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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. Investigating The Neutron Flux Distribution Of The Miniature Neutron Source Reactor MNSR Type

    International Nuclear Information System (INIS)

    Neutron flux distribution is the important characteristic of nuclear reactor. In this article, four energy group neutron flux distributions of the miniature neutron source reactor MNSR type versus radial and axial directions are investigated in case the control rod is fully withdrawn. In addition, the effect of control rod positions on the thermal neutron flux distribution is also studied. The group constants for all reactor components are generated by the WIMSD code, and the neutron flux distributions are calculated by the CITATION code. The results show that the control rod positions only affect in the planning area for distribution in the region around the control rod. (author)

  3. 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.)

  4. 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.)

  5. 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

  6. 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)

  7. 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.

  8. Investigating The Integral Control Rod Worth Of The Miniature Neutron Source Reactor MNSR

    International Nuclear Information System (INIS)

    Determining control rod characteristics is an essential problem of nuclear reactor analysis. In this research, the integral control rod worth of the miniature neutron source reactor MNSR is investigated. Some other parameters of the nuclear reactor, such as core excess reactivity, shut down margin, are also calculated. Group constants for all reactor components are generated by the WIMSD code and then are used in the CITATION code to solve the neutron diffusion equations. The maximum relative error of the calculated results compared with the measurement data is about 3.5%. (author)

  9. 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)

  10. 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)

  11. 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

  12. 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.

  13. Radionuclide survey for Ji'nan miniature neutron source reactor decommissioning

    International Nuclear Information System (INIS)

    Radionuclides in upper reactor container, pool materials. resin and waste were surveyed on the decommissioning site of Ji'nan Miniature Neutron Source Reactor, and the survey of pool materials was described in detail. The radionuclides in surveyed materials include 60Co, 152Eu, 137Cs and 54Mn with higher activity concentrations in materials at pool bottom just below the reactor container. The survey results indicate that upper reactor container, resin for reactor water purification, waste and a part of pool materials are low level waste. A certified reference material GBW08304a was used in this survey for quality control. The measured values agree well with the standard values within ±15%. (authors)

  14. 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)

  15. 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

  16. 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

  17. 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

  18. 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)

  19. 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.

  20. 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.

  1. 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.

  2. 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.

  3. 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

  4. Cold neutron source at the Budapest reactor

    International Nuclear Information System (INIS)

    The installation of a liquid hydrogen cold neutron source assembly with a single closed circuit feed by two cryogenerators and utilizing the thermosyphon principle is in progress at the reconstructed Budapest reactor. The end of the in-pile part is a nearly tangential horizontal channel with a moderator cell of 250 cm3 volume made of aluminium alloy located in a hole inside the Be-reflector. The cold neutrons will be directed to the user positions by three mirror guide tubes. (orig.)

  5. Applications of the advanced neutron source reactor

    International Nuclear Information System (INIS)

    When the technique of neutron scattering was pioneered at the X-10 graphite reactor at Oak Ridge National Laboratory about 50 years ago, it was used to study certain important, but fairly esoteric, properties of crystals. From this modest beginning, neutron scattering has become a major tool in every branch of science, from the astrophysics of the early universe to human biology, and in many important industrial and engineering applications. In a typical modern research reactor it is not unusual to find one instrument studying new polymeric materials, while its neighbor is measuring residual stress in a jet turbine, sometimes with the jet operating. Most of this development has taken place outside of the United States, primarily in Western Europe, Japan and Russia, and it is generally recognized that we are a decade behind our competitors in this important field. The Advanced Neutron Source (ANS), planned to become operational as a user-facility at Oak Ridge at the end of this decade, will regain our leadership in neutron-based research and will be a major center for attracting new students into science. This paper discusses some of the research and development applications of the ANS, with an emphasis on applied materials science and engineering

  6. 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

  7. Crystal Driven Neutron Source: A New Paradigm for Miniature Neutron Sources

    International Nuclear Information System (INIS)

    Neutron interrogation techniques have specific advantages for detection of hidden, shielded, or buried threats over other detection modalities in that neutrons readily penetrate most materials providing backscattered gammas indicative of the elemental composition of the potential threat. Such techniques have broad application to military and homeland security needs. Present neutron sources and interrogation systems are expensive and relatively bulky, thereby making widespread use of this technique impractical. Development of a compact, high intensity crystal driven neutron source is described. The crystal driven neutron source approach has been previously demonstrated using pyroelectric crystals that generate extremely high voltages when thermal cycled. Placement of a sharpened needle on the positively polarized surface of the pyroelectric crystal results in sufficient field intensification to field ionize background deuterium molecules in a test chamber, and subsequently accelerate the ions to energies in excess of ∼100 keV, sufficient for either D-D or D-T fusion reactions with appropriate target materials. Further increase in ion beam current can be achieved through optimization of crystal thermal ramping, ion source and crystal accelerator configuration. The advantage of such a system is the compact size along with elimination of large, high voltage power supplies. A novel implementation discussed incorporates an independently controlled ion source in order to provide pulsed neutron operation having microsecond pulse width.

  8. Crystal Driven Neutron Source: A New Paradigm for Miniature Neutron Sources

    Science.gov (United States)

    Tang, V.; Morse, J.; Meyer, G.; Falabella, S.; Guethlein, G.; Kerr, P.; Park, H. G.; Rusnak, B.; Sampayan, S.; Schmid, G.; Spadaccini, C.; Wang, L.

    2009-03-01

    Neutron interrogation techniques have specific advantages for detection of hidden, shielded, or buried threats over other detection modalities in that neutrons readily penetrate most materials providing backscattered gammas indicative of the elemental composition of the potential threat. Such techniques have broad application to military and homeland security needs. Present neutron sources and interrogation systems are expensive and relatively bulky, thereby making widespread use of this technique impractical. Development of a compact, high intensity crystal driven neutron source is described. The crystal driven neutron source approach has been previously demonstrated using pyroelectric crystals that generate extremely high voltages when thermal cycled [1-4]. Placement of a sharpened needle on the positively polarized surface of the pyroelectric crystal results in sufficient field intensification to field ionize background deuterium molecules in a test chamber, and subsequently accelerate the ions to energies in excess of ˜100 keV, sufficient for either D-D or D-T fusion reactions with appropriate target materials. Further increase in ion beam current can be achieved through optimization of crystal thermal ramping, ion source and crystal accelerator configuration. The advantage of such a system is the compact size along with elimination of large, high voltage power supplies. A novel implementation discussed incorporates an independently controlled ion source in order to provide pulsed neutron operation having microsecond pulse width.

  9. The 'RB' Reactor as a Source of Fast Neutrons

    International Nuclear Information System (INIS)

    A study of the RB reactor as possible source of fast neutrons began in 1976 and four different version of fast neutron sources are designed up to 1990: an external neutron converter - ENC (1976), an experimental fuel channel - EFC (1982), an internal neutron converter - INC (1983), and a coupled fast-thermal core - HERBE (1990). An overview of applications and characteristics of each particular source of fast neutrons, including available irradiation space, neutron spectra and equivalent neutron and gamma dose rates is presented in the paper. Control and safety-related implications of these modifications of the reactor are emphasised. Computer codes and nuclear data libraries, used in calculations, are described. (author)

  10. Numerical modeling of velocity distribution of coolant flow in fuel channel of a miniature neutron source reactor using navier-stokes equations

    International Nuclear Information System (INIS)

    One major requirement for safe operation of a nuclear reactor is adequate cooling system during normal and emergency conditions. Analysis of the heat and mass transfer, and coolant flow rate variables during operations of nuclear reactors are required for performing risk and hazard management to plan optimum reactor recovery strategies. The velocity distribution of coolant in the channel of the Ghana Research Reactor -1 (GHARR-1) is of major concern because when the velocity of the coolant is too fast it results in pool cooling. On the contrary when the flow rate is also slow, the possibility of boiling occurs and hence poor cooling. This is because the virtually stagnant coolant will take much of the heat generated from the core and would begin to boil which can also lead to core meltdown. These parameters can be determined through experimental setup and computer simulations using models based on the laws of fluid dynamics and thermodynamics. The research took into consideration a computer based model which used Navier Stokes equations of continuity, momentum and energy conservation to simulate flow patterns in the channels. The Navier Stokes equation was then expressed in algorithm using the Marker and Cell (MAC) finite different technique. The algorithm equations were then developed into matrix form (algebraic equations) by discretization. MATLAB has been used to code and solve the resulting finite different equations numerically to determine the velocity fields of coolant in the GHARR-1 reactor core channel. The velocity distribution in the Ghana research reactor (GHARR-1) was determined to be in the range of 0.9 m/s to 1.9 m/s. It is observed that the flow in the hottest channel was faster than the channel which were far from the center of the core and hence helped in removing much heat from the core of the reactor and ensuring reactor safety. (au)

  11. 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.)

  12. Project and supply agreement: Agreement of 28 February 1992 between the International Atomic Energy Agency and the Governments of the Syrian Arab Republic and the People's Republic of China concerning the transfer of a miniature neutron source reactor and enriched uranium

    International Nuclear Information System (INIS)

    The document reproduces the text of the Project and Supply Agreement between the Agency and the Governments of the Syrian Arab Republic and the People's Republic of China for the transfer of a 30 KW miniature neutron source reactor for radioisotope production, research and tracing and of approximately 980.40 grams of uranium enriched to approximately 90.2 percent by weight in the isotope uranium-235 contained in fuel elements for the supplied reactor. The Agreement was approved by the Agency's Board of Governors on 25 February 1992, signed in Vienna on 28 February 1992, and entered into force on 18 May 1992. 1 tab

  13. 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.)

  14. 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.

  15. 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

  16. 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)

  17. Intense neutron source requirements for fusion reactor materials development

    International Nuclear Information System (INIS)

    Materials research should precede machine construction by at least ten years because considerable time is required for the materials development. When the next generation machine is under discussion, materials scientists and engineers should consider next-next generation device as DEMO for establishing the materials database in time. In this sense, development of an intense high energy neutron source is an urgent problem. Characteristic features of radiation effects with 14 MeV neutrons will be briefly reviewed. Then, the reasons why we need intense source will be discussed. These discussions will lead to identify requirements for the intense neutron sources. There are both near term and long term materials issues which can be studied with such intense neutron sources depending on their capacity. One should also recognize that development of such an intense source will require considerable time and maximum use of existing intense fission reactor neutrons will be one of the practical options for the moment. In other words, the intense neutron sources under discussion should be superior for the study of fusion radiation effects than the existing fission reactors. Items are listed for the evaluation of the sources and some critical comments will be made on several kinds of sources currently being proposed. (author)

  18. Cold neutron source at the Budapest WWR-SM reactor

    International Nuclear Information System (INIS)

    Upgrading and complete reconstruction of the KFKI WWR-SM reactor includes the installation of a cold neutron source in order to improve neutron scattering facilities for condensed matter research. The principles of cold neutron moderators are given, and the operation as well as the main elements of a small size cell liquid hydrogen cold source planned to be installed are presented describing also the installation and testing procedures. The most important hazard factors and safety problems are analyzed. (author) 24 refs.; 8 figs.; 1 tab

  19. Advanced Neutron Source Reactor thermal analysis of fuel plate defects

    International Nuclear Information System (INIS)

    The Advanced Neutron Source Reactor (ANSR) is a research reactor designed to provide the highest continuous neutron beam intensity of any reactor in the world. The present technology for determining safe operations were developed for the High Flux Isotope Reactor (HFIR). These techniques are conservative and provide confidence in the safe operation of HFIR. However, the more intense requirements of ANSR necessitate the development of more accurate, but still conservative, techniques. This report details the development of a Local Analysis Technique (LAT) that provides an appropriate approach. Application of the LAT to two ANSR core designs are presented. New theories of the thermal and nuclear behavior of the U3Si2 fuel are utilized. The implications of lower fuel enrichment and of modifying the inspection procedures are also discussed. Development of the computer codes that enable the automate execution of the LAT is included

  20. Canadian Neutron Source (CNS): a research reactor solution for medical isotopes and neutrons for science

    International Nuclear Information System (INIS)

    This presentation describes a dual purpose research facility at the University of Saskatchewan for Canada for the production of medical isotopes and neutrons for scientific research. The proposed research reactor is intended to supply most of Canada's medical isotope requirements and provide a neutron source for Canada's research community. Scientific research would include materials research, biomedical research and imaging.

  1. Neutron measurements with miniature fission chambers

    International Nuclear Information System (INIS)

    This report analyses the use and qualification of miniature fission chambers for two types of neutron measurements: 1) relative measurements in nuclear power reactors, or nuclear irradiation reactors. We consider the problem of burnable fissile material in detectors, under important neutron exposures and conclude on recommending the use of regenerating neutron detectors. 2) Measurements of integral physical parameters in experimental reactors. This method has been applied to measurements of fission rates in detectors with strong alpha-rays emitter deposits in order to establish standard tables of physical parameters. It has been used in particular for detectors including very radioactive actinium series. After a discussion of the instrumentation choice and data processing, specific to each measurements, we provide results on: 1) the technological and neutron behaviour of regenerating miniature fission chambers, in the swimming pool reactor 'TRITON', under a neutron exposure of 7.1020 n cm-2 and a gamma-rays exposure of 1,5.1011 rads about. The fissile material is 235U and the fertile material 234U. 2) The integral measurements of effective cross sections for 241Am and 238Pu in the experimental reactor 'MINERVE' use for study of problems of fast neutron reactors

  2. 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

  3. 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.

  4. 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)

  5. 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-11 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 of corrosion or entrained moisture in large aircraft panels

  6. Advanced Neutron Source reactor control and plant protection systems design

    International Nuclear Information System (INIS)

    This paper describes the reactor control and plant protection systems' conceptual design of the Advanced Neutron Source (ANS). The Plant Instrumentation, Control, and Data Systems and the Reactor Instrumentation and Control System of the ANS are planned as an integrated digital system with a hierarchical, distributed control structure of qualified redundant subsystems and a hybrid digital/analog protection system to achieve the necessary fast response for critical parameters. Data networks transfer information between systems for control, display, and recording. Protection is accomplished by the rapid insertion of negative reactivity with control rods or other reactivity mechanisms to shut down the fission process and reduce heat generation in the fuel. The shutdown system is designed for high functional reliability by use of conservative design features and a high degree of redundance and independence to guard against single failures. Two independent reactivity control systems of different design principles are provided, and each system has multiple independent rods or subsystems to provide appropriate margin for malfunctions such as stuck rods or other single failures. Each system is capable of maintaining the reactor in a cold shutdown condition independently of the functioning of the other system. A highly reliable, redundant channel control system is used not only to achieve high availability of the reactor, but also to reduce challenges to the protection system by maintaining important plant parameters within appropriate limits. The control system has a number of contingency features to maintain acceptable, off-normal conditions in spite of limited control or plant component failures thereby further reducing protection system challenges

  7. Advanced neutron source reactor probabilistic flow blockage assessment

    International Nuclear Information System (INIS)

    The Phase I Level I Probabilistic Risk Assessment (PRA) of the conceptual design of the Advanced Neutron Source (ANS) Reactor identified core flow blockage as the most likely internal event leading to fuel damage. The flow blockage event frequency used in the original ANS PRA was based primarily on the flow blockage work done for the High Flux Isotope Reactor (HFIR) PRA. This report examines potential flow blockage scenarios and calculates an estimate of the likelihood of debris-induced fuel damage. The bulk of the report is based specifically on the conceptual design of ANS with a 93%-enriched, two-element core; insights to the impact of the proposed three-element core are examined in Sect. 5. In addition to providing a probability (uncertainty) distribution for the likelihood of core flow blockage, this ongoing effort will serve to indicate potential areas of concern to be focused on in the preliminary design for elimination or mitigation. It will also serve as a loose-parts management tool

  8. Reactor installation and maintenance for the Advanced Neutron Source

    International Nuclear Information System (INIS)

    Advanced Neutron Source (ANS) reactor assembly components have been modeled in great detail in IGRIP in order to realistically simulate preliminary installation and maintenance processes. Animation of these processes has been captured in a 15-minute video with narration. Approximately 90% of the parts were initially translated from CADAM (a two-dimensional drawing package) to IGRIP and then revolved or extruded. IGRIP's IGES translator greatly reduced the time required to perform this operation. The interfacing of devices in the work cell has identified numerous design inconsistencies. Most of the modeled reactor components are devices with a single degree of freedom (DOF) however, some of the slanted experiments required 6 DOF so that they could be removed at an angle in order to clear the reflector vessel flanges. IGRIP's collision detection feature proved to be extremely helpful in determining interferences when removing the experiments. The combination of three-dimensional visualization and collision detection allows engineers to clearly and easily visualize potential design problems before the construction phase of the project

  9. Capabilities of a DT tokamak fusion neutron source for driving a spent nuclear fuel transmutation reactor

    International Nuclear Information System (INIS)

    The capabilities of a DT fusion neutron source for driving a spent nuclear fuel transmutation reactor are characterized by identifying limits on transmutation rates that would be imposed by tokamak physics and engineering limitations on fusion neutron source performance. The need for spent nuclear fuel transmutation and the need for a neutron source to drive subcritical fission transmutation reactors are reviewed. The likely parameter ranges for tokamak neutron sources that could produce an interesting transmutation rate of 100s to 1000s of kg/FPY (where FPY stands for full power year) are identified (Pfus ∼ 10-100 MW, βN ∼ 2-3, Qp ∼ 2-5, R ∼ 3-5 m, I ∼ 6-10 MA). The electrical and thermal power characteristics of transmutation reactors driven by fusion and accelerator spallation neutron sources are compared. The status of fusion development vis-a-vis a neutron source is reviewed. (author)

  10. Capabilities of a DT tokamak fusion neutron source for driving a spent nuclear fuel transmutation reactor

    Science.gov (United States)

    Stacey, W. M.

    2001-02-01

    The capabilities of a DT fusion neutron source for driving a spent nuclear fuel transmutation reactor are characterized by identifying limits on transmutation rates that would be imposed by tokamak physics and engineering limitations on fusion neutron source performance. The need for spent nuclear fuel transmutation and the need for a neutron source to drive subcritical fission transmutation reactors are reviewed. The likely parameter ranges for tokamak neutron sources that could produce an interesting transmutation rate of 100s to 1000s of kg/FPY (where FPY stands for full power year) are identified (Pfus approx 10-100 MW, βN approx 2-3, Qp approx 2-5, R approx 3-5 m, I approx 6-10 MA). The electrical and thermal power characteristics of transmutation reactors driven by fusion and accelerator spallation neutron sources are compared. The status of fusion development vis-à-vis a neutron source is reviewed.

  11. 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.

  12. 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.

  13. Calculations of neutron source at the KYIV research reactor for the boron neutron capture therapy aims

    International Nuclear Information System (INIS)

    Calculation results of an epithermal neutron source which can be created at the Kyiv Research Reactor (KRR) by means of placing of specially selected moderators, filters, collimators, and shielding into the 10-th horizontal experimental tube (so-called thermal column) are presented. The general Monte-Carlo radiation transport code MCNP4C [1], the Oak Ridge isotope generation code ORIGEN2 [2] and the NJOY99 [3] nuclear data processing system have been used for these calculations

  14. Flow blockage analysis for the advanced neutron source reactor

    International Nuclear Information System (INIS)

    The Advanced Neutron Source (ANS) reactor was designed to provide a research tool with capabilities beyond those of any existing reactors. One portion of its state-of-the-art design required high-speed fluid flow through narrow channels between the fuel plates in the core. Experience with previous reactors has shown that fuel plate damage can occur when debris becomes lodged at the entrance to these channels. Such debris disrupts the fluid flow to the plate surfaces and can prevent adequate cooling of the fuel. Preliminary ANS designs addressed this issue by providing an unheated entrance length for each fuel plate so that any flow disruption would recover, thus providing adequate heat removal from the downstream, heated portions of the fuel plates. As part of the safety analysis, the adequacy of this unheated entrance length was assessed using both analytical models and experimental measurements. The Flow Blockage Test Facility (FBTF) was designed and built to conduct experiments in an environment closely matching the ANS channel geometry. The FBTF permitted careful measurements of both heat transfer and hydraulic parameters. In addition to these experimental efforts, a thin, rectangular channel was modeled using the Fluent computational fluid dynamics computer code. The numerical results were compared with the experimental data to benchmark the hydrodynamics of the model. After this comparison, the model was extended to include those elements of the safety analysis that were difficult to measure experimentally. These elements included the high wall heat flux pattern and variable fluid properties. The results were used to determine the relationship between potential blockage sizes and the unheated entrance length required

  15. 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)

  16. Neutron and gamma ray streaming experiments at the fast neutron source reactor 'YAYOI'

    International Nuclear Information System (INIS)

    Neutron and gamma ray streaming experiments were performed in the ducts and cavities that were located in the heavy concrete shields of the fast neutron source reactor YAYOI of University of Tokyo. The configurations have the feature that the streaming through the ducts are occurred following the scattering in the cavity. The axes of the ducts are perpendicular to the source radiation from the core. The spectrum of the source was modified by putting a plug in the beam hole of the core. An aluminum plug and the plug which contains paraffin were used. The decay in the ducts, however, hardly depends on the source spectrum. The decay in the ducts is nearly exponential. (author)

  17. 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

  18. Consideration of LH2 and LD2 cold neutron sources in heavy water reactor reflector

    International Nuclear Information System (INIS)

    The reactor power, the required CNS dimensions and power of the cryogenic equipment define the CNS type with maximized cold neutron production. Cold neutron fluxes from liquid hydrogen (LH2) and liquid deuterium (LD2) cold neutron sources (CNS) are analyzed. Different CNS volumes, presents and absence of reentrant holes inside the CNS, different adjustment of beam tube and containment are considered. (orig.)

  19. The reference neutron field - a standard neutron source for neutron measurements at the research reactor IRT-2000 in Sofia

    International Nuclear Information System (INIS)

    A reference neutron field (RFN) is used as a standard neutron source (SNS) that is influenced by the changes in the reactor core due to recharging or other causes. A whole range of measurements is carried out in a full scope, to specify its characteristics precisely. The SNS comprises: 1) the RNF certificated to the neutron energy spectrum, its location in the reactor field, being a reference measure of the differential energy distribution in the neutron flux; 2) exposure monitoring tools (detectors revealing the certified physical characteristics); 3) functional measurement apparatus (revealing the spectral characteristics). The following basic metrological characteristics are given: differential neutron energy spectrum, described by F(E) [1/cm2.s.MeV], normalized by 1 in the range 3-19 MeV and the measurement error; the conventional neutron flux density and its error. The methodology of measuring the neutron flux integral density comprises the following six steps: 1) assessment of the influence of the changes in the core configuration on the stability of the RNF (estimated in six energy ranges); 2) demonstration of RNF application in reactor physics studies; 3) irradiation of two sets of activation detectors (Au, Sc and Au, Sc, S in Al and Cd shields); 4) measurement of the detector activities by calibrated gamma- and beta- spectrometric apparatus; 5) determination of the neutron field characteristics at a certain point of the RNF by the method of activating ratios; 6) the result accuracy assessment and probabilistic error limits determination with 95% upper bound frequency. The RNF neutron energy range have been measured 6 times for a period of two years. 6 refs., 8 figs. (M.A.)

  20. 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.

  1. Source driven breeding thermal power reactors using D-T fusion neutron sources

    International Nuclear Information System (INIS)

    Improvements in the performance of fission power reactors made possible by designing them subcritical driven by D-T neutron sources are investigated. Light-water thermal systems are found to be most promising, neutronically and energetically, for the source driven mode of operation. The range of performance characteristics expected from breeding Light Water Hybrid Reactors (LWHR) is defined. Several promising types of LWHR blankets are identified. Options opened for the nuclear energy strategy by four types of the LWHRs are examined, and the potential contribution of these LWHRs to the nuclear energy economy are discussed. The power systems based on these LWHRs are found to enable a high utilization of the energy content of the uranium resources in all forms available-including depleted uranium and spent fuel from LWRs, while being free from the need for uranium enrichment and plutonium separation capabilities. (orig.)

  2. Source neutron multiplication in subcritical reactors: Its dependency on core design

    International Nuclear Information System (INIS)

    The effectiveness of the external neutron source has been evaluated in terms of its multiplication in subcritical reactors. Mathematical formulas for the source neutron multiplication are reviewed and their implications are addressed. In order to identify dependency of the spallation neutron multiplication on the core design features, numerical experiments are performed with a Monte Carlo code for an accelerator-driven system, HYPER (HYbrid Power Extraction Reactor). For various core conditions, the actual source multiplication factors are compared with the conventional critical mode expectation. Sensitivity of the multiplication factor to the radial power distribution is investigated to find an optimum radial power distribution from the point of the spallation neutron multiplication. Also, impacts of neutron absorbers on the source multiplication are considered in this work. In addition, the efficiency of external spallation neutrons is evaluated for proton target designs. (author)

  3. 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)

  4. 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%

  5. 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)

  6. 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)

  7. 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.

  8. 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...

  9. Reactor physics analyses of the advanced neutron source three-element core

    International Nuclear Information System (INIS)

    A reactor physics analysis was performed for the Advanced Neutron Source reactor with a three-element core configuration. The analysis was performed with a two-dimensional r-z 20-energy-group finite-difference diffusion theory model of the 17-d fuel cycle. The model included equivalent r-z geometry representations of the central control rods, the irradiation and production targets, and reflector components. Calculated quantities include fuel cycle parameters, fuel element power distributions, unperturbed neutron fluxes in the reflector and target regions, reactivity perturbations, and neutron kinetics parameters

  10. Subcriticality Evaluation of AGN-201 Reactor Using Modified Neutron Source Multiplication Method

    International Nuclear Information System (INIS)

    One of the main issues in nuclear criticality safety is to measure subcriticality accurately at nuclear facility containing fissile materials. In order to verify the feasibility and safety of reactor, reactor physics test is performed in the commercial reactor. Among these test items, the measurement of control rod worth is taken most of period of reactor physics test. For that reason, the new methods have been introduced for subcriticality measurement to reduce the test period from the economic point of view : for example, pulse neutron method, neutron noise analysis method, Neutron Source Multiplication (NSM) method and so on. In 1980's, the research for subcriticality measurement methodology was performed about accelerator driven system, fast breeder reactor and critical experiment reactor. In this study, subcritcality is evaluated by modified NSM method. It is based on the conventional NSM method adding two correction processes: extraction of the fundamental mode from measuring neutron count rate data that contains not only fundamental mode but also higher modes in real situation and spatial corrections for perturbation induced by a reactivity addition in the distributions of the fundamental mode and a neutron importance field. In the previous studies, the verification of this method has been firstly performed for the subcriticality measurement of critical assembly of Kyoto University Critical Assembly (KUCA) at Kyoto University Research Reactor Institute in Japan. Recently subcriticality measurement study for the Pressurized Water Reactor (PWR) has been carried out. In the present study, the subcriticality was evaluated for Aerojet General Nucleonics (AGN)-201 reactor by the modified NSM method with two correction processes. The AGN-201 reactor is the graphite moderated homogeneous type research reactor and is used for reactor experiments such as critical mass approach, control rod calibration, measurement of neutron flux and so on. For subcriticality

  11. The design of the cold neutron source of the OPAL reactor

    International Nuclear Information System (INIS)

    The present work describes the conceptual design process of the first cold neutron source developed by INVAP for the nuclear research reactor OPAL. The analysis begins from the requirements given by the client and continues with the chosen solutions. Furthermore, we studied how impact in the design the fully illuminated constraint with the finite remote source model. (author)

  12. Pulsed TRIGA reactor as substitute for long pulse spallation neutron source

    International Nuclear Information System (INIS)

    TRIGA reactor cores have been used to demonstrate various pulsing applications. The TRIGA reactor fuel (U-ZrHx) is very robust especially in pulsing applications. The features required to produce 50 pulses per second have been successfully demonstrated individually, including pulse tests with small diameter fuel rods. A partially optimized core has been evaluated for pulses at 50 Hz with peak pulsed power up to 100 MW and an average power up to 10 MW. Depending on the design, the full width at half power of the individual pulses can range between 2000 μsec to 3000 μsec. Until recently, the relatively long pulses (2000 μsec to 3000 μsec) from a pulsed thermal reactor or a long pulse spallation source (LPSS) have been considered unsuitable for time-of-flight measurements of neutron scattering. More recently considerable attention has been devoted to evaluating the performance of long pulse (1000 to 4000 μs) spallation sources for the same type of neutron measurements originally performed only with short pulses from spallation sources (SPSS). Adequate information is available to permit meaningful comparisons between CW, SPSS, and LPSS neutron sources. Except where extremely high resolution is required (fraction of a percent), which does require short pulses, it is demonstrated that the LPSS source with a 1000 msec or longer pulse length and a repetition rate of 50 to 60 Hz gives results comparable to those from the 60 MW ILL (CW) source. For many of these applications the shorter pulse is not necessarily a disadvantage, but it is not an advantage over the long pulse system. In one study, the conclusion is that a 5 MW 2000 μsec LPSS source improves the capability for structural biology studies of macromolecules by at least a factor of 5 over that achievable with a high flux reactor. Recent studies have identified the advantages and usefulness of long pulse neutron sources. It is evident that the multiple pulse TRIGA reactor can produce pulses comparable to those

  13. 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.

  14. The TAPIRO fast-neutron source reactor as a support to nuclear data assessment

    International Nuclear Information System (INIS)

    TAPIRO is a fast neutron source reactor operating at CASACCIA Research Center since 1971. The project, entirely developed by ENEA's staff, is based on the general concept of AFSR (Argonne Fast Source Reactor - Idaho Falls). The reactor is equipped with a homogeneous cylindrical core having 6.29 cm as radius and 10.87 cm as height; cladding is provided by stainless steel (0.5 mm thickness) placed on a cylindrical copper reflector having (30 cm as thickness). All components assembled in a stainless steel tank, are placed inside a near spherical borated concrete shielding system having 1.75 m as thickness. Channels of various dimension and with different neutron spectra are distributed around the core. A large thermal column is manufactured by graphite blocks, suitable to be removed and replaced with experimental assemblies for any research purpose. The TAPIRO possibilities for reactor experiments with energies up to 1.35 MeV will be illustrated. (author)

  15. New approach to handle neutron startup sources in a high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    In a high temperature gas-cooled reactor, a neutron startup source (NS) cannot be handled simply. An appropriate transfer vessel, connected with a 241Am-Be source for startup core physics tests through an NS coupler and drive wire, was newly designed and installed in the high temperature engineering test reactor (HTTR). A result of tests using the HTTR revealed that the NS could be loaded simply and certainly into the reactor core through the transfer vessel below the allowable limit of effective dose equivalent for operators

  16. Several dosimetry studies in the fast neutron source reactor ''YAYOI''

    International Nuclear Information System (INIS)

    Neutron flux and spectra in the YAYOI core-center field have been well characterized through the detector intercomparison studies. Spectra are summarized in an efficient parameter representation form which is useful for neutron dose evaluation. An intermediate neutron field with near 1/E spectrum shape has been developed in the center of the octagonal lead pile driven by YAYOI core. Calculated neutron flux and spectra by two dimensional transport code have been in good agreement with the experimental results within +- 10% above 0.2 MeV and +- 20% below 0.2 MeV. Its usefulness as a standard neutron field is discussed. Absorbed dose measurements have been also carried out at the core-center field with three different techniques; calorimeter, Fricke-dosimeter and foil detector set used with a thermo-luminescent dosimeter, and a good agreement has been obtained especially about the neutron dose value

  17. 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

  18. Power burst reactor facility as an epithermal neutron source for brain cancer therapy

    International Nuclear Information System (INIS)

    The Power Burst Facility (PBF) reactor is considered for modification to provide an intense, clean source of intermediate-energy (epithermal) neutrons desirable for clinical studies of neutron capture therapy (NCT) for malignant tumors. The modifications include partial replacement of the reflector, installation of a neutron-moderating, shifting region, addition shielding, and penetration of the present concrete shield with a collimating and (optionally) filtering region. The studies have indicated that the reactor, after these modifications, will be safely operable at full power (28 MW) within the acceptable limits of the plant protection systems. The neutron beam existing from the collimator port is predicted to be of sufficient intensity (∼ 1010) neutrons/cm2-s) to provide therapeutic doses in very short irradiation times. The beam would be relatively free of undesirable fast neutrons, thermal neutrons and gamma rays. The calculated neutron energy spectrum and associated gamma rays in the beam were provided as input in simulation studies that used a computer model of a patient with a brain tumor to determine predicted dose rates to the tumor and healthy tissue. The results of this conceptual study indicate an intense, clean beam of epithermal neutrons for NCT clinical trials is attainable in the PBF facility with properly engineered design modifications. 9 references, 11 figures, 3 tables

  19. 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)

  20. 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

  1. 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

  2. Analytical solution of neutron transport equation in an annular reactor with a rotating pulsed source

    International Nuclear Information System (INIS)

    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) δ (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) δ (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)

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

    International Nuclear Information System (INIS)

    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 4.12×109 and 1.21×109 n/cm2/s, respectively, which can provide an appropriate neutron beam for BNCT by designing and constructing a proper Beam Shaping Assembly (BSA) structure. - Highlights: • The feasibility of using of TRR for BNCT has been investigated. • Neutron energy spectrum at all external irradiation facilities of TRR have been measured and calculated. • Spatial distribution of neutrons have been measured using copper wire activation method

  4. 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

  5. Neutron sources

    International Nuclear Information System (INIS)

    As neutron scattering experiments have grown more and more demanding with respect to resolution and quality, it became more and more necessary to include the neutron source itself in the design of an experimental setup. In this sense the generic representation of a neutron scattering arrangement includes the primary neutron source and the associated spectrum shifter (or moderator). In fact, the design of a modern neutron source will start from a set of users requirements and will proceed 'inwards' through a selection of the moderators (spectrum shifters) to the primary source best suited to meet these often conflicting needs. This paper aims at explaining the options source designers have to match the neutron source performance to the users' demands. (author)

  6. Single crystal diffractometers for bio-macromolecules using neutrons at steady-state reactor sources

    International Nuclear Information System (INIS)

    In order to overcome low flux of neutron sources as well as weak diffraction intensity from bio-macromolecule crystals, several devices have been developed in neutron biological crystallography. Elastically bent Si monochromator has contributed the increase of incident beam intensity, and neutron imaging plate (NIP) has provided large detecting area. In particular, the successful development of the NIP made a breakthrough in this research field. Additionally, recent advances in techniques for cryogenic temperature measurement, growth of large crystal and sample deuteration have made a contribution to efficient measurement performance. Currently, a total of six diffractometers for bio-macromolecule are available at research reactors in the world. Neutron crystallography is on the verge of becoming a prevalent method for structural study on bio-macromolecules. (author)

  7. 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

  8. 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

  9. 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

  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. Performance of the advanced cold neutron source and optics upgrades at the NIST Research Reactor

    International Nuclear Information System (INIS)

    On March 6, 2002, the NIST Research Reactor resumed routine operation following a six-month shutdown for facility upgrades and maintenance. During the shutdown, the original liquid hydrogen cold neutron source was removed, and the advanced cold source was installed. An optical filter was installed on one of the neutron guides, NG-3, replacing a crystal filter for the 30-m SANS instrument and the guide used between the chopper disks of the Disk Chopper time-of-flight Spectrometer (DCS) installed on NG-4 has been recently reconfigured. Additional improvements in the neutron optics of various instruments are being made. The advanced liquid hydrogen cold neutron source performs as expected, nearly doubling the flux available to most instruments. The measured gains range from about 1.4 at 2 A, to over a factor of two at 15 A. Also as expected, the heat load in the new source increased to 1200 watts, but the previously existing refrigerator has easily accommodated the increase. With intensity gains of a factor of two in the important long wavelength region of the spectrum, the advanced cold source significantly enhances the measurement capability of the cold neutron scattering instrumentation at NIST. The optical filter on NG-3 is also very successful; the 30-m SANS has an additional gain of two at 17 A. A system of refracting lenses and prisms near the SANS sample position has made possible measurements at low Q (0.0005 A-1) that were previously not feasible. The DCS has also seen additional intensity gain factors in excess of two for the majority of experiments and at short neutron wavelengths the gains exceed three. In addition, two new triple axis spectrometers will feature double-focusing monochromators in order to exploit the full size of the available thermal and cold neutron beam tubes. The success of the advanced cold source and enhanced neutron optics contributed to the recognition of the NIST Center for Neutron Research as 'the premiere neutron scattering

  12. Development of an asymmetric multiple-position neutron source (AMPNS) method to monitor the criticality of a degraded reactor core

    International Nuclear Information System (INIS)

    An analytical/experimental method has been developed to monitor the subcritical reactivity and unfold the k/sub infinity/ distribution of a degraded reactor core. The method uses several fixed neutron detectors and a Cf-252 neutron source placed sequentially in multiple positions in the core. Therefore, it is called the Asymmetric Multiple Position Neutron Source (AMPNS) method. The AMPNS method employs nucleonic codes to analyze the neutron multiplication of a Cf-252 neutron source. An optimization program, GPM, is utilized to unfold the k/sub infinity/ distribution of the degraded core, in which the desired performance measure minimizes the error between the calculated and the measured count rates of the degraded reactor core. The analytical/experimental approach is validated by performing experiments using the Penn State Breazeale TRIGA Reactor (PSBR). A significant result of this study is that it provides a method to monitor the criticality of a damaged core during the recovery period

  13. Advanced Neutron Source Cross Section Libraries (ANSL-V): ENDF/B-V based multigroup cross-section libraries for advanced neutron source (ANS) reactor studies

    International Nuclear Information System (INIS)

    Pseudo-problem-independent, multigroup cross-section libraries were generated to support Advanced Neutron Source (ANS) Reactor design studies. The ANS is a proposed reactor which would be fueled with highly enriched uranium and cooled with heavy water. The libraries, designated ANSL-V (Advanced Neutron Source Cross Section Libraries based on ENDF/B-V), are data bases in AMPX master format for subsequent generation of problem-dependent cross-sections for use with codes such as KENO, ANISN, XSDRNPM, VENTURE, DOT, DORT, TORT, and MORSE. Included in ANSL-V are 99-group and 39-group neutron, 39-neutron-group 44-gamma-ray-group secondary gamma-ray production (SGRP), 44-group gamma-ray interaction (GRI), and coupled, 39-neutron group 44-gamma-ray group (CNG) cross-section libraries. The neutron and SGRP libraries were generated primarily from ENDF/B-V data; the GRI library was generated from DLC-99/HUGO data, which is recognized as the ENDF/B-V photon interaction data. Modules from the AMPX and NJOY systems were used to process the multigroup data. Validity of selected data from the fine- and broad-group neutron libraries was satisfactorily tested in performance parameter calculations

  14. The first main steps for development of BNCT neutron sources at the Ukrainian and Uzbek Research Reactors

    International Nuclear Information System (INIS)

    Both in Ukraine and in Uzbekistan, epithermal neutron irradiation facilities for Boron Neutron Capture Therapy (BNCT) are under consideration, as the need for them is very large. Based on information from medical cancer treatment institutions of the total number of patients identified with cancer, about 5000 have brain tumours. The most prospective method of their treatment is BNCT. Both in Ukraine and in Uzbekistan, this method can be implemented on existing research reactors. Modification of research reactors may be a relatively straightforward and inexpensive way to develop a BNCT neutron source, especially in comparison with construction of new reactors specialized for BNCT. However, prior to any reactor modification, careful calculations need to be performed, which take into account all the peculiarities of the specific reactor system. Based on the world experience in epithermal neutron beam development, it is very clear that the research reactors in Kyiv (Kyiv Research Reactor-KRR) and Tashkent (Tashkent Research Reactor-TRR) may be reconstructed into epithermal irradiation facilities. Selection of the most suitable materials for moderator, collimator, shielding, etc., demands carrying out calculations considering their individual characteristics. Since the KRR and TRR are the same kind of research reactors, with for example similar thermal columns, the development of a BNCT neutron source at these research reactors may be achieved in a like manner. The development plan and the first experience in this direction (using preliminary MCNP calculation results) are presented here. (author)

  15. First results on testing of the ultra-cold neutrons pulsed source at BIGR reactor (VNIIEF, Sarov)

    International Nuclear Information System (INIS)

    The high-intensity ultracold neutron source constructed on the basis of the aperiodic BIGR reactor (All-Russian Institute of Experimental Physics, Sarov, Nizhnij Novgorod region) is described. The neutron density equal to 25 ± 5 n x cm-3 is obtained to the moment of the reactor pulse. The further development of the installation requires the modernization of the source that will allow to increase the ultracold neutron density by 2-3 orders as expected. The source modernization should include the optimization of the moderator block, namely, the thickness reduction of the forward wall which should increase the thermal neutron flux density in the moderator channel; the use of the polyethylene converter cooled-down to 80 K and the use of the reactor pulses with maximal parameters, namely, the energy release of 250 MJ with the pulse duration of the order of 2 ms. Realization of the ideas discussed opens opportunities for the essentially new experiment dealing with the neutron lifetime measurement which consists in the simultaneous registration of the β decay neutron products (electrons and protons) and the heated-up neutrons (ultracold neutrons scattered inelastically on the storage volume walls that results in the neutron get-away from the trap). The numerical modeling of this experiment shows that the accuracy of about seconds per reactor pulse may be achieved

  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. 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.

  19. 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.

  20. 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

  1. 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

  2. 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.

  3. Small-angle scattering at a pulsed neutron source: comparison with a steady-state reactor

    International Nuclear Information System (INIS)

    A time-of-flight small-angle diffractometer employing seven tapered collimator elements and a two-dimensional gas proportional counter was successfully utilized to collect small-angle scattering data from a solution sample of the lipid salt cetylpyridinium chloride, C21H38N+.Cl-, at the Argonne National Laboratory prototype pulsed spallation neutron source, ZING-P'. Comparison of the small-angle scattering observed from the same compound at the University of Missouri Research Reactor corroborated the ZING-P' results. The results are used to compare the neutron flux available from the ZING-P' source relative to the well characterized University of Missouri source. Calculations based on experimentally determined parameters indicated the time-averaged rate of detected neutrons at the ZING-P' pulsed spallation source to have been at least 33% higher than the steady-state count rate from the same sample. Differences between time-of-flight techniques and conventional steady-state techniques are discussed. (Auth.)

  4. 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.

  5. 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.

  6. 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.

  7. 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

  8. 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.

  9. 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

  10. 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)

  11. 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.

  12. 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

  13. Experimental investigation of thermal limits in parallel plate configuration for the Advanced Neutron Source Reactor

    International Nuclear Information System (INIS)

    The Advanced Neutron Source Reactor (ANSR) is currently being designed to become the world's highest-flux, steady-state, thermal neutron source for scientific experiments. Highly subcooled, heavy-water coolant flows vertically upward at a very high velocity of 25 m/s through parallel aluminum fuel-plates. The core has average and peak heat fluxes of 5.9 and 12 MW/m2, respectively. In this configuration, both flow excursion (FE) and true critical heat flux (CHF), represent potential thermal limitations. The availability of experimental data for both FE and true CHF at the conditions applicable to the ANSR is very limited. A Thermal Hydraulic Test Loop (THTL) facility was designed and built to simulate a full-length coolant subchannel of the core, allowing experimental determination of both thermal limits under the expected ANSR T/H conditions. A series of FE tests with water flowing vertically upward was completed over a nominal heat flux range of 6 to 14 MW/m2 and a corresponding velocity range of 8 to 21 m/s. Both the exit pressure (1.7 MPa) and inlet temperature (45 degrees C) were maintained constant for these tests, while the loop was operated in a ''stiff''(constant flow) mode. Limited experiments were also conducted at 12 MW/m2 using a ''soft'' mode (near constant pressure-drop) for actual FE burnout tests and using a ''stiff' mode for true CHF tests, to compare with the original FE experiments

  14. 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

  15. 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

  16. Study on recriticality of fuel debris during hypothetical severe accidents in the Advanced Neutron Source reactor

    International Nuclear Information System (INIS)

    A study has been performed to measure the potential of recriticality during hypothetical severe accident in Advanced Neutron Source (ANS). For the lumped debris configuration in the Reactor Coolant System (RCS), as found in the previous study, recriticality potential may be very low. However, if fuel debris is dispersed and mixed with heavy water in RCS, recriticality potential has been predicted to be substantial depending on thermal-hydraulic conditions surrounding fuel debris mixture. The recriticality potential in RCS is substantially reduced for the three element core design with 50% enrichment. Also, as observed in the previous study, strong dependencies of keff on key thermal hydraulic parameters are shown. Light water contamination is shown to provide a positive reactivity, and void formation due to boiling of mixed water provides enough negative reactivity and to bring the system down to subcritical. For criticality potential in the subpile room, the lumped debris configuration does not pose a concern. Dispersed configuration in light water pool of the subpile room is also unlikely to result in criticality. However, if the debris is dispersed in the pool that is mixed with heavy water, the results indicate that a substantial potential exists for the debris to reach the criticality. However, if prompt recriticality disperses the debris completely in the subpile room pool, subsequent recriticality may be prevented since neutron leakage effects become large enough

  17. 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.)

  18. 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.

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

    International Nuclear Information System (INIS)

    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 KQ 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

  20. 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

  1. 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.

  2. Home brew technetium : clinical scale desktop plasma fusion neutron source to produce Tc99m as an alternative to industrial scale fission reactor sources

    International Nuclear Information System (INIS)

    Full text: Tc-99m (decay product of Mo-99) accounts for ∼ 90% of world's production of radiopharmaceuticals. Recent unexpected shutdowns of two fission reactors and routine maintenance closures .e created a global shortage of Tc-99m, hence the large global effort to find alternative sources. This project aims to design and produce a novel prototype Mo-99/Tc-99m source. An operational desktop neutron source is available at the University of Sydney, employing a deuterium fusion-plasma to create 2.45 MeV neutrons. These neutrons will be used to activate Mo-98 thin an activation vessel. In one embodiment, the activation vessel contains an aqueous slurry or gel containing Mo-98 which converts to 0-99 upon activation. The decay product Tc-99m could then be milked, similar to existing Tc-99m generators. Monte Carlo will be :ed to assess yield versus size and geometry for various vessel designs. The neutron source filled with deuterium operating at 250 W, produces 3 x 106 neutrons continuously. The neutron flux can be increased ∼ 100-fold if the fill gas is 50% tritium and by another ∼ 100-1000-fold by increasing the power. This is being designed for local use, perhaps on the scale f one or a few hospitals, so the yield would not need to be industrial ;ale as with fission reactor sources. This device is low cost <$300 K) compared with cyclotrons and fission reactors.

  3. Application of the modified neutron source multiplication method for a measurement of sub-criticality in AGN-201K reactor

    International Nuclear Information System (INIS)

    Measurement of sub-criticality is a challenging and required task in nuclear industry both for nuclear criticality safety and physics test in nuclear power plant. A relatively new method named as Modified Neutron Source Multiplication Method (MNSM) was proposed in Japan. This method is an improvement of traditional Neutron Source Multiplication (NSM) Method, in which three correction factors are applied additionally. In this study, MNSM was tested in calculation of rod worth using an educational reactor in Kyung Hee University, AGN-201K. For this study, a revised nuclear data library and a neutron transport code system TRANSX-PARTISN were used for the calculation of correction factors for various control rod positions and source locations. Experiments were designed and performed to enhance errors in NSM from the location effects of source and detectors. MNSM can correct these effects but current results showed not much correction effects. (author)

  4. Modeling ampersand analysis of criticality-induced severe accidents during refueling for the Advanced Neutron Source Reactor

    International Nuclear Information System (INIS)

    This paper describes work done at the Oak Ridge National Laboratory (ORNL) for evaluating the potential and resulting consequences of a hypothetical criticality accident during refueling of the 330-MW Advanced Neutron Source (ANS) research reactor. The development of an analytical capability is described. Modeling and problem formulation were conducted using concepts of reactor neutronic theory for determining power level escalation, coupled with ORIGEN and MELCOR code simulations for radionuclide buildup and containment transport Gaussian plume transport modeling was done for determining off-site radiological consequences. Nuances associated with modeling this blast-type scenario are described. Analysis results for ANS containment response under a variety of postulated scenarios and containment failure modes are presented. It is demonstrated that individuals at the reactor site boundary will not receive doses beyond regulatory limits for any of the containment configurations studied

  5. Modeling Advanced Neutron Source reactor station blackout accident using RELAP5

    International Nuclear Information System (INIS)

    The Advanced Neutron Source (ANS) system model using RELAP5 has been developed to perform loss-of-coolant accident (LOCA) and non-LOCA transients as safety-related input for early design considerations. The transients studies include LOCA, station blackout, and reactivity insertion accidents. The small-, medium-, and large-break LOCA results were presented and documented. This paper will focus on the station blackout scenario. The station blackout analyses have concentrated on thermal-hydraulic system response with and without accumulators. Five transient calculations were performed to characterize system performance using various numbers and sizes of accumulators at several key sites. The main findings will be discussed with recommendations for conceptual design considerations. ANS is a state-of-the-art research reactor to be built and operated at high heat flux, high mass flux, and high coolant subcooling. To accommodate these features, three ANS-specific changes were made in the RELAP5 code by adding: the Petukhov heat transfer correlation for single-phase forced convection in the thin coolant channel; the Gambill additive method with the Weatherhead wall superheat for the critical heat flux; and the Griffith drift flux model for the interfacial drag in the slug flow regime. 7 refs., 6 figs., 1 tab

  6. 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.)

  7. 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%....

  8. 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.

  9. 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

  10. Selected thermal and hydraulic experimentation in support of the advanced neutron source reactor

    International Nuclear Information System (INIS)

    The ANS Reactor has unique thermal-hydraulic characteristics in comparison to other research and commercial reactors: Heavy water coolant, Parallel Rectangular channels (involute), Very small channel gap (1.27 mm), Very high velocity (25 m/s), Very high exit subcooling, Moderately high heat flux, High average power density. The objective was to determine experimentally the appropriate core thermal hydraulic limits at ANS conditions. Advanced Neutron Source (ANS) Thermal Hydraulic Test Loop (THTL) was designed to operate in 'Stiff', 'Soft' and 'Modified Stiff' Modes.Summary of Thermal Hydraulic Limit Testing and Analysis shows: FE data has been acquired at ANS typical flow velocities; An extensive OSV/OFI data base has been developed with a very broad parameter range, A modification of the Saha-Zuber correlation was proposed to account for reduced subcooling effects; Closeout activities include continued investigation of wider span test channels; Some testing for HFIR will be performed to evaluate the effect of reduced channel gap; Future plans called for additional testing at 3-core conditions, hot spot testing, etc. The Objective of Fuel Plate Stability Testing was to experimentally evaluate the structural response of ANS fuel plates to hydraulic loads. Summary of Fuel Plate Stability Testing shows: A Method Has Been Developed to Predict Structural Response of Fuel Plates to Hydraulic Loading Prediction of AP across plates Determine deflection/stress levels using structural analysis; ANS, Specific Conclusions are: no evidence of potential plate collapse in the coolant velocity range from 050 m/s, no evidence of plate flutter with coolant velocities below 33 m/s, local stress levels appear to dictate plate limits as opposed to plate deflection. The objective of Flow Blockage Testing was to experimentally determine local thermal and fluid. Summary of Flow Blockage Testing and Analysis showed: CFD code has been benchmarked against prototypic ANS flow conditions and

  11. 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...

  12. 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

  13. 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.

  14. Severe accident risk minimization studies for the Advanced Neutron Source (ANS) reactor plant at the Oak Ridge National Laboratory

    International Nuclear Information System (INIS)

    This paper discusses salient aspects of severe accident related phenomenological considerations, scoping studies, and mitigative design features being studied for incorporation into a high-power research reactor plant. Key results of scoping studies on steam explosions, recriticality, core-concrete interactions, and containment transport are highlighted. Evolving design features of the containment are described. Containment response calculations for a site-suitability basis transient are presented that demonstrate acceptable source term values and superior containment performance. Oak Ridge National Laboratory's (ORNL) Advanced Neutron Source (ANS) will be a new user facility for all kinds of neutron research, centered around a research reactor of unprecedented neutron beam flux. A defense-in-depth philosophy has been adopted. In response to this commitment, ANS Project management initiated severe accident analysis and related technology development early-on in the design phase itself. This was done to aid in designing a sufficiently robust containment for retention and controlled release of radionuclides in the event of such an accident. It also provides a means for satisfying on- and off-site regulatory requirements, accident-related dose exposures, and containment response and source-term best-estimate analyses for level-2 and -3 Probabilistic Risk Analysis (PRAs) that will be produced. Moreover, it will provide the best possible understanding of the ANS under severe accident conditions and consequently provide insights for the development of strategies and design philosophies for accident mitigation, management, and emergency preparedness efforts

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

    Energy Technology Data Exchange (ETDEWEB)

    Cook, J.C.; Barker, J.G.; Rowe, J.M.; Williams, R.E. [NIST Center for Neutron Research, National Institute of Standards and Technology, 100 Bureau Drive, Mail Stop 6100, Gaithersburg, MD 20899-6100 (United States); Gagnon, C. [Department of Materials Science and Engineering, University of Maryland, College Park, MD 20742 (United States); Lindstrom, R.M. [Scientist Emeritus, Chemical Sciences Division, National Institute of Standards and Technology, 100 Bureau Drive, Mail Stop 8395, Gaithersburg, MD 20899-8395 (United States); Ibberson, R.M.; Neumann, D.A. [NIST Center for Neutron Research, National Institute of Standards and Technology, 100 Bureau Drive, Mail Stop 6100, Gaithersburg, MD 20899-6100 (United States)

    2015-08-21

    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.

  16. 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.

  17. 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.

  18. Summary of dynamic analyses of the advanced neutron source reactor inner control rods

    International Nuclear Information System (INIS)

    A summary of the structural dynamic analyses that were instrumental in providing design guidance to the Advanced Neutron source (ANS) inner control element system is presented in this report. The structural analyses and the functional constraints that required certain performance parameters were combined to shape and guide the design effort toward a prediction of successful and reliable control and scram operation to be provided by these inner control rods

  19. 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

  20. 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

  1. 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.)

  2. Modeling ampersand analysis of core debris recriticality during hypothetical severe accidents in the Advanced Neutron Source Reactor

    International Nuclear Information System (INIS)

    This paper discusses salient aspects of severe-accident-related recriticality modeling and analysis in the Advanced Neutron Source (ANS) reactor. The development of an analytical capability using the KEN05A-SCALE system is described including evaluation of suitable nuclear cross-section sets to account for the effects of system geometry, mixture temperature, material dispersion and other thermal-hydraulic conditions. Benchmarking and validation efforts conducted with KEN05-SCALE and other neutronic codes against critical experiment data are described. Potential deviations and biases resulting from use of the 16-group Hansen-Roach library are shown. A comprehensive test matrix of calculations to evaluate the threat of a criticality event in the ANS is described. Strong dependencies on geometry, material constituents, and thermal-hydraulic conditions are described. The introduction of designed mitigative features are described

  3. 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.

  4. Fission-Fusion Neutron Source

    International Nuclear Information System (INIS)

    Full text of publication follows: In order to meet the requirement of fusion reactor developing and nuclear waste treatment, a concept of fission-fusion neutron source has been proposed with LiD cylinder in heavy water region of China Advanced Research Reactor (CARR) by slow neutrons to transfer to fusion neutron. The principal is the reaction of 6Li(n,α) to produce energetic tritium ion with 2.739 MeV in LiD by slow neutron, which will be bombarding the deuteron of LiD to induce fusion reaction to produce 14 MeV neutron. The fusion reaction rate will increase with the accumulation of tritium in LiD by the reaction between tritium and deuteron recoils produced by 14 MeV neutrons. When the concentration of tritium in LiD reaches O.5 x 1022 T/cm3 and the fraction of fusion reaction induced by deuteron recoils with tritium approaches to 1, the 14 MeV neutron flux will be doubled and redoubled increasing to approach saturation in which the produced tritium at time t is exhausted by fusion reaction to keep the constant of tritium concentration in LiD. At this case the 14 MeV neutron production rate is too high, it has to decrease the slow neutron flux with decreasing CARR reactor power progressively when the fusion neutron flux approaches to presetting value, for example 3.5 x 1014 n/cm2 sec and will approach to saturation at the low level of neutron flux. This paper describes the principle of fission-fusion neutron source, including the production rate of fusion neutron, the accumulation rate and concentration of tritium, the fusion reaction rate induced by deuteron recoils with tritium, the 14 MeV neutron flux of inner surface of LiD cylinder in the heavy water region of CARR reactor without neutron depression and the influence factors. To consider the neutron depression an assembly of LiD rods in 20 x 20 cm with a centre hole in CARR reactor must be designed to optimize the fusion neutron flux in centre hole. (author)

  5. 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)

  6. Investigation of primary cooling water chemistry following the partial meltdown of Pu-Be neutron source in Tehran Research Reactor Core (TRR)

    International Nuclear Information System (INIS)

    Research highlights: → Effect of Pu-Be neutron source meltdown in core on reactor water chemistry. → Water chemistry of primary cooling before, during and after of above incident was compared. → Training importance. → Management of nuclear incident and accident. - Abstract: Effect of Pu-Be neutron source meltdown in core on reactor water chemistry was main aim of this study. Leaving the neutron source in the core after reactor power exceeds a few hundred Watts was the main reason for its partial meltdown. Water chemistry of primary cooling before, during and after of above incident was compared. Activity of some radio-nuclides such as Ba-140, La-140, I-131, I-132, Te-132 and Xe-135 increased. Other radio-nuclides such as Nd-147, Xe-133, Sr-91, I-133 and I-135 are also detected which were not existed before this incident.

  7. Reactor physics methods, models, and applications used to support the conceptual design of the Advanced Neutron Source

    Energy Technology Data Exchange (ETDEWEB)

    Gehin, J.C.; Worley, B.A.; Renier, J.P. [Oak Ridge National Lab., TN (United States); Wemple, C.A.; Jahshan, S.N.; Ryskammp, J.M. [Idaho National Engineering Lab., Idaho Falls, ID (United States)

    1995-08-01

    This report summarizes the neutronics analysis performed during 1991 and 1992 in support of characterization of the conceptual design of the Advanced Neutron Source (ANS). The methods used in the analysis, parametric studies, and key results supporting the design and safety evaluations of the conceptual design are presented. The analysis approach used during the conceptual design phase followed the same approach used in early ANS evaluations: (1) a strong reliance on Monte Carlo theory for beginning-of-cycle reactor performance calculations and (2) a reliance on few-group diffusion theory for reactor fuel cycle analysis and for evaluation of reactor performance at specific time steps over the fuel cycle. The Monte Carlo analysis was carried out using the MCNP continuous-energy code, and the few- group diffusion theory calculations were performed using the VENTURE and PDQ code systems. The MCNP code was used primarily for its capability to model the reflector components in realistic geometries as well as the inherent circumvention of cross-section processing requirements and use of energy-collapsed cross sections. The MCNP code was used for evaluations of reflector component reactivity effects and of heat loads in these components. The code was also used as a benchmark comparison against the diffusion-theory estimates of key reactor parameters such as region fluxes, control rod worths, reactivity coefficients, and material worths. The VENTURE and PDQ codes were used to provide independent evaluations of burnup effects, power distributions, and small perturbation worths. The performance and safety calculations performed over the subject time period are summarized, and key results are provided. The key results include flux and power distributions over the fuel cycle, silicon production rates, fuel burnup rates, component reactivities, control rod worths, component heat loads, shutdown reactivity margins, reactivity coefficients, and isotope production rates.

  8. 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)

  9. 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)

  10. 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)

  11. 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

  12. Modeling and analysis of hydrogen detonation events in the Advanced Neutron Source reactor containment

    International Nuclear Information System (INIS)

    This paper describes salient aspects of the modeling, analyses, and evaluations for hydrogen detonation in selected regions of the Advanced Neutron Source (ANS) containment during hypothetical severe accident conditions. Shock wave generation and transport modeling and analyses were conducted for two stratified configurations in the dome region of the high bay. Principal tools utilized for these purposes were the CTH and CET89 computer codes. Dynamic pressure loading functions were generated for key locations and used for evaluating structural response behavior for which a finite-element model was developed using the ANSYS code. For the range of conditions analyzed in the two critical dome regions, it was revealed that the ANS containment would be able to withstand detonation loads without failure

  13. Modeling and analysis of hydrogen detonation events in the advanced neutron source reactor containment

    International Nuclear Information System (INIS)

    This paper describes salient aspects of the modeling, analyses, and evaluations for hydrogen detonation in selected regions of the Advanced Neutron Source (ANS) containment during hypothetical severe accident conditions. Shock wave generation and transport modeling and analyses were conducted for two stratified configurations in the dome region of the high bay. Principal tools utilized for these purposes were the CTH and CET89 computer codes. Dynamic pressure loading functions were generated for key locations and used for evaluating structural response behavior for which a finite-element model was developed using the ANSYS code. For the range of conditions analyzed in the two critical dome regions, it was revealed that the ANS containment would be able to withstand detonation loads without failure. (author)

  14. Characteristics of fast neutron sources

    International Nuclear Information System (INIS)

    The contributions of a poster session from a clinical radiotherapy conference are reviewed and discussed with respect to economic aspects. The contributions were concerned with the optimum neutron treatment source for neutron therapy. The neutron sources considered were D-T generators with either metal hydride or gaseous targets, cyclotrons, nuclear reactors, proton linear accelerators and a pion facility. All facilities would appear to cost more than cobalt units or 4-6 MeV electron accelerators. From the radiobiological studies to date, there is little data to support the selection of one energy cyclotron over another. It is concluded that no neutron source will achieve the desirable physics characteristics of 4-6 MeV electrons and only the more expensive sources will achieve a depth dose similar to a cobalt unit. (UK)

  15. Sources of ultracold neutrons

    International Nuclear Information System (INIS)

    The results of comparative experimental investigations to study ultracold neutron yields from different neutron moderator-converters are presented. The installation is described which is based on a WWR-K reactor once-through beam hole. The neutron yields were measured using Al, Mg, ZrHsub(1.9), H2O and H2 neutron converters at 80 and 300 K. For H2 converters pressure dependences of the neutron yield were also measured in the 0.1-1.5 atm. pressure range. Among solid neutron converters the ZrHsub(1.9) one possesses the highest ultracold neutron yield, whereas among all the converters tested the best performance was shown by the frozen water one, the ultracold neutron count with the proportional He3 counter being about 500ssup(-1)

  16. 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...

  17. Review of the scientific results obtained at the research reactor-booster IBR-30 and the Program of investigations at the neutron source IREN

    International Nuclear Information System (INIS)

    Brief review of the main scientific results obtained at research reactor booster IBR-30 and its predecessor IBR and IBR-1 for the period 1960 - 2001 is presented. The thesis of the scientific program for the upgrade of IBR-30 resonance neutron source IREN are adduced

  18. Review of the scientific results obtained at the research reactor-booster IBR-30 and the Program of investigations at the neutron source IREN

    CERN Document Server

    Furman, W

    2002-01-01

    Brief review of the main scientific results obtained at research reactor booster IBR-30 and its predecessor IBR and IBR-1 for the period 1960 - 2001 is presented. The thesis of the scientific program for the upgrade of IBR-30 resonance neutron source IREN are adduced

  19. Cryogenic refrigeration for cold neutron sources

    International Nuclear Information System (INIS)

    Neutron moderation by means of a fluid at cryogenic temperature is a very interesting way to obtain cold neutrons. Today, a number of nuclear research reactors are using this technology. This paper deals with thermodynamics and technology which are used for cooling Cold Neutron Sources

  20. Conceptual design of thorium-fuelled Mitrailleuse accelerator-driven subcritical reactor using D-Be neutron source

    International Nuclear Information System (INIS)

    A distributed accelerator is a charged-particle accelerator that uses a new acceleration method based on repeated electrostatic acceleration. This method offers outstanding benefits not possible with the conventional radio-frequency acceleration method, including: (1) high acceleration efficiency, (2) large acceleration current, and (3) lower failure rate made possible by a fully solid-state acceleration field generation circuit. A 'Mitrailleuse Accelerator' is a product we have conceived to optimize this distributed accelerator technology for use with a high-strength neutron source. We have completed the conceptual design of a Mitrailleuse Accelerator and of a thorium-fuelled subcritical reactor driven by a Mitrailleuse Accelerator. This paper presents the conceptual design details and approach to implementing the subcritical reactor core. We will spend the next year or so on detailed design work, and then will start work on developing a prototype for demonstration. If there are no obstacles in setting up a development organization, we expect to finish verifying the prototype's performance by the third quarter of 2015. (authors)

  1. OPAL REACTOR: Calculation/Experiment comparison of Neutron Flux Mapping in Flux Coolant Channels

    Energy Technology Data Exchange (ETDEWEB)

    Barbot, L.; Domergue, C.; Villard, J. F.; Destouches, C. [CEA, Paris (France); Braoudakis, G.; Wassink, D.; Sinclair, B.; Osborn, J. C.; Huayou, Wu [ANSTO, Syeney (Australia)

    2013-07-01

    The measurement and calculation of the neutron flux mapping of the OPAL research reactor are presented. Following an investigation of fuel coolant channels using sub-miniature fission chambers to measure thermal neutron flux profiles, neutronic calculations were performed. Comparison between calculation and measurement shows very good agreement.

  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. The advanced neutron source (ANS) project

    International Nuclear Information System (INIS)

    The Advanced Neutron Source (ANS) is a new user experimental facility for neutron research planned at Oak Ridge. The centerpiece of the facility will be a steady-state source of neutrons from a reactor of unprecedented flux. In addition, extensive and comprehensive equipment and facilities for neutron research will be included. The scientific fields to be served include neutron scattering with cold, thermal, and hot neutrons (the most important scientific justification for the project); engineering materials irradiation; isotope production (including transuranium isotopes); materials analysis; and nuclear science

  5. 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)

  6. Securing the future of medical isotopes and neutron science in Canada: the Canadian Neutron Source (CNS)

    International Nuclear Information System (INIS)

    This presentation discusses establishment of the Canadian Neutron Source (CNS) that could be utilized for production of medical isotopes and neutron science in Canada. The Canadian Neutron Source would be 20 MWth research reactor optimized for delivery of medical isotopes and neutron beams for neutron science to serve both industry and the public sector. Employing existing reactor and isotope technology minimizes the risk and schedule. Neutron beams could be used in materials science research, biomedical research as well as imaging.

  7. Pulsed spallation Neutron Sources

    International Nuclear Information System (INIS)

    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

  8. Pulsed spallation neutron sources

    International Nuclear Information System (INIS)

    This paper reviews the early history of pulsed spallation neutron source development ar 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 provide a few examples of applications in fundamental condensed matter physics, materials science and technology

  9. 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.

  10. 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...

  11. 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

  12. 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.

  13. RELAP5 analyses of two hypothetical flow reversal events for the Advanced Neutron Source Reactor

    International Nuclear Information System (INIS)

    The reactor design features 4 independent cooling loops (3 active, 1 standby), each containing a main circulation pump (with battery powered pony motor), heat exchanger, an accumulator, and a check valve. The first transient assumes one of these pumps fails, and also that the check valve in that loop remains stuck open. This accident is considered extremely unlikely. Flow reverses in this loop, reducing 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-dia instantaneous pipe break near the core inlet (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 4 thermal limits: Twall = Tsat, incipient boiling, onset of significant void, and critical heat flux. For the first transient, results show that these limits are not exceeded (at 95% non-exceedance probability level) if the pony motor battery lasts 30 minutes (present design value). For the second transient, the closest approach of the fuel surface temperature to local saturation temperature during core flow reversal is about 39 C, so the fuel remains cool during this transient. Although this work is for the ANSR geometry and operating conditions, the general conclusion may be applicable to other highly subcooled reactor systems

  14. Neutronic conceptual design of the ETRR-2 cold neutron source

    International Nuclear Information System (INIS)

    The conceptual neutronic design of the cold neutron source (CNS) for the Egyptian second research reactor (ETRR-2) was carried out using the MCNP code. A parametric analysis was also performed to choose the type and geometry of the moderator and the required CNS dimensions to maximize the cold neutron production. The moderator cell is a spherical annulus containing liquid hydrogen. The cold neutron gain and brightness are calculated together with the nuclear heat load of the CNS. The effects of void fraction in the moderator cell and the ortho:para ratios on cold neutron gain were calculated. (orig.)

  15. Neutronic analysis for core conversion (HEU-LEU) of Pakistan research reactor-2 (PARR-2)

    International Nuclear Information System (INIS)

    Neutronic analyses for the core conversion of Pakistan research reactor-2 (PARR-2) from high enriched uranium (HEU) fuel to low enriched uranium (LEU) fuel has been performed. Neutronic model has been verified for 90.2% enriched HEU fuel (UAl4-Al). For core conversion, UO2 fuel was chosen as an appropriate fuel option because of higher uranium density. Clad has been changed from aluminum to zircalloy-4. Uranium enrichment of 12.6% has been optimized based on the design basis criterion of excess reactivity 4 mk in miniature neutron source reactor (MNSR). Lattice calculations for cross-section generation have been performed utilizing WIMS while core modeling was carried out employing three dimensions option of CITATION. Calculated neutronic parameters were compared for HEU and LEU fuels. Comparison shows that to get same thermal neutron flux at inner irradiation sites, reactor power has to be increased from 30 to 33 kW for LEU fuel. Reactivity coefficients calculations show that doppler and void coefficient values of LEU fuel are higher while moderator coefficient of HEU fuel is higher. It is concluded that from neutronic point of view LEU fuel UO2 of 12.6% enrichment with zircalloy-4 clad is suitable to replace the existing HEU fuel provided that dimensions of fuel pin and total number of fuel pins are kept same as for HEU fuel

  16. 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

  17. 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.

  18. 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)

  19. 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)

  20. Cold neutron source at CMRR

    International Nuclear Information System (INIS)

    As an effective means to study structure of many materials and law of microscopic movements on atomic or molecular scale, neutron scattering technique is paid more and more attention by many countries. To promote its development in China, a set of advanced Neutron Scattering Experimental Facilities (NSEF) will be installed at China Mianyang Research Reactor (CMRR), currently under construction. The cold neutron source (CNS) on CMRR, one of the most important components of NSEF, is of vertical thermosiphon type, and uses single-phase liquid hydrogen moderator. Nice working capacity and safety are the benefit features of CNS on CMRR. Cooling helium from refrigerator removes the total heat load from CNS in the heat exchanger. In this paper, the in-pile parts, parameters and safety features of CNS are given in detail. At the same time, the utilization of the CNS is briefly described. (author)

  1. 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

  2. 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

  3. Fast neutron benchmark proposal at TRIGA-ACPR Reactor

    International Nuclear Information System (INIS)

    The development of fast neutron benchmarks is a historical aim of reactor physics. The dry experimental tube situated in the central region of the core in TRIGA Annular-Core Pulsing Reactor (ACPR) offers a suitable neutron source for fast neutron benchmark development. Our proposal consists in mounting a high-enriched uranium annular converter into the dry channel of the core. Preliminary computations and measurements are presented in this paper. Neutron flux computations in the dry channel and the uranium converter were performed using MCNP and WIMS codes. Also neutron flux spectrum measurements and fast and thermal neutron flux distribution measurements were performed using foil activation techniques. (authors)

  4. Locally manufactured films for neutron flux measurement in the MNSR type reactor

    International Nuclear Information System (INIS)

    Highlights: • Metal films deposited on Teflon are prepared to use as neutron monitors in the MNSR. • Ti, Al, V, and Ag films have been locally prepared by two different methods. • The thermal neutron flux was measured using Ti, Al, V, and Ag films. • V and Ag films were used as neutron monitors for the first time in MNSR type reactor. • With compared to published results in literature our neutron monitors are validated. - Abstract: Metal films deposited on Teflon are used in the Miniature Neutron Source Reactor (MNSR) for the first time to study their usability as neutron activation detectors for the thermal neutron flux measurements in the reactor. For this purpose Titanium, Aluminum, Vanadium, and Silver films deposited on Teflon have been locally prepared at room temperature using two methods: the vacuum arc deposition and DC Magnetron sputtering techniques. The thermal neutron flux in the MNSR inner irradiation site was measured using the prepared metal films. The results at the 95% level of confidence of the neutron flux using the metal films deposited on Teflon by the vacuum arc deposition for Titanium, Aluminum, and Vanadium were: (9.9 ± 0.3) × 1011, (1.4 ± 0.3) × 1012, (1.2 ± 0.2) × cm−2 s−1, respectively. The result at the same level of confidence of the neutron flux using the metal films deposited on Teflon by the DC Magnetron sputtering for Silver was: (1.5 ± 0.2) × 1011 cm−2 s−1. Good agreements are noticed between our obtained mean value (9.3 ± 0.9) × 1011 cm−2 s−1 and the previous published results

  5. Coded source neutron imaging

    Energy Technology Data Exchange (ETDEWEB)

    Bingham, Philip R [ORNL; Santos-Villalobos, Hector J [ORNL

    2011-01-01

    Coded aperture techniques have been applied to neutron radiography to address limitations in neutron flux and resolution of neutron detectors in a system labeled coded source imaging (CSI). By coding the neutron source, a magnified imaging system is designed with small spot size aperture holes (10 and 100 m) for improved resolution beyond the detector limits and with many holes in the aperture (50% open) to account for flux losses due to the small pinhole size. An introduction to neutron radiography and coded aperture imaging is presented. A system design is developed for a CSI system with a development of equations for limitations on the system based on the coded image requirements and the neutron source characteristics of size and divergence. Simulation has been applied to the design using McStas to provide qualitative measures of performance with simulations of pinhole array objects followed by a quantitative measure through simulation of a tilted edge and calculation of the modulation transfer function (MTF) from the line spread function. MTF results for both 100um and 10um aperture hole diameters show resolutions matching the hole diameters.

  6. 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)

  7. 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.

  8. 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)

  9. Neutron source multiplication method

    International Nuclear Information System (INIS)

    Extensive use has been made of neutron source multiplication in thousands of measurements of critical masses and configurations and in subcritical neutron-multiplication measurements in situ that provide data for criticality prevention and control in nuclear materials operations. There is continuing interest in developing reliable methods for monitoring the reactivity, or k/sub eff/, of plant operations, but the required measurements are difficult to carry out and interpret on the far subcritical configurations usually encountered. The relationship between neutron multiplication and reactivity is briefly discussed and data presented to illustrate problems associated with the absolute measurement of neutron multiplication and reactivity in subcritical systems. A number of curves of inverse multiplication have been selected from a variety of experiments showing variations observed in multiplication during the course of critical and subcritical experiments where different methods of reactivity addition were used, with different neutron source detector position locations. Concern is raised regarding the meaning and interpretation of k/sub eff/ as might be measured in a far subcritical system because of the modal effects and spectrum differences that exist between the subcritical and critical systems. Because of this, the calculation of k/sub eff/ identical with unity for the critical assembly, although necessary, may not be sufficient to assure safety margins in calculations pertaining to far subcritical systems. Further study is needed on the interpretation and meaning of k/sub eff/ in the far subcritical system

  10. Neutron cooling and cold-neutron sources (1962)

    International Nuclear Information System (INIS)

    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)

  11. 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...

  12. Accelerators for Driving Intense spallation Neutron Sources

    International Nuclear Information System (INIS)

    A worldwide trend to replace aging research reactors with accelerator driven neutron sources is currently underway. The ''SARAF'' program at Soreq NRC is a notable example. Setting the background to this trend, a review of the history of accelerator based spallation neutron sources is presented. We follow the evolution of ideas and projects for intense spallation neutron sources. The survey is mainly focused on the properties of the accelerators chosen as drivers throughout the evolution of spallation neutron sources. Since the late 1940s, high-energy proton and deuteron accelerators were developed in view of producing intense neutron sources for various applications related to the nuclear industry, i.e. breeding fissile isotopes, driving nuclear reactors using alternative fuels (like the 'Energy Amplifier') and nuclear waste incineration. However, these projects never progressed beyond the R and D stage. In recent years there is a trend to replace aging reactor-based strong cw neutron sources by pulsed intense spallation sources. Their main applications are in the fields of physics research, material sciences, biology and medicine. Prominent examples of successful projects are ISIS at RAL in Great Britain and SINQ at PSI in Switzerland. Other successful projects are noted in Japan and the US. The clear success of these spallation sources prompted the development of a new generation of more intense spallation neutron sources, notably in Europe (ESS), US (SNS) and Japan (JAERI). Generally, the pulsed spallation neutron sources are based on high-energy proton accelerators. Initially, the proton accelerators were room temperature linacs. In view of the progress relating to properties of RF superconducting resonators and the excellent accumulated experience with cryogenic accelerators, future accelerators for spallation sources will be mostly cryogenic linacs

  13. Advanced Neutron Source operating philosophy

    International Nuclear Information System (INIS)

    An operating philosophy and operations cost estimate were prepared to support the Conceptual Design Report for the Advanced Neutron Source (ANS), a new research reactor planned for the Oak Ridge National Laboratory (ORNL). The operating philosophy was part of the initial effort of the ANS Human Factors Program, was integrated into the conceptual design, and addressed operational issues such as remote vs local operation; control room layout and responsibility issues; role of the operator; simulation and training; staffing levels; and plant computer systems. This paper will report on the overall plans and purpose for the operations work, the results of the work done for conceptual design, and plans for future effort

  14. 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.

  15. 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.

  16. 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

  17. 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.

  18. Accelerator based steady state neutron source

    International Nuclear Information System (INIS)

    Using high current, cw linear accelerator technology, a spallation neutron source can achieve much higher average intensities than existing or proposed pulsed spallation sources. With about 100 mA of 300 MeV protons or deuterons, the accelerator based neutron research facility (ABNR) would initially achieve the 1016 n/cm2s thermal flux goal of the advanced steady state neutron source, and upgrading could provide higher steady state fluxes. The relatively low ion energy compared to other spallation sources has an important impact on R and D requirements as well as capital cost, for which a range of $300-450 M is estimated by comparison to other accelerator-based neutron source facilities. The source is similar to a reactor source is most respects. It has some higher energy neutrons but fewer gamma rays, and the moderator region is free of many of the design constraints of a reactor, which helps to implement sources for various neutron energy spectra, many beam tubes, etc., with the development of a multibeam concept and the basis for currents greater than 100 mA that is assumed in the R and D plan, the ABNR would serve many additional uses, such as fusion materials development, production of proton-rich isotopes, and other energy and defense program needs

  19. An accelerator based steady state neutron source

    International Nuclear Information System (INIS)

    Using high current, cw linear accelerator technology, a spallation neutron source can achieve much higher average intensities than existing or proposed pulsed spallation sources. With about 100 mA of 300 MeV protons or deuterons, the accelerator based neutron research facility (ABNR) would initially achieve the 1016 n/cm2 s themal flux goal of the advanced steady state neutron source, and upgrading could provide higher steady state fluxes. The relatively low ion energy compared to other spallation sources has an important impact on R and D requirements as well as capital cost, for which a range of Dollar 300-450 is estimated by comparison to other accelerator-based neutron source facilities. The source is similar to a reactor source in most respects. It has some higher energy neutrons but fewer gamma rays, and the moderator region is free of many of the design constraints of a reactor, which helps to implement sources for various neutron energy spectra, many beam tubes, etc. With the development of a multibeam concept and the basis for currents greater than 100 mA that is assumed in the R and D plan, the ABNR would serve many additional uses, such as fusion materials development, production of proton-rich isotopes, and other energy and defense program needs. (orig.)

  20. Cold neutron irradiation facility for the Brazilian research reactors

    International Nuclear Information System (INIS)

    Neutron irradiation in research reactors and accelerators can be realized at appropriated neutron guides or beam holes shared around a cold neutron source (CNS) with neutron of variable intensity and energy. An irradiation facility for multiple applications including an intense CNS was calculated for the three Brazilian research reactors and can be utilized as a first concept for the new research reactor to be designed, the Brazilian multiple purpose research reactor (RMB). A study about coolant and moderators properties, and simulations with neutron physics and thermal codes, may be important for the definition of the type of the CNS to be utilized. Some earlier results of MCNP simulations and a discussion about the different factors involved in the definition of its installation in the Brazilian research reactors are here presented. One suggests an international cooperation for the design development of this system and posterior construction of a prototype in the Argonauta reactor at the Instituto de Engenharia Nuclear (IEN-CNEN/RJ). It is also being considered the inclusion of other devices as a neutron fiber to guide the neutron beams away of the gamma radiation and fast neutron background. The cold neutron facility increases the intensity of cold neutrons, without the need of additional fuel burn up. (author)

  1. Neutron beam facilities at the Australian Replacement Research Reactor

    International Nuclear Information System (INIS)

    Australia is building a research reactor to replace the HIFAR reactor at Lucas Heights by the end of 2005. Like HIFAR, the Replacement Research Reactor will be multipurpose with capabilities for both neutron beam research and radioisotope production. It will be a pool-type reactor with thermal neutron flux (unperturbed) of 4 x 1014 n/cm2/sec and a liquid D2 cold neutron source. Cold and thermal neutron beams for neutron beam research will be provided at the reactor face and in a large neutron guide hall. Supermirror neutron guides will transport cold and thermal neutrons to the guide hall. The reactor and the associated infrastructure, with the exception of the neutron beam instruments, is to be built by INVAP S.E. under contract. The neutron beam instruments will be developed by ANSTO, in consultation with the Australian user community. This status report includes a review the planned scientific capabilities, a description of the facility and a summary of progress to date. (author)

  2. Parallel Monte Carlo reactor neutronics

    International Nuclear Information System (INIS)

    The issues affecting implementation of parallel algorithms for large-scale engineering Monte Carlo neutron transport simulations are discussed. For nuclear reactor calculations, these include load balancing, recoding effort, reproducibility, domain decomposition techniques, I/O minimization, and strategies for different parallel architectures. Two codes were parallelized and tested for performance. The architectures employed include SIMD, MIMD-distributed memory, and workstation network with uneven interactive load. Speedups linear with the number of nodes were achieved

  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. Thermo-hydraulic test of the moderator cell of liquid hydrogen cold neutron source for the Budapest research reactor

    International Nuclear Information System (INIS)

    Thermo-hydraulic experiment was carried out in order to test performance of the direct cooled liquid hydrogen moderator cell to be installed at the research reactor of the Budapest Neutron Center. Two electric hearers up to 300 W each imitated the nuclear heat release in the liquid hydrogen as well as in construction material. The test moderator cell was also equipped with temperature gauges to measure the hydrogen temperature at different positions as well as the inlet and outlet temperature of cooling he gas. The hydrogen pressure in the connected buffer volume was also controlled. At 140 w expected total heat load the moderator cell was filled with liquid hydrogen within 4 hours. The heat load and hydrogen pressure characteristics of the moderator cell are also presented. (author)

  5. Neutron beam facilities at the Replacement Research Reactor, ANSTO

    International Nuclear Information System (INIS)

    The exciting development for Australia is the construction of a modern state-of-the-art 20-MW Replacement Research Reactor which is currently under construction to replace the aging reactor (HIFAR) at ANSTO in 2006. To cater for advanced scientific applications, the replacement reactor will provide not only thermal neutron beams but also a modern cold-neutron source moderated by liquid deuterium at approximately -250 deg C, complete with provision for installation of a hot-neutron source at a later stage. The latest 'supermirror' guides will be used to transport the neutrons to the Reactor Hall and its adjoining Neutron Guide Hall where a suite of neutron beam instruments will be installed. These new facilities will expand and enhance ANSTO's capabilities and performance in neutron beam science compared with what is possible with the existing HIFAR facilities, and will make ANSTO/Australia competitive with the best neutron facilities in the world. Eight 'leading-edge' neutron beam instruments are planned for the Replacement Research Reactor when it goes critical in 2006, followed by more instruments by 2010 and beyond. Up to 18 neutron beam instruments can be accommodated at the Replacement Research Reactor, however, it has the capacity for further expansion, including potential for a second Neutron Guide Hall. The first batch of eight instruments has been carefully selected in conjunction with a user group representing various scientific interests in Australia. A team of scientists, engineers, drafting officers and technicians has been assembled to carry out the Neutron Beam Instrument Project to successful completion. Today, most of the planned instruments have conceptual designs and are now being engineered in detail prior to construction and procurement. A suite of ancillary equipment will also be provided to enable scientific experiments at different temperatures, pressures and magnetic fields. This paper describes the Neutron Beam Instrument Project and gives

  6. 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

  7. 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.

  8. 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)

  9. Radiological consequence analyses under severe accident conditions for the Advanced Neutron Source Reactor at the Oak Ridge National Laboratory

    International Nuclear Information System (INIS)

    This paper discusses salient aspects of methodology, assumptions, modeling of various features related to radiation exposure, and health consequences from source terms resulting from two conservatively scoped severe accident scenarios. Radiological consequences for a site-suitability scenario based on 10 CFR 100 guidelines also are presented. Consequences arising from severe accidents involving steaming pools and core-concrete interaction (CCI) events combined with several different containment configurations are presented. Results are presented in the form of mean cumulative values for prompt and latent cancer fatality estimates and related cumulative, complementary distribution functions as a function of distance from the reactor site. It is shown that the reactor-site-suitability risk goals are met by a large margin and that overall risk is dominated by early containment failure combined with CCI events

  10. 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

  11. Cryogenic hydrogen circulation system of neutron source

    International Nuclear Information System (INIS)

    Cold neutron sources of reactors and spallation neutron sources are classic high flux neutron sources in operation all over the world. Cryogenic fluids such as supercritical or supercooled hydrogen are commonly selected as a moderator to absorb the nuclear heating from proton beams. By comparing supercritical hydrogen circulation systems and supercooled hydrogen circulation systems, the merits and drawbacks in both systems are summarized. When supercritical hydrogen circulates as the moderator, severe pressure fluctuations caused by temperature changes will occur. The pressure control system used to balance the system pressure, which consists of a heater as an active controller for thermal compensation and an accumulator as a passive volume controller, is preliminarily studied. The results may provide guidelines for design and operation of other cryogenic hydrogen system for neutron sources under construction

  12. Identification of pollution sources of total suspended air particulates over the Zhongguancun area of Beijing, China

    International Nuclear Information System (INIS)

    The concentrations of twenty six elements were determined by instrumental neutron activation analysis for samples of air particulates collected over a twelve month period in the Zhongguancun area of Beijing. This set of data is analyzed for underlying structure by the method of common factor analysis. The data can be interpreted of the basis of four common factors accounting for 90.3% of the total variance in the system. These factors are attributed to various sources of particulate material by noting the dependence of the factors on the elements

  13. 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

  14. 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.

  15. 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

  16. Neutronics of a D-Li neutron source: An overview

    International Nuclear Information System (INIS)

    The importance of having a high energy (14 MeV) neutron source for fusion materials testing is widely recognized. The availability of a test volume with easy accessibility, with a radiation environment similar to the one expected for a fusion reactor, and with dimensions large enough to accommodate several small samples or a small blanket mock-up are requirements impossible to meet with the existing reactors and irradiation facilities. A D-Li neutron source meets the above mentioned requirements and can be built today with well known technology. This paper describes some relevant topics related to beam target configuration, neutron flux spectrum, and nuclear responses for a D-Li neutron source. The target-beam configuration is analyzed for different beam cross sectional areas and trade-offs between the area of the beam and related quantities such as available volume for testing, peak fluxes, and flux or nuclear responses gradient are presented. The conclusion is that the D-Li neutron source has the necessary characteristics to be the option of choice for IFMIF

  17. Monte Carlo neutron transport simulation of the Ghana Research Reactor-1

    International Nuclear Information System (INIS)

    Stochastic Monte Carlo neutron particle transport methods have been applied to successfully model in 3-D, the HEU-fueled Ghana Research Reactor-1 (GHARR-1), a commercial version of the Miniature Neutron Source Reactor (MNSR) using the MCNP version 4c3 particle transport code. The preliminary multigroup neutronic criticality calculations yielded a keff is contained in 1.00449 with a corresponding cold clean excess reactivity of 4.47mk (447pcm) compared with experimental values of keff is contained in 1.00402 and excess reactivity of 4.00mk (400pcm). The Monte Carlo simulations also show comparable results in the neutron fluxes in the HEU core and some regions of interest. The observed trends in the radial and axial flux distributions in the core, beryllium annular reflector and the water region in the top shim reflector tray were reproduced, indicating consistency of the results, accuracy of the model, precision of the MCNP transport code and the comparability of the Monte Carlo simulations. The results further illustrate the close agreement between stochastic transport theory and the experimental measurements conducted during off-site zero power cold tests. (author)

  18. 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.

  19. 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

  20. Neutronic design of small reactors

    International Nuclear Information System (INIS)

    Small reactors design is one of the main activities of AREVA TA. At the time, AREVA TA main projects are oriented towards research reactors and reactors for military naval propulsion. Due to differences in the physics and performances to meet, each kind of small reactor leads to specific modelling needs. Many computing tools have been developed in order to successfully carry out these projects. These schemes are mainly based on the use of TRIPOLI, MCNP, APOLLO2 and CRONOS2 codes. In that framework, a multi-purpose pre/post processing tool named CHARM is being developed by AREVA NP in partnership with AREVA TA in order to integrate small reactors specification. CHARM is used to elaborate APOLLO2 input data while various dedicated tools are used to automatically generate TRIPOLI and MNCP input data. These 3D numerical models need a very accurate spatial description to perform specific calculations. As an example, for the JHR design, after calculating 3D burn up by APOLLO2/MOC models, the data is fed back into a TRIPOLI model used for safety analyses. This paper presents our methodology for the small core design and 3 examples: 1) The calculation scheme for the JHR (Jules Horowitz Reactor) neutronic studies. These design studies are a recent illustration of combined use of both deterministic and probabilistic codes, 2) The use of CHARM, with the modelling of a JHR core. The purpose of CHARM- V2, based on Open Cascade Technology, is to provide a pre/post processing tool for APOLLO2/MOC, TRIPOLI4 and MCNP solvers, 3) The depletion Monte Carlo calculation of a MTR core. (author)

  1. Measurements of neutron flux in the RA reactor

    International Nuclear Information System (INIS)

    This report includes results of the following measurements performed at the RA reactor: thermal neutron flux in the experimental channels, epithermal and fast neutron flux, neutron flux in the biological shield, neutron flux distribution in the reactor cell

  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. 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)

  4. Sodium fast neutron reactors. Status and perspective of development

    International Nuclear Information System (INIS)

    This report reveals data on development history of domestic fast neutron reactors cooled with sodium (BN reactors). It also shows BN reactors' unique role in expanding source of nuclear power raw materials and in solving ecological problems relating to radioactive wastes. There is brief information on characteristics and operation experience of research reactors BR-10, BOR-60, pilot-industrial reactors BN-350 and BN-600. As well there is data on BN-800 reactor designing that obtained a license for building. There are considered BN reactor peculiarities in regard of safety and design decisions on safety provision at the level meeting standard document requirements. BN reactor technical and economic indices and the ways of their improvement are evaluated. There is brief information on alternative perspective technologies of fast reactors, in particular regarding 'BREST-300' reactor cooled with lead coolant

  5. Glow discharge electron impact ionization source for miniature mass spectrometers.

    Science.gov (United States)

    Gao, Liang; Song, Qingyu; Noll, Robert J; Duncan, Jason; Cooks, R Graham; Ouyang, Zheng

    2007-05-01

    A glow discharge electron impact ionization (GDEI) source was developed for operation using air as the support gas. An alternative to the use of thermoemission from a resistively heated filament electron source for miniature mass spectrometers, the GDEI source is shown to have advantages of long lifetime under high-pressure operation and low power consumption. The GDEI source was characterized using our laboratory's handheld mass spectrometer, the Mini 10. The effects of the discharge voltage and pressure were investigated. Design considerations are illustrated with calculations. Performance is demonstrated in a set of experimental tests. The results show that the low power requirements, mechanical ruggedness, and quality of the data produced using the small glow discharge ion source make it well-suited for use with a portable handheld mass spectrometer. PMID:17441220

  6. 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.

  7. 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.

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

    International Nuclear Information System (INIS)

    The nucleus of 208Pb, a system 18 orders of magnitude 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. Indeed, models with a thicker neutron skin in 208Pb generate larger neutron stars that have a lower liquid-to-solid transition density. Further, it is shown how the correlation between the neutron skin in 208Pb and the radius of a 1.4 solar-mass neutron star may be used to place important constraints on the equation of state of neutron-rich matter and how it may help elucidate the existence of a phase transition at the core of the star. (author)

  9. 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

  10. Survey of research reactors

    International Nuclear Information System (INIS)

    A survey of reasearch reactors based on the IAEA Nuclear Research Reactor Data Base (RRDB) was done. This database includes information on 273 operating research reactors ranging in power from zero to several hundred MW. From these 273 operating research reactors 205 reactors have a power level below 5 MW, the remaining 68 reactors range from 5 MW up to several 100 MW thermal power. The major reactor types with common design are: Siemens Unterrichtsreaktors, 1.2 Argonaut reactors, Slowpoke reactors, the miniature neutron source reactors, TRIGA reactors, material testing reactors and high flux reactors. Technical data such as: power, fuel material, fuel type, enrichment, maximum neutron flux density and experimental facilities for each reactor type as well as a description of their utilization in physics and chemistry, medicine and biology, academic research and teaching, training purposes (students and physicists, operating personnel), industrial application (neutron radiography, silicon neutron transmutation doping facilities) are provided. The geographically distribution of these reactors is also shown. As conclusions the author discussed the advantages (low capital cost, low operating cost, low burn up, simple to operate, safe, less restrictive containment and sitting requirements, versatility) and disadvantages (lower sensitivity for NAA, limited radioisotope production, limited use of neutron beams, limited access to the core, licensing) of low power research reactors. 24 figs., refs. 15, Tab. 1 (nevyjel)

  11. 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

  12. 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.

  13. 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.

  14. [Regional Source Apportionment of PM2.5 in Beijing in January 2013].

    Science.gov (United States)

    Li, Xuan; Nie, Teng; Qi, Jun; Zhou, Zhen; Sun, Xue-song

    2015-04-01

    In January 2013, Beijing area experienced several severe haze weather events. The pollution of fine particles has become an important problem in Beijing. Understanding the sources of PM2.5 in Beijing is essential for solutions and related policy-formulations. Three-dimensional air quality modelling system was established to analyze the PM2.5 pollution during 20-24 January in 2013. PSAT technology was used to study the regional sources of Beijing PM2.5 pollution. The results showed that local emission was the major source of PM2.5 in Beijing City, with an average contribution rate of 34% . The average contribution rates of Hebei and Tianjin were 26% and 4%, respectively. The neighboring area and the boundary conditions contributed 12% and 24% to PM2.5 in Beijing. In the heavy pollution period, the influence of regional transportation increased significantly, and became the major source of PM2.5 pollution in Beijing. Nitrate in PM2.5 in Beijing mainly came from the surrounding area of Beijing City, while sulfate and secondary organic aerosols showed characteristics of long-distance transportation. Ammonium salt and other components were mainly from Beijing local contribution. PMID:26164884

  15. IBR-2: A periodically operating pulsed reactor for neutron research

    International Nuclear Information System (INIS)

    An extensive description of the IBR-2 fast neutron pulse reactor is given. The operation and construction of the reactor, the control and safety systems and the cooling system are described in details. In the booster mode of operation a linear induction accelerator is applied as injector. The shielding problems and the experimental possibilities are also discussed. The particular problems in the design, construction and operation of such reactors are reviewed. The general characteristics of the IBR-2 as a neutron source for spectrometry as well as a list of the experimental apparatus under construction are given. (R.J.)

  16. 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

  17. 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

  18. 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

  19. 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.

  20. Advanced Neutron Sources: Plant Design Requirements

    Energy Technology Data Exchange (ETDEWEB)

    1990-07-01

    The Advanced Neutron Source (ANS) is a new, world class facility for research using hot, thermal, cold, and ultra-cold neutrons. At the heart of the facility is a 350-MW{sub th}, heavy water cooled and moderated reactor. The reactor is housed in a central reactor building, with supporting equipment located in an adjoining reactor support building. An array of cold neutron guides fans out into a large guide hall, housing about 30 neutron research stations. Office, laboratory, and shop facilities are included to provide a complete users facility. The ANS is scheduled to begin operation at the Oak Ridge National Laboratory at the end of the decade. This Plant Design Requirements document defines the plant-level requirements for the design, construction, and operation of the ANS. This document also defines and provides input to the individual System Design Description (SDD) documents. Together, this Plant Design Requirements document and the set of SDD documents will define and control the baseline configuration of the ANS.

  1. 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

  2. Advanced Neutron Sources: Plant Design Requirements

    International Nuclear Information System (INIS)

    The Advanced Neutron Source (ANS) is a new, world class facility for research using hot, thermal, cold, and ultra-cold neutrons. At the heart of the facility is a 350-MWth, heavy water cooled and moderated reactor. The reactor is housed in a central reactor building, with supporting equipment located in an adjoining reactor support building. An array of cold neutron guides fans out into a large guide hall, housing about 30 neutron research stations. Office, laboratory, and shop facilities are included to provide a complete users facility. The ANS is scheduled to begin operation at the Oak Ridge National Laboratory at the end of the decade. This Plant Design Requirements document defines the plant-level requirements for the design, construction, and operation of the ANS. This document also defines and provides input to the individual System Design Description (SDD) documents. Together, this Plant Design Requirements document and the set of SDD documents will define and control the baseline configuration of the ANS

  3. 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)

  4. 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)

  5. 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

  6. Neutron sources in Canada - Present and future

    Science.gov (United States)

    Dolling, G.; Lidstone, R. F.

    Canada's pre-eminent neutron source since 1957 has been the NRU reactor at Chalk River. It is unlikely to operate beyond the year 2005. In 1994, AECL prepared the case and concept for a new research reactor, the Irradiation Research Facility (IRF), to replace NRU. The IRF was developed with the dual purpose of meeting the needs of both R&D programs to support existing and advanced CANDU® designs and also of condensed matter science and materials research using extracted neutron beams. In November 1995, AECL began a pre-project engineering programme to develop the design of the facility and to begin the safety analysis and “up-front” licensing process. The dual-purpose concept continues to be pursued and the design modified, to achieve maximum performance in the most cost-effective manner. The planned neutron-beam facilities, which include a cold source and a guide hall, will greatly enhance Canada's programs of neutron-beam research and applications. The current status of the IRF design and of efforts to secure funding for the neutron-beam components will be presented.

  7. 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 (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 geochemical models of the solubility-limiting phase. A study of the competing effects of radiation damage and annealing in a U-bearing crystal of zircon shows that low temperature annealing in actinide-bearing phases is significant in the annealing of radiation damage.

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

    Directory of Open Access Journals (Sweden)

    Aldawahra Saadou

    2015-06-01

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

  9. The reactor and cold neutron research facility at NIST

    International Nuclear Information System (INIS)

    The NIST Reactor (NBSR) is a 20 MW research reactor located at the Gaithersburg, MD site, and has been in operation since 1969. It services 26 thermal neutron facilities which are used for materials science, chemical analysis, nondestructive evaluation, neutron standards work, and irradiations. In 1987 the Department of Commerce and NIST began development of the CNRF - a $30M National Facility for cold neutron research -which will provide fifteen new experimental stations with capabilities currently unavailable in this country. As of May 1992, four of the planned seven guides and a cold port were installed, eight cold neutron experimental stations were operational, and the Call for Proposals for the second cycle of formally-reviewed guest-researcher experiments had been sent out. Some details of the performance of instrumentation are described, along with the proposed design of the new hydrogen cold source which will replace the present D2O/H2O ice cold source. (author)

  10. 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)

  11. Ion source requirements for pulsed spallation neutron sources

    International Nuclear Information System (INIS)

    The neutron scattering community has endorsed the need for a high- power (1 to 5 MW) accelerator-driven source of neutrons for materials research. Properly configured, the accelerator could produce very short (sub-microsecond) bursts of cold neutrons, said time structure offering advantages over the continuous flux from a reactor for a large class of experiments. The recent cancellation of the ANS reactor project has increased the urgency to develop a comprehensive strategy based on the best technological scenarios. Studies to date have built on the experience from ISIS (the 160 KW source in the UK), and call for a high-current (approx. 100 mA peak) H- source-linac combination injecting into one or more accumulator rings in which beam may be further accelerated. The 1 to 5 GeV proton beam is extracted in a single turn and brought to the target-moderator stations. The high current, high duty-factor, high brightness and high reliability required of the ion source present a very large challenge to the ion source community. A workshop held in Berkeley in October 1994, analyzed in detail the source requirements for proposed accelerator scenarios, the present performance capabilities of different H- source technologies, and identified necessary R ampersand D efforts to bridge the gap

  12. 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...

  13. 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.

  14. Neutronic models for the HIFAR reactor

    International Nuclear Information System (INIS)

    Standard neutronic models have been developed for the AAEC's materials testing reactor HIFAR, and are available as members of a partitioned data set. The models have been used to calculate reactor physics parameters related to operation and safety. Results from the calculations are presented

  15. 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

  16. Materials for spallation neutron sources

    International Nuclear Information System (INIS)

    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

  17. 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.

  18. Neutron scattering facilities at the research reactor DR3

    International Nuclear Information System (INIS)

    DR3 is a heavy-water-moderated 10 MW thermal neutron research reactor. The 26 fuel elements contain 2.5-3.5 kg uranium enriched to less than 20% 235U. Neutron beams emerge from four horizontal through-tubes tangential to the reactor core. Two of the horizontal tubes are used for neutron scattering experiments in the field of materials research. The vertical tubes are predominantly used for isotope production and materials testing. The thermal neutron flux is about 3.5x1013 n/cm2/s in the centre of the 7-inch diameter horizontal through-tubes. The thermal neutron flux in equilibrium with the D2O moderator (50 deg. C) has a nearly Maxwellian distribution peaking at 1.1 A. At the maximum flux position in the two horizontal through-tubes used for materials research are installed scatterers designed with a considerably higher scattering power for thermal neutrons than for fast neutrons and gamma-rays. The scatterer is either a 10 mm slab of light water, providing a nearly thermal Maxweellian spectrum at the beam port, or a chamber filled with supercritical hydrogen gas at 16 atmospheres and 38 K, a so-called cold neutron source. The spectrum from a cold source has a considerable flux enhancement in the long wavelength region when compared to the thermal water scatterer. Neutron beams are available for materials research from two thermal and two cold beam ports in the Reactor Hall. One of the cold beams is shared with a 20 meter long cold-neutron guide-tube which provides three beam ports in a separate building, the Neutron House, with could neutrons. Only neutrons that have undergone total reflection from the Ni-coated glass plates in the bent guide-tube arrive at the end of the guide tube in the Neutron House. The angle of total reflection is proportional to the neutron wavelength. Therefore almost no neutrons of wavelength below a certain ''critical'' wavelength are transmitted through the guide-tube, and the experimental equipment installed in the Neutron Houyse

  19. Source characterization of Purnima Neutron Generator (PNG)

    International Nuclear Information System (INIS)

    The use of 14.1 MeV neutron generators for the applications such as elemental analysis, Accelerated Driven System (ADS) study, fast neutron radiography requires the characterization of neutron source i.e neutron yield (emission rate in n/sec), neutron dose, beam spot size and energy spectrum. In this paper, a series of experiments carried out to characterize this neutron source. The neutron source has been quantified with neutron emission rate, neutron dose at various source strength and beam spot size at target position

  20. PARR-2: reactor description and experiments

    International Nuclear Information System (INIS)

    PARR-2 is a miniature neutron source reactor (MNSR) research reactor has been designed at the rate of 27 kW. Reactor assembly comprises of peaking characteristics with a self limiting flux. In this report reactor description with its assembly and instrumentation control system has been explained. The reactor engineering and physics experiments which can be performed on this reactor are explained in this report. PARR-2 is fueled with HEU fuel pins which are about 90% enriched in U-235. Specific requirements for the safety of the reactor, its building and the personnel, normal instrumentation as required in an industrial environment is sufficient. (A.B.)

  1. 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)

  2. 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.

  3. 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.

  4. 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

  5. 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...

  6. An inductively coupled miniature plasma jet source at microwave frequencies

    International Nuclear Information System (INIS)

    A miniature double plasma jet source driven at microwave frequencies (∼2.45 GHz) was developed and analyzed. The source consists of a copper resonator (screened within an aluminum housing) that excites plasma simultaneously in two alumina tubes of 5 mm internal diameter. Field and plasma simulations were performed using the software Comsol. Assuming a homogeneous electron distribution we calculate the plasma impedance as a function of its conductivity. The electron density and the plasma conductivity are estimated as a function of the absorbed power in plasma for argon and oxygen. Experimentally it was shown that the microwave energy is coupled into oxygen plasma with an efficiency of >85% and into argon plasma with ∼30%. The source efficiently produces atomic oxygen and nitrogen as is demonstrated by plasma-enhanced atomic layer deposition. Finally, the time evolution during ignition, the transition from low efficient capacitive to highly efficient inductive coupling, the free electron distribution as a function of time and other parameters are analyzed. (paper)

  7. Neutronic Design of a Cold Neutron Source with MCNP

    International Nuclear Information System (INIS)

    The neutronic design of a cold neutron source (CNS) requires the use of powerful tools to model neutron transport as accurately as possible. For this purpose, nowadays, the increase in hardware calculation power makes it possible to make use of Monte Carlo techniques, even during the design stage. For design purposes, the goal is to find the optimal combination between positioning and geometry of the moderator chamber and composition of the moderator material to produce the maximum cold neutron flux at the experimental location. Close to the optimum balance, the influence of each of these parameters on the cold flux can be expected to be about 1-5%. These small effects must be discriminated from statistical errors without a strong increase of the calculation time. A short description of the calculation line, leading to a fast and reliable method to perform these optimization calculations with low statistical errors and times compatible with a design schedule is presented. Several parametric analyses of the design variables are presented in order to show how this calculation methodology works and how consistent their results are. The analysis was done during the design of the replacement research reactor (RRR) CNS for the Australian Nuclear Science and Technology Organisation (ANSTO). As a conclusion to the paper, we demonstrate the possibility to apply Monte Carlo techniques in a design project framework to obtain an optimized CNS neutronic design. (author)

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

    International Nuclear Information System (INIS)

    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

  9. 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.

  10. Status report of the program on neutron beam utilization at the Dalat Nuclear Research Reactor

    International Nuclear Information System (INIS)

    The thermal reactor is an intense source not only of thermal neutron, but also intermediate as well as fast neutrons. Using the filtered neutron beam technique at steady state atomic reactor allows receiving the neutrons in the intermediate energy region with the most available intense flux at present. In the near time at the Dalat reactor the filtered neutron beam technique has been applied. Utilization of the filtered neutron beams in basic and applied researches has been a important activity of the Dalat Nuclear Research Institute (DNRI). This report presents some relevant characteristics of the filtered neutron beams and their utilization in nuclear data measurements, neutron capture gamma ray spectroscopy, neutron radiography, neutron dose calibration and other applications. (author). 3 refs, 2 figs

  11. Low dimensional neutron moderators for enhanced source brightness

    DEFF Research Database (Denmark)

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

    2014-01-01

    In a recent numerical optimization study we have found that liquid para-hydrogen coupled cold neutron moderators deliver 3–5 times higher cold neutron brightness at a spallation neutron source if they take the form of a flat, quasi 2-dimensional disc, in contrast to the conventional more voluminous...... 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 to the...... shapes used by now. In the present paper we describe a simple theoretical explanation of this unexpected behavior, which is based on the large difference in para-hydrogen between the values of the scattering mean free path for thermal neutrons (in the range of 1 cm) and its much larger equivalent for...

  12. 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.

  13. 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.

  14. 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

  15. EL-2 reactor: Thermal neutron flux distribution

    International Nuclear Information System (INIS)

    The flux distribution of thermal neutrons in EL-2 reactor is studied. The reactor core and lattices are described as well as the experimental reactor facilities, in particular, the experimental channels and special facilities. The measurement shows that the thermal neutron flux increases in the central channel when enriched uranium is used in place of natural uranium. However the thermal neutron flux is not perturbed in the other reactor channels by the fuel modification. The macroscopic flux distribution is measured according the radial positioning of fuel rods. The longitudinal neutron flux distribution in a fuel rod is also measured and shows no difference between enriched and natural uranium fuel rods. In addition, measurements of the flux distribution have been effectuated for rods containing other material as steel or aluminium. The neutron flux distribution is also studied in all the experimental channels as well as in the thermal column. The determination of the distribution of the thermal neutron flux in all experimental facilities, the thermal column and the fuel channels has been made with a heavy water level of 1825 mm and is given for an operating power of 1000 kW. (M.P.)

  16. 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

  17. 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.

  18. 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.

  19. TOF powder diffractometer on a reactor source

    International Nuclear Information System (INIS)

    Complete text of publication follows. The performance of time-of-flight (TOF) methods on Long Pulse Spallation Sources can be studied at a reactor source. For this purpose a prototype TOF monochromator instrument will be installed at the KFKI reactor in Budapest. The initial setup will be a powder diffractometer with a resolution of δd/d down to 2 x 10-3 at a wavelength of 1 A. The instrument uses choppers to produce neutron pulses of down to 10 μs FWHM. The optimal neutron source for a chopper instrument is a Long Pulse Spallation Source, but even on a continuous source simulations have shown that this instrument outperforms a conventional crystal monochromator powder diffractometer at high resolution. The main components of the TOF instrument are one double chopper defining the time resolution and two single choppers to select the wavelength range and to prevent frame overlap. For inelastic experiments a further chopper can be added in front of the sample. The neutron guide has a super-mirror coating and a curvature of 3500m. The total flight path is 20m and there are 24 single detectors in backscattering geometry. (author)

  20. Design of neutron beams for boron neutron capture therapy in a fast reactor

    International Nuclear Information System (INIS)

    The BNCT (Boron Neutron Capture Therapy) technique makes use of thermal or epithermal neutrons to irradiate tumours previously loaded with 10B. Reactors are currently seen as a suitable neutron source for BNCT implementation, due to the high intensity of the flux they can provide. The TAPIRO reactor, that is located at the ENEA Casaccia Centre near Rome, is a low-power fast-flux research reactor that can be usefully employed for this application. In this work computer simulations were carried out on this reactor to obtain epithermal and thermal neutron beams for the application of BNCT in Italy in the framework of a specific research program. Comparisons with measurements are also reported. Using the MCNP-4B code, Monte Carlo calculations were carried out to determine the materials suitable for the design of the thermal and epithermal columns. Various arrangements of reflector and moderator materials have been investigated to achieve the desired experimental constraints. On the basis of these calculations, a thermal column was designed and installed in the TAPIRO reactor to perform preliminary experiments on small laboratory animals. For the planning of a therapy treatment of gliomas on larger size animals, several material configurations were investigated in the search for an optimal epithermal facility. The aim of the present study is to indicate how a fast research reactor can be successfully modified for generating neutron beams suitable for BNCT applications. (author)

  1. 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.

  2. Neutrons down-under: Australia's research reactor review

    International Nuclear Information System (INIS)

    Australian research reactor review commenced in September 1992, the Review had the following Terms of Reference: Whether, on review of the benefits and costs for scientific, commercial, industrial and national interest reasons, Australia has a need for a new reactor; a review of the present reactor, HIFAR, to include: an assessment of national and commercial benefits and costs of operations, its likely remaining useful life and its eventual closure and decommissioning; if Australia has a need for a new nuclear research reactor, the Review will consider: possible locations for a new reactor, its environmental impact at alternative locations, recommend a preferred location, and evaluate matters associated with regulation of the facility and organisational arrangements for reactor-based research. From the Review findings the following recommendations were stated: keep HIFAR going; commission a PRA to ascertain HIFAR's remaining life and refurbishment possibilities; identify and establish a HLW repository; accept that neither HIFAR nor a new reactor can be completely commercial; any decision on a new neutron source must rest primarily on benefits to science and Australia's national interest; make a decision on a new neutron source in about five years' time (1998). Design Proposals for a New Reactor are specified

  3. Attenuation of reactor thermal neutrons in a bulk shield of ordinary concrete

    International Nuclear Information System (INIS)

    This work is concerned with the study of the distribution attenuation of doses of thermal neutrons emitted directly from the core of research reactor in ordinary concrete shield. In practice it is not possible to identify the reactor thermal neutrons in the emitted continuos neutron spectrum. Therefore, measurement was carried out by using a direct and cadmium filtered beam of reactor neutrons. All measurements were performed using Li2B4O7:Mn thermoluminescent dosimeters. The data obtained were analyzed and the dose distributions of reactor thermal neutrons were evaluated. A group of isodose curves constructed which give directly the shape and thickness of the shield required to attenuate the intensity of doses of reactor thermal neutrons to specific values. In addition, the thermal neutron relaxation lengths in ordinary concrete were derived for disc-collimated beam and infinite plane mono-directional sources

  4. Hot Fuel Examination Facility's neutron radiography reactor

    International Nuclear Information System (INIS)

    Argonne National Laboratory-West is located near Idaho Falls, Idaho, and is operated by the University of Chicago for the United States Department of Energy in support of the Liquid Metal Fast Breeder Reactor Program, LMFBR. The Hot Fuel Examination Facility, HFEF, is one of several facilities located at the Argonne Site. HFEF comprises a large hot cell where both nondestructive and destructive examination of highly-irradiated reactor fuels are conducted in support of the LMFBR program. One of the nondestructive examination techniques utilized at HFEF is neutron radiography, which is provided by the NRAD reactor facility (a TRIGA type reactor) below the HFEF hot cell

  5. McCARD for neutronics design and analysis of research reactor cores

    International Nuclear Information System (INIS)

    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 research reactor, and YALINA subcritical facility. (authors)

  6. 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

  7. 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

  8. 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

  9. Neutron flux measurements in PUSPATI Triga Reactor

    International Nuclear Information System (INIS)

    Neutron flux measurement in the PUSPATI TRIGA Reactor (PTR) was initiated after its commissioning on 28 June 1982. Initial measured thermal neutron flux at the bottom of the rotary specimen rack (rotating) and in-core pneumatic terminus were 3.81E+11 n/cm2 sec and 1.10E+12n/cm2 sec respectively at 100KW. Work to complete the neutron flux data are still going on. The cadmium ratio, thermal and epithermal neutron flux are measured in the reactor core, rotary specimen rack, in-core pneumatic terminus and thermal column. Bare and Cadmium covered gold foils and wires are used for the above measurement. The activities of the irradiated gold foils and wires are determined using Ge(Li) and hyperpure germinium detectors. (author)

  10. 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

  11. 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.

  12. Neutron sources for the research of condensed matter

    International Nuclear Information System (INIS)

    Neutron scattering experiments are a powerful technique to study the microscopic behavior of matter for physic, chemistry, material research, biology and geology. The need to investigate the structure and dynamics on a microscopic level implies the need of a new high flux neutron source (ANS). For the future, high flux sources are necessary to measure novel and unforeseen results with high resolution instruments. Small reactors can be used for extensive and detailed neutron measurements. The neutrons which are at disposal, should be effectively used by improvement of the facility

  13. 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.

  14. International Seminar on Advanced Pulsed Neutron Sources PANS-II. Invited talks

    International Nuclear Information System (INIS)

    A conceptual design of creating intense pulsed neutron sources based on high-current accelerators and pulsed reactors for neutron scattering experiments is considered. The progress in high-efficiency moderator developments is shown. Results of diffraction studied are presented

  15. Utilisation of a low power reactor for instrumental neutron activation analysis of 40 elements in coal

    International Nuclear Information System (INIS)

    Coal is mostly used as an energy source for power generation and in local brick kilns in Pakistan, which causes environmental pollution problems due to the release of toxic constituents. Elemental characterisation provides useful information regarding the nature of environmental pollutants to which coal workers and adjacent terrain are exposed and circumscribes different elements as they exist in the coal. Instrumental neutron activation analysis employing a low power miniature neutron source reactor has been used for the determination of 40 major, minor, and trace elements in coal. Bituminous, sub-bituminous and lignite coal varieties of Pakistan were analysed which show that bituminous coal from Salt Range contains lower amount of toxic elements. The quality of the analysed data has been assured by a simultaneous analysis of the IAEA and NBS/NIST certified reference materials. The data will be useful for extrapolating the extent of elemental emission through the combustion of these coals. Enrichment factors calculated for these elements in coal show high values for As, Br, Cl, Dy, Hg, Mo, Sb, and Se, indicating difference in geochemistry and growth environment of the coal deposits. Elemental concentrations of our coal varieties have been compared with those of other countries. (orig.)

  16. Neutronic conceptual design of the ETRR-2 cold-neutron source using the MCNP code

    Science.gov (United States)

    Khalil, M. Y.; Shaat, M. K.; Abdelfattah, A. Y.

    2005-04-01

    A conceptual neutronic design of the cold-neutron source (CNS) for the Egyptian second research reactor (ETRR-2) was done using the MCNP code. Parametric analysis to chose the type and geometry of the moderator, and the required CNS dimensions to maximize the cold neutron production was performed. The moderator cell has a spherical annulus structure containing liquid hydrogen. The cold neutron gain and cold neutron brightness are calculated together with the nuclear heat load of the CNS. Analysis of the estimated performance of the CNS has been done regarding the effect of void fraction in the moderator cell together with the ortho: para ratio.

  17. Neutronic conceptual design of the ETRR-2 cold-neutron source using the MCNP code

    International Nuclear Information System (INIS)

    A conceptual neutronic design of the cold-neutron source (CNS) for the Egyptian second research reactor (ETRR-2) was done using the MCNP code. Parametric analysis to chose the type and geometry of the moderator, and the required CNS dimensions to maximize the cold neutron production was performed. The moderator cell has a spherical annulus structure containing liquid hydrogen. The cold neutron gain and cold neutron brightness are calculated together with the nuclear heat load of the CNS. Analysis of the estimated performance of the CNS has been done regarding the effect of void fraction in the moderator cell together with the ortho: para ratio

  18. New neutron physics using spallation sources

    International Nuclear Information System (INIS)

    The extraordinary neutron intensities available from the new spallation pulsed neutron sources open up exciting opportunities for basic and applied research in neutron nuclear physics. The energy range of neutron research which is being explored with these sources extends from thermal energies to almost 800 MeV. The emphasis here is on prospective experiments below 100 keV neutron energy using the intense neutron bursts produced by the Proton Storage Ring (PSR) at Los Alamos. 30 refs., 10 figs

  19. 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.

  20. 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

  1. Intense neutron source facility for the fusion energy program

    International Nuclear Information System (INIS)

    The Intense Neutron Source Facility, INS, has been proposed to provide a neutronic environment similar to that anticipated in a fully operational fusion-power reactor. The neutron generator will produce an intense flux of 14-MeV neutrons greater than 1014 neutrons per cm2/sec from the collision of two intersecting beams, one of 1.1 A of 270 keV tritium ions and the other of a supersonic jet of deuterium gas. Using either the pure 14-MeV primary neutron spectrum or by tailoring the spectrum with appropriate moderators, crucial radiation-damage effects which are likely to occur in fusion reactors can be thoroughly explored and better understood

  2. 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

  3. HANARO Cold Neutron Source Design

    International Nuclear Information System (INIS)

    The cold neutron source (CNS) design has been completed and confirmed by the full scale mock-up test. When its licensing is expected to be issued within 2007, the CNS will be installed in HANARO in 2009 and be operated from 2010 after the commissioning. The production of cold neutrons from 2009 will enable the neutron guides and the scattering instruments to be commissioned in parallel. From 2010, a new era of neutron science will be open in the area of biotechnology, nano-technology, and material science through the probing capability of cold neutrons with nano-wavelength. The prominent research output that will be created from this cold neutron research facility will ensure the basic science and technology, which will provide the strong foundation for the advanced engineering and technology. This paper presents the design of in-pool assembly including the nuclear design of moderator cell, the manufacturing test of in-pool assembly, the full scale mock-up test, and the safety analysis

  4. 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.

  5. UCN Source at an External Beam of Thermal Neutrons

    International Nuclear Information System (INIS)

    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

  6. 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

  7. The new neutron imaging facility at TRIGA reactor in Morocco

    International Nuclear Information System (INIS)

    A new neutron imaging facility is currently developed around 2 MW TRIGA MARK-II reactor at Maamora Nuclear research centre (CENM). Neutron imaging combined to X-ray or gamma radiography offers the opportunity to extend Non Destructive Testing (NDT) activities DT in Morocco to new fields of applications such as space and aircraft Moroccan industry, mining, wood industry and Archeology. The facility is planed to be completed in the end of 2011. In order to reduce the gamma-ray content in the neutron beam, the reactor tangential channel is selected. For power of 2 MW, the corresponding thermal neutron flux at the inlet of the tangential channel is around 1.1013ncm2/s. The facility will be based on a conical neutron collimator with a flight tube of 8m and offers three circular diaphragms with diameters of 1cm, 2 cm and 4 cm corresponding to L/D-ratio varying between 200 and 600. The holes will be housed in the primary shutter. These diaphragms' sizes allow to perform neutron radiography with high resolution (L/D = 600) and high speed (L/D= 200). Monte Carlo calculations by a fully 3D numerical code GEANT4 are used to optimize the whole neutron beam line and to reach a shorten distance between the source and detector and reduce as possible the exposure time. (author)

  8. The new neutron imaging facility at TRIGA reactor in Morocco

    Energy Technology Data Exchange (ETDEWEB)

    Ouardi, A.; Alami, R.; Bensitel, A. [Centre National de l' Energie des Science et des Techniques Nucleaires, PB.1382 R.P 10001 Rabat (Morocco)

    2011-07-01

    A new neutron imaging facility is currently developed around 2 MW TRIGA MARK-II reactor at Maamora Nuclear research centre (CENM). Neutron imaging combined to X-ray or gamma radiography offers the opportunity to extend Non Destructive Testing (NDT) activities DT in Morocco to new fields of applications such as space and aircraft Moroccan industry, mining, wood industry and Archeology. The facility is planed to be completed in the end of 2011. In order to reduce the gamma-ray content in the neutron beam, the reactor tangential channel is selected. For power of 2 MW, the corresponding thermal neutron flux at the inlet of the tangential channel is around 1.10{sup 13}ncm{sup 2}/s. The facility will be based on a conical neutron collimator with a flight tube of 8m and offers three circular diaphragms with diameters of 1cm, 2 cm and 4 cm corresponding to L/D-ratio varying between 200 and 600. The holes will be housed in the primary shutter. These diaphragms' sizes allow to perform neutron radiography with high resolution (L/D = 600) and high speed (L/D= 200). Monte Carlo calculations by a fully 3D numerical code GEANT4 are used to optimize the whole neutron beam line and to reach a shorten distance between the source and detector and reduce as possible the exposure time. (author)

  9. 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.

  10. Current status for TRR-II Cold Neutron Source

    International Nuclear Information System (INIS)

    The Taiwan Research Reactor (TRR) project (TRR-II) is carrying out at Institute of Nuclear Energy Research (INER) from October 1998 to December 2006. The purpose of Cold Neutron Source (CNS) project is to build entire CNS facility to generate cold neutrons within TRR-II reactor. The objective of CNS design is to install CNS facility with a competitive brightness of cold neutron beam to other facilities in the world. Based on the TRR-II CNS project schedule, the conceptual design for TRR-II CNS facility has been completed and the mock-up test facility for full-scale hydrogen loop has been designed. (author)

  11. Studies and modeling of cold neutron sources

    International Nuclear Information System (INIS)

    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

  12. 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.

  13. Neutron spectra produced by moderating an isotopic neutron source

    International Nuclear Information System (INIS)

    A Monte Carlo study has been carried out to determine the neutron spectra produced by an isotopic neutron source inserted in moderating media. Most devices used for radiation protection have a response strongly dependent on neutron energy. ISO recommends several neutron sources and monoenergetic neutron radiations, but actual working situations have broad spectral neutron distributions extending from thermal to MeV energies, for instance, near nuclear power plants, medical applications accelerators and cosmic neutrons. To improve the evaluation of the dosimetric quantities, is recommended to calibrate the radiation protection devices in neutron spectra which are nearly like those met in practice. In order to complete the range of neutron calibrating sources, it seems useful to develop several wide spectral distributions representative of typical spectra down to thermal energies. The aim of this investigation was to use an isotopic neutron source in different moderating media to reproduce some of the neutron fields found in practice. MCNP code has been used during calculations, in these a 239PuBe neutron source was inserted in H2O, D2O and polyethylene moderators. Moderators were modeled as spheres and cylinders of different sizes. In the case of cylindrical geometry the anisotropy of resulting neutron spectra was calculated from 0 to 2. From neutron spectra dosimetric features were calculated. MCNP calculations were validated by measuring the neutron spectra of a 239PuBe neutron source inserted in a H2O cylindrical moderator. The measurements were carried out with a multisphere neutron spectrometer with a 6LiI(Eu) scintillator. From the measurements the neutron spectrum was unfolded using the BUNKIUT code and the UTA4 response matrix. Some of the moderators with the source produce a neutron spectrum close to spectra found in actual applications, then can be used during the calibration of radiation protection devices

  14. Neutronic study of fusion reactor blanket

    International Nuclear Information System (INIS)

    The problem of effective regeneration is a crucial issue for the fusion reactor, specially for the power reactor because of the conflicting requirements of heat removal and tritium breeding. For that, calculations are performed to evaluate blanket materials. Precise techniques are herein developed to improve the accuracy of the tritium production and the neutron and gamma transport calculations. Many configurations are studied with realistic breeder, structure, and coolant proportions. Accuracy of the results are evaluated from the sensitivity theory and uncertainty study using covariance matricies. At the end of this work, we presented the needs of nuclear data for fusion reactors and we give some advices for improving our knowledge of these data

  15. 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)

  16. Neutronic study of fusion reactor blanket

    International Nuclear Information System (INIS)

    The problem of effective regeneration is a crucial issue for the fusion reactor, specially for the power reactor because of the conflicting requirements of heat removal and tritium breeding. For that, calculations are performed to evaluate blanket materials. Precise techniques are herein developed to improve the accuracy of the tritium production and the neutron and gamma transport calculations. Many configurations are studied with realistic breeder, structure, and coolant proportions. Accuracy of the results are evaluated from the sensitivity theory and uncertainty study using covariance matrices. At the end of this work, we presented the needs of nuclear data for fusion reactors and we give some advices for improving our knowledge of these data

  17. 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)

  18. The neutron radiography facility designed for TRIGA reactors and its results

    International Nuclear Information System (INIS)

    The two TRIGA reactors of INR, the Steady State Reactor (SSR) having a power of 14 MW and Annular Core Pulsing Reactor (ACPR) having in steady state a power of 500 kW and being capable of a pulse to the peak power of 20000 MW, are placed in the same pool. The neutron flux ranging at the edges of those reactors cores is suitable for neutron radiography. The neutron radiography facility is placed in the pool of the TRIGA reactors. Till now as neutron source only the ACPR, in steady state or pulsing mode has been used. For the future one intends to use also the neutron flux of SSR. The aim of this facility is to achieve neutron radiographs of the nuclear fuel elements. (authors)

  19. Neutron measurements in MASURCA reactor: preliminary experience

    International Nuclear Information System (INIS)

    In the context of the studies on ADS, an experimental program is in progress in collaboration with the CEA team of the small nuclear reactor MASURCA at Cadarache. We report the first test of neutron flux measurement and the status of the program. (authors)

  20. Fast neutron flux in heavy water reactors

    International Nuclear Information System (INIS)

    The possibility of calculating the fast neutron flux in a natural uranium-heavy water lattice by superposition of the individual contributions of the different fuel elements was verified using a one-dimension Monte-Carlo code. The results obtained are in good agreement with experimental measurements done in the core and reflector of the reactor AQUILON. (author)

  1. 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

  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. Measurement of thermal neutron fluence rate of in-hospital neutron irradiator by SSNTD

    International Nuclear Information System (INIS)

    In-hospital neutron irradiator (IHNI) is an especially designed nuclear device based on Miniature Neutron Source Reactor (MNSR) for boron neutron capture therapy (BNCT). Its rated power is 30 kW. There are a thermal neutron beam and an epithermal neutron beam for treating patients at the opposite of the core. From the thermal neutron beam, a test beam is fetched out for measurement of boron concentration in blood by prompt γ neutron activation analysis (PGNAA) method. The neutron fluence rates at the end of thermal, epithermal and test neutron beam were measured by 235U fissile target and mica slice detector. At rated power, they are 1.67 × 109, 2.44 × 107 and 3.03 × 106 cm-2 · s-1, respectively. The results show that the thermal and epithermal neutron fluence rate can meet the requirement of BNCT and test neutron fluence rate meets the requirement of PGNAA. (authors)

  4. Development of Systems for Cold Neutron Source

    International Nuclear Information System (INIS)

    The design technology of CNS(Cold Neutron Source) facility system is a high technology which only a few advanced countries possess and is considered as a core technology in this particular situation that we are trying to move into higher level among nuclear energy countries. Especially, the very low temperature control and the vacuum control technology will be the basic important technique in high-tech field and furthermore, this will raise up the national power with the core neutron dispersion research center in the Northeast Asia. This original design technique will contribute to generate new other original technology through the fusion with RT, NT and BT, and improve the export competitiveness of the research reactor

  5. Pulsed Neutron Sources from Low Energy Proton Beams

    International Nuclear Information System (INIS)

    The efficiency with which neutrons may be produced using (p,n) reactions in Be and Li is substantially less than that of spallation. Only about 1 neutron for every 100 or more protons for these reactions in contrast to 10’s of neutrons per proton in the case of spallation. Nevertheless, the large currents available from linear accelerators with energies in the range from 3 to 30MeV allow the construction of a pulsed neutron source with reasonable flux based on these reactions because of their low threshold energies. At least one line of commercial neutron sources is presently being marketed for use in radiography medical applications and various research applications using these reactions. These sources provide neutrons at rates up to 1x1013 n/s and couple the source to a simple room temperature moderator. At Indiana University we are taking this concept slightly further in constructing the Low Energy Neutron Source (LENS) to provide neutrons at rates up to 1x1014 n/s and combining the source with a cryogenic moderator. LENS is designed to be a very flexible facility fulfilling three missions: to provide a rich educational environment for students to learn the details of neutron techniques, to develop new types of neutron instrumentation, and to conduct materials research using neutrons. The source will have a variable pulse structure (from as short as 5 μsec to as long as 1.2msec) and variable frequency (up to 100 Hz when using shorter pulses). We envision that sources such as LENS will provide a viable model for constructing networks of small sources that can support the major new spallation sources under construction in the USA and Japan in a manner similar to the support that national reactor sources presently provide for the ILL and ISIS in Europe. In this sense, LENS will serve as a prototype for the type of source this meeting was convened to discuss

  6. Design and safety aspects of the Cornell cold neutron source

    International Nuclear Information System (INIS)

    The cold neutron beam facility at the Cornell University TRIGA Mark II reactor will begin operational testing in early 1993. It is designed to provide a low background subthermal neutron beam that is as free as possible of fast neutrons and gamma rays for applied research and graduate-level instruction. The Cornell cold neutron source differs from the more conventional types of cold sources in that it is inherently safer because it uses a safe handling material (mesitylene) as the moderator instead of hydrogen or methane, avoids the circulation of cryogenic fluids by removing heat from the system by conduction through a 99.99% pure copper rod attached to a cryogenic refrigerator, and is much smaller in its size and loads. The design details and potential hazards are described, where it is concluded that no credible accident involving the cold source could cause damage to the reactor or personnel, or cause release of radioactivity. (author)

  7. 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.

  8. Automated Microdosing System for Integration With a Miniaturized High-pressure Reactor System

    OpenAIRE

    Kerstin Thurow; Norbert Stoll; Ihsan Hawali

    2005-01-01

    We present a new automated dosing system developed by the Institute for Automation of the University of Rostock, Germany. The new system is designed for the dosing of chemical liquids in the range of 50 μL–2.5 mL. It is integrated into a miniaturized reactor system to be used in the field of combinatorial synthesis. The reactor system can be pressurized up to 150 bar and tempered up to 200∘C. A wide range of liquids with different physical properties can be handled with the new d...

  9. Linac-driven spallation-neutron source

    International Nuclear Information System (INIS)

    Strong interest has arisen in accelerator-driven spallation-neutron sources that surpass existing facilities (such as ISIS at Rutherford or LANSCE at Los Alamos) by more than an order of magnitude in beam power delivered to the spallation target. The approach chosen by Los Alamos (as well as the European Spallation Source) provides the full beam energy by acceleration in a linac as opposed to primary acceleration in a synchrotron or other circular device. Two modes of neutron production are visualized for the source. A short-pulse mode produces 1 MW of beam power (at 60 pps) in pulses, of length less than 1 ms, by compression of the linac macropulse through multi-turn injection in an accumulator ring. A long-pulse mode produces a similar beam power with 1-ms-long pulses directly applied to a target. This latter mode rivals the performance of existing reactor facilities to very low neutron energies. Combination with the short-pulse mode addresses virtually all applications

  10. 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.

  11. Advanced neutron source three-element-core fuel grading

    International Nuclear Information System (INIS)

    The proposed advanced neutron source (ANS) neutron research facility's purpose is to provide unprecedented experimental capabilities in the areas of neutron scattering, materials research, and isotope production. The primary goals of the ANS project are to obtain neutron flux levels that are 5 to 10 times larger than any current existing facility and to provide isotope irradiation facilities that are at least as good as the High-Flux Isotope Reactor at Oak Ridge National Laboratory. The design changes in the ANS are described

  12. 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

  13. 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

  14. Pulsed neutron source very intense, Booster

    International Nuclear Information System (INIS)

    A compact Accelerator-Booster (fast, pulsed and modulate reactivity research reactor) is a new and appropriate conception to use as a very intense thermal neutrons source. Its definition and feasibility have been already described in several studies showing its relative advantages in comparison with others kinds of facilities. This work, wich is part of one of those studies, contains a general analysis on the meis facility parameters and core and shielding theoretical calculations. The following results were obtained: Selection and test of a calculation system suitable to use in compact fast reactors; Development a method to perform estimations in some safety and shielding problems and obtainment of adequate theoretical predictions on the general performance. Moreover, final results for importent parameters of the feasibility study and predesign (critical mass and volume, lifetime, etc.) and others related to the use of plutonium oxide as fuel are given and then evaluations of different basic functions are showed. (author)

  15. 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.

  16. Fuel cycle for a fusion neutron source

    International Nuclear Information System (INIS)

    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

  17. 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.

  18. 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

  19. Activation analysis with reactor neutrons

    International Nuclear Information System (INIS)

    The potentialities of neutron as an analytical probe are indicated, pointing out the need for development of other approaches, besides the conventional activation method. Development of instrumental approach to activation and applications, carried out at Analytical Chemistry Division are outlined. The role of, and the need for, the development and application of mathematical methods in enhancing the information content, and in turn the interpretation of the analytical results, is demonstrated. (author)

  20. Development of Cold Neutron Activation Station at HANARO Cold Neutron Source

    International Nuclear Information System (INIS)

    A new cold neutron source at the HANARO Research Reactor had been constructed in the framework of a five-year project, and ended in 2009. It has seven neutron guides, among which five guides were already allocated for a number of neutron scattering instruments. A new two-year project to develop a Cold Neutron Activation Station (CONAS) was carried out at the two neutron guides since May 2010, which was supported by the program of the Ministry of Education, Science and Technology, Korea. Fig. 1 shows the location of CONAS. CONAS is a complex facility including several radioanalytical instruments utilizing neutron capture reaction to analyze elements in a sample. It was designed to include three instruments like a CN-PGAA (Cold Neutron - Prompt Gamma Activation Analysis), a CN-NIPS (Cold Neutron - Neutron Induced Pair Spectrometer), and a CN-NDP (Cold Neutron - Neutron-induced prompt charged particle Depth Profiling). Fig. 2 shows the conceptual configuration of the CONAS concrete bioshield and the instruments. CN-PGAA and CN-NIPS measure the gamma-rays promptly emitted from the sample after neutron capture, whereas CN-NDP is a probe to measure the charged particles emitted from the sample surface after neutron capture. For this, we constructed two cold neutron guides called CG1 and CG2B guides from the CNS

  1. 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.)

  2. 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

  3. Sensibility studies of the equivalent thermal neutron flux on the heat exchanger of a sodium cooled fast reactor. (1. Pt.)

    International Nuclear Information System (INIS)

    This paper reports on sensibility studies of the equivalent thermal neutron flux on the heat exchanger for a sodium cooled fast reactor. Graphs and diagrams of the neutron flux in function of the reactor geometry, contribution of the fission sources in the core and the blanket of the reactor are given

  4. 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...

  5. IBR-2 - pulsed reactor for neutron investigations

    International Nuclear Information System (INIS)

    A brief theory and design are presented of IBR-2 fast neutron pulse reactor (with a periodic operation) of an average power of 3 MW constructed in Dubna and intended for investigation of structure and dynamics of liquids and solids in nuclear and neutron physics. The core of the reactor has the volume of 22 l and is filled with a liquid sodium. Fuel elements are made of sintered plutonium dioxide tablets and placed into a stainless steel cylinder 8.6 mm in diameter with wall thickness of 0.45 mm. The height of the fuel element active part is 445 mm, and the total length of the fuel element is about 780 mm. Fuel elements are separated from each other by a wire 0.5 mm in diameter wound around each element like a spiral. Fuel elements are assembled in cassets (their total number is 78), 7 elements in each casset. To achieve periodic operation and produce power pulses to be applied to the reactor two movable reflectors are used. The booster-type operation is assumed with using line induction accelerator. The peak density of thermal neutron flux in the reactor is 1016 cm-2s-1, for power pulse repetition rate 5 Hz and duration approximately μs. The scope of investigations to be performed on IBR-2 using the flight-of-time spectrometry method is also considered

  6. 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。

  7. 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

  8. 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.

  9. 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)

  10. The Advanced Neutron Source Facility: A new user facility for neutron research

    International Nuclear Information System (INIS)

    The Advanced Neutron Source (ANS) is a new reactor-based research facility being planned by Oak Ridge National Laboratory (ORNL) to meet the need for an intense steady state source of neutrons and for associated research space and equipment. The ANS will be open for use by scientists from universities, industry, and other federal laboratories. The ANS will be built around a new research reactor of unprecedented flux; that is, it will produce the most intense continuous beams of neutrons in the world. The goal is to reach a thermal neutron flux for beam experiments of 5 /times/ 1019 to 10 /times/ 1019 neutrons/(m2/center dot/s/sup /minus/1/). By combining the higher source flux with improved experimental facilities, the ANS will surpass current US high flux reactors---the High Flux Isotope Reactor (HFIR) at ORNL and the High Flux Beam Reactor (HFBR) at Brookhaven National Laboratory---by a factor of 10 to 20. The safety analysis of the ANS facility will include a complete probabilistic risk assessment (PRA), which will provide a systematic assessment of dependencies among systems at the malfunctions. For the current generation of nuclear power plants that have recently undergone the licensing review process, PRA has been used an an analysis tool after completion of the plant designs. For the ANS Project, the PRA effort has already begun, before the facility conceptual design. This allows safety insights from the PRA to be incorporated into the evolving plant design. 4 refs., 6 figs

  11. Fission-Fusion Neutron Source Progress Report July 31, 2009

    Energy Technology Data Exchange (ETDEWEB)

    Chapline, G; Daffin, F; Clarke, R

    2010-02-19

    In this report the authors describe progress in evaluating the feasibility of a novel concept for producing intense pulses of 14 MeV neutrons using the DT fusion reaction. In this new scheme the heating of the DT is accomplished using fission fragments rather than ion beams as in conventional magnet fusion schemes or lasers in ICF schemes. This has the great advantage that there is no need for any large auxiliary power source. The scheme does require large magnetic fields, but generating these fields, e.g. with superconducting magnets, requires only a modest power source. As a source of fission fragments they propose using a dusty reactor concept introduced some time ago by one of us (RC). The version of the dusty reactor that they propose using for our neutron source would operate as a thermal neutron reactor and use highly enriched uranium in the form of micron sized pellets of UC. Our scheme for using the fission fragments to produce intense pulses of 14 MeV neutrons is based on the fission fragment rocket idea. In the fission fragment rocket scheme it was contemplated that the fission fragments produced in a low density reactor core would then be guided out of the reactor by large magnetic fields. A simple version of this idea would be to use the fission fragments escaping from one side of a tandem magnet mirror to heat DT gas confined in the adjacent magnetic trap.

  12. Californium-252 Neutron Sources for Medical Applications

    International Nuclear Information System (INIS)

    Californium-252 neutron sources are being prepared to investigate the value of this radionuclide in diagnosing and treating diseases. A source resembling a cell-loaded radium needle was developed for neutron therapy. Since therapy needles are normally implanted in the body, very conservative design criteria were established to prevent leakage of radioactive. Methods are being developed to prepare very intense californium sources that could be used eventually for neutron radiography and for diagnosis by neutron activation analysis. This paper discusses these methods

  13. Design of a medical reactor generating high quality neutron beams for boron neutron capture therapy

    International Nuclear Information System (INIS)

    Boron neutron capture therapy (BNCT) is a binary treatment modality that can selectively irradiate tumor tissue. BNCT uses drugs containing a stable isotope of boron, B-10, that are capable of preferentially accumulating in the tumor, which is then irradiated with thermal neutrons. The interaction of the B-10 with a thermal neutron causes the B-10 nucleus to split, releasing an alpha particle and a lithium nucleus. These products of the boron neutron capture reaction are very damaging to cells but have a path length in tissue of approximately 14 micrometers, or roughly the diameter of one or two cells. Thus, most of the ionizing energy imparted to tissue is localized to B-10-loaded cells. Since the early 1980s, there have been considerable improvements in boron compounds and neutron beams. More is known now about the radiation biology of BNCT, which has reemerged as a potentially useful method for preferential irradiation of tumors. Clinical trials have been initiated at BNL and MIT, with an improved boron compound and epithermal neutrons. At this time, nuclear reactors are the only demonstrated satisfactory sources of epithermal neutrons. While some reactors are available and within reach of cancer treatment centers, a question arises as to the feasibility and practicality of placing new epithermal neutron sources in hospitals. In this thesis, we design a square reactor (that can easily be reconfigured into polygonal reactors as the need arises) with four slab type assemblies to produce two epithermal neutron beams and two thermal neutron beams for use in neutron capture therapy. This square reactor with four large-area faces consists of 1056 U3Si-Al fuel elements and 36 B4C control rods. The proposed facility, based on this square reactor core with a maximum operating power of 300kW, provides an epithermal neutron beam of 3.2x109 nepi/cm2 · s intensity with low contamination by fast neutrons (<1.6x10-13 Gy · cm2/nepi) and gamma rays (<1.0x10-13 Gy · cm2/nepi

  14. Outline of spallation neutron source engineering

    International Nuclear Information System (INIS)

    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)

  15. 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)

  16. 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.

  17. Fusion reactor blanket: neutronic studies in France

    International Nuclear Information System (INIS)

    The problem of effective tritium regeneration is a crucial issue for the fusion reactor, especially for the power reactor because of the conflicting requirements of heat removal and tritium breeding. For that, calculations are performed to evaluate blanket materials. Precise techniques are herein developed to improve the accuracy of the tritium production and the neutron and gamma transport calculations. Many configurations are studied with realistic breeder, structure, and coolant proportions. Accuracy of the results are evaluated from the sensitivity theory and uncertainty analysis. The results of these studies permit us to conclude that it is possible to expect an adequate tritium breeding ratio

  18. 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)

  19. 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

  20. 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.

  1. Designs and Experiments for Studies of Fast Neutron Fields at the RB Reactor

    International Nuclear Information System (INIS)

    The RB reactor is a heavy water critical assembly that has been in operation since 1958 using, at different times, natural metal uranium, 2% enriched metal uranium, and 80% enriched aluminium dioxide fuel of Soviet origin. A feasibility study of the RB reactor as a fast neutron source began in 1976, and four versions of fast neutron fields around or in the reactor were designed through 1990: an external neutron converter (ENC) in 1976; an experimental fuel channel (EPC) in 1982, an internal neutron converter (lNC) in 1983, and a coupled fast-thermal core (HERBE) in 1990. This paper presents an overview of the characteristics and experimental applications of each particular fast neutron field mentioned above, including available irradiation space, neutron spectra, and equivalent neutron and gamma dose rates. Control and safety-related implications of these modifications are emphasized. The computer codes and nuclear data libraries used in calculations are described briefly. (author)

  2. A National Spallation Neutron Source for neutron scattering

    International Nuclear Information System (INIS)

    The National Spallation Neutron Source is a collaborative project or perform the conceptual design for a next generation neutron source for the Department of Energy. This paper reviews the need and justification for a new neutron source, the origins and structure of the collaboration formed to address this need, and the community input leading up to the current design approach. A reference design is presented for an accelerator based spallation neutron source that would begin operation at about 1 megawatt of power but designed so that it could be upgraded to significantly higher powers in the future. The technology approach, status, and progress on the conceptual design to date are presented

  3. Neutron excitation function guide for reactor dosimetry

    International Nuclear Information System (INIS)

    Neutron Excitation Function Guide for Reactor Dosimetry (NEFGRD) has been prepared in the Ukrainian Nuclear Data Center (UKRNDC) using ZVV 9.2 code for graphical data presentation. The data can be retrieved through Web or obtained on CD-ROM or as hard copy report. NEFGRD contains graphical and text information for 56 nuclides (81 dosimetry reactions). Each reaction is provided by the information part and several graphical function blocks (from one to nine). (author)

  4. Utilization of the Dalat Research Reactor for Radioisotope Production, Neutron Activation Analysis, Research and Training

    International Nuclear Information System (INIS)

    The Dalat Nuclear Research Reactor (DNRR) is a 500 kW pool type reactor loaded with a mixed core of HEU (36% enrichment) and LEU (19.75% enrichment) fuel assemblies. The reactor is used as a neutron source for the purposes of radioisotopes production, neutron activation analysis, basic and applied research and training. The reactor is operated mainly in continuous runs of 108 hours for cycles of 3–4 weeks for the above mentioned purposes. The current status of safety, operation and utilization of the reactor is given and some aspects for improvement of commercial products and services of the DNRR are also discussed in this paper. (author)

  5. 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)

  6. 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

  7. 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.

  8. 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)

  9. 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.

  10. Fission-Fusion Neutron Source Progress Report Sept 30, 2009

    Energy Technology Data Exchange (ETDEWEB)

    Chapline, G F; Daffin, F; Clark, R

    2010-02-19

    In this report the authors describe the progress made in FY09 in evaluating the feasibility of a new concept for using the DT fusion reaction to produce intense pulses of 14 MeV neutrons. In this new scheme the heating of the DT is accomplished using fission fragments rather than ion beams as in conventional magnet confinement fusion schemes or lasers in inertial confinement schemes. As a source of fission fragments they propose using a dust reactor concept introduced some time ago by one of us (RC). An attractive feature of this approach is that there is no need for a large auxiliary power source to heat the DT plasma to the point where self-sustaining fusion become possible. Their scheme does require pulsed magnetic fields, but generating these fields requires only a modest power source. The dust reactor that they propose using for their neutron source would use micron-sized UC pellets suspended in a vacuum as the reactor fuel. Surrounding the fuel with a moderator such as heavy water (D{sub 2}O) would allow the reactor to operate as a thermal reactor and require only modest amounts of HEU. The scheme for using fission fragments to generate intense pulses of 14 MeV neutrons is based on the fission fragment rocket idea. In the fission fragment rocket scheme it was contemplated that the fission fragments produced in a low density reactor core could be guided out of the reactor by large magnetic fields used to form a 'rocket exhaust'. Their adaptation of this idea for the purposes of making a neutron source involves using the fission fragments escaping from one side of a tandem magnet mirror to heat DT gas confined in the adjacent magnetic trap.

  11. Fission-Fusion Neutron Source Progress Report Sept 30, 2009

    International Nuclear Information System (INIS)

    In this report the authors describe the progress made in FY09 in evaluating the feasibility of a new concept for using the DT fusion reaction to produce intense pulses of 14 MeV neutrons. In this new scheme the heating of the DT is accomplished using fission fragments rather than ion beams as in conventional magnet confinement fusion schemes or lasers in inertial confinement schemes. As a source of fission fragments they propose using a dust reactor concept introduced some time ago by one of us (RC). An attractive feature of this approach is that there is no need for a large auxiliary power source to heat the DT plasma to the point where self-sustaining fusion become possible. Their scheme does require pulsed magnetic fields, but generating these fields requires only a modest power source. The dust reactor that they propose using for their neutron source would use micron-sized UC pellets suspended in a vacuum as the reactor fuel. Surrounding the fuel with a moderator such as heavy water (D2O) would allow the reactor to operate as a thermal reactor and require only modest amounts of HEU. The scheme for using fission fragments to generate intense pulses of 14 MeV neutrons is based on the fission fragment rocket idea. In the fission fragment rocket scheme it was contemplated that the fission fragments produced in a low density reactor core could be guided out of the reactor by large magnetic fields used to form a 'rocket exhaust'. Their adaptation of this idea for the purposes of making a neutron source involves using the fission fragments escaping from one side of a tandem magnet mirror to heat DT gas confined in the adjacent magnetic trap.

  12. Calculations on neutron irradiation damage in reactor materials

    International Nuclear Information System (INIS)

    Neutron irradiation damage calculations were made for Mo, Nb, V, Fe, Ni and Cr. Firstly, damage functions were calculated as a function of neutron energy with neutron cross sections of elastic and inelastic scatterings, and (n,2n) and (n,γ) reactions filed in ENDF/B-III. Secondly, displacement damage expressed in displacements per atom (DPA) was estimated for neutron environments such as fission spectrum, thermal neutron reactor (JMTR), fast breeder reactor (MONJU) and two fusion reactors (The Conceptual Design of Fusion Reactor in JAERI and ORNL-Benchmark). then, damage cross section in units of dpa. barn was defined as a factor to convert a given neutron fluence to the DPA value, and was calculated for the materials in the above neutron environments. Finally, production rates of helium and hydrogen atoms were calculated with (n,α) and (n,p) cross sections in ENDF/B-III for the materials irradiated in the above reactors. (auth.)

  13. Polarized neutron reflectometry at the IBR-2 pulsed reactor

    Science.gov (United States)

    Aksenov, V. L.; Nikitenko, Yu. V.

    2007-05-01

    Polarized neutron reflectometry as a method for investigating layered nanostructures and its implementation at the IBR-2 pulsed reactor (Dubna, Russia) are described. The experimental data illustrating the studies of magnetic layered nanostructures and the development of the method of polarized neutron reflectometry on the polarized neutron spectrometer are presented. The directions of further development of the method of polarized neutron reflectometry are analyzed.

  14. Neutron multiplier alternative for fusion reactor blankets

    International Nuclear Information System (INIS)

    A proposal is given to replace neutron multiplier needed to enable low lithium and tritium inventories simultaneously assuring sufficient production of tritium, by an efficient moderator (7LiH or 7LiD). The advantageous effect of the intensified neutron energy degradation is due to the 1/v character of the main tritium producing reaction. The slowing-down medium is designed to be the source of moderated neutrons for the surrounding Li (6Li enriched) region where the most of tritium is to be produced. The surplus tritium production remains stored in the moderator zone. Some preliminary calculations illustrating the above concept were carried out and the neutron flux and tritium production distributions are presented. The indications regarding further studies are also suggested. (author)

  15. 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

  16. 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.

  17. 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.

  18. 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.)

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

    OpenAIRE

    Lee, Seung Kyu; Kang, Byoung-Hwi; 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 spe...

  20. 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.

  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. The spallation neutron source SINQ and related dosimetry problems

    International Nuclear Information System (INIS)

    The spallation neutron source SINQ, presently under construction at Switzerland's Paul Scherrer Institut, will handle the highest proton current of any comparable facility in the world: a continuous beam of 1.5 mA, 590 MeV protons from an isochronous ring cyclotron. For the users, SINQ as a neutron source should resemble closely a medium flux research reactor; the presence of high and medium energy particles creates new technical problems in design and operation. The engineering design for the major components is based on the results of neutronic calculations, using a code package built around the HETC program. At present, opportunities to verify the theoretical calculations experimentally are very limited. Safety factors have to be built in which conflict with the optimization of SINQ as a neutron source. To benchmark the calculational methods, a wide ranging diagnostic system will be required

  3. A mobile D-T neutron source for neutron radiography

    International Nuclear Information System (INIS)

    There has been an increasing need for a reliable and high flux and monoenergetic neutron source facility for radiographic applications both in basic research and industry. The neutron generator based on D-T reaction is a prolific source of 14 MeV neutrons which can be suitably moderated for providing a collimated beam of thermal neutrons. The main features of the D-T generator incorporating major changes in size reduction for converting it into a mobile unit is discussed. Structural details regarding ion source, accelerator design and tritium target system is highlighted. A built-in deuterium gas supply unit provides uninterrupted deuteron beam for on-line measurements. A neutron yield of 10E12 n/sec would ensure that thermal neutron radiography as well as activation analysis could be considered. The salient features of the different subsystems and their design as well as operational characteristics are presented. (author)

  4. 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.

  5. 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

  6. 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)

  7. Fast Reactor Physics Parameters from a Pulsed Source

    International Nuclear Information System (INIS)

    One of the more important integral parameters in fast reactor physics analysis is the neutron spectrum of a particular composition reactor core. Various methods, such as proton recoil counters and nuclear emulsion analysis, have been used to study fast reactor spectra. With the development of high intensity short-duration pulsed neutron sources, the time-of-flight technique has become suitable for fast reactor spectrum determination. To evaluate the feasibility of measuring fast neutron spectra from a core using time-of-flight techniques, an experiment has been performed to measure the equilibrium spectmm in a large block of depleted uranium using the General Atomics Linac facilities. A ten-metric-ton block of depleted uranium was assembled to form a 81-cm cube. This block of uranium was pulsed by electron bombardment of a uranium target imbedded in the block. The spectra from various sections of the block were measured using time-of-flight techniques for a 50-m flight path. Spectral indices, such as the ratio of the fission rates of U238/U235, U233/U235, U234/U235, Np237/U235, Pu239/U235 were also measured. In addition, measurements of the U238 capture rates were obtained in various parts of the block. This paper describes the techniques used to obtain these reactor physics parameters. The experimental results such as the spectra and spectral indices are also compared with those obtained from theoretical considerations using multigroup transport theory analysis. The pulsed neutron technique is also applicable for the measurement of such parameters as: β/ℓ, where β is the effective delayed neutron fraction and ℓ is the lifetime; neutron importance; and keff. This paper concludes with a discussion on the proposed application of a pulsed neutron source for the measurement of some of these parameters on fast reactor cores constructed on ZPR-VI, the Argonne Fast Critical Facility. (author)

  8. Exploratory studies on neutron radiography with a small neutron source using a nuclear scintillation imaging technique

    International Nuclear Information System (INIS)

    Neutron radiography based on mobile neutron sources need optimum utilization of available neutron fluxes which are usually lower compared to those available from reactors. For optimum utilization of such low flux devices, a sensitive neutron imaging technique is required. Such a neutron imaging system based on a Li6F-ZnS scintillator screen has been developed using a pair of image intensifier tubes and a charge coupled device. This detector system has been employed to study the feasibility of neutron radiography using low neutron fluences. The main feature of this imaging system is its ability to detect individual neutron scintillation events with a higher degree of spatial resolution. In order to test the efficiency of this imaging system, a small scale moderator-collimator assembly was designed using a Pu-Be neutron source of strength ∼2.107 n/s. Details of this imaging system and results of some exploratory experiments for low fluence neutron imaging are presented in this paper. (orig.)

  9. Neutronic study of the two french heavy water reactors

    International Nuclear Information System (INIS)

    The two french reactors - the reactor of Chatillon, named Zoe, and the reactor of Saclay - P2 - were the object of detailed neutronic studies which the main ideas are exposed in this report. These studies were mostly done by the Department of the Reactor Studies (D.E.P.). We have thus studied the distribution of neutronic fluxes; the factors influencing reactivity; the link between reactivity and divergence with the formula of Nordheim; the mean time life of neutrons; neutron spectra s of P2; the xenon effect; or the effect of the different adjustments of the plates and controls bar. (M.B.)

  10. 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.)

  11. 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

  12. Compact, energy EFFICIENT neutron source: enabling technology for various applications

    Energy Technology Data Exchange (ETDEWEB)

    Hershcovitch, A.; Roser, T.

    2009-12-01

    A novel neutron source comprising of a deuterium beam (energy of about 100 KeV) injected into a tube filled with tritium gas and/or tritium plasma that generates D-T fusion reactions, whose products are 14.06 MeV neutrons and 3.52 MeV alpha particles, is described. At the opposite end of the tube, the energy of deuterium ions that did not interact is recovered. Beryllium walls of proper thickness can be utilized to absorb 14 MeV neutrons and release 2-3 low energy neutrons. Each ion source and tube forms a module. Larger systems can be formed from multiple units. Unlike currently proposed methods, where accelerator-based neutron sources are very expensive, large, and require large amounts of power for operation, this neutron source is compact, inexpensive, easy to test and to scale up. Among possible applications for this neutron source concept are sub-critical nuclear breeder reactors and transmutation of radioactive waste.

  13. 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.

  14. 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

  15. 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)

  16. Radiation sources generated by TRIGA - INR reactor operation

    International Nuclear Information System (INIS)

    The main radioisotopes occurring in TRIGA reactor and in its accessories and irradiation devices during reactor operation, that determine the radiation fields in the adjacent technological halls are presented. The source data covering, the period November 1979 to May 200, were gamma spectrometric analysis reports for the liquid radioactive waste as well as analysis reports of water, gas or refuse samples and filters for radioactive aerosols retained from installations and adjacent rooms. The main radiation sources inside the reactor building are: - fission products; - radioactive wastes; - from the reactor cooling water and water additions (intrinsic activation products); - activated products of corrosion leavings. These radiation sources are analyzed in details and their occurrence and strength interpreted as probes of reactor operation. For instance, occurrence of delayed neutrons in cooling systems indicates can failure

  17. 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.

  18. Noise Thermometer at the FRM II Hot Neutron Source

    International Nuclear Information System (INIS)

    The 20 MW research reactor FRM II operated by the Technische Universitaet Muenchen is equipped with a hot neutron source (HNS). The source is aimed to shift the well thermalized neutron spectrum in the heavy water moderator tank to higher energies as requested by the experimental users. The main component of the HNS is a solid graphite cylinder being heated by gamma radiation from the reactor core up to a temperature of about 2000 oC. The hot graphite cylinder is surrounded by a high-temperature insulation of carbon fiber, to achieve and maintain the high temperature. Due to the extremely harsh environment, the high temperature and the nuclear radiation, the temperature inside the graphite cylinder is measured by a purpose-built noise thermometer. It measures the white noise of an electrical resistor and determines the absolute temperature of the graphite cylinder. During nuclear commissioning of the hot neutron source, the temperature of the graphite cylinder was measured by the noise thermometer at several power steps of the reactor. The following relevant parameters of the HNS had been determined: the maximum temperature, the heating rate and the cooling rate after shut down of the reactor. The relative long time needed to reach the maximum temperature was used to measure the heat-up effect of the HNS. Since the nuclear start-up of the reactor the noise thermometer of the HNS is operated without significant problems. (author)

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

    International Nuclear Information System (INIS)

    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 this 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 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

  20. 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)

  1. Neutron Noise Analysis with Flash-Fourier Algorithm at the IBR-2M Reactor

    OpenAIRE

    Dima, Mihai O.; Pepelyshev, Yuri N.(Joint Institute for Nuclear Research, FLNP, Dubna, Moscow 141980, Russia); Lachin Tayibov

    2014-01-01

    Neutron noise spectra in nuclear reactors are a convolution of multiple effects. For the IBR-2M pulsed reactor (JINR, Dubna), one part of these is represented by the reactivities induced by the two moving auxiliary reflectors and another part of these by other sources that are moderately stable. The study of neutron noise involves, foremostly, knowing its frequency spectral distribution, hence Fourier transforms of the noise. Traditional methods compute the Fourier transform of the autocorrel...

  2. Neutron sources: Present practice and future potential

    International Nuclear Information System (INIS)

    The present capability and future potential of accelerator-based monoenergetic and white neutron sources are outlined in the context of fundamental and applied neutron-nuclear research. The neutron energy range extends from thermal to 500 MeV, and the time domain from steady-state to pico-second pulsed sources. Accelerator technology is summarized, including the production of intense light-ion, heavy-ion and electron beams. Target capabilities are discussed with attention to neutron-producing efficiency and power-handling capabilities. The status of underlying neutron-producing reactions is summarized. The present and future use of neutron sources in: fundamental neutron-nuclear research, nuclear data acquisition, materials damage studies, engineering tests, and biomedical applications are discussed. Emphasis is given to current status, near-term advances well within current technology, and to long-range projections. 90 refs., 4 figs

  3. Tajoura reactor core conversion neutrons analysis

    International Nuclear Information System (INIS)

    This paper presents the preliminary neutronics studies and results of the Tajoura reactor core conversion calculations from currently used highly enriched (80% U235) fuel to low enriched fuel (36% U''2''3''5) by using the TAJN computer package. The compact core loading consists of 16 fuel assemblies type IRT-2M surrounded by removable and stationary beryllium reflector and ordinary water for moderation and cooling. The study was undertaken to compare results of TAJN computer package and the vendor documented results. The results of these calculations at the BOL and EOL conditions with equilibrium Xe at 10 MWt are shown. (author)

  4. 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.

  5. 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.

  6. Accelerator based neutron source for neutron capture therapy

    International Nuclear Information System (INIS)

    Full text: The Budker Institute of Nuclear Physics (Novosibirsk) and the Institute of Physics and Power Engineering (Obninsk) have proposed an accelerator based neutron source for neutron capture and fast neutron therapy for hospital. Innovative approach is based upon vacuum insulation tandem accelerator (VITA) and near threshold 7Li(p,n)7Be neutron generation. Pilot accelerator based neutron source for neutron capture therapy is under construction now at the Budker Institute of Nuclear Physics, Novosibirsk, Russia. In the present report, the pilot facility design is presented and discussed. Design features of facility components are discussed. Results of experiments and simulations are presented. Complete experimental tests are planned by the end of the year 2005

  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. Quantitative evaluation of emission controls on primary and secondary organic aerosol sources during Beijing 2008 Olympics

    OpenAIRE

    Guo, S.; M. Hu; Guo, Q.; Zhang, X; Schauer, J. J.; Zhang, R

    2013-01-01

    To assess the primary and secondary sources of fine organic aerosols after the aggressive implementation of air pollution controls during the 2008 Beijing Olympic Games, 12 h PM2.5 values were measured at an urban site at Peking University (PKU) and an upwind rural site at Yufa during the CAREBEIJING-2008 (Campaigns of Air quality REsearch in BEIJING and surrounding region) summer field campaign. The average PM2.5 concentrations were 72.5 ± 43.6 μg m−3 and 64.3 ± 36.2 μg ...

  9. Neutron emissions from a lunar-based reactor: Excavation siting

    International Nuclear Information System (INIS)

    Lunar surface operations, such as deep space observatories, mineral extraction facilities, and oxygen production factories, will require large amounts of electrical power. Compact nuclear reactor power sources are possible for such applications where the demand exceeds 25 kW(electric) and will be required above 500 kW(electric). One scenario for the siting of a lunar reactor power system is to place the reactor at the bottom of an excavation and locate the power conversion and heat rejection systems on the lunar surface. The lunar soil provides the necessary radiation shielding for equipment and personnel, and this configuration minimizes the amount of shielding material that must be transported from Earth. Such a configuration, however, allows neutrons to be scattered around the primary instrument shields and to be emitted from the top of the hole. A parametric study of excavation configurations has been completed using MCNP. Variations of the excavation design include increasing the depth and the diameter of the excavation and the inclusion of neutron absorbing materials in the bulkheads, which support the reactor power system and restrain the lunar soil

  10. Intercalibration of physical neutron dosimetry for the RA-3 and MURR thermal neutron sources for BNCT small-animal research

    International Nuclear Information System (INIS)

    New thermal neutron irradiation facilities to perform cell and small-animal irradiations for Boron Neutron Capture Therapy research have been installed at the Missouri University Research Reactor and at the RA-3 research reactor facility in Buenos Aires, Argentina. Recognizing the importance of accurate and reproducible physical beam dosimetry as an essential tool for combination and intercomparisons of preclinical and clinical results from the different facilities, we have conducted an experimental intercalibration of the neutronic performance of the RA-3 and MURR thermal neutron sources.

  11. Intercalibration of physical neutron dosimetry for the RA-3 and MURR thermal neutron sources for BNCT small-animal research.

    Science.gov (United States)

    Pozzi, Emiliano C C; Thorp, Silvia; Brockman, John; Miller, Marcelo; Nigg, David W; Hawthorne, M Frederick

    2011-12-01

    New thermal neutron irradiation facilities to perform cell and small-animal irradiations for Boron Neutron Capture Therapy research have been installed at the Missouri University Research Reactor and at the RA-3 research reactor facility in Buenos Aires, Argentina. Recognizing the importance of accurate and reproducible physical beam dosimetry as an essential tool for combination and intercomparisons of preclinical and clinical results from the different facilities, we have conducted an experimental intercalibration of the neutronic performance of the RA-3 and MURR thermal neutron sources. PMID:21330143

  12. 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

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

    International Nuclear Information System (INIS)

    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

  14. 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.

  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. 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)

  17. 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.

  18. 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.

  19. Neutron scattering at the high-flux isotope reactor

    International Nuclear Information System (INIS)

    The title facilities offer the brightest source of neutrons in the national user program. Neutron scattering experiments probe the structure and dynamics of materials in unique and complementary ways as compared to x-ray scattering methods and provide fundamental data on materials of interest to solid state physicists, chemists, biologists, polymer scientists, colloid scientists, mineralogists, and metallurgists. Instrumentation at the High- Flux Isotope Reactor includes triple-axis spectrometers for inelastic scattering experiments, a single-crystal four diffractometer for crystal structural studies, a high-resolution powder diffractometer for nuclear and magnetic structure studies, a wide-angle diffractometer for dynamic powder studies and measurements of diffuse scattering in crystals, a small-angle neutron scattering (SANS) instrument used primarily to study structure-function relationships in polymers and biological macromolecules, a neutron reflectometer for studies of surface and thin-film structures, and residual stress instrumentation for determining macro- and micro-stresses in structural metals and ceramics. Research highlights of these areas will illustrate the current state of neutron science to study the physical properties of materials

  20. 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)

  1. 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)

  2. 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.

  3. 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.

  4. 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)

  5. Seminar on Heat-transfer fluids for fast neutron reactors

    International Nuclear Information System (INIS)

    This book reports the content of a two-day meeting held by the Academy of Sciences on the use of heat-transfer fluids in fast neutron reactors. After a first part which proposes an overview of scientific and technical problems related to these heat-transfer fluids (heat transfer process, nuclear properties, chemistry, materials, risks), a contribution proposes a return on experience on the use of heat-transfer fluids in the different design options of reactors of fourth generation: from mercury to NaK in the first fast neutron reactor projects, specific assets and constraints of sodium used as heat-transfer fluid, concepts of fast neutron reactors cooled by something else than sodium, perspectives for projects and research in fast neutron reactors. The next contribution discusses the specifications of future fast-neutron reactors: expectations for fourth-generation reactors, expectations in terms of performance and of safety, specific challenges. The last contribution addresses actions to be undertaken in the field of research and development: actions regarding all reactor types or specific types as sodium-cooled reactors, lead cooled reactors, molten salt reactors, and gas-cooled fast reactors

  6. Neutron beam applications using low power research reactor Malaysia perspectives

    International Nuclear Information System (INIS)

    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 area currently focussed on is the utilisation of neutron beam ports available at this 1MW reactor. Projects undertaken are the development and utilisation of the Neutron Radiography (myNR), Small Angle Neutron Scattering (mySANS) and Boron Neutron Capture Therapy (BNCT) - preliminary study. In order to implement active research programmes, a group comprised of researcher from research institutes and academic institutions, has formed: known as Malaysian Reactor Interest Group (MRIG). This paper describes the recent status the above neutron beam facilities and their application in industrial, health and material technology research and education. The related activities of MRIG are also highlighted. (author)

  7. 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

  8. Neutron flux and fluence determination for BWR reactors

    International Nuclear Information System (INIS)

    Measurements of gamma emission rates from Fe and Cu dosimeters extracted from a BWR type reactor vessel were carried out in order to determine their total activity. The dosimeter's activity is related to the neutron flux there by taking into account the reactor material's embrittlement caused by neutron bombardment. The dosimeters were taken out after the first reactor operation cycle. From gamma radioactivity measurements of these dosimeters, neutron flux and fluence were calculated. These parameters are used in the determination of shift and adjusted reference temperature values needed for the development of pressure-temperature curves used during reactor operation

  9. Neutron spectrum characterization at Pneumatic Fast Transfer System (PFTS) of Kamini reactor

    International Nuclear Information System (INIS)

    Kalpakkam mini (KAMINI) reactor is a research reactor operated at Indira Gandhi Centre for Atomic Research (IGCAR), Kalpakkam. Neutron spectrum is an important characteristic feature of a given neutron source. This paper discusses about the neutron spectrum characterization at PFTS position of KAMINI reactor at 20 kW power using multi-foil irradiation method. Several activation foils (Au, Ti, Ni, In, Mo, Fe, Co and Al foils and Y2O3 powder) were irradiated at PFTS position of KAMINI reactor and assayed by high resolution gamma spectrometry. Saturated activity per nuclide and nuclear reaction rates were calculated and used to unfold the neutron spectrum at the irradiation location based on Least Square Minimization approach. (author)

  10. Observation of Neutron Skyshine from an Accelerator Based Neutron Source

    Science.gov (United States)

    Franklyn, C. B.

    2011-12-01

    A key feature of neutron based interrogation systems is the need for adequate provision of shielding around the facility. Accelerator facilities adapted for fast neutron generation are not necessarily suitably equipped to ensure complete containment of the vast quantity of neutrons generated, typically >1011 nṡs-1. Simulating the neutron leakage from a facility is not a simple exercise since the energy and directional distribution can only be approximated. Although adequate horizontal, planar shielding provision is made for a neutron generator facility, it is sometimes the case that vertical shielding is minimized, due to structural and economic constraints. It is further justified by assuming the atmosphere above a facility functions as an adequate radiation shield. It has become apparent that multiple neutron scattering within the atmosphere can result in a measurable dose of neutrons reaching ground level some distance from a facility, an effect commonly known as skyshine. This paper describes a neutron detection system developed to monitor neutrons detected several hundred metres from a neutron source due to the effect of skyshine.

  11. Measurement of Neutron Tissue Dose Outside the Reactor Shielding

    International Nuclear Information System (INIS)

    Intermediate neutrons form an important part of the neutron-tissue dose outside the reactor shielding. Equipment developed in the RUS series makes it possible to measure the flux and the tissue dose rate of intermediate neutrons. In experiments on the 1RT-1000 reactor the neutron-dose composition was studied and it was shown that this depends greatly on the composition of the shielding. It was found that the neutron-tissue dose calculated from data obtained by means of RPN-1 apparatus is in reality too low by a factor of up to 1.5 for water shielding and 5 for concrete shielding. (author)

  12. Burnup influence on the VVER-1000 reactor vessel neutron fluence evaluation

    International Nuclear Information System (INIS)

    The neutron fluence of the vessels of the reactors is determined regularly accordingly the RPV Surveillance Program of the Kozloduy NPP Unit 5 and 6 in order to assess the state of the metal vessel and their radiation damaging. The calculations are carried out by the method of discrete ordinates used in the TORT program for operated reactor cycles. An average reactor spectrum corresponding to fresh U-235 fuel is used as an input neutron source. The impact of the burn up of the fuel on the neutron fluence of VVER-1000 reactor vessel is evaluated. The calculations of isotopic concentrations of U-235 and Pu-239 corresponding to 4 years burn up were performed by the module SAS2H of the code system SCALE 4.4. Since fresh fuel or 4 years burn up fuel assembly are placed in periphery of reactor core the contribution of Pu-239 of first year burn up and of 4 years burn up is taken in consideration. Calculations of neutron fluence were performed with neutron spectrum for fresh fuel, for 1 year and for 4 years burn up fuel. Correction factors for neutron fluence at the inner surface of the reactor vessel, in 1/4 depth of the vessel and in the air behind the vessel were obtained. The correction coefficient could be used when the neutron fluence is assessed so in verification when the measured activity of ex-vessel detectors is compared with calculated ones. (authors)

  13. Development of nuclear design criteria for neutron spallation sources

    International Nuclear Information System (INIS)

    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

  14. 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

  15. Control circuit for a pulsed neutron source

    International Nuclear Information System (INIS)

    A pulsed neutron source is operated with a control circuit which produces neutron pulses very sharply defined with reference to time. A relatively steep rising high voltage control pulse for the ion source is produced by means of a low voltage input control pulse. Simultaneously, a control pulse is generated for a delayed quenching circuit, which quenches the high voltage control pulse for the ion source after a fixed time delay for a short time. The control voltage obtained for the ion source and for the neutron output is sharply defined as regards time. (orig.)

  16. 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

  17. 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

  18. 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)

  19. 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)

  20. Emissions from slow pyrolysis of solid fuels studied by GC/MS in a miniature reactor

    Energy Technology Data Exchange (ETDEWEB)

    Pedersen, J.R.; Stroemberg, K.; Olsson, J.O. [Chalmers University of Technology, Gothenburg (Sweden). Dept. of Physical Chemistry

    1996-12-31

    In an effort to increase knowledge of slow pyrolysis processes of coal and biomass a miniature pyrolysis reactor suited for GC/MS analysis was designed. Three solid fuels were investigated: bituminous coal, sawdust from spruce and chips from eucalyptus. Coal pyrolysis produced comparatively high amounts of alkanes, naphthalene and polycyclic aromatic hydrocarbon not found in the wood samples. Coal also produced alkyl substituted phenols. Sawdust from spruce produced almost the same products as eucalyptus. However, the product distributions from spruce and eucalyptus were more substituted than these from spruce. The products from wood samples were dominated by oxygenates: fatty acids, cyclic oxygenates and methoxy substituted phenols. This approach can be used for selecting solid fuels suitable for clean combustion and production of bio-oils. 9 refs., 4 figs.

  1. Status of the FRM-II hot neutron source

    International Nuclear Information System (INIS)

    The new research reactor FRM-II will be equipped with a hot neutron source. This secondary source will shift a part of the thermal neutron energy spectrum in the D2O moderator to energies from 0.1 to 1 eV. The hot neutron source consists of a graphite cylinder (200 mm diameter, 300 mm high), which is heated by gamma radiation up to a maximum temperature of about 2400 C. The graphite cylinder is surrounded by a high-temperature insulation of carbon fiber, to achieve this high temperature. We have accomplished mock-up tests of the carbon fiber in a high temperature furnace, to investigate the insulation properties of the material. The graphite cylinder and the insulation are covered with two vessels made out of Zircaloy 4. The space between the vessels is filled with helium. The hot neutron source is permanent under control by pressure and temperature measurements. The temperature inside the graphite cylinder will be measured by a purpose-built noise thermometer due to the extremely harsh environment conditions (temperature and nuclear radiation). The hot neutron source is designed and manufactured according to the general specification basic safety and to the German nuclear atomic rules (KTA). The source will be installed in year 2001. (orig.)

  2. 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.)

  3. Rotating target neutron source II: progress report

    International Nuclear Information System (INIS)

    The RTNS-II Facility at Livermore was authorized in the FY76 ERDA budget. This facility will house two 4 x 1013 n/s sources of 14-MeV neutrons for materials damage experimentation. RTNS-II will be the first of DCTR's dedicated neutron source facilities. Initial operation is currently scheduled for March 1978. Engineering design of buildings and neutron sources started in March 1976 with construction scheduled to begin in August 1976. Design of the 150 mA D+ accelerators is based upon LLL experience with the MATS-III ion source and with the ICT accelerator of the RTNS-I source. Hardware design for the 50 cm, 5000 rpm tritium-in-titanium targets was guided by computer modeling of the target system now in use on RTNS-I. The final design of neutron sources and building layout will be discussed

  4. 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.

  5. 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.

  6. 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...

  7. 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)

  8. 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.)

  9. Fission fragment driven neutron source

    Science.gov (United States)

    Miller, Lowell G.; Young, Robert C.; Brugger, Robert M.

    1976-01-01

    Fissionable uranium formed into a foil is bombarded with thermal neutrons in the presence of deuterium-tritium gas. The resulting fission fragments impart energy to accelerate deuterium and tritium particles which in turn provide approximately 14 MeV neutrons by the reactions t(d,n).sup.4 He and d(t,n).sup.4 He.

  10. Fission-neutrons source with fast neutron-emission timing

    Science.gov (United States)

    Rusev, G.; Baramsai, B.; Bond, E. M.; Jandel, M.

    2016-05-01

    A neutron source with fast timing has been built to help with detector-response measurements. The source is based on the neutron emission from the spontaneous fission of 252Cf. The time is provided by registering the fission fragments in a layer of a thin scintillation film with a signal rise time of 1 ns. The scintillation light output is measured by two silicon photomultipliers with rise time of 0.5 ns. Overall time resolution of the source is 0.3 ns. Design of the source and test measurements using it are described. An example application of the source for determining the neutron/gamma pulse-shape discrimination by a stilbene crystal is given.

  11. 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.

  12. 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...

  13. 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

  14. 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

  15. 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)

  16. Extraction of Mechanical-Reactivity Influences from Neutron Noise Spectra at the IBR-2 Reactor

    Science.gov (United States)

    Dima, M.; Pepelyshev, Yu.

    2013-07-01

    Neutron noise spectra in nuclear reactors are a convolution of multiple-induced reactivities. For the IBR-2 pulsed reactor (JINR-Dubna), one part is represented by the reactivities induced by the two moving reflectors, and the other part by other sources that are moderately stable. In the present study, using recordings of the mechanical noise of the two moving reflectors, their non-linear correlations into the power spectra of the reactor are extracted using statistical analysis. The remaining noise sources are moderately stable noise and can be further monitored by other automated reactor diagnoses.

  17. The Reactor and Cold Neutron Research Facility at NIST

    International Nuclear Information System (INIS)

    The NIST Reactor (NBSR) and Cold Neutron Research Facility (CNRF) are located at the Gaithersburg, MD site, and have been in operation since 1969 and 1991, respectively. A total of 26 thermal neutron facilities and 11 cold neutron stations are operating for studies in condensed matter physics and chemistry, materials science, chemical analysis, nondestructive evaluation, neutron standards, fundamental neutron physics, and irradiations. Thermal and cold neutron instruments which have become operational since the 2d IGORR Conference will be described. Major facility upgrades to be implemented in early 1994 will be outlined. (author)

  18. 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)

  19. Neutron wall loading of Tokamak reactors

    International Nuclear Information System (INIS)

    Neutron wall loading (Γn) is a key parameter for the selection of fusion power core component materials. It also impacts the economic, performance, design, safety and environmental aspect of the fusion power plant. This paper reports the determination of the range of Γn for economically competitive fusion power plants based on the analysis that couples the MHD stability physics results to a system design code. Cost of electricity (COE) was selected as the parameter to be minimized. For both normal conducting and superconducting coil options, at thermal efficiency of 46% and at the power output range of 1-2 GW(e) the average neutron wall loading is 4-7 MW/m2. For a given power output, higher thermal efficiency will allow lower Γn. At the above range of Γn, in order to have economical fusion power reactors, for the solid first wall design option, high thermal efficiency of 46% to 57.5% requires the use of alloys like V and W-alloy, respectively. The corresponding COE can be projected to be in the economically competitive range of 62-54.6 mill/kWh

  20. 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)

  1. 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

  2. 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.

  3. 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

  4. Consumption and sources of dietary salt in family members in Beijing.

    Science.gov (United States)

    Zhao, Fang; Zhang, Puhong; Zhang, Lu; Niu, Wenyi; Gao, Jianmei; Lu, Lixin; Liu, Caixia; Gao, Xian

    2015-04-01

    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. PMID:25867952

  5. Source profiles of volatile organic compounds associated with solvent use in Beijing, China

    Science.gov (United States)

    Yuan, Bin; Shao, Min; Lu, Sihua; Wang, Bin

    2010-05-01

    Compositions of volatile organic compound (VOC) emissions from painting applications and printing processes were sampled and measured by gas chromatography-mass spectrometry/flame ionization detection (GC-MS/FID) in Beijing. Toluene and C8 aromatics were the most abundant species, accounting for 76% of the total VOCs emitted from paint applications. The major species in printing emissions included heavier alkanes and aromatics, such as n-nonane, n-decane, n-undecane, toluene, and m/p-xylene. Measurements of VOCs obtained from furniture paint emissions in 2003 and 2007 suggest a quick decline in benzene levels associated with formulation changes in furniture paints during these years. A comparison of VOC source profiles for painting and printing between Beijing and other parts of the world showed significant region-specific discrepancies, probably because of different market demands and environmental standards. We conducted the evaluation of the source reactivities for various VOC emission sources. The ozone formation potential (OFP) for unit mass of VOCs source emissions is the highest for paint applications. Substituting solvent-based paints by water-based in Beijing will lead to an OFP reduction of 152,000 tons per year, which is more than 1/4 of the OFPs for VOCs emissions from vehicle exhaust in the city.

  6. 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.

  7. [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

  8. High Fluence Neutron Source for Nondestructive Characterization of Nuclear Waste

    International Nuclear Information System (INIS)

    We are addressing the need to measure nuclear wastes, residues, and spent fuel in order to process these for final disposition. For example, TRU wastes destined for the WIPP must satisfy extensive characterization criteria outlined in the Waste Acceptance Criteria, the Quality Assurance Program Plan, and the Performance Demonstration Plan. Similar requirements exist for spent fuel and residues. At present, no nondestructive assay (NDA) instrumentation is capable of satisfying all of the PDP test cycles (particularly for Remote-Handled TRU waste). One of the primary methods for waste assay is by active neutron interrogation. We plan to improve the capability of all active neutron systems by providing a higher intensity neutron source (by about a factor of 1,000) for essentially the same cost, power, and space requirements as existing systems. This high intensity neutron source will be an electrostatically confined (IEC) plasma device. The IEC is a symmetric sphere that was originally developed in the 1950s as a possible fusion reactor. It operates as D-T neutron generator. Although it was not believed to scale to fusion reactor levels, these experiments demonstrated a neutron yield of 2 x 1010 neutrons/second on table-top experiments that could be powered from ordinary laboratory circuits (10 kilowatts). Subsequently, the IEC physics has been extensively studied at the University of Illinois and other locations. We have established theoretically the basis for scaling the output up to 1x1011 neutrons / second. In addition, IEC devices have run for cumulative times approaching 10,000 hours, which is essential for practical application to NDA. They have been operated in pulsed and continuous mode. The essential features of the IEC plasma neutron source, compared to existing sources of the same cost, size and power consumption, are: Table 1: Present and Target Operating Parameters for Small Neutron Generators Parameter Present IEC Target or Already Proven Neutron Yield

  9. 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)

  10. Analysis of municipal waste dump soil using a low power reactor. A study by instrumental neutron activation analysis

    Energy Technology Data Exchange (ETDEWEB)

    Waheed, S.; Siddique, N. [Pakistan Institute of Nuclear Science and Technology, Islamabad (Pakistan). Chemistry Div.; Hamid, Q. [Pakistan Institute of Engineering and Applied Sciences, Islamabad (Pakistan)

    2010-07-01

    Soil samples from an open waste dump site and a background sample from a site located beside Kashmir Highway near sector H-11, Islamabad, were analyzed for 36 major, minor and trace elements. Suitable instrumental neutron activation analysis (INAA) methodology was developed through optimization of irradiation, cooling and counting protocols using a 27 kW low power miniature neutron source reactor (MNSR). Large variations in the composition of soil samples taken from the dump site were observed and it was found that most of the elements including Al, Ca, Ti, As, Br, Ce, Co, Cs, Ga, Hf, Mn, Nd, La, Sb, Sc, Sm, Sn, Ta, Tb Th, V, Yb, Zn, Eu, Se, and Zr, in the background soil lie in the concentration range of the dump soil sample. However, Fe, K, Mg, Na, Ba, Rb, Sr, U, Zn, Hg and Sb were found to be in higher amounts in the dump site samples. To validate the developed INAA methodology and to ensure the accuracy and precision of the results, IAEA Lake Sediment (SL-1) and IAEA Soil-7 (S-7) were analyzed as control materials. (orig.)

  11. International workshop on plasma-based neutron sources

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1996-12-09

    The workshop was devoted to discussion of the status and future directions of work on plasma-based neutron sources. The workshop presentations demonstrated significant progress in development of the concepts of these sources and in broadening the required data base. Two main groups of neutron source designs were presented at the workshop: tokamak-based and mirror-based. Designs of the tokamak- based devices use the extensive data base generated during decades of tokamak research. Their plasma physics performance can be predicted with a high degree of confidence. On the other hand, they are relatively large and expensive, and best suited for Volumetric Neutron Sources (VNSes) or other large scale test facilities. They also have the advantage of being on the direct path to a power- producing reactor as presently conceived, although alternatives to the tokamak are presently receiving serious consideration for a reactor. The data base for the mirror-based group of plasma sources is less developed, but they are generally more flexible and, with appropriate selection of parameters, have the potential to be developed as compact Accelerated Test Facilities (ATFs) as well as full-scale VNSes. Also discussed at the workshop were some newly proposed but potentially promising concepts, like those based on the flow-through pinch and electrostatic ion-beam sources.

  12. 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)

  13. 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

  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 Yuanhang; TANG Xiaoyan

    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. Modeling a neutron rich nuclei source

    International Nuclear Information System (INIS)

    The deuteron break-up process in a suitable converter gives rise to intense neutron beams. A source of neutron rich nuclei based on the neutron induced fission can be realised using these beams. A theoretical optimization of such a facility as a function of the incident deuteron energy is reported. The model used to determine the fission products takes into account the excitation energy of the target nucleus and the evaporation of prompt neutrons. Results are presented in connection with a converter-target specific geometry. (authors)

  16. Modeling a neutron-rich nuclei source

    International Nuclear Information System (INIS)

    The deuteron break-up process in a suitable converter gives rise to intense neutron beams. A source of neutron-rich nuclei based on the neutron-induced fission can be realised using these beams. A theoretical optimization of such a facility as a function of the incident deuteron energy is reported. The model used to determine the fission products takes into account the excitation energy of the target nucleus and the evaporation of prompt neutrons. Results are presented in connection with a converter-target specific geometry. (orig.)

  17. HEXANN-EVALU, Neutron Irradiation of Reactor Pressure Vessels

    International Nuclear Information System (INIS)

    1 - Description of program or function: HEXANN-EVALU calculates the neutron irradiation of a pressure vessel surrounding a nuclear reactor core composed of hexagonal assemblies. The area outside the core may contain hexagonal shielding assemblies of non-multiplying materials, a core liner and annular material zones. 2 - Method of solution: The Monte Carlo method is used. The neutrons start at the core boundary with a given distribution in space, angle and energy. The angular distribution is calculated by HEXANN itself from data describing the spatial distribution of the neutron source in the core. Survival biasing is used in all collisions. To increase efficiency the following options are included: region-wise importance, Russian roulette, low energy Russian roulette, splitting, path stretching with explicit exponential transform and automatic importance correction. 3 - Restrictions on the complexity of the problem: 30-degree symmetry assumed; Maximum number of energy groups = 30; Maximum group number change in down-scattering = 19; No up-scattering; Maximum number of materials = 10; Maximum number of cylindrical surfaces = 20; Maximum number of source faces (hexagon faces at edge of core in 30-degree sector)= 20; Vacuum boundary conditions are used at the top, bottom and outer boundary

  18. Tajoura reactor conversion and preliminary neutronic experimental data

    International Nuclear Information System (INIS)

    Reactor LEU fuel conversion means that a certain procedures must be done according to this process. This paper presents the local experience of this work and focuses on the experimental data of neutron flux measurements along the fuel and reflector cells of Tajoura research reactor immediately followed the reactor building up critical mass. The study was done by applying the neutron activation technique with the irradiation of thin gold foils along the selected core cells and the results shows agreement with that publications based on the comparison between low and high enrichment fuels neutron fluxes. (author)

  19. Tajoura reactor conversion and preliminary neutronic experimental data

    Energy Technology Data Exchange (ETDEWEB)

    Aburwes, M.E.N. [Reactor Department, Nuclear measurements lab, Renewable Energies And Water Desalinaton Research Center, Tajoura (Libyan Arab Jamahiriya)

    2008-10-29

    Reactor LEU fuel conversion means that a certain procedures must be done according to this process. This paper presents the local experience of this work and focuses on the experimental data of neutron flux measurements along the fuel and reflector cells of Tajoura research reactor immediately followed the reactor building up critical mass. The study was done by applying the neutron activation technique with the irradiation of thin gold foils along the selected core cells and the results shows agreement with that publications based on the comparison between low and high enrichment fuels neutron fluxes. (author)

  20. Proposal of a wide-band mirror polarizer of slow neutrons at a pulsed neutron source

    International Nuclear Information System (INIS)

    The new type of wide-band mirror-based neutron polarizer, which is to be operated at a pulsed neutron source, is suggested. The idea is to use a movable polarizing mirror system, which, with the incoming beam monochromatized by the time-of-flight, would allow one to tune glancing angles in time so that the total reflection condition is always fulfilled only for one of the two neutron spin eigenstates. Estimates show that with the pulsed reactor IBR-2 such a polarizer allows one to build a small angle neutron scattering instrument capable of effectively using the wavelength band from 2 A with a rather high luminosity (time-averaged flux at sample position being up to 107 n/s/cm-2). (orig.)

  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. [Distribution Characteristics and Source Analysis of Dustfall Trace Elements During Winter in Beijing].

    Science.gov (United States)

    Xiong, Qiu-lin; Zhao, Wen-ji; Guo, Xiao-yu; Chen, Fan-tao; Shu, Tong-tong; Zheng, Xiao-xia; Zhao, Wen-hui

    2015-08-01

    The dustfall content is one of the evaluation indexes of atmospheric pollution. Trace elements especially heavy metals in dustfall can lead to risks to ecological environment and human health. In order to study the distribution characteristics of trace elements, heavy metals pollution and their sources in winter atmospheric dust, 49 dustfall samples were collected in Beijing City and nearby during November 2013 to March 2014. Then the contents (mass percentages) of 40 trace elements were measured by Elan DRC It type inductively coupled plasma mass (ICP-MS). Test results showed that more than half of the trace elements in the dust were less than 10 mg x kg(-1); about a quarter were between 10-100 mg x kg-1); while 7 elements (Pb, Zr, Cr, Cu, Zn, Sr and Ba) were more than 100 mg x kg(-1). The contents of Pb, Cu, Zn, Bi, Cd and Mo of winter dustfall in Beijing city.were respectively 4.18, 4.66, 5.35, 6.31, 6.62, and 8.62 times as high as those of corresponding elements in the surface soil in the same period, which went beyond the soil background values by more than 300% . The contribution of human activities to dustfall trace heavy metals content in Beijing city was larger than that in the surrounding region. Then sources analysis of dustfall and its 20 main trace elements (Cd, Mo, Nb, Ga, Co, Y, Nd, Li, La, Ni, Rb, V, Ce, Pb, Zr, Cr, Cu, Zn, Sr, Ba) was conducted through a multi-method analysis, including Pearson correlation analysis, Kendall correlation coefficient analysis and principal component analysis. Research results indicated that sources of winter dustfall in Beijing city were mainly composed of the earth's crust sources (including road dust, construction dust and remote transmission of dust) and the burning of fossil fuels (vehicle emissions, coal combustion, biomass combustion and industrial processes). PMID:26591998

  3. 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...

  4. Status report on the SIN neutron source

    International Nuclear Information System (INIS)

    The present status is as follows: injector II is under construction, improvement of the proton channel for high current operation is in the design stage, and the spallation neutron source design is progressing

  5. 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

  6. 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

  7. Neutron shielding for a 252 Cf source

    International Nuclear Information System (INIS)

    To determine the neutron shielding features of water-extended polyester a Monte Carlo study was carried out. Materials with low atomic number are predominantly used for neutron shielding because these materials effectively attenuate neutrons, mainly through inelastic collisions and absorption reactions. During the selection of materials to design a neutron shield, prompt gamma production as well as radionuclide production induced by neutron activation must be considered. In this investigation the Monte Carlo method was used to evaluate the performance of a water-extended polyester shield designed for the transportation, storage, and use of a 252Cf isotopic neutron source. During calculations a detailed model for the 252Cf and the shield was utilized. To compare the shielding features of water extended polyester, the calculations were also made for the bare 252Cf in vacuum, air and the shield filled with water. For all cases the calculated neutron spectra was utilized to determine the ambient equivalent neutron dose at four sites around the shielding. In the case of water extended polyester and water shielding the calculations were extended to include the prompt gamma rays produced during neutron interactions, with this information the Kerma in air was calculated at the same locations where the ambient equivalent neutron dose was determined. (Author)

  8. 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

  9. 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)

  10. 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)

  11. Absorbed neutron doses in air holes of fast neutron fields at the RB reactor

    International Nuclear Information System (INIS)

    Different experimental fast neutron fields are created at the RB reactor. The absorbed neutron doses in their air holes are determined on the basis of intermediate and fast neutron spectra measurements. The obtained results are analyzed in connection with application of these fields. (author)

  12. Spatial fluxes and energy distributions of reactor fast neutrons in two types of heat resistant concretes

    International Nuclear Information System (INIS)

    Measurements have been carried out to study the spatial fluxes and energy distributions of reactor fast neutrons transmitted through two types of heat resistant concretes, serpentine concrete and magnetic lemonite concrete. The physical, chemical and mechanical properties of these concretes were checked by well known techniques. In addition, the effect of heating at temperatures up to 500deg C on the crystaline water content was checked by the method of differential thermal analysis. Measurements were performed using a collimated beam of reactor neutrons emitted from a 10 MW research reactor. The neutron spectra transmitted through concrete barriers of different thickness were measured by a scintillation spectrometer with NE-213 liquid organic scintillator. Discrimination against undesired pulses due to gamma-rays was achieved by a method based on pulse shape discrimination technique. The operating principle of this technique is based on the comparison of two weighted time integrals of the detector signal. The measured pulse amplitude distribution was converted to neutron energy distribution by a computational code based on double differentiation technique. The spectrometer workability and the accuracy of the unfolding technique were checked by measuring the neutron spectra of neutrons from Pu-α-Be and 252Cf neutron sources. The obtained neutron spectra for the two concretes were used to derive the total cross sections for neutrons of different energies. (orig.)

  13. Biomedical irradiation system for boron neutron capture therapy at the Kyoto University reactor

    International Nuclear Information System (INIS)

    Physics studies related to radiation source, spectroscopy, beam quality, dosimetry, and biomedical applications using the Kyoto University Reactor Heavy Water Facility are described. Also, described are a Nickel Mirror Neutron Guide Tube and a Super Mirror Neutron Guide Tube that are used both for the measurement of boron concentration in phantom and living tissue and for precise measurements of neutron flux in phantom in the presence of both light and heavy water. Discussed are: (1) spectrum measurements using the time of flight technique, (2) the elimination of gamma rays and fast neutrons from a thermal neutron irradiation field, (3) neutron collimation without producing secondary gamma rays, (4) precise neutron flux measurements, dose estimation, and the measurement of boron concentration in tumor and its periphery using guide tubes, (5) the dose estimation of boron-10 for the first melanoma patient, and (6) special-purpose biological irradiation equipment. Other related subjects are also described

  14. 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.

  15. 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)

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

    International Nuclear Information System (INIS)

    The measurements carried out on reactor dosimetry are applied mainly at the study on the effects of the radiation in the materials of the reactor; a 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 tracks manufactured in the ININ is presented, for the environmental monitoring in penetrations around the primary container of the Unit 1 of the Laguna Verde power plant. The monitoring of neutrons carried out with ends of radiological protection, during those operational tests of the reactor. (author)

  17. Quantitative evaluation of emission control of primary and secondary organic aerosol sources during Beijing 2008 Olympics

    Directory of Open Access Journals (Sweden)

    S. Guo

    2012-12-01

    Full Text Available To explore the primary and secondary sources of fine organic particles after the aggressive implementation of air pollution controls during 2008 Beijing Olympic Games, 12-h PM2.5 concentrations were measured at one urban and one upwind rural site during the CAREBeijing-2008 (Campaigns of Air quality REsearch in Beijing and surrounding region summer field campaign. The PM2.5 concentrations were 72.5±43.6μg m3 and 64.3±36.2μg m−3 at the urban site and rural site, respectively, which were the lowest in recent years due to the implementation of drastic control measures and favorable weather conditions. Five primary and four secondary fine organic particle sources were quantified using a CMB (chemical mass balance model and tracer-yield method. Compared with previous studies in Beijing, the contribution of vehicle emission increased, with diesel engines contributing 16.2±5.9% and 14.5±4.1% to the total organic carbon (OC concentrations and gasoline vehicles accounting for 10.3±8.7% and 7.9±6.2% of the OC concentrations at two sites. Due to the implementation of emission control measures, the OC concentrations from important primary sources have been reduced, and secondary formation has become an important contributor to fine organic aerosols. Compared with the non-controlled period, primary vehicle contributions were reduced by 30% and 24% in the urban and regional area, and reductions in the contribution from coal combustion were 57% and 7%, respectively. These results demonstrate the emission control measures significantly alleviated the primary organic particle pollution in and around Beijing. However, the control effectiveness of secondary organic particles was not significant.

  18. 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

  19. Status report on the cold neutron source of the Garching neutron research facility FRM-II

    International Nuclear Information System (INIS)

    The new high flux research reactor of the Technical University of Munich (Technische Universitaet Muenchen, TUM) will be equipped with a cold neutron source (CNS). The centre of the CNS will be located in the D2O-reflector tank at 400 mm from the reactor core axis, close to the thermal neutron flux maximum. The power of 4000 W developed by the nuclear heating in the 16 litres of liquid deuterium at 25 K, and in the structures, is evacuated by a two phase thermal siphon avoiding film boiling and flooding. The thermal siphon is a single tube with counter current flow. It is inclined by 10 deg from vertical, and optimised for a deuterium flow rate of 14 g/s. Optimisation of structure design and material, as well as safety aspects will be discussed. Those parts of the structure, which are exposed to high thermal neutron flux, are made from Zircaloy 4 and 6061T6 aluminium. Structure failure due to embrittlement of the structure material under high rapid neutron flux is very improbable during the life time of the CNS (30 years). Double, in pile even triple, containment with inert gas liner guarantees lack of explosion risk and of tritium contamination to the environment. Adding a few percent of hydrogen (H2) to the deuterium (D2) will improve the moderating properties of our relatively small moderator volume. Nearly all of the hydrogen is bound in the form of HD molecules. A long term change of the hydrogen content in the deuterium is avoided be storing the mixture not in a gas buffer volume but as a metal hydride at low pressure. The metal hydride storage system contains two getter beds, one with 250 kg of LaCo3Ni2, the other one with 150 kg of ZrCo(0.8)Ni(0.2). Each bed can take the total gas inventory, both beds together can absorb the total gas inventory in less than 6 minutes at a pressure < 3 bar. The new reactor will have 13 beam tubes, 4 of which are looking at the cold neutron source (CNS), including two for very cold (VCN) and ultra-cold neutron (UCN

  20. 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.