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Sample records for beijing miniature neutron source reactor

  1. Miniature neutron source reactor burnup calculations using IRBURN code system

    International Nuclear Information System (INIS)

    Highlights: ► Fuel consumption of Iranian MNSR during 15 years of operation has been investigated. ► Calculations have been performed by the IRBURN code. Precision and accuracy of the implemented model has been validated. ► Our study shows the consumption rate of MNSR is about 1%. - Abstract: Fuel consumption of Iranian miniature neutron source reactor (MNSR) during 15 years of operation has been investigated. Reactor core neutronic parameters such as flux and power distributions, control rod worth and effective multiplication factor at BOL and after 15 years of irradiation has been calculated. The Monte Carlo-based depletion code system IRBURN has been used for studying the reactor core neutronic parameters as well as the isotopic inventory of the fuel during burnup. The precision and accuracy of the implemented model has been verified via validation the results for neutronic parameters in the MNSR final safety analysis report. The results show that keff decreases from 1.0034 to 0.9897 and the total U-235 consumption in the core is about 13.669 g after 15 years of operational time. Finally, our studying shows the consumption rate of MNSR is about 1%.

  2. Determination of neutron generation time in miniature neutron source reactor by measurement of neutronics transfer function

    Energy Technology Data Exchange (ETDEWEB)

    Hainoun, A.; Khamis, I. [Atomic Energy Commission, Damascus (Syria). Dept. of Physics

    2000-02-01

    The prompt neutron generation time {lambda} and the total effective fraction of delayed neutrons (including the effect of photoneutrons) {beta} have been experimentally determined for the miniature neutron source reactor (MNSR) of Syria. The neutron generation time was found by taking measurements of the reactor open-loop transfer function using newly devised reactivity-step-ejection method by the reactor pneumatic rabbit system. Small reactivity perturbations i.e. step changes of reactivity starting from steady state, were introduced into the reactor during operation at low power level i.e. zero-power. Relative neutron flux and reactivity versus time were obtained. Using transfer function analysis as well as least square fitting techniques and measuring the delayed neutrons fraction, the neutron generation time was determined to be 74.6{+-}1.57 {mu}s. Using the prompt jump approximation of neutron flux, the total effective fraction of delayed neutrons was measured and found to be 0.00783{+-}0.00017. Measured values of {lambda} and {beta} were found to be very consistent with calculated ones reported in the safety analysis report. (orig.)

  3. Determination of neutron generation time in miniature neutron source reactor by measurement of neutronics transfer function

    International Nuclear Information System (INIS)

    The prompt neutron generation time Λ and the total effective fraction of delayed neutrons (including the effect of photoneutrons) β have been experimentally determined for the miniature neutron source reactor (MNSR) of Syria. The neutron generation time was found by taking measurements of the reactor open-loop transfer function using newly devised reactivity-step-ejection method by the reactor pneumatic rabbit system. Small reactivity perturbations i.e. step changes of reactivity starting from steady state, were introduced into the reactor during operation at low power level i.e. zero-power. Relative neutron flux and reactivity versus time were obtained. Using transfer function analysis as well as least square fitting techniques and measuring the delayed neutrons fraction, the neutron generation time was determined to be 74.6±1.57 μs. Using the prompt jump approximation of neutron flux, the total effective fraction of delayed neutrons was measured and found to be 0.00783±0.00017. Measured values of Λ and β were found to be very consistent with calculated ones reported in the safety analysis report. (orig.)

  4. Neutron energy spectrum adjustment using deposited metal films on Teflon in the miniature neutron source reactor.

    Science.gov (United States)

    Nassan, L; Abdallah, B; Omar, H; Sarheel, A; Alsomel, N; Ghazi, N

    2016-01-01

    The focus of this article was on the experimental estimation of the neutron energy spectrum in the inner irradiation site of the miniature neutron source reactor (MNSR), using, for the first time, a selected set of deposited metal films on Teflon (DMFTs) neutron detectors. Gold, copper, zinc, titanium, aluminum, nickel, silver, and chromium were selected because of the dependence of their neutron cross-sections on neutron energy. Emphasis was placed on the usability of this new type of neutron detectors in the total neutron energy spectrum adjustment. The measured saturation activities per target nucleus values of the DMFTs, and the calculated neutron spectrum in the inner irradiation site using the MCNP-4C code were used as an input for the STAY'SL computer code during the adjustment procedure. The agreement between the numerically calculated and experimentally adjusted spectra results was discussed. PMID:26562448

  5. Evaluation of the photo-neutron source and delayed neutrons in the Syrian miniature neutron source reactor

    International Nuclear Information System (INIS)

    A mathematical model has been developed to simulate the dynamic behavior of the Syrian Miniature Neutron Source Reactor (MNSR). The model is used to assess and evaluate the core average temperature as a function of the overall reactivity load in the core on one hand. On the other hand, the model is utilized to evaluate dynamically the photo and delayed neutron effects in MNSR. The model considers relevant physical phenomena that govern the core such as reactor kinetics, reactivity feedbacks due to coolant temperature and xenon, and thermalhydraulics. Natural convection and point kinetics including the prompt jump and complete mixing approximations were employed. Peak power, reactivity core load, core outlet temperature, and other variables are predicted during self-limiting power excursions. Direct photo-neutron sources strength was dynamically evaluated for the MNSR in subcritical condition. Two different static methods were applied for comparison. In addition, measurement of the photo-neutron source was made using neutron flux monitors and neutron activation analysis technique. Results for both methods were in good agreement. Dynamics effect of the photo neutron source on reactor response to reactivity insertions was demonstrated. Photo-neutron source existence due to beryllium reflector was realized. Compared to related references, close results have been obtained. Core average temperature was studied as a function of reactivity during reactor operation and transients. An overall rough estimate of core average temperature as a function of reactivity load is presented; hence, a procedure to measure such temperature is suggested. (author)

  6. Neutron activation analysis of essential elements in Multani mitti clay using miniature neutron source reactor

    International Nuclear Information System (INIS)

    Multani mitti clay was studied for 19 essential and other elements. Four different radio-assay schemes were adopted for instrumental neutron activation analysis (INAA) using miniature neutron source reactor. The estimated weekly intakes of Cr and Fe are high for men, women, pregnant and lactating women and children while intake of Co is higher in adult categories and Mn by pregnant women. Comparison of MM clay with other type of clays shows that it is a good source of essential elements. - Highlights: ► Multani mitti clay has been studied for 19 essential elements for human adequacy and safety using INAA and AAS. ► Weekly intakes for different consumer categories have been calculated and compared with DRIs. ► Comparison of MM with other type of clays depict that MM clay is a good source of essential elements.

  7. Calculation of the moderator temperature coefficient of reactivity for miniature neutron source reactors

    International Nuclear Information System (INIS)

    This paper presents results of the evaluated group constants for fuel and other important materials of the Miniature Neutron Source Reactor (Mnr) and the moderator temperature coefficient of reactivity through global reactor calculation. In this study the group constants were calculated with the WIMSD code and the global reactor calculation is accomplished by the CITATION code. This work also presents a method for evaluation of the moderator temperature coefficient of reactivity at different temperatures and it's average value in a range of temperature directly through the values of moderator temperature for MNSRs. This method provides simple analytical representation convenient for reactor kinetics calculation and reactor safety assessment. (author)

  8. A safe private nuclear tool-the miniature neutron source reactor

    International Nuclear Information System (INIS)

    The prototype miniature neutron source reactor (MNSR) designed by China Institute of Atomic Energy has been operated successfully for more than 3 years and the practical experience enriches the original design idea. The commercial MNSR is under study design and develop in following aspects: 1. Prolonging the control rod cycle duration and core burn-up life; 2. Increasing the neutron flux per unit power. Obviously, the MNSR will show more advantages in extending application area and in providing the core using low envichment fuel. (Liu)

  9. Xenon poisoning calculation code for miniature neutron source reactor (MNSR)

    International Nuclear Information System (INIS)

    In line with the actual requirements and based upon the specific characteristics of MNSR, a revised point-reactor model was adopted to model MNSR's xenon poisoning. The corresponding calculation code, MNSRXPCC (Xenon Poisoning Calculation Code for MNSR), was developed and tested by the Shanghai MNSR data

  10. Xenon poisoning calculation code for miniature neutron source reactor (MNSR)

    Institute of Scientific and Technical Information of China (English)

    2001-01-01

    In line with the actual requirements and based upon the specific char acteristics of MNSR, a revised point-reactor model was adopted to model MNSR's xenon poisoning. The corresponding calculation code, MNSRXPCC (Xenon Poison ing Calculation Code for MNSR), was developed and tested by the Shanghai MNSR data.

  11. The low power miniature neutron source reactors: Design, safety and applications

    International Nuclear Information System (INIS)

    The Chinese Miniature Neutron Source Reactor (MNSR) is a low power research reactor with maximum thermal neutron flux of 1 x 1012 n.cm-2.s-1 in one of its inner irradiation channels and thermal power of approximately 30kW. The MNSR is designed based on the Canadian SLOWPOKE reactor and is one of the smallest commercial research reactors presently available in the world. Its commercial versions currently in operation in China, Ghana, Iran, Nigeria, Pakistan and Syria, is considered as an excellent tool for Neutron Activation Analysis (NAA), training of Scientist, and Engineers in nuclear science and technology and small scale radioisotope production. The paper highlights the basic design and theory of the commercial MNSR, its safety features, applications and advantages over the Chinese Prototype. The experimental flux characteristics determined in this work and in similar studies by other authors reveal that the commercial MNSR has more flux stability, longer life span, higher negative temperature coefficient of reactivity and low under-moderation compared to its prototype in China. The result shows that the facility is safe for reactor physics experiments, teaching and training of students and also ideal for application of NAA for the determination of elemental composition of biological and environmental samples. It can also be a useful tool for geochemical and soil fertility mapping. (author)

  12. The behavior of reactor power and flux resulting from changes in core-coolant temperature for a miniature neutron source reactor

    International Nuclear Information System (INIS)

    In this work, measurements were performed to verify the theoretical predictions of reactor power and flux parameters that result from changes in core inlet temperature (Tin) and the temperature difference between the coolant inlet and outlet (ΔT) in the Nigeria Research Reactor-1 (NIRR-1), which is a Miniature Neutron Source Reactor (MNSR). The measured data shows that there is a strong dependence of the reactor power on coolant temperature in agreement with the design of MNSR. The experimental parameters were found to be in good agreement with data obtained using a semi-empirical relationship between the reactor power, flux parameters, core inlet temperature, and the coolant temperature rise. The relationship was therefore used to predict the power level of NIRR-1 from its neutron flux parameters to which it has been found to be proportional. The variation of Tin and ΔT with the reactor power and flux was also investigated and the results obtained are hereby discussed

  13. Installation of permanent cadmium-lined channel as a means for increasing epithermal NAA capabilities of miniature neutron source reactors

    International Nuclear Information System (INIS)

    Highlights: • High demand for epithermal neutrons necessitated the need of a permanent cadmium-line. • We reported the design specifications, preliminary studies done and steps followed. • Reactivity worth of the old channel = 0.12 mk and the new channel = 0.336 mk. • Temperature coefficient = −0.1 mk/°C and control rod worth coefficient = 0.023 mk/mm. • The work is a useful tool to the MNSR community for upgrading their reactors. -- Abstract: High demand for epithermal neutrons by the clients of the Nigerian Research Reactor-1 (NIRR-1), a Miniature Neutron Source Reactor (MNSR) has necessitated the need to explore avenues for increasing epithermal Neutron Activation Analysis (NAA) capabilities of the reactor. Safety and flux stability simulations were done by our group using Monte Carlo Transport Code MCNP5 for permanent cadmium line inside the irradiation channel of NIRR-1 and compared with the ones reported by other MNSR groups. The results of all these simulations revealed that the effect of cadmium-line on safety and flux stability is very minimal in the outer channel than in the inner channel. We have reported here the design specifications, preliminary studies done, steps followed in installation and measurements done in the pre and post installation of the permanent cadmium-line in outer channel of the reactor. We measured the reactivity worth of the old and new channel and readjusted the reactor's core excess reactivity after the installation. Results obtained are: reactivity worth of the old channel (0.12 mk), reactivity worth of the new channel = 0.336 mk, temperature coefficient = −0.1 mk/°C, control rod worth coefficient = 0.023 mk/mm and the core excess reactivity = 3.85 mk. We have also measured the radial and axial flux distribution in the channels of the reactor after the installation. The installation of the permanent cadmium-lined channel reported here will not only boost the sample handling capabilities of NIRR-1 but will also

  14. Simulation of the Syrian miniature neutron source reactor for training operators on the analysis of its anticipated operational accidents

    International Nuclear Information System (INIS)

    For the purpose of training operators and other educational aspects, a mathematical model capable of assessing potential accidents and safety implications of the research Miniature Neutron Source Reactor (MNSR) has been developed. The model considers relevant physical phenomena that govern the core such as reactor kinetics, reactivity feedbacks due to coolant temperature and xenon, and thermal hydraulics. Natural convection and point kinetics including the prompt jump and complete mixing approximations were employed. Peak power, reactivity core load, core outlet temperature, and other variables are predicted during self-limiting power excursions. Compared to related references, close results have been obtained. The simulating model proves to be a useful tool to train operators and students to assess qualitatively the transient behaviour of the MNSR as a result of sudden reactivity insertion in the core. In addition, the model was utilized to verify some of the design basis accidents already presented in both the Safety Analysis Report (SAR) and the Commissioning Report (CR) of the reactor. Furthermore, the dynamic model generates other core variables that are of interest to update the SAR on one side, and confirms others measured and reported in the CR

  15. Simulation of the Syrian Miniature Neutron Source Reactor for training operators on the analysis of its anticipated operational accidents

    International Nuclear Information System (INIS)

    For the purpose of training operators and other educational aspects, a mathematical model capable of assessing potential accidents and safety implications of the research Miniature Neutron Source Reactor (MNSR) has been developed. The model considers relevant physical phenomena that govern the core such as reactor kinetics, reactivity feed-backs due to coolant temperature and xenon, and thermal hydraulics. Natural convection and point kinetics including the prompt jump and complete mixing approximations were employed. Peak power, reactivity core load, core outlet temperature, and other variables are predicted during self-limiting power excursions. Compared to related references, close results have been obtained. The simulating model proves to be a useful tool to train operators and students to assess qualitatively the transient behaviour of the MNSR as a result of sudden reactivity insertion in the core. In addition, the model was utilized to verify some of the design basis accidents already presented in both the Safety Analysis Report (SAR) and the Commissioning Report (CR) of the reactor, as can be seen in Table 1. Furthermore, the dynamic model generates other core variables that are of interest to update the SAR on one side, and confirms others measured and reported in the CR. (author)

  16. Graphite reflecting characteristics and shielding factors for Miniature Neutron Source Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Albarhoum, M., E-mail: pscientific1@aec.org.s [Department of Nuclear Engineering, Atomic Energy Commission, P.O. Box 6091, Damascus (Syrian Arab Republic)

    2011-01-15

    The usability of graphite as a reflector for MNSRs is investigated in this paper. Its use is optimized and shielding factors are calculated. Graphite seems to be compatible with liquid water. As a reflector, graphite proves to be usable as well, but it decreases the fuel cycle lifetime by about 7%. To optimize its use the average worth reactivity of the unit volume was assessed for the different modes of filling the shim tray of the reactor with graphite which were: RIOS, RIOC, ROIS, and ROIC modes for the radial direction, and ASM, and ACM modes for the axial one. This quantity was found to be maximum for the ROIC mode reaching more than 0.01 mk/cm{sup 3}. The shielding factors for the radial and axial filling modes were found to be 0.7101 and 0.6266, respectively.

  17. Graphite reflecting characteristics and shielding factors for Miniature Neutron Source Reactors

    International Nuclear Information System (INIS)

    The usability of graphite as a reflector for MNSRs is investigated in this paper. Its use is optimized and shielding factors are calculated. Graphite seems to be compatible with liquid water. As a reflector, graphite proves to be usable as well, but it decreases the fuel cycle lifetime by about 7%. To optimize its use the average worth reactivity of the unit volume was assessed for the different modes of filling the shim tray of the reactor with graphite which were: RIOS, RIOC, ROIS, and ROIC modes for the radial direction, and ASM, and ACM modes for the axial one. This quantity was found to be maximum for the ROIC mode reaching more than 0.01 mk/cm3. The shielding factors for the radial and axial filling modes were found to be 0.7101 and 0.6266, respectively.

  18. An investigation of the effect of the upper beryllium reflector on the moderator temperature coefficient of reactivity of miniature neutron source reactors

    Energy Technology Data Exchange (ETDEWEB)

    Binh, Do Quang [Univ. of Technical Education, Ho Chi Minh City (Viet Nam); Hai, Nguyen Hoang [Centre for Research and Development of Radiation Technology, Ho Chi Minh City (Viet Nam)

    2014-11-15

    In this paper, an investigation on the dependence of the effective multiplication factor, k{sub eff}, on moderator temperature for various thicknesses of the upper beryllium reflector in reactor conditions with different fuel burnups for the Miniature Neutron Source Reactor is carried out. Based on the linear dependence of k{sub eff} on moderator temperature, an approach to calculate the moderator temperature coefficient of reactivity, α{sub T}, at different temperatures and its average value, anti α{sub T}, in a range of temperatures directly through the moderator temperature is developed. Calculations are performed to evaluate the effect of change in the upper reflector thickness on the moderator temperature coefficient of reactivity for the fresh core and reactor conditions with different fuel burnups. Calculated results indicate that anti α{sub T} increases with the increased beryllium thickness, but decreases with the increasing fuel burnup. Analysis of calculated results provides an additional insight into the relation of the upper reflector thickness, the neutron energy spectrum in the reactor core, and the moderator temperature coefficient of reactivity.

  19. Equalisation of Transient Temperature Profile Within the Fuel Pin of a Miniature Neutron Source Reactor (MNSR During Total Loss of Coolant

    Directory of Open Access Journals (Sweden)

    Christian Amevi Adjei

    2010-10-01

    Full Text Available Transient temperature distributions in cylindrical fuel element of Ghana Research Reactor-1 (GHARR-1 Miniature Neutron Source Reactor (MNSR following sudden total loss of cooling have been investigated. The loss of cooling in the reactor core resulting from a blockage of the inner orifice of coolant flow channels was assumed to occur during normal operations and led to sudden shut dow n of the reactor. The objective was to analyse the transient behaviour by solving analytically the heat transfer equation using Bessel functions and also develop from first principle the transient temperature equations for the fuel element. Results obtained during a sudden total lost of cooling showed a high transient temperature distribution at the centre of the fuel element, with the surface of the fuel clad recording the least temperature. The transient temperature distribution decreased from the centre of the fuel element to the surface of the fuel clad and followed a parabolic decay pattern which after increase in tim e follow ed an equalisation pattern. During sudden shut down, since there w as no heat generated and decay heat , the rate at which the fuel elem ent was cooled w as directly proportional to time.

  20. Miniature neutron sources: Thermal neutron sources and their uses in the academic field

    International Nuclear Information System (INIS)

    The three levels of thermal neutron sources are introduced: university laboratory sources; infrastructure sources; and world-class sources; and the needs for each kind and their inter-dependence will be emphasized. A description of the possibilities for university sources based on α-Be reactions or spontaneous fission emission is given, and current experience with them is described. A new generation of infrastructure sources is needed to continue the regional programs based on small reactors. Some possibilities for accelerator sources that could meet this need are considered

  1. Agreement of 10 September 1991 between the International Atomic Energy Agency and the Government of the Islamic Republic of Pakistan for the application of safeguards in connection with the supply of a miniature neutron source reactor from the People's Republic of China

    International Nuclear Information System (INIS)

    The document reproduces the text of the Agreement of 10 September 1991, between the Government of the Islamic Republic of Pakistan and the International Atomic Energy Agency for the application of safeguards in connection with the supply of a miniature neutron source reactor from the People's Republic of China. The Agreement was approved by the Agency's Board of Governors on 20 February 1990 and entered into force upon signature on 10 September 1991

  2. Development of miniaturized specimens for the study of neutron irradiation/plasma exposure synergistic effects on candidate fusion reactor materials

    International Nuclear Information System (INIS)

    The aim of this work is to choose a miniaturized specimen version relevant for testing candidate fusion reactor materials including mechanical testing after combined neutron irradiation/plasma exposure in a fission reactor. The material examined was reactor pressure vessel type steel in irradiated and aged (unirradiated) conditions. Comparative standard impact, three point bend and small punch tests were conducted. It is established that there is a possibility of miniaturization of irradiated steel experimental specimens by means of proper specimens type choice with mass reducing from ∼40 (Charpy) to 0.4 g (small plates). (orig.)

  3. Neutron source structure for nuclear reactors

    International Nuclear Information System (INIS)

    Purpose: To improve the compatibility between metal beryllium forming a neutron source and a metal cladding material at a high temperature. Constitution: An intermediate layer made of silicon or silicone-beryllium alloy is put between metal beryllium forming a neutron source and a metal cladding material containing the metal beryllium in a tightly sealed manner. By the disposition of the intermediate layer, the compatibility between the metal beryllium and the metal cladding material is improved, by which the neutron source can be operated satisfactorily over a long time use at a high temperature of 500 - 7000C. (Moriyama, K.)

  4. Commissioning of the Opal reactor cold neutron source

    International Nuclear Information System (INIS)

    Full text: At OPAL, Australia's first cold neutron facility will form an essential part of the reactor's research programs. Fast neutrons, born in the core of a reactor, interact with a cryogenic material, in this case liquid deuterium, to give them very low energies (10meV). A cold neutron flux of 1.4 10E14n/cm2/s is expected, with a peak in the energy spectrum at 4.2meV. The cold neutron source reached cryogenic conditions for the first time in late 2005. The cold neutron source operates with a sub-cooled liquid Deuterium moderator at 24K. The moderator chamber, which contains the deuterium, has been constructed from AlMg5. The thermosiphon and moderator chamber are cooled by helium gas, in a natural convection thermosiphon loop. The helium refrigeration system utilises the Brayton cycle, and is fully insulated within a high vacuum environment. Despite the proximity of the cold neutron source to the reactor core, it has been considered as effectively separate to the reactor system, due to the design of its special vacuum containment vessel. As OPAL is a multipurpose research reactor, used for beam research as well as radiopharmaceutical production and industrial irradiations, the cold neutron source has been designed with a stand-by mode, to maximise production. The stand-by mode is a warm operating mode using only gaseous deuterium at ambient temperatures (∼ 300K), allowing for continued reactor operations whilst parts of the cold source are unavailable or in maintenance. This is the first time such a stand-by feature has been incorporated into a cold source facility

  5. Condensed matter and materials research using neutron diffraction and spectroscopy: reactor and pulsed neutron sources

    International Nuclear Information System (INIS)

    The paper provides a short, and partial view of the neutron scattering technique applied to condensed matter and materials research. Reactor and accelerator-based neutron spectrometers are discussed, together with examples of research projects that illustrate the puissance and modern applications of neutron scattering. Some examples are chosen to show the range of facilities available at the medium flux reactor operated by Casaccia ENEA, Roma and the advanced, pulsed spallation neutron source at the Rutherford Appleton Laboratory, Oxfordshire. (author)

  6. Battery powered tabletop pulsed neutron source based on a sealed miniature plasma focus device

    Science.gov (United States)

    Rout, R. K.; Mishra, P.; Rawool, A. M.; Kulkarni, L. V.; Gupta, Satish C.

    2008-10-01

    The development of a novel and portable tabletop pulsed neutron source is presented. It is a battery powered neutron tube based on a miniature plasma focus (PF) device having all metal-sealed components. The tube, fuelled with deuterium gas, generates neutrons because of D-D fusion reactions. The inner diameter and the length of the tube are 3.4 cm and 8 cm, respectively. A single capacitor (200 J, 4.0 µF, 10 nH) of compact size (17 cm × 15 cm × 13 cm, 6.5 kg) is used as the energy driver. A power supply system charges the capacitor to 10 kV in 10 s and also provides a 30 kV trigger pulse to the spark gap. An input of 24 V dc (7.5 A) to the power supply system is provided by two rechargeable batteries (each 12 V, 7.5 A, 20 h). The device has produced neutrons for 150 shots within a period of 120 days in a very reliable manner without purging the deuterium gas between the shots. For the first 50 shots, the average yield is (1.6 ± 0.3) × 106 neutrons/shot in 4π sr with a pulse width of 23.4 ± 3.3 ns. The estimated neutron energy is 2.47 ± 0.22 MeV. The neutron production reduces slowly and reaches the detection threshold value of 3 × 105 neutrons/shot towards the last shots. The device produces neutrons in a similar manner on evacuation and refilling. The height of the mounted PF tube with the capacitor and the spark gap is 35 cm. The complete setup comprising the capacitor with spark gap, the PF tube, the power supply system with two batteries and the control panel weighs only 23 kg.

  7. High Flux Isotope Reactor cold neutron source reference design concept

    Energy Technology Data Exchange (ETDEWEB)

    Selby, D.L.; Lucas, A.T.; Hyman, C.R. [and others

    1998-05-01

    In February 1995, Oak Ridge National Laboratory`s (ORNL`s) deputy director formed a group to examine the need for upgrades to the High Flux Isotope Reactor (HFIR) system in light of the cancellation of the Advanced neutron Source Project. One of the major findings of this study was that there was an immediate need for the installation of a cold neutron source facility in the HFIR complex. In May 1995, a team was formed to examine the feasibility of retrofitting a liquid hydrogen (LH{sub 2}) cold source facility into an existing HFIR beam tube. The results of this feasibility study indicated that the most practical location for such a cold source was the HB-4 beam tube. This location provides a potential flux environment higher than the Institut Laue-Langevin (ILL) vertical cold source and maximizes the space available for a future cold neutron guide hall expansion. It was determined that this cold neutron beam would be comparable, in cold neutron brightness, to the best facilities in the world, and a decision was made to complete a preconceptual design study with the intention of proceeding with an activity to install a working LH{sub 2} cold source in the HFIR HB-4 beam tube. During the development of the reference design the liquid hydrogen concept was changed to a supercritical hydrogen system for a number of reasons. This report documents the reference supercritical hydrogen design and its performance. The cold source project has been divided into four phases: (1) preconceptual, (2) conceptual design and testing, (3) detailed design and procurement, and (4) installation and operation. This report marks the conclusion of the conceptual design phase and establishes the baseline reference concept.

  8. High Flux Isotope Reactor cold neutron source reference design concept

    International Nuclear Information System (INIS)

    In February 1995, Oak Ridge National Laboratory's (ORNL's) deputy director formed a group to examine the need for upgrades to the High Flux Isotope Reactor (HFIR) system in light of the cancellation of the Advanced neutron Source Project. One of the major findings of this study was that there was an immediate need for the installation of a cold neutron source facility in the HFIR complex. In May 1995, a team was formed to examine the feasibility of retrofitting a liquid hydrogen (LH2) cold source facility into an existing HFIR beam tube. The results of this feasibility study indicated that the most practical location for such a cold source was the HB-4 beam tube. This location provides a potential flux environment higher than the Institut Laue-Langevin (ILL) vertical cold source and maximizes the space available for a future cold neutron guide hall expansion. It was determined that this cold neutron beam would be comparable, in cold neutron brightness, to the best facilities in the world, and a decision was made to complete a preconceptual design study with the intention of proceeding with an activity to install a working LH2 cold source in the HFIR HB-4 beam tube. During the development of the reference design the liquid hydrogen concept was changed to a supercritical hydrogen system for a number of reasons. This report documents the reference supercritical hydrogen design and its performance. The cold source project has been divided into four phases: (1) preconceptual, (2) conceptual design and testing, (3) detailed design and procurement, and (4) installation and operation. This report marks the conclusion of the conceptual design phase and establishes the baseline reference concept

  9. Application of a triga research reactor as the neutron source for a production neutron radiography facility

    International Nuclear Information System (INIS)

    GA Technologies Inc. (GA) has developed a Stationary Neutron Radiography System (SNRS) using a 250-1000 KW TRIGA reactor as the neutron source. The partially below ground reactor will be equipped with four vertical beam tubes originating in the reactor graphite reflector and installed tangential to the core to provide a strong current of thermal neutrons with minimum gamma-ray contamination. The vertical beam tubes interface with rugged component positioning systems designed to handle intact F-111 aircraft wings, partial A-10 aircraft wings, pyrotechnics, and other honeycomb aircraft structures. The SNRS will be equipped with real-time, near-real-time, and film-radiographic imaging systems to provide a broad spectrum of capability for detection or corrosion of entrained moisture in large aircraft panels. (author)

  10. Monte Carlo calculation of neutron generation time in critical reactor and subcritical reactor with an external source

    International Nuclear Information System (INIS)

    The neutron generation time Λ plays an important role in the reactor kinetics. However, it is not straightforward nor standard in most continuous energy Monte Carlo codes which are able to calculate the prompt neutron lifetime lp directly. The difference between Λ and lp are sometimes very apparent. As very few delayed neutrons are produced in the reactor, they have little influence on Λ. Thus on the assumption that no delayed neutrons are produced in the system, the prompt kinetics equations for critical system and subcritical system with an external source are proposed. And then the equations are applied to calculating Λ with pulsed neutron technique using Monte Carlo. Only one fission neutron source is simulated with Monte Carlo in critical system while two neutron sources, including a fission source and an external source, are simulated for subcritical system. Calculations are performed on both critical benchmarks and subcritical system with an external source and the results are consistent with the reference values. (author)

  11. Source driven breeding thermal power reactors, Pt. 2. Using lithium-free neutron sources

    International Nuclear Information System (INIS)

    The feasibility of fusion devices operating in the semi-catalyzed deuterium (SCD) mode and of high energy proton accelerators to provide the neutron sources for driving subcritical breeding light water power reactors is assessed. The assessment is done by studying the energy balance of the resulting source driven light water reactors (SDLWR) and comparing it with the energy balance of the reference light water hybrid reactors (LWHR) driven by a D-T neutron source (DT-LWHR). The conditions the non-DT neutron sources should satisfy in order to make the SDLWR viable power reactors are identified. It is found that in order for a SCD-LWHR to have the same overall efficiency as a DT-LWHR, the fusion energy gain of the SCD device should be at least one half that of the DT device. The efficiency of ADLWRs using uranium targets is comparable with that of DT-LWHRs having a fusion energy gain of unity. Advantages and disadvantages of the DT-LWHR, SCD-LWHR and ADLWR are discussed

  12. Burn up calculations for the Iranian miniature reactor: A reliable and safe research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Faghihi, F. [Department of Nuclear Engineering, School of Engineering, Shiraz University, Shiraz 71345 (Iran, Islamic Republic of); Research Center for Radiation Protection, Shiraz University, Shiraz (Iran, Islamic Republic of)], E-mail: faghihif@shirazu.ac.ir; Mirvakili, S.M. [Department of Nuclear Engineering, School of Engineering, Shiraz University, Shiraz 71345 (Iran, Islamic Republic of)

    2009-06-15

    Presenting neutronic calculations pertaining to the Iranian miniature research reactor is the main goal of this article. This is a key to maintaining safe and reliable core operation. The following reactor core neutronic parameters were calculated: clean cold core excess reactivity ({rho}{sub ex}), control rod and shim worth, shut down margin (SDM), neutron flux distribution of the reactor core components, and reactivity feedback coefficients. Calculations for the fuel burnup and radionuclide inventory of the Iranian miniature neutron source reactor (MNSR), after 13 years of operational time, are carried out. Moreover, the amount of uranium burnup and produced plutonium, the concentrations and activities of the most important fission products, the actinide radionuclides accumulated, and the total radioactivity of the core are estimated. Flux distribution for both water and fuel temperature increases are calculated and changes of the central control rod position are investigated as well. Standard neutronic simulation codes WIMS-D4 and CITATION are employed for these studies. The input model was validated by the experimental data according to the final safety analysis report (FSAR) of the reactor. The total activity of the MNSR core is calculated including all radionuclides at the end of the core life and it is found to be equal to 1.3 x 10{sup 3}Ci. Our investigation shows that the reactor is operating under safe and reliable conditions.

  13. Burn up calculations for the Iranian miniature reactor: A reliable and safe research reactor

    International Nuclear Information System (INIS)

    Presenting neutronic calculations pertaining to the Iranian miniature research reactor is the main goal of this article. This is a key to maintaining safe and reliable core operation. The following reactor core neutronic parameters were calculated: clean cold core excess reactivity (ρex), control rod and shim worth, shut down margin (SDM), neutron flux distribution of the reactor core components, and reactivity feedback coefficients. Calculations for the fuel burnup and radionuclide inventory of the Iranian miniature neutron source reactor (MNSR), after 13 years of operational time, are carried out. Moreover, the amount of uranium burnup and produced plutonium, the concentrations and activities of the most important fission products, the actinide radionuclides accumulated, and the total radioactivity of the core are estimated. Flux distribution for both water and fuel temperature increases are calculated and changes of the central control rod position are investigated as well. Standard neutronic simulation codes WIMS-D4 and CITATION are employed for these studies. The input model was validated by the experimental data according to the final safety analysis report (FSAR) of the reactor. The total activity of the MNSR core is calculated including all radionuclides at the end of the core life and it is found to be equal to 1.3 x 103Ci. Our investigation shows that the reactor is operating under safe and reliable conditions.

  14. China Experimental Fast Reactor(CEFR)——Criterion of Criticality for Reactor With External Neutron Source

    Institute of Scientific and Technical Information of China (English)

    ZHAOYu-sen

    2003-01-01

    There is a neutron source with 109 s-1 neutrons in core of CEFR during start up test and operation of CEFR. For judging the criticality of reactor with external neutron source and near criticality, it is important that the neutron level changes in core with time must be understood after introducing positive reactivity to core with external neutron source.

  15. INR TRIGA Research Reactors: A Neutron Source for Radioisotopes and Materials Investigation

    International Nuclear Information System (INIS)

    At the INR there are 2 high intensity neutron sources. These sources are in fact the two nuclear TRIGA reactors: TRIGA SSR 14 MW and TRIGA ACPR. TRIGA stationary reactor is provided with several in-core irradiation channels. Other several out-of-core irradiation channels are located in the vertical channels in the beryllium reflector blocks. The maximum value of the thermal neutron flux (E14 cm-2s-1 and of fast neutron flux (E>1 MeV) is 6.89×1013 cm-2s-1. For neutron activation analysis both reactors are used and k0-NAA method has been implemented. At INR Pitesti a prompt gamma ray neutron activation analysis devices has been designed, manufactured ant put into operation. For nuclear materials properties investigation neutron radiography methods was developed in INR. For these purposes two neutron radiography devices were manufacture, one of them underwater and other one dry. The neutron beams are used for investigation of materials properties and components produced or under development for applications in the energy sector (fission and fusion). At TRIGA 14 MW reactor a neutron difractormeter and a SANS devices are available for material residual stress and texture measurements. TRIGA 14 MW reactor is used for medical and industrial radioisotopes production (131I, 125I, 192Ir, etc) and a method for 99Mo-99Tc production from fission is under developing. At INR Pitesti several special programmes for new types of nuclear fuel behavior characterization are under development. (author)

  16. Detailed flux calculations for the conceptual design of the Advanced Neutron Source Reactor

    International Nuclear Information System (INIS)

    A detailed MCNP model of the Advanced Neutron Source Reactor has been developed. All reactor components inside the reflector tank were included, and all components were highly segmented. Neutron and photon multigroup flux spectra have been calculated for each segment in the model, and thermal-to-fast neutron flux ratios were determined for each component segment. Axial profiles of the spectra are provided for all components of the reactor. Individual segment statistical uncertainties were limited wherever possible, and the group fluxes for all important reflector components have a standard deviation below 10%

  17. Experimental determination of the neutron source for the Argonauta reactor subcritical assembly

    Energy Technology Data Exchange (ETDEWEB)

    Renke, Carlos A.C.; Furieri, Rosanne C.A.A.; Pereira, Joao C.S.; Voi, Dante L.; Barbosa, Andre L.N., E-mail: renke@ien.gov.b [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

    2011-07-01

    The utilization of a subcritical assembly for the determination of nuclear parameters in a multiplier medium requires a well defined neutron source to carry out the experiments necessary for the acquisition of the desired data. The Argonauta research reactor installed at the Instituto de Engenharia Nuclear has a subcritical assembly, under development, to be coupled at the upper part of the reactor core that will provide the needed neutrons emerging from its internal thermal column made of graphite. In order to perform neutronic calculations to compare with the experimental results, it is necessary a precise knowledge of the emergent neutron flux that will be used as neutron source in the subcritical assembly. In this work, we present the thermal neutron flux profile determined experimentally via the technique of neutron activation analysis, using dysprosium wires uniformly distributed at the top of the internal thermal neutron column of the Argonauta reactor and later submitted to a detection system using Geiger-Mueller detector. These experimental data were then compared with those obtained through neutronic calculation using HAMMER and CITATION codes in order to validate this calculation system and to define a correct neutron source distribution to be used in the subcritical assembly. This procedure avoids a coupled neutronic calculation of the subcritical assembly and the reactor core. It has also been determined the dimension of the graphite pedestal to be used in the bottom of the subcritical assembly tank in order to smooth the emergent neutron flux at the reactor top. Finally, it is estimated the thermal neutron flux inside the assembly tank when filled with water. (author)

  18. Implementation and training methodology of subcritical reactors neutronic calculations triggered by external neutron source and applications

    International Nuclear Information System (INIS)

    This works had as goal to investigate calculational methodologies on subcritical source driven reactor, such as Accelerator Driven Subcritical Reactor (ADSR) and Fusion Driven Subcritical Reactor (FDSR). Intense R and D has been done about these subcritical concepts, mainly due to Minor Actinides (MA) and Long Lived Fission Products (LLFP) transmutation possibilities. In this work, particular emphasis has been given to: (1) complement and improve calculation methodology with neutronic transmutation and decay capabilities and implement it computationally, (2) utilization of this methodology in the Coordinated Research Project (CRP) of the International Atomic Energy Agency Analytical and Experimental Benchmark Analysis of ADS and in the Collaborative Work on Use of Low Enriched Uranium in ADS, especially in the reproduction of the experimental results of the Yalina Booster subcritical assembly and study of a subcritical core of IPEN / MB-01 reactor, (3) to compare different nuclear data libraries calculation of integral parameters, such as keff and ksrc, and differential distributions, such as spectrum and flux, and nuclides inventories and (4) apply the develop methodology in a study that may help future choices about dedicated transmutation system. The following tools have been used in this work: MCNP (Monte Carlo N particle transport code), MCB (enhanced version of MCNP that allows burnup calculation) and NJOY to process nuclear data from evaluated nuclear data files. (author)

  19. UCN sources at external beams of thermal neutrons. An example of PIK reactor

    Science.gov (United States)

    Lychagin, E. V.; Mityukhlyaev, V. A.; Muzychka, A. Yu.; Nekhaev, G. V.; Nesvizhevsky, V. V.; Onegin, M. S.; Sharapov, E. I.; Strelkov, A. V.

    2016-07-01

    We consider ultracold neutron (UCN) sources based on a new method of UCN production in superfluid helium (4He). The PIK reactor is chosen as a perspective example of application of this idea, which consists of installing 4He UCN source in the beam of thermal or cold neutrons and surrounding the source with moderator-reflector, which plays the role of cold neutron (CN) source feeding the UCN source. CN flux in the source can be several times larger than the incident flux, due to multiple neutron reflections from the moderator-reflector. We show that such a source at the PIK reactor would provide an order of magnitude larger density and production rate than an analogous source at the ILL reactor. We estimate parameters of 4He source with solid methane (CH4) or/and liquid deuterium (D2) moderator-reflector. We show that such a source with CH4 moderator-reflector at the PIK reactor would provide the UCN density of ~1·105 cm-3, and the UCN production rate of ~2·107 s-1. These values are respectively 1000 and 20 times larger than those for the most intense UCN user source. The UCN density in a source with D2 moderator-reflector would reach the value of ~2·105 cm-3, and the UCN production rate would be equal ~8·107 s-1. Installation of such a source in a beam of CNs would slightly increase the density and production rate.

  20. UCN sources at external beams of thermal neutrons. An example of PIK reactor

    CERN Document Server

    Lychagin, E V; Muzychka, A Yu; Nekhaev, G V; Nesvizhevsky, V V; Onegin, M S; Sharapov, E I; Strelkov, A V

    2015-01-01

    We consider ultracold neutron (UCN) sources based on a new method of UCN production in superfluid helium (4He). The PIK reactor is chosen as a perspective example of the application of this idea, which consists of installing a 4He UCN source in a beam of thermal or cold neutrons and surrounding the source with a moderator-reflector, which plays the role of a source of cold neutrons (CNs) feeding the UCN source. The CN flux in the source can be several times larger than the incident flux, due to multiple neutron reflections from the moderator-reflector. We show that such a source at the PIK reactor would provide an order of magnitude larger density and production rate than an analogous source at the ILL reactor. We estimate parameters of a 4He source with solid methane (CH4) or/and liquid deuterium (D2) moderator-reflector. We show that such a source with CH4 moderator-reflector at the PIK reactor would provide the UCN density of ~1x10^5 1/cm^3, and the UCN production rate of ~2x10^7 1/s. These values are resp...

  1. Diffraction Experiments at the IBR-2 Pulsed Reactor with Methane Cold Neutron Source

    CERN Document Server

    Balagurov, A M; Mironova, G M; Pole, A V; Simkin, V G

    2000-01-01

    A new methane cold neutron source has been tested at the IBR-2 pulsed reactor at the Frank Laboratory of Neutron Physics. In a paper the results of experiments at neutron diffractometers HRFD and DN-2 which are placed at the IBR-2 from the methane moderator side are given. A comparison with the results obtained with the conventional water comb-like moderator is performed. The perspectives of the cold source for various kinds of neutron diffraction experiments, including atomic and magnetic structural analysis and real time experiments are discussed. It is shown, that for a huge number of the experiments which are performing at both HRFD and DN-2 the methane cold neutron source provides the better conditions than water comb-like moderator.

  2. Advanced Neutron Source Reactor zoning, shielding, and radiological optimization guide

    International Nuclear Information System (INIS)

    In the design of major nuclear facilities, it is important to protect both humans and equipment excessive radiation dose. Past experience has shown that it is very effective to apply dose reduction principles early in the design of a nuclear facility both to specific design features and to the manner of operation of the facility, where they can aid in making the facility more efficient and cost-effective. Since the appropriate choice of radiological controls and practices varies according to the case, each area of the facility must be analyzed for its radiological impact, both by itself and in interactions with other areas. For the Advanced Neutron Source (ANS) project, a large relational database will be used to collect facility information by system and relate it to areas. The database will also hold the facility dose and shielding information as it is produced during the design process. This report details how the ANS zoning scheme was established and how the calculation of doses and shielding are to be done

  3. IRPhE-TAPIRO-ARCHIVE, Fast neutron source reactor primary documents, reactor physics experiments

    International Nuclear Information System (INIS)

    Description of program or function: The TAPIRO reactor, located in the ENEA Casaccia Centre near Rome, is a highly enriched uranium fast neutron facility. The nominal power is 5 kW (thermal) and the core centre neutron flux is 4. E12/cm2/s. The reactor has a cylindrical core (12.6 cm diameter and 10.9 cm height) made of 93.5 % enriched uranium metal in a uranium-molybdenum alloy which is totally reflected by copper. The copper reflector (cylindrical-shaped) is divided into two concentric zones: the inner zone, up to 17.4 cm radius, and the outer zone up to 40.0 cm. Radius. The height of the reflector is 72.0 cm. The reactor is surrounded by borate concrete shielding about 170 cm thick. The maximum depth available for the epithermal column is 160 cm, reserved for filter/moderator materials. The graphite column extends to the external reflector boundary where a sector of the outer copper reflector has been removed and then characterized by a very hard neutron spectrum. Along the column the spectrum gradually softens up to thermal values - Different materials can be interposed, such as U-nat, Pb, Fe, etc. to reproduce spectrum transition conditions at interface points between regions with different compositions. - Activation foils can be used for activation analysis with threshold energies in the fast, intermediate and epithermal regions. The archive contains reports characterising the reactor and describes experiments carried out, together with the corresponding data

  4. Power spectral analysis for a subcritical reactor system driven by a pulsed spallation neutron source

    International Nuclear Information System (INIS)

    A series of power spectral analyses for a thermal subcritical reactor system driven by a pulsed spallation neutron source was carried out at Kyoto University Critical Assembly (KUCA), to determine the prompt-neutron decay constant of the Accelerator-Driven System (ADS). High-energy protons (100 MeV) obtained from the fixed field alternating gradient accelerator were injected onto a lead-bismuth target, whereby the spallation neutrons were generated. In the cross-power spectral density between time-sequence signal data of two neutron detectors, many delta-function-like peaks at the integral multiple of pulse repetition frequency could be observed. However, no continuous reactor-noise component could be measured. This is because these detectors have too high count-rate to be placed closely to the core. From the point data of these delta-function-like peaks, the prompt-neutron decay constant could be determined. At a slightly subcritical state, the decay constant was consistent with that obtained by a previous power spectral analysis for a pulsed 14 MeV neutron source and by a pulsed neutron experiment. At another deeply subcritical state, however, the present analysis leads to an underestimate of the decay constant. (author)

  5. Conceptual Study of Transmutation Reactor Based on LAR Tokamak Fusion Neutron Source

    International Nuclear Information System (INIS)

    A compact tokamak reactor concept as a 14 MeV neutron source is desirable from an economic viewpoint for a fusion-driven transmutation reactor. For the optimal design of a reactor, a radial build of reactor components has to be determined by considering the plasma physics and engineering constraints which interrelate various reactor components. In a transmutation reactor, design of blanket and shield play a key role in determining the size of a reactor: the blanket should produce enough tritium for tritium self-sufficiency, the transmutation rate of waste has to be maximized, and the shield should provide sufficient protection for the superconducting toroidal field (TF) coil. To determine the radial build of the blanket and the shield, not only a radiation transport analysis but also a burn-up calculation were coupled with the system analysis and it allowed the self-consistent determination of the design parameters of a transmutation reactor. For neutronic optimization of the blanket and the shield, the quantities such as the tritium breeding ratio (TBR), nuclear heating, radiation damage to the toroidal field coil have to be calculated and burn-up rates of Li, actinides and fission products have to be calculated. Thus the neutronic analysis need to be coupled in the system analysis. In most of the previous system studies, neutronic calculation and plasma analysis are performed separately, so blanket and shield size was determined independently from the reactor size. In this work, to account for the interrelation of blanket and shield with the other components of a reactor system, we coupled the system analysis with one-dimensional neutronic calculation to determine the reactor parameters in self consistent manner. LAR (Low Aspect Ratio) tokamak plasma has the potential of high β operation with high bootstrap current fractions. In the LAR tokamak reactor, the radial build of TF coil(TFC) and the shield play a key role in determining the size of a reactor since it

  6. A Liquid Deuterium Cold Neutron Source for the NIST Research Reactor - Conceptual Design

    International Nuclear Information System (INIS)

    The NBSR is a 20 MW research reactor operated by the NIST Center for Neutron Research (NCNR) as a neutron source providing beams of thermal and cold neutrons for research in materials science, fundamental physics and nuclear chemistry. A large, 550 mm diameter beam port was included in the design for the installation of a cold neutron source, and the NCNR has been steadily improving its cold neutron facilities for more than 25 years. Monte Carlo Simulations have shown that a liquid deuterium (LD2) source will provide a gain of 1.5 to 2 for neutron wavelengths between 4 A and 10 A with respect to the existing liquid hydrogen cold source. The conceptual design for the LD2 source will be presented. To achieve these gains, a large volume (35 litres) of LD2 is required. The expected nuclear heat load in this moderator and vessel is 4000 W. A new, 7 kW helium refrigerator is being built to provide the necessary cooling capacity; it will be completely installed and tested early in 2014. The source will operate as a naturally circulating thermosiphon, very similar to the horizontal cold source in the High Flux Reactor at the Institut Laue-Langevin (ILL) in Grenoble. A condenser will be mounted on the reactor face about 2 m above the source providing the gravitational head to supply the source with LD2. The system will always be open to a 16 m3 ballast tank to store the deuterium at 500 kPa when the refrigerator is not operating, and providing a passively safe response to a refrigerator trip. It is expected the source will operate at 23 K, the boiling point of LD2 at 100 kPa. All components will be surrounded by a blanket of helium to prevent the possibility of creating a flammable mixture of deuterium and air. A design for the cryostat assembly, consisting of the moderator chamber, vacuum jacket, helium containment and a heavy water cooling water jacket, has been completed and sent to procurement to solicit bids. It is expected that installation of the LD2 cold source will

  7. Preseving Old Beijing-in Miniature

    Institute of Scientific and Technical Information of China (English)

    WuWei

    2003-01-01

    The other day I went to an artgallery near Yongdingmen for a painting exhibition. To my pleasant surprise, an exhibition of models of old Beijing architecture was also on display. The exquisite diminutive pavilions,towers, palace halls and folk dwellings were amazing. ] was so eager to know who made the delicate

  8. Moderators for the design of a cold neutron source for the RA 3 reactor

    International Nuclear Information System (INIS)

    The cold neutron production of hydrogenous materials was studied, taking into account their radiation resistance, for the conceptual design of a cold neutron source for the RA-3 reactor.Low spontaneous release of chemical energy was found in mesitylene.Libraries for hidrogen in mesitylene were generated using the NJOY nuclear processing system and the resulting cross sections were compared with experimental data.Good agreement between measurements and calculations was found in those cases where data are available.New calculations using the RA-3 geometry and these validated libraries will be performed

  9. Steady-state thermal-hydraulic design analysis of the Advanced Neutron Source reactor

    International Nuclear Information System (INIS)

    The Advanced Neutron Source (ANS) is a research reactor that is planned for construction at Oak Ridge National Laboratory. This reactor will be a user facility with the major objective of providing the highest continuous neutron beam intensities of any reactor in the world. Additional objectives for the facility include providing materials irradiation facilities and isotope production facilities as good as, or better than, those in the High Flux Isotope Reactor. To achieve these objectives, the reactor design uses highly subcooled heavy water as both coolant and moderator. Two separate core halves of 67.6-L total volume operate at an average power density of 4.5 MW(t)/L, and the coolant flows upward through the core at 25 m/s. Operating pressure is 3.1 MPa at the core inlet with a 1.4-MPa pressure drop through the core region. Finally, in order to make the resources available for experimentation, the fuel is designed to provide a 17-d fuel cycle with an additional 4 d planned in each cycle for the refueling process. This report examines the codes and models used to develop the thermal-hydraulic design for ANS, as well as the correlations and physical data; evaluates thermal-hydraulic uncertainties; reports on thermal-hydraulic design and safety analysis; describes experimentation in support of the ANS reactor design and safety analysis; and provides an overview of the experimental plan

  10. Awareness, Preference, Utilization, and Messaging Research for the Spallation Neutron Source and High Flux Isotope Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Bryant, Rebecca [Bryant Research, LLC; Kszos, Lynn A [ORNL

    2011-03-01

    Oak Ridge National Laboratory (ORNL) offers the scientific community unique access to two types of world-class neutron sources at a single site - the Spallation Neutron Source (SNS) and the High Flux Isotope Reactor (HFIR). The 85-MW HFIR provides one of the highest steady-state neutron fluxes of any research reactor in the world, and the SNS is one of the world's most intense pulsed neutron beams. Management of these two resources is the responsibility of the Neutron Sciences Directorate (NScD). NScD commissioned this survey research to develop baseline information regarding awareness of and perceptions about neutron science. Specific areas of investigative interest include the following: (1) awareness levels among those in the scientific community about the two neutron sources that ORNL offers; (2) the level of understanding members of various scientific communities have regarding benefits that neutron scattering techniques offer; and (3) any perceptions that negatively impact utilization of the facilities. NScD leadership identified users of two light sources in North America - the Advanced Photon Source (APS) at Argonne National Laboratory and the National Synchrotron Light Source (NSLS) at Brookhaven National Laboratory - as key publics. Given the type of research in which these scientists engage, they would quite likely benefit from including the neutron techniques available at SNS and HFIR among their scientific investigation tools. The objective of the survey of users of APS, NSLS, SNS, and HFIR was to explore awareness of and perceptions regarding SNS and HFIR among those in selected scientific communities. Perceptions of SNS and FHIR will provide a foundation for strategic communication plan development and for developing key educational messages. The survey was conducted in two phases. The first phase included qualitative methods of (1) key stakeholder meetings; (2) online interviews with user administrators of APS and NSLS; and (3) one

  11. Awareness, Preference, Utilization, and Messaging Research for the Spallation Neutron Source and High Flux Isotope Reactor

    International Nuclear Information System (INIS)

    Oak Ridge National Laboratory (ORNL) offers the scientific community unique access to two types of world-class neutron sources at a single site - the Spallation Neutron Source (SNS) and the High Flux Isotope Reactor (HFIR). The 85-MW HFIR provides one of the highest steady-state neutron fluxes of any research reactor in the world, and the SNS is one of the world's most intense pulsed neutron beams. Management of these two resources is the responsibility of the Neutron Sciences Directorate (NScD). NScD commissioned this survey research to develop baseline information regarding awareness of and perceptions about neutron science. Specific areas of investigative interest include the following: (1) awareness levels among those in the scientific community about the two neutron sources that ORNL offers; (2) the level of understanding members of various scientific communities have regarding benefits that neutron scattering techniques offer; and (3) any perceptions that negatively impact utilization of the facilities. NScD leadership identified users of two light sources in North America - the Advanced Photon Source (APS) at Argonne National Laboratory and the National Synchrotron Light Source (NSLS) at Brookhaven National Laboratory - as key publics. Given the type of research in which these scientists engage, they would quite likely benefit from including the neutron techniques available at SNS and HFIR among their scientific investigation tools. The objective of the survey of users of APS, NSLS, SNS, and HFIR was to explore awareness of and perceptions regarding SNS and HFIR among those in selected scientific communities. Perceptions of SNS and FHIR will provide a foundation for strategic communication plan development and for developing key educational messages. The survey was conducted in two phases. The first phase included qualitative methods of (1) key stakeholder meetings; (2) online interviews with user administrators of APS and NSLS; and (3) one-on-one interviews

  12. A feasibility study of the Tehran research reactor as a neutron source for BNCT.

    Science.gov (United States)

    Kasesaz, Yaser; Khalafi, Hossein; Rahmani, Faezeh; Ezati, Arsalan; Keyvani, Mehdi; Hossnirokh, Ashkan; Shamami, Mehrdad Azizi; Monshizadeh, Mahdi

    2014-08-01

    Investigation on the use of the Tehran Research Reactor (TRR) as a neutron source for Boron Neutron Capture Therapy (BNCT) has been performed by calculating and measuring energy spectrum and the spatial distribution of neutrons in all external irradiation facilities, including six beam tubes, thermal column, and the medical room. Activation methods with multiple foils and a copper wire have been used for the mentioned measurements. The results show that (1) the small diameter and long length beam tubes cannot provide sufficient neutron flux for BNCT; (2) in order to use the medical room, the TRR core should be placed in the open pool position, in this situation the distance between the core and patient position is about 400 cm, so neutron flux cannot be sufficient for BNCT; and (3) the best facility which can be adapted for BNCT application is the thermal column, if all graphite blocks can be removed. The epithermal and fast neutron flux at the beginning of this empty column are 4.12×10(9) and 1.21×10(9) n/cm(2)/s, respectively, which can provide an appropriate neutron beam for BNCT by designing and constructing a proper Beam Shaping Assembly (BSA) structure.

  13. NEUTRONIC REACTOR

    Science.gov (United States)

    Anderson, H.L.

    1960-09-20

    A nuclear reactor is described comprising fissionable material dispersed in graphite blocks, helium filling the voids of the blocks and the spaces therebetween, and means other than the helium in thermal conductive contact with the graphite for removing heat.

  14. NEUTRONIC REACTORS

    Science.gov (United States)

    Anderson, J.B.

    1960-01-01

    A reactor is described which comprises a tank, a plurality of coaxial steel sleeves in the tank, a mass of water in the tank, and wire grids in abutting relationship within a plurality of elongated parallel channels within the steel sleeves, the wire being provided with a plurality of bends in the same plane forming adjacent parallel sections between bends, and the sections of adjacent grids being normally disposed relative to each other.

  15. Performance of the solid deuterium ultra-cold neutron source at the pulsed reactor TRIGA Mainz

    CERN Document Server

    Karch, J; Beck, M; Eberhardt, K; Hampel, G; Heil, W; Kieser, R; Reich, T; Trautmann, N; Ziegner, M

    2013-01-01

    The performance of the solid deuterium ultra-cold neutron source at the pulsed reactor TRIGA Mainz with a maximum peak energy of 10 MJ is described. The solid deuterium converter with a volume of V=160 cm3 (8 mol), which is exposed to a thermal neutron fluence of 4.5x10^13 n/cm2, delivers up to 550 000 UCN per pulse outside of the biological shield at the experimental area. UCN densities of ~ 10/cm3 are obtained in stainless steel bottles of V ~ 10 L resulting in a storage efficiency of ~20%. The measured UCN yields compare well with the predictions from a Monte Carlo simulation developed to model the source and to optimize its performance for the upcoming upgrade of the TRIGA Mainz into a user facility for UCN physics.

  16. Performance of the solid deuterium ultra-cold neutron source at the pulsed reactor TRIGA Mainz

    Science.gov (United States)

    Karch, J.; Sobolev, Yu.; Beck, M.; Eberhardt, K.; Hampel, G.; Heil, W.; Kieser, R.; Reich, T.; Trautmann, N.; Ziegner, M.

    2014-04-01

    The performance of the solid deuterium ultra-cold neutron (UCN) source at the pulsed reactor TRIGA Mainz with a maximum peak energy of 10MJ is described. The solid deuterium converter with a volume of cm3 (8mol), which is exposed to a thermal neutron fluence of n/cm2, delivers up to 240000 UCN ( m/s) per pulse outside the biological shield at the experimental area. UCN densities of 10 cm3 are obtained in stainless-steel bottles of 10 L. The measured UCN yields compare well with the predictions from a Monte Carlo simulation developed to model the source and to optimize its performance for the upcoming upgrade of the TRIGA Mainz into a user facility for UCN physics.

  17. Performance of the solid deuterium ultra-cold neutron source at the pulsed reactor TRIGA Mainz

    International Nuclear Information System (INIS)

    The performance of the solid deuterium ultra-cold neutron (UCN) source at the pulsed reactor TRIGA Mainz with a maximum peak energy of 10MJ is described. The solid deuterium converter with a volume of V=160 cm3 (8mol), which is exposed to a thermal neutron fluence of 4.5 x 1013 n/cm2, delivers up to 240000 UCN (v ≤ 6 m/s) per pulse outside the biological shield at the experimental area. UCN densities of ∼ 10 cm3 are obtained in stainless-steel bottles of V ∼ 10 L. The measured UCN yields compare well with the predictions from a Monte Carlo simulation developed to model the source and to optimize its performance for the upcoming upgrade of the TRIGA Mainz into a user facility for UCN physics. (orig.)

  18. Study of the External Neutron Source Effect on TRU Burning in a Sub-critical Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Zafar, Zafar Iqbal; Kim, Myung Hyun [Kyung Hee University, Yongin (Korea, Republic of)

    2015-05-15

    One of the drawback points of nuclear power is the production of highly radioactive and long lasting waste isotopes during power production. Therefore, most important design requirement of future nuclear option should have a potential to burn selectively long-lived fission products (LLFP) and long-lived minor actinides (LLMA). However, there is no way to burn them selectively in the reactor core. Practical method of waste transmutation should rely on selective separation of them from spent nuclear fuel of power plants. Under the proliferation concern, direct separation of trans-uranic isotopes (TRU) from pyro-reprocessing plant became a feasible option in our country. Even though social political agreement is not matured as well as technical feasibility, current study is done based on basic assumptions; TRU and LLFP is separated from spent fuel of nuclear power plants. The remaining neutrons (among the external 3%) very few in number (less than 1% in any case) being very energetic (above three MeV or so) do cause much more fissions per neutron than their counterparts but, because of their overall low population they do not have any significant and decisive influence in the overall reactor performance. Currently, entire study is limited to the source neutron energy of 20 MeV only. In future, it is expected to get reasonably plausible fixed source dependent difference in the TRU burning by using tabulated data for the neutrons of higher energy (up to 250 MeV at least). Secondly, a clearer picture is expected if the TRU loading was increased from the current value of 133 kg to few metric tons, as is the case in most of the existing reactors.

  19. Study of the External Neutron Source Effect on TRU Burning in a Sub-critical Reactor

    International Nuclear Information System (INIS)

    One of the drawback points of nuclear power is the production of highly radioactive and long lasting waste isotopes during power production. Therefore, most important design requirement of future nuclear option should have a potential to burn selectively long-lived fission products (LLFP) and long-lived minor actinides (LLMA). However, there is no way to burn them selectively in the reactor core. Practical method of waste transmutation should rely on selective separation of them from spent nuclear fuel of power plants. Under the proliferation concern, direct separation of trans-uranic isotopes (TRU) from pyro-reprocessing plant became a feasible option in our country. Even though social political agreement is not matured as well as technical feasibility, current study is done based on basic assumptions; TRU and LLFP is separated from spent fuel of nuclear power plants. The remaining neutrons (among the external 3%) very few in number (less than 1% in any case) being very energetic (above three MeV or so) do cause much more fissions per neutron than their counterparts but, because of their overall low population they do not have any significant and decisive influence in the overall reactor performance. Currently, entire study is limited to the source neutron energy of 20 MeV only. In future, it is expected to get reasonably plausible fixed source dependent difference in the TRU burning by using tabulated data for the neutrons of higher energy (up to 250 MeV at least). Secondly, a clearer picture is expected if the TRU loading was increased from the current value of 133 kg to few metric tons, as is the case in most of the existing reactors

  20. Physics design for the Brookhaven Medical Research Reactor epithermal neutron source

    International Nuclear Information System (INIS)

    A collaborative effort by researchers at the Idaho National Engineering Laboratory and the Brookhaven National Laboratory has resulted in the design and implementation of an epithermal-neutron source at the Brookhaven Medical Research Reactor (BMRR). Large aluminum containers, filled with aluminum oxide tiles and aluminum spacers, were tailored to pre-existing compartments on the animal side of the reactor facility. A layer of cadmium was used to minimize the thermal-neutron component. Additional bismuth was added to the pre-existing bismuth shield to minimize the gamma component of the beam. Lead was also added to reduce gamma streaming around the bismuth. The physics design methods are outlined in this paper. Information available to date shows close agreement between calculated and measured beam parameters. The neutron spectrum is predominantly in the intermediate energy range (0.5 eV - 10 keV). The peak flux intensity is 6.4E + 12 n/(m2.s.MW) at the center of the beam on the outer surface of the final gamma shield. The corresponding neutron current is 3.8E + 12 n/(m2.s.MW). Presently, the core operates at a maximum of 3 MW. The fast-neutron KERMA is 3.6E-15 cGy/(n/m2) and the gamma KERMA is 5.0E-16 cGY/(n/m2) for the unperturbed beam. The neutron intensity falls off rapidly with distance from the outer shield and the thermal flux realized in phantom or tissue is strongly dependent on the beam-delimiter and target geometry

  1. High-density ultracold neutron sources for the WWR-M and PIK reactors

    Energy Technology Data Exchange (ETDEWEB)

    Serebrov, A. P., E-mail: serebrov@pnpi.spb.ru; Fomin, A. K.; Kharitonov, A. G.; Lyamkin, V. A.; Prudnikov, D. V.; Ivanov, S. A.; Erykalov, A. N.; Onegin, M. S. [National Research Centre “Kurchatov Institute”, Petersburg Nuclear Physics Institute (Russian Federation); Gridnev, K. A. [St. Petersburg State University (Russian Federation)

    2016-01-15

    It is proposed to equip the PIK and WWR-M research reactors at the Petersburg Nuclear Physics Institute (PNPI) with high-density ultracold neutron (UCN) sources, where UCNs will be obtained based on the effect of their accumulation in superfluid helium (due to the specific features of this quantum fluid). The maximum UCN storage time in superfluid helium is obtained at temperatures on the order of 1 K. These sources are expected to yield UCN densities of 10{sup 3}–10{sup 4} cm{sup –3}, i.e., approximately three orders of magnitude higher than the density from existing UCN sources throughout the world. The development of highest intensity UCN sources will make PNPI an international center of fundamental UCN research.

  2. High-density ultracold neutron sources for the WWR-M and PIK reactors

    Science.gov (United States)

    Serebrov, A. P.; Fomin, A. K.; Kharitonov, A. G.; Lyamkin, V. A.; Prudnikov, D. V.; Ivanov, S. A.; Erykalov, A. N.; Onegin, M. S.; Gridnev, K. A.

    2016-01-01

    It is proposed to equip the PIK and WWR-M research reactors at the Petersburg Nuclear Physics Institute (PNPI) with high-density ultracold neutron (UCN) sources, where UCNs will be obtained based on the effect of their accumulation in superfluid helium (due to the specific features of this quantum fluid). The maximum UCN storage time in superfluid helium is obtained at temperatures on the order of 1 K. These sources are expected to yield UCN densities of 103-104 cm-3, i.e., approximately three orders of magnitude higher than the density from existing UCN sources throughout the world. The development of highest intensity UCN sources will make PNPI an international center of fundamental UCN research.

  3. Containment performance analyses for the Advanced Neutron Source Reactor at the Oak Ridge National Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Kim, S.H.; Taleyarkhan, R.P.; Georgevich, V.

    1992-10-01

    This paper discusses salient aspects of methodology, assumptions, and modeling of various features related to estimation of source terms from two conservatively scoped severe accident scenarios in the Advanced Neutron Source (ANS) reactor at the Oak Ridge National Laboratory. Various containment configurations are considered for steaming-pool-type accidents and an accident involving molten core-concrete interaction. Several design features (such as rupture disks) are examined to study containment response during postulated severe accidents. Also, thermal-hydraulic response of the containment and radionuclide transport and retention in the containment are studied. The results are described as transient variations of source terms for each scenario, which are to be used for studying off-site radiological consequences and health effects for these postulated severe accidents. Also highlighted will be a comparison of source terms estimated by two different versions of the MELCOR code.

  4. Neutron sources and applications

    Energy Technology Data Exchange (ETDEWEB)

    Price, D.L. [ed.] [Argonne National Lab., IL (United States); Rush, J.J. [ed.] [National Inst. of Standards and Technology, Gaithersburg, MD (United States)

    1994-01-01

    Review of Neutron Sources and Applications was held at Oak Brook, Illinois, during September 8--10, 1992. This review involved some 70 national and international experts in different areas of neutron research, sources, and applications. Separate working groups were asked to (1) review the current status of advanced research reactors and spallation sources; and (2) provide an update on scientific, technological, and medical applications, including neutron scattering research in a number of disciplines, isotope production, materials irradiation, and other important uses of neutron sources such as materials analysis and fundamental neutron physics. This report summarizes the findings and conclusions of the different working groups involved in the review, and contains some of the best current expertise on neutron sources and applications.

  5. Neutron sources and applications

    International Nuclear Information System (INIS)

    Review of Neutron Sources and Applications was held at Oak Brook, Illinois, during September 8--10, 1992. This review involved some 70 national and international experts in different areas of neutron research, sources, and applications. Separate working groups were asked to (1) review the current status of advanced research reactors and spallation sources; and (2) provide an update on scientific, technological, and medical applications, including neutron scattering research in a number of disciplines, isotope production, materials irradiation, and other important uses of neutron sources such as materials analysis and fundamental neutron physics. This report summarizes the findings and conclusions of the different working groups involved in the review, and contains some of the best current expertise on neutron sources and applications

  6. Reactor neutron dosimetry

    International Nuclear Information System (INIS)

    An analysis of requirements and possibilities for experimental neutron spectrum determination during the reactor pressure vessel surveil lance programme is given. Fast neutron spectrum and neutron dose rate were measured in the Fast neutron irradiation facility of our TRIGA reactor. It was shown that the facility can be used for calibration of neutron dosimeters and for irradiation of samples sensitive to neutron radiation. The investigation of the unfolding algorithm ITER was continued. Based on this investigations are two specialized unfolding program packages ITERAD and ITERGS written this year. They are able to unfold data from activation detectors and NaI(T1) gamma spectrometer respectively

  7. A new method to determine in situ the transmission of a neutron-guide system at a reactor source

    CERN Document Server

    Haan, V O D; Gommers, R M; Labohm, F; Well, A A V; De Leege, P F A; Schebetov, A; Pusenkov, V

    2002-01-01

    In this paper, a description of a new method to determine the transmission of neutron guides after they are installed in a beam-tube at a reactor source is given. The method is based on activation measurements of gold foils at the entrance of the beam-tube and at the exit of the neutron guides compared to Monte-Carlo calculations. In this method, a quality factor is defined as the ratio between the actual transmission and the theoretical maximum attainable transmission. This method is used to determine the quality of an optimised neutron-guide system developed for beam-tube R2 of the HOR. The HOR is a pool-type nuclear research reactor at the Interfaculty Reactor Institute of the Delft University of Technology. It is shown that the quality factors of the newly installed neutron guides are between 0.49 and 0.63.

  8. External neutron source anomalies analysis using Hurst's exponent for the Myrrha reactor

    Energy Technology Data Exchange (ETDEWEB)

    Henrice Junior, Edson; Goncalves, Alessandro C., E-mail: ejunior@con.ufrj.br [Coordenacao dos Cursos de Pos-Graduacao em Engenharia (COPPE/UFRJ), Rio de Janeiro, RJ (Brazil). Programa de Engenharia Nuclear; Palma, Daniel A.P., E-mail: dapalma@cnen.gov.br [Comissao Nacional de Energia Nuclear (CNEN), Rio de Janeir, RJ (Brazil)

    2015-07-01

    Anomalous diffusion is usually marked by the non-linear growth of the variance in time, that is, the diffusion will be considered as anomalous if there is a deviation in the behaviour described before. This paper aims to identify anomalies in the neutron flux during the operation of an ADS (Accelerator Driven System) nuclear reactor as a result of a trip that originates in the proton accelerator as per project bases, from two different calculation methods for Hurst exponents. These methods are the R/S Method and the Detrended Fluctuation Analysis (DFA) Method. For that the Myrrha Reactor will be simulated using the SERPENT code and the neutron source will be subjected to a production peak at a given instant. The Hurst exponent has a direct application on determining the order of derivatives in fractional point-kinetics equations and the estimate for the fractional derivative can be related as being twice that of Hurst's exponent, according to the co-variance function in the Gauss' processes. After getting the Hurst's exponent a numerically solution is proposed. This subject being a theme very much in focus nowadays. (author)

  9. Pulsed neutron sources for epithermal neutrons

    International Nuclear Information System (INIS)

    It is shown how accelerator based neutron sources, giving a fast neutron pulse of short duration compared to the neutron moderation time, promise to open up a new field of epithermal neutron scattering. The three principal methods of fast neutron production: electrons, protons and fission boosters will be compared. Pulsed reactors are less suitable for epithermal neutrons and will only be briefly mentioned. The design principle of the target producing fast neutrons, the moderator and reflector to slow them down to epithermal energies, and the cell with its beam tubes and shielding will all be described with examples taken from the new Harwell electron linac to be commissioned in 1978. A general comparison of pulsed neutron performance with reactors is fraught with difficulties but has been attempted. Calculation of the new pulsed source fluxes and pulse widths is now being performed but we have taken the practical course of basing all comparisons on extrapolations from measurements on the old 1958 Harwell electron linac. Comparisons for time-of-flight and crystal monochromator experiments show reactors to be at their best at long wavelengths, at coarse resolution, and for experiments needing a specific incident wavelength. Even existing pulsed sources are shown to compete with the high flux reactors in experiments where the hot neutron flux and the time-of-flight methods can be best exploited. The sources under construction can open a new field of inelastic neutron scattering based on energy transfer up to an electron volt and beyond

  10. The Design and Construction of a Cold Neutron Source for Use in the Cornell University Triga Reactor

    Science.gov (United States)

    Young, Lydia Jane

    A cold neutron source has been designed and constructed for insertion into the 6"-radial beam port of the Cornell University TRIGA reactor for use with a neutron guide tube system. The main differences between this cold source and other existing sources are the use of heat conduction as the method of cooling and the use of mesitylene (1,3,5 -trimethylbenzene; melting point, 228(DEGREES)K; boiling point, 437(DEGREES)K) as the moderating material. This thesis describes the design and construction details of the cold neutron source, discusses its safety aspects, and presents its cryogenic performance curves and also the results of a test of its neutron moderating ability. A closed-cycle helium gas refrigerator, located outside the reactor shielding, cools the 500 cm('3) moderator chamber and its surrounding heat shield by heat conduction through two meters of copper and rod tubing. Moderator temperatures of 23 (+OR-) 3(DEGREES)K have been achieved. Mesitylene, a hydrocarbon, is an effective cold moderator because even at low temperatures the weakly hindered rotational motions of its methyl groups enable the absorption of small amounts of energy ((LESSTHEQ) 0.005 eV) from neutrons. The use of mesitylene simplifies the cold source design because it is a liquid at room temperature and thus, the usual design safeguards required for sources using gaseous moderators are not necessary. Moreover, the flammability of mesitylene is much smaller than that of hydrogen and methane, which are the commonly used cold moderators. A method of transferring and handling the mesitylene, a carcinogen, was devised to ensure minimal contact with this substance. To test the neutron moderating ability of the cold neutron source, an out-of-reactor neutron transmission experiment was performed with the moderator chamber first at room temperature and then at about 23(DEGREES)K. The results indicate that the neutron energy spectrum is strongly shifted to lower energies when the chamber is cold

  11. NEUTRONIC REACTOR FUEL COMPOSITION

    Science.gov (United States)

    Thurber, W.C.

    1961-01-10

    Uranium-aluminum alloys in which boron is homogeneously dispersed by adding it as a nickel boride are described. These compositions have particular utility as fuels for neutronic reactors, boron being present as a burnable poison.

  12. Neutron measurements performed with miniature fission chambers

    International Nuclear Information System (INIS)

    This research aims at proposing solutions regarding instruments to perform neutron flow measurements in nuclear power reactors and to perform measurements of the reaction rates of highly radioactive transuranic fissile elements in experimental reactors. This research is also part of a program aimed at the adjustment of the Cadarache cross section set. The report defines the instrumentation, recalls the operation of fission chambers, discusses the implemented instrumentation, and discusses the obtained measurements

  13. Preliminary fracture analysis of the core pressure boundary tube for the Advanced Neutron Source Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Schulz, K.C. [Univ. of Turabo, Gurabo, Puerto (Puerto Rico). College of Engineering; Yahr, G.T. [Oak Ridge National Lab., TN (United States)

    1995-08-01

    The outer core pressure boundary tube (CPBT) of the Advanced neutron Source (ANS) reactor being designed at Oak Ridge National Laboratory is currently specified as being composed of 6061-T6 aluminum. ASME Boiler and Pressure Vessel Code fracture analysis rules for nuclear components are based on the use of ferritic steels; the expressions, tables, charts and equations were all developed from tests and analyses conducted for ferritic steels. Because of the nature of the Code, design with thin aluminum requires analytical approaches that do not directly follow the Code. The intent of this report is to present a methodology comparable to the ASME Code for ensuring the prevention of nonductile fracture of the CPBT in the ANS reactor. 6061-T6 aluminum is known to be a relatively brittle material; the linear elastic fracture mechanics (LEFM) approach is utilized to determine allowable flaw sizes for the CPBT. A J-analysis following the procedure developed by the Electric Power Research Institute was conducted as a check; the results matched those for the LEFM analysis for the cases analyzed. Since 6061-T6 is known to embrittle when irradiated, the reduction in K{sub Q} due to irradiation is considered in the analysis. In anticipation of probable requirements regarding maximum allowable flaw size, a survey of nondestructive inspection capabilities is also presented. A discussion of probabilistic fracture mechanics approaches, principally Monte Carlo techniques, is included in this report as an introduction to what quantifying the probability of nonductile failure of the CPBT may entail.

  14. RELAP5 analyses of two hypothetical flow reversal events for the advanced neutron source reactor

    Energy Technology Data Exchange (ETDEWEB)

    Chen, N.C.J.; Wendel, M.W.; Yoder, G.L. Jr. [Oak Ridge National Lab., TN (United States)

    1995-09-01

    This paper presents RELAP5 results of two hypothetical, low flow transients analyzed as part of the Advanced Neutron Source Reactor safety program. The reactor design features four independent coolant loops (three active and one in standby), each containing a main curculation pump (with battery powered pony motor), heat exchanger, an accumulator, and a check valve. The first transient assumes one of these pumps fails, and additionally, that the check valve in that loop remains stuck in the open position. This accident is considered extremely unlikely. Flow reverses in this loop, reducing the core flow because much of the coolant is diverted from the intact loops back through the failed loop. The second transient examines a 102-mm-diam instantaneous pipe break near the core inlet (the worst break location). A break is assumed to occur 90 s after a total loss-of-offsite power. Core flow reversal occurs because accumulator injection overpowers the diminishing pump flow. Safety margins are evaluated against four thermal limits: T{sub wall}=T{sub sat}, incipient boiling, onset of significant void, and critical heat flux. For the first transient, the results show that these limits are not exceeded (at a 95% non-exceedance probability level) if the pony motor battery lasts 30 minutes (the present design value). For the second transient, the results show that the closest approach of the fuel surface temperature to the local saturation temperature during core flow reversal is about 39{degrees}C. Therefore the fuel remains cool during this transient. Although this work is done specifically for the ANSR geometry and operating conditions, the general conclusions may be applicable to other highly subcooled reactor systems.

  15. Californium-252 neutron sources

    International Nuclear Information System (INIS)

    Major production programs for the Savannah River reactors and the High Flux Isotopes Reactor at Oak Ridge have made 252Cf one of the most available and, at the USAEC's sales price of $10/μg, one of the least-expensive isotopic neutron sources. Reactor production has totaled approximately 2 g, and, based on expected demand, an additional 10 g will be produced in the next decade. The approximately 800 mg chemically separated to date has been used to prepare over 600 neutron sources. Most, about 500, have been medical sources containing 1 to 5 μg of 252Cf plated in needles for experimental cancer therapy studies. The remainder have generally been point sources containing 10 μg to 12 mg of oxide for activation, well logging, or radiography uses. Bulk sources have also been supplied to the commercial encapsulators. The latest development has been the production of 252Cf cermet wire which can be cut into almost contamination-free lengths of the desired 252Cf content. Casks are available for transport of sources up to 50 mg. Subcritical assemblies have been developed to multiply the source neutrons by a factor of 10 to 40, and collimators and thermalizers have also been extensively developed to shape the neutron flux and energy distributions for special applications. (U.S.)

  16. The research reactor TRIGA Mainz. A neutron source for versatile applications in research and education

    International Nuclear Information System (INIS)

    Currently, four research reactors with a thermal power ranging from 0.1 to 23 MWth are in operation in Germany and one new reactor (20 MWth) is under construction. The TRIGA Mark II reactor at the Institut fuer Kernchemie became first critical on August 3, 1965. It can be operated in the steady state mode with a maximum power of 100 kWth and in the pulse mode with a peak power of 250 MWth. A survey of the research programmes carried out at the TRIGA Mainz is given covering a wide range of applications in basic and applied science in nuclear chemistry, nuclear- and particle physics. Furthermore, the reactor is used for neutron activation analysis and for education and training of students and technical personal. (orig.)

  17. V2:Performance of the solid deuterium ultra-cold neutron source at the pulsed reactor TRIGA Mainz

    OpenAIRE

    Karch, J.; Sobolev, Yu.; M. Beck; Eberhardt, K.; Hampel, G.; Heil, W.; Kieser, R.; Reich, T.; Trautmann, N.; Ziegner, M.

    2013-01-01

    The performance of the solid deuterium ultra-cold neutron source at the pulsed reactor TRIGA Mainz with a maximum peak energy of 10 MJ is described. The solid deuterium converter with a volume of V=160 cm3 (8 mol), which is exposed to a thermal neutron fluence of 4.5x10^13 n/cm2, delivers up to 550 000 UCN per pulse outside of the biological shield at the experimental area. UCN densities of ~ 10/cm3 are obtained in stainless steel bottles of V ~ 10 L resulting in a storage efficiency of ~20%....

  18. Neutron spatial flux profile measurement in compact subcritical system using miniature neutron detectors

    International Nuclear Information System (INIS)

    A zero power multiplying assembly in subcritical regime serves as a benchmark for validating subcritical reactor physics. The utilization of a subcritical assembly for the determination of nuclear parameters in a multiplying medium requires a well-defined neutron flux to carry out the experiments. For this it is necessary to know the neutron flux profile inside a subcritical system. A compact subcritical assembly BRAHMMA has been developed in India. The experimental channels in this assembly are typically less than 8 mm diameter. This requires use of miniature detectors that can be mounted in these experimental channels. In this article we present the thermal neutron flux profile measurement in a compact subcritical system using indigenously developed miniature gas filled neutron detectors. These detectors were specially designed and fabricated considering the restrictive dimensional requirements of the subcritical core. Detectors of non-standard size with various sensitivities, from 0.4 to 0.001 cps/nv were used for neutron flux of interest ranging from 103 to 107 n-cm−2 s−1. A comparison of measured neutron flux using these detectors and simulated Monte Carlo calculations are also presented in this article

  19. Analyses of the reflector tank, cold source, and beam tube cooling for ANS reactor. [Advanced Neutron Source (ANS)

    Energy Technology Data Exchange (ETDEWEB)

    Marland, S. (Tennessee Univ., Knoxville, TN (United States))

    1992-07-01

    This report describes my work as an intern with Martin Marietta Energy Systems, Inc., in the summer of 1991. I was assigned to the Reactor Technology Engineering Department, working on the Advanced Neutron Source (ANS). My first project was to select and analyze sealing systems for the top of the diverter/reflector tank. This involved investigating various metal seals and calculating the forces necessary to maintain an adequate seal. The force calculations led to an analysis of several bolt patterns and lockring concepts that could be used to maintain a seal on the vessel. Another project involved some pressure vessel stress calculations and the calculation of the center of gravity for the cold source assembly. I also completed some sketches of possible cooling channel patterns for the inner vessel of the cold source. In addition, I worked on some thermal design analyses for the reflector tank and beam tubes, including heat transfer calculations and assisting in Patran and Pthermal analyses. To supplement the ANS work, I worked on other projects. I completed some stress/deflection analyses on several different beams. These analyses were done with the aid of CAASE, a beam-analysis software package. An additional project involved bending analysis on a carbon removal system. This study was done to find the deflection of a complex-shaped beam when loaded with a full waste can.

  20. Progress in the neutronic core conversion (HEU-LEU) analysis of Ghana research reactor-1.

    Energy Technology Data Exchange (ETDEWEB)

    Anim-Sampong, S.; Maakuu, B. T.; Akaho, E. H. K.; Andam, A.; Liaw, J. J. R.; Matos, J. E.; Nuclear Engineering Division; Ghana Atomic Energy Commission; Kwame Nkrumah Univ. of Science and Technology

    2006-01-01

    The Ghana Research Reactor-1 (GHARR-1) is a commercial version of the Miniature Neutron Source Reactor (MNSR) and has operated at different power levels since its commissioning in March 1995. As required for all nuclear reactors, neutronic and thermal hydraulic analysis are being performed for the HEU-LEU core conversion studies of the Ghana Research Reactor-1 (GHARR-1) facility, which is a commercial version of the Miniature Neutron Source Reactor (MNSR). Stochastic Monte Carlo particle transport methods and tools (MCNP4c/MCNP5) were used to fine-tune a previously developed 3-D MCNP model of the GHARR-1 facility and perform neutronic analysis of the 90.2% HEU reference and candidate LEU (UO{sub 2}, U{sub 3}Si{sub 2}, U-9Mo) fresh cores with varying enrichments from 12.6%-19.75%. In this paper, the results of the progress made in the Monte Carlo neutronic analysis of the HEU reference and candidate LEU fuels are presented. In particular, a comparative performance assessment of the LEU with respect to neutron flux variations in the fission chamber and experimental irradiation channels are highlighted.

  1. Neutronics methods, models, and applications at the Idaho National Engineering Laboratory for the advanced neutron source reactor three-element core design

    International Nuclear Information System (INIS)

    A summary of the methods and models used to perform neutronics analyses on the Advanced Neutron Source reactor three-element core design is presented. The applications of the neutral particle Monte Carlo code MCNP are detailed, as well as the expansion of the static role of MCNP to analysis of fuel cycle depletion calculations. Results to date of these applications are presented also. A summary of the calculations not yet performed is also given to provide a open-quotes to-doclose quotes list if the project is resurrected

  2. NEUTRONIC REACTOR FUEL ELEMENT

    Science.gov (United States)

    Picklesimer, M.L.; Thurber, W.C.

    1961-01-01

    A chemically nonreactive fuel composition for incorporation in aluminum- clad, plate type fuel elements for neutronic reactors is described. The composition comprises a mixture of aluminum and uranium carbide particles, the uranium carbide particles containing at least 80 wt.% UC/sub 2/.

  3. Global shielding analysis for the three-element core advanced neutron source reactor under normal operating conditions

    Energy Technology Data Exchange (ETDEWEB)

    Slater, C.O.; Bucholz, J.A.

    1995-08-01

    Two-dimensional discrete ordinates radiation transport calculations were performed for a model of the three-element core Advanced Neutron Source reactor design under normal operating conditions. The core consists of two concentric upper elements and a lower element radially centered in the annulus between the upper elements. The initial radiation transport calculations were performed with the DORT two-dimensional discrete ordinates radiation transport code using the 39-neutron-group/44-gamma-ray-group ANSL-V cross-section library, an S{sub 6} quadrature, and a P{sub 1} Legendre polynomial expansion of the cross sections to determine the fission neutron source distribution in the core fuel elements. These calculations were limited to neutron groups only. The final radiation transport calculations, also performed with DORT using the 39-neutron-group/44-gamma-ray-group ANSL-V cross-section library, an S{sub l0} quadrature, and a P{sub 3} Legendre polynomial expansion of the cross sections, produced neutron and gamma-ray fluxes over the full extent of the geometry model. Responses (or activities) at various locations in the model were then obtained by folding the appropriate response functions with the fluxes at those locations. Some comparisons were made with VENTURE-calculated (diffusion theory) 20-group neutron fluxes that were summed into four broad groups. Tne results were in reasonably good agreement when the effects of photoneutrons were not included, thus verifying the physics model upon which the shielding model was based. Photoneutrons increased the fast-neutron flux levels deep within the D{sub 2}0 several orders of magnitude. Results are presented as tables of activity values for selected radial and axial traverses, plots of the radial and axial traverse data, and activity contours superimposed on the calculational geometry model.

  4. Investigation of a superthermal ultracold neutron source based on a solid deuterium converter for the TRIGA Mainz reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lauer, Thorsten

    2010-12-22

    Research in fundamental physics with the free neutron is one of the key tools for testing the Standard Model at low energies. Most prominent goals in this field are the search for a neutron electric dipole moment (EDM) and the measurement of the neutron lifetime. Significant improvements of the experimental performance using ultracold neutrons (UCN) require reduction of both systematic and statistical errors.The development and construction of new UCN sources based on the superthermal concept is therefore an important step for the success of future fundamental physics with ultracold neutrons. Significant enhancement of today available UCN densities strongly correlates with an efficient use of an UCN converter material. The UCN converter here is to be understood as a medium which reduces the velocity of cold neutrons (CN, velocity of about 600 m/s) to the velocity of UCN (velocity of about 6 m/s).Several big research centers around the world are presently planning or constructing new superthermal UCN sources, which are mainly based on the use of either solid deuterium or superfluid helium as UCN converter.Thanks to the idea of Yu.Pokotilovsky, there exists the opportunity to build competitive UCN sources also at small research reactors of the TRIGA type. Of course these smaller facilities don't promise high UCN densities of several 1000 UCN/cm{sup 3}, but they are able to provide densities around 100 UCN/cm{sup 3} for experiments.In the context of this thesis, it was possible to demonstrate succesfully the feasibility of a superthermal UCN source at the tangential beamport C of the research reactor TRIGA Mainz. Based on a prototype for the future UCN source at the Forschungs-Neutronenquelle Heinz Maier-Leibnitz (FRMII) in Munich, which was planned and built in collaboration with the Technical University of Munich, further investigations and improvements were done and are presented in this thesis. In parallel, a second UCN source for the radial beamport D was

  5. Investigation of a superthermal ultracold neutron source based on a solid deuterium converter for the TRIGA Mainz reactor

    International Nuclear Information System (INIS)

    Research in fundamental physics with the free neutron is one of the key tools for testing the Standard Model at low energies. Most prominent goals in this field are the search for a neutron electric dipole moment (EDM) and the measurement of the neutron lifetime. Significant improvements of the experimental performance using ultracold neutrons (UCN) require reduction of both systematic and statistical errors.The development and construction of new UCN sources based on the superthermal concept is therefore an important step for the success of future fundamental physics with ultracold neutrons. Significant enhancement of today available UCN densities strongly correlates with an efficient use of an UCN converter material. The UCN converter here is to be understood as a medium which reduces the velocity of cold neutrons (CN, velocity of about 600 m/s) to the velocity of UCN (velocity of about 6 m/s).Several big research centers around the world are presently planning or constructing new superthermal UCN sources, which are mainly based on the use of either solid deuterium or superfluid helium as UCN converter.Thanks to the idea of Yu.Pokotilovsky, there exists the opportunity to build competitive UCN sources also at small research reactors of the TRIGA type. Of course these smaller facilities don't promise high UCN densities of several 1000 UCN/cm3, but they are able to provide densities around 100 UCN/cm3 for experiments.In the context of this thesis, it was possible to demonstrate succesfully the feasibility of a superthermal UCN source at the tangential beamport C of the research reactor TRIGA Mainz. Based on a prototype for the future UCN source at the Forschungs-Neutronenquelle Heinz Maier-Leibnitz (FRMII) in Munich, which was planned and built in collaboration with the Technical University of Munich, further investigations and improvements were done and are presented in this thesis. In parallel, a second UCN source for the radial beamport D was designed and built

  6. The proposed spallation neutron source and modernized reactor as possible sites for a low temperature irradiation facility in Germany*1

    Science.gov (United States)

    Böning, K.; Gläser, W.; Golub, R.; Meier, J.

    1982-07-01

    A feasibility study for a Spallation Neutron Source (SNQ) in Germany was completed in June 1981. In this project an intensity-modulated LINAC (100 pps) would provide a proton beam of energy 1100 MeV and time-average current Īp = 5 mA . Spallation neutrons are produced in the lead material of a rotating target wheel and moderated in a hybrid arrangement consisting of both a small H 2O volume and a large D 2O tank. Here the maximum values of the peak and time-average thermal fluxes are ̂gf th ≈ 1.3 × 10 16 cm -2 s -1 and ¯gf th ≈ 6.5 × 10 14 cm -2 s -1, respectively. A low temperature irradiation facility (LTIF) has been proposed to allow irradiations in the temperature range of 4.5 to 450 K with either thermal neutrons ( ¯gf th ≈ 1 × 10 14 cm -2 s -1) or fast neutrons ( ¯gf f ≈ 2 × 10 13 cm -2 s -1). The advantages and disadvantages of having this LTIF at the SNQ are discussed with respect to the alternative of installing it at a fission reactor. Finally, the example of a possible modernization and upgrading of the Munich research reactor FRM is used to discuss the performance of such a reactor and the concept of a LTIF in this case, and to point out the complementarity of an optimized SNQ (high- ̂gf applications) and such a modernized reactor (high- ¯gf applications).

  7. Spallation Neutron Source (SNS)

    Data.gov (United States)

    Federal Laboratory Consortium — The SNS at Oak Ridge National Laboratory is a next-generation spallation neutron source for neutron scattering that is currently the most powerful neutron source in...

  8. Experimental characterization of the Advanced Liquid Hydrogen Cold Neutron Source spectrum of the NBSR reactor at the NIST Center for Neutron Research

    Science.gov (United States)

    Cook, J. C.; Barker, J. G.; Rowe, J. M.; Williams, R. E.; Gagnon, C.; Lindstrom, R. M.; Ibberson, R. M.; Neumann, D. A.

    2015-08-01

    The recent expansion of the National Institute of Standards and Technology (NIST) Center for Neutron Research facility has offered a rare opportunity to perform an accurate measurement of the cold neutron spectrum at the exit of a newly-installed neutron guide. Using a combination of a neutron time-of-flight measurement, a gold foil activation measurement, and Monte Carlo simulation of the neutron guide transmission, we obtain the most reliable experimental characterization of the Advanced Liquid Hydrogen Cold Neutron Source brightness to date. Time-of-flight measurements were performed at three distinct fuel burnup intervals, including one immediately following reactor startup. Prior to the latter measurement, the hydrogen was maintained in a liquefied state for an extended period in an attempt to observe an initial radiation-induced increase of the ortho (o)-hydrogen fraction. Since para (p)-hydrogen has a small scattering cross-section for neutron energies below 15 meV (neutron wavelengths greater than about 2.3 Å), changes in the o- p hydrogen ratio and in the void distribution in the boiling hydrogen influence the spectral distribution. The nature of such changes is simulated with a continuous-energy, Monte Carlo radiation-transport code using 20 K o and p hydrogen scattering kernels and an estimated hydrogen density distribution derived from an analysis of localized heat loads. A comparison of the transport calculations with the mean brightness function resulting from the three measurements suggests an overall o- p ratio of about 17.5(±1) % o- 82.5% p for neutron energies<15 meV, a significantly lower ortho concentration than previously assumed.

  9. Advanced neutron source reactor thermal-hydraulic test loop facility description

    Energy Technology Data Exchange (ETDEWEB)

    Felde, D.K.; Farquharson, G.; Hardy, J.H.; King, J.F.; McFee, M.T.; Montgomery, B.H.; Pawel, R.E.; Power, B.H.; Shourbaji, A.A.; Siman-Tov, M.; Wood, R.J.; Yoder, G.L.

    1994-02-01

    The Thermal-Hydraulic Test Loop (THTL) is a facility for experiments constructed to support the development of the Advanced Neutron Source Reactor (ANSR) at Oak Ridge National Laboratory. The ANSR is both cooled and moderated by heavy water and uses uranium silicide fuel. The core is composed of two coaxial fuel-element annuli, each of different diameter. There are 684 parallel aluminum-clad fuel plates (252 in the inner-lower core and 432 in the outer-upper core) arranged in an involute geometry that effectively creates an array of thin rectangular flow channels. Both the fuel plates and the coolant channels are 1.27 mm thick, with a span of 87 mm (lower core), 70 mm (upper core), and 507-mm heated length. The coolant flows vertically upwards at a mass flux of 27 Mg/m{sup 2}s (inlet velocity of 25 m/s) with an inlet temperature of 45{degrees}C and inlet pressure of 3.2 MPa. The average and peak heat fluxes are approximately 6 and 12 MW/m{sup 2}, respectively. The availability of experimental data for both flow excursion (FE) and true critical heat flux (CHF) at the conditions applicable to the ANSR is very limited. The THTL was designed and built to simulate a full-length coolant subchannel of the core, allowing experimental determination of thermal limits under the expected ANSR thermal-hydraulic conditions. For these experimental studies, the involute-shaped fuel plates of the ANSR core with the narrow 1.27-mm flow gap are represented by a narrow rectangular channel. Tests in the THTL will provide both single- and two-phase thermal-hydraulic information. The specific phenomena that are to be examined are (1) single-phase heat-transfer coefficients and friction factors, (2) the point of incipient boiling, (3) nucleate boiling heat-transfer coefficients, (4) two-phase pressure-drop characteristics in the nucleate boiling regime, (5) flow instability limits, and (6) CHF limits.

  10. Analysis of the multiplication of D-T neutron sources in blanket materials of fusion reactors

    International Nuclear Information System (INIS)

    A D-T neutron source is amplified when emitted into a body of material with appreciable (n,2n), (n,3n) or (n,f) cross sections. This amplification is described by a simple theory, approximating the strict integral transport description of the process. The distribution of neutrons in energy, from 14 MeV down to the (n,2n) threshold, is approximated by effective one-group cross sections for amplifiers of high and medium mass numbers; two-group cross sections are needed for Be. The spatial character of the multiplication is described by average collision probabilities for non-flat collision sources. The probabilities are approximated for spherical shell geometry with a small number of geometrical parameters. The theory enables a very accurate determination of sigmasub(n,2n) + 2sigmasub(n,3n) + (vsub(f)-1)sigmasub(f) at the source energy from measurements of total multiplications. If total leakages above the (n,2n) threshold are also measured, then the hardness of the secondary neutron spectra can be estimated. The accuracy of the approximate theory was ascertained by energy-space detailed transport comparison calculations for Be, Cu, Zr, Fe, Pb and U238. (orig.)

  11. Neutronics calculation of a reactor cold neutron source%反应堆冷中子源中子物理学计算

    Institute of Scientific and Technical Information of China (English)

    胡春明; 余朝举; 童剑飞

    2011-01-01

    用MCNP软件计算反应堆冷中子源,慢化剂室内平均中子注量率为6.69× 1013/cm-2.s-1,波长为0.4 nm和0.6 nm的冷中子增益因子~16和32.冷源慢化剂中正仲氢比例对输出的冷中子能谱有较大影响,而在3K范围内慢化剂温度变化对冷中子能谱的影响很小.计算结果表明,冷中子源性能达到基本设计要求.%The construction of a reactor cold neutron source (CNS) will be completed in the near future. To evaluate performance of the CNS, a neutronics calculation using MCNP4C code has been carried out. The results show that the average neutron flux in the moderator is 6.69×1013/cm2·s, and the cold neutron gain factors corresponding to 4-(A) and 6-(A) wavelengths are 16 and 32, respectively. The results also indicate that different ratios of ortho-H2/para-H2 have an obvious impact on cold neutron spectrum in the moderator, but within 3 K of the moderator temperature changes, the spectrum varies slightly.

  12. BR2 reactor neutron beams

    International Nuclear Information System (INIS)

    The use of reactor neutron beams is becoming increasingly more widespread for the study of some properties of condensed matter. It is mainly due to the unique properties of the ''thermal'' neutrons as regards wavelength, energy, magnetic moment and overall favorable ratio of scattering to absorption cross-sections. Besides these fundamental reasons, the impetus for using neutrons is also due to the existence of powerful research reactors (such as BR2) built mainly for nuclear engineering programs, but where a number of intense neutron beams are available at marginal cost. A brief introduction to the production of suitable neutron beams from a reactor is given. (author)

  13. Project and supply agreement. The text of the agreement of 14 October 1994 among the International Atomic Energy Agency and the Governments of the Republic of Ghana and the People's Republic of China concerning the transfer of a miniature neutron research reactor and enriched uranium

    International Nuclear Information System (INIS)

    The text of the Project and Supply Agreement, which was approved by the Agency's Board of Governors on 5 December 1991, among the Agency and the Governments of the Republic of Ghana and the People's Republic of China concerning the transfer of a miniature neutron research reactor and enriched uranium is reproduced for the information of all Members. The agreement entered into force on 14 October 1994, pursuant to Article XIII

  14. Project and supply agreement. The text of the agreement of 29 August 1996 among the International Atomic Energy Agency and the governments of the Republic of Nigeria and the People's Republic of China concerning the transfer of a miniature neutron research reactor and enriched uranium

    International Nuclear Information System (INIS)

    The document reproduces the text of the Project and Supply Agreement, which was approved by the Board of Governors on 19 March 1996, among the IAEA and the Governments of the Republic of Nigeria and the People's Republic of China concerning the transfer of a 30 kw miniature neutron research reactor and approximately 1000 grams of uranium enriched to approximately 90% by weight in the isotope uranium-235

  15. Production and Supplies of 99Mo: Lessons Learnt and New Options within Research Reactors and Neutron Sources Community

    International Nuclear Information System (INIS)

    During the past few years, the research reactor (RR) topic has occupied the centre stage being the major factor in the crisis faced world over in the supplies of medical isotopes, molybdenum-99 in particular. It is therefore an important aspect for discussion at the quadrennial international conference on research reactors organised by the IAEA. The November 2011 IAEA conference at Rabat, Morocco comes at a time when the international availability of 99Mo has fairly stabilised following the excellent technological efforts in terms of repairs done in the two large reactors, in Canada (NRU) and The Netherlands (HFR), serving the bulk of 99Mo users. The author, who had led and coordinated the IAEA activities in addressing the various issues and extending support to international efforts and initiatives during the period until March 2011, shares in this article his professional analysis of the field of 99Mo production, lessons and experience from the crisis as well as the aspects to be addressed to securing sustainable supplies of both 99Mo and 99mTc in future. In line with the suggestion of the International Programme Committee of the IAEA Conference, the scope of coverage is confined to sourcing 99Mo and 99mTc from RR and other neutron sources, while accelerator-based options are not included in this article. (author)

  16. Pulsed neutron sources at Dubna

    International Nuclear Information System (INIS)

    In 1960 the first world repetitively pulsed reactor IBR was put into operation. It was the beginning of the story how fission based pulsed neutron sources at Dubna have survived. The engineers involved have experienced many successes and failures in the course of new sources upgrading to finally come to possess the world's brightest neutron source - IBR-2. The details are being reviewed through the paper. The fission based pulsed neutron sources did not reach their final state as yet- the conceptual views of IBR prospects are being discussed with the goal to double the thermal neutron peak flux (up to 2x1016) and to enhance the cold neutron flux by 10 times (with the present one being as high that of the ISIS cold moderator). (author)

  17. Results of neutron-propagation experiments performed in iron-sodium mixtures using the source reactor HARMONIE and TAPIRO

    International Nuclear Information System (INIS)

    The results of neutron-propagation experiments performed in iron-sodium mixtures, using the source reactor HARMONIE and TAPIRO are presented: with HARMONIE: five media were studied: pure sodium and pure iron, and mixtures of iron and sodium (30 V/o, 50 V/o, 70 V/o); with TAPIRO the presently available results concern a pure sodium medium; two types of results are given: first, the raw results; then, the correction factors for heterogeneity and leakage to convert the raw results into the ideal homogeneous and spherical model used in the calculations. Finally, a comparison is made between results obtained with HARMONIE and TAPIRO for the pure sodium; the aim of this comparison is to verify that the differences in geometry and source spectra are properly accounted for the calculations. (author)

  18. Miniaturizing the miniature: Liquid droplets as miniscule reactors

    Directory of Open Access Journals (Sweden)

    Pratap R Patnaik

    2012-08-01

    Full Text Available Even though microreactors are performing an increasing number ofchemical and biological reactions better than conventionalmacroreactors, their weaknesses are being observed and overcome.A recent method is to conduct reactions inside liquid microdroplets.Each droplet contains all the reagents and catalysts required and theconditions are controlled to promote the most favorable processes.By populating a microreactor with millions of microdroplets, eachperforming as a complete reactor or sub-reactor, it is possible todistribute different reactions at different rates. Then the dropletsmay be fused or fissioned as required so that complementaryprocesses are brought into contact with one another at theappropriate points in time. The present article provides an overviewof the width and potential of this new and exciting technologyacross chemical and biological applications.

  19. Georgia Tech Studies of Sub-Critical Advanced Burner Reactors with a D-T Fusion Tokamak Neutron Source for the Transmutation of Spent Nuclear Fuel

    Science.gov (United States)

    Stacey, W. M.

    2009-09-01

    The possibility that a tokamak D-T fusion neutron source, based on ITER physics and technology, could be used to drive sub-critical, fast-spectrum nuclear reactors fueled with the transuranics (TRU) in spent nuclear fuel discharged from conventional nuclear reactors has been investigated at Georgia Tech in a series of studies which are summarized in this paper. It is found that sub-critical operation of such fast transmutation reactors is advantageous in allowing longer fuel residence time, hence greater TRU burnup between fuel reprocessing stages, and in allowing higher TRU loading without compromising safety, relative to what could be achieved in a similar critical transmutation reactor. The required plasma and fusion technology operating parameter range of the fusion neutron source is generally within the anticipated operational range of ITER. The implications of these results for fusion development policy, if they hold up under more extensive and detailed analysis, is that a D-T fusion tokamak neutron source for a sub-critical transmutation reactor, built on the basis of the ITER operating experience, could possibly be a logical next step after ITER on the path to fusion electrical power reactors. At the same time, such an application would allow fusion to contribute to meeting the nation's energy needs at an earlier stage by helping to close the fission reactor nuclear fuel cycle.

  20. Properties of neutron sources

    International Nuclear Information System (INIS)

    The Conference presentations were divided into sessions devoted to the following topics: white neutron sources, primarily pulsed (6 papers); fast neutron fields (5 papers); Californium-252 prompt fission neutron spectra (14 papers); monoenergetic sources and filtered beams (11 papers); 14 MeV neutron sources (10 papers); selected special application (one paper); and a general interest session (4 papers). Individual abstracts were prepared separately for the papers

  1. Investigation of a superthermal ultracold neutron source based on a solid deuterium converter for the TRIGA Mainz reactor

    OpenAIRE

    Lauer, Thorsten

    2009-01-01

    Research in fundamental physics with the free neutron is one of the key tools for testing the Standard Model at low energies. Most prominent goals in this field are the search for a neutron electric dipole moment (EDM) and the measurement of the neutron lifetime. Significant improvements of the experimental performance using ultracold neutrons (UCN) require reduction of both systematic and statistical errors.rnThe development and construction of new UCN sources based on the superthermal conce...

  2. Advanced Neutron Source (ANS) Project progress report

    International Nuclear Information System (INIS)

    This report discusses the following topics on the advanced neutron source: quality assurance (QA) program; reactor core development; fuel element specification; corrosion loop tests and analyses; thermal-hydraulic loop tests; reactor control concepts; critical and subcritical experiments; material data, structural tests, and analysis; cold source development; beam tube, guide, and instrument development; hot source development; neutron transport and shielding; I ampersand C research and development; facility concepts; design; and safety

  3. Advanced Neutron Source (ANS) Project progress report

    Energy Technology Data Exchange (ETDEWEB)

    McBee, M.R.; Chance, C.M. (eds.) (Oak Ridge National Lab., TN (USA)); Selby, D.L.; Harrington, R.M.; Peretz, F.J. (Oak Ridge National Lab., TN (USA))

    1990-04-01

    This report discusses the following topics on the advanced neutron source: quality assurance (QA) program; reactor core development; fuel element specification; corrosion loop tests and analyses; thermal-hydraulic loop tests; reactor control concepts; critical and subcritical experiments; material data, structural tests, and analysis; cold source development; beam tube, guide, and instrument development; hot source development; neutron transport and shielding; I C research and development; facility concepts; design; and safety.

  4. Different spectra with the same neutron source

    International Nuclear Information System (INIS)

    Using as source term the spectrum of a 239Pu-Be source several neutron spectra have been calculated using Monte Carlo methods. The source term was located in the centre of spherical moderators made of light water, heavy water and polyethylene of different diameters. Also a 239Pu-Be source was used to measure its neutron spectrum, bare and moderated by water. The neutron spectra were measured at 100 cm with a Bonner spheres spectrometer. Monte Carlo calculations were used to calculate the neutron spectra of bare and water-moderated spectra that were compared with those measured with the spectrometer. Resulting spectra are similar to those found in power plants with PWR, BWR and Candu nuclear reactors. Beside the spectra the dosimetric features were determined. Using moderators and a single neutron source can be produced neutron spectra alike those found in workplaces, this neutron fields can be utilized to calibrate neutron dosimeters and area monitors. (Author)

  5. Miniaturized radioisotope solid state power sources

    Science.gov (United States)

    Fleurial, J.-P.; Snyder, G. J.; Patel, J.; Herman, J. A.; Caillat, T.; Nesmith, B.; Kolawa, E. A.

    2000-01-01

    Electrical power requirements for the next generation of deep space missions cover a wide range from the kilowatt to the milliwatt. Several of these missions call for the development of compact, low weight, long life, rugged power sources capable of delivering a few milliwatts up to a couple of watts while operating in harsh environments. Advanced solid state thermoelectric microdevices combined with radioisotope heat sources and energy storage devices such as capacitors are ideally suited for these applications. By making use of macroscopic film technology, microgenrators operating across relatively small temperature differences can be conceptualized for a variety of high heat flux or low heat flux heat source configurations. Moreover, by shrinking the size of the thermoelements and increasing their number to several thousands in a single structure, these devices can generate high voltages even at low power outputs that are more compatible with electronic components. Because the miniaturization of state-of-the-art thermoelectric module technology based on Bi2Te3 alloys is limited due to mechanical and manufacturing constraints, we are developing novel microdevices using integrated-circuit type fabrication processes, electrochemical deposition techniques and high thermal conductivity substrate materials. One power source concept is based on several thermoelectric microgenerator modules that are tightly integrated with a 1.1W Radioisotope Heater Unit. Such a system could deliver up to 50mW of electrical power in a small lightweight package of approximately 50 to 60g and 30cm3. An even higher degree of miniaturization and high specific power values (mW/mm3) can be obtained when considering the potential use of radioisotope materials for an alpha-voltaic or a hybrid thermoelectric/alpha-voltaic power source. Some of the technical challenges associated with these concepts are discussed in this paper. .

  6. Study on void fraction distribution in the moderator cell of Cold Neutron Source systems in China Advanced Research Reactor

    Science.gov (United States)

    Li, Liangxing; Li, Huixiong; Hu, Jinfeng; Bi, Qincheng; Chen, Tingkuan

    2007-04-01

    A physical model is developed for analyzing and evaluating the void fraction profiles in the moderator cell of the Cold Neutron Source (CNS) of the China Advanced Research Reactor (CARR), which is now constructing in the China Institute of Atomic Energy (CIAE). The results derived from the model are compared with the related experimental data and its propriety is verified. The model is then used to explore the influence of various factors, including the diameter of boiling vapor bubbles, liquid density, liquid viscosity and the total heating power acted on the moderator cell, on the void fraction profiles in the cell. The results calculated with the present model indicate that the void fraction in the moderator cell increases linearly with heating power, and increases with the liquid viscosity, but decreases as the size of bubbles increases, and increases linearly with heating power. For the case where hydrogen is being used as a moderator, calculation results show that the void fraction in the moderator cell may be less than 30%, which is the maximum void fraction permitted from the nuclear physics point of view. The model and the calculation results will help to obtain insight of the mechanism that controls the void fraction distribution in the moderator cell, and provide theoretical supports for the moderator cell design.

  7. Neutronics of nuclear power reactors

    International Nuclear Information System (INIS)

    This review, prepared on the occasion of 25th ETAN Conference describes the research activities in the field of neutronics which started in 1947. A number of researchers in Yugoslav Institutes was engaged in development of neutronics theory and calculation methods related to power reactors since 1960. To illustrate the activities of Yugoslav authors, this review contains the list of the most important relevant papers published in international journals

  8. Pulsed spallation Neutron Sources

    Energy Technology Data Exchange (ETDEWEB)

    Carpenter, J.M. [Argonne National Lab., IL (United States)

    1994-12-31

    This paper reviews the early history of pulsed spallation neutron source development at Argonne and provides an overview of existing sources world wide. A number of proposals for machines more powerful than currently exist are under development, which are briefly described. The author reviews the status of the Intense Pulsed Neutron Source, its instrumentation, and its user program, and provides a few examples of applications in fundamental condensed matter physics, materials science and technology.

  9. Advanced Neutron Source (ANS) Project

    International Nuclear Information System (INIS)

    This report covers the progress made in 1993 in the following sections: (1) project management; (2) research and development; (3) design and (4) safety. The section on research and development covers the following: (1) reactor core development; (2) fuel development; (3) corrosion loop tests and analysis; (4) thermal-hydraulic loop tests; (5) reactor control and shutdown concepts; (6) critical and subcritical experiments; (7) material data, structure tests, and analysis; (8) cold source development; (9) beam tube, guide, and instrument development; (10) neutron transport and shielding; (11) I and C research and development; and (12) facility concepts

  10. Intense pulsed neutron sources

    Energy Technology Data Exchange (ETDEWEB)

    Kustom, R.L.

    1981-01-01

    Accelerator requirements for pulsed spallation neutron sources are stated. Brief descriptions of the Argonne IPNS-I, the Japanese KENS, Los Alamos Scientific Laboratory WNR/PSR, the Rutherford Laboratory SNS, and the West German SNQ facilities are presented.

  11. Detailed heat load calculations at the beginning, middle, and end of cycle for the conceptual design of the Advanced Neutron Source Reactor

    International Nuclear Information System (INIS)

    The Advanced Neutron Source (ANS) is a world-class research reactor and experimental center for neutron research, presently being designed at the Oak Ridge National Laboratory (ORNL). The reactor consists of a 330-MW(f) highly enriched uranium core, which is cooled, moderated, and reflected with heavy water. When completed, it will be the preeminent ultrahigh neutron flux reactor in the world, with facilities for research programs in biology, materials science, chemistry, fundamental and nuclear physics, and analytical chemistry. Irradiation facilities are provided for a variety of isotope production capabilities, as well as materials irradiation. The ANS reactor design, at the time of this report, has completed the conceptual design phase and entered the advanced conceptual design phase. This report is part of an effort to fully document the analysis methods and results for the conceptual design. It details the methods used to perform heat load calculations on the ANS reactor design, describes the model used, and gives the resulting heat loads in all components of the reactor, in both a differential (by segment) and integral (by component) fashion. These heat load data are provided at three times within the ANS fuel cycle - at beginning (0 days), middle (8.5 days), and end (17 days) of cycle. The remainder of the report is dedicated to this description. In Chapter 2, some necessary background on the reactor design is provided. Chapters 3 and 4 give details of the depletion methods used and revisions to previous MCNP models. Chapter 5 analyzes the results of these calculations, and Chapter 6 provides a summary and conclusions

  12. International workshop on cold neutron sources

    Energy Technology Data Exchange (ETDEWEB)

    Russell, G.J.; West, C.D. (comps.) (Los Alamos National Lab., NM (United States))

    1991-08-01

    The first meeting devoted to cold neutron sources was held at the Los Alamos National Laboratory on March 5--8, 1990. Cosponsored by Los Alamos and Oak Ridge National Laboratories, the meeting was organized as an International Workshop on Cold Neutron Sources and brought together experts in the field of cold-neutron-source design for reactors and spallation sources. Eighty-four people from seven countries attended. Because the meeting was the first of its kind in over forty years, much time was spent acquainting participants with past and planned activities at reactor and spallation facilities worldwide. As a result, the meeting had more of a conference flavor than one of a workshop. The general topics covered at the workshop included: Criteria for cold source design; neutronic predictions and performance; energy deposition and removal; engineering design, fabrication, and operation; material properties; radiation damage; instrumentation; safety; existing cold sources; and future cold sources.

  13. International workshop on cold neutron sources

    International Nuclear Information System (INIS)

    The first meeting devoted to cold neutron sources was held at the Los Alamos National Laboratory on March 5--8, 1990. Cosponsored by Los Alamos and Oak Ridge National Laboratories, the meeting was organized as an International Workshop on Cold Neutron Sources and brought together experts in the field of cold-neutron-source design for reactors and spallation sources. Eighty-four people from seven countries attended. Because the meeting was the first of its kind in over forty years, much time was spent acquainting participants with past and planned activities at reactor and spallation facilities worldwide. As a result, the meeting had more of a conference flavor than one of a workshop. The general topics covered at the workshop included: Criteria for cold source design; neutronic predictions and performance; energy deposition and removal; engineering design, fabrication, and operation; material properties; radiation damage; instrumentation; safety; existing cold sources; and future cold sources

  14. Analytical solution of neutron transport equation in an annular reactor with a rotating pulsed source; Resolucao analitica da equacao de transporte de neutrons em um reator anelar com fonte pulsada rotativa

    Energy Technology Data Exchange (ETDEWEB)

    Teixeira, Paulo Cleber Mendonca

    2002-12-01

    In this study, an analytical solution of the neutron transport equation in an annular reactor is presented with a short and rotating neutron source of the type S(x) {delta} (x- Vt), where V is the speed of annular pulsed reactor. The study is an extension of a previous study by Williams [12] carried out with a pulsed source of the type S(x) {delta} (t). In the new concept of annular pulsed reactor designed to produce continuous high flux, the core consists of a subcritical annular geometry pulsed by a rotating modulator, producing local super prompt critical condition, thereby giving origin to a rotating neutron pulse. An analytical solution is obtained by opening up of the annular geometry and applying one energy group transport theory in one dimension using applied mathematical techniques of Laplace transform and Complex Variables. The general solution for the flux consists of a fundamental mode, a finite number of harmonics and a transient integral. A condition which limits the number of harmonics depending upon the circumference of the annular geometry has been obtained. Inverse Laplace transform technique is used to analyse instability condition in annular reactor core. A regenerator parameter in conjunction with perimeter of the ring and nuclear properties is used to obtain stable and unstable harmonics and to verify if these exist. It is found that the solution does not present instability in the conditions stated in the new concept of annular pulsed reactor. (author)

  15. Advanced spallation neutron sources for condensed matter research

    International Nuclear Information System (INIS)

    Advanced spallation neutron sources afford significant advantages over existing high flux reactors. The effective flux is much greater than that currently available with reactor sources. A ten-fold increase in neutron flux will be a major benefit to a wide range of condensed matter studies, and it will realise important experiments that are marginal at reactor sources. Moreover, the high intensity of epithermal neutrons open new vistas in studies of electronic states and molecular vibrations. (author)

  16. Neutron measurements at nuclear power reactors [55

    CERN Document Server

    Scherpelz, R I

    2002-01-01

    Staff from the Pacific Northwest National Laboratory (operated by Battelle Memorial Institute), have performed neutron measurements at a number of commercial nuclear power plants in the United States. Neutron radiation fields at light water reactor (LWR) power plants are typically characterized by low-energy distributions due to the presence of large amounts of scattering material such as water and concrete. These low-energy distributions make it difficult to accurately monitor personnel exposures, since most survey meters and dosimeters are calibrated to higher-energy fields such as those produced by bare or D sub 2 O-moderated sup 2 sup 5 sup 2 Cf sources. Commercial plants typically use thermoluminescent dosimeters in an albedo configuration for personnel dosimetry and survey meters based on a thermal-neutron detector inside a cylindrical or spherical moderator for dose rate assessment, so their methods of routine monitoring are highly dependent on the energy of the neutron fields. Battelle has participate...

  17. Status of spallation neutron source

    Energy Technology Data Exchange (ETDEWEB)

    Oyama, Yukio [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1997-03-01

    Existing and planned facilities using proton accelerator driven spallation neutron source are reviewed. These include new project of neutron science proposed from Japan Atomic Energy Research Institute. The present status of facility requirement and accelerator technology leads us to new era of neutron science such as neutron scattering research and nuclear transmutation study using very intense neutron source. (author)

  18. Plasma focus neutron source

    International Nuclear Information System (INIS)

    A neutron source not permanently active is obtained from an electric discharge plasma focus (PF) device. A small PF device, a Mather model device, works in the limit of low energy, 100 to 200 J at charging voltage of 20 to 30 kV with a capacitor bank of 160 nF, and a characteristic inductance of 25 to 50 nH. A theoretical model leads us to estimate the optimum values of capacitance, inductance, initial charging voltage and electrode geometry. In this work is presented the design evolution and construction of a first PF neutron source prototype, preliminary measures of current, voltage and temporal evolution of the current with the end of have an electric characterization. This parameters must be optimized with the objective of geeting an emission of 104 to 105 neutrons per pulse when Deuterium is used like filled gas (C.W)

  19. Low dimensional neutron moderators for enhanced source brightness

    DEFF Research Database (Denmark)

    Mezei, Ferenc; Zanini, Luca; Takibayev, Alan;

    2014-01-01

    for cold neutrons. This model leads to the conclusions that the optimal shape for high brightness para-hydrogen neutron moderators is the quasi 1-dimensional tube and these low dimensional moderators can also deliver much enhanced cold neutron brightness in fission reactor neutron sources, compared...

  20. Dhruva reactor -- a high flux facility for neutron beam research

    International Nuclear Information System (INIS)

    Dhruva reactor, the highest flux thermal neutron source in India has been operating at full power of 100 MW over the past two years. Several advanced facilities like the cold source, guides, etc. are being installed for neutron beam research in condensed matter. A large number and variety of neutron spectrometers are operational. This paper deals with the basic advantages that one can derive from neutron scattering investigations and gives a brief description of the instruments that are developed and commissioned at Dhruva for neutron beam research. (author). 3 figs

  1. The University of Texas Cold Neutron Source

    Science.gov (United States)

    Ünlü, Kenan; Ríos-Martínez, Carlos; Wehring, Bernard W.

    1994-12-01

    A cold neutron source has been designed, constructed, and tested by the Nuclear Engineering Teaching Laboratory (NETL) at The University of Texas at Austin. The Texas Cold Neutron Source (TCNS) is located in one of the beam ports of the NETL 1-MW TRIGA Mark II research reactor. The main components of the TCNS are a cooled moderator, a heat pipe, a cryogenic refrigerator, and a neutron guide. 80 ml of mesitylene moderator are maintained at about 30 K in a chamber within the reactor graphite reflector by the heat pipe and cryogenic refrigerator. The heat pipe is a 3-m long aluminum tube that contains neon as the working fluid. The cold neutrons obtained from the moderator are transported by a curved 6-m long neutron guide. This neutron guide has a radius of curvature of 300 m, a 50 × 15 mm cross-section, 58Ni coating, and is separated into three channels. The TCNS will provide a low-background subthermal neutron beam for neutron capture and scattering research. After the installation of the external portion of the neutron guide, a neutron focusing system and a Prompt Gamma Activation Analysis facility will be set up at the TCNS.

  2. Miniature X-ray Source for Planetary Exploration Instruments Project

    Data.gov (United States)

    National Aeronautics and Space Administration — The objective of the proposed work is to develop a rugged, low power, passively cooled X-Ray source that is compatible with miniaturized XRD systems. The XRD...

  3. Neutron source for Neutron Capture Synovectomy

    International Nuclear Information System (INIS)

    Monte Carlo calculations were performed to obtain a thermal neutron field from a 239PuBe neutron source inside a cylindrical heterogeneous moderators for Neutron Capture Synovectomy. Studied moderators were light water and heavy water, graphite and heavy water, lucite and polyethylene and heavy water. The neutron spectrum of polyethylene and heavy water moderator was used to determine neutron spectra inside a knee model. In this model the elemental composition of synovium and synovial liquid was assumed like blood. Kerma factors for synovium and synovial liquid were calculated to compare with water Kerma factors, in this calculations the synovium was loaded with two different concentrations of Boron

  4. Cyclotron-based neutron source for BNCT

    Energy Technology Data Exchange (ETDEWEB)

    Mitsumoto, T.; Yajima, S.; Tsutsui, H.; Ogasawara, T.; Fujita, K. [Sumitomo Heavy Industries, Ltd (Japan); Tanaka, H.; Sakurai, Y.; Maruhashi, A. [Kyoto University Research Reactor Institute (Japan)

    2013-04-19

    Kyoto University Research Reactor Institute (KURRI) and Sumitomo Heavy Industries, Ltd. (SHI) have developed a cyclotron-based neutron source for Boron Neutron Capture Therapy (BNCT). It was installed at KURRI in Osaka prefecture. The neutron source consists of a proton cyclotron named HM-30, a beam transport system and an irradiation and treatment system. In the cyclotron, H- ions are accelerated and extracted as 30 MeV proton beams of 1 mA. The proton beams is transported to the neutron production target made by a beryllium plate. Emitted neutrons are moderated by lead, iron, aluminum and calcium fluoride. The aperture diameter of neutron collimator is in the range from 100 mm to 250 mm. The peak neutron flux in the water phantom is 1.8 Multiplication-Sign 109 neutrons/cm{sup 2}/sec at 20 mm from the surface at 1 mA proton beam. The neutron source have been stably operated for 3 years with 30 kW proton beam. Various pre-clinical tests including animal tests have been done by using the cyclotron-based neutron source with {sup 10}B-p-Borono-phenylalanine. Clinical trials of malignant brain tumors will be started in this year.

  5. Neutron scattering and spallation neutron sources

    International Nuclear Information System (INIS)

    Neutron scattering as a probe of microscopic structure and dynamics is a powerful tool for research in a wide variety of fields, and an accelerator-based spallation neutron source can supply high flux pulses for neutron scattering. The characteristics of neutron scattering, the principle and development of spallation neutron sources, and their advantages in multidisciplinary applications are summarized. In the proposed project of the Chinese Spallation Neutron Source the target station will consist of a piece-stacked tungsten target, a Be/Fe reflector and an Fe/heavy concrete bio-protected shelter. The pulsed neutron flux will be up to 2.4 x 1016 n/cm2/s under a nuclear power of 100 kW. Five neutron scattering instruments--a high flux powder diffractometer, a high resolution powder diffractometer, small angle diffractometer, multi-functional reflectometer and direct geometry inelastic spectrometer, will be constructed as the first step to cover most neutron scattering applications. (authors)

  6. Isotopic neutron sources for neutron activation analysis

    International Nuclear Information System (INIS)

    This User's Manual is an attempt to provide for teaching and training purposes, a series of well thought out demonstrative experiments in neutron activation analysis based on the utilization of an isotopic neutron source. In some cases, these ideas can be applied to solve practical analytical problems. 19 refs, figs and tabs

  7. Effect of cathode structure on neutron yield performance of a miniature plasma focus device

    International Nuclear Information System (INIS)

    In this Letter we report the effect of two different cathode structures - tubular and squirrel cage, on neutron output from a miniature plasma focus device. The squirrel cage cathode is typical of most DPF sources, with an outer, tubular envelope that serves as a vacuum housing, but does not carry current. The tubular cathode carries the return current and also serves as the vacuum envelope, thereby minimizing the size of the DPF head. The maximum average neutron yield of (1.82±0.52)x105 n/shot for the tubular cathode at 4 mbar was enhanced to (1.15±0.2)x106 n/shot with squirrel cage cathode at 6 mbar operation. These results are explained on the basis of a current sheath loading/mass choking effect. The penalty for using a non-transparent cathode negates the advantage of the smaller size of the DPF head.

  8. An Accelerator Neutron Source for BNCT

    Energy Technology Data Exchange (ETDEWEB)

    Blue, Thomas, E

    2006-03-14

    The overall goal of this project was to develop an accelerator-based neutron source (ABNS) for Boron Neutron Capture Therapy (BNCT). Specifically, our goals were to design, and confirm by measurement, a target assembly and a moderator assembly that would fulfill the design requirements of the ABNS. These design requirements were 1) that the neutron field quality be as good as the neutron field quality for the reactor-based neutron sources for BNCT, 2) that the patient treatment time be reasonable, 3) that the proton current required to treat patients in reasonable times be technologially achievable at reasonable cost with good reliability, and accelerator space requirements which can be met in a hospital, and finally 4) that the treatment be safe for the patients.

  9. An Accelerator Neutron Source for BNCT

    International Nuclear Information System (INIS)

    The overall goal of this project was to develop an accelerator-based neutron source (ABNS) for Boron Neutron Capture Therapy (BNCT). Specifically, our goals were to design, and confirm by measurement, a target assembly and a moderator assembly that would fulfill the design requirements of the ABNS. These design requirements were (1) that the neutron field quality be as good as the neutron field quality for the reactor-based neutron sources for BNCT, (2) that the patient treatment time be reasonable, (3) that the proton current required to treat patients in reasonable times be technologically achievable at reasonable cost with good reliability, and accelerator space requirements which can be met in a hospital, and finally (4) that the treatment be safe for the patients

  10. From reactors to long pulse sources

    International Nuclear Information System (INIS)

    We will show, that by using an adapted instrumentation concept, the performance of a continuous source can be emulated by one switch on in long pulses for only about 10% of the total time. This 10 fold gain in neutron economy opens up the way for building reactor like sources with an order of magnitude higher flux than the present technological limits. Linac accelerator driven spallation lends itself favorably for the realization of this kind of long pulse sources, which will be complementary to short pulse spallation sources, the same way continuous reactor sources are

  11. Neutron scattering instrumentation for biology at spallation neutron sources

    Energy Technology Data Exchange (ETDEWEB)

    Pynn, R. [Los Alamos National Laboratory, NM (United States)

    1994-12-31

    Conventional wisdom holds that since biological entities are large, they must be studied with cold neutrons, a domain in which reactor sources of neutrons are often supposed to be pre-eminent. In fact, the current generation of pulsed spallation neutron sources, such as LANSCE at Los Alamos and ISIS in the United Kingdom, has demonstrated a capability for small angle scattering (SANS) - a typical cold- neutron application - that was not anticipated five years ago. Although no one has yet built a Laue diffractometer at a pulsed spallation source, calculations show that such an instrument would provide an exceptional capability for protein crystallography at one of the existing high-power spoliation sources. Even more exciting is the prospect of installing such spectrometers either at a next-generation, short-pulse spallation source or at a long-pulse spallation source. A recent Los Alamos study has shown that a one-megawatt, short-pulse source, which is an order of magnitude more powerful than LANSCE, could be built with today`s technology. In Europe, a preconceptual design study for a five-megawatt source is under way. Although such short-pulse sources are likely to be the wave of the future, they may not be necessary for some applications - such as Laue diffraction - which can be performed very well at a long-pulse spoliation source. Recently, it has been argued by Mezei that a facility that combines a short-pulse spallation source similar to LANSCE, with a one-megawatt, long-pulse spallation source would provide a cost-effective solution to the global shortage of neutrons for research. The basis for this assertion as well as the performance of some existing neutron spectrometers at short-pulse sources will be examined in this presentation.

  12. A telescope for monitoring fast neutron sources

    International Nuclear Information System (INIS)

    In the framework of nuclear waste management, highly radiotoxic long-lived fission products and minor actinides are planned to be transmuted in a sub-critical reactor coupled with an intense external neutron source. The latter source would be created by a high-energy proton beam hitting a high atomic number target. Such a new system, termed an accelerator-driven system (ADS), requires on-line and robust reactivity monitoring. The ratio between the beam current delivered by the accelerator and the reactor power level, or core neutron flux, is the basis of one method which could give access to a core reactivity change. In order to test reactivity measurement technique, some experimental programs use 14-MeV neutrons originating from the interaction of a deuteron beam with a tritium target as an external neutron source. In this case, the target tritium consumption over time precludes use of the beam current for reactivity monitoring and the external neutron source intensity must be monitored directly. A range telescope has been developed for this purpose, consisting of the assembly of a hydrogenous neutron converter and three silicon stages where the recoiling protons are detected. In this article, the performances of such a telescope are presented and compared to Monte-Carlo simulations

  13. Neutron cooling and cold-neutron sources (1962); Refroidissement des neutrons et sources de neutrons froids (1962)

    Energy Technology Data Exchange (ETDEWEB)

    Jacrot, B. [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1962-07-01

    Intense cold-neutron sources are useful in studying solids by the inelastic scattering of neutrons. The paper presents a general survey covering the following aspects: a) theoretical considerations put forward by various authors regarding thermalization processes at very low temperatures; b) the experiments that have been carried out in numerous laboratories with a view to comparing the different moderators that can be used; c) the cold neutron sources that have actually been produced in reactors up to the present time, and the results obtained with them. (author) [French] Des sources intenses de neutrons froids sont utiles pour l'etude des solides par diffusion inelastique des neutrons. On presente une revue d'ensemble: a) des considerations theoriques faites par divers auteurs sur les processus de thermalisation a tres basse temperature; b) des experiences faites dans de nombreux laboratoires pour comparer les divers moderateurs possibles; c) des sources de neutrons froids effectivement realisees dans des piles a ce jour, et des resultats obtenus avec ces sources. (auteur)

  14. Optimization of He-II UCN source with spallation neutron source

    Energy Technology Data Exchange (ETDEWEB)

    Mishima, K.; Choi, E.; Yoshimura, M. [Osaka Univ., Ibaraki (Japan). Research Center for Nuclear Physics; Ooi, M.; Kiyanagi, Y. [Hokkaido Univ., Sapporo (Japan); Masuda, Y.; Muto, S. [High Energy Accelerator Research Organization, Tsukuba, Ibaraki (Japan); Tanaka, M. [Kobe Tokiwa Collage, Kobe, Hyogo (Japan)

    2001-03-01

    A spallation neutron source was designed for super thermal UCN production in He-II. The configuration of neutron production target, moderator and He-II bottle was optimized in order to obtain high neutron flux with low {gamma} heating in He-II. In the optimization the advantage of the spallation neutron source is used: The spallation neutron source has high n/{gamma} ratio and freedom in target moderator configuration in comparison with the reactor. As a result, a great improvement in UCN density is expected compared with the present most intense UCN source at the Grenoble reactor. (authors)

  15. On the Development of a Miniature Neutron Generator for the Brachytherapy Treatment of Cancer

    Science.gov (United States)

    Forman, L.

    2009-03-01

    Brachytherapy refers to application of an irradiation source within a tumor. 252Cf needles used in brachytherapy have been successfully applied to treatment of some of the most virulent cancers but it is doubtful that it will be widely used because of difficulty in dealing with unwanted dose (source cannot be turned off) and in adhering to stringent NRC regulations that have been exacerbated in our post 911 environment. We have been working on the development of a miniature neutron generator with the reaction target placed at the end of a needle (tube) for brachytherapy applications. Orifice geometries are most amenable, e.g. rectum and cervix, but interstitial use is possible with microsurgery. This paper dicusses the results of a 30 watt DD neutron generator SBU project that demonstrates that sufficient hydrogen isotope current can be delivered down a small diameter needle required for a DT neutron treatment device, and, will summarize the progress of building a commercial device pursued by the All Russian Institute for Automatics (VNIIA) supported by the DOE's Industrial Proliferation Prevention Program (IPP). It is known that most of the fast neutron (FN) beam cancer treatment facilities have been closed down. It appears that the major limitation in the use of FN beams has been damage to healthy tissue, which is relatively insensitive to photons, but this problem is alleviated by brachytherapy. Moreover, recent clinical results indicate that fast neutrons in the boost mode are most highly effective in treating large, hypoxic, and rapidly repopulating diseases. It appears that early boost application of FN may halt angiogenesis (development and repair of tumor vascular system) and shrink the tumor resulting in lower hypoxia. The boost brachytherapy application of a small, low cost neutron generator holds promise of significant contribution to the treatment of cancer.

  16. Mechanical behavior of AISI 304SS determined by miniature test methods after neutron irradiation to 28 dpa

    Energy Technology Data Exchange (ETDEWEB)

    Ellen M. Rabenberg; Brian J. Jaques; Bulent H. Sencer; Frank A. Garner; Paula D. Freyer; Taira Okita; Darryl P. Butt

    2014-05-01

    The mechanical properties of AISI 304 stainless steel irradiated for over a decade in the Experimental Breeder Reactor (EBR-II) were measured using miniature mechanical testing methods. The shear punch method was used to evaluate the shear strengths of the neutron-irradiated steel and a correlation factor was empirically determined to predict its tensile strength. The strength of the stainless steel slightly decreased with increasing irradiation temperature, and significantly increased with increasing dose until it saturated above approximately 5 dpa. Ferromagnetic measurements were used to observe and deduce the effects of the stress-induced austenite to martensite transformation as a result of shear punch testing.

  17. Miniature Incandescent Lamps as Fiber-Optic Light Sources

    Science.gov (United States)

    Tuma, Margaret; Collura, Joe; Helvajian, Henry; Pocha, Michael; Meyer, Glenn; McConaghy, Charles F.; Olsen, Barry L.

    2008-01-01

    Miniature incandescent lamps of a special type have been invented to satisfy a need for compact, rapid-response, rugged, broadband, power-efficient, fiber-optic-coupled light sources for diverse purposes that could include calibrating spectrometers, interrogating optical sensors, spot illumination, and spot heating.

  18. Options for neutron scattering instruments on long pulse neutron sources

    International Nuclear Information System (INIS)

    Instrumenttion on long pulse sources can be approached either by instruments from short pulse sources and hence using mainly inverted time of flight techniques or by adopting reactor type instruments and making use of the time dependence of the source flux to enhance their performance substantially. While the first approach requires more or less single use of a beam line by one instrument, the second one allows multiple use of neutron guides, as customary on reactors and hence can make much better use of the source with gains up to 100 for time of flight spectrometers. To a certain extent, the design parameters of the source depend on which of the two approaches is chosen. (author) 8 figs., 1 tab., 16 refs

  19. Neutron source strength monitors for ITER

    International Nuclear Information System (INIS)

    There are several goals for the neutron source strength monitor system for the International Thermonuclear Experimental Reactor (ITER). Desired is a stable, reliable, time-dependent neutron detection system which exhibits a wide dynamic range and broad energy response to incident neutrons while being insensitive to gamma rays and having low noise characteristics in a harsh reactor environment. This system should be able to absolutely calibrated in-situ using various neutron sources. An array of proportional counters of varying sensitivities is proposed along with the most promising possible locations. One proposed location is in the pre-shields of the neutron camera collimators which would allow an integrated design of neutron systems with good detector access. As part of an ongoing conceptual design for this system, the detector-specific issues of dynamic range, performance monitoring, and sensitivity will be presented. The location options of the array will be discussed and most importantly, the calibration issues associated with a heavily shielded vessel will be presented

  20. Neutron source strength monitors for ITER

    Energy Technology Data Exchange (ETDEWEB)

    Barnes, C.W. [Sandia National Labs., Albuquerque, NM (United States); Roquemore, A.L. [Princeton Univ., NJ (United States). Plasma Physics Lab.

    1996-05-07

    There are several goals for the neutron source strength monitor system for the International Thermonuclear Experimental Reactor (ITER). Desired is a stable, reliable, time-dependent neutron detection system which exhibits a wide dynamic range and broad energy response to incident neutrons while being insensitive to gamma rays and having low noise characteristics in a harsh reactor environment. This system should be able to absolutely calibrated in-situ using various neutron sources. An array of proportional counters of varying sensitivities is proposed along with the most promising possible locations. One proposed location is in the pre-shields of the neutron camera collimators which would allow an integrated design of neutron systems with good detector access. As part of an ongoing conceptual design for this system, the detector-specific issues of dynamic range, performance monitoring, and sensitivity will be presented. The location options of the array will be discussed and most importantly, the calibration issues associated with a heavily shielded vessel will be presented.

  1. Reference Neutron Radiographs of Nuclear Reactor Fuel

    OpenAIRE

    Domanus, Joseph Czeslaw

    1986-01-01

    Reference neutron radiographs of nuclear reactor fuel were produced by the Euraton Neutron Radiography Working Group and published in 1984 by the Reidel Publishing Company. In this collection a classification is given of the various neutron radiographic findings, that can occur in different parts of pelletized, annular and vibro-conpacted nuclear fuel pins. Those parts of the pins are shown where changes of appearance differ from those for the parts as fabricated. Also radiographs of those as...

  2. Shielding the LANSCE [Los Alamos Neutron Scattering Center] 800-MeV spallation neutron source

    International Nuclear Information System (INIS)

    Neutrons produced by medium-energy (800-MeV) proton reactions at the Los Alamos Neutron Scattering Center spallation neutron source cause a variety of difficult shield problems. We describe the general shielding questions encountered at such a spallation source, and contrast spallation and reactor source shielding issues using an infinite slab-shield composed of 100 cm of iron and 15 cm of borated polyethylene. The calculations show that (for an incident spallation spectrum characteristic of neutrons leaking at 90 degrees from a tungsten target) high-energy neutrons dominate the dose at the shield surface. Secondary low-energy neutrons (produced by high-energy neutron attenuation) and attendant gamma-rays add significantly to the dose. The primary low-energy neutrons produced directly at the tungsten source contribute negligibly to the dose, and behave similarly to neutrons with a fission spectrum distribution. 8 refs., 10 figs

  3. Neutronic of heterogenous gas cooled reactors

    International Nuclear Information System (INIS)

    At present, one of the main technical features of the advanced gas cooled reactor under development is its fuel element concept, which implies a neutronic homogeneous design, thus requiring higher enrichment compared with present commercial nuclear power plants.In this work a neutronic heterogeneous gas cooled reactor design is analyzed by studying the neutronic design of the Advanced Gas cooled Reactor (AGR), a low enrichment, gas cooled and graphite moderated nuclear power plant.A search of merit figures (some neutronic parameter, characteristic dimension, or a mixture of both) which are important and have been optimized during the reactor design stage is been done, to aim to comprise how a gas heterogeneous reactor is been design, given that semi-infinity arrangement criteria of rods in LWRs and clusters in HWRs can t be applied for a solid moderator and a gas refrigerator.The WIMS code for neutronic cell calculations is been utilized to model the AGR fuel cell and to calculate neutronic parameters such as the multiplication factor and the pick factor, as function of the fuel burnup.Also calculation is been done for various nucleus characteristic dimensions values (fuel pin radius, fuel channel pitch) and neutronic parameters (such as fuel enrichment), around the design established parameters values.A fuel cycle cost analysis is carried out according to the reactor in study, and the enrichment effect over it is been studied.Finally, a thermal stability analysis is been done, in subcritical condition and at power level, to study this reactor characteristic reactivity coefficients.Present results shows (considering the approximation used) a first set of neutronic design figures of merit consistent with the AGR design.

  4. International seminar on structural investigations on pulsed neutron sources. Proceedings

    International Nuclear Information System (INIS)

    The proceedings of the International seminar on structural investigations using pulsed neutron sources are presented. The seminar is dedicated to the memory of Dr. Yu.M. Ostanevich, a world acknowledged physicist. The problems of structural analysis using pulsed neutron source at the IBR-2 reactor are discussed

  5. The measurement of subcritical reactivity in nuclear reactors by use of a high frequency sine-wave modulated neutron source

    International Nuclear Information System (INIS)

    In this report the frequency response characteristics for phase and gain of the fundamental reactor mode of the zero power kinetics are given for various subcritical reactivities in a fast reactor and in a thermal reactor. Results, of a study on harmonic effects based on a small zero energy thermal reactor are presented which demonstrate the importance of spatial harmonic effects. A harmonic theory for thermal reactors is developed. A new method of measuring, subcritical reactivity at moderately high frequencies is suggested which circumvents the harmonic problem. It is shown that at high frequencies there is more sensitivity than at low frequencies and that this could lead to an increased range over which subcritical reactivity can be measured. (author)

  6. Design and Construct of In-Hospital Neutron Irradiator

    International Nuclear Information System (INIS)

    The In-hospital neutron irradiator (IHNI) is designed based on the design of the Miniature Neutron Source Reactor (MNSR) for boron neutron capture therapy (BNCT), NAA, physics experiments, training and teaching. The reactor of the IHNI with thermal power 30 kW is an undermoderated reactor of pool-tank type, UO2 with enrichment of 12.5% as fuel, light water as coolant and moderator, and metal beryllium as reflector. The fission heat produced by the reactor is removed by the natural circulation. On the both sides of the reactor core, there are two neutron beams, one is a thermal neutron beam, and the other, opposite to the thermal beam, is an epithermal neutron beam. An experimental thermal neutron beam is specially designed for the prompt gamma neutron activation analysis (PGNAA). In this paper, the design and experiment results of IHNI will be introduced. (author)

  7. Advanced Neutron Source: Plant Design Requirements

    Energy Technology Data Exchange (ETDEWEB)

    1990-07-01

    The Advanced Neutron Source will be a new world-class facility for research using hot, thermal, cold, and ultra-cold neutrons. The heart of the facility will be a 330-MW (fission), heavy-water cooled and heavy-water moderated reactor. The reactor will be housed in a central reactor building, with supporting equipment located in an adjoining reactor support building. An array of cold neutron guides will fan out into a large guide hall, housing about 30 neutron research stations. Appropriate office, laboratory, and shop facilities will be included to provide a complete facility for users. The ANS is scheduled to begin operation at the Oak Ridge National Laboratory early in the next decade. This PDR document defines the plant-level requirements for the design, construction, and operation of ANS. It also defines and provides input to the individual System Design Description (SDD) documents. Together, this PDR document and the set of SDD documents will define and control the baseline configuration of ANS.

  8. Advanced Neutron Source: Plant Design Requirements

    International Nuclear Information System (INIS)

    The Advanced Neutron Source will be a new world-class facility for research using hot, thermal, cold, and ultra-cold neutrons. The heart of the facility will be a 330-MW (fission), heavy-water cooled and heavy-water moderated reactor. The reactor will be housed in a central reactor building, with supporting equipment located in an adjoining reactor support building. An array of cold neutron guides will fan out into a large guide hall, housing about 30 neutron research stations. Appropriate office, laboratory, and shop facilities will be included to provide a complete facility for users. The ANS is scheduled to begin operation at the Oak Ridge National Laboratory early in the next decade. This PDR document defines the plant-level requirements for the design, construction, and operation of ANS. It also defines and provides input to the individual System Design Description (SDD) documents. Together, this PDR document and the set of SDD documents will define and control the baseline configuration of ANS

  9. Blankets for fusion reactors : materials and neutronics

    International Nuclear Information System (INIS)

    The studies about Fusion Reactors have lead to several problems for which there is no general agreement about the best solution. Nevertheless, several points seem to be well defined, at least for the first generation of reactors. The fuel, for example, should be a mixture of deuterium and tritium. Therefore, the reactor should be able to generate the tritium to be burned and also to transform kinetic energy of the fusion neutrons into heat in a process similar to the fission reactors. The best materials for the composition of the blanket were first selected and then the neutronics for the proposed system was developed. The neutron flux in the blanket was calculated using the discrete ordinates transport code, ANISN. All the nuclides cross sections came from the DLC-28/CTR library, that processed the ENDF/B data, using the SUPERTOG Program. (Author)

  10. Measurements of miniature ionization chamber currents in the JSI TRIGA Mark II reactor demonstrate the importance of the delayed contribution to the photon field in nuclear reactors

    Science.gov (United States)

    Radulović, Vladimir; Fourmentel, Damien; Barbot, Loïc; Villard, Jean-François; Kaiba, Tanja; Gašper, Žerovnik; Snoj, Luka

    2015-12-01

    The characterization of experimental locations of a research nuclear reactor implies the determination of neutron and photon flux levels within, with the best achievable accuracy. In nuclear reactors, photon fluxes are commonly calculated by Monte Carlo simulations but rarely measured on-line. In this context, experiments were conducted with a miniature gas ionization chamber (MIC) based on miniature fission chamber mechanical parts, recently developed by the CEA (French Atomic Energy and Alternative Energies Commission) irradiated in the core of the Jožef Stefan Institute TRIGA Mark II reactor in Ljubljana, Slovenia. The aim of the study was to compare the measured MIC currents with calculated currents based on simulations with the MCNP6 code. A discrepancy of around 50% was observed between the measured and the calculated currents; in the latter taking into consideration only the prompt photon field. Further experimental measurements of MIC currents following reactor SCRAMs (reactor shutdown with rapid insertions of control rods) provide evidence that over 30% of the total measured signal is due to the delayed photon field, originating from fission and activation products, which are untreated in the calculations. In the comparison between the measured and calculated values, these findings imply an overall discrepancy of less than 20% of the total signal which is still unexplained.

  11. The measurement of neutron and neutron induced photon spectra in fusion reactor related assemblies

    CERN Document Server

    Unholzer, S; Klein, H; Seidel, K

    2002-01-01

    The spectral neutron and photon fluence (or flux) measured outside and inside of assemblies related to fusion reactor constructions are basic quantities of fusion neutronics. The comparison of measured spectra with the results of MCNP neutron and photon transport calculations allows a crucial test of evaluated nuclear data as generally used in fusion applications to be carried out. The experiments concern mixed neutron/photon fields with about the same intensity of the two components. An NE-213 scintillation spectrometer, well described by response matrices for both neutrons and photons, is used as proton-recoil and Compton spectrometer. The experiments described here in more detail address the background problematic of two applications, an iron benchmark experiment with an ns-pulsed neutron source and a deep penetration mock-up experiment for the investigation of the ITER in-board shield system. The measured spectral neutron and photon fluences are compared with spectra calculated with the MCNP code on the b...

  12. Reference Neutron Radiographs of Nuclear Reactor Fuel

    DEFF Research Database (Denmark)

    Domanus, Joseph Czeslaw

    1986-01-01

    Reference neutron radiographs of nuclear reactor fuel were produced by the Euraton Neutron Radiography Working Group and published in 1984 by the Reidel Publishing Company. In this collection a classification is given of the various neutron radiographic findings, that can occur in different parts...... of pelletized, annular and vibro-conpacted nuclear fuel pins. Those parts of the pins are shown where changes of appearance differ from those for the parts as fabricated. Also radiographs of those as fabricated parts are included. The collection contains 158 neutron radiographs, reproduced on photographic paper...

  13. High Brightness Neutron Source for Radiography

    Energy Technology Data Exchange (ETDEWEB)

    Cremer, J. T.; Piestrup, Melvin, A.; Gary, Charles, K.; Harris, Jack, L. Williams, David, J.; Jones, Glenn, E.; Vainionpaa, J. , H.; Fuller, Michael, J.; Rothbart, George, H.; Kwan, J., W.; Ludewigt, B., A.; Gough, R.., A..; Reijonen, Jani; Leung, Ka-Ngo

    2008-12-08

    This research and development program was designed to improve nondestructive evaluation of large mechanical objects by providing both fast and thermal neutron sources for radiography. Neutron radiography permits inspection inside objects that x-rays cannot penetrate and permits imaging of corrosion and cracks in low-density materials. Discovering of fatigue cracks and corrosion in piping without the necessity of insulation removal is possible. Neutron radiography sources can provide for the nondestructive testing interests of commercial and military aircraft, public utilities and petrochemical organizations. Three neutron prototype neutron generators were designed and fabricated based on original research done at the Lawrence Berkeley National Laboratory (LBNL). The research and development of these generators was successfully continued by LBNL and Adelphi Technology Inc. under this STTR. The original design goals of high neutron yield and generator robustness have been achieved, using new technology developed under this grant. In one prototype generator, the fast neutron yield and brightness was roughly 10 times larger than previously marketed neutron generators using the same deuterium-deuterium reaction. In another generator, we integrate a moderator with a fast neutron source, resulting in a high brightness thermal neutron generator. The moderator acts as both conventional moderator and mechanical and electrical support structure for the generator and effectively mimics a nuclear reactor. In addition to the new prototype generators, an entirely new plasma ion source for neutron production was developed. First developed by LBNL, this source uses a spiral antenna to more efficiently couple the RF radiation into the plasma, reducing the required gas pressure so that the generator head can be completely sealed, permitting the possible use of tritium gas. This also permits the generator to use the deuterium-tritium reaction to produce 14-MeV neutrons with increases

  14. Measurement of photon flux with a miniature gas ionization chamber in a Material Testing Reactor

    Science.gov (United States)

    Fourmentel, D.; Filliatre, P.; Villard, J. F.; Lyoussi, A.; Reynard-Carette, C.; Carcreff, H.

    2013-10-01

    Nuclear heating measurements in Material Testing Reactors (MTR) are crucial for the design of the experimental devices and the prediction of the temperature of the hosted samples. Nuclear heating in MTR materials (except fuel) is mainly due to the energy deposition by the photon flux. Therefore, the photon flux is a key input parameter for the computer codes which simulate nuclear heating and temperature reached by samples/devices under irradiation. In the Jules Horowitz MTR under construction at the CEA Cadarache, the maximal expected nuclear heating levels will be about 15 to 18 W g-1 and it will be necessary to assess this parameter with the best accuracy. An experiment was performed at the OSIRIS reactor to combine neutron flux, photon flux and nuclear heating measurements to improve the knowledge of the nuclear heating in MTR. There are few appropriate sensors for selective measurement of the photon flux in MTR even if studies and developments are ongoing. An experiment, called CARMEN-1, was conducted at the OSIRIS MTR and we used in particular a gas ionization chamber based on miniature fission chamber design to measure the photon flux. In this paper, we detail Monte-Carlo simulations to analyze the photon fluxes with ionization chamber measurements and we compare the photon flux calculations to the nuclear heating measurements. These results show a good accordance between photon flux measurements and nuclear heating measurement and allow improving the knowledge of these parameters.

  15. Measurement of photon flux with a miniature gas ionization chamber in a Material Testing Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Fourmentel, D., E-mail: damien.fourmentel@cea.fr [CEA, DEN, DER, Instrumentation Sensors and Dosimetry Laboratory, Cadarache, F-13108 St-Paul-Lez-Durance (France); Filliatre, P.; Villard, J.F.; Lyoussi, A. [CEA, DEN, DER, Instrumentation Sensors and Dosimetry Laboratory, Cadarache, F-13108 St-Paul-Lez-Durance (France); Reynard-Carette, C. [Aix-Marseille Université, LISA EA 4672, cedex 20, Marseille 13397 (France); Carcreff, H. [CEA, DEN, DRSN, Saclay, F-91191 Gif-sur-Yvette (France)

    2013-10-01

    Nuclear heating measurements in Material Testing Reactors (MTR) are crucial for the design of the experimental devices and the prediction of the temperature of the hosted samples. Nuclear heating in MTR materials (except fuel) is mainly due to the energy deposition by the photon flux. Therefore, the photon flux is a key input parameter for the computer codes which simulate nuclear heating and temperature reached by samples/devices under irradiation. In the Jules Horowitz MTR under construction at the CEA Cadarache, the maximal expected nuclear heating levels will be about 15 to 18 W g{sup −1} and it will be necessary to assess this parameter with the best accuracy. An experiment was performed at the OSIRIS reactor to combine neutron flux, photon flux and nuclear heating measurements to improve the knowledge of the nuclear heating in MTR. There are few appropriate sensors for selective measurement of the photon flux in MTR even if studies and developments are ongoing. An experiment, called CARMEN-1, was conducted at the OSIRIS MTR and we used in particular a gas ionization chamber based on miniature fission chamber design to measure the photon flux. In this paper, we detail Monte-Carlo simulations to analyze the photon fluxes with ionization chamber measurements and we compare the photon flux calculations to the nuclear heating measurements. These results show a good accordance between photon flux measurements and nuclear heating measurement and allow improving the knowledge of these parameters.

  16. High Intensity Accelerator and Neutron Source in China

    Science.gov (United States)

    Guan, Xialing; Wei, J.; Loong, Chun

    2011-06-01

    High intensity Accelerator is being studied all over world for numerous applications, which includes the waste transmutation, spallation neutron source and material irradiation facilities. The R/D activities of the technology of High intensity accelerator are also developed in China for some year, and have some good facilities around China. This paper will reports the status of some high intensity accelerators and neutron source in China, which including ADS/RFQ; CARR; CSNS; PKUNIFTY & CPHS. This paper will emphatically report the Compact Pulsed Hadron Source (CPHS) led by the Department of Engineering Physics of Tsinghua University in Beijing, China.

  17. How to Produce a Reactor Neutron Spectrum Using a Proton Accelerator

    Science.gov (United States)

    Burns, K.; Wootan, D.; Gates, R.; Schmitt, B.; Asner, D. M.

    A method for reproducing the neutron energy spectrum present in the core of an operating nuclear reactor using an engineered target in an accelerator proton beam is proposed. The protons interact with a target to create neutrons through various (p,n) type reactions. Spectral tailoring of the emitted neutrons can be used to modify the energy of the generated neutron spectrum to represent various reactor spectra. Through the use of moderators and reflectors, the neutron spectrum can be modified to reproduce many different spectra of interest including spectra in small thermal test reactors, large pressurized water reactors, and fast reactors. The particular application of this methodology is the design of an experimental approach for using an accelerator to measure the betas produced during fission to be used to reduce uncertainties in the interpretation of reactor antineutrino measurements. This approach involves using a proton accelerator to produce a neutron field representative of a power reactor, and using this neutron field to irradiate fission foils of the primary isotopes contributing to fission in the reactor, creating unstable, neutron rich fission products that subsequently beta decay and emit electron antineutrinos. A major advantage of an accelerator neutron source over a neutron beam from a thermal reactor is that the fast neutrons can be slowed down or tailored to approximate various power reactor spectra. An accelerator based neutron source that can be tailored to match various reactor neutron spectra provides an advantage for control in studying how changes in the neutron spectra affect parameters such as the resulting fission product beta spectrum.

  18. How to produce a reactor neutron spectrum using a proton accelerator

    Energy Technology Data Exchange (ETDEWEB)

    Burns, Kimberly A.; Wootan, David W.; Gates, Robert O.; Schmitt, Bruce E.; Asner, David M.

    2015-01-01

    A method for reproducing the neutron energy spectrum present in the core of an operating nuclear reactor using an engineered target in an accelerator proton beam is proposed. The protons interact with a target to create neutrons through various (p,n) type reactions. Spectral tailoring of the emitted neutrons can be used to modify the energy of the generated neutron spectrum to represent various reactor spectra. Through the use of moderators and reflectors, the neutron spectrum can be modified to reproduce many different spectra of interest including spectra in small thermal test reactors, large pressurized water reactors, and fast reactors. The particular application of this methodology is the design of an experimental approach for using an accelerator to measure the betas produced during fission to be used to reduce uncertainties in the interpretation of reactor antineutrino measurements. This approach involves using a proton accelerator to produce a neutron field representative of a power reactor, and using this neutron field to irradiate fission foils of the primary isotopes contributing to fission in the reactor, creating unstable, neutron rich fission products that subsequently beta decay and emit electron antineutrinos. A major advantage of an accelerator neutron source over a neutron beam from a thermal reactor is that the fast neutrons can be slowed down or tailored to approximate various power reactor spectra. An accelerator based neutron source that can be tailored to match various reactor neutron spectra provides an advantage for control in studying how changes in the neutron spectra affect parameters such as the resulting fission product beta spectrum.

  19. The Chinese Spallation Neutron Source Project

    International Nuclear Information System (INIS)

    The proposal of the Chinese Spallation Neutron Source (CSNS) project was granted in the beginning of 2002 after three review meetings, organized by the Chinese Academy of Sciences (CAS) and other scientific organizations. Physicists from the Institute of Physics (IP) and the Institute of High Energy Physics (IHEP), both belonging to CAS, consequently started a conceptual design and feasibility study. The CSNS plan calls for a 70-MeV H- linac and a 1.6 GeV rapid cycling synchrotron producing a proton current of 62.5 μA (100kW) at a 25 Hz repetition rate. It should be able to be upgraded to a higher beam power in its second phase. The CSNS target station design team, has initiated to conceptual design of the targetmoderator system based on the suggestions and comments from an international advisory team, in the first moderator-target planning meeting of CSNS project (21-26, April 2002 in Beijing). In consideration of the characteristics of the spallation neutron source, the budgets and possible requests for future users in China, five multi-purpose neutron scattering spectrometers were proposed as the first step

  20. Evaluated neutron data for thermal reactor calculations

    International Nuclear Information System (INIS)

    The paper describes a library of evaluated neutron data designed for thermal reactor calculations and other low energy neutron physics applications. The name of the library is KORT (Evaluated Thermal Reactor Constants). The following information is given in KORT: a general characterization of the nucleus (mass, energy of capture and fission reactions, parameters of radioactive decay); partial cross-sections for neutrons of thermal energy, and the number of secondary fission neutrons (estimated errors in the measurements of these quantities are indicated); coefficients defining the deviation of capture and fission cross-sections from the 1/v law in a Maxwellian spectrum; resonance capture and fission integrals and the estimated errors in these quantities (for nuclei with Z>=90); detailed energy dependence of the cross-sections in the 10-4-5 eV region at T=300 K

  1. Statistically based uncertainty analysis for ranking of component importance in the thermal-hydraulic safety analysis of the Advanced Neutron Source Reactor

    International Nuclear Information System (INIS)

    The Analytic Hierarchy Process (AHP) has been used to help determine the importance of components and phenomena in thermal-hydraulic safety analyses of nuclear reactors. The AHP results are based, in part on expert opinion. Therefore, it is prudent to evaluate the uncertainty of the AHP ranks of importance. Prior applications have addressed uncertainty with experimental data comparisons and bounding sensitivity calculations. These methods work well when a sufficient experimental data base exists to justify the comparisons. However, in the case of limited or no experimental data the size of the uncertainty is normally made conservatively large. Accordingly, the author has taken another approach, that of performing a statistically based uncertainty analysis. The new work is based on prior evaluations of the importance of components and phenomena in the thermal-hydraulic safety analysis of the Advanced Neutron Source Reactor (ANSR), a new facility now in the design phase. The uncertainty during large break loss of coolant, and decay heat removal scenarios is estimated by assigning a probability distribution function (pdf) to the potential error in the initial expert estimates of pair-wise importance between the components. Using a Monte Carlo sampling technique, the error pdfs are propagated through the AHP software solutions to determine a pdf of uncertainty in the system wide importance of each component. To enhance the generality of the results, study of one other problem having different number of elements is reported, as are the effects of a larger assumed pdf error in the expert ranks. Validation of the Monte Carlo sample size and repeatability are also documented

  2. The tokamak as a neutron source

    International Nuclear Information System (INIS)

    This paper describes the tokamak in its role as a neutron source, with emphasis on experimental results for D-D neutron production. The sections summarize tokamak operation, sources of fusion and non-fusion neutrons, principal neutron detection methods and their calibration, neutron energy spectra and fluxes outside the tokamak plasma chamber, history of neutron production in tokamaks, neutron emission and fusion power gain from JET and TFTR (the largest present-day tokamaks), and D-T neutron production from burnup of D-D tritons. This paper also discusses the prospects for future tokamak neutron production and potential applications of tokamak neutron sources. 100 refs., 16 figs., 4 tabs

  3. Modeling and analysis framework for core damage propagation during flow-blockage-initiated accidents in the Advanced Neutron Source Reactor at Oak Ridge National Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Kim, S.H.; Taleyarkhan, R.P.; Navarro-Valenti, S.; Georgevich, V.

    1995-09-01

    This paper describes modeling and analysis to evaluate the extent of core damage during flow blockage events in the Advanced Neutron Source (ANS) reactor planned to be built at the Oak Ridge National Laboratory (ORNL). Damage propagation is postulated to occur from thermal conduction between damaged and undamaged plates due to direct thermal contact. Such direct thermal contact may occur because of fuel plate swelling during fission product vapor release or plate buckling. Complex phenomena of damage propagation were modeled using a one-dimensional heat transfer model. A scoping study was conducted to learn what parameters are important for core damage propagation, and to obtain initial estimates of core melt mass for addressing recriticality and steam explosion events. The study included investigating the effects of the plate contact area, the convective heat transfer coefficient, thermal conductivity upon fuel swelling, and the initial temperature of the plate being contacted by the damaged plate. Also, the side support plates were modeled to account for their effects on damage propagation. The results provide useful insights into how various uncertain parameters affect damage propagation.

  4. Neutronic analysis for conversion of the Ghana Research Reactor-1 facility using Monte Carlo methods and UO{sub 2} LEU fuel

    Energy Technology Data Exchange (ETDEWEB)

    Anim-Sampong, S.; Akaho, E.H.K.; Maakuu, B.T.; Gbadago, J.K. [Ghana Research Reactor-1 Centre, Dept. of Nuclear Engineering and Materials Science, National Nuclear Research Institute, Ghana Atomic Energy Commission, Legon, Accra (Ghana); Andam, A. [Kwame Nkrumah Univ. of Science and Technology, Dept. of Physics (Ghana); Liaw, J.J.R.; Matos, J.E. [Argonne National Lab., RERTR Programme, Div. of Nuclear Engineering (United States)

    2007-07-01

    Monte Carlo particle transport methods and software (MCNP) have been applied to the modelling, simulation and neutronic analysis for the conversion of the HEU-fuelled (high enrichment uranium) core of the Ghana Research Reactor-1 (GHARR-1) facility. The results show that the MCNP model of the GHARR-1 facility, which is a commercial version of the Miniature Neutron Source Reactor (MNSR) is good as the simulated neutronic and other reactor physics parameters agree with very well with experimental and zero power results. Three UO{sub 2} LEU (low enrichment uranium) fuels with different enrichments (12.6% and 19.75%), core configurations, core loadings were utilized in the conversion studies. The nuclear criticality and kinetic parameters obtained from the Monte Carlo simulation and neutronic analysis using three UO{sub 2} LEU fuels are in close agreement with results obtained for the reference 90.2% U-Al HEU core. The neutron flux variation in the core, fission chamber and irradiation channels for the LEU UO{sub 2} fuels show the same trend as the HEU core as presented in the paper. The Monte Carlo model confirms a reduction (8% max) in the peak neutron fluxes simulated in the irradiation channels which are utilized for experimental and commercial activities. However, the reductions or 'losses' in the flux levels neither affects the criticality safety, reactor operations and safety nor utilization of the reactor. Employing careful core loading optimization techniques and fuel loadings and enrichment, it is possible to eliminate the apparent reductions or 'losses' in the neutron fluxes as suggested in this paper. Concerning neutronics, it can be concluded that all the 3 LEU fuels qualify as LEU candidates for core conversion of the GHARR-1 facility.

  5. Polarized neutron reflectometry at Dhruva reactor

    Indian Academy of Sciences (India)

    Surendra Singh; Saibal Basu

    2004-08-01

    Polarized neutron reflectometry (PNR) is an ideal non-destructive tool for chemical and magnetic characterization of thin films and multilayers. We have installed a position sensitive detector-based polarized neutron reflectometer at Dhruva reactor, Trombay. In this paper we will discuss the results obtained from this instrument for two multilayer samples. The first sample is a (Ni–Mo alloy)/Ti multilayer sample. We have determined the chemical structure of this multilayer by unpolarized neutron reflectometry (NR). The other sample is a Fe/Ge multilayer sample for which we obtained the chemical structure by NR and magnetic moment per Fe atom by PNR.

  6. Introduction to the neutron kinetics of nuclear power reactors

    CERN Document Server

    Tyror, J G; Grant, P J

    2013-01-01

    An Introduction to the Neutron Kinetics of Nuclear Power Reactors introduces the reader to the neutron kinetics of nuclear power reactors. Topics covered include the neutron physics of reactor kinetics, feedback effects, water-moderated reactors, fast reactors, and methods of plant control. The reactor transients following faults are also discussed, along with the use of computers in the study of power reactor kinetics. This book is comprised of eight chapters and begins with an overview of the reactor physics characteristics of a nuclear power reactor and their influence on system design and

  7. Materials for spallation neutron sources

    Energy Technology Data Exchange (ETDEWEB)

    Sommer, W.F.; Daemen, L.L. [comps.

    1996-03-01

    The Workshop on Materials for Spallation Neutron Sources at the Los Alamos Neutron Science Center, February 6 to 10, 1995, gathered scientists from Department of Energy national laboratories, other federal institutions, universities, and industry to discuss areas in which work is needed, successful designs and use of materials, and opportunities for further studies. During the first day of the workshop, speakers presented overviews of current spallation neutron sources. During the next 3 days, seven panels allowed speakers to present information on a variety of topics ranging from experimental and theoretical considerations on radiation damage to materials safety issues. An attempt was made to identify specific problems that require attention within the context of spallation neutron sources. This proceedings is a collection of summaries from the overview sessions and the panel presentations.

  8. Elemental activation analysis with decay and prompt gamma ray techniques, using isotopic neutron sources and a nuclear research reactor. Part of a coordinated programme on nuclear-based techniques in geology and mineral prospecting

    International Nuclear Information System (INIS)

    This is to review the research activities carried out through the IAEA research Project 1697/RB ''Elemental activation analysis with decay and prompt gamma ray techniques, using isotopic neutron sources and the nuclear research reactor. The programme of work includes: a) Development of decay and prompt gamma ray activation techniques for mineral exploration. b) Development of epithermal NAA in addition to thermal NAA especially for gold ore. c) Development of non-destructive insitu elemental analysis with decay and prompt gamma ray techniques using isotopic neutron sources. A joint programme has been established with the Egyptian Geological Surrey and Mining Authority for using nuclear techniques in evaluating gold prospects of several ancient gold mines and investigating several tin-tantalum deposits, which were discovered over the last few years. Two sources of neutrons were used for irradiation, one of the dry channels of the two megawatts research reactor, ET-RR-1 for laboratory studies, and a Pu-Be neutron source in paraffin assembly for possible insitu work

  9. Beam characterization at the Neutron Radiography Reactor

    International Nuclear Information System (INIS)

    Highlights: • The project characterized the beam at the Neutron Radiography Reactor. • Experiments indicate that the neutron energy spectrum model may not be accurate. • The facility is a category I radiography facility. • The beam divergence and effective collimation ratio are 0.3 ± 0.1° and >125. • The predicted total neutron flux at the image plane is 5.54 × 106 n/cm2 s. -- Abstract: The quality of a neutron-imaging beam directly impacts the quality of radiographic images produced using that beam. Fully characterizing a neutron beam, including determination of the beam's effective length-to-diameter ratio, neutron flux profile, energy spectrum, potential image quality, and beam divergence, is vital for producing quality radiographic images. This paper provides a characterization of the east neutron imaging beamline at the Idaho National Laboratory Neutron Radiography Reactor (NRAD). The experiments which measured the beam's effective length-to-diameter ratio and potential image quality are based on American Society for Testing and Materials (ASTM) standards. An analysis of the image produced by a calibrated phantom measured the beam divergence. The energy spectrum measurements consist of a series of foil irradiations using a selection of activation foils, compared to the results produced by a Monte Carlo n-Particle (MCNP) model of the beamline. The NRAD has an effective collimation ratio greater than 125, a beam divergence of 0.3 ± 0.1°, and a gold foil cadmium ratio of 2.7. The flux profile has been quantified and the facility is an ASTM Category 1 radiographic facility. Based on bare and cadmium covered foil activation results, the neutron energy spectrum used in the current MCNP model of the radiography beamline over-samples the thermal region of the neutron energy spectrum

  10. BNL Activities in Advanced Neutron Source Development: Past and Present

    Energy Technology Data Exchange (ETDEWEB)

    Hastings, J.B.; Ludewig, H.; Montanez, P.; Todosow, M.; Smith, G.C.; Larese, J.Z.

    1998-06-14

    Brookhaven National Laboratory has been involved in advanced neutron sources almost from its inception in 1947. These efforts have mainly focused on steady state reactors beginning with the construction of the first research reactor for neutron beams, the Brookhaven Graphite Research Reactor. This was followed by the High Flux Beam Reactor that has served as the design standard for all the subsequent high flux reactors constructed worldwide. In parallel with the reactor developments BNL has focused on the construction and use of high energy proton accelerators. The first machine to operate over 1 GeV in the world was the Cosmotron. The machine that followed this, the AGS, is still operating and is the highest intensity proton machine in the world and has nucleated an international collaboration investigating liquid metal targets for next generation pulsed spallation sources. Early work using the Cosmotron focused on spallation product studies for both light and heavy elements into the several GeV proton energy region. These original studies are still important today. In this report we discuss the facilities and activities at BNL focused on advanced neutron sources. BNL is involved in the proton source for the Spallation Neutron source, spectrometer development at LANSCE, target studies using the AGS and state-of-the-art neutron detector development.

  11. BNL ACTIVITIES IN ADVANCED NEUTRON SOURCE DEVELOPMENT: PAST AND PRESENT

    Energy Technology Data Exchange (ETDEWEB)

    HASTINGS,J.B.; LUDEWIG,H.; MONTANEZ,P.; TODOSOW,M.; SMITH,G.C.; LARESE,J.Z.

    1998-06-14

    Brookhaven National Laboratory has been involved in advanced neutron sources almost from its inception in 1947. These efforts have mainly focused on steady state reactors beginning with the construction of the first research reactor for neutron beams, the Brookhaven Graphite Research Reactor. This was followed by the High Flux Beam Reactor that has served as the design standard for all the subsequent high flux reactors constructed worldwide. In parallel with the reactor developments BNL has focused on the construction and use of high energy proton accelerators. The first machine to operate over 1 GeV in the world was the Cosmotron. The machine that followed this, the AGS, is still operating and is the highest intensity proton machine in the world and has nucleated an international collaboration investigating liquid metal targets for next generation pulsed spallation sources. Early work using the Cosmotron focused on spallation product studies for both light and heavy elements into the several GeV proton energy region. These original studies are still important today. In the sections below the authors discuss the facilities and activities at BNL focused on advanced neutron sources. BNL is involved in the proton source for the Spallation Neutron source, spectrometer development at LANSCE, target studies using the AGS and state-of-the-art neutron detector development.

  12. Neutronic parameters calculations of a CANDU reactor

    International Nuclear Information System (INIS)

    Neutronic calculations that reproduce in a simplified way some aspects of a CANDU reactor design were performed. Starting from some prefixed reactor parameters, cylindrical and uniform iron adjuster rods were designed. An appropriate refueling scheme was established, defininig in a 2 zones model their dimensions and exit burnups. The calculations have been done using the codes WIMS-D4 (cell), SNOD (reactivity device simulations) and PUMA (reactor). Comparing with similar calculations done with codes and models usually employed for CANDU design, it is concluded that the models and methods used are appropriate. (Author)

  13. RA-0 reactor. New neutronic calculations

    International Nuclear Information System (INIS)

    An updating of the neutronic calculations performed at the RA-0 reactor, located at the Natural, Physical and Exact Sciences Faculty of Cordoba National University, are herein described. The techniques used for the calculation of a reactor like the RA-0 allows prediction in detail of the flux behaviour in the core's interior and in the reflector, which will be helpful for experiments design. In particular, the use of WIMSD4 code to make calculations on the reactor implies a novelty in the possible applications of this code to solve the problems that arise in practice. (Author)

  14. RELAP5 model for advanced neutron source reactor thermal-hydraulic transients, three-element-core design

    International Nuclear Information System (INIS)

    In order to utilize reduced enrichment fuel, the three-element-core design has been proposed. The proposed core configuration consists of inner, middle, and outer elements, with the middle element offset axially beneath the inner and outer elements, which are axially aligned. The three-element-core RELAP5 model assumes that the reactor hardware is changed only within the core region, so that the loop piping, heat exchangers, and pumps remain as assumed for the two-element-core configuration. However, the total flow rate through the core is greater and the pressure drop across the core is less so that the primary coolant pumps and heat exchangers are operating at a different point in their performance curves. This report describes the new RELAP5 input for the core components

  15. The lead nucleus as a miniature surrogate for a neutron star

    International Nuclear Information System (INIS)

    Full context: The nucleus of 208 Pb, a system 18 order of magnitudes smaller and 55 orders of magnitude lighter than a neutron star, may be used as a miniature surrogate to establish important correlations between its neutron skin and several neutron star properties. Relativistic models that reproduce a variety of ground-state observables can not determine uniquely the neutron skin of a heavy nucleus. Thus, a large range of neutron skins may be generated by supplementing the models with nonlinear couplings between isoscalar and isovector mesons. These studies are particularly timely as an accurate measurement of the neutron radius in 208 Pb via parity violating electron scattering has been proposed at the Thomas Jefferson Laboratory. Moreover, a number of improved radii measurements on isolated neutron stars, such as Geminga and RX J185635-3754 are now available. Theories with a thicker neutron skin in 208 Pb generate larger neutron stars that have a higher electron fraction and a lower liquid-to-solid transition density for neutron rich matter. These properties are determined by the density dependence of the symmetry energy which we modify by varying the couplings between isoscalar and isovector mesons. Finally, we illustrate how the correlation between the neutron skin and the radius of the star can be used to place important constraints on the equation of state and how it may help elucidate the existence of a phase transition in the interior of the neutron star. (Author)

  16. Nuclear data guide for reactor neutron metrology

    International Nuclear Information System (INIS)

    Part I lists numerical data on activation detector materials, on detection reactions involved, and on radionuclides produced. The data are mainly taken from existing evaluations. Where evaluations were missing, unknown or obsolete, data were taken from recent literature. The detection reactions considered were selected on the basis of available information of its use for reactor neutron measurements. In the opinion of the Euratom Working Group on Reactor Dosimetry (EWGRD) this document may contribute to the creation of a common data set for all laboratories working in the field of reactor neutron metrology. Part II lists numerical data on fission reactions, the cross-section data and the decay data of some fission product radionuclides, suitable for gamma counting. The selection of data is based on the usefulness to determine neutron flux densities, neutron fluences and neutron spectra. The data are mainly taken from existing review papers and evaluations. It is expected that after one or two years an updated and revised guide will be issued. The present document supersedes the previous report ECN-37

  17. Use of Zircaloy 4 material for the pressure vessels of hot and cold neutron sources and beam tubes for research reactors

    International Nuclear Information System (INIS)

    The material Zircaloy 4 can be used for the pressure retaining walls for the cold and hot neutron sources and beam tubes. For the research reactor FRM-II of the Technical University Munich, Germany, the material Zircaloy 4 were chosen for the vessels of the cold and hot neutron source and for the beam tube No. 6. The sheets and forgings of Zircaloy 4 were examined in the temperature range between -256degC and 250degC. The thickness of the sheets are 3, 4, 5 and 10 mm, the maximum diameter of the forgings was 560 mm. This great forging diameters are not be treated in the ASTM rule B 351 for nuclear material, so a special approval with independent experts was necessary. The requirements for the material examinations were specified in a material specification and material test sheets which based on the ASTM rules B 351 and B 352 with additional restriction and additional requirements of the basic safety concept for nuclear power plants in Germany, which was taken into consideration in the nuclear licensing procedure. Charpy-V samples were carried out in the temperature range between -256degC and 150degC to get more information on the ductile behaviour of the Zircaloy 4. The results of the sheet examination confirm the requirements of the specifications, the results of the forging examination in the tangential testing direction are lower than specified and expected for the tensile strength. The axial and transverse values confirm the specification requirements. For the strength calculation of the pressure retaining wall a reduced material value for the forgings has to taken into consideration. The material behaviour of Zircaloy 4 under irradiation up to a fluence of ∼1·1022 n/cm2 was investigated. The loss of ductility was determined. As an additional criteria the variation of the fracture toughness was studies. Fracture mechanic calculations of the material were carried out in the licensing procedure with the focus to fulfill the leak before rupture criteria of the

  18. Use of Zircaloy 4 material for the pressure vessels of hot and cold neutron sources and beam tubes for research reactors

    International Nuclear Information System (INIS)

    The material Zircaloy 4 can be used for the pressure retaining walls for the cold and hot neutron sources and beam tubes. For the research reactor FRM-II of the Technical University Munich, Germany, the material Zircaloy 4 were chosen for the vessels of the cold and hot neutron source and for the beam tube No. 6. The sheets and forgings of Zircaloy 4 were examined in the temperature range between -256 deg. C and 250 deg. C. The thickness of the sheets are 3, 4, 5 and 10 mm, the maximum diameter of the forgings was 560 mm. This great forging diameters are not be treated in the ASTM rule B 351 for nuclear material, so a special approval with independent experts was necessary. The requirements for the material examinations were specified in a material specification and material test sheets which based on the ASTM rules B 351 and B 352 with additional restriction and additional requirements of the basic safety concept for nuclear power plants in Germany, which was take into consideration in the nuclear licensing procedure. Charpy-V samples were carried out in the temperature range between -256 deg. C and 150 deg. C to get more information on the ductile behaviour of the Zircaloy 4. The results of the sheet examination confirm the requirements of the specifications, the results of the forging examination in the tangential testing direction are lower than specified and expected for the tensile strength. The axial and transverse values confirm the specification requirements. For the strength calculation of the pressure retaining wall a reduced material value for the forgings has to take into consideration. The material behaviour of Zircaloy 4 under irradiation up to a fluence of ∼ 1x1022 n/cm2 was investigated. The loss of ductility was determined. As an additional criteria the variation of the fracture toughness was studies. Fracture mechanic calculations of the material were carried out in the licensing procedure with the focus to fulfill the leak before rupture

  19. Advanced Neutron Source (ANS) Project Progress report, FY 1991

    Energy Technology Data Exchange (ETDEWEB)

    Campbell, J.H. (ed.) (Oak Ridge National Lab., TN (United States)); Selby, D.L.; Harrington, R.M. (Oak Ridge National Lab., TN (United States)); Thompson, P.B. (Martin Marietta Energy Systems, Inc., (United States). Engineering Division)

    1992-01-01

    This report discusses the following about the Advanced Neutron Source: Project Management; Research and Development; Fuel Development; Corrosion Loop Tests and Analyses; Thermal-Hydraulic Loop Tests; Reactor Control and Shutdown Concepts; Critical and Subcritical Experiments; Material Data, Structural Tests, and Analysis; Cold-Source Development; Beam Tube, Guide, and Instrument Development; Hot-Source Development; Neutron Transport and Shielding; I C Research and Development; Design; and Safety.

  20. Advanced Neutron Source (ANS) Project Progress report, FY 1991

    Energy Technology Data Exchange (ETDEWEB)

    Campbell, J.H. [ed.] [Oak Ridge National Lab., TN (United States); Selby, D.L.; Harrington, R.M. [Oak Ridge National Lab., TN (United States); Thompson, P.B. [Martin Marietta Energy Systems, Inc., (United States). Engineering Division

    1992-01-01

    This report discusses the following about the Advanced Neutron Source: Project Management; Research and Development; Fuel Development; Corrosion Loop Tests and Analyses; Thermal-Hydraulic Loop Tests; Reactor Control and Shutdown Concepts; Critical and Subcritical Experiments; Material Data, Structural Tests, and Analysis; Cold-Source Development; Beam Tube, Guide, and Instrument Development; Hot-Source Development; Neutron Transport and Shielding; I & C Research and Development; Design; and Safety.

  1. Advanced Neutron Source (ANS) Project Progress report, FY 1991

    International Nuclear Information System (INIS)

    This report discusses the following about the Advanced Neutron Source: Project Management; Research and Development; Fuel Development; Corrosion Loop Tests and Analyses; Thermal-Hydraulic Loop Tests; Reactor Control and Shutdown Concepts; Critical and Subcritical Experiments; Material Data, Structural Tests, and Analysis; Cold-Source Development; Beam Tube, Guide, and Instrument Development; Hot-Source Development; Neutron Transport and Shielding; I ampersand C Research and Development; Design; and Safety

  2. Isotopic composition and neutronics of the Okelobondo natural reactor

    Science.gov (United States)

    Palenik, Christopher Samuel

    The Oklo-Okelobondo and Bangombe uranium deposits, in Gabon, Africa host Earth's only known natural nuclear fission reactors. These 2 billion year old reactors represent a unique opportunity to study used nuclear fuel over geologic periods of time. The reactors in these deposits have been studied as a means by which to constrain the source term of fission product concentrations produced during reactor operation. The source term depends on the neutronic parameters, which include reactor operation duration, neutron flux and the neutron energy spectrum. Reactor operation has been modeled using a point-source computer simulation (Oak Ridge Isotope Generation and Depletion, ORIGEN, code) for a light water reactor. Model results have been constrained using secondary ionization mass spectroscopy (SIMS) isotopic measurements of the fission products Nd and Te, as well as U in uraninite from samples collected in the Okelobondo reactor zone. Based upon the constraints on the operating conditions, the pre-reactor concentrations of Nd (150 ppm +/- 75 ppm) and Te (<1 ppm) in uraninite were estimated. Related to the burnup measured in Okelobondo samples (0.7 to 13.8 GWd/MTU), the final fission product inventories of Nd (90 to 1200 ppm) and Te (10 to 110 ppm) were calculated. By the same means, the ranges of all other fission products and actinides produced during reactor operation were calculated as a function of burnup. These results provide a source term against which the present elemental and decay abundances at the fission reactor can be compared. Furthermore, they provide new insights into the extent to which a "fossil" nuclear reactor can be characterized on the basis of its isotopic signatures. In addition, results from the study of two other natural systems related to the radionuclide and fission product transport are included. A detailed mineralogical characterization of the uranyl mineralogy at the Bangombe uranium deposit in Gabon, Africa was completed to improve

  3. Miniature field emission light sources for bio-chips

    International Nuclear Information System (INIS)

    A concept based on preparation of miniature field emission light sources (FELS) for integration with bio-chips is presented. Glass and silicon-glass micro-fluidic systems (biochips) with spectrofluorometric detection are designated for this solution. Planar, miniature silicon-glass field emission light sources were designed and fabricated for this conception. Carbon nanotubes (CNTs) have been used as a low-voltage electron emissive layer. Nanocrystalline yttria matrices doped with rare earth (Re) ions (Re: Eu3+, Tb3+) have been synthesized and utilized as phosphor layers. Light emission spectral characteristics of fabricated sources allow to couple them with typical fluorescent markers as e.g. Alexa, Fluorescein or TO-PRO, used on the wide scale in biochemical researches. Fabricated FELSs are characterized by the intensive and homogenous light emission with well defined sharp emission lines. The efficient and stable field emission from carbon nanotubes has also been noticed. Fabricated FELS are technologically compatible with highly developing micromachined fluidic systems and are able to direct on-chip integration with these microsystems.

  4. Monte Carlo Model of TRIGA Reactor to Support Neutron Activation Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Zerovnik, G.; Snoj, L.; Trkov, A. [Reactor Physics Department, Jozef Stefan Institute, Jamova cesta 39, SI-1000 Ljubljana (Slovenia)

    2011-07-01

    The TRIGA reactor at Jozef Stefan Institute is used as a neutron source for neutron activation analysis. The accuracy of the method depends on the accuracy of the neutron spectrum characterization. Therefore, computational models on different scales have been developed: Monte Carlo full reactor model, model of an irradiation channel and deterministic code for self-shielding factor calculations. The models have been validated by comparing against experiment and thus provide a very strong support for neutron activation analysis of samples irradiated at the TRIGA reactor. (author)

  5. 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.

  6. Generalization of the analytical solution of neutron point kinetics equations with time-dependent external source

    Science.gov (United States)

    Seidi, M.; Behnia, S.; Khodabakhsh, R.

    2014-09-01

    Point reactor kinetics equations with one group of delayed neutrons in the presence of the time-dependent external neutron source are solved analytically during the start-up of a nuclear reactor. Our model incorporates the random nature of the source and linear reactivity variation. We establish a general relationship between the expectation values of source intensity and the expectation values of neutron density of the sub-critical reactor by ignoring the term of the second derivative for neutron density in neutron point kinetics equations. The results of the analytical solution are in good agreement with the results obtained with numerical solution.

  7. Sensitivity analysis of the spectra of the core neutronic source in the calculation of radiation damage in internal of PWR reactor vessel. Internal; Analisis de sensibilidad a los espectros de la fuente neutronica del nucleo en el calculo del dano por irradiacion en los internos de la vasija de un reactor PWR

    Energy Technology Data Exchange (ETDEWEB)

    Cadenas Mendicoa, A. M.; Benito Hernandez, M.; Barrerira Pereira, P.

    2012-07-01

    This study is to analyze the sensitivity to the expected differences in the energy spectra characterizing the neutron source that radiates the vessel internals of a commercial PWR reactor, in order to quantify their influence in the quantities that determine the damage in materials metal.

  8. The Possibilities of Fission Material Reproduction Increase in Thermal Reactor with the Assemblies with a Hard Neutron Spectrum

    Directory of Open Access Journals (Sweden)

    Vladimir M. Kotov

    2011-01-01

    The possibility of additional neutron source development with the use of fast neutrons with an energy distribution close to the fission spectrum in the major part of thermal reactor core is researched in this paper.

  9. Neutron spectrometer for fast nuclear reactors

    CERN Document Server

    Osipenko, M; Ricco, G; Caiffi, B; Pompili, F; Pillon, M; Angelone, M; Verona-Rinati, G; Cardarelli, R; Mila, G; Argiro, S

    2015-01-01

    In this paper we describe the development and first tests of a neutron spectrometer designed for high flux environments, such as the ones found in fast nuclear reactors. The spectrometer is based on the conversion of neutrons impinging on $^6$Li into $\\alpha$ and $t$ whose total energy comprises the initial neutron energy and the reaction $Q$-value. The $^6$LiF layer is sandwiched between two CVD diamond detectors, which measure the two reaction products in coincidence. The spectrometer was calibrated at two neutron energies in well known thermal and 3 MeV neutron fluxes. The measured neutron detection efficiency varies from 4.2$\\times 10^{-4}$ to 3.5$\\times 10^{-8}$ for thermal and 3 MeV neutrons, respectively. These values are in agreement with Geant4 simulations and close to simple estimates based on the knowledge of the $^6$Li(n,$\\alpha$)$t$ cross section. The energy resolution of the spectrometer was found to be better than 100 keV when using 5 m cables between the detector and the preamplifiers.

  10. Accelerator driven reactors, - the significance of the energy distribution of spallation neutrons on the neutron statistics

    International Nuclear Information System (INIS)

    In order to make correct predictions of the second moment of statistical nuclear variables, such as the number of fissions and the number of thermalized neutrons, the dependence of the energy distribution of the source particles on their number should be considered. It has been pointed out recently that neglecting this number dependence in accelerator driven systems might result in bad estimates of the second moment, and this paper contains qualitative and quantitative estimates of the size of these efforts. We walk towards the requested results in two steps. First, models of the number dependent energy distributions of the neutrons that are ejected in the spallation reactions are constructed, both by simple assumptions and by extracting energy distributions of spallation neutrons from a high-energy particle transport code. Then, the second moment of nuclear variables in a sub-critical reactor, into which spallation neutrons are injected, is calculated. The results from second moment calculations using number dependent energy distributions for the source neutrons are compared to those where only the average energy distribution is used. Two physical models are employed to simulate the neutron transport in the reactor. One is analytical, treating only slowing down of neutrons by elastic scattering in the core material. For this model, equations are written down and solved for the second moment of thermalized neutrons that include the distribution of energy of the spallation neutrons. The other model utilizes Monte Carlo methods for tracking the source neutrons as they travel inside the reactor material. Fast and thermal fission reactions are considered, as well as neutron capture and elastic scattering, and the second moment of the number of fissions, the number of neutrons that leaked out of the system, etc. are calculated. Both models use a cylindrical core with a homogenous mixture of core material. Our results indicate that the number dependence of the energy

  11. Material selection for spallation neutron source windows

    Energy Technology Data Exchange (ETDEWEB)

    Sordo, F. [ETSII/Universidad Politecnica de Madrid, J. Gutierrez Abascal, 2-28006 Madrid (Spain); Abanades, A. [ETSII/Universidad Politecnica de Madrid, J. Gutierrez Abascal, 2-28006 Madrid (Spain)], E-mail: abanades@etsii.upm.es; Lafuente, A.; Martinez-Val, J.M. [ETSII/Universidad Politecnica de Madrid, J. Gutierrez Abascal, 2-28006 Madrid (Spain); Perlado, M. [Instituto de Fusion Nuclear (DENIM)/ETSII/Universidad Politecnica, Madrid, J. Gutierrez Abascal, 2-28006 Madrid (Spain)

    2009-11-15

    High performance neutron sources are being proposed for many scientific and industrial applications, ranging from material studies, hybrid reactors and transmutation of nuclear wastes. In the case of transmutation of nuclear wastes, accelerator driven systems (ADS) are considered as one of the main technical options for such purpose. In ADS a high performance spallation neutron source becomes an essential element for its operation and control. This spallation source must fulfil very challenging nuclear and thermo-mechanical requirements, because of the high neutron rates needed in ADS. The material selection for this key component becomes of paramount importance, particularly the source window that separates the vacuum accelerator tube from the spallation material where the accelerated protons impinge. In this paper, an integral analysis of spallation sources is done, taking as a reference the projects in this field proposal in the framework of European projects. Our analysis and calculations show that titanium and vanadium alloys are more suitable than steel as structural material for an industrial ADS beam window, mostly due to its irradiation damage resistance.

  12. Source apportionment studies on particulate matter in Beijing/China

    Science.gov (United States)

    Suppan, P.; Shen, R.; Shao, L.; Schrader, S.; Schäfer, K.; Norra, S.; Vogel, B.; Cen, K.; Wang, Y.

    2013-05-01

    More than 15 million people in the greater area of Beijing are still suffering from severe air pollution levels caused by sources within the city itself but also from external impacts like severe dust storms and long range advection from the southern and central part of China. Within this context particulate matter (PM) is the major air pollutant in the greater area of Beijing (Garland et al., 2009). PM did not serve only as lead substance for air quality levels and therefore for adverse health impact effects but also for a strong influence on the climate system by changing e.g. the radiative balance. Investigations on emission reductions during the Olympic Summer Games in 2008 have caused a strong reduction on coarser particles (PM10) but not on smaller particles (PM2.5). In order to discriminate the composition of the particulate matter levels, the different behavior of coarser and smaller particles investigations on source attribution, particle characteristics and external impacts on the PM levels of the city of Beijing by measurements and modeling are performed: a) Examples of long term measurements of PM2.5 filter sampling in 2010/2011 with the objectives of detailed chemical (source attribution, carbon fraction, organic speciation and inorganic composition) and isotopic analyses as well as toxicological assessment in cooperation with several institutions (Karlsruhe Institute of Technology (IfGG/IMG), Helmholtz Zentrum München (HMGU), University Rostock (UR), Chinese University of Mining and Technology Beijing, CUMTB) will be discussed. b) The impact of dust storm events on the overall pollution level of particulate matter in the greater area of Beijing is being assessed by the online coupled comprehensive model system COSMO-ART. First results of the dust storm modeling in northern China (2011, April 30th) demonstrates very well the general behavior of the meteorological parameters temperature and humidity as well as a good agreement between modeled and

  13. Prompt Neutron Lifetime for the NBSR Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hanson, A.L.; Diamond, D.

    2012-06-24

    In preparation for the proposed conversion of the National Institute of Standards and Technology (NIST) research reactor (NBSR) from high-enriched uranium (HEU) to low-enriched uranium (LEU) fuel, certain point kinetics parameters must be calculated. We report here values of the prompt neutron lifetime that have been calculated using three independent methods. All three sets of calculations demonstrate that the prompt neutron lifetime is shorter for the LEU fuel when compared to the HEU fuel and longer for the equilibrium end-of-cycle (EOC) condition when compared to the equilibrium startup (SU) condition for both the HEU and LEU fuels.

  14. Interferences in reactor neutron activation analyses

    International Nuclear Information System (INIS)

    It has been shown that interfering reactions may occur in neutron activation analyses of aluminum and zinc matrixes, commonly used in nuclear areas. The interferences analysed were: Al2713 (n, α) Na2411 and Zn6430 (n, p) Cu6429. The method used was the non-destructive neutron activation analysis and the spectra were obtained in a 1024 multichannel system coupled with a Ge(Li) detector. Sodium was detected in aluminum samples from the reactor tank and pneumatic transfer system. The independence of the sodium concentration in samples in the range of 0 - 100 ppm is shown by the attenuation obtained with the samples encapsulated in cadmium. (Author)

  15. Source Apportionment of Fine Particles in Xinzhen, Beijing

    Institute of Scientific and Technical Information of China (English)

    JIN; Xiang-chun; ZHANG; Gui-ying; XIAO; Cai-jin; WANG; Ping-sheng; WANG; Xing-hua; HUA; Long; YAO; Yong-gang; YUAN; Guo-jun; NI; Bang-fa

    2013-01-01

    As the capital city of China,Beijing often suffers from hazy weather recently.In order to improve air quality,it is of great importance to perform source apportionment and source trajectory.A total of 140airborne particulate matter samples were collected at Xinzhen from May,2007 to July,2013 and their chemical compositions were analyzed by Particle Induced X-ray Emission(PIXE)and Energy Disperse-X

  16. Optical polarizing neutron devices designed for pulsed neutron sources

    Energy Technology Data Exchange (ETDEWEB)

    Takeda, M.; Kurahashi, K.; Endoh, Y. [Tohoku Univ, Sendai (Japan); Itoh, S. [National Lab. for High Energy Physics, Tsukuba (Japan)

    1997-09-01

    We have designed two polarizing neutron devices for pulsed cold neutrons. The devices have been tested at the pulsed neutron source at the Booster Synchrotron Utilization Facility of the National Laboratory for High Energy Physics. These two devices proved to have a practical use for experiments to investigate condensed matter physics using pulsed cold polarized neutrons.

  17. Neutron-emission measurements at a white neutron source

    Energy Technology Data Exchange (ETDEWEB)

    Haight, Robert C [Los Alamos National Laboratory

    2010-01-01

    Data on the spectrum of neutrons emittcd from neutron-induced reactions are important in basic nuclear physics and in applications. Our program studies neutron emission from inelastic scattering as well as fission neutron spectra. A ''white'' neutron source (continuous in energy) allows measurements over a wide range of neutron energies all in one experiment. We use the tast neutron source at the Los Alamos Neutron Science Center for incident neutron energies from 0.5 MeV to 200 MeV These experiments are based on double time-of-flight techniques to determine the energies of the incident and emitted neutrons. For the fission neutron measurements, parallel-plate ionization or avalanche detectors identify fission in actinide samples and give the required fast timing pulse. For inelastic scattering, gamma-ray detectors provide the timing and energy spectroscopy. A large neutron-detector array detects the emitted neutrons. Time-of-flight techniques are used to measure the energies of both the incident and emitted neutrons. Design considerations for the array include neutron-gamma discrimination, neutron energy resolution, angular coverage, segmentation, detector efficiency calibration and data acquisition. We have made preliminary measurements of the fission neutron spectra from {sup 235}U, {sup 238}U, {sup 237}Np and {sup 239}Pu. Neutron emission spectra from inelastic scattering on iron and nickel have also been investigated. The results obtained will be compared with evaluated data.

  18. UCN Source at an External Beam of Thermal Neutrons

    Directory of Open Access Journals (Sweden)

    E. V. Lychagin

    2015-01-01

    Full Text Available We propose a new method for production of ultracold neutrons (UCNs in superfluid helium. The principal idea consists in installing a helium UCN source into an external beam of thermal or cold neutrons and in surrounding this source with a solid methane moderator/reflector cooled down to ~4 K. The moderator plays the role of an external source of cold neutrons needed to produce UCNs. The flux of accumulated neutrons could exceed the flux of incident neutrons due to their numerous reflections from methane; also the source size could be significantly larger than the incident beam diameter. We provide preliminary calculations of cooling of neutrons. These calculations show that such a source being installed at an intense source of thermal or cold neutrons like the ILL or PIK reactor or the ESS spallation source could provide the UCN density 105 cm−3, the production rate 107 UCN/s−1. Main advantages of such an UCN source include its low radiative and thermal load, relatively low cost, and convenient accessibility for any maintenance. We have carried out an experiment on cooling of thermal neutrons in a methane cavity. The data confirm the results of our calculations of the spectrum and flux of neutrons in the methane cavity.

  19. Mechanical behavior of AISI 304SS determined by miniature test methods after neutron irradiation to 28 dpa

    Energy Technology Data Exchange (ETDEWEB)

    Rabenberg, Ellen M.; Jaques, Brian J. [Department of Materials Science and Engineering, Boise State University, 1910 University Dr., Boise, ID 83725 (United States); Center for Advanced Energy Studies, 995 University Blvd., Idaho Falls, ID 83401 (United States); Sencer, Bulent H. [Center for Advanced Energy Studies, 995 University Blvd., Idaho Falls, ID 83401 (United States); Idaho National Laboratory, Idaho Falls, ID 83415 (United States); Garner, Frank A. [Radiation Effects Consulting, 2003 Howell Ave., Richland, WA 99354 (United States); Freyer, Paula D. [Westinghouse Electric Company LLC, Pittsburgh, PA 15235 (United States); Okita, Taira [Research Into Artifacts Dept., Center for Engineering, University of Tokyo, Tokyo (Japan); Butt, Darryl P., E-mail: DarrylButt@BoiseState.edu [Department of Materials Science and Engineering, Boise State University, 1910 University Dr., Boise, ID 83725 (United States); Center for Advanced Energy Studies, 995 University Blvd., Idaho Falls, ID 83401 (United States)

    2014-05-01

    The mechanical properties of AISI 304 stainless steel irradiated for over a decade in the Experimental Breeder Reactor (EBR-II) were measured using miniature mechanical testing methods. The shear punch method was used to evaluate the shear strengths of the neutron-irradiated steel and a correlation factor was empirically determined to predict its tensile strength. The strength of the stainless steel slightly decreased with increasing irradiation temperature, and significantly increased with increasing dose until it saturated above approximately 5 dpa. An effective tensile strain hardening exponent was also obtained from the data which shows a relative decrease in ductility of steel with increased irradiation damage. Ferromagnetic measurements were used to observe and deduce the effects of the stress-induced austenite to martensite transformation as a result of shear punch testing.

  20. Design of a neutron source for calibration

    International Nuclear Information System (INIS)

    The neutron spectra produced by an isotopic neutron source located at the center of moderating media were calculated using Monte Carlo method in the aim to design a neutron source for calibration purposes. To improve the evaluation of the dosimetric quantities, is recommended to calibrate the radiation protection devices with calibrated neutron sources whose neutron spectra being similar to those met in practice. Here, a 239Pu-Be neutron source was inserted in H2O, D2O and polyethylene cylindrical moderators in order to produce neutron spectra that resembles spectra found in workplaces

  1. Beijing

    Institute of Scientific and Technical Information of China (English)

    WANGPEI

    2004-01-01

    As the nation's capital, Beijing hasunderstandably been positioned as China's political and cultural centel As the second largest economy among China's cities according to figures for 2003, Beijing also earns the title of an economic center. In the past two years Beijing has started to realize the indispensable value of finance for its overall economic development and set out to build a financial area in the city.

  2. Neutron sources for the medical use

    International Nuclear Information System (INIS)

    Recently encouraging results of the neutron radiation therapy have been obtained in clinical trials. In addition to the therapy, the neutrons are applied to the diagnosis besides the production of radioisotopes, that is, in-vivo activation analysis and neutron radiograph. In the medicine, high energy neutrons are effectively used. The necessary conditions, especially neutron source reactions, angular distributions, etc., and the neutron dosimetry including neutron kerma factors are discussed. Finally the requirements for neutron sources, their related problems and nuclear data are enumerated. (author)

  3. Compact ion source neutron generator

    Energy Technology Data Exchange (ETDEWEB)

    Schenkel, Thomas; Persaud, Arun; Kapadia, Rehan; Javey, Ali; Chang-Hasnain, Constance; Rangelow, Ivo; Kwan, Joe

    2015-10-13

    A neutron generator includes a conductive substrate comprising a plurality of conductive nanostructures with free-standing tips and a source of an atomic species to introduce the atomic species in proximity to the free-standing tips. A target placed apart from the substrate is voltage biased relative to the substrate to ionize and accelerate the ionized atomic species toward the target. The target includes an element capable of a nuclear fusion reaction with the ionized atomic species to produce a one or more neutrons as a reaction by-product.

  4. Uses of reactor neutrons for studying the microcomposition of materials

    International Nuclear Information System (INIS)

    Reactor neutrons constitute excellents 'probes' for exploring and measuring a wide range both of minor and trace constituents in solids and liquids with high sensitivity because of their transparency in materials. Nondestructive neutron prompt-gamma analysis (PGA) utilizing either cold or thermal neutrons, such as at JRR-3M, is compared and contrasted to the more common (delayed) instrumental neutron activation analysis (INAA) and epithermal NAA. Clearly PGA offers high sensitivity for selected elements: B, H, Cd and REE's in suitable matrices, and is therefore, complementary to INAA which is not as useful for them, or for Ni, Sn, Fe, C or N. Recent INAA applications in our laboratory that demonstrate some of the uniqueness of neutron methods include use of epithermal neutrons for small biological specimens to measure Cd, K, As, Zn and, multielemental INAA for environmental pollution studies. The latter involves large data sets of multielemental concentrations which are subjected to statistical multivariant factor analysis to reveal unknown or unsuspected quantitative relationships among groups of trace constituents. These patterns, or 'factors' are shown to be uniquely related to pollution sources and can be utilized to compute the relative source contributions at a given receptor site. (author)

  5. Irradiation testing of miniature fuel plates for the RERTR program. [Reduced Enrichment Research and Test Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Senn, R.L.; Martin, M.M.

    1981-07-01

    An irradiation test facility, which provides a test bed for irradiating a variety of miniature fuel plates (miniplates) for the Reduced Enrichment Research and Test Reactors (RERTR) program, has been placed into operation. These tests screen various candidate fuel materials on their suitability for replacing the highly enriched uranium fuel materials currently used by the world's test and research reactors with a lower enrichment fuel material, without significantly degrading reactor operating characteristics and power levels. The use of low uranium enrichment of about 20% /sup 235/U in place of highly enriched fuel for these reactors would reduce the potential for /sup 235/U diversion. The irradiation test facility, designated as HFED, is operating in core position E-7 in the Oak Ridge Research Reactor (ORR), a 30-MW water-moderated reactor. The miniplates will achieve burnups of up to approx. 2.2 x 10/sup 27/ fissions/m/sup 3/ of fuel.

  6. Measurements with a Pulsed and Modulated Source in a Reactor

    International Nuclear Information System (INIS)

    A generator with a neutron level variable in terms of any time factor has been developed by Philips Research Laboratories. Its practical use. in reactor physics has been demonstrated through a series of measurements carried out in the BRO2 reactor when subcritical. The stability of this generator, and the possibility of introducing sharp variations in the neutron intensity and of pulsing the flux or modulating it sinusoidally, makes it a very versatile instrument. It enables reactivity (ρ = Δk/β) and neutron lifetime (ℓ/β) to be determined by different independent methods. An exact comparison can be made of these methods since they can be employed without changing the conditions under which measurements are carried out. The following were determined: (1) ρ based on delayed neutrons, by a sudden reduction of neutron level, (2) ρ based on prompt neutrons by neutron pulses, (3) (ℓ/β) by a combination of (1) and (2) for 0.5$ < ρ < 2$; and (4) ℓ/β based on the transfer function of the reactor for a modulated source. The transfer functions for a reactivity oscillator and for a sinusoidally modulated source are discussed. It is shown that the measurement of ℓ/β is possible for 0.1 $ < ρ < 10 $ by using a modulated source. The same method also gives the reactivity on the basis of the ratio of prompt neutrons to delayed neutrons for an optimal frequency, practically independently of the data for delayed neutrons and of the value of ℓ/β. By accumulating a large number of cycles in the multi-channel analyser, better statistics for each method can be obtained. Since the neutron level from the generator is in fact sinusoidal, the response of the reactor may be integrated over each quarter of a period, as the measurement sequence is controlled by the generator; measurement time is then minimal. Observations recorded on a perforated tape are analysed by a digital computer

  7. 反应堆冷中子源的调试运行%Commissioning Operation of Reactor Based Cold Neutron Source Facility

    Institute of Scientific and Technical Information of China (English)

    兰晓华; 郑洲; 刘聪; 刘显坤

    2014-01-01

    14.5K cryogenic Helium is the refrigerating medium of the Cold Neutron Source ( CNS) Facility.Hy-drogen is liquefied from normal atmospheric temperature by the operation of CNS Refrigeration Cryogenic System (RCS).Liquid hydrogen overflows the Moderator Chamber .Thermal neutrons around liquid hydrogen deliver their energy to liquid hydrogen and then they become cold neutrons .Cold neutrons travel along the neutron guide tube and then arrive to the neutron spectrometers in the scattering hall .Stably operating of all CNS sub-systems is the base for obtaining cold neutron .And RCS debugging is the most complex one in all works .Trou-ble and problem solving during the commissioning operation of RCS is the primary coverage of this paper .%冷中子源装置利用14.5 K的低温氦气作为制冷剂。通过制冷系统的不停运转,将冷源氢系统内的氢液化。液态的氢充满着整个堆内部件系统的慢化剂室,使得其周围的热中子与液氢慢化剂进行能量交换,变为冷中子,再通过中子导管将冷中子输送到散射大厅各台谱仪上。冷源装置所有子系统调试运行的成功是获得冷中子的基础。在冷源调试中,氦制冷系统调试运行最为繁琐。着重介绍氦制冷系统在调试过程中遇到的问题及解决措施。

  8. Novel neutron focusing mirrors for compact neutron sources

    OpenAIRE

    Gubarev, M. V.; Zavlin, V. E.; Katz, R.; Resta, G.; Robertson, L; Crow, L.; Ramsey, B. D.; Khaykovich, Boris; Liu, DaZhi; Moncton, David E.

    2012-01-01

    We demonstrated neutron beam focusing and neutron imaging using axisymmetric optics, based on pairs of confocal ellipsoid and hyperboloid mirrors. Such systems, known as Wolter mirrors, are commonly used in x-ray telescopes. A system containing four nested Ni mirror pairs was implemented and tested by focusing a polychromatic neutron beam at the MIT Reactor and conducting an imaging experiment at HFIR. The major advantage of the Wolter mirrors is the possibility of nesting for large angular c...

  9. The spallation neutron source: New opportunities

    Indian Academy of Sciences (India)

    Ian S Anderson

    2008-11-01

    The spallation neutron source (SNS) facility became operational in the spring of 2006, and is now well on its way to become the world-leading facility for neutron scattering. Furthermore, the SNS and the HFIR reactor facility, newly outfitted with a brilliant cold source and guide hall, were brought together within a single Neutron Sciences Directorate at ORNL providing the opportunity to develop science and instrumentation programs which take advantage of the unique characteristics of each source. SNS and HFIR will both operate as scientific user facilities. Access to these facilities is being managed under an integrated proposal system, which also includes the Center for Nanophase Materials Sciences (CNMS) and the electron microscopes in the Shared Research Equipment (SHARE) program. Presently, SNS has three instruments operating in the user program and seven more will begin operations in 2008. When complete, the facility will accommodate 25 instruments enabling researchers from the United States and abroad to study materials science that forms the basis for new technologies in telecommunications, manufacturing, transportation, information technology, biotechnology, and health.

  10. Production, distribution and applications of californium-252 neutron sources

    International Nuclear Information System (INIS)

    The radioisotope 252Cf is routinely encapsulated into compact, portable, intense neutron sources with a 2.6-yr half-life. A source the size of a person's little finger can emit up to 1011 neutrons s-1. Californium-252 is used commercially as a reliable, cost-effective neutron source for prompt gamma neutron activation analysis (PGNAA) of coal, cement and minerals, as well as for detection and identification of explosives, land mines and unexploded military ordnance. Other uses are neutron radiography, nuclear waste assays, reactor start-up sources, calibration standards and cancer therapy. The inherent safety of source encapsulations is demonstrated by 30 yr of experience and by US Bureau of Mines tests of source survivability during explosions. The production and distribution center for the US Department of Energy (DOE) Californium Program is the Radiochemical Engineering Development Center (REDC) at Oak Ridge National Laboratory (ORNL). DOE sells 252Cf to commercial reencapsulators domestically and internationally. Sealed 252Cf sources are also available for loan to agencies and subcontractors of the US government and to universities for educational, research and medical applications. The REDC has established the Californium User Facility (CUF) for Neutron Science to make its large inventory of 252Cf sources available to researchers for irradiations inside uncontaminated hot cells. Experiments at the CUF include a land mine detection system, neutron damage testing of solid-state detectors, irradiation of human cancer cells for boron neutron capture therapy experiments and irradiation of rice to induce genetic mutations

  11. Production, distribution and applications of californium-252 neutron sources.

    Science.gov (United States)

    Martin, R C; Knauer, J B; Balo, P A

    2000-01-01

    The radioisotope 252Cf is routinely encapsulated into compact, portable, intense neutron sources with a 2.6-yr half-life. A source the size of a person's little finger can emit up to 10(11) neutrons s(-1). Californium-252 is used commercially as a reliable, cost-effective neutron source for prompt gamma neutron activation analysis (PGNAA) of coal, cement and minerals, as well as for detection and identification of explosives, land mines and unexploded military ordinance. Other uses are neutron radiography, nuclear waste assays, reactor start-up sources, calibration standards and cancer therapy. The inherent safety of source encapsulations is demonstrated by 30 yr of experience and by US Bureau of Mines tests of source survivability during explosions. The production and distribution center for the US Department of Energy (DOE) Californium Program is the Radiochemical Engineering Development Center (REDC) at Oak Ridge National Laboratory (ORNL). DOE sells 252Cf to commercial reencapsulators domestically and internationally. Sealed 252Cf sources are also available for loan to agencies and subcontractors of the US government and to universities for educational, research and medical applications. The REDC has established the Californium User Facility (CUF) for Neutron Science to make its large inventory of 252Cf sources available to researchers for irradiations inside uncontaminated hot cells. Experiments at the CUF include a land mine detection system, neutron damage testing of solid-state detectors, irradiation of human cancer cells for boron neutron capture therapy experiments and irradiation of rice to induce genetic mutations. PMID:11003521

  12. Fuel cycle for a fusion neutron source

    Energy Technology Data Exchange (ETDEWEB)

    Ananyev, S. S., E-mail: Ananyev-SS@nrcki.ru; Spitsyn, A. V., E-mail: spitsyn-av@nrcki.ru; Kuteev, B. V., E-mail: Kuteev-BV@nrcki.ru [National Research Center Kurchatov Institute (Russian Federation)

    2015-12-15

    The concept of a tokamak-based stationary fusion neutron source (FNS) for scientific research (neutron diffraction, etc.), tests of structural materials for future fusion reactors, nuclear waste transmutation, fission reactor fuel production, and control of subcritical nuclear systems (fusion–fission hybrid reactor) is being developed in Russia. The fuel cycle system is one of the most important systems of FNS that provides circulation and reprocessing of the deuterium–tritium fuel mixture in all fusion reactor systems: the vacuum chamber, neutral injection system, cryogenic pumps, tritium purification system, separation system, storage system, and tritium-breeding blanket. The existing technologies need to be significantly upgraded since the engineering solutions adopted in the ITER project can be only partially used in the FNS (considering the capacity factor higher than 0.3, tritium flow up to 200 m{sup 3}Pa/s, and temperature of reactor elements up to 650°C). The deuterium–tritium fuel cycle of the stationary FNS is considered. The TC-FNS computer code developed for estimating the tritium distribution in the systems of FNS is described. The code calculates tritium flows and inventory in tokamak systems (vacuum chamber, cryogenic pumps, neutral injection system, fuel mixture purification system, isotope separation system, tritium storage system) and takes into account tritium loss in the fuel cycle due to thermonuclear burnup and β decay. For the two facility versions considered, FNS-ST and DEMO-FNS, the amount of fuel mixture needed for uninterrupted operation of all fuel cycle systems is 0.9 and 1.4 kg, consequently, and the tritium consumption is 0.3 and 1.8 kg per year, including 35 and 55 g/yr, respectively, due to tritium decay.

  13. Fuel cycle for a fusion neutron source

    Science.gov (United States)

    Ananyev, S. S.; Spitsyn, A. V.; Kuteev, B. V.

    2015-12-01

    The concept of a tokamak-based stationary fusion neutron source (FNS) for scientific research (neutron diffraction, etc.), tests of structural materials for future fusion reactors, nuclear waste transmutation, fission reactor fuel production, and control of subcritical nuclear systems (fusion-fission hybrid reactor) is being developed in Russia. The fuel cycle system is one of the most important systems of FNS that provides circulation and reprocessing of the deuterium-tritium fuel mixture in all fusion reactor systems: the vacuum chamber, neutral injection system, cryogenic pumps, tritium purification system, separation system, storage system, and tritium-breeding blanket. The existing technologies need to be significantly upgraded since the engineering solutions adopted in the ITER project can be only partially used in the FNS (considering the capacity factor higher than 0.3, tritium flow up to 200 m3Pa/s, and temperature of reactor elements up to 650°C). The deuterium-tritium fuel cycle of the stationary FNS is considered. The TC-FNS computer code developed for estimating the tritium distribution in the systems of FNS is described. The code calculates tritium flows and inventory in tokamak systems (vacuum chamber, cryogenic pumps, neutral injection system, fuel mixture purification system, isotope separation system, tritium storage system) and takes into account tritium loss in the fuel cycle due to thermonuclear burnup and β decay. For the two facility versions considered, FNS-ST and DEMO-FNS, the amount of fuel mixture needed for uninterrupted operation of all fuel cycle systems is 0.9 and 1.4 kg, consequently, and the tritium consumption is 0.3 and 1.8 kg per year, including 35 and 55 g/yr, respectively, due to tritium decay.

  14. NEUTRONIC REACTOR HAVING LOCALIZED AREAS OF HIGH THERMAL NEUTRON DENSITIES

    Science.gov (United States)

    Newson, H.W.

    1958-06-01

    A nuclear reactor for the irradiation of materials designed to provide a localized area of high thermal neutron flux density in which the materials to be irradiated are inserted is described. The active portion of the reactor is comprised of a cubicle graphite moderator of about 25 feet in length along each axis which has a plurality of cylindrical channels for accommodatirg elongated tubular-shaped fuel elements. The fuel elements have radial fins for spacing the fuel elements from the channel walls, thereby providing spaces through which a coolant may be passed, and also to serve as a heatconductirg means. Ducts for accommnodating the sample material to be irradiated extend through the moderator material perpendicular to and between parallel rows of fuel channels. The improvement is in the provision of additional fuel element channels spaced midway between 2 rows of the regular fuel channels in the localized area surrounding the duct where the high thermal neutron flux density is desired. The fuel elements normally disposed in the channels directly adjacent the duct are placed in the additional channels, and the channels directly adjacent the duct are plugged with moderator material. This design provides localized areas of high thermal neutron flux density without the necessity of providing additional fuel material.

  15. Target technology of high energy neutron source

    International Nuclear Information System (INIS)

    As a facility of high energy neutron source for materials research and development, Fusion Materials Irradiation Test Facility (FMIT) is a strong candidate. The FMIT is designed to study the irradiation effect of fusion neutron on a fusion reactor materials. The FMIT generates a high-flux, high-energy neutron, which is produced in a stripping reaction by impinging a 3.5 MeV-0.1A beam of deuterons on a flowing lithium target. Target technology obtained in the FMIT will be useful for Energy Selective Neutron Irradiation Test Facility (ESNIT) and IFMIF of D-Li stripping reaction facility. In the first report (I), the flowing lithium target of the FMIT was reviewed, and some technical considerations in design were pointed out. In the second report (II), the target assembly and target material were proposed as the option of the HEDEL reference design of FMIT in order to improve the hazard and economy for the Li system: Firstly, the exchangeable target back wall and the measures to minimize the outside device damage in case of back wall breaking, and secondly, the option of molten fluoride salt as target material were proposed. (M.T.)

  16. Production, Distribution, and Applications of Californium-252 Neutron Sources

    Energy Technology Data Exchange (ETDEWEB)

    Balo, P.A.; Knauer, J.B.; Martin, R.C.

    1999-10-03

    The radioisotope {sup 252}Cf is routinely encapsulated into compact, portable, intense neutron sources with a 2.6-year half-life. A source the size of a person's little finger can emit up to 10{sup 11} neutrons/s. Californium-252 is used commercially as a reliable, cost-effective neutron source for prompt gamma neutron activation analysis (PGNAA) of coal, cement, and minerals, as well as for detection and identification of explosives, laud mines, and unexploded military ordnance. Other uses are neutron radiography, nuclear waste assays, reactor start-up sources, calibration standards, and cancer therapy. The inherent safety of source encapsulations is demonstrated by 30 years of experience and by U.S. Bureau of Mines tests of source survivability during explosions. The production and distribution center for the U. S Department of Energy (DOE) Californium Program is the Radiochemical Engineering Development Center (REDC) at Oak Ridge National Laboratory (ORNL). DOE sells The radioisotope {sup 252}Cf is routinely encapsulated into compact, portable, intense neutron sources with a 2.6- year half-life. A source the size of a person's little finger can emit up to 10 neutrons/s. Californium-252 is used commercially as a reliable, cost-effective neutron source for prompt gamma neutron activation analysis (PGNAA) of coal, cement, and minerals, as well as for detection and identification of explosives, laud mines, and unexploded military ordnance. Other uses are neutron radiography, nuclear waste assays, reactor start-up sources, calibration standards, and cancer therapy. The inherent safety of source encapsulations is demonstrated by 30 years of experience and by U.S. Bureau of Mines tests of source survivability during explosions. The production and distribution center for the U. S Department of Energy (DOE) Californium Program is the Radiochemical Engineering Development Center (REDC) at Oak Ridge National Laboratory(ORNL). DOE sells {sup 252}Cf to commercial

  17. Thermal hydraulic analysis of two-phase closed thermosyphon cooling system for new cold neutron source moderator of Breazeale research reactor at Penn State

    Science.gov (United States)

    Habte, Melaku

    A cold neutron source cooling system is required for the Penn State's next generation cold neutron source facility that can accommodate a variable heat load up to about ˜10W with operating temperature of about 28K. An existing cold neutron source cooling system operating at the University of Texas Cold Neutron Source (TCNS) facility failed to accommodate heat loads upwards of 4W with the moderator temperature reaching a maximum of 44K, which is the critical temperature for the operating fluid neon. The cooling system that was used in the TCNS cooling system was a two-phase closed thermosyphon with a reservoir (TPCTR). The reservoir containing neon gas is kept at room temperature. In this study a detailed thermal analysis of the fundamental operating principles of a TPCTR were carried out. A detailed parametric study of the various geometric and thermo-physical factors that affect the limits of the operational capacity of the TPCTR investigated. A CFD analysis is carried out in order to further refine the heat transfer analysis and understand the flow structure inside the thermosyphon and the two-phase nucleate boiling in the evaporator section of the thermosyphon. In order to help the new design, a variety of ways of increasing the operating range and heat removal capacity of the TPCTR cooling system were analyzed so that it can accommodate the anticipated heat load of 10W or more. It is found, for example, that doubling the pressure of the system will increase the capacity index zeta by 50% for a system with an initial fill ratio FR of 1. A decrease in cryorefrigeration performance angle increases the capacity index. For example taking the current condition of the TCNS system and reducing the angle from the current value of ˜700 by half (˜350) will increase the cooling power 300%. Finally based on detailed analytic and CFD analysis the best operating condition were proposed.

  18. A bright neutron source driven by relativistic transparency of solids

    Science.gov (United States)

    Roth, M.; Jung, D.; Falk, K.; Guler, N.; Deppert, O.; Devlin, M.; Favalli, A.; Fernandez, J.; Gautier, D. C.; Geissel, M.; Haight, R.; Hamilton, C. E.; Hegelich, B. M.; Johnson, R. P.; Kleinschmidt, A.; Merrill, F.; Schaumann, G.; Schoenberg, K.; Schollmeier, M.; Shimada, T.; Taddeucci, T.; Tybo, J. L.; Wagner, F.; Wender, S. A.; Wilde, C. H.; Wurden, G. A.

    2016-03-01

    Neutrons are a unique tool to alter and diagnose material properties and excite nuclear reactions with a large field of applications. It has been stated over the last years, that there is a growing need for intense, pulsed neutron sources, either fast or moderated neutrons for the scientific community. Accelerator based spallation sources provide unprecedented neutron fluxes, but could be complemented by novel sources with higher peak brightness that are more compact. Lasers offer the prospect of generating a very compact neutron source of high peak brightness that could be linked to other facilities more easily. We present experimental results on the first short pulse laser driven neutron source powerful enough for applications in radiography. For the first time an acceleration mechanism (BOA) based on the concept of relativistic transparency has been used to generate neutrons. This mechanism not only provides much higher particle energies, but also accelerated the entire target volume, thereby circumventing the need for complicated target treatment and no longer limited to protons as an intense ion source. As a consequence we have demonstrated a new record in laser-neutron production, not only in numbers, but also in energy and directionality based on an intense deuteron beam. The beam contained, for the first time, neutrons with energies in excess of 100 MeV and showed pronounced directionality, which makes then extremely useful for a variety of applications. The results also address a larger community as it paves the way to use short pulse lasers as a neutron source. They can open up neutron research to a broad academic community including material science, biology, medicine and high energy density physics as laser systems become more easily available to universities and therefore can complement large scale facilities like reactors or particle accelerators. We believe that this has the potential to increase the user community for neutron research largely.

  19. The physics experimental study for in-hospital neutron irradiator

    International Nuclear Information System (INIS)

    MNSRs (Miniature Neutron Source Reactor) are low power research reactors designed and manufactured by China Institute of Atomic Energy (CIAE). MNSRs are mainly used for NAA, training and teaching, testing of nuclear instrumentation. The first MNSR, the prototype MNSR, was put into operation in 1984, later, eight other MNSRs had been built both at home and abroad. For MNSRs, highly enriched uranium (90%) is used as the fuel material. The In-Hospital Neutron Irradiator (IHNI) is designed for Boron Neutron Capture Therapy (BNCT) based on Miniature Neutron Source Reactor(MNSR). On both sides of the reactor core, there are two neutron beams, one is thermal neutron beam, and the other opposite to the thermal beam, is epithermal neutron beam. A small thermal neutron beam is specially designed for the measurement of blood boron concentration by the prompt gamma neutron activation analysis (PGNAA). In this paper, the experimental results of critical mass worth of the top Be reflectors worth of the control rod, neutron flux distribution and other components worth were measured, the experiment was done on the Zero Power Experiment equipment of MNSR. (author)

  20. Fabrication of electron sources for a miniature scanning electron microscope

    Energy Technology Data Exchange (ETDEWEB)

    Chen, L

    1999-10-01

    The thesis describes the fabrication of a micro field emitter, which is designed as an electron source for application in a miniature scanning electron microscope. A micro field emitter has a cathode tip, insulating layer, metal gate, sandwich structure. This structure is fabricated on a silicon substrate. The tip normally has a tenth of a nanometer apex radius, and is separated from the surrounding gate electrode by a space, ranging from sub-micron to several microns in width, where a high electric field is formed. The micro field emitter is capable of delivering electrons into the surrounding vacuum. The emitted electron current is governed by the Fowler-Nordheim theory. A new fabrication technique for producing arrays of silicon nanotips with precise control of tip size and geometry, without the need for a conventional oxidation sharpening process, was developed. Gated micro field emitter arrays were fabricated using an electron beam evaporation technique. Also volcano shaped micro field emitters were fabricated using plasma enhanced chemical vapor deposition and etch back techniques. Silicon based gated field emitters with tungsten coating, silicon carbide coating and diamond like carbide coating were produced for evaluation. Field emission studies from these micro field emitters were carried out in ultra high vacuum systems. The current/voltage characteristic, emission current stability, field emission spatial distribution, and the field emission pressure dependence were investigated. Failures of devices were investigated and the failure mechanisms were drawn out. A feedback back control circuit was designed and built to control the field emission current fluctuation and stabilized emission current from the micro field emitters was obtained. Various suggestions for developing the entire miniature scanning electron microscope are made and future work for studying the micro field emitters is proposed. (author)

  1. Outline of spallation neutron source engineering

    Energy Technology Data Exchange (ETDEWEB)

    Watanabe, Noboru [Center for Neutron Science, Tokai Research Establishment, Japan Atomic Energy Research Institute, Tokai, Ibaraki (Japan)

    2001-01-01

    Slow neutrons such as cold and thermal neutrons are unique probes which can determine structures and dynamics of condensed matter in atomic scale. The neutron scattering technique is indispensable not only for basic sciences such as condensed matter research and life science, but also for basic industrial technology in 21 century. It is believed that to survive in the science-technology competition in 21 century would be almost impossible without neutron scattering. However, the intensity of neutrons presently available is much lower than synchrotron radiation sources, etc. Thus, R and D of intense neutron sources become most important. The High-Intensity Proton Accelerator Project is now being promoted jointly by Japan Atomic Energy Research Institute and High Energy Accelerator Research Organization, but there has so far been no good text which covers all the aspects of pulsed spallation neutron sources. The present review was prepare aiming at giving a better understanding on pulsed spallation neutron sources not only to neutron source researchers but also more widely to neutron scattering researchers and accelerator scientists in this field. The contents involve, starting from what is neutron scattering and what neutrons are necessary for neutron scattering, what is the spallation reaction, how to produce neutrons required for neutron scattering more efficiently, target-moderator-reflector neutronics and its engineering, shielding, target station, material issues, etc. The author have engaged in R and D of pulsed apallation neutron sources and neutron scattering research using them over 30 years. The present review is prepared based on the author's experiences with useful information obtained through ICANS collaboration and recent data from the JSNS (Japanese Spallation Neutron Source) design team. (author)

  2. Compositions and pollutant sources of haze in Beijing urban sites.

    Science.gov (United States)

    Wang, Junmei; Song, Yujun; Zuo, Jiangnan; Wu, Hongwen

    2016-05-01

    Haze from urban sites in Beijing was collected with a self-assembled electrostatic dust collector. The sizes and morphologies, thermal properties, and compositions of the particles in the haze were characterized by scanning electronic microscopy (SEM), differential scanning calorimetry (DSC), thermal gravimetric analysis (TGA), and X-ray photoelectric spectroscopy (XPS), respectively. Based on these results, the causes and pollutant sources of the chemicals in the haze were analyzed, and some countermeasures were further advanced to reduce the related pollutant sources. Graphical abstract ᅟ. PMID:26810665

  3. Compositions and pollutant sources of haze in Beijing urban sites.

    Science.gov (United States)

    Wang, Junmei; Song, Yujun; Zuo, Jiangnan; Wu, Hongwen

    2016-05-01

    Haze from urban sites in Beijing was collected with a self-assembled electrostatic dust collector. The sizes and morphologies, thermal properties, and compositions of the particles in the haze were characterized by scanning electronic microscopy (SEM), differential scanning calorimetry (DSC), thermal gravimetric analysis (TGA), and X-ray photoelectric spectroscopy (XPS), respectively. Based on these results, the causes and pollutant sources of the chemicals in the haze were analyzed, and some countermeasures were further advanced to reduce the related pollutant sources. Graphical abstract ᅟ.

  4. Activity report of the fusion neutronics source from April 1, 2001 to March 31, 2004

    International Nuclear Information System (INIS)

    The Fusion Neutronics Source (FNS) is an accelerator based 14 MeV neutron generator established in 1981. FNS is a powerful tool for neutronics research aiming the fusion reactor development such as neutron cross section measurements, integral experiments and blanket neutronics experiments. This report reviews the FNS activities in the period from April 1, 2001 to March 31, 2004, including collaboration with universities and other research institutes. The 35 papers are indexed individually. (J.P.N.)

  5. New applications of neutron noise theory in power reactor physics

    Energy Technology Data Exchange (ETDEWEB)

    Arzhanov, Vasiliy

    2000-04-01

    The present thesis deals with neutron noise theory as applied to three comparatively different topics (or problems) in power reactor physics. Namely they are: theoretical investigation of the possibility to use a newly proposed current-flux (C/F) detector in Pressurized Water Reactors (PWRs) for the localisation of anomalies; both definition and studies on the point kinetic and adiabatic approximations for the relatively recently proposed Accelerator Driven Systems (ADS); development of the general theory of linear reactor kinetics and neutron noise in systems with varying size. One important practical problem is to detect and localise a vibrating control rod pin. The significance comes from the operational experience which indicates that individual pins can execute excessive mechanical vibrations that may lead to damage. Such mechanical vibrations induce neutron noise that can be detected. While the detection is relatively easy, the localisation of a vibrating control rod is much more complicated because only one measuring position is available and one needs to have at least three measured quantities. Therefore it has currently been proposed that the fluctuations of the neutron current vector, called the current noise, can be used in addition to the scalar noise in reactor diagnostic problems. The thesis investigates the possibility of the localization of a vibrating control rod pin in a PWR control assembly by using the scalar neutron noise and the 2-D radial current noise as measured at one central point in the control assembly. An explicit localisation technique is elaborated in which the searched position is determined as the absolute minimum of a minimisation function. The technique is investigated in numerical simulations. The results of the simulation tests show the potential applicability of the method. By design accelerator-driven systems would operate in a subcritical mode with a strong external source. This calls for a revision of many concepts and

  6. Physics and technology of spallation neutron sources

    Energy Technology Data Exchange (ETDEWEB)

    Bauer, G.S.

    1998-08-01

    Next to fission and fusion, spallation is an efficient process for releasing neutrons from nuclei. Unlike the other two reactions, it is an endothermal process and can, therefore, not be used per se in energy generation. In order to sustain a spallation reaction, an energetic beam of particles, most commonly protons, must be supplied onto a heavy target. Spallation can, however, play an important role as a source of neutrons whose flux can be easily controlled via the driving beam. Up to a few GeV of energy, the neutron production is roughly proportional to the beam power. Although sophisticated Monte Carlo codes exist to compute all aspects of a spallation facility, many features can be understood on the basis of simple physics arguments. Technically a spallation facility is very demanding, not only because a reliable and economic accelerator of high power is needed to drive the reaction, but also, and in particular, because high levels of radiation and heat are generated in the target which are difficult to cope with. Radiation effects in a spallation environment are different from those commonly encountered in a reactor and are probably even more temperature dependent than the latter because of the high gas production rate. A commonly favored solution is the use of molten heavy metal targets. While radiation damage is not a problem in this case, except for the container, a number of other issues are discussed. (author)

  7. Physics and technology of spallation neutron sources

    International Nuclear Information System (INIS)

    Next to fission and fusion, spallation is an efficient process for releasing neutrons from nuclei. Unlike the other two reactions, it is an endothermal process and can, therefore, not be used per se in energy generation. In order to sustain a spallation reaction, an energetic beam of particles, most commonly protons, must be supplied onto a heavy target. Spallation can, however, play an important role as a source of neutrons whose flux can be easily controlled via the driving beam. Up to a few GeV of energy, the neutron production is roughly proportional to the beam power. Although sophisticated Monte Carlo codes exist to compute all aspects of a spallation facility, many features can be understood on the basis of simple physics arguments. Technically a spallation facility is very demanding, not only because a reliable and economic accelerator of high power is needed to drive the reaction, but also, and in particular, because high levels of radiation and heat are generated in the target which are difficult to cope with. Radiation effects in a spallation environment are different from those commonly encountered in a reactor and are probably even more temperature dependent than the latter because of the high gas production rate. A commonly favored solution is the use of molten heavy metal targets. While radiation damage is not a problem in this case, except for the container, a number of other issues are discussed. (author)

  8. Optimal Neutron Source and Beam Shaping Assembly for Boron Neutron Capture Therapy

    CERN Document Server

    Vujic, J L; Greenspan, E; Guess, S; Karni, Y; Kastenber, W E; Kim, L; Leung, K N; Regev, D; Verbeke, J M; Waldron, W L; Zhu, Y

    2003-01-01

    There were three objectives to this project: (1) The development of the 2-D Swan code for the optimization of the nuclear design of facilities for medical applications of radiation, radiation shields, blankets of accelerator-driven systems, fusion facilities, etc. (2) Identification of the maximum beam quality that can be obtained for Boron Neutron Capture Therapy (BNCT) from different reactor-, and accelerator-based neutron sources. The optimal beam-shaping assembly (BSA) design for each neutron source was also to e obtained. (3) Feasibility assessment of a new neutron source for NCT and other medical and industrial applications. This source consists of a state-of-the-art proton or deuteron accelerator driving and inherently safe, proliferation resistant, small subcritical fission assembly.

  9. Optimal Neutron Source and Beam Shaping Assembly for Boron Neutron Capture Therapy

    International Nuclear Information System (INIS)

    There were three objectives to this project: (1) The development of the 2-D Swan code for the optimization of the nuclear design of facilities for medical applications of radiation, radiation shields, blankets of accelerator-driven systems, fusion facilities, etc. (2) Identification of the maximum beam quality that can be obtained for Boron Neutron Capture Therapy (BNCT) from different reactor-, and accelerator-based neutron sources. The optimal beam-shaping assembly (BSA) design for each neutron source was also to e obtained. (3) Feasibility assessment of a new neutron source for NCT and other medical and industrial applications. This source consists of a state-of-the-art proton or deuteron accelerator driving and inherently safe, proliferation resistant, small subcritical fission assembly

  10. Optimal Neutron Source & Beam Shaping Assembly for Boron Neutron Capture Therapy

    Energy Technology Data Exchange (ETDEWEB)

    J. Vujic; E. Greenspan; W.E. Kastenber; Y. Karni; D. Regev; J.M. Verbeke, K.N. Leung; D. Chivers; S. Guess; L. Kim; W. Waldron; Y. Zhu

    2003-04-30

    There were three objectives to this project: (1) The development of the 2-D Swan code for the optimization of the nuclear design of facilities for medical applications of radiation, radiation shields, blankets of accelerator-driven systems, fusion facilities, etc. (2) Identification of the maximum beam quality that can be obtained for Boron Neutron Capture Therapy (BNCT) from different reactor-, and accelerator-based neutron sources. The optimal beam-shaping assembly (BSA) design for each neutron source was also to e obtained. (3) Feasibility assessment of a new neutron source for NCT and other medical and industrial applications. This source consists of a state-of-the-art proton or deuteron accelerator driving and inherently safe, proliferation resistant, small subcritical fission assembly.

  11. Small Angle Neutron Scattering instrument at Malaysian TRIGA reactor

    Energy Technology Data Exchange (ETDEWEB)

    Shukri Mohd; Razali Kassim; Zal Uyun Mahmood [Malaysian Inst. for Nuclear Technology Research (MINT), Bangi, Kajang (Malaysia); Shahidan Radiman

    1998-10-01

    The TRIGA MARK II Research reactor at the Malaysian Institute for Nuclear Research (MINT) was commissioned in July 1982. Since then various works have been performed to utilise the neutrons produced from this steady state reactor. One of the project involved the Small Angle Neutron Scattering (SANS). (author)

  12. Destructive analysis of neutron sources

    International Nuclear Information System (INIS)

    Fuel-liner reactions in Pu--Be neutron sources were examined. The source is contained in an outer jacket of 304 stainless steel and an inner Ta container incorporating a TIG welded Ta plug. Small cracks were observed in some of the outer stainless steel containers as well as in some of the tantalum inner liners. Major cracking was observed as well as penetration of the reaction product through the tantalum sidewalls in two sources. High temperatures aided and accelerated the degradation and ultimate failure of the tantalum inner liner. Traces of beryllium metal as indicated from x-ray results of the fuel and large concentration gradients between tantalum and plutonium as shown in microprobe analysis were found to exist. The fuel was inhomogeneous in nature and the data suggest the possibility of tantalum-beryllium compounds, free unreacted plutonium, and potentially a ternary phase of tantalum, beryllium, plutonium as being present in the fuel

  13. Monochromatic Neutron Tomography Using 1-D PSD Detector at Low Flux Research Reactor

    Science.gov (United States)

    Ashari, N. Abidin; Saleh, J. Mohamad; Abdullah, M. Zaid; Mohamed, A. Aziz; Azman, A.; Jamro, R.

    2008-03-01

    This paper describes the monochromatic neutron tomography experiment using the 1-D Position Sensitive Neutron Detector (PSD) located at Nuclear Malaysia TRIGA MARK II Research reactor. Experimental work was performed using monochromatic neutron source from beryllium filter and HOPG crystal monochromator. The principal main aim of this experiment was to test the detector efficiency, image reconstruction algorithm and the usage of 0.5 nm monochromatic neutrons for the neutron tomography setup. Other objective includes gathering important parameters and features to characterize the system.

  14. 医院中子照射器Ⅰ型堆热中子束流孔道等效平面源的模拟计算%Numerical calculation for the equivalent surface source of the thermal neutron duct of in-hospital neutron irradiator mark 1 reactor

    Institute of Scientific and Technical Information of China (English)

    朱养妮; 江新标; 赵柱民; 张良; 周永茂

    2012-01-01

    采用蒙特卡罗程序MCNP模拟计算了医院中子照射器Ⅰ型堆(IHNI-1)热中子束流孔道出口处的等效平面源.对B堆芯进行了临界搜索计算,模拟计算了热中子束流孔道及出口处中子、γ的束流参数,应用等效平面源模型建立了BNCT等效中子、γ平面源.为人体头颅等效模型剂量分布的快速计算提供了较为可靠的平面源.%Numerical calculation for the equivalent surface source of the thermal neutron duct of in-hospital neutron irradiator mark 1 (IHNI-1) reactor is carried out using MCNP Monte Carlo code. Cold clean criticality of B-core is searched. Neutron beam parameters at the exit of thermal neutron duct are calculated. Equivalent neutron and -y surface sources for BNCT are built using equivalent surface source model. And these sources are reliable to calculate absorbed dose distribution in equivalent model of head quickly.

  15. Expectation for energy selective neutron source based on the current neutron irradiation study of materials

    International Nuclear Information System (INIS)

    For an effective utilization of superior characteristics of the energy selective high energy neutron source, a consideration was made. Electron irradiation with high voltage electron microscopes (HVEM), D-T fusion neutron irradiation with rotating target neutron source (RTNS-II), and fission neutron irradiation with fission reactors were referred. The expected Energy Selective Neutron Source (ESNS) were compared with different types of irradiation facilities in regard to energy spectrum, flux stability, temperature control, and possibility of in-situ experiments. The excellent performance of HVEM electron irradiation, and of RTNS-II D-T fusion neutron irradiation was exemplified. The possibility of extending these excellent performances to the future ESNS experiment was discussed. Difficulties in the neutron irradiation experiment with fission reactors were exemplified. Shrinkage and growth of these difficulties in the ESNS experiment was discussed. Expected advantage and limitation of the ESNS was evaluated. Finally the positioning of ESNS was made, and the importance of its complementality with other facilities was pointed out. (M.T.)

  16. Measurement of the Syrian MNSR delayed neutron fraction and neutron generation time by noise analysis

    Energy Technology Data Exchange (ETDEWEB)

    Khamis, I. E-mail: ikhamis@aec.org.sy; Hainoun, A.; Suleiman, W

    2003-02-01

    Delayed neutron fraction {beta} and prompt neutron generation time {lambda} were determined for the Miniature Neutron Source Reactor of Syria using noise analysis technique. Small reactivity perturbations, step-wise and impulse in time, were introduced into the reactor at low power level i.e. zero-power. Power and reactivity versus time were obtained. Using the generalized least square algorithm and transfer function analysis, measurement of both the delayed neutron fraction and the neutron generation time were made. The MNSR values obtained for the prompt neutron generation time and delayed neutron fraction are 78.3{+-}1.3 {mu}s and 7.94{+-}0.11x10{sup -3} respectively. Both measured values of {beta} and {lambda} were found to be very consistent with previously measured and calculated ones reported in the Safety Analysis Report.

  17. Measurement of the Syrian MNSR delayed neutron fraction and neutron generation time by noise analysis

    International Nuclear Information System (INIS)

    Delayed neutron fraction β and prompt neutron generation time Λ were determined for the Miniature Neutron Source Reactor of Syria using noise analysis technique. Small reactivity perturbations, step-wise and impulse in time, were introduced into the reactor at low power level i.e. zero-power. Power and reactivity versus time were obtained. Using the generalized least square algorithm and transfer function analysis, measurement of both the delayed neutron fraction and the neutron generation time were made. The MNSR values obtained for the prompt neutron generation time and delayed neutron fraction are 78.3±1.3 μs and 7.94±0.11x10-3 respectively. Both measured values of β and Λ were found to be very consistent with previously measured and calculated ones reported in the Safety Analysis Report

  18. Measurement of the Syrian MNSR delayed neutron fraction and neutron generation time by noise analysis

    International Nuclear Information System (INIS)

    Delayed neutron fraction beta and prompt neutron generation time LAMBDA were determined for the Miniature Neutron Source Reactor of Syria using noise analysis technique. Small reactivity perturbations, step-wise and impulse in time, were introduced into the reactor at low power level i.e. zero-power. Power and reactivity versus time were obtained. Using the generalized least square algorithm and transfer function analysis, measurement of both the delayed neutron fraction and the neutron generation time were made. The MNSR values obtained for the prompt neutron generation time and delayed neutron fraction are 78.3+-1.3 mu s and 7.94+-0.11x10 sup - sup 3 respectively. Both measured values of beta and LAMBDA were found to be very consistent with previously measured and calculated once reported in the Safety Analysis Report. (author)

  19. The Frankfurt neutron source FRANZ

    Science.gov (United States)

    Alzubaidi, Suha; Bartz, Ulrich; Basten, Markus; Bechtold, Alexander; Chau, Long Phi; Claessens, Christine; Dinter, Hannes; Droba, Martin; Fix, Christopher; Hähnel, Hendrik; Heilmann, Manuel; Hinrichs, Ole; Huneck, Simon; Klump, Batu; Lotz, Marcel; Mäder, Dominik; Meusel, Oliver; Noll, Daniel; Nowottnick, Tobias; Obermayer, Marcus; Payir, Onur; Petry, Nils; Podlech, Holger; Ratzinger, Ulrich; Schempp, Alwin; Schmidt, Stefan; Schneider, Philipp; Seibel, Anja; Schwarz, Malte; Schweizer, Waldemar; Volk, Klaus; Wagner, Christopher; Wiesner, Christoph

    2016-05-01

    A 2MeV proton beam will produce a quasi-Maxwellian neutron spectrum of around 30 keV by the 7Li(p, n)7Be reaction. The experiments are mainly focused on the measurement of differential neutron capture cross sections relevant for the astrophysical s-process in nuclear synthesis. Moreover, proton capture cross sections for the astrophysical p-process can be measured directly with the proton beam. For an efficient time of flight measurement of the neutron energies along the 0.7 m long drift from the Li-target to the sample, 1ns short, intense proton pulses are needed at the target. Additionally, to reach 107 n/cm2/s at the sample, a pulse repetition rate of 250 kHz is intended. After completion and successful running in, FRANZ will become a user facility with internal and external users. The 120 kV injector terminal and the 200mA proton source as well as the low-energy beam transport section and the FRANZ cave have been realized successfully. The 1.9 MV RF accelerator consists of a combined 4-Rod-RFQ/IH-DTL-resonator and is in the RF tuning and power testing phase. The 2 MeV transport and rebuncher section is ready for installation. In a first step FRANZ will offer experimental areas for neutron activation experiments and for proton beam experiments, as mentioned above. From the accelerator physics point of view, FRANZ will be an excellent facility for high current beam investigations and for beam wall interaction studies.

  20. Theory of neutron slowing down in nuclear reactors

    CERN Document Server

    Ferziger, Joel H; Dunworth, J V

    2013-01-01

    The Theory of Neutron Slowing Down in Nuclear Reactors focuses on one facet of nuclear reactor design: the slowing down (or moderation) of neutrons from the high energies with which they are born in fission to the energies at which they are ultimately absorbed. In conjunction with the study of neutron moderation, calculations of reactor criticality are presented. A mathematical description of the slowing-down process is given, with particular emphasis on the problems encountered in the design of thermal reactors. This volume is comprised of four chapters and begins by considering the problems

  1. Nuclear and dosimetric features of an isotopic neutron source

    Science.gov (United States)

    Vega-Carrillo, H. R.; Hernández-Dávila, V. M.; Rivera, T.; Sánchez, A.

    2014-02-01

    A multisphere neutron spectrometer was used to determine the features of a 239PuBe neutron source that is used to operate the ESFM-IPN Subcritical Reactor. To determine the source main features it was located a 100 cm from the spectrometer which was a 6LiI(Eu) scintillator and 2, 3, 5, 8, 10 and 12 in.-diameter polyethylene spheres. Count rates obtained with the spectrometer were unfolded using the NSDUAZ code and neutron spectrum, total fluence, and ambient dose equivalent were determined. A Monte Carlo calculation was carried out to estimate the spectrum and integral features being less than values obtained experimentally due to the presence of 241Pu in the Pu used to fabricate the source. Actual neutron yield and the mass fraction of 241Pu was estimated.

  2. Studies and modeling of cold neutron sources; Etude et modelisation des sources froides de neutron

    Energy Technology Data Exchange (ETDEWEB)

    Campioni, G

    2004-11-15

    With the purpose of updating knowledge in the fields of cold neutron sources, the work of this thesis has been run according to the 3 following axes. First, the gathering of specific information forming the materials of this work. This set of knowledge covers the following fields: cold neutron, cross-sections for the different cold moderators, flux slowing down, different measurements of the cold flux and finally, issues in the thermal analysis of the problem. Secondly, the study and development of suitable computation tools. After an analysis of the problem, several tools have been planed, implemented and tested in the 3-dimensional radiation transport code Tripoli-4. In particular, a module of uncoupling, integrated in the official version of Tripoli-4, can perform Monte-Carlo parametric studies with a spare factor of Cpu time fetching 50 times. A module of coupling, simulating neutron guides, has also been developed and implemented in the Monte-Carlo code McStas. Thirdly, achieving a complete study for the validation of the installed calculation chain. These studies focus on 3 cold sources currently functioning: SP1 from Orphee reactor and 2 other sources (SFH and SFV) from the HFR at the Laue Langevin Institute. These studies give examples of problems and methods for the design of future cold sources.

  3. Research reactors as sources of atmospheric radioxenon

    International Nuclear Information System (INIS)

    Radioxenon emissions of the TRIGA Mark II research reactor in Vienna were investigated with respect to a possible impact on the verification of the Comprehensive Nuclear Test-Ban-Treaty. Using the Swedish Automatic Unit for Noble Gas Acquisition (SAUNA II), five radioxenon isotopes 125Xe, 131mXe, 133mXe, 133Xe and 135Xe were detected, of which 125Xe is solely produced by neutron capture in stable atmospheric 124Xe and hence acts as an indicator for neutron activation processes. The other nuclides are produced in both fission and neutron capture reactions. The detected activity concentrations ranged from 0.0010 to 190 Bq/m3. The source of the radioxenon is not yet fully clarified, but it could be micro-cracks in the fuel cladding, fission of 235U contaminations on the outside of the fuel elements or neutron activation of atmospheric Xe. Neutron deficient 125Xe with its highly complex decay scheme was seen for the first time in a SAUNA system. In many experiments the activity ratios of the radioxenon nuclides carry the signature of nuclear explosions, if 131mXe is omitted. Only if 131mXe is included into the calculations of the isotopic activity ratios, the majority of the measurements revealed a 'civil' signature (typical for a NPP). A significant contribution of the TRIGA Vienna to the global or European radioxenon inventory can be excluded. Due to the very low activities, the emissions are far below any concern for human health. (author)

  4. Reactor neutron activation for multielemental analysis

    International Nuclear Information System (INIS)

    Neutron Activation Analysis using single comparator (K0 NAA method) has been used for obtaining multielemental profiles in a variety of matrices related to environment. Gold was used as the comparator. Neutron flux was characterised by determining f, the epithermal to thermal neutron flux ratio and cc, the deviation from ideal shape of the neutron spectrum. The f and a were determined in different irradiation positions in APSARA reactor, PCF position in CIRUS reactor and tray rod position in Dhruva reactor using both cadmium cut off and multi isotope detector methods. High resolution gamma ray spectrometry was used for radioactive assay of the activation products. This technique is being used for multielement analysis in a variety of matrices like lake sediments, sea nodules and crusts, minerals, leaves, cereals, pulses, leaves, water and soil. Elemental profiles of the sediments corresponding to different depths from Nainital lake were determined and used to understand the history of natural absorption/desorption pattern of the previous 160 years. Ferromanganese crusts from different locations of Indian Ocean were analysed with a view to studying the distribution of some trace elements along with Fe and Mn. Variation of Mn/Fe ratio was used to identify the nature of the crusts as hydrogenous or hydrothermal. Fe-rich and Fe-depleted nodules from Indian Ocean were analysed to understand the REE patterns and it is proposed that REE-Th associated minerals could be the potential Th contributors to the sea water and thus reached ferromanganese nodules. Dolomites (unaltered and altered), two types of serpentines and intrusive rock dolerite from the asbestos mines of Cuddapah basin were analysed for major, minor and trace elements. The elemental concentrations are used for distinguishing and characterising these minerals. From our investigations, it was concluded that both dolomite and dolerite contribute elements in the serpentinisation process. Chemical neutron

  5. Investigation of neutron distribution in training reactor VR-1

    International Nuclear Information System (INIS)

    The VR-1 training reactor is a pool-type light-water reactor with the low-enriched uranium and maximum thermal power of 1 kW. The reactor is mainly used for students' training. The training is aimed to areas such as the reactor physics, neutronics, dosimetry, nuclear safety and I and C systems. Since neutron flux in the VR-1 core is well measured, this work focuses on one part of the reactor - its Radial experimental Channel (RC). This paper deals with the measurement of the neutron distribution by means of gold-foil neutron-activation technique and continual measurement with 3He-filled detector. Obtained experimental results were verified with the simulation in the Monte-Carlo N-Particle Transport Code. Results and conclusions from this paper will be used for further investigation of neutrons and their spatial distribution inside the low-power training reactor. Also, the data obtained in this paper can be used as a basis for future detailed measurements of neutron flux and its distribution in other hard accessible areas inside the reactor. The paper gives a simple theoretical introduction concerning neutron measurement procedures and available techniques in this field, which is particularly important for improving training courses and a content of offered experiments in the VR-1 reactor. (author)

  6. Slow neutron leakage spectra from spallation neutron sources

    International Nuclear Information System (INIS)

    An efficient technique is described for Monte Carlo simulation of neutron beam spectra from target-moderator-reflector assemblies typical of pulsed spallation neutron sources. The technique involves the scoring of the transport-theoretical probability that a neutron will emerge from the moderator surface in the direction of interest, at each collision. An angle-biasing probability is also introduced which further enhances efficiency in simple problems. These modifications were introduced into the VIM low energy neutron transport code, representing the spatial and energy distributions of the source neutrons approximately as those of evaporation neutrons generated through the spallation process by protons of various energies. The intensity of slow neutrons leaking from various reflected moderators was studied for various neutron source arrangements. These include computations relating to early measurements on a mockup-assembly, a brief survey of moderator materials and sizes, and a survey of the effects of varying source and moderator configurations with a practical, liquid metal cooled uranium source Wing and slab, i.e., tangential and radial moderator arrangements, and Be vs CH2 reflectors are compared. Results are also presented for several complicated geometries which more closely represent realistic arrangements for a practical source, and for a subcritical fission multiplier such as might be driven by an electron linac. An adaptation of the code was developed to enable time dependent calculations, and investigated the effects of the reflector, decoupling and void liner materials on the pulse shape

  7. Development opportunities for small and medium scale accelerator driven neutron sources. Proceedings of a technical meeting

    International Nuclear Information System (INIS)

    Neutron applications in the life sciences will be a rapidly growing research area in the near future, as neutrons can provide unique information on the reaction dynamics of complex biomolecular systems, complementing other analytical techniques such as electron microscopy, X rays and nuclear magnetic resonance. Small and medium power spallation neutron sources will become more important, as many small neutron producing research reactors are being phased out. Recent developments in accelerator technology have made it possible to produce useful neutron fluxes at accelerator facilities suitable for universities and industrial laboratories. In addition to basic research these alternative neutron sources will be important for educational and training purposes. In a wider perspective this technology should make it possible to introduce neutron research and applications to industrial and national research centres in IAEA Member States that are unable to afford a high energy spallation neutron source and have no access to a research reactor

  8. The neutron utilization and promotion program of TRR-II research reactor project in Taiwan

    International Nuclear Information System (INIS)

    The objective of the Taiwan research reactor system improvement and utilization promotion project is to reconstruct the old Taiwan research reactor (TRR), which was operated by the Institute of Nuclear Energy Research (INER) between 1973 and 1988, into a multi-purpose medium flux research reactor (TRR-II). The project started in 1998, and the new reactor is scheduled to have its first critical in June of 2006. The estimated maximum unperturbed thermal neutron flux (E14 n/cm2sec, and it is about one order of magnitude higher than other operating research reactors in Taiwan. The new reactor will equip with secondary neutron sources to provide neutrons with different energies, which will be an essential tool for advanced material researches in Taiwan. One of the major tasks of TRR-II project is to promote domestic utilization of neutrons generated at TRR-II. The traditional uses of neutrons in fuel/material research, trace element analysis, and isotope production has been carried out at INER for many years. On the other hand, it is obvious that promotions of neutron spectrometric technique will be a major challenge for the project team. The limited neutron flux from operating research reactors had discouraged domestic users in developing neutron spectrometric technique for many years, and only few researchers in Taiwan are experienced in using spectrometers. It is important for the project team to encourage domestic researchers to use neutron spectrometers provided by TRR-II as a tool for their future researches in various fields. This paper describes the current status of TRR-II neutron utilization and promotion program. The current status and future plans for important issues such as staff recruiting, personnel training, international collaboration, and promotion strategy will be described. (orig.)

  9. Intense neutron sources for cancer treatment

    International Nuclear Information System (INIS)

    Significant progress has been made in the development of small, solid-target, pulsed neutron sources for nuclear weapons applications. The feasibility of using this type of neutron source for cancer treatment is discussed. Plans for fabrication and testing of such a source is briefly described

  10. Neutron Beam Characterization for Neutron Radiography Facility at the Thai Research Reactor TRR-1/M1

    International Nuclear Information System (INIS)

    The aim of this research is to characterize the present status of neutron beam coming out from the reactor core of Thai Research Reactor TRR-1/M1 through neutron radiography facility. In this study, the neutron beam profiles at different positions along the beam exit were recorded using digital imaging devices. In addition, thin foil activation technique, with and without cadmium cover, was employed to determine thermal neutron flux and Cd ratio. An acrylic step wedge was exposed to neutron at different time. In parallel to image construction, neutron detection was carried out using a BF3 gas-filled detector. Then, the image intensities at particular thicknesses were normalized by neutron counts from the BF3 detector to determine relative neutron intensity. The obtained information of neutron beam characterization will be useful not only for monitoring the present status of neutron radiography facility but also for determining the optimum exposure conditions for particular samples in the future.

  11. Neutronic studies of the coupled moderators for spallation neutron sources

    Institute of Scientific and Technical Information of China (English)

    Yin Wen; Liang Jiu-Qing

    2005-01-01

    We investigate the neutronic performance of coupled moderators to be implemented in spallation neutron sources by Monte-Carlo simulation and give the slow neutron spectra for the cold and thermal moderators. CH4 moderator can provide slow neutrons with highly desirable characteristics and will be used in low-power spallation neutron soureces. The slow neutron intensity extracted from different angles has been calculated. The capability of moderation of liquid H2 is lower than H2O and liquid CH4 due to lower atomic number density of hydrogen but we can compensate for this disadvantage by using a premoderator. The H2O premoderator of 2cm thickness can reduce the heat deposition in the cold moderator by about 33% without spoiling the neutron pulse.

  12. PGNAA neutron source moderation setup optimization

    CERN Document Server

    Zhang, Jinzhao

    2013-01-01

    Monte Carlo simulations were carried out to design a prompt {\\gamma}-ray neutron activation analysis (PGNAA) thermal neutron output setup using MCNP5 computer code. In these simulations the moderator materials, reflective materials and structure of the PGNAA 252Cf neutrons of thermal neutron output setup were optimized. Results of the calcuations revealed that the thin layer paraffin and the thick layer of heavy water moderated effect is best for 252Cf neutrons spectrum. The new design compared with the conventional neutron source design, the thermal neutron flux and rate were increased by 3.02 times and 3.27 times. Results indicate that the use of this design should increase the neutron flux of prompt gamma-ray neutron activation analysis significantly.

  13. Materials and neutronic research at the Low Energy Neutron Source

    Science.gov (United States)

    Baxter, David V.

    2016-04-01

    In the decade since the Low Energy Neutron Source (LENS) at Indiana University Center for Exploration of Energy and Matter (CEEM) produced its first neutrons, the facility has made important contributions to the international neutron scattering community. LENS employs a 13MeV proton beam at up to 4kW beam power onto one of two Be targets to produce neutrons for research in fields ranging from radiation effects in electronics to studies of the structure of fluids confined in nanoporous materials. The neutron source design at the heart of LENS facilitates relatively rapid hands-on access to most of its components which provides a foundation for a research program in experimental neutronics and affords numerous opportunities for novel educational experiences. We describe in some detail a number of the unique capabilities of this facility.

  14. New sources and instrumentation for neutron science

    Science.gov (United States)

    Gil, Alina

    2011-04-01

    Neutron-scattering research has a lot to do with our everyday lives. Things like medicine, food, electronics, cars and airplanes have all been improved by neutron-scattering research. Neutron research also helps scientists improve materials used in a multitude of different products, such as high-temperature superconductors, powerful lightweight magnets, stronger, lighter plastic products etc. Neutron scattering is one of the most effective ways to obtain information on both, the structure and the dynamics of condensed matter. Most of the world's neutron sources were built decades ago, and although the uses and demand for neutrons have increased throughout the years, few new sources have been built. The new construction, accelerator-based neutron source, the spallation source will provide the most intense pulsed neutron beams in the world for scientific research and industrial development. In this paper it will be described what neutrons are and what unique properties make them useful for science, how spallation source is designed to produce neutron beams and the experimental instruments that will use those beams. Finally, it will be described how past neutron research has affected our everyday lives and what we might expect from the most exciting future applications.

  15. ANSL-V: ENDF/B-V based multigroup cross-section libraries for Advanced Neutron Source (ANS) reactor studies. Supplement 1

    International Nuclear Information System (INIS)

    The original ANSL-V cross-section libraries (ORNL-6618) were developed over a period of several years for the physics analysis of the ANS reactor, with little thought toward including the materials commonly needed for shielding applications. Materials commonly used for shielding applications include calcium barium, sulfur, phosphorous, and bismuth. These materials, as well as 6Li, 7Li, and the naturally occurring isotopes of hafnium, have been added to the ANSL-V libraries. The gamma-ray production and gamma-ray interaction cross sections were completely regenerated for the ANSL-V 99n/44g library which did not exist previously. The MALOCS module was used to collapse the 99n/44g coupled library to the 39n/44g broad- group library. COMET was used to renormalize the two-dimensional (2- D) neutron matrix sums to agree with the one-dimensional (1-D) averaged values. The FRESH module was used to adjust the thermal scattering matrices on the 99n/44g and 39n/44g ANSL-V libraries. PERFUME was used to correct the original XLACS Legendre polynomial fits to produce acceptable distributions. The final ANSL-V 99n/44g and 39n/44g cross-section libraries were both checked by running RADE. The AIM module was used to convert the master cross-section libraries from binary coded decimal to binary format (or vice versa)

  16. ANSL-V: ENDF/B-V based multigroup cross-section libraries for Advanced Neutron Source (ANS) reactor studies. Supplement 1

    Energy Technology Data Exchange (ETDEWEB)

    Wright, R.Q.; Renier, J.P.; Bucholz, J.A.

    1995-08-01

    The original ANSL-V cross-section libraries (ORNL-6618) were developed over a period of several years for the physics analysis of the ANS reactor, with little thought toward including the materials commonly needed for shielding applications. Materials commonly used for shielding applications include calcium barium, sulfur, phosphorous, and bismuth. These materials, as well as {sup 6}Li, {sup 7}Li, and the naturally occurring isotopes of hafnium, have been added to the ANSL-V libraries. The gamma-ray production and gamma-ray interaction cross sections were completely regenerated for the ANSL-V 99n/44g library which did not exist previously. The MALOCS module was used to collapse the 99n/44g coupled library to the 39n/44g broad- group library. COMET was used to renormalize the two-dimensional (2- D) neutron matrix sums to agree with the one-dimensional (1-D) averaged values. The FRESH module was used to adjust the thermal scattering matrices on the 99n/44g and 39n/44g ANSL-V libraries. PERFUME was used to correct the original XLACS Legendre polynomial fits to produce acceptable distributions. The final ANSL-V 99n/44g and 39n/44g cross-section libraries were both checked by running RADE. The AIM module was used to convert the master cross-section libraries from binary coded decimal to binary format (or vice versa).

  17. Plans for an Ultra Cold Neutron source at Los Alamos

    Energy Technology Data Exchange (ETDEWEB)

    Seestrom, S.J.; Bowles, T.J.; Hill, R.; Greene, G.L. [Los Alamos National Lab., NM (United States)

    1996-08-01

    Ultra Cold Neutrons (UCN) can be produced at spallation sources using a variety of techniques. To date the technique used has been to Bragg scatter and Doppler shift cold neutrons into UCN from a moving crystal. This is particularly applicable to short-pulse spallation sources. We are presently constructing a UCN source at LANSCE using method. In addition, large gains in UCN density should be possible using cryogenic UCN sources. Research is under way at Gatchina to demonstrate technical feasibility of be a frozen deuterium source. If successful, a source of this type could be implemented at future spallation source, such as the long pulse source being planned at Los Alamos, with a UCN density that may be two orders of magnitude higher than that presently available at reactors. (author)

  18. Dose estimations of fast neutrons from a nuclear reactor by micronuclear yields in onion seedlings.

    Science.gov (United States)

    Fujikawa, K; Endo, S; Itoh, T; Yonezawa, Y; Hoshi, M

    1999-12-01

    Irradiations of onion seedlings with fission neutrons from bare, Pb-moderated, and Fe-moderated 252Cf sources induced micronuclei in the root-tip cells at similar rates. The rate per cGy averaged for the three sources, , was 19 times higher than rate induced by 60Co gamma-rays. When neutron doses, Dn, were estimated from frequencies of micronuclei induced in onion seedlings after exposure to neutron-gamma mixed radiation from a 1 W nuclear reactor, using the reciprocal of as conversion factor, resulting Dn values agreed within 10% with doses measured with paired ionizing chambers. This excellent agreement was achieved by the high sensitivity of the onion system to fast neutrons relative to gamma-rays and the high contribution of fast neutrons to the total dose of mixed radiation in the reactor's field.

  19. Dose estimations of fast neutrons from a nuclear reactor by micronuclear yields in onion seedlings

    International Nuclear Information System (INIS)

    Irradiations of onion seedlings with fission neutrons from bare, Pb-moderated, and Fe-moderated 252Cf sources induced micronuclei in the root-tip cells at similar rates. The rate per cGy averaged for the three sources, n>, was 19 times higher than rate induced by 60Co γ-rays. When neutron doses, Dn, were estimated from frequencies of micronuclei induced in onion seedlings after exposure to neutron-γmixed radiation from a 1 W nuclear reactor, using the reciprocal of n> as conversion factor, resulting Dn values agreed within 10% with doses measured with paired ionizing chambers. This excellent agreement was achieved by the high sensitivity of the onion system to fast neutrons relative to γ-rays and the high contribution of fast neutrons to the total dose of mixed radiation in the reactor's field. (author)

  20. Research opportunities with compact accelerator-driven neutron sources

    Science.gov (United States)

    Anderson, I. S.; Andreani, C.; Carpenter, J. M.; Festa, G.; Gorini, G.; Loong, C.-K.; Senesi, R.

    2016-10-01

    Since the discovery of the neutron in 1932 neutron beams have been used in a very broad range of applications, As an aging fleet of nuclear reactor sources is retired the use of compact accelerator-driven neutron sources (CANS) is becoming more prevalent. CANS are playing a significant and expanding role in research and development in science and engineering, as well as in education and training. In the realm of multidisciplinary applications, CANS offer opportunities over a wide range of technical utilization, from interrogation of civil structures to medical therapy to cultural heritage study. This paper aims to provide the first comprehensive overview of the history, current status of operation, and ongoing development of CANS worldwide. The basic physics and engineering regarding neutron production by accelerators, target-moderator systems, and beam line instrumentation are introduced, followed by an extensive discussion of various evolving applications currently exploited at CANS.

  1. The replacement of research reactors with a compact proton linac for neutron radiography

    International Nuclear Information System (INIS)

    The work presented here examines a neutron radiography facility, based on accelerator-driven compact neutron source, in order to find a suitable replacement for the facilities, which is based on research nuclear reactors. High-quality neutrons beam can be produced via the 9Be(p,n)9B reaction at proton energies of about 4 MeV. Except for the Be target the materials considered were compatible with the European Union Directive on ‘Restriction of Hazardous Substances’ (RoHS) 2002/95/EC. According to this directive some common materials in radiography units such as lead and cadmium have been excluded. The suggested facility has been simulated with an extensive range of parameters, which characterizes the neutron radiography, and the results specify that an accelerator-based neutron source is an attractive alternative to research nuclear reactors. - Highlights: ► Accelerator-driven source as an alternative to nuclear reactors for neutron radiography. ► Neutron beam can be produced via the 9Be(p,n)9B reaction at proton energies of 4 MeV. ► The presence of sapphire filter improves parameters associated with neutron radiography

  2. Code system for fast reactor neutronics analysis

    International Nuclear Information System (INIS)

    A code system for analysis of fast reactor neutronics has been developed for the purpose of handy use and error reduction. The JOINT code produces the input data file to be used in the neutronics calculation code and also prepares the cross section library file with an assigned format. The effective cross sections are saved in the PDS file with an unified format. At the present stage, this code system includes the following codes; SLAROM, ESELEM5, EXPANDA-G for the production of effective cross sections and CITATION-FBR, ANISN-JR, TWOTRAN2, PHENIX, 3DB, MORSE, CIPER and SNPERT. In the course of the development, some utility programs and service programs have been additionaly developed. These are used for access of PDS file, edit of the cross sections and graphic display. Included in this report are a description of input data format of the JOINT and other programs, and of the function of each subroutine and utility programs. The usage of PDS file is also explained. In Appendix A, the input formats are described for the revised version of the CIPER code. (author)

  3. Neutrons characterization of the nuclear reactor Ian-R1 of Colombia; Caracterizacion de los neutrones del reactor nuclear IAN-R1 de Colombia

    Energy Technology Data Exchange (ETDEWEB)

    Gonzalez P, L. X.; Martinez O, S. A. [Universidad Pedagogica y Tecnologica de Colombia, Grupo de Fisica Nuclear Aplicada y Simulacion, Carretera Central del Norte Km. 1, Via Paipa, 150003 Tunja, Boyaca (Colombia); Vega C, H. R., E-mail: s.agustin.martinez@uptc.edu.co [Universidad Autonoma de Zacatecas, Unidad Academica de Estudios Nucleares, Cipres No. 10, Fracc. La Penuela, 98068 Zacatecas (Mexico)

    2014-08-15

    By means of Monte Carlo methods, with the code MCNPX, the neutron characteristics of the research nuclear reactor Ian-R1 of Colombia, in power off but with the neutrons source in their start position, have been valued. The neutrons spectra, the total flow and their average power were calculated in the irradiation spaces inside the graphite reflector, as well as in the cells with air. Also the spectra, the total flow and the absorbed dose were calculated in several places distributed along the radial shaft inside the water moderator. The neutrons total flow was also considered to the long of the axial shaft. The characteristics of the neutrons spectra vary depending on their position regarding the source and the material that surrounds to the cell where the calculation was made. (Author)

  4. Neutron spin echo spectroscopy on the spallation neutron source

    International Nuclear Information System (INIS)

    An investigation has been made into the practicability of combining the neutron spin echo and time-of-flight techniques on the Rutherford Laboratory Spallation Neutron Source. Preliminary specifications are presented for a quasielastic instrument with an energy resolution down to approximately 10 neV and an inelastic spectrometer for measuring excitation widths approximately 1 μ eV. (author)

  5. (International Collaboration on Advanced Neutron Sources)

    Energy Technology Data Exchange (ETDEWEB)

    Hayter, J.B.

    1990-11-08

    The International Collaboration on Advanced Neutron Sources was started about a decade ago with the purpose of sharing information throughout the global neutron community. The collaboration has been extremely successful in optimizing the use of resources, and the discussions are open and detailed, with reasons for failure shared as well as reasons for success. Although the meetings have become increasingly oriented toward pulsed neutron sources, many of the neutron instrumentation techniques, such as the development of better monochromators, fast response detectors and various data analysis methods, are highly relevant to the Advanced Neutron Source (ANS). I presented one paper on the ANS, and another on the neutron optical polarizer design work which won a 1989 R D-100 Award. I also gained some valuable design ideas, in particular for the ANS hot source, in discussions with individual researchers from Canada, Western Europe, and Japan.

  6. Neutron Sources for Standard-Based Testing

    Energy Technology Data Exchange (ETDEWEB)

    Radev, Radoslav [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); McLean, Thomas [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2014-11-10

    The DHS TC Standards and the consensus ANSI Standards use 252Cf as the neutron source for performance testing because its energy spectrum is similar to the 235U and 239Pu fission sources used in nuclear weapons. An emission rate of 20,000 ± 20% neutrons per second is used for testing of the radiological requirements both in the ANSI standards and the TCS. Determination of the accurate neutron emission rate of the test source is important for maintaining consistency and agreement between testing results obtained at different testing facilities. Several characteristics in the manufacture and the decay of the source need to be understood and accounted for in order to make an accurate measurement of the performance of the neutron detection instrument. Additionally, neutron response characteristics of the particular instrument need to be known and taken into account as well as neutron scattering in the testing environment.

  7. Graphene liquid marbles as photothermal miniature reactors for reaction kinetics modulation.

    Science.gov (United States)

    Gao, Wei; Lee, Hiang Kwee; Hobley, Jonathan; Liu, Tianxi; Phang, In Yee; Ling, Xing Yi

    2015-03-23

    We demonstrate the fabrication of graphene liquid marbles as photothermal miniature reactors with precise temperature control for reaction kinetics modulation. Graphene liquid marbles show rapid and highly reproducible photothermal behavior while maintaining their excellent mechanical robustness. By tuning the applied laser power, swift regulation of graphene liquid marble's surface temperature between 21-135 °C and its encapsulated water temperature between 21-74 °C are demonstrated. The temperature regulation modulates the reaction kinetics in our graphene liquid marble, achieving a 12-fold superior reaction rate constant for methylene blue degradation than at room temperature.

  8. Development of neutron beam projects at the University of Texas TRIGA Mark II Reactor

    International Nuclear Information System (INIS)

    Recently, the UT-TRIGA research reactor was licensed and has become fully operational. This reactor, the first new US university reactor in 17 years, is the focus of a new reactor laboratory facility which is located on the Balcones Research Center at The University of Texas at Austin. The TRIGA Mark II reactor is licensed for 1.1 MW steady power operation, 3 dollar pulsing, and includes five beam ports. Various neutron beam-line projects have been assigned to each beam port. Neutron Depth Profiling (NDP) and the Texas Cold Neutron Source (TCNS) are close to completion and will be operational in the near future. The design of the NDP instrument has been completed, a target chamber has been built, and the thermal neutron collimator, detectors, data acquisition electronics, and data processing computers have been acquired. The target chamber accommodates wafers up to 12'' in diameter and provides remote positioning of these wafers. The design and construction of the TCNS has been completed. The TCNS consists of a moderator (mesitylene), a neon heat pipe, a cryogenic refrigerator, and neutron guide tubes. In addition, fission-fragment research (HIAWATHA), Neutron Capture Therapy, and Neutron Radiography are being pursued as projects for the other three beam ports. (author)

  9. Fast Neutron Detector for Fusion Reactor KSTAR Using Stilbene Scintillator

    CERN Document Server

    Lee, Seung Kyu; Kim, Gi-Dong; Kim, Yong-Kyun

    2011-01-01

    Various neutron diagnostic tools are used in fusion reactors to evaluate different aspects of plasma performance, such as fusion power, power density, ion temperature, fast ion energy, and their spatial distributions. The stilbene scintillator has been proposed for use as a neutron diagnostic system to measure the characteristics of neutrons from the Korea Superconducting Tokamak Advanced Research (KSTAR) fusion reactor. Specially designed electronics are necessary to measure fast neutron spectra with high radiation from a gamma-ray background. The signals from neutrons and gamma-rays are discriminated by the digital charge pulse shape discrimination (PSD) method, which uses total to partial charge ratio analysis. The signals are digitized by a flash analog-to-digital convertor (FADC). To evaluate the performance of the fabricated stilbene neutron diagnostic system, the efficiency of 10 mm soft-iron magnetic shielding and the detection efficiency of fast neutrons were tested experimentally using a 252Cf neutr...

  10. Nested Focusing Optics for Compact Neutron Sources

    Science.gov (United States)

    Nabors, Sammy A.

    2015-01-01

    NASA's Marshall Space Flight Center, the Massachusetts Institute of Technology (MIT), and the University of Alabama Huntsville (UAH) have developed novel neutron grazing incidence optics for use with small-scale portable neutron generators. The technology was developed to enable the use of commercially available neutron generators for applications requiring high flux densities, including high performance imaging and analysis. Nested grazing incidence mirror optics, with high collection efficiency, are used to produce divergent, parallel, or convergent neutron beams. Ray tracing simulations of the system (with source-object separation of 10m for 5 meV neutrons) show nearly an order of magnitude neutron flux increase on a 1-mm diameter object. The technology is a result of joint development efforts between NASA and MIT researchers seeking to maximize neutron flux from diffuse sources for imaging and testing applications.

  11. Simulation of neutron radiograph images at the Neutron Radiography Reactor

    International Nuclear Information System (INIS)

    Highlights: ► The project developed a method to simulate neutron radiograph images. ► A characteristic curve relation activation foil activity to film optical density. ► The simulation uses the characteristic curve to predict radiographic images. ► Radiographs of a polyethylene step block validate the methodology. - Abstract: The ability to accurately simulate potential radiographic images produced by a radiographic facility can improve the facility’s ability to design experiments and evaluate images. The image simulation methods detailed in this paper predict the radiographic image of an object based on the foil reaction rate data obtained by placing a model of the object in front of the image plane in a Monte Carlo beamline model. The image simulation method utilizes a characteristic curve relating foil activity to optical density for the film and foil combination in use at the Neutron Radiography Reactor. The simulation validation compared a radiograph of a polyethylene step block to a simulated radiograph of the same step block. The simulation accurately predicts the optical density in each region of a radiograph of the step block. The simulated radiograph predicts the average optical density of the actual radiograph more accurately for the thinner steps, resulting in step averaged optical density differences between the actual and simulated images of −11.6% for the thinnest step versus a difference of −34.7% for the thickest step, possibly due to the greater accuracy of the higher optical density region of the characteristic curve. Applying the scanner calibration curve to the calculated optical density values decreases the difference between the actual radiograph pixel values and the simulated pixel values for each step except the thinnest step. The step averaged differences between the corrected and actual images increase from −11.6% to −17.0% for the thinnest step and decrease from −34.7% to +7.7% for the thickest step after the

  12. Experimental determination of neutron lifetimes through macroscopic neutron noise in the IPEN/MB-01 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Gonnelli, Eduardo; Diniz, Ricardo [Instituto de Pesquisas Energeticas e Nucleares - IPEN/CNEN-SP Travessa R-400, 05508-900, Cidade Universitaria, Sao Paulo (Brazil)

    2013-05-06

    The neutron lifetimes of the core, reflector, and global were experimentally obtained through macroscopic neutron noise in the IPEN/MB-01 reactor for five levels of subcriticality. The theoretical Auto Power Spectral Densities were derived by point kinetic equations taking the reflector effect into account, and one of the approaches consider an additional group of delayed neutrons.

  13. Simplified neutron detector for angular distribution measurement of p-Li neutron source

    International Nuclear Information System (INIS)

    Boron Neutron Capture Therapy (BNCT) is one of the most promising cancer therapies using 10B(n, α)7Li nuclear reaction. Because nuclear reactor is currently used for BNCT, the therapy is much restricted. Many kinds of accelerator based neutron sources for BNCT are being investigated worldwide and p-Li reaction is one of the most promising candidates because the emitted neutron energy is comparatively low and no gamma-ray is produced. To use p-Li neutron source for BNCT, measurement of the angular distribution is important. However, the energy of neutrons changes depending on the angle with respect to the proton beam, e.g., the energy of forward emitted neutrons are about 700 keV and it is 100 keV for backward direction. So a neutron detector, the efficiency of which is not dependent on energy, is needed. Though so-called “Long Counter” is known to be available, its structure is complicated and moreover it is expensive. Thus we have designed and developed a simplified neutron detector using Monte Carlo simulation. We verified the developed detector experimentally and measured the angular distribution in detail for p-Li reaction by using it. The obtained results were compared with analytical calculations. (author)

  14. Neutron dosimetry. Environmental monitoring in a BWR type reactor

    International Nuclear Information System (INIS)

    The measurements carried out on reactor dosimetry are applied mainly to the study on the effects of the radiation in 108 materials of the reactor; little is on the environmental dosimetry outside of the primary container of BWR reactors. In this work the application of a neutron spectrometer formed by plastic detectors of nuclear traces manufactured in the ININ, for the environmental monitoring in penetrations around the primary container of the unit I of the Laguna Verde central is presented. The neutron monitoring carries out with purposes of radiological protection, during the operational tests of the reactor. (Author)

  15. Development of nuclear design criteria for neutron spallation sources

    Energy Technology Data Exchange (ETDEWEB)

    Sordo, F.; Abanades, A. [E.T.S. Industriales, Madrid Polytechnic University, UPM, J.Gutierrez Abascal, 2 -28006 Madrid (Spain)

    2008-07-01

    Spallation neutron sources allow obtaining high neutronic flux for many scientific and industrial applications. In recent years, several proposals have been made about its use, notably the European Spallation Source (ESS), the Japanese Spallation Source (JSNS) and the projects of Accelerator-Driven Subcritical reactors (ADS), particularly in the framework of EURATOM programs. Given their interest, it seems necessary to establish adequate design basis for guiding the engineering analysis and construction projects of this kind of installations. In this sense, all works done so far seek to obtain particular solutions to a particular design, but there has not been any general development to set up an engineering methodology in this field. In the integral design of a spallation source, all relevant physical processes that may influence its behaviour must be taken into account. Neutronic aspects (emitted neutrons and their spectrum, generation performance..), thermomechanical (energy deposition, cooling conditions, stress distribution..), radiological (spallation waste activity, activation reactions and residual heat) and material properties alteration due to irradiation (atomic displacements and gas generation) must all be considered. After analysing in a systematic manner the different options available in scientific literature, the main objective of this thesis was established as making a significant contribution to determine the limiting factors of the main aspects of spallation sources, its application range and the criteria for choosing optimal materials. To achieve this goal, a series of general simulations have been completed, covering all the relevant physical processes in the neutronic and thermal-mechanical field. Finally, the obtained criteria have been applied to the particular case of the design of the spallation source of subcritical reactors PDX-ADS and XT-ADS. These two designs, developed under the European R and D Framework Program, represent nowadays

  16. Use of Zircaloy 4 material for the pressure vessels of hot and cold neutron sources and beam tubes for research reactors results of forming the material and welding processes

    International Nuclear Information System (INIS)

    The material Zircaloy 4 can be used for the pressure retaining walls for the cold and hot neutron sources and beam tubes. For the research reactor FRM-II of the Technical University Munich, Germany, the material Zircaloy 4 were chosen for the vessels of the cold and hot neutron source and for the beam tube no. 6. For fabrication of the vessels and piping parts it was necessary to form the base material and calibrate the sheets or welded parts with necessary heat treatments. Additional to the technical specifications preliminary material investigations and production test of welded and unwelded material were carried out of the formed parts up to a cold work of 5%. Further one with respect to the material thickness of 3, 4, 5 and 10 mm of the used sheets, welding procedure test before the fabrication and welding production tests during fabrication were carried out of the base material combination sheet/sheet and sheet/forging. Electronic beam welding was used for the welding process. Material tests as tensile tests, charpy-V-tests, bend tests, metallographic tests, hardness tests, radiographic tests a.s.o. were carried out. The results of the examinations confirm the specified requirements. For the material forming process an optimization was necessary after the preliminary results to get final sufficient material behaviour results. (orig.)

  17. Neutron scattering instruments for the Spallation Neutron Source (SNS)

    Energy Technology Data Exchange (ETDEWEB)

    Crawford, R.K.; Fornek, T. [Argonne National Lab., IL (United States); Herwig, K.W. [Oak Ridge National Lab., TN (United States)

    1998-07-01

    The Spallation Neutron Source (SNS) is a 1 MW pulsed spallation source for neutron scattering planned for construction at Oak Ridge National Laboratory. This facility is being designed as a 5-laboratory collaboration project. This paper addresses the proposed facility layout, the process for selection and construction of neutron scattering instruments at the SNS, the initial planning done on the basis of a reference set of ten instruments, and the plans for research and development (R and D) to support construction of the first ten instruments and to establish the infrastructure to support later development and construction of additional instruments.

  18. Advanced Neutron Source: Plant Design Requirements. Revision 4

    Energy Technology Data Exchange (ETDEWEB)

    1990-07-01

    The Advanced Neutron Source will be a new world-class facility for research using hot, thermal, cold, and ultra-cold neutrons. The heart of the facility will be a 330-MW (fission), heavy-water cooled and heavy-water moderated reactor. The reactor will be housed in a central reactor building, with supporting equipment located in an adjoining reactor support building. An array of cold neutron guides will fan out into a large guide hall, housing about 30 neutron research stations. Appropriate office, laboratory, and shop facilities will be included to provide a complete facility for users. The ANS is scheduled to begin operation at the Oak Ridge National Laboratory early in the next decade. This PDR document defines the plant-level requirements for the design, construction, and operation of ANS. It also defines and provides input to the individual System Design Description (SDD) documents. Together, this PDR document and the set of SDD documents will define and control the baseline configuration of ANS.

  19. Radionuclide 252Cf neutron source

    International Nuclear Information System (INIS)

    Characteristics of radionuclide neutron sourses of 252Cf base with the activity from 106 to 109 n/s have been investigated. Energetic distributions of neutrons and gamma-radiation have been presented. The results obtained have been compared with other data available. The hardness parameter of the neutron spectrum for the energy range from 3 to 15 MeV is 1.4 +- 0.02 MeV

  20. Simulation of neutron generation in short pulsed X-band linac neutron source

    International Nuclear Information System (INIS)

    It is important to improve the accuracy of nuclear cross section for waste reprocessing and design of new reactors, but facilities where nuclear fuel materials can be measured are limited. So, in Tokyo University, there is a plan of development of electron linac neutron source for analyzing nuclear data. The research reactor “Yayoi” was decommissioned in Tokyo University and by introducing this linac in the core of the reactor, measurement of nuclear fuel materials can be expected. 30 MeV X-band linac is used so this system will be compact and this enables us to introduce the system into the core of “Yayoi.” Pulse width is short in order to measure high energy neutron with TOF method. In this research, generation of electrons, acceleration and interaction with several kinds of target and moderator are simulated. (author)

  1. Neutron diagnostic investigations with a research reactor

    International Nuclear Information System (INIS)

    Some aspects of the use of neutron transmission analysis in applied research, as pursued at McMaster University (Canada), are examined. Examples considered are void measurements in two-phase flow, neutron conversion enhancement in neutron radiography, reconstruction of interior bulk heterogenities in solids and temperature sensing with neutrons. (author)

  2. Neutron dosimetry. Environmental monitoring in a BWR type reactor; Dosimetria de neutrones. Monitoreo ambiental en un reactor del tipo BWR

    Energy Technology Data Exchange (ETDEWEB)

    Tavera D, L.; Camacho L, M.E

    1991-01-15

    The measurements carried out on reactor dosimetry are applied mainly to the study on the effects of the radiation in 108 materials of the reactor; little is on the environmental dosimetry outside of the primary container of BWR reactors. In this work the application of a neutron spectrometer formed by plastic detectors of nuclear traces manufactured in the ININ, for the environmental monitoring in penetrations around the primary container of the unit I of the Laguna Verde central is presented. The neutron monitoring carries out with purposes of radiological protection, during the operational tests of the reactor. (Author)

  3. New modelling method for fast reactor neutronic behaviours analysis

    International Nuclear Information System (INIS)

    Due to safety rules running on fourth generation reactors' core development, neutronics simulation tools have to be as accurate as never before. First part of this report enumerates every step of fast reactor's neutronics simulation implemented in current reference code: ECCO. Considering the field of fast reactors that meet criteria of fourth generation, ability of models to describe self-shielding phenomenon, to simulate neutrons leakage in a lattice of fuel assemblies and to produce representative macroscopic sections is evaluated. The second part of this thesis is dedicated to the simulation of fast reactors' core with steel reflector. These require the development of advanced methods of condensation and homogenization. Several methods are proposed and compared on a typical case: the ZONA2B core of MASURCA reactor. (author)

  4. Thermal and neutronic calculation for fast breeder reactor FBR

    International Nuclear Information System (INIS)

    This research included studying of thermal and neutronic calculation for fast breeder nuclear reactor, to putting the optimum design for this reactor. So a Soviet type (BN-350) was chosen, which has its core composed of two enrichment zones, and with blanket that contains depleted uranium. A group of thermal calculation programs was made by using personal computer, to obtain core and blanket reactor dimensions and volume fractions of reaction input material and number and dimensions of fuel rods which were used for neutron calculations. Several core and blanket enrichments were used to study neutron flux behaviour for two reactors different conditions. First when control rods exist in the core reactor and second when the rods are out of the core. Breeding ratio was also studied for different core and blanket enrichment. 30 tabs.; 24 figs.; 34 refs.; 3 apps

  5. Post-shutdown decay power and radionuclide inventories in the discharged fuels of HEU and potential LEU miniature neutron source reactors

    Energy Technology Data Exchange (ETDEWEB)

    Mirza, Sikander M.; Khan, Abdullah [Department of Physics and Applied Mathematics, Pakistan Institute of Engineering and Applied Sciences (PIEAS), P.O. Nilore, Islamabad 45650 (Pakistan); Mirza, Nasir M., E-mail: nasirmm@yahoo.co [Department of Physics and Applied Mathematics, Pakistan Institute of Engineering and Applied Sciences (PIEAS), P.O. Nilore, Islamabad 45650 (Pakistan)

    2010-05-15

    Assessment of fission product and actinide content along with the time variation of decay power of discharged fuels of both HEU and LEU cores of MNSRs have been carried out for once-through cycle using the ORIGEN2 computer code. The results for the LEU core have been compared with the corresponding values for the current HEU core of MNSRs. For the HEU and the potential LEU UO{sub 2}, U-9Mo discharged fuels, the ORIGEN2 computed isotopic and total activity values have been found in good agreement with the corresponding results obtained by using the WIMSD4 code. All three MNSR fuels show fission product dominated activity behavior for post-shutdown periods up to about 10{sup 3} years during which, the total activity decreases by as much as 10{sup 6} times. The residual actinide activity shows smaller variations as the three discharged fuels decay thru 10{sup 6} years. The time variation of the decay power follows the same behavior as the corresponding total activity values during the fission product dominated period. A decrease from initial values of 154.76, 162.6,160.39 W to the final values 9.35 x 10{sup -5}, 2.1 x 10{sup -3}, 1.7 x 10{sup -3} W has been found for the standard HEU, and potential UO{sub 2}, U-9Mo LEU fuels correspondingly during this time. The standard HEU fuel shows smallest decay power values while the UO{sub 2} and U-9Mo LEU fuels have comparable values for time spans from 10{sup 3} to about 10{sup 6} years.

  6. High Brightness Neutron Source for Radiography. Final report

    International Nuclear Information System (INIS)

    This research and development program was designed to improve nondestructive evaluation of large mechanical objects by providing both fast and thermal neutron sources for radiography. Neutron radiography permits inspection inside objects that x-rays cannot penetrate and permits imaging of corrosion and cracks in low-density materials. Discovering of fatigue cracks and corrosion in piping without the necessity of insulation removal is possible. Neutron radiography sources can provide for the nondestructive testing interests of commercial and military aircraft, public utilities and petrochemical organizations. Three neutron prototype neutron generators were designed and fabricated based on original research done at the Lawrence Berkeley National Laboratory (LBNL). The research and development of these generators was successfully continued by LBNL and Adelphi Technology Inc. under this STTR. The original design goals of high neutron yield and generator robustness have been achieved, using new technology developed under this grant. In one prototype generator, the fast neutron yield and brightness was roughly 10 times larger than previously marketed neutron generators using the same deuterium-deuterium reaction. In another generator, we integrate a moderator with a fast neutron source, resulting in a high brightness thermal neutron generator. The moderator acts as both conventional moderator and mechanical and electrical support structure for the generator and effectively mimics a nuclear reactor. In addition to the new prototype generators, an entirely new plasma ion source for neutron production was developed. First developed by LBNL, this source uses a spiral antenna to more efficiently couple the RF radiation into the plasma, reducing the required gas pressure so that the generator head can be completely sealed, permitting the possible use of tritium gas. This also permits the generator to use the deuterium-tritium reaction to produce 14-MeV neutrons with increases

  7. Materials research with neutron beams from a research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Root, J.; Banks, D. [Canadian Neutron Beam Centre, Chalk River Laboratories, Chalk River, Ontario (Canada)

    2015-03-15

    Because of the unique ways that neutrons interact with matter, neutron beams from a research reactor can reveal knowledge about materials that cannot be obtained as easily with other scientific methods. Neutron beams are suitable for imaging methods (radiography or tomography), for scattering methods (diffraction, spectroscopy, and reflectometry) and for other possibilities. Neutron-beam methods are applied by students and researchers from academia, industry and government to support their materials research programs in several disciplines: physics, chemistry, materials science and life science. The arising knowledge about materials has been applied to advance technologies that appear in everyday life: transportation, communication, energy, environment and health. This paper illustrates the broad spectrum of materials research with neutron beams, by presenting examples from the Canadian Neutron Beam Centre at the NRU research reactor in Chalk River. (author)

  8. Measurements of thermal- and slow-neutron dose distributions in ordinary concrete shield using a reactor neutron beam of different energy ranges

    Energy Technology Data Exchange (ETDEWEB)

    Megahid, R.M.; Makarious, A.S.; El-Kolaly, M.A.; Afifi, Y.A.

    1980-01-01

    Experimental studies on the distribution and attenuation of thermal and slow neutron doses in ordinary concrete shield have been carried-out. A collimated beam of reactor neutrons emitted from one of the horizontal channels of the ET-RR-1 reactor was used. Measurements were performed using, a direct beam, cadmium filtered beam and boron carbide filtered beam. The neutron doses were measured using thermolumin-escent Li/sub 2/B/sub 4/O/sub 7/ detectors. The measured data have been analyzed and a group of attenuation curves were given for beams of reactor neutrons of different energy. These curves show that cadmium and boron carbide filters tend to decrease the neutron doses specially at the beginning of penetration. The data were transformed to that which would be obtained using neutron sources of different geometries.

  9. Evaluation of impact properties of weld joint of reactor pressure vessel steels with the use of miniaturized specimens

    International Nuclear Information System (INIS)

    The effects of specimen size and location of V-notch on the Charpy impact properties were investigated with different sizes of specimens, standard, CVN-1/2, CVN-1/3, and CVN-1.5 mm, for A533B steel, low Mn, high Cu, high phosphorus (P), and high Cu/P steel weld joint. A part of the specimens was irradiated with neutron at 563 K up to 8x1019 n/cm2. The heat affected zone (HAZ) specimen is the best in the impact properties among the specimens of base metal, HAZ, and weld metal in the steels with 0.003 wt.% P, while it is the worst in the steels with ∼ 0.3 wt.% P. This indicates that the surveillance test of HAZ specimen can be represented by base metal in the case of A533B steels with lower P content (∼ 0.003 wt.%). The effects of notch location and chemical contents on ductile to brittle transition temperature (DBTT) are almost independent of specimen size within an error of ±5 K, indicating that the miniaturized Charpy specimens are applicable and effective in the surveillance tests of reactor pressure vessel steel of extended operation period. After irradiation, the highest DBTT was observed for the specimen with V-notch in base metal in the case of A533B steel and high Cu steel with 0.003 wt.% P. (author)

  10. Characteristics and sources of 2002 super dust storm in Beijing

    Institute of Scientific and Technical Information of China (English)

    SUN Yele; ZHUANG Guoshun; YUAN Hui; ZHANG Xingying; GUO Jinghua

    2004-01-01

    On 20 March, 2002, a super dust storm attacked Beijing, which was stronger than any dust storm ever recorded. The concentration of total suspended particulates air quality standard. The concentrations of major crustal elements, such as Ca, Al, Fe, Na, Mg and Ti, were 30-58times higher than those in non-dust storm days. The concentrations of pollution elements, such as Zn, Cu, Pb, As, Cd and S, were also about several or even nearly ten times higher than those in normal days. The enrichment factors of Pb, As, Cd and S in PM2.5 were as high as 12.7, 29.6, 43.5,28.4, indicating that these pollutants came from the mixing of mineral aerosol with pollution aerosol emitted by pollution sources on the way of dust storm's long-range transport. The overlap of invaded air mass from dust with pollution air mass from Beijing local area was another reason for the enhancement of pollutants. During dust storm, fine particles (PM2.5) accounted for 30% of TSP and pollutants in PM2.5accounted for even as high as 45%-69% of TSP. The increase of pollutants after dust storm proved further that mineral aerosol, especially the fine particles from dust storm favored the transformation and accumulation of pollutants.It must be noted that Fe (Ⅱ) was detected again in this dust storm, which provided new evidence for the mechanism of coupling and feedback between iron and sulfur in the atmosphere and the ocean. The increase of both pollutants and nutrient, Fe(Ⅱ), during dust storm illuminated that dust storm is an important factor affecting the global environment change.

  11. Review of the Advanced Neutron Source (ANS) materials irradiation facilities

    International Nuclear Information System (INIS)

    The purpose of the workshop was to document as accurately as possible the present and future needs for neutron irradiation capacity and facilities as related to the design of the Advanced Neutron Source (ANS) which will be the next generation steady-state research reactor. The report provides the findings and recommendations of the working group. After introductory and background information is presented, the discussion includes the status of the ANS design, in particular in-core materials irradiation facilities design and important experimental parameters. The summary of workshop discussions describes a survey of irradiation-effects research community and opportunities for ex-core irradiation facilities. 20 refs., 2 figs., 4 tabs

  12. Description of the CAREM Reactor Neutronic Calculation Codes

    International Nuclear Information System (INIS)

    In this work is described the neutronic calculation line used to design the CAREM reactor.A description of the codes used and the interfaces between the different programs are presented.Both, the normal calculation line and the alternative or verification calculation line are included.The calculation line used to obtain the kinetics parameters (effective delayed-neutron fraction and prompt-neutron lifetime) is also included

  13. Detection of Neutron Sources in Cargo Containers

    OpenAIRE

    Katz, J. I.

    2007-01-01

    We investigate the problem of detecting the presence of clandestine neutron sources, such as would be produced by nuclear weapons containing plutonium, within cargo containers. Small, simple and economical semiconductor photodiode detectors affixed to the outsides of containers are capable of producing statistically robust detections of unshielded sources when their output is integrated over the durations of ocean voyages. It is possible to shield such sources with thick layers of neutron-abs...

  14. Socio-economic Analysis of the Beijing-Tianjin Sandstorm Source Control Project Phase Ⅱ

    Institute of Scientific and Technical Information of China (English)

    Zhangfenglai

    2012-01-01

    With the purpose of improving the air quality as well as curbing sand and dust hazard in the Beijing-Tianjin Area,our country urgently started Beijing-Tianjin Sandstorm Source Control Project in 2000,which,more then ten years' construction,has achieved enormous social,economic and ecological results.

  15. Design considerations for neutron activation and neutron source strength monitors for ITER

    Energy Technology Data Exchange (ETDEWEB)

    Barnes, C.W. [Los Alamos National Lab., NM (United States); Jassby, D.L.; LeMunyan, G.; Roquemore, A.L. [Princeton Univ., NJ (United States). Plasma Physics Lab.; Walker, C. [ITER Joint Central Team, Garching (Germany)

    1997-12-31

    The International Thermonuclear Experimental Reactor will require highly accurate measurements of fusion power production in time, space, and energy. Spectrometers in the neutron camera could do it all, but experience has taught us that multiple methods with redundancy and complementary uncertainties are needed. Previously, conceptual designs have been presented for time-integrated neutron activation and time-dependent neutron source strength monitors, both of which will be important parts of the integrated suite of neutron diagnostics for this purpose. The primary goals of the neutron activation system are: to maintain a robust relative measure of fusion energy production with stability and wide dynamic range; to enable an accurate absolute calibration of fusion power using neutronic techniques as successfully demonstrated on JET and TFTR; and to provide a flexible system for materials testing. The greatest difficulty is that the irradiation locations need to be close to plasma with a wide field of view. The routing of the pneumatic system is difficult because of minimum radius of curvature requirements and because of the careful need for containment of the tritium and activated air. The neutron source strength system needs to provide real-time source strength vs. time with {approximately}1 ms resolution and wide dynamic range in a robust and reliable manner with the capability to be absolutely calibrated by in-situ neutron sources as done on TFTR, JT-60U, and JET. In this paper a more detailed look at the expected neutron flux field around ITER is folded into a more complete design of the fission chamber system.

  16. About possibilities of obtaining focused beams of thermal neutrons of radionuclide source

    International Nuclear Information System (INIS)

    Full text: In the last years significant progress is achieved in development of neutron focusing methods (concentrating neutrons in a given direction and a small area). In this, main attention is given to focusing of neutron beams of reactor, particularly cold neutrons and their applications. [1,2]. However, isotope sources also let obtain intensive neutron beams and solve quite important (tasks) problems (e.g. neutron capture therapy for malignant tumors) [3], and an actual problems is focusing of neutrons. We developed a device on the basis of californium source of neutrons, allowing to obtain focused (preliminarily) beam of thermal neutrons with the aid of respective choice of moderators, reflectors and geometry of their disposition. Here, fast neutrons and gamma rays in the beam are minimized. With the aid of the model we developed on the basis of Monte-Carlo method, it is possible to modify aforementioned device and dynamics of output neutrons in wide energy range and analyze ways of optimization of neutron beams of isotope sources with different neutron outputs. Device of preliminary focusing of thermal neutrons can serve as a basis for further focus of neutrons using micro- and nano-capillar systems. It is known that, capillary systems performed with certain technology can form beam of thermal neutrons increasing its density by more than two orders of magnitude and effectively divert beams up to 20o with length of system 15 cm

  17. Calculation of neutron flux in the presence of a source

    International Nuclear Information System (INIS)

    Neutron sources are introduced into the reactors to initiate the chain reaction. For safety reasons, we have to know the distribution and evolution of the flux throughout the startup phase. The flux is calculated iteratively but convergence of the process can slow down arbitrarily as we approach criticality. A calculation method is presented, with a convergence speed which does not depend on the negative reactivity when it is small. (author). 7 refs

  18. Neutron noise analysis techniques in nuclear power reactors

    International Nuclear Information System (INIS)

    The main techniques used in neutron noise analysis of BWR and PWR nuclear reactors are reviewed. Several applications such as control of vibrations in both reactor types, determination of two phase flow parameters in BWR and stability control in BWR are discussed with some detail. The paper contains many experimental results obtained by the main author of this paper. (author)

  19. Pulsed neutron source and instruments at neutron facility

    Energy Technology Data Exchange (ETDEWEB)

    Teshigawara, Makoto; Aizawa, Kazuya; Suzuki, Jun-ichi; Morii, Yukio; Watanabe, Noboru [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1997-11-01

    We report the results of design studies on the optimal target shape, target - moderator coupling, optimal layout of moderators, and neutron instruments for a next generation pulsed spallation source in JAERI. The source utilizes a projected high-intensity proton accelerator (linac: 1.5 GeV, {approx}8 MW in total beam power, compressor ring: {approx}5 MW). We discuss the target neutronics, moderators and their layout. The sources is designed to have at least 30 beam lines equipped with more than 40 instruments, which are selected tentatively to the present knowledge. (author)

  20. A new neutron noise technique for fast reactors

    International Nuclear Information System (INIS)

    This paper gives a new neutron noise technique for fast reactors, which is known as thermalization measurement technique of the neutron noise. The theoretical formulas of the technique were developed, and a digital delayed coincidence time analyzer consisted of TTL integrated circuits was constructed for the study of this technique. The technique has been tested and applied practically at Df-VI fast zero power reactor. It was shown that the provided technique in this work has a number of significant advantages in comparison with the conventional neutron noise method

  1. Advanced Neutron Source (ANS) Project. Progress report FY 1993

    Energy Technology Data Exchange (ETDEWEB)

    Campbell, J.H. [ed.; Selby, D.L.; Harrington, R.M. [Oak Ridge National Lab., TN (United States); Thompson, P.B. [Martin Marietta Energy Systems, Inc., Oak Ridge, TN (United States). Engineering Div.

    1994-01-01

    This report covers the progress made in 1993 in the following sections: (1) project management; (2) research and development; (3) design and (4) safety. The section on research and development covers the following: (1) reactor core development; (2) fuel development; (3) corrosion loop tests and analysis; (4) thermal-hydraulic loop tests; (5) reactor control and shutdown concepts; (6) critical and subcritical experiments; (7) material data, structure tests, and analysis; (8) cold source development; (9) beam tube, guide, and instrument development; (10) neutron transport and shielding; (11) I and C research and development; and (12) facility concepts.

  2. Design, construction, and demonstration of a neutron beamline and a neutron imaging facility at a Mark-I TRIGA reactor

    Science.gov (United States)

    Craft, Aaron E.

    The fleet of research and training reactors is aging, and no new research reactors are planned in the United States. Thus, there is a need to expand the capabilities of existing reactors to meet users' needs. While many research reactors have beam port facilities, the original design of the United States Geological Survey TRIGA Reactor (GSTR) did not include beam ports. The MInes NEutron Radiography (MINER) facility developed by this thesis and installed at the GSTR provides new capabilities for both researchers and students at the Colorado School of Mines. The facility consists of a number of components, including a neutron beamline and beamstop, an optical table, an experimental enclosure and associated interlocks, a computer control system, a multi-channel plate imaging detector, and the associated electronics. The neutron beam source location, determined through Monte Carlo modeling, provides the best mixture of high neutron flux, high thermal neutron content, and low gamma radiation content. A Monte Carlo n-Particle (MCNP) model of the neutron beam provides researchers with a tool for designing experiments before placing objects in the neutron beam. Experimental multi-foil activation results, compared to calculated multi-foil activation results, verify the model. The MCNP model predicts a neutron beamline flux of 2.2*106 +/- 6.4*105 n/cm2-s based on a source particle rate determined from the foil activation experiments when the reactor is operating at a power of 950 kWt with the beam shutter fully open. The average cadmium ratio of the beamline is 7.4, and the L/D of the neutron beam is approximately 200+/-10. Radiographs of a sensitivity indicator taken using both the digital detector and the transfer foil method provide one demonstration of the radiographic capabilities of the new facility. Calibration fuel pins manufactured using copper and stainless steel surrogate fuel pellets provide additional specimens for demonstration of the new facility and offer a

  3. The high intensity neutron source FRANZ

    CERN Document Server

    Lederer, Claudia

    2014-01-01

    The Frankfurt neutron source of Stern Gerlach Zentrum FRANZ is currently under construction at the University of Frankfurt. At FRANZ, a high intensity neutron beam in the keV energy region will be produced by bombarding a $^7$Li target with a proton beam of several mA. These unprecedented high neutron fluxes will allow a number of neutron induced cross section measurements for the first time. Measurements can be performed by the time-of-flight and by the activation technique.

  4. Measurement of the neutron spectrum of a Pu-C source with a liquid scintillator

    Institute of Scientific and Technical Information of China (English)

    WANG Song-Lin; HUANG Han-Xiong; RUAN Xi-Chao; LI Xia; BAO Jie; NIE Yang-So; ZHONG Qi-Ping; ZHOU Zu-Ying; KONG Xiang-Zhong

    2009-01-01

    The neutron response function for a BC501A liquid scintillator (LS) has been measured using a series of monoenergetic neutrons produced by the p-T reaction. The proton energies were chosen such as to produce neutrons in the energy range of 1 to 20 MeV. The principles of the technique of unfolding a neutron energy spectrum by using the measured neutron response function and the measured Pulse Height (PH) spectrum is briefly described. The PH spectrum of neutrons from the Pu-C source, which will be used for the calibration of the reactor antineutrino detectors for the Daya Bay neutrino experiment, was measured and analyzed to get the neutron energy spectrum. Simultaneously the neutron energy spectrum of an Am-Be source was measured and compared with other measurements as a check of the result for the Pu-C source. Finally, an error analysis and a discussion of the results are given.

  5. 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)

  6. Evaluation of thermal neutron irradiation field using a cyclotron-based neutron source for alpha autoradiography

    International Nuclear Information System (INIS)

    It is important to measure the microdistribution of 10B in a cell to predict the cell-killing effect of new boron compounds in the field of boron neutron capture therapy. Alpha autoradiography has generally been used to detect the microdistribution of 10B in a cell. Although it has been performed using a reactor-based neutron source, the realization of an accelerator-based thermal neutron irradiation field is anticipated because of its easy installation at any location and stable operation. Therefore, we propose a method using a cyclotron-based epithermal neutron source in combination with a water phantom to produce a thermal neutron irradiation field for alpha autoradiography. This system can supply a uniform thermal neutron field with an intensity of 1.7×109 (cm−2 s−1) and an area of 40 mm in diameter. In this paper, we give an overview of our proposed system and describe a demonstration test using a mouse liver sample injected with 500 mg/kg of boronophenyl-alanine. - Highlights: • We developed a thermal neutron irradiation field using cyclotron based epithermal neutron source combination with a water phantom for alpha autoradiography. • The uniform thermal neutron irradiation field with an intensity of 1.7×109 (cm−2 s−1) with a size of 40 mm in diameter was obtained. • Demonstration test of alpha autoradiography using a liver sample with the injection of BPA was performed. • Boron image discriminated with the background event of protons was clearly shown by means of the particle identification

  7. Nuclear and dosimetric features of an isotopic neutron source

    International Nuclear Information System (INIS)

    A multisphere neutron spectrometer was used to determine the features of a 239PuBe neutron source that is used to operate the ESFM-IPN Subcritical Reactor. To determine the source main features it was located a 100 cm from the spectrometer which was a 6LiI(Eu) scintillator and 2, 3, 5, 8, 10 and 12 in.-diameter polyethylene spheres. Count rates obtained with the spectrometer were unfolded using the NSDUAZ code and neutron spectrum, total fluence, and ambient dose equivalent were determined. A Monte Carlo calculation was carried out to estimate the spectrum and integral features being less than values obtained experimentally due to the presence of 241Pu in the Pu used to fabricate the source. Actual neutron yield and the mass fraction of 241Pu was estimated. - Highlights: • The neutron spectrum of a 239PuBe was measured. • With the spectrum integral features were determined. • It was estimated 0.23 w/o of 241Pu in the Pu used to make the source

  8. Prompt γ-ray analysis with reactor neutrons (review)

    International Nuclear Information System (INIS)

    A neutron-induced prompt γ-ray analysis (PGA) has recently evolved as a non-destructive analytical method which can determine light elements, such as H and B, and multi-elements as a result of increased analytical sensitivities by using low-energy guided neutron beams of nuclear reactors. Firstly, the principle, characteristics and history of reactor neutron-based PGA is described; then, the characteristics of PGA systems which are classified into internal, beam and guided beam types are described. Moreover, the basics and standardization for the elemental determination in PGA are described and various applications of PGA are reviewed. Finally, micro-sample, spot spatial-distribution analyses of elements using a cold-neutron micro beam produced by a neutron lens are described and the future development of the method is predicted. (author)

  9. Neutronic performance issues for the Spallation Neutron Source moderators

    Energy Technology Data Exchange (ETDEWEB)

    Iverson, E.B.; Murphy, B.D. [Oak Ridge National Laboratory, Spallation Neutron Source, Oak Ridge, TN (United States)

    2001-03-01

    We continue to develop the neutronic models of the Spallation Neutron Source target station and moderators in order to better predict the neutronic performance of the system as a whole and in order to better optimize that performance. While we are not able to say that every model change leads to more intense neutron beams being predicted, we do feel that such changes are advantageous in either performance or in the accuracy of the prediction of performance. We have computationally and experimentally studied the neutronics of hydrogen-water composite moderators such as are proposed for the SNS Project. In performing these studies, we find that the composite moderator, at least in the configuration we have examined, does not provide performance characteristics desirable for the instruments proposed and being designed for this neutron scattering facility. The pulse width as a function of energy is significantly broader than for other moderators, limiting attainable resolution-bandwidth combinations. Furthermore, there is reason to expect that higher-energy (0.1-1 eV) applications will be significantly impacted by bimodal pulse shapes requiring enormous effort to parameterize. As a result of these studies, we have changed the SNS design, and will not use a composite moderator at this time. We have analyzed the depletion of a gadolinium poison plate in a hydrogen moderator at the Spallation Neutron Source, and found that conventional poison thicknesses will be completely unable to last the desired component lifetime of three operational years. A poison plate 300-600 {mu}m thick will survive for the required length of time, but will somewhat degrade the intensity (by as much as 15% depending on neutron energy) and the consistency of the neutron source performance. Our results should scale fairly easily to other moderators on this or any other spallation source. While depletion will be important for all highly-absorbing materials in high-flux regions, we feel it likely

  10. Materials Selection for the HFIR Cold Neutron Source

    Energy Technology Data Exchange (ETDEWEB)

    Farrell, K.

    2001-08-24

    In year 2002 the High Flux Isotope Reactor (HFIR) will be fitted with a source of cold neutrons to upgrade and expand its existing neutron scattering facilities. The in-reactor components of the new source consist of a moderator vessel containing supercritical hydrogen gas moderator at a temperature of 20K and pressure of 15 bar, and a surrounding vacuum vessel. They will be installed in an enlarged beam tube located at the site of the present horizontal beam tube, HB-4; which terminates within the reactor's beryllium reflector. These components must withstand exceptional service conditions. This report describes the reasons and factors underlying the choice of 6061-T6 aluminum alloy for construction of the in-reactor components. The overwhelming considerations are the need to minimize generation of nuclear heat and to remove that heat through the flowing moderator, and to achieve a minimum service life of about 8 years coincident with the replacement schedule for the beryllium reflector. 6061-T6 aluminum alloy offers the best combination of low nuclear heating, high thermal conductivity, good fabricability, compatibility with hydrogen, superior cryogenic properties, and a well-established history of satisfactory performance in nuclear environments. These features are documented herein. An assessment is given of the expected performance of each component of the cold source.

  11. Future opportunities with pulsed neutron sources

    Energy Technology Data Exchange (ETDEWEB)

    Taylor, A.D. [Rutherford Appleton Lab., Chilton (United Kingdom)

    1996-05-01

    ISIS is the world`s most powerful pulsed spallation source and in the past ten years has demonstrated the scientific potential of accelerator-driven pulsed neutron sources in fields as diverse as physics, earth sciences, chemistry, materials science, engineering and biology. The Japan Hadron Project gives the opportunity to build on this development and to further realize the potential of neutrons as a microscopic probe of the condensed state. (author)

  12. Targets for neutron beam spallation sources

    International Nuclear Information System (INIS)

    The meeting on Targets for Neutron Beam Spallation Sources held at the Institut fuer Festkoerperforschung at KFA Juelich on June 11 and 12, 1979 was planned as an informal get-together for scientists involved in the planning, design and future use of spallation neutron sources in Europe. These proceedings contain the papers contributed to this meeting. For further information see hints under relevant topics. (orig./FKS)

  13. Role of the Tapiro Fast Research Reactor in Neutron Capture Therapy in Italy Calculations and Measurements

    International Nuclear Information System (INIS)

    Thermal-neutron research reactors are currently the most common source of neutron beams for both research and clinical trials of neutron capture therapy (NCT). Neutron spectra suitable for NCT are typically produced either by beam filtering or spectrum shifting techniques. However, fast-neutron reactors are also being considered for NCT application as it is recognized that they may allow for improved beam quality. TAPIRO is a low power, high flux, highly enriched (93.5% 235U) fast reactor. The power is 5 kW and the maximum neutron flux in the core is 3x1012 cm-2.s-1. Both a thermal and an epithermal column have been designed and constructed, aimed at dosimetry and animal experiments. The configurations of the columns have been designed by means of Monte Carlo calculations. The columns have been characterized by means of measurements performed with activation techniques and thermoluminescence and gel dosimeters. Experimental results have shown good consistency with calculations. Moreover, they have confirmed the good quality of the beams obtainable with such a reactor. An epithermal column for clinical trials of patients with brain gliomas has been designed and is under construction. The treatment planning figures-of-merit in an anthropomorphic phantom look very satisfactory. (author)

  14. Characteristics comparison between a cyclotron-based neutron source and KUR-HWNIF for boron neutron capture therapy

    Science.gov (United States)

    Tanaka, H.; Sakurai, Y.; Suzuki, M.; Masunaga, S.; Kinashi, Y.; Kashino, G.; Liu, Y.; Mitsumoto, T.; Yajima, S.; Tsutsui, H.; Maruhashi, A.; Ono, K.

    2009-06-01

    At Kyoto University Research Reactor Institute (KURRI), 275 clinical trials of boron neutron capture therapy (BNCT) have been performed as of March 2006, and the effectiveness of BNCT has been revealed. In order to further develop BNCT, it is desirable to supply accelerator-based epithermal-neutron sources that can be installed near the hospital. We proposed the method of filtering and moderating fast neutrons, which are emitted from the reaction between a beryllium target and 30-MeV protons accelerated by a cyclotron accelerator, using an optimum moderator system composed of iron, lead, aluminum and calcium fluoride. At present, an epithermal-neutron source is under construction from June 2008. This system consists of a cyclotron accelerator, beam transport system, neutron-yielding target, filter, moderator and irradiation bed. In this article, an overview of this system and the properties of the treatment neutron beam optimized by the MCNPX Monte Carlo neutron transport code are presented. The distribution of biological effect weighted dose in a head phantom compared with that of Kyoto University Research Reactor (KUR) is shown. It is confirmed that for the accelerator, the biological effect weighted dose for a deeply situated tumor in the phantom is 18% larger than that for KUR, when the limit dose of the normal brain is 10 Gy-eq. The therapeutic time of the cyclotron-based neutron sources are nearly one-quarter of that of KUR. The cyclotron-based epithermal-neutron source is a promising alternative to reactor-based neutron sources for treatments by BNCT.

  15. Summary report of the consultants' meeting on neutron sources spectra for EXFOR

    International Nuclear Information System (INIS)

    The participants highlighted the importance of complementing the averaged cross section data already stored in EXFOR by the incident neutron energy spectra. They shared their experience on measurement and simulation of neutron fields produced at reactors and accelerators over a wide energy range. The source characteristics, format and rules needed for storage in EXFOR were discussed. The participants submitted the numerical information on spectra that will essentially increase the number of 'complete' data sets in EXFOR. The report additionally provides an overview of (i) neutron production cross sections and thick target yields missing from the EXFOR database; (ii) codes for neutron spectra calculations; (iii) informational resources for reactor, radioactive and spallation neutron sources; (iv) codes for spectrum unfolding and (v) EXFOR compilation rules for the Maxwellian averaged cross sections measured for the reactor and astrophysical applications. (author)

  16. A fast and flexible reactor physics model for simulating neutron spectra and depletion in fast reactors

    Science.gov (United States)

    Recktenwald, Geoff; Deinert, Mark

    2010-03-01

    Determining the time dependent concentration of isotopes within a nuclear reactor core is central to the analysis of nuclear fuel cycles. We present a fast, flexible tool for determining the time dependent neutron spectrum within fast reactors. The code (VBUDS: visualization, burnup, depletion and spectra) uses a two region, multigroup collision probability model to simulate the energy dependent neutron flux and tracks the buildup and burnout of 24 actinides, as well as fission products. While originally developed for LWR simulations, the model is shown to produce fast reactor spectra that show high degree of fidelity to available fast reactor benchmarks.

  17. Optimizing a neutron-beam focusing device for the direct geometry time-of-flight spectrometer TOFTOF at the FRM II reactor source

    Science.gov (United States)

    Rasmussen, N. G.; Simeoni, G. G.; Lefmann, K.

    2016-04-01

    A dedicated beam-focusing device has been designed for the direct geometry thermal-cold neutron time-of-flight spectrometer TOFTOF at the neutron facility FRM II (Garching, Germany). The prototype, based on the compressed Archimedes' mirror concept, benefits from the adaptive-optics technology (adjustable supermirror curvature) and the compact size (only 0.5 m long). We have simulated the neutron transport across the entire guide system. We present a detailed computer characterization of the existing device, along with the study of the factors mostly influencing the future improvement. We have optimized the simulated prototype as a function of the neutron wavelength, accounting also for all relevant features of a real instrument like the non-reflecting side edges. The results confirm the "chromatic" displacement of the focal point (flux density maximum) at fixed supermirror curvature, and the ability of a variable curvature to keep the focal point at the sample position. Our simulations are in excellent agreement with theoretical predictions and the experimentally measured beam profile. With respect to the possibility of a further upgrade, we find that supermirror coatings with m-values higher than 3.5 would have only marginal influence on the optimal behaviour, whereas comparable spectrometers could take advantage of longer focusing segments, with particular impact for the thermal region of the neutron spectrum.

  18. Moderator choice for the cold neutron source of the BER II

    International Nuclear Information System (INIS)

    A Cold Neutron Source is foreseen in the modernization program of the Berlin Experimental Reactor BER II. The moderator will be hydrogen at 25 K. The phase and shape of the moderator, its neutron-physical and safety aspects are discussed in detail. (orig.)

  19. Powerful pulsed neutron sources for research with a pulsed magnetic field

    International Nuclear Information System (INIS)

    The prospects for neutron investigations into the magnetic properties of condensed matter with the use of powerful pulsed neutron sources [Japan Spallation Neutron Source (JSNS) (Tokai, Japan), TIRAN (Zababakhin All-Russia Research Institute of Technical Physics, Russian Federal Nuclear Center, Snezhinsk, Russia), Large Hadron Collider (LHC) (CERN)] and pulsed magnetic fields are considered. It is demonstrated that the diffraction measurements of the magnetic states induced by a magnetic field of up to 700-1000 kOe can be performed at the TIRAN reactor and the neutron source that can be developed on the basis of the LHC accelerator.

  20. Digital Square-Wave Frequency Modulated Microwave Sources for a Miniature Optically Pumped Cesium Beam Clock

    Institute of Scientific and Technical Information of China (English)

    CHEN Jingbiao; ZHU Chengjin; LIU Ge; WANG Fengzhi; WANG Yiqiu; YANG Donghai

    2001-01-01

    Three different digital frequencymodulated microwave sources have been designed andapplied to our miniature optically pumped cesiumbeam clock.The main features and their influenceon clock accuracy have been experimentally tested.Itis proved that a digital square-wave frequency modu-lated microwave source using a microprocessor con-trolled direct-digital frequency synthesizer (DDFS)for our miniature optically pumped cesium beamclock works well,the frequency short term stability2 × 10 11/x r and the long term stability 3.5 x 10-13 forone day sample time have been obtained.

  1. Tokamak D-T neutron source models for different plasma physics confinement modes

    Energy Technology Data Exchange (ETDEWEB)

    Fausser, Clement, E-mail: clement.fausser@cea.fr [CEA, DEN, Saclay, DANS/DM2S/SERMA, F-91191 Gif-sur-Yvette (France); Puma, Antonella Li; Gabriel, Franck [CEA, DEN, Saclay, DANS/DM2S/SERMA, F-91191 Gif-sur-Yvette (France); Villari, Rosaria [Associazione EURATOM-ENEA sulla Fusione, Via Enrico Fermi 45, 00044 Frascati, Rome (Italy)

    2012-08-15

    Highlights: Black-Right-Pointing-Pointer HCLL DEMO neutronics is based on plasma physics L-mode, but may use H or A mode. Black-Right-Pointing-Pointer Based on Plasma Physics 0D code, H and A-mode D-T neutron sources formulae are proposed. Black-Right-Pointing-Pointer TRANSGEN code is built to create 2D source maps as input for Monte-Carlo codes. Black-Right-Pointing-Pointer A-mode neutronic impact is compared to L-mode at same power on a HCLL DEMO design. Black-Right-Pointing-Pointer Results show TBR and Me slight changes, contrary to NWL profile: from -22% to +11%. - Abstract: Neutronic studies of European demonstration fusion power plant (DEMO) have been so far based on plasma physics low confinement mode (L-mode). Future tokamaks, nevertheless, may likely use alternative confinement modes such as high or advanced confinement modes (H and A-mode). Based on analytical formulae used in plasma physics, H and A-modes D-T neutron sources formulae are proposed in this paper. For that purpose, a tokamak random neutron source generator, TRANSGEN, has been built generating bidimensional (radial and poloidal) neutron source maps to be used as input for neutronics Monte-Carlo codes (TRIPOLI-4 and MCNP5). The impact of such a source on the neutronic behavior of the European DEMO-2007 Helium-cooled lithium-lead reactor concept has been assessed and compared with previous results obtained using a L-mode neutron source. An A-mode neutron source map from TRANSGEN has been used with the code TRIPOLI-4. Assuming the same fusion power, results show that main reactor global neutronic parameters, e.g. tritium breeding ratio and neutron multiplication factor, evolved slightly when compared to present uncertainties margin. However, local parameters, such as the neutron wall loading (NWL), change significantly compared to L-mode shape: from -22% to +11% for NWL.

  2. Optimizing Laser-accelerated Ion Beams for a Collimated Neutron Source

    Energy Technology Data Exchange (ETDEWEB)

    C.L. Ellison and J. Fuchs

    2010-09-23

    High-flux neutrons for imaging and materials analysis applications have typically been provided by accelerator- and reactor-based neutron sources. A novel approach is to use ultraintense (>1018W/cm2) lasers to generate picosecond, collimated neutrons from a dual target configuration. In this article, the production capabilities of present and upcoming laser facilities are estimated while independently maximizing neutron yields and minimizing beam divergence. A Monte-Carlo code calculates angular and energy distributions of neutrons generated by D-D fusion events occurring within a deuterated target for a given incident beam of D+ ions. Tailoring of the incident distribution via laser parameters and microlens focusing modifies the emerging neutrons. Projected neutron yields and distributions are compared to conventional sources, yielding comparable on-target fluxes per discharge, shorter time resolution, larger neutron energies and greater collimation.

  3. Neutronic predesign tool for fusion power reactors system assessment

    Energy Technology Data Exchange (ETDEWEB)

    Jaboulay, J.-C., E-mail: jean-charles.jaboulay@cea.fr [CEA, DEN, Saclay, DM2S, SERMA, F-91191 Gif-sur-Yvette (France); Li Puma, A. [CEA, DEN, Saclay, DM2S, SERMA, F-91191 Gif-sur-Yvette (France); Martínez Arroyo, J. [ETSEIB, Internship in CEA (Spain)

    2013-10-15

    SYCOMORE, a fusion reactor system code based on a modular approach, is under development at CEA. In this framework, this paper describes a methodology developed to build the neutronic module of SYCOMORE. This neutronic module aims to evaluate main neutronic parameters characterising a fusion reactor (tokamak): tritium breeding ratio, multiplication factor, nuclear heating as a function of the reactor main geometrical parameters (major radius, elongation, etc.), of the radial build, Li enrichment, blanket and shield thickness, etc. It is based on calculations carried out with APOLLO2 and TRIPOLI-4 CEA transport code on simplified 1D and 2D neutronic models. These models are validated versus a more detailed 3D Monte-Carlo model (using TRIPOLI-4). To ease the integration of this neutronic module in SYCOMORE and provide results instantly, a surrogate model that replicates the 1D and 2D neutronic model results was used. Among the different surrogate models types (polynomial interpolation, responses functions, interpolating by Kriging, artificial neural network, etc.) the neural networks were selected for their efficiency and flexibility. The methodology described in this paper to build SYCOMORE neutronic module is devoted to HCLL blanket, but it could be applied to any breeder blanket concept provided that appropriate validation could be carried out.

  4. Thermal-hydraulic studies of the Advanced Neutron Source cold source

    International Nuclear Information System (INIS)

    The Advanced Neutron Source (ANS), in its conceptual design phase at Oak Ridge National Laboratory, was to be a user-oriented neutron research facility producing the most intense steady-state flux of thermal and cold neutrons in the world. Among its many scientific applications, the production of cold neutrons was a significant research mission for the ANS. The cold neutrons come from two independent cold sources positioned near the reactor core. Contained by an aluminum alloy vessel, each cold source is a 410-mm-diam sphere of liquid deuterium that functions both as a neutron moderator and a cryogenic coolant. With nuclear heating of the containment vessel and internal baffling, steady-state operation requires close control of the liquid deuterium flow near the vessel's inner surface. Preliminary thermal-hydraulic analyses supporting the cold source design were performed with heat conduction simulations of the vessel walls and multidimensional computational fluid dynamics simulations of the liquid deuterium flow and heat transfer. This report presents the starting phase of a challenging program and describes the cold source conceptual design, the thermal-hydraulic feasibility studies of the containment vessel, and the future computational and experimental studies that were planned to verify the final design

  5. D-3He fuel cycles for neutron lean reactors

    International Nuclear Information System (INIS)

    The intrinsic potential of D-3He as a reactor fuel is investigated for a large range of 3He to D density ratios. A steady-state zero-dimensional reactor model is developed in which much care is attributed to a proper treatment of fast fusion products. Useful ranges of reactor parameters as well as temperature-density windows for driven and ignited operation are identified. Various figures of merit are calculated, such as power densities, net power production, neutron production, tritium load and radiative power. These results suggest several optimistic conclusions about the performance of D-3He as a reactor fuel

  6. PREFACE: Neutrino physics at spallation neutron sources

    Science.gov (United States)

    Avignone, F. T.; Chatterjee, L.; Efremenko, Y. V.; Strayer, M.

    2003-11-01

    Unique because of their super-light masses and tiny interaction cross sections, neutrinos combine fundamental physics on the scale of the miniscule with macroscopic physics on the scale of the cosmos. Starting from the ignition of the primal p-p chain of stellar and solar fusion reactions that signal star-birth, these elementary leptons (neutrinos) are also critical players in the life-cycles and explosive deaths of massive stars and the production and disbursement of heavy elements. Stepping beyond their importance in solar, stellar and supernova astrophysics, neutrino interactions and properties influence the evolution, dynamics and symmetries of the cosmos as a whole. Further, they serve as valuable probes of its material content at various levels of structure from atoms and nuclei to valence and sea quarks. In the light of the multitude of physics phenomena that neutrinos influence, it is imperative to enhance our understanding of neutrino interactions and properties to the maximum. This is accentuated by the recent evidence of finite neutrino mass and flavour mixing between generations that reverberates on the plethora of physics that neutrinos influence. Laboratory experiments using intense neutrino fluxes would allow precision measurements and determination of important neutrino reaction rates. These can then complement atmospheric, solar and reactor experiments that have enriched so valuably our understanding of the neutrino and its repertoire of physics applications. In particular, intermediate energy neutrino experiments can provide critical information on stellar and solar astrophysical processes, along with advancing our knowledge of nuclear structure, sub-nuclear physics and fundamental symmetries. So where should we look for such intense neutrino sources? Spallation neutron facilities by their design are sources of intense neutrino pulses that are produced as a by-product of neutron spallation. These neutrino sources could serve as unique laboratories

  7. Ozone source attribution during a severe photochemical smog episode in Beijing,China

    Institute of Scientific and Technical Information of China (English)

    2009-01-01

    Beijing,the capital of China,frequently suffers from the high levels of ozone in summer.A 3-D regional chemical transport model,the Comprehensive Air Quality Model with extensions(CAMx),has been used to simulate a heavy O3 pollution episode in Beijing during June 26―July 2,2000.Ozone Source Apportionment Technology(OSAT) and Geographic Ozone Assessment Technology(GOAT) were applied to quantify the contributions of the precursor emissions from different regions to O3 concentrations in Beijing,to identify the relative importance of different ways by which regional sources affected the O3 levels in Beijing urban areas,and to investigate the sensitivity of O3 formation to the precursors during the episode.The O3 pollution in Beijing showed a significant spatial distribution with strong regional contribution.The results suggested that the plume originating from Beijing urban areas greatly affected the O3 concentrations at the Dingling site,accounting for 55% of elevated O3 there,while O3 pollution in the Beijing urban areas resulted from both local emissions and those from Tianjin and the south of Hebei Province.Transport of O3 was responsible for about 70% of the regional O3 contribution to Beijing urban areas,while transport of O3 precursors accounted for the remainder.The formation of O3 was limited by volatile organic compounds(VOCs) in the urban areas of Beijing,while being more sensitive to NOx levels in the suburban and more remote areas.Therefore,it is necessary to consider a large number of factors,including impacts of emissions from different regions,the two modes of regional contribution as well as the sensitivity of O3 formation to precursors,in the design of emissions control strategies for O3 reduction in Beijing.

  8. Ozone source attribution during a severe photochemical smog episode in Beijing,China

    Institute of Scientific and Technical Information of China (English)

    WANG XueSong; LI JinLong; ZHANG YuanHang; XIE ShaoDong; TANG XiaoYan

    2009-01-01

    Beijing,the capital of China,frequently suffers from the high levels of ozone in summer.A 3-D regional chemical transport model,the Comprehensive Air Quality Model with extensions (CAMx),has been used to simulate a heavy O3 pollution episode in Beijing during June 26-July 2,2000.Ozone Source Apportionment Technology (OSAT) and Geographic Ozone Assessment Technology (GOAT) were applied to quantify the contributions of the precursor emissions from different regions to O3 concentrations in Beijing,to identify the relative importance of different ways by which regional sources affected the O3 levels in Beijing urban areas,and to investigate the sensitivity of O3 formation to the precursors during the episode.The O3 pollution in Beijing showed a significant spatial distribution with strong regional contribution.The results suggested that the plume originating from Beijing urban areas greatly affected the O3 concentrations at the Dingling site,accounting for 55% of elevated O3 there,while O3 pollution in the Beijing urban areas resulted from both local emissions and those from Tianjin and the south of Hebei Province.Transport of O3 was responsible for about 70% of the regional O3 contribution to Beijing urban areas,while transport of O3 precursors accounted for the remainder.The formation of O3 was limited by volatile organic compounds (VOCs) in the urban areas of Beijing,while being more sensitive to NO,levels in the suburban and more remote areas.Therefore,it is necessary to consider a large number of factors,including impacts of emissions from different regions,the two modes of regional contribution as well as the sensitivity of O3 formation to precursors,in the design of emissions control strategies for O3 reduction in Beijing.

  9. Neutron science opportunities at pulsed spallation neutron sources

    International Nuclear Information System (INIS)

    Using the IPNS Upgrade plan developed at Argonne National Laboratory as a worked example of the design of a pulsed spallation neutron source, this paper explores some of the scientific applications of an advanced facility for materials science studies and the instrumentation for those purposes

  10. Nitrogen determination in wheat by neutron activation analysis using fast neutron flux from a thermal nuclear reactor

    International Nuclear Information System (INIS)

    This is a study of the technique for the determination of nitrogen and other elements in wheat flour through activation analysis with fast neutrons from a thermal nuclear reactor. The study begins with an introduction about the basis of the analytical methods, the equipment used in activation analysis and a brief description of the neutrons source. In the study are included the experiments carried out in order to determine the flux form in the site of irradiation, the N-13 half life and the interference due to the sample composition. (author)

  11. Inelastic neutron scattering spectrometer for the IN-06 neutron source at the Moscow meson factory

    International Nuclear Information System (INIS)

    In the spectrometers in which the so-called inverse geometry of scattering is used, the scattering neutron energy, E2, is fixed and the initial neutron energy, E1, is determined by source to sample time of flight. The inverse geometry method is a unique reliable means for determination of the absolute intensity of spectral lines. It is possible because this method allows experiments in a wide range transferred energies to be carried out without changing the experiment geometry. Such spectrometers possess a good definite dependence of the transfer energy on momentum transfer. This allows one to extract complementary to optical methods information from the experimental data. The KDSOG-M spectrometer was created at the IBR-2 pulsed reactor of JINR for investigation of the lattice dynamics of solids and simultaneous analysis of the phase structure of samples. For analysis of the scattering neutron energy polycrystal filters and single crystals are used. Creation of the IN-06 neutron source based on the proton accelerator of the Moscow meson factory leads to the necessary formation of a scientific research program for condensed matter physics. Re-equipment of the KDSOG-M spectrometer for use at the IN-06 neutron source includes preparation of an instrumental basis for carrying out experiments in solid state physics and other fields (biophysics, applied science and so on). This work is a project for transfer and updating the inverse geometry inelastic neutron scattering spectrometer KDSOG-M for use at the IN-06 neutron source of Moscow meson factory. (author) 10 figs., 4 refs

  12. Investigation of water films on fuel rods in boiling water reactors using neutron tomography

    Energy Technology Data Exchange (ETDEWEB)

    Lanthen, Jonas

    2006-09-15

    In a boiling water reactor, thin films of liquid water around the fuel rods play a very important role in cooling the fuel, and evaporation of the film can lead to fuel damage. If the thickness of the water film could be measured accurately the reactor operation could be both safer and more economical. In this thesis, the possibility to use neutron tomography, to study thin water films on fuel rods in an experimental nuclear fuel set-up, has been investigated. The main tool for this has been a computer simulation software. The simulations have shown that very thin water films, down to around 20 pm, can be seen on fuel rods in an experimental set-up using neutron tomography. The spatial resolution needed to obtain this result is around 300 pm. A suitable detector system for this kind of experiment would be plastic fiber scintillators combined with a CCD camera. As a neutron source it would be possible to use a D-D neutron generator, which generates neutrons with energies of 2.5 MeV. Using a neutron generator with a high enough neutron yield and a detector with high enough detection efficiency, a neutron tomography to measure thin water films should take no longer than 25 - 30 minutes.

  13. A Small-Animal Irradiation Facility for Neutron Capture Therapy Research at the RA-3 Research Reactor

    International Nuclear Information System (INIS)

    The National Atomic Energy Commission of Argentina (CNEA) has constructed a thermal neutron source for use in Boron Neutron Capture Therapy (BNCT) applications at the RA-3 research reactor facility located in Buenos Aires. The Idaho National Laboratory (INL) and CNEA have jointly conducted some initial neutronic characterization measurements for one particular configuration of this source. The RA-3 reactor (Figure 1) is an open pool type reactor, with 20% enriched uranium plate-type fuel and light water coolant. A graphite thermal column is situated on one side of the reactor as shown. A tunnel penetrating the graphite structure enables the insertion of samples while the reactor is in normal operation. Samples up to 14 cm height and 15 cm width are accommodated

  14. Advanced in the neutron feedback ICF reactor concept

    International Nuclear Information System (INIS)

    Results are reviewed and updated from an earlier design study of a novel nuclear-pumped flashlamp laser (NP-FL) inertial fusion energy (IFE) power reactor based on the neutron feedback concept for IFE. This concept includes nuclear pumping of the laser flashlamp, a D-T seeded D-3He target and magnetic protection of the first wall of the reactor chamber coupled with direct conversion of deflected charged particles. Advantages include an increased overall plant efficiency due to improved energy coupling via neutron feedback, increased thermal-to-electric energy conversion efficiency, and lower neutron activation and waste. These factors are reflected in a driver energy of 5 MJ and a target gain of only 50 for a 53 % efficient 1000-MWe power plant operating at 6 Hz, novel components involved. However, they require further technological development. Consequently, the NP-FL plant appears to provide a very attractive 'second-generation' IFE reactor. (authors)

  15. Miniaturized LED sources for in vivo optogenetic experimentation

    Science.gov (United States)

    Clements, Isaac P.; Gnade, Andrew G.; Rush, Alexander D.; Patten, Craig D.; Twomey, Mark C.; Kravitz, Alexxai V.

    2013-03-01

    Recently developed optogenetics techniques have enabled researchers to modulate the activity of specific cell types. As a result, complex neural pathways previously regarded as black boxes can now be directly probed, yielding a steadily increasing understanding of the basic neural circuits that underlie health and disease. For in vivo experimentation, fiber-coupled lasers have traditionally been used to illuminate internal brain regions, via an optical fiber that penetrates through overlying tissue. Though able to deliver intense fiber-coupled light, lasers are costly, bulky, and face limitations in output beam stability and temporal precision during modulated outputs. For experiments on unrestricted, behaving animals, a laser-based system also necessitates the use of fiber optic rotary joints, which come with costs and limitations of their own. Here, we report and characterize an alternative light delivery solution, based on high intensity fiber-coupled LEDs that are miniaturized for placement on the end of custom electrical commutators. This design allows for enhanced control of output light and expanded capabilities for optical stimulation as well as simultaneous electrical neural recordings, as with an optrode array. Temporal response of light outputs and light stability during commutator rotation were assessed. The influence of high current optical control signals on adjacent neural recording channels was also explored. To validate the function of this LED based system in in vivo recording scenarios, chronic stimulation experiments were performed.

  16. Research reactors for the social safety and prosperous neutron use

    International Nuclear Information System (INIS)

    The present status of nuclear reactors in Japan and the world was briefly described in this report. Aiming to construct a background of stable future society dependent on nuclear energy, the necessity to establish an organization for research reactors in Japan was pointed out. There are a total of 468 reactors in the world, but only 248 of them are running at present and most of them are superannuated. In Japan, 15 research reactors are running and 8 of them are under collaborative utilization, but not a few of them have various problems. In the education of atomic energy, a reactor is dispensable for understanding its working principle through practice learning. Furthermore, a research reactor has important roles for development of power reactor in addition to various basic studies such as activation analysis, fission track, biological irradiation, neutron scattering, etc. Application of a reactor has been also progressing in industrial and medical fields. However, operation of the reactors has become more and more difficult in Japan because of a large running cost and a lack of residential consensus for nuclear reactor. Here, the author proposed an establishment of organization of research reactor in order to promote utilization of a reactor in the field of education, rearing of professionals and science and engineering. (M.N.)

  17. Experimental Study on Neutron Radiography Device Based on Reactor

    Institute of Scientific and Technical Information of China (English)

    LU; Jin; PENG; Dan; HAO; Qian; YU; Bo-xiang; LI; Yi-guo

    2012-01-01

    <正>Neutron radiography is a non-destructive testing developing fast recently, which requires stable and proper neutron source with low γ background. Neutrons from In-hospital Neutron Irradiator (IHNI) could meet this requirement. Based on the neutron beams of IHNI, a collimator is designed and built for neutron radiography. The experiment results show that in the case of IHNI working at normal rated power, the neutron flux at the end of the collimator is 1.43×106 cm-2·s-1; The max collimation ratio (L/D) is 58; the γ dose rate is 6.3×106 mSv/s. In a word, the collimator could be used for neutron radiography.

  18. Concept of DT fuel cycle for a fusion neutron source

    International Nuclear Information System (INIS)

    A concept of DT-fusion neutron source (FNS) with the neutron yield higher than 1018 neutrons per second is under design in Russia. Such a FNS is of interest for many applications: 1) basic and applied research (neutron scattering, etc); 2) testing the structural materials for fusion reactors; 3) control of sub-critical nuclear systems and 4) nuclear waste processing (including transmutation of minor actinides). This paper describes the fuel cycle concept of a compact fusion neutron source based on a small spherical tokamak (FNS-ST) with a MW range of DT fusion power and considers the key physics issues of this device. The major and minor radii are ∼0.5 and ∼0.3 m, magnetic field ∼1.5 T, heating power less than 15 MW and plasma current 1-2 MA. The system provides the fuel mixture with equal fractions of D and T (D:T = 1:1) for all FNS technology systems. (authors)

  19. THERMAL NEUTRON INTENSITIES IN SOILS IRRADIATED BY FAST NEUTRONS FROM POINT SOURCES. (R825549C054)

    Science.gov (United States)

    Thermal-neutron fluences in soil are reported for selected fast-neutron sources, selected soil types, and selected irradiation geometries. Sources include 14 MeV neutrons from accelerators, neutrons from spontaneously fissioning 252Cf, and neutrons produced from alp...

  20. Comparing neutronics codes performance in analyzing a fresh-fuelled research reactor core

    International Nuclear Information System (INIS)

    Highlights: • Calculation of neutron fluence rate with different neutronic codes is examined. • MCNP, TRIPOLI and CITATION were used for neutron fluence rate calculations. • The recently converted core of the Portuguese Research Reactor (RPI) was used. • Fresh fuel of low enrichment in U-235 was assumed. • Thermal, epithermal and fast neutron fluence rates were computed. - Abstract: In this paper the relative performance of different simulation approaches is examined, focusing on the neutron fluence rate distribution in a nuclear reactor core. The main scope of the work is to benchmark and validate the neutronics code systems utilized in the Greek Research Reactor (GRR-1) for a high-density Low Enriched Uranium (LEU) core of compact size. For this purpose the recently converted core of the Portuguese Research Reactor (RPI), fueled with fresh, low enrichment in U-235 fuel, was simulated with the stochastic code TRIPOLI and the deterministic code system XSDRN/CITATION. RPI was selected on the basis that it is a similar to GRR-1 pool-type reactor, using same fuel and control rods type, as well as same types of coolant, moderator and reflector. The neutron fluence rate in RPI was computed using each numerical approach with changed approximations. In this frame the stochastic code TRIPOLI was tested using two different nuclear data libraries, i.e., ENDF/B-VI versus JEFF3.1, and two different ways of source definition, i.e., “point sources”, placed in the center of each fuel cell, versus a “distributed source”, where each fuel volume was considered as a neutron source. The deterministic code system XSDRN/CITATION was tested with respect to the definition of the transverse leakages associated to each one-dimensional, user-defined core zone, as analyzed by the XSDRN code in order to provide the zone equivalent cross sections. Thermal, epithermal and fast neutron fluence rates were computed and local values found in a 15 cm segment immediately below the

  1. Concrete enclosure to shield a neutron source

    Energy Technology Data Exchange (ETDEWEB)

    Villagrana M, L. E.; Rivera P, E.; De Leon M, H. A.; Soto B, T. G.; Hernandez D, V. M.; Vega C, H. R., E-mail: emmanuelvillagrana@hotmail.com [Universidad Autonoma de Zacatecas, Unidad Academica de Estudios Nucleares, Apdo. Postal 336, 98000 Zacatecas (Mexico)

    2012-10-15

    In the aim to design a shielding for a {sup 239}PuBe isotopic neutron source several Monte Carlo calculations were carried out using MCNP5 code. First, a point-like source was modeled in vacuum and the neutron spectrum and the ambient dose equivalent were calculated at several distances ranging from 5 up to 150 cm, these calculations were repeated including air, and a 1 x 1 x 1 m{sup 3} enclosure that was shielded with 5, 15, 20, 25, 30, 50 and 80 cm-thick Portland type concrete walls. At all the points located inside the enclosure neutron spectra from 10{sup -8} up 0.5 MeV were the same regardless the distance from the source showing the room-return effect, for energies larger than 0.5 MeV neutron spectra are diminished as the distance increases. Outside the enclosure it was noticed that neutron spectra becomes -softer- as the concrete thickness increases due to reduction of mean neutron energy. With the ambient dose values the attenuation curve in terms of concrete thickness was calculated. (Author)

  2. Neutron beam design for low intensity neutron and gamma-ray radioscopy using small neutron sources

    CERN Document Server

    Matsumoto, T

    2003-01-01

    Two small neutron sources of sup 2 sup 5 sup 2 Cf and sup 2 sup 4 sup 1 Am-Be radioisotopes were used for design of neutron beams applicable to low intensity neutron and gamma ray radioscopy (LINGR). In the design, Monte Carlo code (MCNP) was employed to generate neutron and gamma ray beams suited to LINGR. With a view to variable neutron spectrum and neutron intensity, various arrangements were first examined, and neutron-filter, gamma-ray shield and beam collimator were verified. Monte Carlo calculations indicated that with a suitable filter-shield-collimator arrangement, thermal neutron beam of 3,900 ncm sup - sup 2 s sup - sup 1 with neutron/gamma ratio of 7x10 sup 7 , and 25 ncm sup - sup 2 s sup - sup 1 with very large neutron/gamma ratio, respectively, could be produced by using sup 2 sup 5 sup 2 Cf(122 mu g) and a sup 2 sup 4 sup 1 Am-Be(37GBq)radioisotopes at the irradiation port of 35 cm from the neutron sources.

  3. Influencing domain of peripheral sources in the urban heavy pollution process of Beijing

    Institute of Scientific and Technical Information of China (English)

    XU; Xiangde; ZHOU; Li; ZHOU; Xiuji; YAN; Peng; WENG; Yonghu

    2005-01-01

    The effect of city's peripheral pollution sources is one of the key issues urgent to be solved in the decision-making of Beijing's environmental pollution control. This paper comprehensively analyses the surface observations, and the satellite remote sensing data of Moderate Resolution Imaging Spectroradiometer (MODIS) and Total Ozone Mapping Spectrometer (TOMS) during the Beijing City Atmospheric Pollution Experiment (BECAPEX) from January to March, 2001, presents an "upstream" wind field resultant vector method for tracing peripheral pollution sources, and finds that the features of the urban heavy pollution processes of Beijing are significantly correlated with the impact of the emission sources of southern peripheral cities, and the pollutants transferred northwards from distant upstream sources are retarded by the U-shaped "valley" topography in Beijing's periphery. The two factors are responsible for the formation of the S-N zonal influencing domain of pollutants from the southern peripheral areas to Beijing. The paper also comprehensively analyses the features of flow field in the heavy pollution process in the Beijing region, and compares the heavy pollution process with samples of good air-quality days from January to March, 2001. The experiment of Hybrid Single-Particle Lagrangian Integrated Trajectory model (HYSPLIT-4) further reveals the diffusion trajectory of pollutants of the cities in Hebei and Shandong provinces and Tianjin city in the heavy pollution process of Beijing, and the simulations of the Regional Atmosphere Model System (RAMS) confirm the possible contribution of peripheral sources to the exceptionally heavy pollution process of the urban area of Beijing, thus revealing that the input of pollutants from southern peripheral cities is one of the important factors responsible for aggravating urban heavy pollution processes.

  4. Neutron intensity of fast reactor spent fuel

    Energy Technology Data Exchange (ETDEWEB)

    Takamatsu, Misao; Aoyama, Takafumi [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center

    1998-03-01

    Neutron intensity of spent fuel of the JOYO Mk-II core with a burnup of 62,500 MWd/t and cooling time of 5.2 years was measured at the spent fuel storage pond. The measured data were compared with the calculated values based on the JOYO core management code system `MAGI`, and the average C/E approximately 1.2 was obtained. It was found that the axial neutron intensity didn`t simply follow the burnup distribution, and the neutron intensity was locally increased at the bottom end of the fuel region due to an accumulation of {sup 244}Cm. (author)

  5. Resolution in deep inelastic neutron scattering using pulsed neutron sources

    International Nuclear Information System (INIS)

    The principle components of the resolution function for deep inelastic neutron scattering experiments on pulsed neutron sources have been calculated directly in atomic momentum space. Analytical expressions for the relative contributions from the energy, angular and time resolutions are presented for both direct and indirect geometry spectrometers. The general trend in the behaviour of the resolution as a function of neutron energy and atomic mass is presented, and the results of numerical calculations for recoil scattering from hydrogen, helium and beryllium using the ISIS spectrometers HET and eVS, are given. It is shown that the resolution difference between HET and eVS is significantly reduced when compared in atomic momentum space rather than in energy space. Moreover, the contribution from the angular resolution term is only significant for atomic masses <4 au. (author)

  6. Neutron radiography and tomography facility at IBR-2 reactor

    Science.gov (United States)

    Kozlenko, D. P.; Kichanov, S. E.; Lukin, E. V.; Rutkauskas, A. V.; Belushkin, A. V.; Bokuchava, G. D.; Savenko, B. N.

    2016-05-01

    An experimental station for investigations using neutron radiography and tomography was developed at the upgraded high-flux pulsed IBR-2 reactor. The 20 × 20 cm neutron beam is formed by the system of collimators with the characteristic parameter L/D varying from 200 to 2000. The detector system is based on a 6LiF/ZnS scintillation screen; images are recorded using a high-sensitivity video camera based on the high-resolution CCD matrix. The results of the first neutron radiography and tomography experiments at the developed facility are presented.

  7. Fast neutron capture with a white neutron source

    International Nuclear Information System (INIS)

    A system has been developed at the Los Alamos National Laboratory to measure gamma-rays following fast neutron reactions. The neutron beam is produced by bombarding a thick tantalum target with the 800 MeV proton beam from the LAMPF accelerator. Incident neutron energies, from 1 to over 200 MeV, are determined by their times of flight over a 7.6-m flight path. The gamma-rays are detected in five 7.6 x 7.6-cm cylindrical bismuth germanate (BGO) detectors which span an angular range from 450 to 1450 in the reaction plane. With this system it is possible to simultaneously measure the cross section and angular distribution of gamma-rays as a function of neutron energy. The results for the cross section of the 12C(n,n'γ=4.44 MeV) reaction at 900 and 1250 show good agreement with previous measurements while the complete angular distributions show the need for a large a4 coefficient which was not previously observed. Preliminary results for the 12C(n,n'γ=15.1 MeV) reaction have also been obtained. The data obtained for the 40Ca(n,γ0) reaction in the region of the giant dipole resonance demonstrate the unique capabilities of this system. Future developments to the neutron source which will enhance the capabilities of the system are presented. 14 references

  8. Study of the RP-10 reactor neutron beam applied to the neutron radiography

    International Nuclear Information System (INIS)

    We have studied the RP-10 reactor radial neutron beam No. 3, which is used for neutron radiographies, by comparing radiograph's with and without the inner duct, and neutron flux determination with in flakes along the external duct, being the presence of photons creating signals at comparable levels of neutron effects, which reduce the quality of the analysis, values around 106 and 104 n/cm2s for thermal and epithermal flux were obtained respectively. It is recommended evaluate the design of the internal duct which presents strong photon emission. (authors).

  9. Time-of-flight diffraction at pulsed neutron sources: An introduction to the symposium

    International Nuclear Information System (INIS)

    In the 25 years since the first low-power demonstration experiments, pulsed neutron sources have become as productive as reactor sources for many types of diffraction experiments. The pulsed neutron sources presently operating in the United States, England, and Japan offer state of the art instruments for powder and single crystal diffraction, small angle scattering, and such specialized techniques as grazing-incidence neutron reflection, as well as quasielastic and inelastic scattering. In this symposium, speakers review the latest advances in diffraction instrumentation for pulsed neutron sources and give examples of some of the important science presently being done. In this introduction to the symposium, I briefly define the basic principles of pulsed neutron sources, review their development, comment in general terms on the development of time-of-flight diffraction instrumentation for these sources, and project how this field will develop in the next ten years

  10. Neutron Capture and the Antineutrino Yield from Nuclear Reactors.

    Science.gov (United States)

    Huber, Patrick; Jaffke, Patrick

    2016-03-25

    We identify a new, flux-dependent correction to the antineutrino spectrum as produced in nuclear reactors. The abundance of certain nuclides, whose decay chains produce antineutrinos above the threshold for inverse beta decay, has a nonlinear dependence on the neutron flux, unlike the vast majority of antineutrino producing nuclides, whose decay rate is directly related to the fission rate. We have identified four of these so-called nonlinear nuclides and determined that they result in an antineutrino excess at low energies below 3.2 MeV, dependent on the reactor thermal neutron flux. We develop an analytic model for the size of the correction and compare it to the results of detailed reactor simulations for various real existing reactors, spanning 3 orders of magnitude in neutron flux. In a typical pressurized water reactor the resulting correction can reach ∼0.9% of the low energy flux which is comparable in size to other, known low-energy corrections from spent nuclear fuel and the nonequilibrium correction. For naval reactors the nonlinear correction may reach the 5% level by the end of cycle.

  11. Neutron Capture and the Antineutrino Yield from Nuclear Reactors

    Science.gov (United States)

    Huber, Patrick; Jaffke, Patrick

    2016-03-01

    We identify a new, flux-dependent correction to the antineutrino spectrum as produced in nuclear reactors. The abundance of certain nuclides, whose decay chains produce antineutrinos above the threshold for inverse beta decay, has a nonlinear dependence on the neutron flux, unlike the vast majority of antineutrino producing nuclides, whose decay rate is directly related to the fission rate. We have identified four of these so-called nonlinear nuclides and determined that they result in an antineutrino excess at low energies below 3.2 MeV, dependent on the reactor thermal neutron flux. We develop an analytic model for the size of the correction and compare it to the results of detailed reactor simulations for various real existing reactors, spanning 3 orders of magnitude in neutron flux. In a typical pressurized water reactor the resulting correction can reach ˜0.9 % of the low energy flux which is comparable in size to other, known low-energy corrections from spent nuclear fuel and the nonequilibrium correction. For naval reactors the nonlinear correction may reach the 5% level by the end of cycle.

  12. The neutron texture diffractometer at the China Advanced Research Reactor

    Science.gov (United States)

    Li, Mei-Juan; Liu, Xiao-Long; Liu, Yun-Tao; Tian, Geng-Fang; Gao, Jian-Bo; Yu, Zhou-Xiang; Li, Yu-Qing; Wu, Li-Qi; Yang, Lin-Feng; Sun, Kai; Wang, Hong-Li; Santisteban, J. r.; Chen, Dong-Feng

    2016-03-01

    The first neutron texture diffractometer in China has been built at the China Advanced Research Reactor, due to strong demand for texture measurement with neutrons from the domestic user community. This neutron texture diffractometer has high neutron intensity, moderate resolution and is mainly applied to study texture in commonly used industrial materials and engineering components. In this paper, the design and characteristics of this instrument are described. The results for calibration with neutrons and quantitative texture analysis of zirconium alloy plate are presented. The comparison of texture measurements with the results obtained in HIPPO at LANSCE and Kowari at ANSTO illustrates the reliability of the texture diffractometer. Supported by National Nature Science Foundation of China (11105231, 11205248, 51327902) and International Atomic Energy Agency-TC program (CPR0012)

  13. The cold neutron source in DR 3

    International Nuclear Information System (INIS)

    A description of the cold neutron source in DR 3 is given. The moderator of the cold neutron source is supercritical hydrogen at about 30degK and 15 bar abs. The necessary cooling capacity is supplied by two Philips Stirling B20 cryogenerators. The hydrogen is circulated between the cryogenerators and the in-pile moderator chamber by small fans. The safety of the facility is based on the use of triple containment preventing contact between hydrogen and air. The triple containment is achieved by enclosing the high vacuum system, surrounging the hydrogen system, in a helium blanket. The achieved spectrum of the thermal neutron flux and the gain factor are given as well as the experience from more than 5 years of operation. Finally some work on extension of the facility to operate two cold sources is reported. (author)

  14. Neutron radiography applications in I.T.U. TRIGA Mark-II reactor

    International Nuclear Information System (INIS)

    Neutron radiography is an important radiographic technique which is supplied different and advanced information according to the X or gamma ray radiography. However, it has a trouble for supplying the convenient neutron sources. Tangential beam tube of Istanbul Technical University (ITU) TRIGA Mark-II Training and Research Reactor has been arranged for using neutron radiography. The neutron radiography set defined as detailed for the application of the technique. Two different techniques for neutron radiography are defined as namely, transfer method and direct method. For the transfer method dysprosium and indium screens are used in the study. But, dysprosium generally was preferred in many studies in the point of view nuclear safety. Gadolinium was used for direct method. Two techniques are compared and explained the preferring of the transfer technique. Firstly, reference composition is prepared for seeing the differences between neutron and X-ray or gamma radiography. In addition of it, some radiograph samples are given neutron and X-ray radiography which shows the different image characters. Lastly, some examples are given from archaeometric studies. One of them the brass plates of Great Mosque door in Cizre. After the neutron radiography application, organic dye traces are noticed. Other study is on a sword that belong to Urartu period at the first millennium B.C. It is seen that some wooden part on it. Some different artefacts are examined with neutron radiography from the Ikiztepe excavation site, then some animal post parts are recognized on them. One of them is sword and sheath which are corroded together. After the neutron radiography application, it can be noticed that there are a cloth between the sword and its sheath. By using neutron radiography, many interesting and detailed results are observed in ITU TRIGA Mark-II Training and Research Reactor. Some of them shouldn't be recognised by using any other technique

  15. Experimental neutronic science and instrumentation: from hybrid reactors to fourth generation reactors

    International Nuclear Information System (INIS)

    After an overview of his academic career and scientific and research activities, the author proposes a rather detailed synthesis and overview of his scientific activities in the fields of cross sections and Doppler effect (development and validation of a code), on the MUSE-4 hybrid reactor (experiments, static and dynamic measurements), on the TRADE hybrid reactor (experimental means, sub-critical reactivity measurement), on the RACE hybrid reactor (experimental results, modelling and interpretation), and on neutron detection (design and modelling of fission chamber, on-line measurement of the fast flow). The next part gives an overview of some research programs (neutron monitoring in sodium-cool fast reactors, research and development on fission chambers, improvement of effective delayed neutron measurements)

  16. Neutron Imaging Using Neutrons From TRIGA MARK II PUSPATI Reactor

    International Nuclear Information System (INIS)

    This article reports about the implementation of neutron imaging work utilizing neutron beam from TRIGA MARK II PUSPATI collimation channels. Two methods have been implemented namely radiography and tomography. Advantage of these methods is the fact that, radiograms are obtained from normal radiographic imaging methodology and they are the projections used for tomographic image reconstruction. Therefore, both radiogram and tomogram are obtained consecutively. The method has been implemented on the round robin test sample for contrast and resolution measurement and also to some archaeological objects. (author)

  17. Development of cold neutron source

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Chang Oong; Cho, M. S.; Park, K. N. and others

    1999-05-01

    The purpose of this study is to develop the CNS facility in Hanaro to extend the scope of the neutron utilization and to carry out the works impossible by thermal neutrons. According to the project schedule, the establishment of the CNS concept and the basic design are performed in the phase 1, and the elementary technologies for basic design will be developed in the phase 2. Finally in the phase 3, the design of CNS will be completed, and the fabrication, the installation will be ended and then the development plan of spectrometers will be decided to establish the foothold to carry out the basic researches. This study is aimed to produce the design data and utilize them in the future basic and detail design, which include the estimation and the measurement of the heat load, the code development for the design of the in pile assembly and the heat removal system, the measurement of the shape of the CN hole, the performance test of thermosiphon and the concept of the general layout of the whole system etc.. (author)

  18. Development of cold neutron source

    International Nuclear Information System (INIS)

    The purpose of this study is to develop the CNS facility in Hanaro to extend the scope of the neutron utilization and to carry out the works impossible by thermal neutrons. According to the project schedule, the establishment of the CNS concept and the basic design are performed in the phase 1, and the elementary technologies for basic design will be developed in the phase 2. Finally in the phase 3, the design of CNS will be completed, and the fabrication, the installation will be ended and then the development plan of spectrometers will be decided to establish the foothold to carry out the basic researches. This study is aimed to produce the design data and utilize them in the future basic and detail design, which include the estimation and the measurement of the heat load, the code development for the design of the in pile assembly and the heat removal system, the measurement of the shape of the CN hole, the performance test of thermosiphon and the concept of the general layout of the whole system etc.. (author)

  19. The advanced neutron source research and development plan

    Energy Technology Data Exchange (ETDEWEB)

    Selby, D.L.

    1995-08-01

    The Advanced Neutron Source (ANS) is being designed as a user-oriented neutron research laboratory centered around the most intense continuous beams of thermal and subthermal neutrons in the world (an order of magnitude more intense than beams available from the most advanced existing reactors). The ANS will be built around a new research reactor of 330-MW fission power, producing an unprecedented peak thermal flux of >7 {center_dot} 10{sup 19} {center_dot} m{sup -2} {center_dot} s{sup -1}. Primarily a research facility, the ANS will accommodate more than 1000 academic, industrial, and government researchers each year. They will conduct basic research in all branches of science as well as applied research leading to better understanding of new materials, including high temperature super conductors, plastics, and thin films. Some 48 neutron beam stations will be set up in the ANS beam rooms and the neutron guide hall for neutron scattering and for fundamental and nuclear physics research. There also will be extensive facilities for materials irradiation, isotope production, and analytical chemistry. The top level work breakdown structure (WBS) for the project. As noted in this figure, one component of the project is a research and development (R&D) program (WBS 1.1). This program interfaces with all of the other project level two WBS activities. Because one of the project guidelines is to meet minimum performance goals without relying on new inventions, this R&D activity is not intended to produce new concepts to allow the project to meet minimum performance goals. Instead, the R&D program will focus on the four objectives described.

  20. The advanced neutron source research and development plan

    International Nuclear Information System (INIS)

    The Advanced Neutron Source (ANS) is being designed as a user-oriented neutron research laboratory centered around the most intense continuous beams of thermal and subthermal neutrons in the world (an order of magnitude more intense than beams available from the most advanced existing reactors). The ANS will be built around a new research reactor of 330-MW fission power, producing an unprecedented peak thermal flux of >7 · 1019 · m-2 · s-1. Primarily a research facility, the ANS will accommodate more than 1000 academic, industrial, and government researchers each year. They will conduct basic research in all branches of science as well as applied research leading to better understanding of new materials, including high temperature super conductors, plastics, and thin films. Some 48 neutron beam stations will be set up in the ANS beam rooms and the neutron guide hall for neutron scattering and for fundamental and nuclear physics research. There also will be extensive facilities for materials irradiation, isotope production, and analytical chemistry. The top level work breakdown structure (WBS) for the project. As noted in this figure, one component of the project is a research and development (R ampersand D) program (WBS 1.1). This program interfaces with all of the other project level two WBS activities. Because one of the project guidelines is to meet minimum performance goals without relying on new inventions, this R ampersand D activity is not intended to produce new concepts to allow the project to meet minimum performance goals. Instead, the R ampersand D program will focus on the four objectives described

  1. Use of research reactors for neutron activation analysis. Report of an advisory group meeting

    International Nuclear Information System (INIS)

    Neutron activation analysis (NAA) is an analytical technique based on the measurement of characteristic radiation from radionuclides formed directly or indirectly by neutron irradiation of the material of interest. In the last three decades, neutron activation analysis has been found to be extremely useful in the determination of trace and minor elements in many disciplines. These include environmental analysis applications, nutritional and health related studies, geological as well as material sciences. The most suitable source of neutrons for NAA is a research reactor. There are several application fields in which NAA has a superior position compared to other analytical methods, and there are good prospects in developing countries for long term growth. Therefore, the IAEA is making concerted efforts to promote neutron activation analysis and at the same time to assist developing Member States in better utilization of their research reactors. The purpose of the meeting was to discuss the benefits and the role of NAA in applications and research areas that may contribute towards improving utilization of research reactors. The participants focused on five specific topics: (1) Current trends in NAA; (2) The role of NAA compared to other methods of chemical analysis; (3) How to increase the number of NAA users through interaction with industries, research institutes, universities and medical institutions; (4) How to reduce costs and to maintain quality and reliability; (5) NAA using low power research reactors. Neutron activation analysis in its various forms is still active and there are good prospects in developing countries for long-term growth. This can be achieved by a more effective use of existing irradiation and counting facilities, a better end-user focus, and perhaps marginal improvements in equipment and techniques. Therefore, it is recommended that the Member States provide financial and other assistance to enhance the effectiveness of their laboratories

  2. Miniature, Rugged, Pulsed Laser Source for LIDAR Application Project

    Data.gov (United States)

    National Aeronautics and Space Administration — Princeton Optronics proposes to develop a high energy pulsed laser source based on a novel approach. The approach consists of a technique to combine a large number...

  3. Power source device for reactor recycling pump

    International Nuclear Information System (INIS)

    The device of the present invention prevents occurrence of an accident of a reactor forecast upon spontaneous power stoppage, loss of power source or trip of the reactor. Namely, a AC/DC converter and a DC/AC connector having an AC voltage frequency controller are connected in series between an AC (bus) in the plant and reactor recycling pumps. A DC voltage controller, a superconductive energy storing device and an excitation power source are connected to the input of the DC/AC converter. The control device receives signals of the spontaneous power stoppage, loss of power source or trip of the reactor to maintain the output voltage of the superconductive energy storing device to a predetermined value. Further, the ratio of AC power voltage and the frequency of AC voltage to be supplied to the reactor recycling pumps is constantly varied to control the flow rate of the pump to a predetermined value. With such procedures, a power source device for the reactor recycling pumps compact in size, easy for maintenance and having high reliability can be realized by adopting a static-type superconductive energy storing device as an auxiliary power source for the reactor recycling pumps. (I.S.)

  4. Review of neutronic assessments applied to small reactor core physics

    International Nuclear Information System (INIS)

    In its design division for material test reactors and research reactors, AREVA TA has to characterize these manufactured cores. This step is sequential with neutronics benchmarks associated with validation (standard Verification and Validation approach). The previous two points are embedded in core projects and can be run separately especially when experimental tests are foreseen for validation database enrichment. Methodological standard is given in order to match validation and benchmark process illustrated alongside with two specific items on critical research reactors (AZUR - JHR) and subcritical mock up (AZUR). (author)

  5. Measurements of effective delayed neutron fraction in a fast neutron reactor using the perturbation method

    Science.gov (United States)

    Hao-Jun, Zhou; Yan-Peng, Yin; Xiao-Qiang, Fan; Zheng-Hong, Li; Yi-Kang, Pu

    2016-06-01

    A perturbation method is proposed to obtain the effective delayed neutron fraction β eff of a cylindrical highly enriched uranium reactor. Based on reactivity measurements with and without a sample at a specified position using the positive period technique, the reactor reactivity perturbation Δρ of the sample in β eff units is measured. Simulations of the perturbation experiments are performed using the MCNP program. The PERT card is used to provide the difference dk of effective neutron multiplication factors with and without the sample inside the reactor. Based on the relationship between the effective multiplication factor and the reactivity, the equation β eff = dk/Δρ is derived. In this paper, the reactivity perturbations of 13 metal samples at the designable position of the reactor are measured and calculated. The average β eff value of the reactor is given as 0.00645, and the standard uncertainty is 3.0%. Additionally, the perturbation experiments for β eff can be used to evaluate the reliabilities of the delayed neutron parameters. This work shows that the delayed neutron data of 235U and 238U from G.R. Keepin’s publication are more reliable than those from ENDF-B6.0, ENDF-B7.0, JENDL3.3 and CENDL2.2. Supported by Foundation of Key Laboratory of Neutron Physics, China Academy of Engineering Physics (2012AA01, 2014AA01), National Natural Science Foundation (11375158, 91326104)

  6. Aerosol composition, sources and processes during wintertime in Beijing, China

    Directory of Open Access Journals (Sweden)

    Y. L. Sun

    2013-01-01

    Full Text Available Air pollution is a major environmental concern among all seasons in megacity Beijing, China. Here we present the results from a winter study that was conducted from 21 November 2011 to 20 January 2012 with an Aerodyne Aerosol Chemical Speciation Monitor (ACSM and various collocated instruments. The non-refractory submicron aerosol (NR-PM1 species vary dramatically with clean periods and pollution episodes alternating frequently. Compared to summer, wintertime submicron aerosols show much enhanced organics and chloride, which on average account for 52% and 5%, respectively of the total NR-PM1 mass. All NR-PM1 species show quite different diurnal behaviors between summer and winter. For example, the wintertime nitrate presents a gradual increase during daytime and correlates well with secondary organic aerosol (OA, indicating a dominant role of photochemical production over gas-particle partitioning. Positive matrix factorization was performed on ACSM OA mass spectra, and identified three primary OA (POA factors, i.e. hydrocarbon-like OA (HOA, cooking OA (COA, and coal combustion OA (CCOA, and one secondary factor, i.e. oxygenated OA (OOA. The POA dominates OA during wintertime, contributing 69% with the rest of 31% being SOA. Further, all POA components show pronounced diurnal cycles with the highest concentrations occurring at nighttime. CCOA is the largest primary source during the heating season, on average accounting for 33% of OA and 17% of NR-PM1. CCOA also plays a significant role in chemically-resolved particulate matter (PM pollution as its mass contribution increases linearly as a function of NR-PM1 mass loadings. The SOA however presents a reversed trend, which might indicate the limited SOA formation during high PM pollution episodes in winter. The effects of meteorology on PM pollution and aerosol processing were also explored. In particular, the sulfate mass is largely enhanced

  7. Neutron dosimetry and damage calculations for the TRIGA MARK-II reactor in Vienna

    Science.gov (United States)

    Weber, H. W.; Böck, H.; Unfried, E.; Greenwood, L. R.

    1986-02-01

    In order to improve the source characterization of the reactor, especially for recent irradiation experiments in the central irradiation thimble, neutron activation experiments were made on 16 nuclides and the neutron flux spectrum was adjusted using the computer code STAY'SL. The results for the total, thermal and fast neutron flux density at a reactor power of 250 kW are as follows: 2.1 × 10 17, 6.1 × 10 16 ( E 0.1 MeV) and 4.0 × 10 16 ( E > 1 MeV) m -2 s -1. respectively. Calculated damage energy cross sections and gas production rates are presented for selected elements.

  8. Technology and use of low power research reactors

    International Nuclear Information System (INIS)

    The report contains a summary of discussions and 10 papers presented at the Consultants' Meeting on the Technology and Use of Low Power Research Reactors organized by the IAEA and held in Beijing (China) during 30 April - 3 May 1985. The following topics have been covered: reactor utilization in medicine and biology, in universities, for training, as a neutron source for radiography and some remarks on the safety of low power research reactors. A separate abstract was prepared for each paper presented at the meeting

  9. Neutron capture and the antineutrino yield from nuclear reactors

    CERN Document Server

    Huber, Patrick

    2015-01-01

    We identify a new, flux-dependent correction to the antineutrino spectrum as produced in nuclear reactors. The abundance of certain nuclides, whose decay chains produce antineutrinos above the threshold for inverse beta decay, has a nonlinear dependence on the neutron flux, unlike the vast majority of antineutrino producing nuclides, whose decay rate is directly related to the fission rate. We have identified four of these so-called nonlinear nuclides and determined that they result in an antineutrino excess at low-energies below 3.2MeV, dependent on the reactor thermal neutron flux. We develop an analytic model for the size of the correction and compare it to the results of detailed reactor simulations for various real existing reactors, spanning 3 orders of magnitude in neutron flux. In a typical pressurized water reactor the resulting correction can reach 0.9% of the low energy flux which is comparable in size to other, known low-energy corrections from spent nuclear fuel and the non-equilibrium correction...

  10. Neutron gamma fraction imaging: Detection, location and identification of neutron sources

    International Nuclear Information System (INIS)

    In this paper imaging of neutron sources and identification and separation of a neutron source from another neutron source is described. The system is based upon organic liquid scintillator detector, tungsten collimator, bespoke fast digitiser and adjustable equatorial mount. Three environments have been investigated with this setup corresponding to an AmBe neutron source, a 252Cf neutron source and both sources together separated in space. In each case, events are detected, digitised, discriminated and radiation images plotted corresponding to the area investigated. The visualised neutron count distributions clearly locate the neutron source and, relative gamma to neutron (or neutron to gamma) fraction images aid in discriminating AmBe sources from 252Cf source. The measurements were performed in the low scatter facility of the National Physical Laboratory, Teddington, UK

  11. 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.

  12. Steady state investigation on neutronics of a molten salt reactor considering the flow effect of fuel salt

    Institute of Scientific and Technical Information of China (English)

    ZHANG Da-Lin; QIU Sui-Zheng; LIU Chang-Liang; SU Guang-Hui

    2008-01-01

    The Molten Salt Reactor (MSR),one of the‘Generation Ⅳ'concepts,is a liquid-fuel reactor,which is different from the conventional reactors using solid fissile materials due to the flow effect of fuel salt.The study on its neutronice considering the fuel salt flow,which is the base of the thermal-hydraulic calculation and safety analysis,must be done.In this paper,the theoretical model on neutronics under steady condition for a single-liquid-fueled MSR is conducted and calculated by numerical method.The neutronics model consists of two group neutron diffusion equations for fast and thermal neutron fluxes,and balance equations for six-group delayed neutron precursors considering the flow effect of fuel salt. The spatial discretization of the above models is based on the finite volume method,and the discretization equations are computed by the source iteration method.The distributions of neutron fluxes and the distributions of the delayed neutron precursors in the core are obtained.The numerical calculated results show that,the fuel salt flow has little effect on the distribution of fast and thermal neutron fluxes and the effective multiplication factor;however,it affects the distribution of the delayed neutron precursors significantly,especially the long-lived one.In addition,it could be found that the delayed neutron precursors influence the nentronics slightly under the steady condition.

  13. Spallation neutron source RF cavity bias system

    International Nuclear Information System (INIS)

    The Spallation Neutron Source r.f. cavity bias system is described under the topic headings: bias system, r.f. cavity, cables, d.c. bias power supply, transistor regulator and control system. Calculation of 4 core 300 mm solid aluminium cable inductance, coaxial shunt frequency response and transistor regulator computed frequency response, are discussed in appendices 1-3. (U.K.)

  14. SUPERCONDUCTING LINAC FOR THE SPALLATION NEUTRON SOURCE

    International Nuclear Information System (INIS)

    The Spallation Neutron Source (SNS) linac is comprised of both normal and superconducting rf (SRF) accelerating structures. The SRF linac accelerates the beam from 186 to 1250 MeV through 117 elliptical, multi-cell niobium cavities. This paper describes the SRF linac architecture, physics design considerations, cavity commissioning, and the expected beam dynamics performance

  15. Optimization of a cold neutron source at the FRG-1

    International Nuclear Information System (INIS)

    This paper describes the optimization of a cold neutron source (CNS) for maximizing the subthermal flux in the energy range 2.27-5.12 MeV at the beam tube of FRG-1 reactor. In addition, the gain by using different reflectors (H2O, C, D2O) around the CNS is discussed. Advantage factors (resulting from the use of CNS) are calculated for various configurations by means of the transport code NEUTRA. The cross-sections for NEUTRA are prepared by the spectral code GGC-4. (orig.)

  16. Study on the transient behaviours of MNSR reactor for control rod withdrawal

    International Nuclear Information System (INIS)

    The transient behaviours of Miniature Neutron Source Reactor MNSR are analyzed and calculated with the reactor thermohydraulics RETRAN-02 program and the reactor physics MARIA program. The obtained event sequence and consequence from the calculation are compared with the experiments. The effective resonance integral for study on Doppler effect is taken into account. The reactivity temperature coefficient weighting factors are computed. The transient parameters related to reactor power peaking, coolant inlet temperatures, outlet temperatures and coolant mass flow, etc. are computed and compared with the experimental results. (6 refs., 2 figs., 5 tabs.)

  17. Oxidative potential and inflammatory impacts of source apportioned ambient air pollution in Beijing.

    Science.gov (United States)

    Liu, Qingyang; Baumgartner, Jill; Zhang, Yuanxun; Liu, Yanju; Sun, Yongjun; Zhang, Meigen

    2014-11-01

    Air pollution exposure is associated with a range of adverse health impacts. Knowledge of the chemical components and sources of air pollution most responsible for these health effects could lead to an improved understanding of the mechanisms of such effects and more targeted risk reduction strategies. We measured daily ambient fine particulate matter (Beijing, and assessed the contribution of its chemical components to the oxidative potential of ambient air pollution using the dithiothreitol (DTT) assay. The composition data were applied to a multivariate source apportionment model to determine the PM contributions of six sources or factors: a zinc factor, an aluminum factor, a lead point factor, a secondary source (e.g., SO4(2-), NO3(2-)), an iron source, and a soil dust source. Finally, we assessed the relationship between reactive oxygen species (ROS) activity-related PM sources and inflammatory responses in human bronchial epithelial cells. In peri-urban Beijing, the soil dust source accounted for the largest fraction (47%) of measured ROS variability. In central Beijing, a secondary source explained the greatest fraction (29%) of measured ROS variability. The ROS activities of PM collected in central Beijing were exponentially associated with in vivo inflammatory responses in epithelial cells (R2=0.65-0.89). We also observed a high correlation between three ROS-related PM sources (a lead point factor, a zinc factor, and a secondary source) and expression of an inflammatory marker (r=0.45-0.80). Our results suggest large differences in the contribution of different PM sources to ROS variability at the central versus peri-urban study sites in Beijing and that secondary sources may play an important role in PM2.5-related oxidative potential and inflammatory health impacts.

  18. Cold moderators for pulsed neutron sources

    International Nuclear Information System (INIS)

    This paper reviews cold moderators in pulsed neutron sources and provides details of the performance of different cold moderator materials and configurations. Analytical forms are presented which describe wavelength spectra and emission time distributions. Several types of cooling arrangements used in pulsed source moderators are described. Choices of materials are surveyed. The author examines some of the radiation damage effects in cold moderators, including the phenomenon of ''burping'' in irradiated cold solid methane. 9 refs., 15 figs., 4 tabs

  19. Survey of Pulsed Neutron Source Methods for Multiplying Media

    International Nuclear Information System (INIS)

    In recent years there have existed two schools of thought on the most effective manner of obtaining measurements of the shutdown reactivity using pulsed neutron generators; these are (i) the conventional pulsed neutron source measurements with a repetitively pulsed source and (ii) methods based on a pseudo-random impulse response technique using cross-correlation between input and output. In both techniques the pertinent information obtained is identical, i.e. ideally both methods serve to determine the response function. The development of pulsed neutron source techniques on thermal systems for the purpose of reactivity measurements is traced from the early efforts of Sjöstrand to the recent (kβℓ) method. In the usual pulsed neutron source method, the Green's function of the subcritical assembly, the reactor response to a delta function source of neutrons, is the sought-after property. The exponential decay, exp(-αt), of the Green's function yields a spatially independent prompt neutron decay constant. The methods by which the reactivity is derived from the ct-measurement, e.g. the a-delayed critical measurement and the recent (kβℓ) method, are discussed. The fundamental modal treatments are examined in the light of the theory of the pulsed neutronsource techniques as developed for the (kβℓ) model. The implications of the pulsed neutron source theory to obtain precise decay constants and suitable data for the analysis of pulsed systems are considered. Experimental work is reviewed that shows the advantages as well as the limitations of the (kβℓ) technique. The use of pseudo-random impulse response methods with cross-correlation between the input and output for the determination of the Green's function of a multiplying assembly is also discussed. It is shown that the information obtained by the pseudo-random method is identical to that obtained from the repetitively pulsed method. Thus, this makes it possible to apply the methods developed for the

  20. Neutron resonance transmission analysis of reactor spent fuel assemblies

    International Nuclear Information System (INIS)

    A method called Neutron Resonance Transmission Analysis (NRTA) is under study which would use a pulsed neutron beam for nondestructive isotopic assay of a complete spent fuel assembly. Neutrons removed from the collimated beam by absorption or scattering in the resonances of the various isotopes in the spent fuel appear as dips in the neutron transmission. The method is completely insensitive to matrix materials such as oxide, fuel cladding, and other structural members. Measurements on spent fuel buttons using the NBS linac as a pulsed neutron source demonstrate a high accuracy capability for the isotopes 234235236238U, 239240241242Pu, 241Am, 243Am, and several fission products. The NRTA method offers high speed and modest operational cost, and it can be implemented with commercially available medical or radiographic γ-ray generators adapted for neutron production. (Auth.)

  1. Ceramics research in a high-energy neutron source

    International Nuclear Information System (INIS)

    The studies on the irradiation effect to ceramics have added much to the basic understanding of their behavior, for example, the amorphous state of ceramics related to radiation-induced metamictization, the radiation-induced strengthening and toughening due to ultrafine defect aggregates, the in situ degradation of electrical resistivity, the role of radiation-induced defects on thermal conductivity and so on. Most of the irradiation testing on ceramics in the fields of structural and thermal properties have been carried out by using fast fission neutrons of about 1 MeV, but if this energy could be significantly changed, the size and nature of damage cascade and the quantity of transmutation gases produced would change. The significance of neutron source parameters, the special test requirement for ceramics such as the use of miniature specimens, the control of test environment, the transient reduction of electrical resistivity and so on are discussed. A special case of ceramic studies is that on new oxide superconductors. These materials can be made into amorphous state at about 1 dpa using 1 MeV electrons, and are considered to be fairly damage-sensitive. (K.I.)

  2. Low-temperature neutron irradiation tests of superconducting magnet materials using reactor neutrons at KUR

    Science.gov (United States)

    Yoshida, M.; Nakamoto, T.; Ogitsu, T.; Xu, Q.; Itahashi, T.; Kuno, Y.; Kuriyama, Y.; Mori, Y.; Qin, B.; Sato, A.; Sato, K.; Yoshiie, T.

    2012-06-01

    Radiation resistant superconducting magnets are required for high intensity particle accelerators and associated secondary particle beamlines, such as the LHC upgrade and the COMET experiment at J-PARC. Expected neutron fluence on the superconducting coils reaches 1021 n/m2 or higher, therefore the magnet should be designed taking into account the irradiation effects. Irradiation tests for superconducting magnet materials have been carried out using reactor neutrons at Kyoto Univ. Research Reactor Institute. As a first step of the experiment, aluminum alloy stabilizer for superconducting cable was exposed to the reactor neutrons at low temperature and the resistance has been measured in situ during neutron exposure. After the irradiation at 12 K-15 K, the sample resistance increase was proportional to the integrated neutron fluence, and reached almost double for a fast-neutron fluence of 2.3×1020 n/m2 (>0.1 MeV). It is also confirmed that the induced resistance is fully recovered by thermal cycling to room temperature. Details of the irradiation test and the prospects are described.

  3. Installation and performance tests of KUR cold neutron source, 2

    International Nuclear Information System (INIS)

    Our cold neutron source is operated by a closed-indirect cooling loop. The hydrogen cryogenic system of the KUR-CNS has been shown a self-regulating characteristic to the thermal disturbances smaller than 30 % of the maximum heat load, which is measured 300 W at 25 K. This power is used to release a nuclear heating. This self-regulating characteristic was confirmed from the amplitude vs frequency curve, so called Bode's diagram, which showed the first order time lag. Due to this property, the liquid level in the moderator cell has been kept almost constant under 5 MW power of the reactor. The measurements of neutron counting rates across the central level of the moderator cell showed that the demanded liquid content was stored in the cell. (author)

  4. Modeling of water radiolysis at spallation neutron sources

    Energy Technology Data Exchange (ETDEWEB)

    Daemen, L.L.; Kanner, G.S.; Lillard, R.S.; Butt, D.P.; Brun, T.O.; Sommer, W.F.

    1998-12-01

    In spallation neutron sources neutrons are produced when a beam of high-energy particles (e.g., 1 GeV protons) collides with a (water-cooled) heavy metal target such as tungsten. The resulting spallation reactions produce a complex radiation environment (which differs from typical conditions at fission and fusion reactors) leading to the radiolysis of water molecules. Most water radiolysis products are short-lived but extremely reactive. When formed in the vicinity of the target surface they can react with metal atoms, thereby contributing to target corrosion. The authors will describe the results of calculations and experiments performed at Los Alamos to determine the impact on target corrosion of water radiolysis in the spallation radiation environment. The computational methodology relies on the use of the Los Alamos radiation transport code, LAHET, to determine the radiation environment, and the AEA code, FACSIMILE, to model reaction-diffusion processes.

  5. Source apportionment for urban PM10 and PM2.5 in the Beijing area

    Institute of Scientific and Technical Information of China (English)

    ZHANG Wei; GUO JingHua; SUN YeLe; YUAN Hui; ZHUANG GuoShun; ZHUANG YaHui; HAO ZhengPing

    2007-01-01

    Airborne particulate matter (PM2.5 and PM10) samples were collected at the Beijing Normal University sampling site in the urban area of Beijing, China in dry and wet seasons during 2001―2004. Concentrations of 23 elements and 14 ions in particulate samples were determined by ICP-AES and IC, respectively. Source apportionment results derived from both Positive Matrix Factorization (PMF) and Chemical Mass Balance (CMB) models indicate that the major contributors of PM2.5 and PM10 in Beijing are: soil dust, fossil fuel combustion, vehicle exhausts, secondary particulate, biomass burning and some industrial sources. We have identified both regional common sources, such as vehicular emissions, particulate of secondary origin and biomass burning, as well as country-specific problems, such as sand storms and soil dust that should be addressed for effective air quality control.

  6. Reactor neutron activation analysis of industrial materials

    International Nuclear Information System (INIS)

    The specific application of neutron activation analysis (n.a.a.) for industrial materials is demonstrated by the determination of impurities in BeO, Al, Si, Cu, Ge, GaP, GaAs, steel, and irradiated uranium. A group scheme gives an orientation about the possibilities of n.a.a. The use of different standards, methods for the measurement of low radioactivities and errors caused by recoil reaction and radiation stimulated diffusion are discussed. (author)

  7. Sweden to host a new neutron source

    CERN Multimedia

    Anaïs Schaeffer

    2012-01-01

    The first European neutron source, currently under development, should commence operations by the end of this decade. Its aim: to produce beams of neutrons that can penetrate into the heart of matter without damaging it and reveal its secrets.   An artist's impression of what the ESS should look like in 2019. At the southern end of Sweden, a town called Lund is preparing for the arrival of the world's most powerful neutron source: the European Spallation Source (ESS). Construction is scheduled to start at the beginning of next year, and the facility is expected to become operational by 2019, when it will produce its first neutron beams. “The ESS is the result of an idea that began 20 years ago!” underlines Mats Lindroos, in charge of the ESS Accelerator Division. “Today, 17 European countries support the project, including Sweden, Denmark and Norway, who together account for 50% of the construction funding.” The ESS, whose design is al...

  8. Neutron dose estimation in a zero power nuclear reactor

    Science.gov (United States)

    Triviño, S.; Vedelago, J.; Cantargi, F.; Keil, W.; Figueroa, R.; Mattea, F.; Chautemps, A.; Santibañez, M.; Valente, M.

    2016-10-01

    This work presents the characterization and contribution of neutron and gamma components to the absorbed dose in a zero power nuclear reactor. A dosimetric method based on Fricke gel was implemented to evaluate the separation between dose components in the mixed field. The validation of this proposed method was performed by means of direct measurements of neutron flux in different positions using Au and Mg-Ni activation foils. Monte Carlo simulations were conversely performed using the MCNP main code with a dedicated subroutine to incorporate the exact complete geometry of the nuclear reactor facility. Once nuclear fuel elements were defined, the simulations computed the different contributions to the absorbed dose in specific positions inside the core. Thermal/epithermal contributions of absorbed dose were assessed by means of Fricke gel dosimetry using different isotopic compositions aimed at modifying the sensitivity of the dosimeter for specific dose components. Clear distinctions between gamma and neutron capture dose were obtained. Both Monte Carlo simulations and experimental results provided reliable estimations about neutron flux rate as well as dose rate during the reactor operation. Simulations and experimental results are in good agreement in every positions measured and simulated in the core.

  9. A neutron tomography facility at a low power research reactor

    CERN Document Server

    Körner, S; Von Tobel, P; Rauch, H

    2001-01-01

    Neutron radiography (NR) provides a very efficient tool in the field of non-destructive testing as well as for many applications in fundamental research. A neutron beam penetrating a specimen is attenuated by the sample material and detected by a two-dimensional (2D) imaging device. The image contains information about materials and structure inside the sample because neutrons are attenuated according to the basic law of radiation attenuation. Contrary to X-rays, neutrons can be attenuated by some light materials, as for example, hydrogen and boron, but penetrate many heavy materials. Therefore, NR can yield important information not obtainable by more traditional methods. Nevertheless, there are many aspects of structure, both quantitative and qualitative, that are not accessible from 2D transmission images. Hence, there is an interest in three-dimensional neutron imaging. At the 250 kW TRIGA Mark II reactor of the Atominstitut in Austria a neutron tomography facility has been installed. The neutron flux at ...

  10. Reactor physics computer code development for neutronic design, fuel-management, reactor operation and safety analysis of PHWRs

    International Nuclear Information System (INIS)

    This report discusses various reactor physics codes developed for neutronic design, fuel-management, reactor operation and safety analysis of PHWRs. These code packages have been utilized for nuclear design of 500 MWe and new 235 MWe PHWRs. (author)

  11. Preliminary design of hybrid energy reactor and integral neutron experiments

    International Nuclear Information System (INIS)

    Highlights: ► We propose a uranium alloy fuel and water coolant hybrid energy reactor blanket. ► Using two-dimensional cylinder model we show the character of the hybrid reactor. ► The ratio of fuel volume to water volume 1.5:1 has better M and TBR results. ► We design integral neutron experiments to verify the credibility of physical design. - Abstract: We propose a preliminary design for a fusion–fission hybrid energy reactor (FFHER), based on current fusion science and technology (with some extrapolations forward from ITER) and well-developed fission technology. We list design rules and put forward a primary concept blanket, with uranium alloy as fuel and water as coolant. The uranium fuel can be natural uranium, LWR spent fuel, or depleted uranium. The FFHER design can increase the utilization rate of uranium in a comparatively simple way to sustain the development of nuclear energy. We study the interaction between the fusion neutron and the uranium fuel with the aim of to achieving greater energy multiplication and tritium sustainability. We also review other concept hybrid reactor designs. We design integral neutron experiments in order to verify the credibility of our proposed physical design. The combination of this program of research with the related thermal hydraulic design, alloy fuel manufacture, and nuclear fuel cycle programs provides the science and technology foundation for the future development of the FFHER concept in China.

  12. Neutronic challenges of advanced boiling water reactor designs

    International Nuclear Information System (INIS)

    The advancement of Boiling Water Reactor technology has been under investigation at the Center for Advance Nuclear Energy Systems at MIT. The advanced concepts under study provide economic incentives through enabling further power uprates (i.e. increasing vessel power density) or better fuel cycle uranium utilization. The challenges in modeling of three advanced concepts with focus on neutronics are presented. First, the Helical Cruciform Fuel rod has been used in some Russian reactors, and studied at MIT for uprating the power in LWRs through increased heat transfer area per unit core volume. The HCF design requires high fidelity 3D tools to assess its reactor physics behavior as well as thermal and fuel performance. Second, an advanced core design, the BWR-HD, was found to promise 65% higher power density over existing BWRs, while using current licensing tools and existing technology. Its larger assembly size requires stronger coupling between neutronics and thermal hydraulics compared to the current practice. Third is the reduced moderation BWRs, which had been proposed in Japan to enable breeding and burning of fuel as an alternative to sodium fast reactors. Such technology suffers from stronger sensitivity of its neutronics to the void fraction than the traditional BWRs, thus requiring exact modeling of the core conditions such as bypass voiding, to correctly characterize its performance. (author)

  13. Sources of radiation from neutron stars

    CERN Document Server

    Schutz, B F

    1998-01-01

    I give a brief introduction to the problem of detecting gravitational radiation from neutron stars. After a review of the mechanisms by which such stars may produce radiation, I consider the different search strategies appropriate to the different kinds of sources: isolated known pulsars, neutron stars in binaries, and unseen neutron stars. The problem of an all-sky survey for unseen stars is the most taxing one that we face in analysing data from interferometers. I describe the kinds of hierarchical methods that are now being investigated to reach the maximal sensitivity, and I suggest a replacement for standard Fourier-transform search methods that requires fewer floating-point operations for Fourier-based searches over large parameter spaces, and in addition is highly parallelizable, working just as well on a loosely coupled network of workstations as on a tightly coupled parallel computer.

  14. A miniaturized, high flux BEC source for precision atom interferometry

    Science.gov (United States)

    Herr, Waldemar; Rudolph, Jan; Popp, Manuel; Rasel, Ernst; Quantus Collaboration

    2013-05-01

    Atom chips have proven to be excellent sources for the fast production of ultra-cold gases due to their outstanding performance in evaporative cooling. However, the total number of atoms has previously been limited by the small volume of their magnetic traps. To overcome this restriction, we have developed a novel loading scheme that allows us to produce Bose-Einstein condensates of a few 105 87Rb atoms every two seconds. The apparatus is designed to be operated in microgravity at the drop tower in Bremen, where even higher numbers of atoms can be achieved in the absence of any gravitational sag. Using the drop tower's catapult mode, our setup will perform atom interferometry during nine seconds in free fall. Thus, the fast loading scheme allows for interferometer sequences of up to seven seconds - interrogation times which are inaccessible for ground based devices. The QUANTUS project is supported by the German Space Agency DLR with funds provided by the Federal Ministry of Economics and Technology (BMWi) under grant number DLR 50WM1131. Leibniz Universitaet Hannover, Universitaet Bremen, HU Berlin, Universitaet Hamburg, Universitaet Ulm, TU Darmstadt, MPQ-Garching.

  15. Practices for Neutronic Design of Research Reactors: Safety and Performances

    International Nuclear Information System (INIS)

    In brief, the design aims to have a facility which is quickly operational and profitable, safe and able to evolve over 40 or 60 years, taking into account both the evolution of the requirements for experiments or production yet to be realized and the safety practices. This paper presents the AREVA current design and safety practices (both cannot be realized without the other) for the neutronic design of the research reactor (RR) cores. It completes the paper and presents the general methodology of neutronic design studies for the safety and performance aspects and only slightly focuses on the reactivity shutdown systems and the neutronic calculation schemes. The main points are illustrated with examples of the Jules Horowitz Reactor (core designer point of view). On this basis of our general methodology, certain problems are separated in order to permit rapid reiteration at an individual level before the final synthesis. For example: to carry out generic studies of fuel management strategies and core reactivity control in order to manage the power peak (need core depletion calculation) and to be able to reason step 0 for certain optimizations of the core geometry and characteristics. For the neutronic calculation scheme, our current practice is to combine the use of the deterministic and stochastic codes. The strong points of each type of code are used to reinforce the safety and the performance of our cores. In this field, AREVA has a R and D framework involving and coordinating the participants from the various sectors (power reactors, research reactor etc) in the development of the general calculation methods and associated tools, in particular for Monte Carlo core depletion calculations. The CEA (along with APOLLO, CRONOS and TRIPOLI codes) largely supports us in this field. Comparisons between MCNP and TRIPOLI and between the various libraries (ENDF, JEF, etc.) are also performed. That includes the recalculation of existing reactors (OSIRIS, ORPHEE, AZUR

  16. Thermophysical properties of saturated light and heavy water for Advanced Neutron Source applications

    Energy Technology Data Exchange (ETDEWEB)

    Crabtree, A.; Siman-Tov, M.

    1993-05-01

    The Advanced Neutron Source is an experimental facility being developed by Oak Ridge National Laboratory. As a new nuclear fission research reactor of unprecedented flux, the Advanced Neutron Source Reactor will provide the most intense steady-state beams of neutrons in the world. The high heat fluxes generated in the reactor [303 MW(t) with an average power density of 4.5 MW/L] will be accommodated by a flow of heavy water through the core at high velocities. In support of this experimental and analytical effort, a reliable, highly accurate, and uniform source of thermodynamic and transport property correlations for saturated light and heavy water were developed. In order to attain high accuracy in the correlations, the range of these correlations was limited to the proposed Advanced Neutron Source Reactor`s nominal operating conditions. The temperature and corresponding saturation pressure ranges used for light water were 20--300{degrees}C and 0.0025--8.5 MPa, respectively, while those for heavy water were 50--250{degrees}C and 0.012--3.9 MPa. Deviations between the correlation predictions and data from the various sources did not exceed 1.0%. Light water vapor density was the only exception, with an error of 1.76%. The physical property package consists of analytical correlations, SAS codes, and FORTRAN subroutines incorporating these correlations, as well as an interactive, easy-to-use program entitled QuikProp.

  17. Test of a prototype neutron spectrometer based on diamond detectors in a fast reactor

    CERN Document Server

    Osipenko, M; Ripani, M; Pillon, M; Ricco, G; Caiffi, B; Cardarelli, R; Verona-Rinati, G; Argiro, S

    2015-01-01

    A prototype of neutron spectrometer based on diamond detectors has been developed. This prototype consists of a $^6$Li neutron converter sandwiched between two CVD diamond crystals. The radiation hardness of the diamond crystals makes it suitable for applications in low power research reactors, while a low sensitivity to gamma rays and low leakage current of the detector permit to reach good energy resolution. A fast coincidence between two crystals is used to reject background. The detector was read out using two different electronic chains connected to it by a few meters of cable. The first chain was based on conventional charge-sensitive amplifiers, the other used a custom fast charge amplifier developed for this purpose. The prototype has been tested at various neutron sources and showed its practicability. In particular, the detector was calibrated in a TRIGA thermal reactor (LENA laboratory, University of Pavia) with neutron fluxes of $10^8$ n/cm$^2$s and at the 3 MeV D-D monochromatic neutron source na...

  18. Polycapillary X-ray lens for the secondary focusing Beijing synchrotron radiation source

    Institute of Scientific and Technical Information of China (English)

    Li Yu-De; Lin Xiao-Yan; Liu Shi-Gang; He Jin-Long; Guo Fei; Sun Tian-Xi; Liu Peng

    2013-01-01

    According to intensity distribution of the synchrotron radiation source focused by a toroidal mirror at the Beijing synchrotron radiation biological macromolecule station,theoretical modeling of the Beijing synchrotron radiation source is developed for capillary optics.Using this theoretical modeling,the influences of the configuration curve of the polycapillary X-ray lens on transmission efficiency and working distance are analyzed.The experimental results of the transmission efficiency and working distance at the biological macromolecule station are in good agreement with the theoretical results.

  19. Upgrading of neutron radiography/tomography facility at research reactor

    International Nuclear Information System (INIS)

    A state-of-the-art neutron tomography imaging system was set up at the neutron radiography beam tube at the Egypt Second Research Reactor (ETRR-2) and was successfully commissioned in 2013. This study presents a set of tomographic experiments that demonstrate a high quality tomographic image formation. A computer technique for data processing and 3D image reconstruction was used to see inside a copy module of an ancient clay article provided by the International Atomic Energy Agency (IAEA). The technique was also able to uncover tomographic imaging details of a mummified fish and provided a high resolution tomographic image of a defective fire valve. (orig.)

  20. Updated neutron spectrum characterization of SNL baseline reactor environments

    International Nuclear Information System (INIS)

    The neutron spectrum characteristics of the primary reactor environments are defined for use by facility customers and to provide an audit trail in support of current quality assurance initiatives. The neutron and gamma environments in the four primary customer environments at SPR-III and ACRR facilities are characterized in detail. Enough detail is provided on other frequently-used environments to support the definition of the 3-MeV and 1-MeV(Si) fluence provided on the Radiation Metrology Laboratory dosimetry reports

  1. Upgrading of neutron radiography/tomography facility at research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Abd El Bar, Waleed; Mongy, Tarek [Atomic Energy Authority, Cairo (Egypt). ETRR-2; Kardjilov, Nikolay [Helmholtz Zentrum Berlin (HZB) for Materials and Energy, Berlin (Germany)

    2014-03-15

    A state-of-the-art neutron tomography imaging system was set up at the neutron radiography beam tube at the Egypt Second Research Reactor (ETRR-2) and was successfully commissioned in 2013. This study presents a set of tomographic experiments that demonstrate a high quality tomographic image formation. A computer technique for data processing and 3D image reconstruction was used to see inside a copy module of an ancient clay article provided by the International Atomic Energy Agency (IAEA). The technique was also able to uncover tomographic imaging details of a mummified fish and provided a high resolution tomographic image of a defective fire valve. (orig.)

  2. Comparison of depth-dose distributions between reactor and accelerator neutron beams proposed by design studies

    International Nuclear Information System (INIS)

    Accelerator epithermal neutron beams produced by 7Li(p,n)7Be reactions were compared with reactor neutron beams using a fission converter (20% enriched 235U 5mm-thick plate) from view points of neutron spectrum and depth-dose distributions in a phantom. It is possible to design accelerator epithermal neutron beams having better depth-dose distributions than reactor neutron beams. (author)

  3. Measurement of subcriticality using delayed neutron source combined with pulsed neutron accelerator

    International Nuclear Information System (INIS)

    A new experimental method for subcriticality measurement was developed by using delayed neutron source which is produced by external pulsed neutron source to increase accuracy of measured results by overcoming the space dependency problem which means difference of measured results in different detector position and often appeared in almost all other subcriticality measurement techniques. Experiments were performed at Kyoto University Critical Assembly (KUCA) combined with a DT accelerator to produce pulsed neutron in outside of the core repeatedly. In this method, neutron detection counts in the prompt neutron time region which are appeared just after injection of pulsed neutron are omitted, whereas neutron counts in the delayed neutron time region which are appeared after disappearance of exponential decay of the prompt neutron are adopted in analysis based on neutron source multiplication method or neutron noise analysis method; the variance to mean ratio method. In the delayed neutron time region, neutron sources to initiate fission chain reactions in subcritical state are delayed neutrons from delayed neutron precursors which are mainly produced by fission chain reactions in the prompt neutron time region, and delayed neutron precursors exist only in the fuel region, which makes possible to decrease the space dependency problem. The obtained results were compared with conventional pulsed neutron method, and it was found that the space dependency problem in subcriticality measurement can be fairly decreased by using the present new method compared with conventional one. (author)

  4. A neutron tomography facility at a low power research reactor

    Science.gov (United States)

    Koerner, S.; Schillinger, B.; Vontobel, P.; Rauch, H.

    2001-09-01

    Neutron radiography (NR) provides a very efficient tool in the field of non-destructive testing as well as for many applications in fundamental research. A neutron beam penetrating a specimen is attenuated by the sample material and detected by a two-dimensional (2D) imaging device. The image contains information about materials and structure inside the sample because neutrons are attenuated according to the basic law of radiation attenuation. Contrary to X-rays, neutrons can be attenuated by some light materials, as for example, hydrogen and boron, but penetrate many heavy materials. Therefore, NR can yield important information not obtainable by more traditional methods. Nevertheless, there are many aspects of structure, both quantitative and qualitative, that are not accessible from 2D transmission images. Hence, there is an interest in three-dimensional neutron imaging. At the 250 kW TRIGA Mark II reactor of the Atominstitut in Austria a neutron tomography facility has been installed. The neutron flux at this beam position is 1.3×10 5 neutrons/cm 2 s and the beam diameter is 8 cm. For a 3D tomographic reconstruction of the sample interior, transmission images of the object taken from different view angles are required. Therefore, a rotary table driven by a step motor connected to a computerized motion control system has been installed at the sample position. In parallel a suitable electronic imaging device based on a neutron sensitive scintillator screen and a CCD-camera has been designed. It can be controlled by a computer in order to synchronize the software of the detector and of the rotary table with the aim of an automation of measurements. Reasonable exposure times can get as low as 20 s per image. This means that a complete tomography of a sample can be performed within one working day. Calculation of the 3D voxel array is made by using the filtered backprojection algorithm.

  5. A neutron tomography facility at a low power research reactor

    International Nuclear Information System (INIS)

    Neutron radiography (NR) provides a very efficient tool in the field of non-destructive testing as well as for many applications in fundamental research. A neutron beam penetrating a specimen is attenuated by the sample material and detected by a two-dimensional (2D) imaging device. The image contains information about materials and structure inside the sample because neutrons are attenuated according to the basic law of radiation attenuation. Contrary to X-rays, neutrons can be attenuated by some light materials, as for example, hydrogen and boron, but penetrate many heavy materials. Therefore, NR can yield important information not obtainable by more traditional methods. Nevertheless, there are many aspects of structure, both quantitative and qualitative, that are not accessible from 2D transmission images. Hence, there is an interest in three-dimensional neutron imaging. At the 250 kW TRIGA Mark II reactor of the Atominstitut in Austria a neutron tomography facility has been installed. The neutron flux at this beam position is 1.3x105 neutrons/cm2 s and the beam diameter is 8 cm. For a 3D tomographic reconstruction of the sample interior, transmission images of the object taken from different view angles are required. Therefore, a rotary table driven by a step motor connected to a computerized motion control system has been installed at the sample position. In parallel a suitable electronic imaging device based on a neutron sensitive scintillator screen and a CCD-camera has been designed. It can be controlled by a computer in order to synchronize the software of the detector and of the rotary table with the aim of an automation of measurements. Reasonable exposure times can get as low as 20 s per image. This means that a complete tomography of a sample can be performed within one working day. Calculation of the 3D voxel array is made by using the filtered backprojection algorithm

  6. The Advanced Neutron Source research and development plan

    International Nuclear Information System (INIS)

    The Advanced Neutron Source (ANS) is being designed as a user-oriented neutron research laboratory centered around the most intense continuous beams of thermal and subthermal neutrons in the world. The ANS will be built around a new research reactor of ∼ 330 MW fission power, producing an unprecedented peak thermal flux of > 7 x 1019 M-2 · S-1. Primarily a research facility, the ANS will accommodate more than 1000 academic, industrial, and government researchers each year. They will conduct basic research in all branches of science-as well as applied research-leading to better understanding of new materials, including high temperature super conductors, plastics, and thin films. Some 48 neutron beam stations will be set up in the ANS beam rooms and the neutron guide hall for neutron scattering and for fundamental and nuclear physics research. There also will be extensive facilities for materials irradiation, isotope production, and analytical chemistry. The R ampersand D program will focus on the four objectives: Address feasibility issues; provide analysis support; evaluate options for improvement in performance beyond minimum requirements; and provide prototype demonstrations for unique facilities. The remainder of this report presents (1) the process by which the R ampersand D activities are controlled and (2) a discussion of the individual tasks that have been identified for the R ampersand D program, including their justification, schedule and costs. The activities discussed in this report will be performed by Martin Marietta Energy Systems, Inc. (MMES) through the Oak Ridge National Laboratory (ORNL) and through subcontracts with industry, universities, and other national laboratories. It should be noted that in general a success path has been assumed for all tasks

  7. Neutron Transmutation Doping of Silicon at Research Reactors

    International Nuclear Information System (INIS)

    This publication details the processes and history of neutron transmutation doping of silicon, particularly its commercial pathway, followed by the requirements for a technologically modern and economically viable production scheme and the current trends in the global market for semiconductor products. It should serve as guidelines on the technical requirements, involved processes and required quality standards for the transmission of sound practices and advice for research reactor managers and operators planning commercial scale production of silicon. Furthermore, a detailed and specific database of most of the world's research reactor facilities in this domain is included, featuring their characteristics for irradiation capabilities, associated production capacities and processing.

  8. Coupled hydro-neutronic calculations for fast burst reactor accidents

    International Nuclear Information System (INIS)

    Methods are described for determining the fully coupled neutronic/hydrodynamic response of fast burst reactors (FBR) under disruptive accident conditions. Two code systems, PAD (1 -D Lagrangian) and NIKE-PAGOSA (3-D Eulerian) were used to accomplish this. This is in contrast to the typical methodology that computes these responses by either single point kinetics or in a decoupled manner. This methodology is enabled by the use of modem supercomputers (CM-200). Two examples of this capability are presented: an unreflected metal fast burst assembly, and a reflected fast burst assembly typical of the Skua or SPR-III class of fast burst reactor

  9. Reference neutron transport calculation note for Korea nuclear power plants with 3-loop PWR reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Byung Cheol; Chang, Ki Oak

    1997-05-01

    Reactor pressure vessel (RPV) steels are subjected to neutron irradiation at a temperature of about 290 deg C. This radiation exposure alters the mechanical properties, leading to a shift of the brittle-to-ductile transition temperature toward higher temperatures and to a diminution of the rupture energy as determined by Charpy V-notch tests. This radiation embrittlement is one of the important aging factors of nuclear power plants. U.S. NRC recommended the basic requirements for the determination of the pressure vessel fluence by regulatory guide DG-1025 in order to reduce the uncertainty in the determination of neutron fluence calculation and measurements. The determination of the pressure vessel fluence is based on both calculations and measurements. The fluence prediction is made with a calculation and the measurements are used to qualify the calculational methodology. Because of the importance and the difficulty of these calculations, the method`s qualification by comparison to measurement must be made to ensure a reliable and accurate vessel fluence determination. This reference calculation note is to provide a series of forward and adjoint neutron transport calculations for use in the evaluation of neutron dosimetry from surveillance capsule irradiations at 3-loop PWR reactor as well as for use in the determination of the neutron exposure of the reactor vessel wall in accordance with U.S Regulatory Guide DG-1025 requirements. The calculations of the pressure vessel fluence consist of the following steps; (1) Determination of the geometrical and material input data, (2) Determination of the core neutron source, and (3) Propagation of the neutron fluence from the core to the vessel and into the cavity. (author). 12 tabs., 3 figs., 7 refs.

  10. Neutron spectrum and radial power distribution measurements in a TRIGA reactor fuel element

    International Nuclear Information System (INIS)

    The neutron spectrum in the Illinois Advanced TRIGA Reactor was measured by a crystal spectrometer utilizing an LiF(1, 1, 1) crystal monochromator whose reflectivity was determined experimentally. The fission heat source distribution in a fuel element was also determined as a function of the fuel element temperature. These two measurements were used to investigate the effects of fuel element temperature and the local core loading on the thermal diffusion length in a fuel element. Changes in the thermal diffusion lengths during a reactor pulse underlie the proposed temperature feedback mechanism for the ZrH fuel material. The results of the measurements confirm, in part, this proposed temperature feedback mechanism

  11. Fundamental design of systems and facilities for cold neutron source in the Hanaro

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Bong Soo; Jeong, H. S.; Kim, Y. K.; Wu, S. I

    2006-01-15

    The CNS(Cold Neutron Source) development project has been carried out as the partial project of the reactor utilization R and D government enterprise since 2003. In the advantage of lower energy and long wave length for the cold neutron, it can be used with the essential tool in order to investigate the structure of protein, amino-acid, DNA, super lightweight composite and advanced materials in the filed of high technology. This report is mainly focused on the basic design of the systems and facilities for the HANARO cold neutron source, performed during the second fiscal project year.

  12. Neutronic study regarding transmutation fuel research at Jules Horowitz Reactor

    International Nuclear Information System (INIS)

    In order to estimate the possibilities for transmutation experiments at the Jules Horowitz Reactor several ideas for neutronic and fuel behaviour studies are investigated at CEA Cadarache. Naturally an exact replication of the burning of minor actinides in fast reactors, as expected in most transmutation scenarios, is impossible, but some key transmutation parameters can be investigated in a MTR neutron spectrum. In this paper a parametric study regarding fuel damage by He and fission products in AmUO2 is presented. By varying flux level, uranium enrichment and americium content of the sample in the JHR reflector a He production to fission ratio comparable to reference samples in the core of a SFR can be achieved. The calculations were done with the depletion code DARWIN2.2 using JEF2.2 data and spectra from a TRIPOLI model of JHR and an ERANOS model for the SFR respectively. (author)

  13. Characterisation of an accelerator-based neutron source for BNCT versus beam energy

    CERN Document Server

    Agosteo, S; D'Errico, F; Nath, R; Tinti, R

    2002-01-01

    Neutron capture in sup 1 sup 0 B produces energetic alpha particles that have a high linear energy transfer in tissue. This results in higher cell killing and a higher relative biological effectiveness compared to photons. Using suitably designed boron compounds which preferentially localize in cancerous cells instead of healthy tissues, boron neutron capture therapy (BNCT) has the potential of providing a higher tumor cure rate within minimal toxicity to normal tissues. This clinical approach requires a thermal neutron source, generally a nuclear reactor, with a fluence rate sufficient to deliver tumorcidal doses within a reasonable treatment time (minutes). Thermal neutrons do not penetrate deeply in tissue, therefore BNCT is limited to lesions which are either superficial or otherwise accessible. In this work, we investigate the feasibility of an accelerator-based thermal neutron source for the BNCT of skin melanomas. The source was designed via MCNP Monte Carlo simulations of the thermalization of a fast ...

  14. Neutronics shielding analysis for the end plug of a tandem mirror fusion reactor

    Science.gov (United States)

    Ragheb, Magdi M. H.; Maynard, Charles W.

    1981-10-01

    A neutronics analysis using the Monte Carlo method is carried out for the end-plug penetration and magnet system of a tandem mirror fusion reactor. Detailed penetration and the magnets' three-dimensional configurations are modeled. A method of position dependent angular source biasing is developed to adequately sample the DT fusion source in the central cell region and obtain flux contributions at the penetration components. To assure cryogenic stability, the barrier cylindrical solenoid is identified as needing substantial shielding of about 1 m of a steel-lead-boron-carbide-water mixture. Heating rates there would require a thermal-hydraulic design similar to that in the central cell blanket region. The transition coils, however, need a minimal 0.2 m thickness shield. The leakage neutron flux at the direct converters is estimated at 1.3×1015 n/(m2·s), two orders of magnitude lower than that reported at the neutral beam injectors for tokamaks around 1017 n/(m2·s) for a 1 MW/m2 14 MeV neutron wall loading. This result is obtained through a coupling between the nuclear and plasma physics designs in which hydrogen ions rather than deuterium atoms are used for energy injection at the end plug, to avoid creating a neutron source there. This lower and controllable radiation leakage problem is perceived as a potential major advantage of tandem mirrors compared to tokamaks and laser reactor systems.

  15. Spallation neutron source and other high intensity froton sources

    Energy Technology Data Exchange (ETDEWEB)

    Weiren Chou

    2003-02-06

    This lecture is an introduction to the design of a spallation neutron source and other high intensity proton sources. It discusses two different approaches: linac-based and synchrotron-based. The requirements and design concepts of each approach are presented. The advantages and disadvantages are compared. A brief review of existing machines and those under construction and proposed is also given. An R&D program is included in an appendix.

  16. Design of the miniaturized free electron laser module as an efficient source of the THz waves

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Young Chul [Department of Optometry, Eulji University, 212 Yangji-dong, Sujeong-gu, Seongnam-si, Gyeonggi-do 461-713 (Korea, Republic of); Ahn, Seong Joon [Department of I and C Engineering, Sun Moon University, Tangjeong-myeon, Asan-si, Chungnam 336-708 (Korea, Republic of); Kim, Ho Seob; Kim, Dae-Wook [Department of Information Display, Sun Moon University, 100 Kalsan-ri, Tangjeong-myun, Asan-si, Chungnam 336-708 (Korea, Republic of); Ahn, Seungjoon, E-mail: sjan@sunmoon.ac.kr [Department of Information Display, Sun Moon University, 100 Kalsan-ri, Tangjeong-myun, Asan-si, Chungnam 336-708 (Korea, Republic of)

    2011-10-21

    Since the tremendous potential of the THz wave for the bio-technological applications has been found, there has been a lot of interest paid to development of the THz-wave sources. The miniaturized free electron laser (FEL) module based on the microcolumn can be a very convenient THz wave emitter because of its compactness. In this work, we tried to design the miniaturized FEL module to achieve the optimized electron beam (e-beam) trajectory in the module by using 3D simulation tool. We found that the accelerator bias, the length and radius of the limiting aperture were important parameters to obtain the strong and parallel e-beam. We have also proposed the ring-type grids to get more symmetrical behavior of the e-beam in the wiggler.

  17. The small angle neutron spectrometer at the HANARO reactor, Korea

    Science.gov (United States)

    Seong, B.-S.; Han, Y.-S.; Lee, C.-H.; Lee, J.-S.; Hong, K.-P.; Park, K.-N.; Kim, H.-J.

    A new small angle neutron spectrometer (SANS) has been installed on the CN beam tube at the 30 MW HANARO Research Reactor in the Korea Atomic Energy Research Institute (KAERI). The SANS is to be used for the study of microstructural inhomogeneities in materials in the 1 nm to 100 nm size range. In this paper, the design characteristics of the spectrometer are presented in detail, and several SANS results for standard samples are presented which illustrate its performance.

  18. Fast neutron reactor fuel elements and power grid duty cycling

    International Nuclear Information System (INIS)

    The PHENIX power grid cycling operation in 1982-1983 will allow verification of the models and criteria developed in the interim. It will provide indispensible statistical data and will open the way to power grid duty for Super PHENIX beginning in 1986. Although at the present time it is impossible to resolve the question of weekly or daily load variations, it is felt that fast neutron reactor fuel subassemblies should provide satisfactory performance for primary and secondary frequency adjustments

  19. A methodology of neutronic-thermodynamics simulation for fast reactor

    International Nuclear Information System (INIS)

    Aiming at a general optimization of the project, controlled fuel depletion and management, this paper develop a neutronic thermodynamics simulator, SIRZ, which besides being sufficiently precise, is also economic. That results in a 75% reduction in CPU time, for a startup calculation, when compared with the same calculation at the CITATION code. The simulation system by perturbation calculations, applied to fast reactors, which produce errors smaller than 1% in all components of the reference state given by the CITATION code was tested. (author)

  20. INJECTION CHOICE FOR SPALLATION NEUTRON SOURCE RING.

    Energy Technology Data Exchange (ETDEWEB)

    WEI,J.; BEEBE-WANG,J.; BLASKIEWICZ,M.; BRODOWSKI,J.; FEDOTOV,A.; GARDNER,C.; LEE,Y.Y.; RAPARIA,D.; DANILOV,V.; HOLMES,J.; PRIOR,C.; REES,G.; MACHIDA,S.

    2001-06-18

    Injection is key in the low-loss design of high-intensity proton facilities like the Spallation Neutron Source (SNS). During the design of both the accumulator and the rapid-cycling-synchrotron version of the SNS, extensive comparison has been made to select injection scenarios that satisfy SNS's low-loss design criteria. This paper presents issues and considerations pertaining to the final choice of the SNS injection systems.

  1. Calculation of the neutron importance and weighted neutron generation time using MCNIC method in accelerator driven subcritical reactors

    International Nuclear Information System (INIS)

    Highlights: • All reactor kinetic parameters are importance weighted quantities. • MCNIC method has been developed for calculating neutron importance in ADSRs. • Mean generation time has been calculated in spallation driven systems. -- Abstract: The difference between non-weighted neutron generation time (Λ) and the weighted one (Λ†) can be quite significant depending on the type of the system. In the present work, we will focus on developing MCNIC method for calculation of the neutron importance (Φ†) and importance weighted neutron generation time (Λ†) in accelerator driven systems (ADS). Two hypothetic bare and graphite reflected spallation source driven system have been considered as illustrative examples for this means. The results of this method have been compared with those obtained by MCNPX code. According to the results, the relative difference between Λ and Λ† is within 36% and 24,840% in bare and reflected illustrative examples respectively. The difference is quite significant in reflected systems and increases with reflector thickness. In Conclusion, this method may be used for better estimation of kinetic parameters rather than the MCNPX code because of using neutron importance function

  2. Conceptual design of a high-intensity positron source for the Advanced Neutron Source

    Energy Technology Data Exchange (ETDEWEB)

    Hulett, L.D.; Eberle, C.C.

    1994-12-01

    The Advanced Neutron Source (ANS) is a planned new basic and applied research facility based on a powerful steady-state research reactor that provides neutrons for measurements and experiments in the fields of materials science and engineering, biology, chemistry, materials analysis, and nuclear science. The useful neutron flux will be at least five times more than is available in the world`s best existing reactor facility. Construction of the ANS provides a unique opportunity to build a positron spectroscopy facility (PSF) with very-high-intensity beams based on the radioactive decay of a positron-generating isotope. The estimated maximum beam current is 1000 to 5000 times higher than that available at the world`s best existing positron research facility. Such an improvement in beam capability, coupled with complementary detectors, will reduce experiment durations from months to less than one hour while simultaneously improving output resolution. This facility will remove the existing barriers to the routine use of positron-based analytical techniques and will be a giant step toward realization of the full potential of the application of positron spectroscopy to materials science. The ANS PSF is based on a batch cycle process using {sup 64}Cu isotope as the positron emitter and represents the status of the design at the end of last year. Recent work not included in this report, has led to a proposal for placing the laboratory space for the positron experiments outside the ANS containment; however, the design of the positron source is not changed by that relocation. Hydraulic and pneumatic flight tubes transport the source material between the reactor and the positron source where the beam is generated and conditioned. The beam is then transported through a beam pipe to one of several available detectors. The design presented here includes all systems necessary to support the positron source, but the beam pipe and detectors have not been addressed yet.

  3. Neutron sources and its dosimetric characteristics

    International Nuclear Information System (INIS)

    By means of Monte Carlo methods the spectra of the produced neutrons 252 Cf, 252 Cf/D2O, 241 Am Be, 239 Pu Be, 140 La Be, 239 Pu18O2 and 226 Ra Be have been calculated. With the information of the spectrum it was calculated the average energy of the neutrons of each source. By means of the fluence coefficients to dose it was determined, for each one of the studied sources, the fluence factors to dose. The calculated doses were H, H*(10), Hp,sIab (10, 00), EAP and EISO. During the phase of the calculations the sources were modeled as punctual and their characteristics were determined to 100 cm in the hole. Also, for the case of the sources of 239 Pu Be and 241 Am Be, were carried out calculations modeling the sources with their respective characteristics and the dosimetric properties were determined in a space full with air. The results of this last phase of the calculations were compared with the experimental results obtained for both sources. (Author)

  4. A novel reactor concept for boron neutron capture therapy: annular low-low power reactor (ALLPR)

    Energy Technology Data Exchange (ETDEWEB)

    Petrovic, B.; Levine, S.H. [Department of Nuclear Engineering, Pennsylvania State University, University Park, PA 16802 (United States)

    1998-07-01

    Boron Neutron Capture Therapy (BNC), originally proposed in 50's, has been getting renewed attention over the last {approx}10 years. This is in particular due to its potential for treating deep-seated brain tumors by employing epithermal neutron beams. Large (several MW) research reactors are currently used to obtain epithermal beams for BNCT, but because of cost and licensing issues it is not likely that such high-power reactors can be placed in regular medical centers. This paper describes a novel reactor concept for BNCT devised to overcome this obstacle. The design objective was to produce a beam of epithermal neutrons of sufficient intensity for BNCT at <50 kW using low enriched uranium. It is achieved by the annular reactor design, which is called Annular Low-Low Power Reactor (ALLPR). Preliminary studies using Monte Carlo simulations are summarized in this paper. The ALLPR should be relatively economical to build, and safe and easy to operate. This novel concept may increase the viability of using BNCT in medical centers worldwide. (author)

  5. Fusion Based Neutron Sources for Security Applications: Neutron Techniques

    OpenAIRE

    Albright, S.; Seviour, Rebecca

    2014-01-01

    The current reliance on X-Rays and intelligence for na- tional security is insufficient to combat the current risks of smuggling and terrorism seen on an international level. There are a range of neutron based security techniques which have the potential to dramatically improve national security. Neutron techniques can be broadly grouped into neutron in/neutron out and neutron in/photon out tech- niques. The use of accelerator based fusion devices will potentially enable to wide spread applic...

  6. Neutronic and thermal design considerations for heat-pipe reactors

    International Nuclear Information System (INIS)

    SABRE (Space-Arena Baseline Reactor) is a 100-kW/sub e/, heat-pipe-cooled, beryllium-reflected, fast reactor that produces heat at a temperature of 15000K and radiatively transmits it to high-temperature thermoelectric (TE) conversion elements. The use of heat pipes for core heat removal eliminates single-point failure mechanisms in the reactor cooling system, and provides minimal temperature drop radiative coupling to the TE array, as well as automatic, self-actuating removal of reactor afterheat. The question of how the failure of a fuel module heat pipe will affect neighboring fuel modules in the core is discussed, as is fission density peaking that occurs at the core/reflector interface. Results of neutronic calculations of the control margin available are described. Another issue that is addressed is that of helium generation in the heat pipes from neutron reactions in the core with the heat pipe fluid. Finally, the growth potential of the SABRE design to much higher powers is examined

  7. Thermophysical properties of saturated light and heavy water for advanced neutron source applications

    Energy Technology Data Exchange (ETDEWEB)

    Crabtree, A.; Siman-Tov, M.

    1993-05-01

    The Advanced Neutron Source is an experimental facility being developed by Oak Ridge National Laboratory. As a new nuclear fission research reactor of unprecedented flux, the Advanced Neutron Source Reactor will provide the most intense steady-state beams of neutrons in the world. The high heat fluxes generated in the reactor [303 MW(t) with an average power density of 4.5 MW/L] will be accommodated by a flow of heavy water through the core at high velocities. In support of this experimental and analytical effort, a reliable, highly accurate, and uniform source of thermodynamic and transport property correlations for saturated light and heavy water were developed. In order to attain high accuracy in the correlations, the range of these correlations was limited to the proposed Advanced Neutron Source Reactor's nominal operating conditions. The temperature and corresponding saturation pressure ranges used for light water were 20--300[degrees]C and 0.0025--8.5 MPa, respectively, while those for heavy water were 50--250[degrees]C and 0.012--3.9 MPa. Deviations between the correlation predictions and data from the various sources did not exceed 1.0%. Light water vapor density was the only exception, with an error of 1.76%. The physical property package consists of analytical correlations, SAS codes, and FORTRAN subroutines incorporating these correlations, as well as an interactive, easy-to-use program entitled QuikProp.

  8. [Spatial changes and sources of nitrate in Beijing urban ecosystem surface water].

    Science.gov (United States)

    Xu, Zhi-wei; Zhang, Xin-yu; Ren, Yu-fen; Sun, Xiao-min; Wang, Xiao-ke; Wang, Sheng-zhong

    2012-08-01

    The spatial variation in nitrate-nitrogen (NO3- -N) concentrations in surface water of ten sampling sites in the Beijing urban ecosystem from Kunminghu Lake to Tonghui River were assessed using monitoring data from 2009 to 2010. Nitrogen sources were examined using a hydro-chemical method. The results showed that the average nitrate-N concentrations of surface water in the Beijing urban ecosystem ranged from 0.7-7.6 mg x L(-1), with concentrations at all sites affected by human activities to a varying degree. The nitrate-N concentrations in the Dongbianmen and Tonghui River located in the southeastern of Beijing ranged from 7.0-7.6 mg x L(-1) and were significantly higher than those in the upper reaches (P waste water, leakage from solid waste disposal and domestic wastewater mainly controlled nitrate distribution in the Beijing urban surface water. The results from this study suggest that surface water management should focus on downstream sites located in the southeastern region of Beijing such as the Dongbianmen and Tonghui River in the future.

  9. Preliminary neutronic design of TRIGA Mark II Reactor

    International Nuclear Information System (INIS)

    It is very important to analyse the behaviour of the research reactors, since, they play a key role in developing the power reactor technology and radiation applications such as isotope generation for medical treatments. In this study, the neutronic behaviour of the TRIGA MARK II reactor, owned and operated by Istanbul Technical University is analysed by using the SCALE code system. In the analysis, in order to overcome the disadvantages of special TRIGA codes, such as TRIGAP, the SCALE code system is chosen to perform the calculations. TRIGAP and similar codes have limited geometrical (one-dimensional geometry) and cross sectional options (two-group calculations), however, SCALE has the capability of wider range of geometrical modelling capability (three-dimensional modelling is possible) and multi-group calculations are possible

  10. System Engineering Status and Design Characteristics of Cold Neutron Sources

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Young Ki; Jung, H. S.; Wu, S. I.; Ahn, S. H.; Kim, Y. J

    2004-06-15

    This report has been issued as a result of the surveys of foreign state of the arts on cold neutron source, which was a activity of the project, 'Development of Systems for Cold Neutron Source' commenced on July, 2003. Review and analysis for systems characteristics of the foreign cold neutron source will be a good reference for the basic and detail design starting on the July, 2004. Since the cold neutron source adopts liquid hydrogen or deuterium as a moderator and special field of techniques such as cryogenic cooling system, high vacuum and gas handling, etc., the project started with the survey work in order to be familiar with the concept of the cold neutron source. The survey work was proceeded on the foreign cold neutron sources. This report includes characteristics of the several foreign cold neutron sources.

  11. Tests of a new CCD-camera based neutron radiography detector system at the reactor stations in Munich and Vienna

    Energy Technology Data Exchange (ETDEWEB)

    Lehmann, E.; Pleinert, H. [Paul Scherrer Inst. (PSI), Villigen (Switzerland); Schillinger, B. [Technische Univ. Muenchen (Germany); Koerner, S. [Atominstitut der Oesterreichischen Universitaeten, Vienna (Austria)

    1997-09-01

    The performance of the new neutron radiography detector designed at PSI with a cooled high sensitive CCD-camera was investigated under real neutronic conditions at three beam ports of two reactor stations. Different converter screens were applied for which the sensitivity and the modulation transfer function (MTF) could be obtained. The results are very encouraging concerning the utilization of this detector system as standard tool at the radiography stations at the spallation source SINQ. (author) 3 figs., 5 refs.

  12. Neutronic Design and Measured Performance of the Low Energy Neutron Source (LENS) Target Moderator Reflector Assembly

    CERN Document Server

    Lavelle, C M; Bogdanov, A; Derenchuk, V P; Kaiser, H; Leuschner, M B; Lone, M A; Lozowski, W; Nann, H; Von Przewoski, B; Remmes, N; Rinckel, T; Shin, Y; Snow, W M; Sokol, P E

    2008-01-01

    The Low Energy Neutron Source (LENS) is an accelerator-based pulsed cold neutron facility under construction at the Indiana University Cyclotron Facility (IUCF). The idea behind LENS is to produce pulsed cold neutron beams starting with ~MeV neutrons from (p,n) reactions in Be which are moderated to meV energies and extracted from a small solid angle for use in neutron instruments which can operate efficiently with relatively broad (~1 msec) neutron pulse widths. Although the combination of the features and operating parameters of this source is unique at present, the neutronic design possesses several features similar to those envisioned for future neutron facilities such as long-pulsed spallation sources (LPSS) and very cold neutron (VCN) sources. We describe the underlying ideas and design details of the target/moderator/reflector system (TMR) and compare measurements of its brightness, energy spectrum, and emission time distribution under different moderator configurations with MCNP simulations. Brightnes...

  13. Advanced Neutron Source (ANS) Project: Annual report, April 1987--March 1988

    International Nuclear Information System (INIS)

    The Advanced Neutron Source (ANS) Project (formerly called the Center for Neutron Research) will provide the world's best facilities for the study of neutron scattering. The ANS high-power density reactor will be fueled with uranium silicide and cooled, moderated, and reflected by deuterium oxide. Peak neutron fluxes in the reflector are expected to be 5 to 10 x 1019 neutrons/center dot/m-2/center dot/s-1 with a power level between 270 and 300 MW. This report describes the status of technical work funded through the ANS Project during the period April 1987 through March 1988. Earlier work is described in Center for Neutron Research Project Status Report and other Oak Ridge National Laboratory reports. 22 refs., 57 figs., 23 tabs

  14. Advanced Neutron Source (ANS) Project: Annual report, April 1987--March 1988

    Energy Technology Data Exchange (ETDEWEB)

    Selby, D.L.; Harrington, R.M.; Peretz, F.J.; McBee, M.R. (comp.)

    1989-02-01

    The Advanced Neutron Source (ANS) Project (formerly called the Center for Neutron Research) will provide the world's best facilities for the study of neutron scattering. The ANS high-power density reactor will be fueled with uranium silicide and cooled, moderated, and reflected by deuterium oxide. Peak neutron fluxes in the reflector are expected to be 5 to 10 x 10/sup 19/ neutrons/center dot/m/sup -2//center dot/s/sup -1/ with a power level between 270 and 300 MW. This report describes the status of technical work funded through the ANS Project during the period April 1987 through March 1988. Earlier work is described in Center for Neutron Research Project Status Report and other Oak Ridge National Laboratory reports. 22 refs., 57 figs., 23 tabs.

  15. Seasonal variations and source estimation of saccharides in atmospheric particulate matter in Beijing, China.

    Science.gov (United States)

    Liang, Linlin; Engling, Guenter; Du, Zhenyu; Cheng, Yuan; Duan, Fengkui; Liu, Xuyan; He, Kebin

    2016-05-01

    Saccharides are important constituents of atmospheric particulate matter (PM). In order to better understand the sources and seasonal variations of saccharides in aerosols in Beijing, China, saccharide composition was measured in ambient PM samples collected at an urban site in Beijing. The highest concentrations of total saccharides in Beijing were observed in autumn, while an episode with abnormal high total saccharide levels was observed from 15 to 23 June, 2011, due to extensive agricultural residue burning in northern China during the wheat harvest season. Compared to the other two categories of saccharides, sugars and sugar alcohols, anhydrosugars were the predominant saccharide group, indicating that biomass burning contributions to Beijing urban aerosol were significant. Ambient sugar and sugar alcohol levels in summer and autumn were higher than those in spring and winter, while they were more abundant in PM2.5 during winter time. Levoglucosan was the most abundant saccharide compound in both PM2.5 and PM10, the annual contributions of which to total measured saccharides in PM2.5 and PM10 were 61.5% and 54.1%, respectively. To further investigate the sources of the saccharides in ambient aerosols in Beijing, the PM10 datasets were subjected to positive matrix factorization (PMF) analysis. Based on the objective function to be minimized and the interpretable factors identified by PMF, six factors appeared to be optimal as to the probable origin of saccharides in the atmosphere in Beijing, including biomass burning, soil or dust, isoprene SOA and the direct release of airborne fungal spores and pollen. PMID:26921589

  16. A study of reactor-neutron-induced reactions

    International Nuclear Information System (INIS)

    Cross sections of the second neutron capture in the double-neutron-capture process were studied in the irradiation of 26Mg, 64Ni, 93Nb, and 164Dy with reactor neutrons. Weak radioactivities produced were determined with sufficient accuracy by taking advantage of high resolution Ge(Li) detectors. The neutron spectrum at the irradiating site was defined with Westcott's epithermal index determined by the use of the cadmium ratio method. The isomer cross-section ratio was obtained with the 94Nb target. The statistical theory was found to be applicable to such an unstable or odd-odd nuclide as 94Nb. Cross sections of isomers of 165Dy were determined. The capture cross section of the ground state nuclei was found greater than that of the metastable state nuclei in this case. This differs from the other four isomers observed so far. Cross sections of 28Al(n, p)28Mg and 58Ni(n, 2n)57Ni reactions with fission neutrons were also measured. A simple formula was proposed for the systematics of (n, 2n) reaction cross sections, which enabled prediction of the cross section more accurately than the previously proposed formulae. (author)

  17. Genotoxicity of neutrons in Drosophila melanogaster. Somatic mutation and recombination induced by reactor neutrons.

    Science.gov (United States)

    Guzmán-Rincón, J; Delfín-Loya, A; Ureña-Núñez, F; Paredes, L C; Zambrano-Achirica, F; Graf, U

    2005-08-01

    This paper describes the observation of a direct relationship between the absorbed doses of neutrons and the frequencies of somatic mutation and recombination using the wing somatic mutation and recombination test (SMART) of Drosophila melanogaster. This test was used for evaluating the biological effects induced by neutrons from the Triga Mark III reactor of Mexico. Two different reactor power levels were used, 300 and 1000 kW, and two absorbed doses were tested for each power level: 1.6 and 3.2 Gy for 300 kW and 0.84 and 1.7 Gy for 1000 kW. A linear relationship was observed between the absorbed dose and the somatic mutation and recombination frequencies. Furthermore, these frequencies were dependent on larval age: In 96-h-old larvae, the frequencies were increased considerably but the sizes of the spots were smaller than in 72-h-old larvae. The analysis of the balancer-heterozygous progeny showed a linear absorbed dose- response relationship, although the responses were clearly lower than found in the marker-trans-heterozygous flies. Approximately 65% of the genotoxicity observed is due to recombinational events. The results of the study indicate that thermal and fast neutrons are both mutagenic and recombinagenic in the D. melanogaster wing SMART, and that the frequencies are dependent on neutron dose, reactor power, and the age of the treated larvae. PMID:16038586

  18. Neutron source based on the TORNADO trap

    International Nuclear Information System (INIS)

    The TORNADO magnetic trap as a source of thermonuclear neutrons with 108 neutron per a pulse in the D-D reaction is considered. The construction of magnetic traps both with stationary and quasistationary modes of their operation is shown to be possible. The results of numerical calculation of the magnetic system parameters are given, analysis of permissible mechanical loads, turns displacements of and magnetic fields in the trap is carried out. Considerable decrease of pondermotive forces affecting the turns of an internal spiral when conserving thermo-insulating properties of the magnetic trap field is shown to be possible. The loads of the trap spiral magnet coils are shown to be also acceptable to form the stationary magnetic field of the 2 Tl order in the magnetic barrier

  19. Inversion of source parameters for moderate and small earthquakes in Beijing region

    Institute of Scientific and Technical Information of China (English)

    LAN Cong-xin; LIU Jie; ZHENG Si-hua; MA Shi-zhen; LI Ju-zhen

    2005-01-01

    According to the geological structural features, Beijing and the adjacent areas can be divided into two regions of plain in the east and mountain in the west. Among the stations covered by the telemetered digital seismic station network of Earthquake Administration of Beijing Municipality, the stations in the plain area are all borehole ones and the stations in the western mountainous region are all located on the surface bedrock. In the paper, 511 waveform data recorded by the network from Oct. 2001 to Oct. 2004 are used in the researches for the entire Beijing region, the western mountainous region and the eastern plain area, respectively. The Q values are calculated for each area by Atkinson's method and compared with the existed data. The reliability of the Q values and the reasons for the difference in the Q values are also discussed. Then, the source parameters and site response are inverted by the Moya's method, in which two models are used. The first model uses the Q values, earthquakes and stations in the sub-areas and the second model uses the Q values, earthquakes and stations in the entire Beijing region. The results indicate that the source parameters and site responses obtained by two models are basically consistent with each other. It also indicates that the source parameters obtained by these methods are not affected by the size of station network.

  20. Neutron diffractometers for structural biology at spallation neutron sources

    International Nuclear Information System (INIS)

    Spallation neutron sources are ideal for diffraction studies of proteins and oriented molecular complexes. With spoliation neutrons and their time dependent wavelength structure, it is easy to electronically select data with an optimal wavelength bandwidth and cover the whole Laue spectrum as time (wavelength) resolved snapshots. This optimized data quality with best peak-to-background ratios and provides adequate spatial and energy resolution to eliminate peak overlaps. The application of this concept will use choppers to select the desired Laue wavelength spectrum and employ focusing optics and large cylindrical 3He detectors to optimize data collection rates. Such a diffractometer will cover a Laue wavelength range from 1 to 5 Angstrom with a flight path length of 10m and an energy resolution of 0.25 Angstrom. Moderator concepts for maximal flux distribution within this energy range will be discussed using calculated flux profiles. Since the energy resolution required for such timed data collection in this super Laue techniques is not very high, the use of a linac only (LAMPF) spoliation target is an exciting possibility with an order of magnitude increase in flux

  1. Neutron diffractometers for structural biology at spallation neutron sources

    Energy Technology Data Exchange (ETDEWEB)

    Schoenborn, B.P.; Pitcher, E. [Los Alamos National Laboratory, NM (United States)

    1994-12-31

    Spallation neutron sources are ideal for diffraction studies of proteins and oriented molecular complexes. With spoliation neutrons and their time dependent wavelength structure, it is easy to electronically select data with an optimal wavelength bandwidth and cover the whole Laue spectrum as time (wavelength) resolved snapshots. This optimized data quality with best peak-to-background ratios and provides adequate spatial and energy resolution to eliminate peak overlaps. The application of this concept will use choppers to select the desired Laue wavelength spectrum and employ focusing optics and large cylindrical {sup 3}He detectors to optimize data collection rates. Such a diffractometer will cover a Laue wavelength range from 1 to 5{Angstrom} with a flight path length of 10m and an energy resolution of 0.25{Angstrom}. Moderator concepts for maximal flux distribution within this energy range will be discussed using calculated flux profiles. Since the energy resolution required for such timed data collection in this super Laue techniques is not very high, the use of a linac only (LAMPF) spoliation target is an exciting possibility with an order of magnitude increase in flux.

  2. Inertial electro-magnetostatic plasma neutron sources

    International Nuclear Information System (INIS)

    Two types of systems are being studied experimentally as D-T plasma neutron sources. In both concepts, spherical convergence of either electrons or ions or both is used to produce a dense central focus within which D-T fusion reactions produce 14 MeV neutrons. One concept uses nonneutral plasma confinement principles in a Penning type trap. In this approach, combined electrostatic and magnetic fields provide a vacuum potential well within which electrons are confined and focused. A small (6 mm radius) spherical machine has demonstrated a focus of 30 microm radius, with a central density of up to 35 times the Brillouin density limit of a static trap. The resulting electron plasma of up to several 1013 cm-3 provides a multi-kV electrostatic well for confining thermonuclear ions as a neutron source. The second concept (Inertial Electrostatic Confinement, or IEC) uses a high-transparence grid to form a global well for acceleration and confinement of ions. Such a system has demonstrated steady neutron output of 2 x 1010 s-1. The present experiment will scale this to >1011 s-1. Advanced designs based on each concept have been developed recently. In these proposed approaches, a uniform-density electron sphere forms an electrostatic well for ions. Ions so trapped may be focused by spherical convergence to produce a dense core. An alternative approach produces large amplitude spherical oscillations of a confined ion cloud by a small, resonant modulation of the background electrons. In both the advanced Penning trap approach and the advanced IEC approach, the electrons are magnetically insulated from a large (up to 100 kV) applied electrostatic field. The physics of these devices is discussed, experimental design details are given, present observations are analyzed theoretically, and the performance of future advanced systems are predicted

  3. 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.

  4. Determination of nitrogen in wheat flour through Activation analysis using Fast neutron flux of a Thermal nuclear reactor; Determinacion de nitrogeno en harina de trigo mediante analisis por activacion empleando el flujo de neutrones rapidos de un reactor nuclear termico

    Energy Technology Data Exchange (ETDEWEB)

    Ramirez G, T

    1976-07-01

    In this work is done a technical study for determining Nitrogen (protein) and other elements in wheat flour Activation analysis, with Fast neutrons from a Thermal nuclear reactor. Initially it is given an introduction about the basic principles of the methods of analysis. Equipment used in Activation analysis and a brief description of the neutron source (Thermal nuclear reactor). The realized experiments for determining the flux form in the irradiation site, the half life of N-13 and the interferences due to the sample composition are included too. Finally, the obtained results by Activation and the Kjeldahl method are tabulated. (Author)

  5. A new small-angle neutron scattering spectrometer at China Mianyang research reactor

    Science.gov (United States)

    Peng, Mei; Sun, Liangwei; Chen, Liang; Sun, Guangai; Chen, Bo; Xie, Chaomei; Xia, Qingzhong; Yan, Guanyun; Tian, Qiang; Huang, Chaoqiang; Pang, Beibei; Zhang, Ying; Wang, Yun; Liu, Yaoguang; Kang, Wu; Gong, Jian

    2016-02-01

    A new pinhole small-angle neutron scattering (SANS) spectrometer, installed at the cold neutron source of the 20 MW China Mianyang Research Reactor (CMRR) in the Institute of Nuclear Physics and Chemistry, China Academy of Engineering Physics, has been put into use since 2014. The spectrometer is equipped with a multi-blade mechanical velocity selector, a multi-beam collimation system, and a two-dimensional He-3 position sensitive neutron detector. The q-range of the spectrometer covers from 0.01 nm-1 to 5.0 nm-1. In this paper, the design and characteristics of the SANS spectrometer are described. The q-resolution calculations, together with calibration measurements of silver behenate and a dispersion of nearly monodisperse poly-methyl-methacrylate nanoparticles indicate that our SANS spectrometer has a good performance and is now in routine service.

  6. New sources and instrumentation for neutrons in biology

    DEFF Research Database (Denmark)

    Teixeira, S.C.M.; Zaccai, G.; Ankner, J.;

    2008-01-01

    Neutron radiation offers significant advantages for the study of biological molecular structure and dynamics. A broad and significant effort towards instrumental and methodological development to facilitate biology experiments at neutron sources worldwide is reviewed.......Neutron radiation offers significant advantages for the study of biological molecular structure and dynamics. A broad and significant effort towards instrumental and methodological development to facilitate biology experiments at neutron sources worldwide is reviewed....

  7. Diagnosis of Hydrogen Plasma in a Miniature Penning Ion Source by Double Probes

    Institute of Scientific and Technical Information of China (English)

    JIN Dazhi; YANG Zhonghai; XIAO Kunxiang; DAI Jingyi

    2009-01-01

    Parameters of hydrogen plasma in a miniature Penning discharge ion source,including the electron temperature and the electron density,were measured by using double probes.The results indicate that the electron density increases and the electron temperature decreases with the increase in gas pressure and the discharge current.The electron temperature is about 5~9 eV and the electron density is 6.0x1013~1.2×1014 m-3 while the discharge current is in a range of 50~12μA.

  8. Neutron dosimetry for radiation damage in fission and fusion reactors

    International Nuclear Information System (INIS)

    The properties of materials subjected to the intense neutron radiation fields characteristic of fission power reactors or proposed fusion energy devices is a field of extensive current research. These investigations seek important information relevant to the safety and economics of nuclear energy. In high-level radiation environments, neutron metrology is accomplished predominantly with passive techniques which require detailed knowledge about many nuclear reactions. The quality of neutron dosimetry has increased noticeably during the past decade owing to the availability of new data and evaluations for both integral and differential cross sections, better quantitative understanding of radioactive decay processes, improvements in radiation detection technology, and the development of reliable spectrum unfolding procedures. However, there are problems caused by the persistence of serious integral-differential discrepancies for several important reactions. There is a need to further develop the data base for exothermic and low-threshold reactions needed in thermal and fast-fission dosimetry, and for high-threshold reactions needed in fusion-energy dosimetry. The unsatisfied data requirements for fission reactor dosimetry appear to be relatively modest and well defined, while the needs for fusion are extensive and less well defined because of the immature state of fusion technology. These various data requirements are examined with the goal of providing suggestions for continued dosimetry-related nuclear data research

  9. Neutron sensors in the SP-100 reactor control system

    International Nuclear Information System (INIS)

    The reference reactor control approach for the mature generic flight system (GFS) utilizes highly reliable and diverse reactor outlet temperature measurements for control and protection. Although system dynamic analyses demonstrated that this approach is satisfactory for various modes of operation (including transients involving failure or degradation of equipment), the use of a neutron monitoring system (NMS) for initial startup and for an early period of power operation has been studied to improve the performance of the reactor control design. Control strategies were developed, simulation analyses were produced, and stability margins were examined. In this updated control approach, the signals from the NMS are used for the initial startup, for restarts, for power range control, and for protection from overpower transients as long as reliable data is available from the NMS. The results show satisfactory performance for the updated controls. If the lifetime of the NMS is shorter than that of the flight system, the reactor control will revert to the reference control approach employing reactor outlet temperature measurements only

  10. A status report on the advanced neutron source project

    International Nuclear Information System (INIS)

    Design work on the Advanced Neutron Source facilities has progressed significantly, with cost saving changes to the buildings and other systems. The cold source design has advanced considerably, and in addition design work has been initiated on the hot neutron source and on a positron source. (J.P.N.)

  11. Measurements of effective delayed neutron fraction in a fast neutron reactor using the perturbation method

    CERN Document Server

    Zhou, Hao-Jun; Fan, Xiao-Qiang; Li, Zheng-Hong; Pu, Yi-Kang

    2015-01-01

    The perturbation method is proposed to obtain the effective delayed neutron fraction (\\b{eta}eff) of a cylindrical highly enriched uranium reactor. Based on the reactivity measurements with and without a sample at a designable position using the positive periodic technique, the reactor reactivity perturbation {\\Delta}\\r{ho} of the sample in \\b{eta}eff units is measured. The simulation of the perturbation experiments are performed by MCNP program. The PERT card is used to provide the difference dk of effective neutron multiplication factors with and without the sample inside the reactor. Based on the relationship between the effective multiplication factor and the reactivity, the equation \\b{eta}eff =dk/{\\Delta}\\r{ho} is derived. In this paper, the reactivity perturbations of 13 metal samples at the designable position of the reactor are measured and calculated. The average \\b{eta}eff value of the reactor is given as 0.00645, and the standard uncertainty is 3.0%. Additionally, the perturbation experiments for ...

  12. Determination of nitrogen in wheat flour through Activation analysis using Fast neutron flux of a Thermal nuclear reactor

    International Nuclear Information System (INIS)

    In this work is done a technical study for determining Nitrogen (protein) and other elements in wheat flour Activation analysis, with Fast neutrons from a Thermal nuclear reactor. Initially it is given an introduction about the basic principles of the methods of analysis. Equipment used in Activation analysis and a brief description of the neutron source (Thermal nuclear reactor). The realized experiments for determining the flux form in the irradiation site, the half life of N-13 and the interferences due to the sample composition are included too. Finally, the obtained results by Activation and the Kjeldahl method are tabulated. (Author)

  13. Neutronics Study of the KANUTER Space Propulsion Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Venneri, Paolo; Nam, Seung Hyun; Kim, Yonghee [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

    2014-05-15

    The Korea Advanced Nuclear Thermal Engine Rocket (KANUTER) has been developed at the Korea Advanced Institute of Science and Technology (KAIST). This space propulsion system is unique in that it implements a HEU fuel with a thermal spectrum system. This allows the system to be designed with a minimal amount of fissile material and an incredibly small and light system. This then allows the implementation of the system in a cluster format which enables redundancy and easy scalability for different mission requirements. This combination of low fissile content, compact size, and thermalized spectrum contribute to an interesting and novel behavior of the reactor system. The two codes were both used for the burn up calculations in order to verify their validity while the static calculations and characterization of the core were done principally with MCNPX. The KANUTER space propulsion reactor is in the process of being characterized and improved. Its basic neutronic characteristics have been studied, and its behavior over time has been identified. It has been shown that this reactor will have difficulty operating as hoped in a bimodal configuration where it is able to provide both propulsion and power throughout mission to Mars. The reason for this has been identified as Xe{sup 135}, and it is believed that a possible solution to this issue does exist, either in the form of an appropriately designed neutron spectrum or the building in of sufficient excess reactivity.

  14. An accelerator-based epithermal photoneutron source for boron neutron capture therapy

    Energy Technology Data Exchange (ETDEWEB)

    Mitchell, H.E.

    1996-04-01

    Boron neutron capture therapy is an experimental binary cancer radiotherapy modality in which a boronated pharmaceutical that preferentially accumulates in malignant tissue is first administered, followed by exposing the tissue in the treatment volume to a thermal neutron field. Current usable beams are reactor-based but a viable alternative is the production of an epithermal neutron beam from an accelerator. Current literature cites various proposed accelerator-based designs, most of which are based on proton beams with beryllium or lithium targets. This dissertation examines the efficacy of a novel approach to BNCT treatments that incorporates an electron linear accelerator in the production of a photoneutron source. This source may help to resolve some of the present concerns associated with accelerator sources, including that of target cooling. The photoneutron production process is discussed as a possible alternate source of neutrons for eventual BNCT treatments for cancer. A conceptual design to produce epithermal photoneutrons by high photons (due to bremsstrahlung) impinging on deuterium targets is presented along with computational and experimental neutron production data. A clinically acceptable filtered epithermal neutron flux on the order of 10{sup 7} neutrons per second per milliampere of electron current is shown to be obtainable. Additionally, the neutron beam is modified and characterized for BNCT applications by employing two unique moderating materials (an Al/AlF{sub 3} composite and a stacked Al/Teflon design) at various incident electron energies.

  15. An accelerator-based epithermal photoneutron source for boron neutron capture therapy

    International Nuclear Information System (INIS)

    Boron neutron capture therapy is an experimental binary cancer radiotherapy modality in which a boronated pharmaceutical that preferentially accumulates in malignant tissue is first administered, followed by exposing the tissue in the treatment volume to a thermal neutron field. Current usable beams are reactor-based but a viable alternative is the production of an epithermal neutron beam from an accelerator. Current literature cites various proposed accelerator-based designs, most of which are based on proton beams with beryllium or lithium targets. This dissertation examines the efficacy of a novel approach to BNCT treatments that incorporates an electron linear accelerator in the production of a photoneutron source. This source may help to resolve some of the present concerns associated with accelerator sources, including that of target cooling. The photoneutron production process is discussed as a possible alternate source of neutrons for eventual BNCT treatments for cancer. A conceptual design to produce epithermal photoneutrons by high photons (due to bremsstrahlung) impinging on deuterium targets is presented along with computational and experimental neutron production data. A clinically acceptable filtered epithermal neutron flux on the order of 107 neutrons per second per milliampere of electron current is shown to be obtainable. Additionally, the neutron beam is modified and characterized for BNCT applications by employing two unique moderating materials (an Al/AlF3 composite and a stacked Al/Teflon design) at various incident electron energies

  16. Design of a plate type fuel based - low power medical reactor for boron neutron capture therapy

    International Nuclear Information System (INIS)

    The interest in the boron neutron capture therapy (BNCT) has been renewed for cancer therapy with some indication of its potential efficacy in recent years. To solve the most important problem that thermal neutrons are attenuated rapidly in tissue due to absorption and scattering, thermal neutron beams are replaced by epithermal neutron beams. Thus, epithermal neutron beams are directed towards a patient's head, during their passage through tissue these neutrons rapidly lose energy by elastic scattering until they end up as thermal neutrons in target tumor volume. The thermal neutrons thus formed, are captured by the 10B atoms which become 11B atoms in the excited state for a very short time 10-12 sec. The 11B atoms then decay producing alpha particles, 7Li recoil nuclei and gamma rays. Tumor cells are killed selectively by the energetic alpha particles and 7Li fission products. We propose a 300kW slab type reactor core having thin and large surface areas so that most of the neutrons emerging from the faces and entering moderator region are fission spectrum neutrons to acquire high intense epithermal neutron beam with high quality. All faces of the slab core, East-West region and North-South region, were considered for epithermal neutron beam collimators. Plate-type U3Si2-Al dispersion fuel having high uranium density is very compatible with composing of a slab type core. The reactor core is loaded with 3.89kg U235 and has the dimension of about 23.46cm width, 31.28cm length and 64.8cm height, with 216 locations to place 204 fuel elements, eight control plates and four safety plates. The general-purpose MCNP 4B code was used to carry out the neutron and photon transport computations. Both keff criticality and fixed source problems were computed. We could reduce at least 7 times long computer time (105 to 140 h in a run) needed to initiate enough neutrons in a run ( 6000 to 8000 cycles in a run with 3000 neutrons per cycle) using the PVM (Parallel Virtual

  17. Physics Analyses in the Design of the HFIR Cold Neutron Source

    International Nuclear Information System (INIS)

    Physics analyses have been performed to characterize the performance of the cold neutron source to be installed in the High Flux Isotope Reactor at the Oak Ridge National Laboratory in the near future. This paper provides a description of the physics models developed, and the resulting analyses that have been performed to support the design of the cold source. These analyses have provided important parametric performance information, such as cold neutron brightness down the beam tube and the various component heat loads, that have been used to develop the reference cold source concept

  18. Current status of neutron activation analysis in HANARO Research Reactor

    International Nuclear Information System (INIS)

    The facilities for neutron activation analysis in the HANARO (Hi-flux Advanced Neutron Application Research Reactor) are described and the main applications of NAA (Neutron Activation Analysis) are reviewed. The sample irradiation tube, automatic and manual pneumatic transfer system were installed at three irradiation holes of HANARO at the end of 1995. The performance of the NAA facility was examined to identify the characteristics of the tube transfer system, irradiation sites and custom-made polyethylene irradiation capsule. The available thermal neutron fluxes at irradiation sites are in the range of 3 x 1013 - 1 x 1014 n/cm2·s and cadmium ratios are in 15 - 250. For an automatic sample changer for gamma-ray counting, a domestic product was designed and manufactured. An integrated computer program (Labview) to analyse the content was developed. In 2001, PGNAA (Prompt Gamma Neutron Activation Analysis) facility has been installed using a diffracted neutron beam of ST1. NAA has been applied in the trace component analysis of nuclear, geological, biological, environmental and high purity materials, and various polymers for research and development. The improvement of analytical procedures and establishment of an analytical quality control and assurance system were studied. Applied research and development for the environment, industry and human health by NAA and its standardization were carried out. For the application of the KOLAS (Korea Laboratory Accreditation Scheme), evaluation of measurement uncertainty and proficiency testing of reference materials were performed. Also to verify the reliability and to validate analytical results, intercomparison studies between laboratories were carried out. (author)

  19. Current status of neutron activation analysis in HANARO Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Yong Sam; Moon, Jong Hwa; Sohn, Jae Min [Korea Atomic Energy Research Institute, Daejeon (Korea)

    2003-03-01

    The facilities for neutron activation analysis in the HANARO (Hi-flux Advanced Neutron Application Research Reactor) are described and the main applications of NAA (Neutron Activation Analysis) are reviewed. The sample irradiation tube, automatic and manual pneumatic transfer system were installed at three irradiation holes of HANARO at the end of 1995. The performance of the NAA facility was examined to identify the characteristics of the tube transfer system, irradiation sites and custom-made polyethylene irradiation capsule. The available thermal neutron fluxes at irradiation sites are in the range of 3 x 10{sup 13} - 1 x 10{sup 14} n/cm{sup 2}{center_dot}s and cadmium ratios are in 15 - 250. For an automatic sample changer for gamma-ray counting, a domestic product was designed and manufactured. An integrated computer program (Labview) to analyse the content was developed. In 2001, PGNAA (Prompt Gamma Neutron Activation Analysis) facility has been installed using a diffracted neutron beam of ST1. NAA has been applied in the trace component analysis of nuclear, geological, biological, environmental and high purity materials, and various polymers for research and development. The improvement of analytical procedures and establishment of an analytical quality control and assurance system were studied. Applied research and development for the environment, industry and human health by NAA and its standardization were carried out. For the application of the KOLAS (Korea Laboratory Accreditation Scheme), evaluation of measurement uncertainty and proficiency testing of reference materials were performed. Also to verify the reliability and to validate analytical results, intercomparison studies between laboratories were carried out. (author)

  20. UCN Source at an External Beam of Thermal Neutrons

    OpenAIRE

    2015-01-01

    We propose a new method for production of ultracold neutrons (UCNs) in superfluid helium. The principal idea consists in installing a helium UCN source into an external beam of thermal or cold neutrons and in surrounding this source with a solid methane moderator/reflector cooled down to ~4 K. The moderator plays the role of an external source of cold neutrons needed to produce UCNs. The flux of accumulated neutrons could exceed the flux of incident neutrons due to their numerous reflections ...

  1. Coupled neutronics and thermal-hydraulics numerical simulations of a Molten Salt Fast Reactor (MSFR)

    International Nuclear Information System (INIS)

    Coupled neutronics and thermalhydraulic numerical analyses of a molten salt fast reactor (MSFR) are presented. These preliminary numerical simulations are carried-out using the Monte Carlo code MCNP and the Computation Fluid Dynamic code OpenFOAM. The main objectives of this analysis performed at steady-reactor conditions are to confirm the acceptability of the current neutronic and thermalhydraulic designs of the reactor, to study the effects of the reactor operating conditions on some of the key MSFR design parameters such as the temperature peaking factor. The effects of the precursor's motion on the reactor safety parameters such as the effective fraction of delayed neutrons have been evaluated. (authors)

  2. The IAEA Collaborating Centre for Neutron Activation Based Methodologies of Research Reactors

    International Nuclear Information System (INIS)

    The Reactor Institute Delft was inaugurated in May 2009 as a new IAEA Collaborating Centre for Neutron Activation Based Methodologies of Research Reactors. The collaboration involves education, research and development in (i) Production of reactor-produced, no-carrier added radioisotopes of high specific activity via neutron activation; (ii) Neutron activation analysis with emphasis on automation as well as analysis of large samples, and radiotracer techniques; and, as a cross-cutting activity, (iii) Quality assurance and management in research and application of research reactor based techniques and in research reactor operations. (author)

  3. Characterization of a {sup 239}PuBe isotopic neutron source

    Energy Technology Data Exchange (ETDEWEB)

    Vega C, H. R.; Hernandez D, V. M. [Universidad Autonoma de Zacatecas, Unidad Academica de Estudios Nucleares, Cipres No. 10, Fracc. La Penuela, 98068 Zacatecas (Mexico); Rivera M, T. [IPN, Centro de Investigacion en Ciencia Aplicada y Tecnologia Avanzada, Calz. Legaria No. 694, Col. Irrigacion, 11500 Mexico D. F. (Mexico); Sanchez, A., E-mail: fermineutron@yahoo.com [IPN, Escuela Superior de Fisica y Matematicas, 07738 Mexico D. F. (Mexico)

    2012-10-15

    A Bonner sphere spectrometer was used to determine the features of a {sup 239}PuBe neutron source used to operate the ESFM-Ipn Subcritical Reactor. The spectrometer is a {sup 6}Lil(Eu) scintillator and 2, 3, 5, 8, 10 and 12 inches-diameter polyethylene spheres, that was located 100 cm from the neutron source. The count rates obtained with the spectrometer were unfolded using the NSDUAZ code and neutron spectrum, total fluence, and ambient dose equivalent were determined. A Monte Carlo calculation, using the MCNP5 code, was carried out to estimate the spectrum and integral features being less that values obtained experimentally due to the presence of {sup 241}Pu in the Pu used to fabricate the source. Using the experimental information the actual neutron yield and the mass fraction of {sup 241}Pu was estimated. (Author)

  4. Californium-252 Neutron Therapy in China

    International Nuclear Information System (INIS)

    Californium-252 brachytherapy, believed to be the most successful source for neutron therapy, gives most of the cures as well as long-term and complication-free survivals. Chinese radiation oncologists were interested in californium neutron therapy (Cf-NT) in the early 1980s, but 252Cf sources for medical use were not available in China until 1992 when a californium joint venture was established by the China Institute of Atomic Energy (Beijing) and the Research Institute for Nuclear Reactors (Dimitrovgrad) of Russia. In 1995, 25 seeds of 252Cf with a strength of 3 μg each were sent to China for preclinical investigation. Three years later, a high dose rate (HDR) 252Cf source was imported and transferred into a home-made remote after-loader for intracavitary treatment in Chongqing, and a clinical trail was started in February 1999. This is the first time that Cf-NT was performed for cancer patients in China. Since then, Cf-NT in China has developed rapidly. It is estimated that one-tenth of those radiation oncology centers with brachytherapy practice will be equipped with californium units in 5 yr. That means more than 30 units will be in use in hospitals. That is significant compared with other countries, but it is just one, on average, for each province or one per 40 million people in China. Progress also has been achieved in the 252Cf treatment delivery equipment. Preliminary clinical trails showed complete response observed in all cases treated, with a rapid clearance of tumors and mild reactions in normal tissues. The short-term results are quite encouraging. To deal with problems due to the demand for Cf-NT in China, attention should be paid to the following particulars: (1) A high-strength miniature source is needed for HDR/MDR interstitial therapy to extend the Cf-NT coverage. (2) Basic work on radiophysics and radiobiology needs to be done, including source calibration, clinical dosimetry, clinical RBE determination, and Cf-NT quality assurance

  5. On the source contribution to Beijing PM2.5 concentrations

    Science.gov (United States)

    Zíková, Naděžda; Wang, Yungang; Yang, Fumo; Li, Xinghua; Tian, Mi; Hopke, Philip K.

    2016-06-01

    Beijing is a city with some of the world's worst particulate air pollution. Although there have been various control strategies implemented since 1998, there are still episodes of PM2.5 concentrations of hundreds of micrograms per cubic meter. In this study, samples were collected over a year in Beijing, chemically characterized, and the resulting data analyzed for source apportionment. The new error analysis capabilities built into EPA PMF V5.0 have been employed to better evaluate the profiles and assign them to source types. Secondary sulfate, local coal combustion and secondary nitrate were the major contributors to the PM2.5 mass. However, in this study, traffic was found to be more important as a PM compared to prior studies. It was actually the largest PM2.5 source in autumn and winter although local coal combustion is also a large source of PM in the winter months. These results demonstrate the value of using the displacement method to assess the variability in source profiles to improve our interpretation of PMF results. They also suggest more attention needs to be paid to traffic emissions in Beijing.

  6. Source apportionment and health risk assessment of trace metals in surface soils of Beijing metropolitan, China.

    Science.gov (United States)

    Chen, Haiyang; Teng, Yanguo; Lu, Sijin; Wang, Yeyao; Wu, Jin; Wang, Jinsheng

    2016-02-01

    Understanding the exposure risks of trace metals in contamination soils and apportioning their sources are the basic preconditions for soil pollution prevention and control. In this study, a detailed investigation was conducted to assess the health risks of trace metals in surface soils of Beijing which is one of the most populated cities in the world and to apportion their potential sources. The data set of metals for 12 elements in 240 soil samples was collected. Pollution index and enrichment factor were used to identify the general contamination characteristic of soil metals. The probabilistic risk model was employed for health risk assessment, and a chemometrics technique, multivariate curve resolution-weighted alternating least squares (MCR-WALS), was applied to apportion sources. Results suggested that the soils in Beijing metropolitan region were contaminated by Hg, Cd, Cu, As, and Pb in varying degree, lying in the moderate pollution level. As a whole, the health risks posed by soil metals were acceptable or close to tolerable. Comparatively speaking, children and adult females were the relatively vulnerable populations for the non-carcinogenic and carcinogenic risks, respectively. Atmospheric deposition, fertilizers and agrochemicals, and natural source were apportioned as the potential sources determining the contents of trace metals in soils of Beijing area with contributions of 15.5%-16.4%, 5.9%-7.7% and 76.0%-78.6%, respectively.

  7. Neutron spectrometric evaluations in the Argentine research reactor RA-1

    International Nuclear Information System (INIS)

    Full text: The determination of the quantities dose equivalent H*(10) and personal dose equivalent Hp(10) in mixed field (n,γ) needs the knowledge of the related spectrum. In order to fulfill this aim spectrometer system has been built based on the combination of polyethylene spheres of different diameters (Bonner Spheres System-BSS) and a He3 proportional counter detector sensitive to thermal neutrons. The detector is located in the geometrical centre of each of the spheres and has an associated electronics with a charge preamplifier, an amplifier and a multichannel system that allows the outgoing spectrum analysis. In order to determine the neutron spectrum a deconvolution method is applied based on the LOUHI82 code. In this work are shown the spectra and the related values of H*(10) that have been got in five places of the reactor and in the command room with the BSS. (author)

  8. Neutron diffraction facilities at the high flux reactor, Petten

    Science.gov (United States)

    Ohms, C.; Youtsos, A. G.; Bontenbal, A.; Mulder, F. M.

    2000-03-01

    The High Flux Reactor in Petten is equipped with twelve beam tubes for the extraction of thermal neutrons for applications in materials and medical science. Beam tubes HB4 and HB5 are equipped with diffractometers for residual stress and powder investigations. Recently at HB4 the Large Component Neutron Diffraction Facility has been installed. It is a unique facility with respect to its capability of handling heavy components up to 1000 kg in residual stress testing. Its basic features are described and the first applications on thick piping welds are shown. The diffractometer at HB5 can be set up for powder and stress measurements. Recent applications include temperature dependent measurements on phase transitions in intermetallic compounds and on Li ion energy storage materials.

  9. Progress report on neutron activation analysis at Dalat Nuclear Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Tuan, Nguyen Ngoc [Nuclear Research Institute, Dalat (Viet Nam)

    2003-03-01

    Neutron Activation Analysis (NAA) is one of most powerful techniques for the simultaneous multi-elements analysis. This technique has been studied and applied to analyze major, minor and trace elements in Geological, Biological and Environmental samples at Dalat Nuclear Research Reactor. At the sixth Workshop, February 8-11, 1999, Yojakarta, Indonesia we had a report on Current Status of Neutron Activation Analysis using Dalat Nuclear Research Reactor. Another report on Neutron Activation Analysis at the Dalat Nuclear Research Reactor also was presented at the seventh Workshop in Taejon, Korea from November 20-24, 2000. So in this report, we would like to present the results obtained of the application of NAA at NRI for one year as follows: (1) Determination of the concentrations of noble, rare earth, uranium, thorium and other elements in Geological samples according to requirement of clients particularly the geologists, who want to find out the mineral resources. (2) The analysis of concentration of radionuclides and nutrient elements in foodstuffs to attend the program on Asian Reference Man. (3) The evaluation of the contents of trace elements in crude oil and basement rock samples to determine original source of the oil. (4) Determination of the elemental composition of airborne particle in the Ho Chi Minh City for studying air pollution. The analytical data of standard reference material, toxic elements and natural radionuclides in seawater are also presented. (author)

  10. Hydride Formation in Neutron Irradiated Material Under In Reactor Conditions

    International Nuclear Information System (INIS)

    The present is a brief summary of the three reports completed within the framework of the SPAR III project. The following is a resume of our aims, techniques used to achieve the objectives and conclusions attained under the guiding thread of the hydride formation in neutron irradiated zirconium alloys and other reactor in operating conditions. As is it known, under reactor operating conditions zirconium components go through transformations which affect their original microstructural and thermodynamical properties. Both concerns are starting points of many research lines for the zirconium alloys used in the nuclear power reactors. Regarding microstructural transformations, one of the most important topics is the phase stability of these alloys. To cite a well-known case, second phase particles of zircaloy-4 shown to be unstable under neutron radiation. Since such phases play a role in the corrosion rate control, this instability became a problem for high burnup fuel claddings design. Similar observations can be made about the β−Zr phase in the Zr-2.5Nb CANDU pressure tubes alloy. On the other hand, there are issues directly involved with thermodynamics, e.g., hydrogen behaviour and its role in the degradation processes of fuel assemblies and other zirconium alloys components, which showed to be affected by neutron radiation. Finally, applied stresses and thermal cycling are part of these operating conditions, which can be simulated performing experiments in situ which allows testing hydrogen solubility behaviour and hydride reorientation. In the context described above, the research topics proposed to SPAR III were aimed to improve the knowledge of these degradation processes. In this scheme, zircaloy-4 which remained more than ten years at full power operation and virgin unirradiated zirconium alloys were suited by the more improved micro analytical techniques to characterize microstructural transformations cited above

  11. On the neutron spatial distribution in ionization chamber channels of the WWER type reactors

    International Nuclear Information System (INIS)

    Results of experimental and calculational studies permitting to estimate the neutron flux spatial distribution in ionization chamber channels of the commercial WWER-1000 and WWER-440 reactors and also of the WWER-440 reactor with water biological shield are presented. The integral neutron flux density distribution along the channel cross section approximately at height of the core middle and the corresponding thermal and fast neutron flux density distributions are measured by the activation detectors. It is shown that the difference in fast neutron flux density exceeds that of thermal neutrons. The commercial WWER-1000 type reactor the fast neutron flux density is decreased by the factor of 1.7, and thermal neutron flux density - by the factor of 1.2, for the commercial WWER-440 reactor these values are 1.37 and 1.18, and for the WWER-440 one with water shield - 1.5 and 1.18

  12. A comparison of different neutron spectroscopy systems at the reactor facility VENUS

    CERN Document Server

    Vanhavere, F; Chartier, J L; Itie, C; Rosenstock, W; Koeble, T; D'Errico, F

    2002-01-01

    The VENUS facility is a zero-power research reactor mainly devoted to studies on LWR fuels. Localised high-neutron rates were found around the reactor, with a neutron/gamma dose equivalent rate ratio as high as three. Therefore, a study of the neutron dosimetry around the reactor was started some years ago. During this study, several methods of neutron spectroscopy were employed and a study of individual and ambient dosemeters was performed. A first spectrometric measurement was done with the IPSN multisphere spectrometer in three positions around the reactor. Secondly, the ROSPEC spectrometer from the Fraunhofer Institut was used. The spectra were also measured with the bubble interactive neutron spectrometer. These measurements were compared with a numerical simulation of the neutron field made with the code TRIPOLI-3. Dosimetric measurements were made with three types of personal neutron dosemeters: an albedo type, a track etch detector and a bubble detector.

  13. Measuring Neutron Spectrum at MIT Research Reactor Utilizing He-3 Bonner Cylinder Approach with an Unfolding Analysis

    Science.gov (United States)

    Leder, Alexander; Ricochet Collaboration

    2016-03-01

    The Ricochet experiment seeks to measure Coherent (neutral-current) Elastic Neutrino-Nucleus Scattering (CENNS) using dark matter style detectors placed near a neutrino source, possibly the MIT research reactor (MITR), which offers a high continuous neutrino flux at high energies. Currently, Ricochet is characterizing the backgrounds at MITR. The main background is the neutrons emitted simultaneously from the core. To characterize this background, we wrapped a Bonner cylinder around a 3He thermal neutron detector, whose data was then unfolded to produce a neutron energy spectrum across several orders of magnitude. We discuss the resulting spectrum as well its implications for deploying Ricochet in the future.

  14. The LICORNE Neutron Source and Measurements of Prompt γ-rays Emitted in Fission

    Science.gov (United States)

    Wilson, J. N.; Lebois, M.; Halipre, P.; Oberstedt, S.; Oberstedt, A.

    The emission of prompt gamma rays is one of the least measured and least well-understood parts of the fission process. Knowledge of prompt fission gamma spectra, mean energies and multiplicities are important for reactor gamma heating and hence linked to reactor safety. At the IPN Orsay we have developed a unique, directional, fast neutron source called LICORNE, intended initially to facilitate prompt fission gamma measurements. The ability of the IPN Orsay tandem accelerator to produce intense beams of 7Li is exploited to produce quasi mono-energetic neutrons between 0.5 - 4 MeV using the p(7Li, 7Be)n inverse reaction. The available fluxes of up to 7×107 neutrons/second/steradian are comparable to existing installations, but with two added advantages: (i) The kinematic focusing produces a natural neutron beam collimation which allows placement of gamma detectors adjacent to the irradiated sample unimpeded by source neutrons. (ii) The background of scattered neutrons in the experimental hall is drastically reduced. The dedicated neutron converter was commissioned in June 2013

  15. Neutron spectra at two beam ports of a TRIGA Mark III reactor loaded with HEU fuel

    International Nuclear Information System (INIS)

    The neutron spectra have been measured in two beam ports, one radial and another tangential, of the TRIGA Mark III nuclear reactor from the National Institute of Nuclear Research in Mexico. Measurements were carried out with the reactor core loaded with high enriched uranium fuel. Two reactor powers, 5 and 10 W, were used during neutron spectra measurements using a Bonner sphere spectrometer with a 6LiI(Eu) scintillator and 2, 3, 5, 8, 10 and 12 in.-diameter high-density polyethylene spheres. The neutron spectra were unfolded using the NSDUAZ unfolding code. For each spectrum total flux, mean energy and ambient dose equivalent were determined. Measured spectra show fission, epithermal and thermal neutrons, being harder in the radial beam port. - Highlights: • Neutron spectra of a TRIGA reactor were measured. • The reactor core is loaded with HEU. • The spectra were measured at two reactor beam ports. • Measurements were carried out at 5 and 10 W

  16. Study of fluid fuel influence on delayed neutron in Molten Salt Reactor

    International Nuclear Information System (INIS)

    Background: The Molten Salt Reactor (MSR) is one of the six advanced reactor types for future nuclear energy systems in the Generation IV International Forum (GIF), possessing the advantages of good neutron economics, inherent safety, processing online, nuclear nonproliferation, etc. Due to the fluid fuel, a part of delayed neutron precursors drifts out of the reactor and decays in the loop, which is different from the solid-fuel reactor. Purpose: The distribution of delayed neutrons with different fuel velocity for Molten Salt Reactor Experiment (MSRE) is analyzed to provide reference for safety design of MSR. Methods: Based on the neutron kinetics and approximate method of homogeneous reactor, a distribution model of delayed neutron was presented. This model is applied to the distribution study of delayed neutrons with different fuel velocity. Results: The number of delayed neutrons in a core radius of 20 cm was more than that in a core radius of 30 cm. When the flow velocity of fuel decreased by a half, the discrepancy of delayed neutrons between the fifth-group and sixth-group precursors is within 6%. Conclusion: The number of delayed neutrons is increasing toward the center of reactor core. The group of delayed neutron precursors with shorter half-time has little effect on the flow velocity of molten salt. (authors)

  17. Neutron physics for nuclear reactors unpublished writings by Enrico Fermi

    CERN Document Server

    Fermi, Enrico; Pisanti, O

    2010-01-01

    This unique volume gives an accurate and very detailed description of the functioning and operation of basic nuclear reactors, as emerging from yet unpublished papers by Nobel Laureate Enrico Fermi. In the first part, the entire course of lectures on Neutron Physics delivered by Fermi at Los Alamos is reported, according to the version made by Anthony P French. Here, the fundamental physical phenomena are described very clearly and comprehensively, giving the appropriate physics grounds for the functioning of nuclear piles. In the second part, all the patents issued by Fermi (and coworkers) on

  18. Neutron radiography with RP-10 reactor and its practical applications

    International Nuclear Information System (INIS)

    Neutrography is a non destructive essay, their principal characteristics are the high neutron absorption by light elements and the high contrast of materials of similar thickness, the typical applications that we can mention are the analysis of nuclear fuels, detection of hydrogenated and organic materials, detection of flaws in turbine blades, corrosion in airships components, ceramic materials quality control, drugs and explosive materials detection (useful in the pyrotechnic industry and ammunitions), study of archaeological materials, detection of lubricating film in bearing systems as well as dynamic processes of lubrication and combustion and so for. In the present work, varied examples of applications obtained with the RP-10 reactor are shown. (orig.)

  19. Field Ion Source Development for Neutron Generators

    Energy Technology Data Exchange (ETDEWEB)

    B. Bargsten Johnson; P. R. Schwoebel; C. E. Holland; P. J. Resnick; K. L. Hertz; D. L. Chichester

    2012-01-01

    An ion source based on the principles of electrostatic field desorption is being developed to improve the performance of existing compact neutron generators. The ion source is an array of gated metal tips derived from field electron emitter array microfabrication technology. A comprehensive summary of development and experimental activities is presented. Many structural modifications to the arrays have been incorporated to achieve higher tip operating fields, while lowering fields at the gate electrode to prevent gate field electron emission which initiates electrical breakdown in the array. The latest focus of fabrication activities has been on rounding the gate electrode edge and surrounding the gate electrode with dielectric material. Array testing results have indicated a steady progression of increased array tip operating fields with each new design tested. The latest arrays have consistently achieved fields beyond those required for the onset of deuterium desorption ({approx}20 V/nm), and have demonstrated the desorption of deuterium at fields up to 36 V/nm. The number of ions desorbed from an array has been quantified, and field desorption of metal tip substrate material from array tips has been observed for the first time. Gas-phase field ionization studies with {approx}10,000 tip arrays have achieved deuterium ion currents of {approx}50 nA. Neutron production by field ionization has yielded {approx}10{sup 2} n/s from {approx}1 mm{sup 2} of array area using the deuterium-deuterium fusion reaction at 90 kV.

  20. Field ion source development for neutron generators

    Energy Technology Data Exchange (ETDEWEB)

    Bargsten Johnson, B. [University of New Mexico, Albuquerque, NM 87131 (United States); Schwoebel, P.R., E-mail: schwoebel@chtm.unm.edu [University of New Mexico, Albuquerque, NM 87131 (United States); Holland, C.E. [SRI International, Menlo Park, CA 94025 (United States); Resnick, P.J. [Sandia National Laboratories, Albuquerque, NM 87123 (United States); Hertz, K.L. [Sandia National Laboratories, Livermore, CA 94551 (United States); Chichester, D.L. [Idaho National Laboratory, Idaho Falls, ID 83415 (United States)

    2012-01-21

    An ion source based on the principles of electrostatic field desorption is being developed to improve the performance of existing compact neutron generators. The ion source is an array of gated metal tips derived from field electron emitter array microfabrication technology. A comprehensive summary of development and experimental activities is presented. Many structural modifications to the arrays have been incorporated to achieve higher tip operating fields, while lowering fields at the gate electrode to prevent gate field electron emission which initiates electrical breakdown in the array. The latest focus of fabrication activities has been on rounding the gate electrode edge and surrounding the gate electrode with dielectric material. Array testing results have indicated a steady progression of increased array tip operating fields with each new design tested. The latest arrays have consistently achieved fields beyond those required for the onset of deuterium desorption ({approx}20 V/nm), and have demonstrated the desorption of deuterium at fields up to 36 V/nm. The number of ions desorbed from an array has been quantified, and field desorption of metal tip substrate material from array tips has been observed for the first time. Gas-phase field ionization studies with {approx}10,000 tip arrays have achieved deuterium ion currents of {approx}50 nA. Neutron production by field ionization has yielded {approx}10{sup 2} n/s from {approx}1 mm{sup 2} of array area using the deuterium-deuterium fusion reaction at 90 kV.

  1. Cronos 2: a neutronic simulation software for reactor core calculations

    International Nuclear Information System (INIS)

    The CRONOS2 software is that part of the SAPHYR code system dedicated to neutronic core calculations. CRONOS2 is a powerful tool for reactor design, fuel management and safety studies. Its modular structure and great flexibility make CRONOS2 an unique simulation tool for research and development for a wide variety of reactor systems. CRONOS2 is a versatile tool that covers a large range of applications from very fast calculations used in training simulators to time and memory consuming reference calculations needed to understand complex physical phenomena. CRONOS2 has a procedure library named CPROC that allows the user to create its own application environment fitted to a specific industrial use. (authors)

  2. Excitation of neutron flux waves in reactor core transients

    International Nuclear Information System (INIS)

    An analysis of the excitation of neutron flux waves in reactor core transients has been performed. A perturbation theory solution has been developed for the time-dependent thermal diffusion equation in which the absorption cross section undergoes a rapid change, as in a PWR rod ejection accident (REA). In this analysis the unperturbed reactor flux states provide the basis for the spatial representation of the flux solution. Using a simplified space-time representation for the cross section change, the temporal integrations have been carried out and analytic expressions for the modal flux amplitudes determined. The first order modal excitation strength is determined by the spatial overlap between the initial and final flux states, and the cross section perturbation. The flux wave amplitudes are found to be largest for rapid transients involving large reactivity perturbations

  3. Measurement and analysis of leakage neutron energy spectra around the Kinki University Reactor, UTR-KINKI

    CERN Document Server

    Ogawa, Y; Sagawa, H; Tsujimoto, T

    2002-01-01

    The highly sensitive cylindrical multi-moderator type neutron spectrometer was constructed for measurement of low level environmental neutrons. This neutron spectrometer was applied for the determination of leakage neutron energy spectra around the Kinki University Reactor. The analysis of the leakage neutron energy spectra was performed by MCNP Monte Carlo code. From the obtained results, the agreement between the MCNP predictions and the experimentally determined values is fairly good, which indicates the MCNP model is correctly simulating the UTR-KINKI.

  4. Chemical characterization and source apportionment of PM2.5 in Beijing: seasonal perspective

    Directory of Open Access Journals (Sweden)

    R. Zhang

    2013-07-01

    Full Text Available In this study, 121 daily PM2.5 (aerosol particle with aerodynamic diameter less than 2.5 μm samples were collected from an urban site in Beijing in four months between April 2009 and January 2010 representing the four seasons. The samples were determined for various compositions, including elements, ions, and organic/elemental carbon. Various approaches, such as chemical mass balance, positive matrix factorization (PMF, trajectory clustering, and potential source contribution function (PSCF, were employed for characterizing aerosol speciation, identifying likely sources, and apportioning contributions from each likely source. Our results have shown distinctive seasonality for various aerosol speciations associated with PM2.5 in Beijing. Soil dust waxes in the spring and wanes in the summer. Regarding the secondary aerosol components, inorganic and organic species may behave in different manners. The former preferentially forms in the hot and humid summer via photochemical reactions, although their precursor gases, such as SO2 and NOx, are emitted much more in winter. The latter seems to favorably form in the cold and dry winter. Synoptic meteorological and climate conditions can overwhelm the emission pattern in the formation of secondary aerosols. The PMF model identified six main sources: soil dust, coal combustion, biomass burning, traffic and waste incineration emission, industrial pollution, and secondary inorganic aerosol. Each of these sources has an annual mean contribution of 16, 14, 13, 3, 28, and 26%, respectively, to PM2.5. However, the relative contributions of these identified sources significantly vary with changing seasons. The results of trajectory clustering and the PSCF method demonstrated that regional sources could be crucial contributors to PM pollution in Beijing. In conclusion, we have unraveled some complex aspects of the pollution sources and formation processes of PM2.5 in Beijing. To our knowledge, this is the first

  5. Chemical characterization and source apportionment of PM2.5 in Beijing: seasonal perspective

    Directory of Open Access Journals (Sweden)

    R. Zhang

    2013-04-01

    Full Text Available In this study, 121 daily PM2.5 (aerosol particle with aerodynamic diameter less than 2.5 μm samples were collected from an urban site in Beijing in four months between April 2009 and January 2010 representing the four seasons. The samples were determined for various compositions, including elements, ions, and organic/elemental carbon. Various approaches, such as chemical mass balance, positive matrix factorization (PMF, trajectory clustering, and potential source contribution function (PSCF, were employed for characterizing aerosol speciation, identifying likely sources, and apportioning contributions from each likely source. Our results have shown distinctive seasonalities for various aerosol speciations associated with PM2.5 in Beijing. Soil dust waxes in the spring and wanes in the summer. Regarding the secondary aerosol components, inorganic and organic species may behave in different manners. The former preferentially forms in the hot and humid summer via photochemical reactions, although their precursor gases, such as SO2 and NOx, are emitted much more in winter. The latter seems to favorably form in the cold and dry winter. Synoptic meteorological and climate conditions can overwhelm the emission pattern in the formation of secondary aerosols. The PMF model identified six main sources: soil dust, coal combustion, biomass burning, traffic and waste incineration emission, industrial pollution, and secondary inorganic aerosol. Each of these sources has an annual mean contribution of 16, 14, 13, 3, 28, and 26%, respectively, to PM2.5. However, the relative contributions of these identified sources significantly vary with changing seasons. The results of trajectory clustering and the PSCF method demonstrated that regional sources could be crucial contributors to PM pollution in Beijing. In conclusion, we have unraveled some complex aspects of the pollution sources and formation processes of PM2.5 in Beijing. To our knowledge, this study is

  6. Chemical characterization and source apportionment of PM2.5 in Beijing: seasonal perspective

    Science.gov (United States)

    Zhang, R.; Jing, J.; Tao, J.; Hsu, S.-C.; Wang, G.; Cao, J.; Lee, C. S. L.; Zhu, L.; Chen, Z.; Zhao, Y.; Shen, Z.

    2013-07-01

    In this study, 121 daily PM2.5 (aerosol particle with aerodynamic diameter less than 2.5 μm) samples were collected from an urban site in Beijing in four months between April 2009 and January 2010 representing the four seasons. The samples were determined for various compositions, including elements, ions, and organic/elemental carbon. Various approaches, such as chemical mass balance, positive matrix factorization (PMF), trajectory clustering, and potential source contribution function (PSCF), were employed for characterizing aerosol speciation, identifying likely sources, and apportioning contributions from each likely source. Our results have shown distinctive seasonality for various aerosol speciations associated with PM2.5 in Beijing. Soil dust waxes in the spring and wanes in the summer. Regarding the secondary aerosol components, inorganic and organic species may behave in different manners. The former preferentially forms in the hot and humid summer via photochemical reactions, although their precursor gases, such as SO2 and NOx, are emitted much more in winter. The latter seems to favorably form in the cold and dry winter. Synoptic meteorological and climate conditions can overwhelm the emission pattern in the formation of secondary aerosols. The PMF model identified six main sources: soil dust, coal combustion, biomass burning, traffic and waste incineration emission, industrial pollution, and secondary inorganic aerosol. Each of these sources has an annual mean contribution of 16, 14, 13, 3, 28, and 26%, respectively, to PM2.5. However, the relative contributions of these identified sources significantly vary with changing seasons. The results of trajectory clustering and the PSCF method demonstrated that regional sources could be crucial contributors to PM pollution in Beijing. In conclusion, we have unraveled some complex aspects of the pollution sources and formation processes of PM2.5 in Beijing. To our knowledge, this is the first systematic study

  7. Regional source identification of atmospheric aerosols in Beijing based on sulfur isotopic compositions

    Science.gov (United States)

    Lianfang, Wei; Pingqing, Fu; Xiaokun, Han; Qingjun, Guo; Yele, Sun; Zifa, Wang

    2016-04-01

    65 daily PM2.5 (aerosol particle with aerodynamic diameter less than 2.5 μm) samples were collected from an urban site in Beijing in four months representing the four seasons between September 2013 and July 2014. Inorganic ions, organic/elemental carbon and stable sulfur isotopes of sulfate aerosols were analyzed systematically. The "fingerprint" characteristics of the stable sulfur isotopic composition, together with trajectory clustering modeled by HYSPLIT-4 and potential source contribution function (PSCF), were employed for identifying potential regional sources. Results obviously exhibited the distinctive seasonality for various aerosol speciation associated with PM2.5 in Beijing with sulfate, nitrate, ammonium, organic matter, and element carbon being the dominant species. Elevated chloride associated with higher concentration of organics were found in autumn and winter, due to enhanced coal combustion emissions. The δ34S values of Beijing aerosol samples ranged from 2.94‰ to 10.2‰ with an average value of 6.18±1.87‰ indicating that the major sulfur source is direct fossil fuel burning-related emissions. Owning to a temperature-dependent fractionation and elevated biogenic sources of isotopically light sulfur in summer, the δ34S values had significant seasonal variations with a winter maximum ( 8.6‰)and a summer minimum ( 5.0‰). The results of trajectory clustering and the PSCF method demonstrated that higher concentrations of sulfate with lower sulfur isotope ratios ( 4.83‰) were associated with air masses from the south, southeast or east, whereas lower sulfate concentrations with higher δ34S values ( 6.69‰) when the air masses were mainly from north or northwest. These results suggested two main different kinds of regional coal combustion sources contributed to the pollution in Beijing.

  8. [Testing of Concentration and Characteristics of Particulate Matters Emitted from Stationary Combustion Sources in Beijing].

    Science.gov (United States)

    Hu, Yue-qi; Wu, Xiao-dong; Wang, Chen; Liang, Yun-ping; Ma, Zhao-hui

    2016-05-15

    A self-built monitoring sampling system on particulate matters and water soluble ions emitted from stationary combustion sources and a size separated sampling system on particulate matters based on FPS4000 and ELPI + were applied to test particulate matters in fumes of typical stationary combustion sources in Beijing. The results showed that the maximum concentration of total particulate matters in fumes of stationary combustion sources in Beijing was 83.68 mg · m⁻³ in standard smoke oxygen content and the minimum was 0.12 mg · m⁻³. And particle number concentration was in the 10⁴-10⁶ cm⁻³ number of grade. Both mass and number concentration ranking order of particulate matters emitted from stationary combustion sources in Beijing was: heating gas fired boilers power plant coal fired boilers coal fired boilers. And two or three peaks existed under 1 µm of particulate size for both number size distribution and mass size distribution. The number concentration for PM₂.₅ accounted for over 99.8% of that for PM₁₀ and that for PM₀.₁ accounted for over 83% of that for PM₂.₅. But the proportions of PM₀.₁, and PM₂.₅ in PM₁₀ were significantly lower in quality analysis,the proportion of PM₂.₅ in PM₁₀ was about 82%, and that of PM₀.₁ in PM₂.₅ was about 27%-33%. PMID:27506016

  9. Consumption and Sources of Dietary Salt in Family Members in Beijing

    Directory of Open Access Journals (Sweden)

    Fang Zhao

    2015-04-01

    Full Text Available In China, few people are aware of the amount and source of their salt intake. We conducted a survey to investigate the consumption and sources of dietary salt using the “one-week salt estimation method” by weighing cooking salt and major salt-containing food, and estimating salt intake during dining out based on established evidence. Nine hundred and three families (1981 adults and 971 children with students in eight primary or junior high schools in urban and suburban Beijing were recruited. On average, the daily dietary salt intake of family members in Beijing was 11.0 (standard deviation: 6.2 g for children and adolescents (under 18 years old, 15.2 (9.1 g for adults (18 to 59 years old, and 10.2 (4.8 g for senior citizens (60 years old and over, respectively. Overall, 60.5% of dietary salt was consumed at home, and 39.5% consumed outside the home. Approximately 90% of the salt intake came from cooking (household cooking and cafeteria or restaurant cooking, while less than 10% came from processed food. In conclusion, the dietary salt intake in Beijing families far surpassed the recommended amounts by World Health Organization, with both household cooking and dining-out as main sources of salt consumption. More targeted interventions, especially education about major sources of salt and corresponding methods for salt reduction should be taken to reduce the risks associated with a high salt diet.

  10. Analysis of calculated neutron flux response at detectors of G.A. Siwabessy multipurpose reactor (RSG-GAS Reactor)

    International Nuclear Information System (INIS)

    Multi Purpose Reactor G.A. Siwabessy (RSG-GAS) reactor core possesses 4 fission-chamber detectors to measure intermediate power level of RSG-GAS reactor. Another detector, also fission-chamber detector, is intended to measure power level of RSG-GAS reactor. To investigate influence of space to the neutron flux values for each detector measuring intermediate and power levels has been carried out. The calculation was carried out using combination of WIMS/D4 and CITATION-3D code and focused on calculation of neutron flux at different detector location of RSG-GAS typical working core various scenarios. For different scenarios, all calculation results showed that each detector, located at different location in the RSG-GAS reactor core, causes different neutron flux occurred in the reactor core due to spatial time effect

  11. Time collimation for elastic neutron scattering at a pulsed source

    International Nuclear Information System (INIS)

    Conditions for carrying out elastic neutron scattering experiments using the time-of-flight technique are considered. It is shown, that the employment of time dependent neutron beam collimation in the source-sample flight path increases the luminosity of the spectrometer under certain resolution restrictions. Time collimation modes are proposed for small-angle scattering and neutron reflection. (author) 8 figs., 3 refs

  12. PGNAA neutron source moderation setup optimization

    OpenAIRE

    Zhang, Jinzhao; Tuo, Xianguo

    2013-01-01

    Monte Carlo simulations were carried out to design a prompt {\\gamma}-ray neutron activation analysis (PGNAA) thermal neutron output setup using MCNP5 computer code. In these simulations the moderator materials, reflective materials and structure of the PGNAA 252Cf neutrons of thermal neutron output setup were optimized. Results of the calcuations revealed that the thin layer paraffin and the thick layer of heavy water moderated effect is best for 252Cf neutrons spectrum. The new design compar...

  13. Fission, spallation or fusion-based neutron sources

    Indian Academy of Sciences (India)

    Kurt N Clausen

    2008-10-01

    In this paper the most promising technology for high power neutron sources is briefly discussed. The conclusion is that the route to high power neutron sources in the foreseeable future is spallation – short or long pulse or even CW – all of these sources will have areas in which they excel.

  14. Major reactive species of ambient volatile organic compounds (VOCs) and their sources in Beijing

    Institute of Scientific and Technical Information of China (English)

    SHAO; Min; FU; Linlin; LIU; Ying; LU; Sihua; ZHANG; Yuanhan

    2005-01-01

    Volatile organic compounds (VOCs) are important precursors of atmospheric chemical processes. As a whole mixture, the ambient VOCs show very strong chemical reactivity. Based on OH radical loss rates in the air, the chemical reactivity of VOCs in Beijing was calculated. The results revealed that alkenes, accounting for only about 15% in the mixing ratio of VOCs, provide nearly 75% of the reactivity of ambient VOCs and the C4 to C5 alkenes were the major reactive species among the alkenes. The study of emission characteristics of various VOCs sources indicated that these alkenes are mainly from vehicle exhaust and gasoline evaporation. The reduction of alkene species in these two sources will be effective in photochemical pollution control in Beijing.

  15. Spectrum of isotopic neutron sources inside concrete wall spherical cavities

    International Nuclear Information System (INIS)

    The neutron spectra of 252Cf/D2O, 140LaBe, 252Cf, 238Pu18O2, 241AmB, 241AmBe, 226RaBe and 239PuBe isotopic neutron sources due to room-return have been determined for various source-to-detector distances in concrete spherical cavities of different radius. Changes in the amount of thermal neutrons (E≤0.414eV) were analyzed to estimate, for each neutron source, the coefficient that relates the neutron source strength and room surface area with the thermal neutron fluence rates. The study was carried out using Monte Carlo methods for 200, 400, 500, 800, 1000, 1200 and 1500-cm-radius spherical cavity in vacuum; cavities are 100-cm-thick concrete. Point sources were located at the center of cavity and neutron spectra were calculated at several source-to-detector distances along the cavity radius. The thermal neutron contribution was thereby evaluated. From these calculations a weighted coefficient value that relates the thermal neutron fluence with the neutron source strength and the total inner area surface of the cavity was estimated to be 3.76±0.03

  16. 改进的源倍增方法测量控制棒价值%The Control Rod Worth Measurement With Improved Neutron Source Multiplication Method

    Institute of Scientific and Technical Information of China (English)

    史永谦; 李义国; 鲁谨; 洪景彦; 吴小波; 彭旦

    2015-01-01

    该文给出了改进的源倍增方法测量控制棒价值的原理,在高富集度235 U 燃料元件转换为低富集度235 U 的微型中子源零功率反应堆上进行研究,实验测量微型中子源零功率反应堆中心控制棒的价值,与周期方法相比在2%内符合,但减少了测量时间。该方法为今后加速器驱动次临界系统 ADS 的次临界在线监督提供一种可能的方法。%Control rod worth measurement principle with improved neutron source multiplication (INSM)method is described and its worth measurement was completed on the miniature neutron source reactor by this method. Worth of control rod located in the center of the core is in accordance with the value measured by period method within the error 2%,but the measuring time is quite short. INSM can be used to monitor the sub-critical reactivity of accelerator driven sub-critical system (ADS)on line in the future.

  17. Nondestructive analysis of the natural uranium mass through the measurement of delayed neutrons using the technique of pulsed neutron source

    International Nuclear Information System (INIS)

    This work presents results of non destructive mass analysis of natural uranium by the pulsed source technique. Fissioning is produced by irradiating the test sample with pulses of 14 MeV neutrons and the uranium mass is calculated on a relative scale from the measured emission of delayed neutrons. Individual measurements were normalised against the integral counts of a scintillation detector measuring the 14 MeV neutron intensity. Delayed neutrons were measured using a specially constructed slab detector operated in anti synchronism with the fast pulsed source. The 14 MeV neutrons were produced via the T(d,n) 4He reaction using a 400 kV Van de Graaff accelerated operated at 200 kV in the pulsed source mode. Three types of sample were analysed, namely: discs of metallic uranium, pellets of sintered uranium oxide and plates of uranium aluminium alloy sandwiched between aluminium. These plates simulated those of Material Testing Reactor fuel elements. Results of measurements were reproducible to within an overall error in the range 1.6 to 3.9%; the specific error depending on the shape, size and mass of the sample. (author)

  18. 快Z箍缩中子源混合堆界面研究进展%Development of interface options of hybrid reactor driven with fast Z-pinch neutron source

    Institute of Scientific and Technical Information of China (English)

    陈敬平; 王雄

    2011-01-01

    评述了快Z箍缩中子产生及诊断的最新进展,介绍了聚变裂变混合堆原理与结构.概述了混合堆界面的磁绝缘传输线(MITL)和碎片防护罩设计,提出了MITL电流压力建模思路,提出了PTS装置上MITL翻转柱孔汇流结构(PHC)及同轴延伸方式,这两种配置方式简便、易行.%The recent development of neutron generation and diagnostics of fast Z-pinch are reviewed. The principle and con-figuration of fusion and fission hybrid reactor are briefly introduced. Current and magnetic impulse modeling and the debris shield design are examined for the interface between Z-pinch driver and hybrid reactor. The conclusion of this work is that the interface of post hole convolute and extended coaxial magnetically insulated transmission line is feasible and easily operated at primary test stand(PTS).

  19. Measurements of the subcriticality using advanced technique of shooting source during operation of NPP reactors

    Science.gov (United States)

    Lebedev, G. V.; Petrov, V. V.; Bobylyov, V. T.; Butov, R. I.; Zhukov, A. M.; Sladkov, A. A.

    2014-12-01

    According to the rules of nuclear safety, the measurements of the subcriticality of reactors should be carried out in the process of performing nuclear hazardous operations. An advanced technique of shooting source of neutrons is proposed to meet this requirement. As such a source, a pulsed neutron source (PNS) is used. In order to realize this technique, it is recommended to enable a PNS with a frequency of 1-20 Hz. The PNS is stopped after achieving a steady-state (on average) number of neutrons in the reactor volume. The change in the number of neutrons in the reactor volume is measured in time with an interval of discreteness of ˜0.1 s. The results of these measurements with the application of a system of point-kinetics equations are used in order to calculate the sought subcriticality. The basic idea of the proposed technique used to measure the subcriticality is elaborated in a series of experiments on the Kvant assembly. The conditions which should be implemented in order to obtain a positive result of measurements are formulated. A block diagram of the basic version of the experimental setup is presented, whose main element is a pulsed neutron generator.

  20. Neutron leakage from Pb and Bc spherical shells with 14 MeV central neutron source

    International Nuclear Information System (INIS)

    Results of measuring neutron leakage from spherical shells of different thickness, made of Pb and Be with a point neutron source in the sphere centrum are presented. The experiment results are compared to calculations according to different programs using data of various nuclear data libraies. The comparison has shown that all the calculations understate the neutron leakage from Pb assmebly. 9 refs.; 2 tabs