non-LWR) concept by the early 2030s, consistent with DOE’s Vision and Strategy for the Development and Deployment of Advanced Reactors. The intent is only to convey the technical steps that would be required to achieve such a goal; the means by which DOE will determine whether to invest in specific tasks will be treated separately. The starting point for the roadmaps is the Technical Readiness Assessment performed as part of an Advanced Test and Demonstration Reactor study released in 2016. The roadmaps were developed based upon a review of technical reports and vendor literature summarizing the technical maturity of each concept and the outstanding research and development needs. Critical path tasks for specific systems were highlighted on the basis of time and resources needed to complete the tasks and the importance of the system to the performance of the reactor concept. The roadmaps are generic, i.e. not specific to a particular vendor’s design but vendor design information may have been used as representative of the concept family. In the event that both near-term and more advanced versions of a concept are being developed, either a single roadmap with multiple branches or separate roadmaps for each version were developed. In each case, roadmaps point to a demonstration reactor (engineering or commercial) and show the activities that must be completed in parallel to support that demonstration in the 2030-2035 window. This report provides the roadmaps for two fast reactor concepts, the Sodium-cooled Fast Reactor (SFR) and the Lead-cooled Fast Reactor (LFR). The SFR technology is mature enough for commercial demonstration by the early 2030s

  8. Preliminary Assessment of the Impact on Reactor Vessel dpa Rates Due to Installation of a Proposed Low Enriched Uranium (LEU) Core in the High Flux Isotope Reactor (HFIR)

    Energy Technology Data Exchange (ETDEWEB)

    Daily, Charles R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-10-01

    An assessment of the impact on the High Flux Isotope Reactor (HFIR) reactor vessel (RV) displacements-per-atom (dpa) rates due to operations with the proposed low enriched uranium (LEU) core described by Ilas and Primm has been performed and is presented herein. The analyses documented herein support the conclusion that conversion of HFIR to low-enriched uranium (LEU) core operations using the LEU core design of Ilas and Primm will have no negative impact on HFIR RV dpa rates. Since its inception, HFIR has been operated with highly enriched uranium (HEU) cores. As part of an effort sponsored by the National Nuclear Security Administration (NNSA), conversion to LEU cores is being considered for future HFIR operations. The HFIR LEU configurations analyzed are consistent with the LEU core models used by Ilas and Primm and the HEU balance-of-plant models used by Risner and Blakeman in the latest analyses performed to support the HFIR materials surveillance program. The Risner and Blakeman analyses, as well as the studies documented herein, are the first to apply the hybrid transport methods available in the Automated Variance reduction Generator (ADVANTG) code to HFIR RV dpa rate calculations. These calculations have been performed on the Oak Ridge National Laboratory (ORNL) Institutional Cluster (OIC) with version 1.60 of the Monte Carlo N-Particle 5 (MCNP5) computer code.

  9. Safety culture and quality management of Kartini research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Syarip [Yogyakarta Nuclear Research Centre, Yogyakarta (Indonesia); Hauptmanns, Ulrich [Department of Plant Design and Safety, Otto-Von-Guericke-University, Magdeburg (Germany)

    1999-10-01

    The evaluation for assessing the safety culture and quality of safety management of Kartini research reactor is presented. The method is based on the concept of management control of safety (audit) as well as by using the developed method i.e. the questionnaires concerning areas of relevance which have to be answered with value statements. There are seven statements or qualifiers in answering the questions. Since such statements are vague, they are represented by fuzzy numbers. The weaknesses can be identified from the different areas contemplated. The evaluation result show that the quality of safety management of Kartini research reactor is globally rated as 'Average'. The operator behavior in the implementation of 'safety culture' concept is found as a weakness, therefore this area should be improved. (author)

  10. IAEA designated international centre based on research reactors (ICERR)

    Energy Technology Data Exchange (ETDEWEB)

    Di Tigliole, Andrea Borio; Bradley, Edward; Khoroshev, Mikhail; Marshall, Frances; Morris, Charles; Tozser, Sandor [International Atomic Energy Agency, Vienna (Austria). Dept. of Nuclear Energy

    2016-04-15

    International activities in the back end of the research reactor (RR) fuel cycle have so far been dominated by the programmes of acceptance of highly-enriched uranium (HEU) spent nuclear fuel (SNF) by the country where it was originally enriched. These programmes will soon have achieved their goals. However, the needs of the nuclear community dictate that the majority of the research reactors continues to operate using low enriched uranium (LEU) fuel in order to meet the varied mission objectives. As a result, inventories of LEU SNF will continue to be created and the back end solution of RR SNF remains a critical issue. In view of this fact, the IAEA drew up a report presenting available reprocessing and recycling services for RR SNF.

  11. Relevant thermal-hydraulic aspects in the design of the RRR (Replacement Research Reactor)

    Energy Technology Data Exchange (ETDEWEB)

    Doval, Alicia S.; Mazufri, Claudio M. [INVAP SE, Bariloche (Argentina)

    2002-07-01

    A description of the main thermal-hydraulic features and challenges of the Replacement Research Reactor, for the Australian Nuclear Science and Technology Organization (ANSTO), is presented. Different hydraulic and thermal-hydraulic aspects are considered, core cooling during full power operation and the way it affects the design, design criteria, engineered safety features and computational tools, amongst others. A special section is devoted to the thermal-hydraulic aspects inside the reflector tank, as well as the cooling of irradiation facilities, particularly, the Molybdenum production facility. (author)

  12. Advanced Test Reactor Core Modeling Update Project Annual Report for Fiscal Year 2011

    Energy Technology Data Exchange (ETDEWEB)

    David W. Nigg; Devin A. Steuhm

    2011-09-01

    Legacy computational reactor physics software tools and protocols currently used for support of Advanced Test Reactor (ATR) core fuel management and safety assurance and, to some extent, experiment management are obsolete, inconsistent with the state of modern nuclear engineering practice, and are becoming increasingly difficult to properly verify and validate (V&V). Furthermore, the legacy staff knowledge required for application of these tools and protocols from the 1960s and 1970s is rapidly being lost due to staff turnover and retirements. In 2009 the Idaho National Laboratory (INL) initiated a focused effort to address this situation through the introduction of modern high-fidelity computational software and protocols, with appropriate V&V, within the next 3-4 years via the ATR Core Modeling and Simulation and V&V Update (or 'Core Modeling Update') Project. This aggressive computational and experimental campaign will have a broad strategic impact on the operation of the ATR, both in terms of improved computational efficiency and accuracy for support of ongoing DOE programs as well as in terms of national and international recognition of the ATR National Scientific User Facility (NSUF). The ATR Core Modeling Update Project, targeted for full implementation in phase with the anticipated ATR Core Internals Changeout (CIC) in the 2014 time frame, began during the last quarter of Fiscal Year 2009, and has just completed its first full year. Key accomplishments so far have encompassed both computational as well as experimental work. A new suite of stochastic and deterministic transport theory based reactor physics codes and their supporting nuclear data libraries (SCALE, KENO-6, HELIOS, NEWT, and ATTILA) have been installed at the INL under various permanent sitewide license agreements and corresponding baseline models of the ATR and ATRC are now operational, demonstrating the basic feasibility of these code packages for their intended purpose. Furthermore

  13. Advanced Test Reactor Core Modeling Update Project Annual Report for Fiscal Year 2012

    Energy Technology Data Exchange (ETDEWEB)

    David W. Nigg, Principal Investigator; Kevin A. Steuhm, Project Manager

    2012-09-01

    Legacy computational reactor physics software tools and protocols currently used for support of Advanced Test Reactor (ATR) core fuel management and safety assurance, and to some extent, experiment management, are inconsistent with the state of modern nuclear engineering practice, and are difficult, if not impossible, to properly verify and validate (V&V) according to modern standards. Furthermore, the legacy staff knowledge required for application of these tools and protocols from the 1960s and 1970s is rapidly being lost due to staff turnover and retirements. In late 2009, the Idaho National Laboratory (INL) initiated a focused effort, the ATR Core Modeling Update Project, to address this situation through the introduction of modern high-fidelity computational software and protocols. This aggressive computational and experimental campaign will have a broad strategic impact on the operation of the ATR, both in terms of improved computational efficiency and accuracy for support of ongoing DOE programs as well as in terms of national and international recognition of the ATR National Scientific User Facility (NSUF). The ATR Core Modeling Update Project, targeted for full implementation in phase with the next anticipated ATR Core Internals Changeout (CIC) in the 2014-2015 time frame, began during the last quarter of Fiscal Year 2009, and has just completed its third full year. Key accomplishments so far have encompassed both computational as well as experimental work. A new suite of stochastic and deterministic transport theory based reactor physics codes and their supporting nuclear data libraries (HELIOS, KENO6/SCALE, NEWT/SCALE, ATTILA, and an extended implementation of MCNP5) has been installed at the INL under various licensing arrangements. Corresponding models of the ATR and ATRC are now operational with all five codes, demonstrating the basic feasibility of the new code packages for their intended purpose. Of particular importance, a set of as-run core

  14. Status report on the Small Secure Transportable Autonomous Reactor (SSTAR) /Lead-cooled Fast Reactor (LFR) and supporting research and development.

    Energy Technology Data Exchange (ETDEWEB)

    Sienicki, J. J.; Moisseytsev, A.; Yang, W. S.; Wade, D. C.; Nikiforova, A.; Hanania, P.; Ryu, H. J.; Kulesza, K. P.; Kim, S. J.; Halsey, W. G.; Smith, C. F.; Brown, N. W.; Greenspan, E.; de Caro, M.; Li, N.; Hosemann, P.; Zhang, J.; Yu, H.; Nuclear Engineering Division; LLNL; LANL; Massachusetts Inst. of Tech.; Ecole des Mines de Paris; Oregon State Univ.; Univ.of California at Berkley

    2008-06-23

    This report provides an update on development of a pre-conceptual design for the Small Secure Transportable Autonomous Reactor (SSTAR) Lead-Cooled Fast Reactor (LFR) plant concept and supporting research and development activities. SSTAR is a small, 20 MWe (45 MWt), natural circulation, fast reactor plant for international deployment concept incorporating proliferation resistance for deployment in non-fuel cycle states and developing nations, fissile self-sufficiency for efficient utilization of uranium resources, autonomous load following making it suitable for small or immature grid applications, and a high degree of passive safety further supporting deployment in developing nations. In FY 2006, improvements have been made at ANL to the pre-conceptual design of both the reactor system and the energy converter which incorporates a supercritical carbon dioxide Brayton cycle providing higher plant efficiency (44 %) and improved economic competitiveness. The supercritical CO2 Brayton cycle technology is also applicable to Sodium-Cooled Fast Reactors providing the same benefits. One key accomplishment has been the development of a control strategy for automatic control of the supercritical CO2 Brayton cycle in principle enabling autonomous load following over the full power range between nominal and essentially zero power. Under autonomous load following operation, the reactor core power adjusts itself to equal the heat removal from the reactor system to the power converter through the large reactivity feedback of the fast spectrum core without the need for motion of control rods, while the automatic control of the power converter matches the heat removal from the reactor to the grid load. The report includes early calculations for an international benchmarking problem for a LBE-cooled, nitride-fueled fast reactor core organized by the IAEA as part of a Coordinated Research Project on Small Reactors without Onsite Refueling; the calculations use the same neutronics

  15. Synthesis of Ni-SiO2/silicalite-1 core-shell micromembrane reactors and their reaction/diffusion performance

    KAUST Repository

    Khan, Easir A.

    2010-12-15

    Core-shell micromembrane reactors are a novel class of materials where a catalyst and a shape-selective membrane are synergistically housed in a single particle. In this work, we report the synthesis of micrometer -sized core-shell particles containing a catalyst core and a thin permselective zeolite shell and their application as a micromembrane reactor for the selective hydrogenation of the 1-hexene and 3,3-dimethyl-1-butene isomers. The bare catalyst, which is made from porous silica loaded with catalytically active nickel, showed no reactant selectivity between hexene isomers, but the core-shell particles showed high selectivities up to 300 for a 1-hexene conversion of 90%. © 2010 American Chemical Society.

  16. Mixed first- and second-order transport method using domain decomposition techniques for reactor core calculations

    Energy Technology Data Exchange (ETDEWEB)

    Girardi, E.; Ruggieri, J.M. [CEA Cadarache, CEA/DEN/CAD/DER/SPRC/LEPH, 13 - Saint-Paul Lez Durance (France)

    2003-07-01

    The aim of this paper is to present the last developments made on a domain decomposition method applied to reactor core calculations. In this method, two kind of balance equation with two different numerical methods dealing with two different unknowns are coupled. In the first part the two balance transport equations (first order and second order one) are presented with the corresponding following numerical methods: Variational Nodal Method and Discrete Ordinate Nodal Method. In the second part, the Multi-Method/Multi-Domain algorithm is introduced by applying the Schwarz domain decomposition to the multigroup eigenvalue problem of the transport equation. The resulting algorithm is then provided. The projection operators used to coupled the two methods are detailed in the last part of the paper. Finally some preliminary numerical applications on benchmarks are given showing encouraging results. (authors)

  17. Gas Core Reactor Numerical Simulation Using a Coupled MHD-MCNP Model

    Science.gov (United States)

    Kazeminezhad, F.; Anghaie, S.

    2008-01-01

    Analysis is provided in this report of using two head-on magnetohydrodynamic (MHD) shocks to achieve supercritical nuclear fission in an axially elongated cylinder filled with UF4 gas as an energy source for deep space missions. The motivation for each aspect of the design is explained and supported by theory and numerical simulations. A subsequent report will provide detail on relevant experimental work to validate the concept. Here the focus is on the theory of and simulations for the proposed gas core reactor conceptual design from the onset of shock generations to the supercritical state achieved when the shocks collide. The MHD model is coupled to a standard nuclear code (MCNP) to observe the neutron flux and fission power attributed to the supercritical state brought about by the shock collisions. Throughout the modeling, realistic parameters are used for the initial ambient gaseous state and currents to ensure a resulting supercritical state upon shock collisions.

  18. Simulation of In-Core Dose Rates for an Offline CANDU Reactor

    Science.gov (United States)

    Gilbert, Jordan

    This thesis describes the development of a Monte Carlo simulation to predict the dose rates that will be encountered by a novel robotic inspection system for the pressure tubes of an offline CANDU reactor. Simulations were performed using the Monte Carlo N-Particle (MCNP) radiation transport code, version 6.1. The radiation fields within the reactor, even when shut down, are very high, and can cause significant damage to certain structural components and the electronics of the inspection system. Given that the robotic system will rely heavily on electronics, it is important to know the dose rates that will be encountered, in order to estimate the component lifetimes. The MCNP simulation was developed and benchmarked against information obtained from Ontario Power Generation and the Canadian Nuclear Laboratories. The benchmarking showed a good match between the simulated values and the expected values. This simulation, coupled with the accompanying user interface, represent a tool in dose field prediction that is currently unavailable. Predicted dose rates for a postulated inspection at 7 days after shutdown, with 2:5 cm of tungsten shielding around the key components, would survive for approximately 7 hours in core. This is anticipated to be enough time to perform an inspection and shows that the use of this tool can aid in designing the new inspection system.

  19. Evaluation of radcal gamma thermometers for in-core monitoring of Savannah River Site production reactors

    Energy Technology Data Exchange (ETDEWEB)

    McCulloch, R.W.; Crowley, J.L. [DELTA M Corp., Oak Ridge, TN (United States); Croft, W.D. [Westinghouse Savannah River Co., Aiken, SC (United States)

    1991-12-31

    The Savannah River Site (SRS) recently obtained a quantity of Radcal Gamma Thermometer Assemblies (RGTAs) for in-core monitoring of local power in their production reactors. The RGTAs, manufactured by DELTA M Corporation in Oak Ridge, Tennessee, contained seven Self Calibrating Gamma Thermometer (SCGT) sensors within a 7.26 mm diameter, 3.06 m length with a total length of 5.6 m. All RGTAs contained an isolated segmented heater cable for in-situ calibration. Each SCGT sensor was subjected to a 40 point calibration at discrete power levels from 0.5 to 6 watts per gram (w/g) under both joule and cable power. Calibration equations were developed from this to predict reactor power at each sensor. Additionally three units were calibrated at combined joule and cable heating conditions from 0.5 to 2.5 w/g cable and 0.5 to 6 w/g joule. A statistical analysis of all data was used to derive prediction equations that enable SRS engineers to precisely track any changes in sensor calibration throughout the lifetime of the instruments. This paper presents the detailed configuration of the 36 units manufactured for SRS, reviews the calibration results, and discusses the utility and accuracy of the statistically derived prediction equations for in-situ calibration.

  20. Evaluation of radcal gamma thermometers for in-core monitoring of Savannah River Site production reactors

    Energy Technology Data Exchange (ETDEWEB)

    McCulloch, R.W.; Crowley, J.L. (DELTA M Corp., Oak Ridge, TN (United States)); Croft, W.D. (Westinghouse Savannah River Co., Aiken, SC (United States))

    1991-01-01

    The Savannah River Site (SRS) recently obtained a quantity of Radcal Gamma Thermometer Assemblies (RGTAs) for in-core monitoring of local power in their production reactors. The RGTAs, manufactured by DELTA M Corporation in Oak Ridge, Tennessee, contained seven Self Calibrating Gamma Thermometer (SCGT) sensors within a 7.26 mm diameter, 3.06 m length with a total length of 5.6 m. All RGTAs contained an isolated segmented heater cable for in-situ calibration. Each SCGT sensor was subjected to a 40 point calibration at discrete power levels from 0.5 to 6 watts per gram (w/g) under both joule and cable power. Calibration equations were developed from this to predict reactor power at each sensor. Additionally three units were calibrated at combined joule and cable heating conditions from 0.5 to 2.5 w/g cable and 0.5 to 6 w/g joule. A statistical analysis of all data was used to derive prediction equations that enable SRS engineers to precisely track any changes in sensor calibration throughout the lifetime of the instruments. This paper presents the detailed configuration of the 36 units manufactured for SRS, reviews the calibration results, and discusses the utility and accuracy of the statistically derived prediction equations for in-situ calibration.

  1. STEADY STATE MODELING OF THE MINIMUM CRITICAL CORE OF THE TRANSIENT REACTOR TEST FACILITY

    Energy Technology Data Exchange (ETDEWEB)

    Anthony L. Alberti; Todd S. Palmer; Javier Ortensi; Mark D. DeHart

    2016-05-01

    With the advent of next generation reactor systems and new fuel designs, the U.S. Department of Energy (DOE) has identified the need for the resumption of transient testing of nuclear fuels. The DOE has decided that the Transient Reactor Test Facility (TREAT) at Idaho National Laboratory (INL) is best suited for future testing. TREAT is a thermal neutron spectrum, air-cooled, nuclear test facility that is designed to test nuclear fuels in transient scenarios. These specific scenarios range from simple temperature transients to full fuel melt accidents. DOE has expressed a desire to develop a simulation capability that will accurately model the experiments before they are irradiated at the facility. It is the aim for this capability to have an emphasis on effective and safe operation while minimizing experimental time and cost. The multi physics platform MOOSE has been selected as the framework for this project. The goals for this work are to identify the fundamental neutronics properties of TREAT and to develop an accurate steady state model for future multiphysics transient simulations. In order to minimize computational cost, the effect of spatial homogenization and angular discretization are investigated. It was found that significant anisotropy is present in TREAT assemblies and to capture this effect, explicit modeling of cooling channels and inter-element gaps is necessary. For this modeling scheme, single element calculations at 293 K gave power distributions with a root mean square difference of 0.076% from those of reference SERPENT calculations. The minimum critical core configuration with identical gap and channel treatment at 293 K resulted in a root mean square, total core, radial power distribution 2.423% different than those of reference SERPENT solutions.

  2. Simulating High Flux Isotope Reactor Core Thermal-Hydraulics via Interdimensional Model Coupling

    Energy Technology Data Exchange (ETDEWEB)

    Travis, Adam R [ORNL

    2014-05-01

    A coupled interdimensional model is presented for the simulation of the thermal-hydraulic characteristics of the High Flux Isotope Reactor core at Oak Ridge National Laboratory. The model consists of two domains a solid involute fuel plate and the surrounding liquid coolant channel. The fuel plate is modeled explicitly in three-dimensions. The coolant channel is approximated as a twodimensional slice oriented perpendicular to the fuel plate s surface. The two dimensionally-inconsistent domains are linked to one another via interdimensional model coupling mechanisms. The coupled model is presented as a simplified alternative to a fully explicit, fully three-dimensional model. Involute geometries were constructed in SolidWorks. Derivations of the involute construction equations are presented. Geometries were then imported into COMSOL Multiphysics for simulation and modeling. Both models are described in detail so as to highlight their respective attributes in the 3D model, the pursuit of an accurate, reliable, and complete solution; in the coupled model, the intent to simplify the modeling domain as much as possible without affecting significant alterations to the solution. The coupled model was created with the goal of permitting larger portions of the reactor core to be modeled at once without a significant sacrifice to solution integrity. As such, particular care is given to validating incorporated model simplifications. To the greatest extent possible, the decrease in solution time as well as computational cost are quantified versus the effects such gains have on the solution quality. A variant of the coupled model which sufficiently balances these three solution characteristics is presented alongside the more comprehensive 3D model for comparison and validation.

  3. Preparation Before Signature of Upgrade of Algeria Heavy Water Research Reactor Contract

    Institute of Scientific and Technical Information of China (English)

    LI; Song; ZAN; Huai-qi; XU; Qi-guo; JIA; Yu-wen

    2012-01-01

    <正>Algeria heavy water research reactor (Birine) is a multiple-purpose research reactor, which was constructed with the help of China more than 20 years ago. By request of Algeria, China will upgrade the research reactor; so as to improve the status of current reactor such as equipment ageing, shortage of spare parts, several systems do not meet requirements of current standards and criteria etc.

  4. Computational simulation of fuel burnup estimation for research reactors plate type

    Energy Technology Data Exchange (ETDEWEB)

    Santos, Nadia Rodrigues dos, E-mail: nadiasam@gmail.com [Instituto Federal de Educacao, Ciencia e Tecnologia do Rio de Janeiro (IFRJ), Paracambi, RJ (Brazil); Lima, Zelmo Rodrigues de; Moreira, Maria de Lourdes, E-mail: zrlima@ien.gov.br, E-mail: malu@ien.gov.br [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

    2015-07-01

    The aim of this study is to estimate the spatial fuel burnup, through computational simulation, in two research reactors plate type, loaded with dispersion fuel: the benchmark Material Test Research - International Atomic Energy Agency (MTR-IAEA) and a typical multipurpose reactor (MR). The first composed of plates with uranium oxide dispersed in aluminum (UAlx-Al) and a second composed with uranium silicide (U{sub 3}Si{sub 2}) dispersed in aluminum. To develop this work we used the deterministic code, WIMSD-5B, which performs the cell calculation solving the neutron transport equation, and the DF3DQ code, written in FORTRAN, which solves the three-dimensional neutron diffusion equation using the finite difference method. The methodology used was adequate to estimate the spatial fuel burnup , as the results was in accordance with chosen benchmark, given satisfactorily to the proposal presented in this work, even showing the possibility to be applied to other research reactors. For future work are suggested simulations with other WIMS libraries, other settings core and fuel types. Comparisons the WIMSD-5B results with programs often employed in fuel burnup calculations and also others commercial programs, are suggested too. Another proposal is to estimate the fuel burnup, taking into account the thermohydraulics parameters and the Xenon production. (author)

  5. Future Directions for Research on Core Competencies

    Science.gov (United States)

    Bradshaw, Catherine P.; Guerra, Nancy G.

    2008-01-01

    This concluding commentary highlights common themes that emerged across the chapters in this volume. We identify strengths and limitations of the core competencies framework and discuss the importance of context, culture, and development for understanding the role of the core competencies in preventing risk behavior in adolescence. We also outline…

  6. Future Directions for Research on Core Competencies

    Science.gov (United States)

    Bradshaw, Catherine P.; Guerra, Nancy G.

    2008-01-01

    This concluding commentary highlights common themes that emerged across the chapters in this volume. We identify strengths and limitations of the core competencies framework and discuss the importance of context, culture, and development for understanding the role of the core competencies in preventing risk behavior in adolescence. We also outline…

  7. Characterization of radioactive aerosols in Tehran research reactor containment

    Directory of Open Access Journals (Sweden)

    Moradi Gholamreza

    2015-01-01

    Full Text Available The objectives of this research were to determine the levels of radioactivity in the Tehran research reactor containment and to investigate the mass-size distribution, composition, and concentration of radionuclides during operation of the reactor. A cascade impactor sampler was used to determine the size-activity distributions of radioactive aerosols in each of the sampling stations. Levels of a and b activities were determined based on a counting method using a liquid scintillation counter and smear tests. The total average mass fractions of fine particles (particle diameter dp < 1 mm in all of the sampling stations were approximately 26.75 %, with the mean and standard deviation of 52.15 ± 19.75 mg/m3. The total average mass fractions of coarse particles were approximately 73.2%, with the mean and standard deviation of 71.34 ± 24.57 mg/m3. In addition to natural radionuclides, artificial radionuclides, such as 24Na, 91Sr, 131I, 133I, 103Ru, 82Br, and 140La, may be released into the reactor containment structure. Maximum activity was associated with accumulation-mode particles with diameters less than 400 nm. The results obtained from liquid scintillation counting suggested that the mean specific activity of alpha particles in fine and coarse-modes were 89.7 % and 10.26 %, respectively. The mean specific activity of beta particles in fine and coarse-modes were 81.15 % and 18.51 %, respectively. A large fraction of the radionuclides' mass concentration in the Tehran research reactor containment was associated with coarse-mode particles, in addition, a large fraction of the activity in the aerosol particles was associated with accumulation-mode particles.

  8. Nuclear reactor physics

    CERN Document Server

    Stacey, Weston M

    2010-01-01

    Nuclear reactor physics is the core discipline of nuclear engineering. Nuclear reactors now account for a significant portion of the electrical power generated worldwide, and new power reactors with improved fuel cycles are being developed. At the same time, the past few decades have seen an ever-increasing number of industrial, medical, military, and research applications for nuclear reactors. The second edition of this successful comprehensive textbook and reference on basic and advanced nuclear reactor physics has been completely updated, revised and enlarged to include the latest developme

  9. The power distribution and neutron fluence measurements and calculations in the VVER-1000 Mock-Up on the LR-0 research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Kostal, M.; Juricek, V.; Rypar, V.; Svadlenkova, M. [Research Center Rez Ltd., 250 68 Husinec-Rez 130 (Czech Republic); Cvachovec, F. [Univ. of Defence, Kounicova 65, 662 10 Brno (Czech Republic)

    2011-07-01

    The power density distribution in a reactor has significant influence on core structures and pressure vessel mechanical resistance, as well as on the physical characteristics of nuclear fuel. This quantity also has an effect on the leakage neutron and photon field. This issue has become of increasing importance, as it touches on actual questions of the VVER nuclear power plant life time extension. This paper shows the comparison of calculated and experimentally determined pin by pin power distributions. The calculation has been performed with deterministic and Monte Carlo approaches. This quantity is accompanied by the neutron and photon flux density calculation and measurements at different points of the light water zero-power (LR-0) research reactor mock-up core, reactor built-in component (core barrel), and reactor pressure vessel and model. The effect of the different data libraries used for calculation is discussed. (authors)

  10. Natural Nuclear Reactor Oklo and Variation of Fundamental Constants Part 1: Computation of Neutronic of Fresh Core

    CERN Document Server

    Petrov, Yu V; Onegin, M S; Petrov, V Yu; Sakhnovskii, E G; Petrov, Yu.V.

    2006-01-01

    Using a modern methods of reactor physics we have performed the full-scale calculations of the natural reactor Oklo. For reliability we have used the recent version of two Monte Carlo codes: the Russian code MCU REA and world wide known code MCNP (USA). Both codes produce close results. We constructed computer model of zone RZ2 of reactor Oklo which takes into account all details of design and composition. The calculations were performed for the three fresh cores with different uranium contents. Multiplication factors, reactivities and neutron fluxes were calculated. We estimated also the temperature and void effects for the fresh core. As would be expected, we have found for the fresh core a great difference between reactor spectra and Maxwell's one, which was used before for averaging cross sections in the Oklo reactor. The averaged cross section of Sm and its dependence on the shift of resonance position (due to variation of fundamental constants) are significantly different from previous results. Contrary...

  11. Present status and future perspectives of research and test reactor in Japan

    Energy Technology Data Exchange (ETDEWEB)

    Kaneko, Yoshihiko [Atomic Energy Research Laboratory, Musashi Institute of Technology, Kawasaki, Kanagawa (Japan); Kaieda, Keisuke [Department of Research Reactor, Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan)

    2000-10-01

    Since 1957, Japan Atomic Energy Research Institute (JAERI) has constructed several research and test reactors to fulfill a major role in the study of nuclear energy and fundamental research. At present four reactors, the Japan Research Reactor No. 3 and No. 4 (JRR-3M and JRR-4 respectively), the Japan Materials Testing Reactor (JMTR) and the Nuclear Safety Research Reactor (NSRR) are in operation, and a new High Temperature Engineering Test Reactor (HTTR) has recently reached first criticality and now in the power up test. In 1966, the Kyoto University built the Kyoto University Reactor (KUR) and started its operation for joint use program of the Japanese universities. This paper introduces these reactors and describes their present operational status and also efforts for aging management. The recent tendency of utilization and future perspectives is also reported. (author)

  12. Constitutive modeling and finite element procedure development for stress analysis of prismatic high temperature gas cooled reactor graphite core components

    Energy Technology Data Exchange (ETDEWEB)

    Mohanty, Subhasish, E-mail: smohanty@anl.gov [Argonne National Laboratory, South Cass Avenue, Argonne, IL 60439 (United States); Majumdar, Saurindranath [Argonne National Laboratory, South Cass Avenue, Argonne, IL 60439 (United States); Srinivasan, Makuteswara [U.S. Nuclear Regulatory Commission, Washington, DC 20555 (United States)

    2013-07-15

    Highlights: • Finite element procedure developed for stress analysis of HTGR graphite component. • Realistic fluence profile and reflector brick shape considered for the simulation. • Also realistic H-451 grade material properties considered for simulation. • Typical outer reflector of a GT-MHR type reactor considered for numerical study. • Based on the simulation results replacement of graphite bricks can be scheduled. -- Abstract: High temperature gas cooled reactors, such as prismatic and pebble bed reactors, are increasingly becoming popular because of their inherent safety, high temperature process heat output, and high efficiency in nuclear power generation. In prismatic reactors, hexagonal graphite bricks are used as reflectors and fuel bricks. In the reactor environment, graphite bricks experience high temperature and neutron dose. This leads to dimensional changes (swelling and or shrinkage) of these bricks. Irradiation dimensional changes may affect the structural integrity of the individual bricks as well as of the overall core. The present paper presents a generic procedure for stress analysis of prismatic core graphite components using graphite reflector as an example. The procedure is demonstrated through commercially available ABAQUS finite element software using the option of user material subroutine (UMAT). This paper considers General Atomics Gas Turbine-Modular Helium Reactor (GT-MHR) as a bench mark design to perform the time integrated stress analysis of a typical reflector brick considering realistic geometry, flux distribution and realistic irradiation material properties of transversely isotropic H-451 grade graphite.

  13. Experimental and analytical investigations of primary coolant pump coastdown phenomena for the Jordan Research and Training Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Alatrash, Yazan [Advanced Nuclear Engineering System Department, Korea University of Science and Technology (UST), 217 Gajeong-ro Yuseong-gu, Daejeon 305-350 (Korea, Republic of); Kang, Han-ok; Yoon, Hyun-gi; Seo, Kyoungwoo; Chi, Dae-Young [Korea Atomic Energy Institute (KAERI), 989-111 Daeduk-daero, Yuseong-gu, Daejeon 305-353 (Korea, Republic of); Yoon, Juhyeon, E-mail: yoonj@kaeri.re.kr [Korea Atomic Energy Institute (KAERI), 989-111 Daeduk-daero, Yuseong-gu, Daejeon 305-353 (Korea, Republic of); Korea University of Science and Technology (UST), Daejeon (Korea, Republic of)

    2015-05-15

    Highlights: • Core flow coastdown phenomena of a research reactor are investigated experimentally. • The experimental dataset is well predicted by a simulation software package, MMS. • The validity and consistency of the experimental dataset are confirmed. • The designed coastdown half time is confirmed to be well above the design requirement. - Abstract: Many low-power research reactors including the Jordan Research and Training Reactor (JRTR) are designed to have a downward core flow during a normal operation mode for many convenient operating features. This design feature requires maintaining the downward core flow for a short period of time right after a loss of off-site power (LOOP) accident to guarantee nuclear fuel integrity. In the JRTR, a big flywheel is installed on a primary cooling system (PCS) pump shaft to passively provide the inertial downward core flow at an initial stage of the LOOP accident. The inertial pumping capability during the coastdown period is experimentally investigated to confirm whether the coastdown half time requirement given by safety analyses is being satisfied. The validity and consistency of the experimental dataset are evaluated using a simulation software package, modular modeling system (MMS).