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Sample records for delayed neutron analysis

  1. The delayed neutron method of uranium analysis

    International Nuclear Information System (INIS)

    Wall, T.

    1989-01-01

    The technique of delayed neutron analysis (DNA) is discussed. The DNA rig installed on the MOATA reactor, the assay standards and the types of samples which have been assayed are described. Of the total sample throughput of about 55,000 units since the uranium analysis service began, some 78% has been concerned with analysis of uranium ore samples derived from mining and exploration. Delayed neutron analysis provides a high sensitivity, low cost uranium analysis method for both uranium exploration and other applications. It is particularly suitable for analysis of large batch samples and for non-destructive analysis over a wide range of matrices. 8 refs., 4 figs., 3 tabs

  2. Kalman filter analysis of delayed neutron nondestructive assay measurements

    International Nuclear Information System (INIS)

    Aumeier, S. E.

    1998-01-01

    The ability to nondestructively determine the presence and quantity of fissile and fertile nuclei in various matrices is important in several nuclear applications including international and domestics safeguards, radioactive waste characterization and nuclear facility operations. Material irradiation followed by delayed neutron counting is a well known and useful nondestructive assay technique used to determine the fissile-effective content of assay samples. Previous studies have demonstrated the feasibility of using Kalman filters to unfold individual isotopic contributions to delayed neutron measurements resulting from the assay of mixes of uranium and plutonium isotopes. However, the studies in question used simulated measurement data and idealized parameters. We present the results of the Kalman filter analysis of several measurements of U/Pu mixes taken using Argonne National Laboratory's delayed neutron nondestructive assay device. The results demonstrate the use of Kalman filters as a signal processing tool to determine the fissile and fertile isotopic content of an assay sample from the aggregate delayed neutron response following neutron irradiation

  3. Neutron stochastic transport theory with delayed neutrons

    International Nuclear Information System (INIS)

    Munoz-Cobo, J.L.; Verdu, G.

    1987-01-01

    From the stochastic transport theory with delayed neutrons, the Boltzmann transport equation with delayed neutrons for the average flux emerges in a natural way without recourse to any approximation. From this theory a general expression is obtained for the Feynman Y-function when delayed neutrons are included. The single mode approximation for the particular case of a subcritical assembly is developed, and it is shown that Y-function reduces to the familiar expression quoted in many books, when delayed neutrons are not considered, and spatial and source effects are not included. (author)

  4. Analytical applications for delayed neutrons

    International Nuclear Information System (INIS)

    Eccleston, G.W.

    1983-01-01

    Analytical formulations that describe the time dependence of neutron populations in nuclear materials contain delayed-neutron dependent terms. These terms are important because the delayed neutrons, even though their yields in fission are small, permit control of the fission chain reaction process. Analytical applications that use delayed neutrons range from simple problems that can be solved with the point reactor kinetics equations to complex problems that can only be solved with large codes that couple fluid calculations with the neutron dynamics. Reactor safety codes, such as SIMMER, model transients of the entire reactor core using coupled space-time neutronics and comprehensive thermal-fluid dynamics. Nondestructive delayed-neutron assay instruments are designed and modeled using a three-dimensional continuous-energy Monte Carlo code. Calculations on high-burnup spent fuels and other materials that contain a mix of uranium and plutonium isotopes require accurate and complete information on the delayed-neutron periods, yields, and energy spectra. A continuing need exists for delayed-neutron parameters for all the fissioning isotopes

  5. Analysis of incident-energy dependence of delayed neutron yields in actinides

    Energy Technology Data Exchange (ETDEWEB)

    Nasir, Mohamad Nasrun bin Mohd, E-mail: monasr211@gmail.com; Metorima, Kouhei, E-mail: kohei.m2420@hotmail.co.jp; Ohsawa, Takaaki, E-mail: ohsawa@mvg.biglobe.ne.jp; Hashimoto, Kengo, E-mail: kengoh@pp.iij4u.or.jp [Graduate School of Science and Engineering, Kindai University, Kowakae, Higashi-Osaka, 577-8502 (Japan)

    2015-04-29

    The changes of delayed neutron yields (ν{sub d}) of Actinides have been analyzed for incident energy up to 20MeV using realized data of precursor after prompt neutron emission, from semi-empirical model, and delayed neutron emission probability data (P{sub n}) to carry out a summation method. The evaluated nuclear data of the delayed neutron yields of actinide nuclides are still uncertain at the present and the cause of the energy dependence has not been fully understood. In this study, the fission yields of precursor were calculated considering the change of the fission fragment mass yield based on the superposition of fives Gaussian distribution; and the change of the prompt neutrons number associated with the incident energy dependence. Thus, the incident energy dependent behavior of delayed neutron was analyzed.The total number of delayed neutron is expressed as ν{sub d}=∑Y{sub i} • P{sub ni} in the summation method, where Y{sub i} is the mass yields of precursor i and P{sub ni} is the delayed neutron emission probability of precursor i. The value of Y{sub i} is derived from calculation of post neutron emission mass distribution using 5 Gaussian equations with the consideration of large distribution of the fission fragments. The prompt neutron emission ν{sub p} increases at higher incident-energy but there are two different models; one model says that the fission fragment mass dependence that prompt neutron emission increases uniformly regardless of the fission fragments mass; and the other says that the major increases occur at heavy fission fragments area. In this study, the changes of delayed neutron yields by the two models have been investigated.

  6. Use of one delayed-neutron precursor group in transient analysis

    International Nuclear Information System (INIS)

    Diamond, D.J.

    1983-01-01

    In most reactor dynamics calculations six groups of delayed-neutron precursors are usually accounted for. However, under certain circumstances it may be advantageous to simplify the calculation and utilize a single delayed-neutron group. The motivation for going to one precursor group is economy. For LWR transient codes that use point kinetics the equations are solved very rapidly and six precursor groups should always be used. However, codes with spatially dependent neutron kinetics are very long running and the use of one precursor group may save computer costs and not impair the accuracy of the results significantly. Furthermore, in some codes, the elimation of five presursor groups makes additional memory available which may be used to give a net increase in the accuracy of the calculations, e.g., by allowing for an increase in mesh density. In order to use one delayed neutron precursor group it is necessary to derive a single decay constant, 6 lambda-, which, along with the total (or one group) delayed neutron fraction β = Σ/sub i = 1/β/sub i/, will adequately describe the transeint precursor behavior. The present summary explains how a recommendation for lambda- was derived

  7. Systematic of delayed neutron parameters

    International Nuclear Information System (INIS)

    Isaev, S.G.; Piksaikin, V.M.

    2000-01-01

    The experimental studies of the energy dependence of the delayed neutron (DN) parameters for various fission systems has shown that the behaviour of a some combination of delayed neutron parameters has a similar features. On the basis of this findings the systematics of delayed neutron experimental data for thorium, uranium, plutonium and americium isotopes have been investigated with the purpose to find a correlation of DN parameters with characteristics of fissioning system as well as a correlation between the delayed neutron parameters themselves. It was presented the preliminary results which were obtained during study the physics interpretation of the results [ru

  8. Proceedings of the specialists' meeting on delayed neutron nuclear data

    International Nuclear Information System (INIS)

    Katakura, Jun-ichi

    1999-07-01

    This report is the Proceedings of the Specialists' Meeting on Delayed Neutron Nuclear Data. The meeting was held on January 28-29, 1999, at the Tokai Research Establishment of Japan Atomic Energy Research Institute with the participation of thirty specialists, who are evaluators, theorist, experimentalists. Although the fraction of the delayed neutron is no more than 1% in the total neutrons emitted in the fission process, it plays an important roll in the control of fission reactor. In the meeting, the following topics were reported: the present status of delayed neutron data in the major evaluated data libraries, measurements of effective delayed neutron fraction using FCA (Fast Critical Assembly) and TCA (Tank-type Critical Assembly) and their analyses, sensitivity analysis for fast reactor, measurements of delayed neutron emission from actinides and so on. As another topics, delayed neutron in transmutation system and fission yield data were also presented. Free discussion was held on the future activity of delayed neutron data evaluation. The discussion was helpful for the future activity of the delayed neutron working group of JNDC aiming to the evaluation of delayed neutron data for JENDL-3.3. The 15 of the presented papers are indexed individually. (J.P.N.)

  9. A general formula considering one group delayed neutron under nonequilibrium condition

    International Nuclear Information System (INIS)

    Li Haofeng; Chen Wenzhen; Zhu Qian; Luo Lei

    2008-01-01

    A general neutron breeder formula is developed when the reactor does not reach the steady state and the reactivity changes in phase. This formula can be used to calculate the results of six groups delayed neutron model through a way of amending λ in one group delayed neutron model. The analysis shows that the solution of amended single group delayed neutron model is approximately equal to that of six-group delayed neutron model, and the amended model meets the engineering accuracy. (authors)

  10. Delayed neutrons in ANSTO

    International Nuclear Information System (INIS)

    Wall, T.

    1988-01-01

    Delayed neutron analysis carried out at the Australian Nuclear Scientific and Technology Organization facilities, provides a fast, high sensitivity, low cost, reliable method, particularly suitable for large batches of samples, and for non destructive analysis of a range of materials. While its main use has been in uranium exploration, other applications include archeological investigations, agriculture, oceanography and biology

  11. An In-Pile Kinetic Method for Determining the Delayed Neutron Fraction βeff

    International Nuclear Information System (INIS)

    Gilad, E.; Rivin, O.; Ettedgui, H.; Yaar, I.; Geslot, B.; Pepino, A.; Di Salvo, J.; Gruel, A.; Blaise, P.

    2014-01-01

    Delayed neutrons are of fundamental importance in the field of nuclear reactor dynamics and control. Although only a small fraction of the neutrons emitted by fission are not prompt, the knowledge of the delayed neutrons parameters is essential for transient analysis, such as startup or shutdown of the reactor, as well as for accidents analysis and control system design [1]. One of the main delayed neutron parameters used in the point reactor model equations is the effective delayed neutron fraction, which incorporates both delayed neutron spectral properties and core geometrical configuration [1,2]. Additional delayed neutron parameters include the fraction of fission neutrons emitted in each delayed group, and the delayed neutron precursors decay constants . Experimental efforts aimed at determining the value ofβ, which provide experimental support for the evaluation of delayed neutron parameters, are extremely valuable. This is due to the fact that unlike other fields in reactor physics, e.g. criticality safety or shielding, the availability of experimental data and benchmark problems for validating delayed neutron parameters and its implementation in different models is highly limited. Furthermore, the existing experimental data exhibit significant discrepancies between the different sets of parameter, which lead to substantial disparity in the analysis of kinetic experiments and reactor dynamic behavior]. In this work, a method for determining the effective delayed neutron fraction using in-pile reactivity oscillation and Fourier analysis is presented. The method is based on measurements of the reactor's power response to small periodic in-pile reactivity perturbations and utilizes Fourier analysis for reconstruction of the reactor zero power transfer function. Knowledge of the reactor transfer function enables the estimation of theβ value using multi-parameter nonlinear fit. The method accounts for higher harmonics, which are excited by the trapezoidal

  12. Proceedings of the specialists' meeting on delayed neutron nuclear data

    Energy Technology Data Exchange (ETDEWEB)

    Katakura, Jun-ichi [ed.] [Japanese Nuclear Data Committee, Tokai, Ibaraki (Japan)

    1999-07-01

    This report is the Proceedings of the Specialists' Meeting on Delayed Neutron Nuclear Data. The meeting was held on January 28-29, 1999, at the Tokai Research Establishment of Japan Atomic Energy Research Institute with the participation of thirty specialists, who are evaluators, theorist, experimentalists. Although the fraction of the delayed neutron is no more than 1% in the total neutrons emitted in the fission process, it plays an important roll in the control of fission reactor. In the meeting, the following topics were reported: the present status of delayed neutron data in the major evaluated data libraries, measurements of effective delayed neutron fraction using FCA (Fast Critical Assembly) and TCA (Tank-type Critical Assembly) and their analyses, sensitivity analysis for fast reactor, measurements of delayed neutron emission from actinides and so on. As another topics, delayed neutron in transmutation system and fission yield data were also presented. Free discussion was held on the future activity of delayed neutron data evaluation. The discussion was helpful for the future activity of the delayed neutron working group of JNDC aiming to the evaluation of delayed neutron data for JENDL-3.3. The 15 of the presented papers are indexed individually. (J.P.N.)

  13. Effective delayed neutron fraction and prompt neutron lifetime of Tehran research reactor mixed-core

    International Nuclear Information System (INIS)

    Lashkari, A.; Khalafi, H.; Kazeminejad, H.

    2013-01-01

    Highlights: ► Kinetic parameters of Tehran research reactor mixed-core have been calculated. ► Burn-up effect on TRR kinetics parameters has been studied. ► Replacement of LEU-CFE with HEU-CFE in the TRR core has been investigated. ► Results of each mixed core were compared to the reference core. ► Calculation of kinetic parameters are necessary for reactivity and power excursion transient analysis. - Abstract: In this work, kinetic parameters of Tehran research reactor (TRR) mixed cores have been calculated. The mixed core configurations are made by replacement of the low enriched uranium control fuel elements with highly enriched uranium control fuel elements in the reference core. The MTR P C package, a nuclear reactor analysis tool, is used to perform the analysis. Simulations were carried out to compute effective delayed neutron fraction and prompt neutron lifetime. Calculation of kinetic parameters is necessary for reactivity and power excursion transient analysis. The results of this research show that effective delayed neutron fraction decreases and prompt neutron lifetime increases with the fuels burn-up. Also, by increasing the number of highly enriched uranium control fuel elements in the reference core, the prompt neutron lifetime increases, but effective delayed neutron fraction does not show any considerable change

  14. Automated uranium analysis by delayed-neutron counting

    International Nuclear Information System (INIS)

    Kunzendorf, H.; Loevborg, L.; Christiansen, E.M.

    1980-10-01

    Automated uranium analysis by fission-induced delayed-neutron counting is described. A short description is given of the instrumentation including transfer system, process control, irradiation and counting sites, and computer operations. Characteristic parameters of the facility (sample preparations, background, and standards) are discussed. A sensitivity of 817 +- 22 counts per 10 -6 g U is found using irradiation, delay, and counting times of 20 s, 5 s, and 10 s, respectively. Presicion is generally less than 1% for normal geological samples. Critical level and detection limits for 7.5 g samples are 8 and 16 ppb, respectively. The importance of some physical and elemental interferences are outlined. Dead-time corrections of measured count rates are necessary and a polynomical expression is used for count rates up to 10 5 . The presence of rare earth elements is regarded as the most important elemental interference. A typical application is given and other areas of application are described. (auther)

  15. Statistical precision of delayed-neutron nondestructive assay techniques

    International Nuclear Information System (INIS)

    Bayne, C.K.; McNeany, S.R.

    1979-02-01

    A theoretical analysis of the statistical precision of delayed-neutron nondestructive assay instruments is presented. Such instruments measure the fissile content of nuclear fuel samples by neutron irradiation and delayed-neutron detection. The precision of these techniques is limited by the statistical nature of the nuclear decay process, but the precision can be optimized by proper selection of system operating parameters. Our method is a three-part analysis. We first present differential--difference equations describing the fundamental physics of the measurements. We then derive and present complete analytical solutions to these equations. Final equations governing the expected number and variance of delayed-neutron counts were computer programmed to calculate the relative statistical precision of specific system operating parameters. Our results show that Poisson statistics do not govern the number of counts accumulated in multiple irradiation-count cycles and that, in general, maximum count precision does not correspond with maximum count as first expected. Covariance between the counts of individual cycles must be considered in determining the optimum number of irradiation-count cycles and the optimum irradiation-to-count time ratio. For the assay system in use at ORNL, covariance effects are small, but for systems with short irradiation-to-count transition times, covariance effects force the optimum number of irradiation-count cycles to be half those giving maximum count. We conclude that the equations governing the expected value and variance of delayed-neutron counts have been derived in closed form. These have been computerized and can be used to select optimum operating parameters for delayed-neutron assay devices

  16. Delayed neutron yield from fast neutron induced fission of 238U

    International Nuclear Information System (INIS)

    Piksaikin, V.M.; Kazakov, L.E.; Isaev, S.G.; Roshchenko, V.A.; Goverdovski, A.A.; Tertytchnyi, R.G.

    2002-01-01

    The measurements of the total delayed neutron yield from fast neutron induced fission of 238 U were made. The experimental method based on the periodic irradiation of the fissionable sample by neutrons from a suitable nuclear reaction had been employed. The preliminary results on the energy dependence of the total delayed neutron yield from fission of 238 U are obtained. According to the comparison of experimental data with our prediction based on correlation properties of delayed neutron characteristics, it is concluded that the value of the total delayed neutron yield near the threshold of (n,f) reaction is not a constant. (author)

  17. Improved Delayed-Neutron Spectroscopy Using Trapped Ions

    Energy Technology Data Exchange (ETDEWEB)

    Norman, Eric

    2018-04-24

    The neutrons emitted following the  decay of fission fragments (known as delayed neutrons because they are emitted after fission on a timescale of the -decay half-lives) play a crucial role in reactor performance and control. Reviews of delayed-neutron properties highlight the need for high-quality data for a wide variety of delayed-neutron emitters to better understand the timedependence and energy spectrum of the neutrons as these properties are essential for a detailed understanding of reactor kinetics needed for reactor safety and to understand the behavior of these reactors under various accident and component-failure scenarios. For fast breeder reactors, criticality calculations require accurate delayed-neutron energy spectra and approximations that are acceptable for light-water reactors such as assuming the delayed-neutron and fission-neutron energy spectra are identical are not acceptable and improved -delayed neutron data is needed for safety and accident analyses for these reactors. With improved nuclear data, the delayedneutrons flux and energy spectrum could be calculated from the contributions from individual isotopes and therefore could be accurately modeled for any fuel-cycle concept, actinide mix, or irradiation history. High-quality -delayed neutron measurements are also critical to constrain modern nuclear-structure calculations and empirical models that predict the decay properties for nuclei for which no data exists and improve the accuracy and flexibility of the existing empirical descriptions of delayed neutrons from fission such as the six-group representation

  18. Correlation properties of delayed neutrons from fast neutron induced fission

    International Nuclear Information System (INIS)

    Piksaikin, V.M.; Isaev, S.G.

    1998-01-01

    The experimental studies of the energy dependence of the delayed neutron parameters for various fissioning systems has shown that the behavior of a some combination of delayed neutron parameters (group relative abundances a i and half lives T i ) has a similar features. On the basis of this findings the systematics of delayed neutron experimental data for thorium, uranium, plutonium and americium isotopes have been investigated with the purpose to find a correlation of DN parameters with characteristics of fissioning system as well as a correlation between the delayed neutron parameters themselves. Below we will present the preliminary results which were obtained during this study omitting the physics interpretation of the results. (author)

  19. The implication of sensitivity analysis on the safety and delayed-neutron parameters for fast breeder reactors

    International Nuclear Information System (INIS)

    Onega, R.J.; Florian, R.J.

    1983-01-01

    The delayed-neutron energy spectra for LMFBRs are not as well known as those for LWRs. These spectra are necessary for kinetics calculations which play an important role in safety and accident analyses. A sensitivity analysis was performed to study the response of the reactor power and power density to uncertainties in the delayed-neutron spectra during a rod-ejection accident. The accidents studied were central control-rod-ejections with ejection times of 2,10 and 30s. A two-energy group and two-precursor group model was formulated for the International Nuclear Fuel Cycle Evaluation (INFCE) reference design MOX-fueled LMFBR. The sensitivity analysis is based on the use of adjoints so that it is not necessary to repeatedly solve the governing (kinetics) equations to obtain the sensitivity derivatives. This is of particular importance when large systems of equations are used. The power and power-density responses were found to be most sensitive to uncertainties in the spectrum of the second delayed-neutron precursor group, resulting from the fission of 238 U, producing neutrons in the first energy group. It was found, for example, that for a rod-ejection time of 30s, and uncertainty of 7.2% in the fast components of the spectra resulted in a 24% uncertainty in the predicted power and power density. These responses were recalculated by repeatedly solving the kinetics equations. The maximum discrepancy between the recalculated and the sensitivity analysis response was only 1.6%. The results of the sensitivity analysis indicate the need for improved delayed-neutron spectral data in order to reduce the uncertainties in accident analyses. (author)

  20. Gamma/neutron competition above the neutron separation energy in delayed neutron emitters

    Directory of Open Access Journals (Sweden)

    Valencia E.

    2014-03-01

    Full Text Available To study the β-decay properties of some well known delayed neutron emitters an experiment was performed in 2009 at the IGISOL facility (University of Jyväskylä in Finland using Total Absorption γ-ray Spectroscopy (TAGS technique. The aim of these measurements is to obtain the full β-strength distribution below the neutron separation energy (Sn and the γ/neutron competition above. This information is a key parameter in nuclear technology applications as well as in nuclear astrophysics and nuclear structure. Preliminary results of the analysis show a significant γ-branching ratio above Sn.

  1. The effective delayed neutron fraction for bare-metal criticals

    International Nuclear Information System (INIS)

    Pearlstein, S.

    1999-01-01

    Given sufficient material, a large number of actinides could be used to form bare-metal criticals. The effective delayed neutron fraction for a bare critical comprised of a fissile material is comparable with the absolute delayed neutron fraction. The effective delayed neutron fraction for a bare critical composed of a fissionable material is reduced by factors of 2 to 10 when compared with the absolute delayed neutron fraction. When the effective delayed neutron fraction is small, the difference between delayed and prompt criticality is small, and extreme caution must be used in critical assemblies of these materials. This study uses an approximate but realistic model to survey the actinide region to compare effective delayed neutron fractions with absolute delayed neutron fractions

  2. Delayed neutrons in liquid metal spallation targets

    International Nuclear Information System (INIS)

    Ridikas, D.; Bokov, P.; David, J.C.; Dore, D.; Giacri, M.L.; Van Lauwe, A.; Plukiene, R.; Plukis, A.; Ignatiev, S.; Pankratov, D.

    2003-01-01

    The next generation spallation neutron sources, neutrino factories or RIB production facilities currently being designed and constructed around the world will increase the average proton beam power on target by a few orders of magnitude. Increased proton beam power results in target thermal hydraulic issues leading to new target designs, very often based on flowing liquid metal targets such as Hg, Pb, Pb-Bi. Radioactive nuclides produced in liquid metal targets are transported into hot cells, past electronics, into pumps with radiation sensitive components, etc. Besides the considerable amount of photon activity in the irradiated liquid metal, a significant amount of the delayed neutron precursor activity can be accumulated in the target fluid. The transit time from the front of a liquid metal target into areas, where delayed neutrons may be important, can be as short as a few seconds, well within one half-life of many delayed neutron precursors. Therefore, it is necessary to evaluate the total neutron flux (including delayed neutrons) as a function of time and determine if delayed neutrons contribute significantly to the dose rate. In this study the multi-particle transport code MCNPX combined with the material evolution program CINDER'90 will be used to evaluate the delayed neutron flux and spectra. The following scientific issues will be addressed in this paper: - Modeling of a typical geometry of the liquid metal spallation target; - Predictions of the prompt neutron fluxes, fission fragment and spallation product distributions; - Comparison of the above parameters with existing experimental data; - Time-dependent calculations of delayed neutron precursors; - Neutron flux estimates due to the prompt and delayed neutron emission; - Proposal of an experimental program to measure delayed neutron spectra from high energy spallation-fission reactions. The results of this study should be directly applicable in the design study of the European MegaPie (1 MW

  3. Neutron delayed choice experiments

    International Nuclear Information System (INIS)

    Bernstein, H.J.

    1986-01-01

    Delayed choice experiments for neutrons can help extend the interpretation of quantum mechanical phenomena. They may also rule out alternative explanations which static interference experiments allow. A simple example of a feasible neutron test is presented and discussed. (orig.)

  4. Rapid uranium analysis by delayed neutron counting of neutron activated samples

    International Nuclear Information System (INIS)

    Papadopoulos, N.N.

    1985-01-01

    The uranium analyzer at the Nuclear Research Center ''Demokritos'' and the delayed neutron method have been used to determine the uranium content in lignite, in chemically enriched samples and in solutions of extractable uranium. The results are compared with data obtained by other methods. In the case of dissolved extractable uranium. The results are in good agreement with X-ray fluorescence data in the range 100 ppm to 2000 ppm while beyond these limits the discrepancies between neutron and spectrophotometric data are observed. The results for lignite samples are in good agreement with gamma spectrometric data. Discrepancies indicate that more extensive intercomparisons are needed to check the reliability of various methods

  5. Systematics in delayed neutron yields

    Energy Technology Data Exchange (ETDEWEB)

    Ohsawa, Takaaki [Kinki Univ., Higashi-Osaka, Osaka (Japan). Atomic Energy Research Inst.

    1998-03-01

    An attempt was made to reproduce the systematic trend observed in the delayed neutron yields for actinides on the basis of the five-Gaussian representation of the fission yield together with available data sets for delayed neutron emission probability. It was found that systematic decrease in DNY for heavier actinides is mainly due to decrease of fission yields of precursors in the lighter side of the light fragment region. (author)

  6. Nondestructive analysis of the natural uranium mass through the measurement of delayed neutrons using the technique of pulsed neutron source

    International Nuclear Information System (INIS)

    Coelho, Paulo Rogerio Pinto

    1979-01-01

    This work presents results of non destructive mass analysis of natural uranium by the pulsed source technique. Fissioning is produced by irradiating the test sample with pulses of 14 MeV neutrons and the uranium mass is calculated on a relative scale from the measured emission of delayed neutrons. Individual measurements were normalised against the integral counts of a scintillation detector measuring the 14 MeV neutron intensity. Delayed neutrons were measured using a specially constructed slab detector operated in anti synchronism with the fast pulsed source. The 14 MeV neutrons were produced via the T(d,n) 4 He reaction using a 400 kV Van de Graaff accelerated operated at 200 kV in the pulsed source mode. Three types of sample were analysed, namely: discs of metallic uranium, pellets of sintered uranium oxide and plates of uranium aluminium alloy sandwiched between aluminium. These plates simulated those of Material Testing Reactor fuel elements. Results of measurements were reproducible to within an overall error in the range 1.6 to 3.9%; the specific error depending on the shape, size and mass of the sample. (author)

  7. Development of a methodology for analysis of delayed-neutron signals

    International Nuclear Information System (INIS)

    Gross, K.C.; Strain, R.V.; Fryer, R.M.

    1980-02-01

    Experimental and analytical techniques have been developed for analysis and characterization of delayed-neutron (DN) signals that can provide diagnostic information to augment data from cover-gas analyses in the detection and identification of breached elements in an LMFBR. Eleven flow-reduction tests have been run in EBR-II to provide base data support for predicting DN signal characteristics during exposed-fuel operation. Results from the tests demonstrate the feasibility and practicability of response-analysis techniques for determining (a) the transit time, T/sub tr/, for DN emitters traveling from the core to the detector and (b) the isotropic holdup time, T/sub h/, of DN precursors in the fuel element

  8. Beta-delayed neutron decay of $^{33}$Na

    CERN Document Server

    Radivojevic, Z; Caurier, E; Cederkäll, J; Courtin, S; Dessagne, P; Jokinen, A; Knipper, A; Le Scornet, G; Lyapin, V G; Miehé, C; Nowacki, F; Nummela, S; Oinonen, M; Poirier, E; Ramdhane, M; Trzaska, W H; Walter, G; Äystö, J

    2002-01-01

    Beta-delayed neutron decay of /sup 33/Na has been studied using the on-line mass separator ISOLDE. The delayed neutron spectra were measured by time-of-flight technique using fast scintillators. Two main neutron groups at 800(60) and 1020(80) keV were assigned to the /sup 33/Na decay, showing evidence for strong feeding of states at about 4 MeV in /sup 33/Mg. By simultaneous beta - gamma -n counting the delayed neutron emission probabilities P/sub 1n/ = 47(6)% and P /sub 2n/ = 13(3)% were determined. The half-life value for /sup 33 /Na, T/sub 1/2/ = 8.0(3) ms, was measured by three different techniques, one employing identifying gamma transitions and two employing beta and neutron counting. (21 refs).

  9. An experimental facility for studying delayed neutron emission

    International Nuclear Information System (INIS)

    Dermendzhiev, E.; Nazarov, V.M.; Pavlov, S.S.; Ruskov, Iv.; Zamyatin, Yu.S.

    1993-01-01

    A new experimental facility for studying delayed neutron emission has been designed and tested. A method based on utilization of the Dubna IBR-2 pulsed reactor, has been proposed and realized for periodical irradiation of targets composed of fissionable isotopes. Such a powerful pulsed neutron source in combination with a slow neutron chopper synchronized with the reactor bursts makes possible variation of the exposure duration and effective suppression of the fast neutron background due to delay neutrons emitted from the reactor core. Detection of delayed neutrons from the target is carried out by a high-efficiency multicounter neutron detector with a near-4π geometry. Some test measurements and results are briefly described. Possible use of the facility for other tasks is also discussed. 14 refs.; 14 figs

  10. Radiochemical Means of Investigating Delayed Neutron Precursors

    International Nuclear Information System (INIS)

    Marmol, P. del

    1968-01-01

    Fast radiochemical methods used now for the determination of delayed neutron precursors are classified and reviewed: precipitations, solvent extractions, range experiments, milking, gas sweeping, isotopic and ion exchange, hot atom reactions and diffusion loss. Advantages and limitations of irradiation systems with respect to fast separations are discussed: external beams which allow faster separations only have low neutron fluxes, internal beams which are mostly fit for gaseous reactions; and rabbits for solution irradiations. Future prospects of radiochemical procedures are presented; among these, studies should be mostly oriented towards gaseous reactions which offer possibilities of isolating very short-lived delayed neutron precursors. Chemical procedures for delayed neutron precursor detection are compared with mass spectrometric and isotope separator techniques; it is concluded that the methods are complementary. (author)

  11. Radiochemical Means of Investigating Delayed Neutron Precursors

    International Nuclear Information System (INIS)

    Marmol, P. del

    1968-01-01

    Fast radiochemical methods used now for the determination of delayed neutron precursors are classified and reviewed: precipitations, solvent extractions, range experiments, milking, gas sweeping, isotopic and ion exchange, hot-atom reactions and diffusion loss. Advantages and limitations of irradiation systems with respect to fast separations are discussed: external beams which allow faster separations only have low neutron fluxes, internal beams which are mostly fit for gaseous reactions; and rabbits for solution irradiations. Future prospects of radiochemical procedures are presented; among these, studies should be mostly oriented towards gaseous reactions which offer possibilities of isolating very short-lived delayed neutron precursors. Chemical procedures for delayed neutron precursor detection are compared with mass spectrometric and isotope-separator techniques; it is concluded that the methods are complementary. (author)

  12. Study of calculated and measured time dependent delayed neutron yields

    International Nuclear Information System (INIS)

    Waldo, R.W.

    1980-05-01

    Time-dependent delayed neutron emission is of interest in reactor design, reactor dynamics, and nuclear physics studies. The delayed neutrons from neutron-induced fission of 232 U, 237 Np, 238 Pu, 241 Am, /sup 242m/Am, 245 Cm, and 249 Cf were studied for the first time. The delayed neutron emission from 232 Th, 233 U, 235 U, 238 U, 239 Pu, 241 Pu, and 242 Pu were measured as well. The data were used to develop an empirical expression for the total delayed neutron yield. The expression gives accurate results for a large variety of nuclides from 232 Th to 252 Cf. The data measuring the decay of delayed neutrons with time were used to derive another empirical expression predicting the delayed neutron emission with time. It was found that nuclides with similar mass-to-charge ratios have similar decay patterns. Thus the relative decay pattern of one nuclide can be established by any measured nuclide with a similar mass-to-charge ratio. A simple fission product yield model was developed and applied to delayed neutron precursors. It accurately predicts observed yield and decay characteristics. In conclusion, it is possible to not only estimate the total delayed neutron yield for a given nuclide but the time-dependent nature of the delayed neutrons as well. Reactors utilizing recycled fuel or burning actinides are likely to have inventories of fissioning nuclides that have not been studied until now. The delayed neutrons from these nuclides can now be incorporated so that their influence on the stability and control of reactors can be delineated. 8 figures, 39 tables

  13. Evaluation of Kalman filters and genetic algorithms for delayed-neutron nondestructive assay data analyses

    International Nuclear Information System (INIS)

    Aumeier, S.E.; Forsmann, J.H.

    1998-01-01

    The ability to nondestructively determine the presence and quantity of fissile/fertile nuclei in various matrices is important in several areas of nuclear applications, including international and domestic safeguards, radioactive waste characterization, and nuclear facility operations. An analysis was performed to determine the feasibility of identifying the masses of individual fissionable isotopes from a cumulative delayed-neutron signal resulting form the neutron irradiation of several uranium and plutonium isotopes. The feasibility of two separate data-processing techniques was studied: Kalman filtering and genetic algorithms. The basis of each technique is reviewed, and the structure of the algorithms as applied to the delayed-neutron analysis problem is presented. The results of parametric studies performed using several variants of the algorithms are presented. The effect of including additional constraining information such as additional measurements and known relative isotopic concentration is discussed. The parametric studies were conducted using simulated delayed-neutron data representative of the cumulative delayed-neutron response following irradiation of a sample containing 238 U, 235 U, 239 Pu, and 240 Pu. The results show that by processing delayed-neutron data representative of two significantly different fissile/fertile fission ratios, both Kalman filters and genetic algorithms are capable of yielding reasonably accurate estimates of the mass of individual isotopes contained in a given assay sample

  14. Information about the new 8-group delayed neutron set preparation

    International Nuclear Information System (INIS)

    Svarny, J.

    1998-01-01

    Some comments to the present state concerning delayed neutron data preparation is given and preliminary analysis of the new 8-group delayed data (relative abundances) is presented. Comparisons of the 8-group to 6-group set is given for rod drop experiment (Unit 1, Cycle 14, NPP Dukovany).(Author)

  15. Calibration of the delayed-gamma neutron activation facility

    International Nuclear Information System (INIS)

    Ma, R.; Zhao, X.; Rarback, H.M.; Yasumura, S.; Dilmanian, F.A.; Moore, R.I.; Lo Monte, A.F.; Vodopia, K.A.; Liu, H.B.; Economos, C.D.; Nelson, M.E.; Aloia, J.F.; Vaswani, A.N.; Weber, D.A.; Pierson, R.N. Jr.; Joel, D.D.

    1996-01-01

    The delayed-gamma neutron activation facility at Brookhaven National Laboratory was originally calibrated using an anthropomorphic hollow phantom filled with solutions containing predetermined amounts of Ca. However, 99% of the total Ca in the human body is not homogeneously distributed but contained within the skeleton. Recently, an artificial skeleton was designed, constructed, and placed in a bottle phantom to better represent the Ca distribution in the human body. Neutron activation measurements of an anthropomorphic and a bottle (with no skeleton) phantom demonstrate that the difference in size and shape between the two phantoms changes the total body calcium results by less than 1%. To test the artificial skeleton, two small polyethylene jerry-can phantoms were made, one with a femur from a cadaver and one with an artificial bone in exactly the same geometry. The femur was ashed following the neutron activation measurements for chemical analysis of Ca. Results indicate that the artificial bone closely simulates the real bone in neutron activation analysis and provides accurate calibration for Ca measurements. Therefore, the calibration of the delayed-gamma neutron activation system is now based on the new bottle phantom containing an artificial skeleton. This change has improved the accuracy of measurement for total body calcium. Also, the simple geometry of this phantom and the artificial skeleton allows us to simulate the neutron activation process using a Monte Carlo code, which enables us to calibrate the system for human subjects larger and smaller than the phantoms used as standards. copyright 1996 American Association of Physicists in Medicine

  16. Delayed power analysis

    International Nuclear Information System (INIS)

    Adamovich, L.A.; Azarov, V.V.

    1999-01-01

    Time dependent core power behavior in a nuclear reactor is described with well-known neutron kinetics equations. At the same time, two portions are distinguished in energy released from uranium nuclei fission; one released directly at fission and another delayed (residual) portion produced during radioactive decay of fission products. While prompt power is definitely described with kinetics equations, the delayed power presentation still remains outstanding. Since in operation the delayed power part is relatively small (about 6%) operation, it can be neglected for small reactivity disturbances assuming that entire power obeys neutron kinetics equations. In case of a high negative reactivity rapidly inserted in core (e.g. reactor scram initiation) the prompt and delayed components can be calculated separately with practically no impact on each other, employing kinetics equations for prompt power and known approximation formulas for delayed portion, named residual in this specific case. Under substantial disturbances the prompt component in the dynamic process becomes commensurable with delayed portion, thus making necessary to take into account their cross impact. A system of differential equations to describe time-dependent behavior of delayed power is presented. Specific NPP analysis shows a way to significantly simplify the task formulation. (author)

  17. The neutron long counter NERO for studies of β-delayed neutron emission in the r-process

    International Nuclear Information System (INIS)

    Pereira, J.; Hosmer, P.; Lorusso, G.; Santi, P.; Couture, A.; Daly, J.; Del Santo, M.; Elliot, T.

    2010-01-01

    The neutron long counter NERO was built at the National Superconducting Cyclotron Laboratory (NSCL), Michigan State University, for measuring β-delayed neutron-emission probabilities. The detector was designed to work in conjunction with a β-delay implantation station, so that β decays and β-delayed neutrons emitted from implanted nuclei can be measured simultaneously. The high efficiency of about 40%, for the range of energies of interest, along with the small background, are crucial for measuring β-delayed neutron emission branchings for neutron-rich r-process nuclei produced as low intensity fragmentation beams in in-flight separator facilities.

  18. The energy spectrum of delayed neutrons from thermal neutron induced fission of 235U and its analytical approximation

    International Nuclear Information System (INIS)

    Doroshenko, A.Yu.; Tarasko, M.Z.; Piksaikin, V.M.

    2002-01-01

    The energy spectrum of the delayed neutrons is the poorest known of all input data required in the calculation of the effective delayed neutron fractions. In addition to delayed neutron spectra based on the aggregate spectrum measurements there are two different approaches for deriving the delayed neutron energy spectra. Both of them are based on the data related to the delayed neutron spectra from individual precursors of delayed neutrons. In present work these two different data sets were compared with the help of an approximation by gamma-function. The choice of this approximation function instead of the Maxwellian or evaporation type of distribution is substantiated. (author)

  19. Deterministic calculation of the effective delayed neutron fraction without using the adjoint neutron flux - 299

    International Nuclear Information System (INIS)

    Talamo, A.; Gohar, Y.; Aliberti, G.; Zhong, Z.; Bournos, V.; Fokov, Y.; Kiyavitskaya, H.; Routkovskaya, C.; Serafimovich, I.

    2010-01-01

    In 1997, Bretscher calculated the effective delayed neutron fraction by the k-ratio method. The Bretscher's approach is based on calculating the multiplication factor of a nuclear reactor core with and without the contribution of delayed neutrons. The multiplication factor set by the delayed neutrons (the delayed multiplication factor) is obtained as the difference between the total and the prompt multiplication factors. Bretscher evaluated the effective delayed neutron fraction as the ratio between the delayed and total multiplication factors (therefore the method is often referred to as k-ratio method). In the present work, the k-ratio method is applied by deterministic nuclear codes. The ENDF/B nuclear data library of the fuel isotopes ( 238 U and 238 U) have been processed by the NJOY code with and without the delayed neutron data to prepare multigroup WIMSD nuclear data libraries for the DRAGON code. The DRAGON code has been used for preparing the PARTISN macroscopic cross sections. This calculation methodology has been applied to the YALINA-Thermal assembly of Belarus. The assembly has been modeled and analyzed using PARTISN code with 69 energy groups and 60 different material zones. The deterministic and Monte Carlo results for the effective delayed neutron fraction obtained by the k-ratio method agree very well. The results also agree with the values obtained by using the adjoint flux. (authors)

  20. Reduction of delayed-neutron contribution to variance-to-mean ratio by application of difference filter technique

    International Nuclear Information System (INIS)

    Hashimoto, Kengo; Mouri, Tomoaki; Ohtani, Nobuo

    1999-01-01

    The difference-filtering correlation analysis was applied to time-sequence neutron count data measured in a slightly subcritical assembly, where the Feynman-α analysis suffered from large contribution of delayed neutron to the variance-to-mean ratio of counts. The prompt-neutron decay constant inferred from the present filtering analysis agreed very closely with that by pulsed neutron experiment, and no dependence on the gate-time range specified could be observed. The 1st-order filtering was sufficient for the reduction of the delayed-neutron contribution. While the conventional method requires a choice of analysis formula appropriate to a gate-time range, the present method is applicable to a wide variety of gate-time ranges. (author)

  1. 8-group relative delayed neutron yields for monoenergetic neutron induced fission of 239Pu

    International Nuclear Information System (INIS)

    Piksaikin, V.M.; Kazakov, L.E.; Isaev, S.G.; Korolev, G.G.; Roshchenko, V.A.; Tertychnyj, R.G

    2002-01-01

    The energy dependence of the relative yield of delayed neutrons in an 8-group model representation was obtained for monoenergetic neutron induced fission of 239 Pu. A comparison of this data with the available experimental data by other authors was made in terms of the mean half-life of the delayed neutron precursors. (author)

  2. Study of beta-delayed neutron with proton-neutron QRPA plus statistical model

    International Nuclear Information System (INIS)

    Minato, Futoshi; Iwamoto, Osamu

    2015-01-01

    β-delayed neutron is known to be important for safety operation of nuclear reactor and prediction of elemental abundance after freeze-out of r-process. A lot of researches on it have been performed. However, the experimental data are far from complete since the lifetime of most of the relevant nuclei is so short that one cannot measure in a high efficiency. In order to estimate half-lives and delayed neutron emission probabilities of unexplored nuclei, we developed a new theoretical method which combines a proton-neutron quasi-particle random-phase-approximation and the Hauser-Feshbach statistical model. The present method reproduces experimentally known β-decay half-lives within a factor of 10 and about 40% of within a factor of 2. However it fails to reproduce delayed neutron emission probabilities. We discuss the problems and remedy for them to be made in future. (author)

  3. Delayed neutron emission near the shell-closures

    Directory of Open Access Journals (Sweden)

    Borzov Ivan

    2016-01-01

    Full Text Available The self-consistent Density Functional + Continuum QRPA approach (DF+CQRPA provides a good description of the recent experimental beta-decay half-lives and delayed neutron emission branchings for the nuclei approaching to (and beyond the neutron closed shells N = 28; 50; 82. Predictions of beta-decay properties are more reliable than the ones of standard global approaches traditionally used for the r-process modelling. An impact of the quasi-particle phonon coupling on the delayed multi-neutron emission rates P2n, P3n,… near the closed shells is also discussed.

  4. Possibilities of delayed neutron fraction (βeff) calculation and measurement

    International Nuclear Information System (INIS)

    Michalek, S.; Hascik, J.; Farkas, G.

    2008-01-01

    The influence of the delayed neutrons on the reactor dynamics can be understood through their impact on the reactor power change rate. In spite of the fact that delayed neutrons constitute only a very small fraction of the total number of neutrons generated from fission, they play a dominant role in the fission chain reaction control. If only the prompt neutrons existed, the reactor operation would become impossible due to the fast reactor power changes. The exact determination of delayed neutrons main parameter, the delayed neutron fraction (β eff ), is very important in the field of reactor physics. The interest in the delayed neutron data accuracy improvement started to increase at the end of 80-ties and the beginning of 90-ties, after discrepancies among the results of calculations and experiments. In consequence of difficulties in β eff experimental measurement, this value in exact state use to be determined by calculations. Subsequently, its reliability depends on the calculation method and the delayed neutron data used. Determination of β eff requires criticality calculations. In the past, k eff used to be traditionally calculated by taking the ratio of the adjoint- and spectrum-weighted delayed neutron production rate to the adjoint- and spectrum- weighted total neutron production rate. An alternative method has also been used in which β eff is calculated from simple k-eigenvalue solutions. In this work, a summary of possible β eff calculation methods can be found and a calculation of β eff for VR-1 training reactor in one operation state is made using the prompt method, by MCNP5 code. Also a method of β eff kinetic measurement on VR-1 training reactor at Czech Technical University in Prague using in-pile kinetic technique is outlined (authors)

  5. Importance of delayed neutron data in transmutation system

    International Nuclear Information System (INIS)

    Tsujimoto, Kazufumi

    1999-01-01

    The accelerator-driven transmutation system has been studied at the Japan Atomic Energy Research Institute. This system is a hybrid system which consists of a high intensity accelerator, a spallation target and a subcritical core region. The subcritical core is driven by neutrons generated by spallation reaction in the target region. There is no control rod in this system, so the power is controlled only by proton beam current. The beam current to keep constant power change with effective multiplication factor of subcritical core. So, the evaluation of delayed neutron fraction which is strongly connected to the measurement of subcritical level is important factor in operation of accelerator-driven system. In this paper, important nuclides for the delayed neutron fraction of ADS will be discussed, moreover, present state of delayed neutron data in evaluated nuclear data library is presented. (author)

  6. Subcritical Neutron Multiplication Measurements of HEU Using Delayed Neutrons as the Driving Source

    International Nuclear Information System (INIS)

    Hollas, C.L.; Goulding, C.A.; Myers, W.L.

    1999-01-01

    A new method for the determination of the multiplication of highly enriched uranium systems is presented. The method uses delayed neutrons to drive the HEU system. These delayed neutrons are from fission events induced by a pulsed 14-MeV neutron source. Between pulses, neutrons are detected within a medium efficiency neutron detector using 3 He ionization tubes within polyethylene enclosures. The neutron detection times are recorded relative to the initiation of the 14-MeV neutron pulse, and subsequently analyzed with the Feynman reduced variance method to extract singles, doubles and triples neutron counting rates. Measurements have been made on a set of nested hollow spheres of 93% enriched uranium, with mass values from 3.86 kg to 21.48 kg. The singles, doubles and triples counting rates for each uranium system are compared to calculations from point kinetics models of neutron multiplicity to assign multiplication values. These multiplication values are compared to those from MC NP K-Code calculations

  7. Elemental analysis of some West Malaysian limestones using neutron activation, delayed neutron and electron microprobe analysis

    International Nuclear Information System (INIS)

    Amin, Y.M.; Kamaluddin, B.; Mahat, R.H.

    1990-01-01

    Limestone stratigraphy in Malaysia has been and is dependent almost entirely in palaeontology. However fossil localities are sporadic and as such a new fossil discovery mean the necessity for a complete re-appraisal of the stratigraphy. The almost complete dependence upon palaeontology results from the difficulties of stratigraphy correlation of isolated outcrops, from the cover of tropical vegetation and from the often complex folding and faulting which has been imposed on the geosyn-clinical rocks by the Indonesian-Thai-Malayan orogeny. So by studying the elemental composition of limestones accurately, we would be able to correlate outcrops and other stratigraphic samples independent of fossil finds. The use of delayed neutron analysis would also determine the concentration of uranium and thorium accurately. This study, in conjunction with thermoluminescence and fission track studies, would able us to date the age of the limestones

  8. Statistical theory for calculating energy spectra of β-delayed neutrons

    International Nuclear Information System (INIS)

    Kawano, Toshihiko; Moeller, Peter; Wilson, William B.

    2008-01-01

    Theoretical β-delayed neutron spectra are calculated based on the Quasi-particle Random Phase Approximation (QRPA) and the Hauser-Feshbach statistical model. Neutron emissions from an excited daughter nucleus after β-decay to the granddaughter residual are more accurately calculated than previous evaluations, including all the microscopic nuclear structure information, such as a Gamow-Teller strength distribution and discrete states in the granddaughter. The calculated delayed-neutron spectra reasonably agree with those evaluations in the ENDF decay library, which are based on experimental data. The model was adopted to generate the delayed-neutron spectra for all 271 precursors. (authors)

  9. Energy dependence of relative abundances and periods of delayed neutron separate groups from neutron induced fission of 239Pu in the virgin neutron energy range 0.37-4.97 MeV

    International Nuclear Information System (INIS)

    Piksajkin, V.M.; Kazakov, L.E.; Isaev, S.T.; Korolev, G.G.; Roshchenko, V.A.; Tertychnyj, R.G.

    2002-01-01

    Relative yield and group period of delayed neutrons induced by the 239 Pu fission in the 0.37-4.97 MeV range were measured. Comparative analysis of experimental data was conducted in terms of middle period of half-life of delayed neutron nuclei-precursors. Character and scale of changing values of delayed neutron group parameters as changing excitation energy of fission compound-nucleus have been demonstrated for the first time. Considerable energy dependence of group parameters under the neutron induced 239 Pu fission that was expressed by the decreasing middle period of half-life of nuclei-precursors by 10 % in the 2.85 eV - 5 MeV range of virgin neutrons was detected [ru

  10. Evaluation method for uncertainty of effective delayed neutron fraction βeff

    International Nuclear Information System (INIS)

    Zukeran, Atsushi

    1999-01-01

    Uncertainty of effective delayed neutron fraction β eff is evaluated in terms of three quantities; uncertainties of the basic delayed neutron constants, energy dependence of delayed neutron yield ν d m , and the uncertainties of the fission cross sections of fuel elements. The uncertainty of β eff due to the delayed neutron yield is expressed by a linearized formula assuming that the delayed neutron yield does not depend on the incident energy, and the energy dependence is supplemented by using the detailed energy dependence proposed by D'Angelo and Filip. The third quantity, uncertainties of fission cross section, is evaluated on the basis of the generalized perturbation theory in relation to reaction rate rations such as central spectral indexes or average reaction rate ratios. Resultant uncertainty of β eff is about 4 to 5%s, in which primary factor is the delayed neutron yield, and the secondary one is the fission cross section uncertainty, especially for 238 U. The energy dependence of ν d m systematically reduces the magnitude of β eff about 1.4% to 1.7%, depending on the model of the energy vs. ν d m correlation curve. (author)

  11. Modeling delayed neutron monitoring systems for fast breeder reactors

    International Nuclear Information System (INIS)

    Bunch, W.L.; Tang, E.L.

    1983-10-01

    The purpose of the present work was to develop a general expression relating the count rate of a delayed neutron monitoring system to the introduction rate of fission fragments into the sodium coolant of a fast breeder reactor. Most fast breeder reactors include a system for detecting the presence of breached fuel that permits contact between the sodium coolant and the mixed oxide fuel. These systems monitor for the presence of fission fragments in the sodium that emit delayed neutrons. For operational reasons, the goal is to relate the count rate of the delayed neutron monitor to the condition of the breach in order that appropriate action might be taken

  12. Two reports: (i) Correlation properties of delayed neutrons from fast neutron induced fission. (ii) Method and set-up for measurements of trace level content of heavy fissionable elements based on delayed neutron counting

    International Nuclear Information System (INIS)

    Piksaikin, V.M.; Isaev, S.G.; Goverdovski, A.A.; Pshakin, G.M.

    1998-10-01

    The document includes the following two reports: 'Correlation properties of delayed neutrons from fast neutron induced fission' and 'Method and set-up for measurements of trace level content of heavy fissionable elements based on delayed neutron counting. A separate abstract was prepared for each report

  13. Recent activities for β-decay half-lives and β-delayed neutron emission of very neutron-rich isotopes

    Energy Technology Data Exchange (ETDEWEB)

    Dillmann, Iris [TRIUMF, Vancouver BC, V6T 2A3, Canada and GSI Helmholtzzentrum für Schwerionenforschung GmbH, D-64291 Darmstadt (Germany); Abriola, Daniel [Laboratorio Tandar, Comisión Nacional de Energía Atómica, B1650KINA, San Martín, Buenos Aires (Argentina); Singh, Balraj [Department of Physics and Astronomy, McMaster University, Hamilton ON, L8S 4M1 (Canada)

    2014-05-02

    Beta-delayed neutron (βn) emitters play an important, two-fold role in the stellar nucleosynthesis of heavy elements in the 'rapid neutron-capture process' (r process). On one hand they lead to a detour of the material β-decaying back to stability. On the other hand, the released neutrons increase the neutron-to-seed ratio, and are re-captured during the freeze-out phase and thus influence the final solar r-abundance curve. A large fraction of the isotopes inside the r-process reaction path are not yet experimentally accessible and are located in the (experimental) 'Terra Incognita'. With the next generation of fragmentation and ISOL facilities presently being built or already in operation, one of the main motivation of all projects is the investigation of these very neutron-rich isotopes. A short overview of one of the planned programs to measure βn-emitters at the limits of the presently know isotopes, the BRIKEN campaign (Beta delayed neutron emission measurements at RIKEN) will be given. Presently, about 600 β-delayed one-neutron emitters are accessible, but only for a third of them experimental data are available. Reaching more neutron-rich isotopes means also that multiple neutron-emission becomes the dominant decay mechanism. About 460 β-delayed two-, three-or four-neutron emitters are identified up to now but for only 30 of them experimental data about the neutron branching ratios are available, most of them in the light mass region below A=30. The International Atomic and Energy Agency (IAEA) has identified the urgency and picked up this topic recently in a 'Coordinated Research Project' on a 'Reference Database for Beta-Delayed Neutron Emission Data'. This project will review, compile, and evaluate the existing data for neutron-branching ratios and half-lives of β-delayed neutron emitters and help to ensure a reliable database for the future discoveries of new isotopes and help to constrain astrophysical and

  14. Proposal for Analysis of the Safeguarded Nuclear Materials 235U and 239Pu by Delayed Neutrons Technique

    International Nuclear Information System (INIS)

    El-Mongy, S.A.

    2000-01-01

    This paper introduces, describes and initiates a very sensitive and rapid non-destructive technique to be used for analysis of the safeguarded nuclear materials 235 U and 239 Pu. The technique is based on fission of the nuclear material by neutrons and then measuring the delayed neutrons produced from the neutron rich fission products. By this technique, fissile isotope content ( 235 U) can be determined in the presence of the other fissile (e.g. 239 Pu) or fertile isotopes (e.g. 238 U) in fresh and spent fuel. The time consumed for analysis of bulk materials by this technique is only 4 minutes. The method is also used for analysis of uranium in rock, sediment, soil, meteorites, lunar, biological, urine, archaeological, zircon sand and seawater samples. The method enables uranium in a sample to be measured without respect to its oxidation state, organic and inorganic elements

  15. Comparison of dynamic compensation methods for delayed self-powered neutron detector

    International Nuclear Information System (INIS)

    In, Wang Kee; Kim, Joon Sung; Auh, Geun Sun; Yoon, Tae Young

    1993-01-01

    Dynamic compensation methods for rhodium self-powered neutron detector have been developed by Banda and Hoppe to compensate for the time delay associated with detector signals. The time delay is due to the decay of the neutron-activated rhodium and results in delayed detector response. Two digital dynamic compensation methods, were compared for step change of neutron flux in this paper. The inverse kinetics method gave slightly better response time and noise gain. However, the inverse kinetics method also showed overshooting of neutron flux for the step change. (Author)

  16. Delayed neutron spectra and their uncertainties in fission product summation calculations

    Energy Technology Data Exchange (ETDEWEB)

    Miyazono, T.; Sagisaka, M.; Ohta, H.; Oyamatsu, K.; Tamaki, M. [Nagoya Univ. (Japan)

    1997-03-01

    Uncertainties in delayed neutron summation calculations are evaluated with ENDF/B-VI for 50 fissioning systems. As the first step, uncertainty calculations are performed for the aggregate delayed neutron activity with the same approximate method as proposed previously for the decay heat uncertainty analyses. Typical uncertainty values are about 6-14% for {sup 238}U(F) and about 13-23% for {sup 243}Am(F) at cooling times 0.1-100 (s). These values are typically 2-3 times larger than those in decay heat at the same cooling times. For aggregate delayed neutron spectra, the uncertainties would be larger than those for the delayed neutron activity because much more information about the nuclear structure is still necessary. (author)

  17. JENDL-4.0 benchmarking for effective delayed neutron fraction with a continuous-energy Monte Carlo code MVP

    International Nuclear Information System (INIS)

    Nagaya, Yasunobu

    2013-01-01

    Benchmark calculations with a continuous-energy Monte Carlo code have been performed for delayed neutron data of JENDL-4.0. JENDL-4.0 gives good prediction for the effective delayed neutron fraction in the present benchmarks but further detailed analysis is required for some cores. (author)

  18. Delayed neutrons as a probe of nuclear charge distribution in fission of heavy nuclei by neutrons

    International Nuclear Information System (INIS)

    Isaev, S.G.; Piksaikin, V.M.; Kazakov, L.E.; Roshchenko, V.A.

    2002-01-01

    A method of the determination of cumulative yields of delayed neutron precursors is developed. This method is based on the iterative least-square procedure applied to delayed neutron decay curves measured after irradiation of 235 U sample by thermal neutrons. Obtained cumulative yields in turns were used for deriving the values of the most probable charge in low-energy fission of the above-mentioned nucleus. (author)

  19. A possible island of beta-delayed neutron precursors in heavy nucleus region

    International Nuclear Information System (INIS)

    Zhang Li

    1991-01-01

    The possible Beta-Delayed neutron precursors in the elements Tl, Hg, and Au were predicted following a systematic research on the known Beta-Delayed neutron precursors. The masses of the unknown nuclei and neutron emission probabilities were calculated

  20. 8-group relative delayed neutron yields for epithermal neutron induced fission of 235U and 239Pu

    International Nuclear Information System (INIS)

    Piksaikin, V.M.; Kazakov, L.E.; Isaev, S.G.; Korolev, G.G.; Roshchenko, V.A.; Tertychnyj, R.G

    2002-01-01

    An 8-group representation of relative delayed neutron yields was obtained for epithermal neutron induced fission of 235 U and 239 Pu. These data were compared with ENDF/B-VI data in terms of the average half- life of the delayed neutron precursors and on the basis of the dependence of reactivity on the asymptotic period. (author)

  1. Leakage monitoring equipment of fuel element by delayed neutron method

    International Nuclear Information System (INIS)

    Ji Changsong; Zhang Shulan; Zhang Shuheng

    1999-01-01

    Based on monitoring results of delayed neutrons from reactor first circle water, the leakage of reactor fuel elements is monitored. A monitoring equipment consisted of an array of 3 He proportional counter tubes with 75 s delay has been developed. The neutron detection efficiency of 6.1% is obtained

  2. Delayed neutrons as a probe of nuclear charge distribution in fission of heavy nuclei by neutrons

    CERN Document Server

    Isaev, S G; Piksaikin, V M; Roshchenko, V A

    2001-01-01

    A method of the determination of cumulative yields of delayed neutron precursors is developed. This method is based on the iterative least-square procedure applied to delayed neutron decay curves measured after irradiation of sup 2 sup 3 sup 5 U sample by thermal neutrons. Obtained cumulative yields in turns were used for deriving the values of the most probable charge in low-energy fission of the above-mentioned nucleus.

  3. Non-destructive isotopic uranium assay by multiple delayed neutron measurements

    International Nuclear Information System (INIS)

    Papadopoulos, N.N.; Tsagas, N.F.

    1991-01-01

    The high accuracy and precision required in nuclear safeguards measurements can be achieved by an improved neutron activation technique based on multiple delayed fission neutron counting under various experimental conditions. For the necessary ultrahigh counting statistics required, cyclic activation of multiple subsamples has been applied. The home-made automated flexible analytical system with neutron flux and spectrum differentiation by irradiation position adjustment and cadmium screening, permits the non-destructive determination of the U235 abundance and the total U element concentration needed in nuclear safeguards sample analysis, with a high throughout and a low operational cost. Careful experimental optimization led to considerable improvement of the results

  4. Some properties of zero power neutron noise in a time-varying medium with delayed neutrons

    International Nuclear Information System (INIS)

    Kitamura, Y.; Pal, L.; Pazsit, I.; Yamamoto, A.; Yamane, Y.

    2008-01-01

    The temporal evolution of the distribution of the number of neutrons in a time-varying multiplying system, producing only prompt neutrons, was treated recently with the master equation technique by some of the present authors. Such a treatment gives account of both the so-called zero power reactor noise and the power reactor noise simultaneously. In particular, the first two moments of the neutron number, as well as the concept of criticality for time-varying systems, were investigated and discussed. The present paper extends these investigations to the case when delayed neutrons are also taken into account. Due to the complexity of the description, only the expectation of the neutron number is calculated. The concept of criticality of a time-varying system is also generalized to systems with delayed neutrons. The temporal behaviour of the expectation of the number of neutrons and its asymptotic properties are displayed and discussed

  5. Calibration of the JET neutron yield monitors using the delayed neutron counting technique

    International Nuclear Information System (INIS)

    van Belle, P.; Jarvis, O.N.; Sadler, G.; de Leeuw, S.; D'Hondt, P.; Pillon, M.

    1990-01-01

    The time-resolved neutron yield is routinely measured on the JET tokamak using a set of fission chambers. At present, the preferred technique is to employ activation reactions to determine the neutron fluence at a well-chosen position and to relate the measured fluence to the total neutron emission by means of neutron transport calculations. The delayed neutron counting method is a particularly convenient method of performing the activation measurement and the fission cross sections are accurately known. This paper outlines the measurement technique as used on JET

  6. Computer-automated neutron activation analysis system

    International Nuclear Information System (INIS)

    Minor, M.M.; Garcia, S.R.

    1983-01-01

    An automated delayed neutron counting and instrumental neutron activation analysis system has been developed at Los Alamos National Laboratory's Omega West Reactor (OWR) to analyze samples for uranium and 31 additional elements with a maximum throughput of 400 samples per day. 5 references

  7. An accurate solution of point reactor neutron kinetics equations of multi-group of delayed neutrons

    International Nuclear Information System (INIS)

    Yamoah, S.; Akaho, E.H.K.; Nyarko, B.J.B.

    2013-01-01

    Highlights: ► Analytical solution is proposed to solve the point reactor kinetics equations (PRKE). ► The method is based on formulating a coefficient matrix of the PRKE. ► The method was applied to solve the PRKE for six groups of delayed neutrons. ► Results shows good agreement with other traditional methods in literature. ► The method is accurate and efficient for solving the point reactor kinetics equations. - Abstract: The understanding of the time-dependent behaviour of the neutron population in a nuclear reactor in response to either a planned or unplanned change in the reactor conditions is of great importance to the safe and reliable operation of the reactor. In this study, an accurate analytical solution of point reactor kinetics equations with multi-group of delayed neutrons for specified reactivity changes is proposed to calculate the change in neutron density. The method is based on formulating a coefficient matrix of the homogenous differential equations of the point reactor kinetics equations and calculating the eigenvalues and the corresponding eigenvectors of the coefficient matrix. A small time interval is chosen within which reactivity relatively stays constant. The analytical method was applied to solve the point reactor kinetics equations with six-groups delayed neutrons for a representative thermal reactor. The problems of step, ramp and temperature feedback reactivities are computed and the results compared with other traditional methods. The comparison shows that the method presented in this study is accurate and efficient for solving the point reactor kinetics equations of multi-group of delayed neutrons

  8. Study of Beta-Delayed Neutron Emission by Neutron-Rich Nuclei and Analysis of the Nuclear Reaction Mechanism responsible for the Yields of these Nuclei

    International Nuclear Information System (INIS)

    Bazin, D.

    1987-07-01

    Among the nuclear mechanisms used for the production of nuclei far from stability, the projectile fragmentation process has recently proved its efficiency. However, at Fermi energies, one has to take into account some collective and relaxation effects which drastically modify the production cross-sections. The spectroscopic study of very neutron-rich nuclei is very dependent of these production rates. A study of beta-delayed neutron emission which leads to new measurements of half-lives and neutron delayed emission probabilities is achieved with a liquid scintillator detector. The results which are then compared to different theories are of interest for the understanding of natural production of heavy elements (r processus) [fr

  9. Delayed Neutron Fraction (beta-effective) Calculation for VVER 440 Reactor

    International Nuclear Information System (INIS)

    Hascik, J.; Michalek, S.; Farkas, G.; Slugen, V.

    2008-01-01

    Effective delayed neutron fraction (β eff ) is the main parameter in reactor dynamics. In the paper, its possible determination methods are summarized and a β eff calculation for a VVER 440 power reactor as well as for training reactor VR1 using stochastic transport Monte Carlo method based code MCNP5 is made. The uncertainties in determination of basic delayed neutron parameters lead to the unwished conservatism in the reactor control system design and operation. Therefore, the exact determination of the β eff value is the main requirement in the field of reactor dynamics. The interest in the delayed neutron data accuracy improvement started to increase at the end of 80-ties and the beginning of 90-ties, after discrepancies among the results of experiments and measurements what do you mean differences between different calculation approaches and experimental results. In consequence of difficulties in β eff experimental measurement, this value in exact state is determined by calculations. Subsequently, its reliability depends on the calculation method and the delayed neutron data used. An accurate estimate of β eff is essential for converting reactivity, as measured in dollars, to an absolute reactivity and/or to an absolute k eff . In the past, k eff has been traditionally calculated by taking the ratio of the adjoint- and spectrum-weighted delayed neutron production rate to the adjoint- and spectrum-weighted total neutron production rate. An alternative method has also been used in which β eff is calculated from simple k-eigenvalue solutions. The summary of the possible β eff determination methods can be found in this work and also a calculation of β eff first for the training reactor VR1 in one operation state and then for VVER 440 power reactor in two different operation states are made using the prompt method, by MCNP5 code.(author)

  10. Total body-calcium measurements: comparison of two delayed-gamma neutron activation facilities

    International Nuclear Information System (INIS)

    Ma, R.; Ellis, K.J.; Shypailo, R.J.; Pierson, R.N. Jr.

    1999-01-01

    This study compares two independently calibrated delayed-gamma neutron activation (DGNA) facilities, one at the Brookhaven National Laboratory (BNL), Upton, New York, and the other at the Children's Nutrition Research Center (CNRC), Houston, Texas that measure total body calcium (TBCa). A set of BNL phantoms was sent to CNRC for neutron activation analysis, and a set of CNRC phantoms was measured at BNL. Both facilities showed high precision (<2%), and the results were in good agreement, within 5%. (author)

  11. Measurement of delayed neutron-emitting fission products in nuclear reactor coolant water during reactor operation

    International Nuclear Information System (INIS)

    Anon.

    1981-01-01

    The method covers the detection and measurement of delayed neutron-emitting fission products contained in nuclear reactor coolant water while the reactor is operating. The method is limited to the measurement of the delayed neutron-emitting bromine isotope of mass 87 and the delayed neutron-emitting iodine isotope of mass 137. The other delayed neutron-emitting fission products cannot be accurately distinguished from nitrogen 17, which is formed under some reactor conditions by neutron irradiation of the coolant water molecules. The method includes a description of significance, measurement variables, interferences, apparatus, sampling, calibration, standardization, sample measurement procedures, system efficiency determination, calculations, and precision

  12. Influence of delayed neutron parameter calculation accuracy on results of modeled WWER scram experiments

    International Nuclear Information System (INIS)

    Artemov, V.G.; Gusev, V.I.; Zinatullin, R.E.; Karpov, A.S.

    2007-01-01

    Using modeled WWER cram rod drop experiments, performed at the Rostov NPP, as an example, the influence of delayed neutron parameters on the modeling results was investigated. The delayed neutron parameter values were taken from both domestic and foreign nuclear databases. Numerical modeling was carried out on the basis of SAPFIR 9 5andWWERrogram package. Parameters of delayed neutrons were acquired from ENDF/B-VI and BNAB-78 validated data files. It was demonstrated that using delay fraction data from different databases in reactivity meters led to significantly different reactivity results. Based on the results of numerically modeled experiments, delayed neutron parameters providing the best agreement between calculated and measured data were selected and recommended for use in reactor calculations (Authors)

  13. Prompt and delay gamma ray measurements for 'in vivo' neutron activation analysis using a cyclic system

    International Nuclear Information System (INIS)

    Matthews, I.P.

    1979-09-01

    Early attempts at determining the elemental composition of the body by radioactive isotope dilution techniques are reviewed. The development and current status of in-vivo neutron activation analysis and the ways in which it supersedes or supplements certain of the former techniques are outlined. An irradiation facility is described which employs a 5 Ci neutron source and is capable of performing prompt and delay γ-ray measurements as well as cyclic activation. The uniformity of thermal neutron flux in a phantom is demonstrated and the neutron spectrum at a depth in the phantom has been obtained by means of threshold detectors. An examination is made of the possible applications of the Monte Carlo method to the design of irradiation and detection facilities and in yielding information about inaccessible areas. Detection limits for the bulk body elements and trace elements are presented. It is shown that the depth of a region of the body can be determined from a prompt gamma ray spectrum. This technique can be used to correct measurements when it is known that activation and detection is non-uniform. The feasibility of using a C.T. whole body scanner to measure bone demineralisation is explored. (author)

  14. 235U Determination using In-Beam Delayed Neutron Counting Technique at the NRU Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Andrews, M. T. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Bentoumi, G. [Canadian Nuclear Labs., Chalk River, ON (Canada); Corcoran, E. C. [Royal Military College of Canada, Kingston, ON (United States); Dimayuga, I. [Canadian Nuclear Labs., Chalk River, ON (Canada); Kelly, D. G. [Royal Military College of Canada, Kingston, ON (United States); Li, L. [Canadian Nuclear Labs., Chalk River, ON (Canada); Sur, B. [Canadian Nuclear Labs., Chalk River, ON (Canada); Rogge, R. B. [Canadian Nuclear Labs., Chalk River, ON (Canada)

    2015-11-17

    This paper describes a collaborative effort that saw the Royal Military College of Canada (RMC)’s delayed neutron and gamma counting apparatus transported to Canadian Nuclear Laboratories (CNL) for use in the neutron beamline at the National Research Universal (NRU) reactor. Samples containing mg quantities of fissile material were re-interrogated, and their delayed neutron emissions measured. This collaboration offers significant advantages to previous delayed neutron research at both CNL and RMC. This paper details the determination of 235U content in enriched uranium via the assay of in-beam delayed neutron magnitudes and temporal behavior. 235U mass was determined with an average absolute error of ± 2.7 %. This error is lower than that obtained at RMCC for the assay of 235U content in aqueous solutions (3.6 %) using delayed neutron counting. Delayed neutron counting has been demonstrated to be a rapid, accurate, and precise method for special nuclear material detection and identification.

  15. Calculation of the pulsed Feynman- and Rossi-alpha formulae with delayed neutrons

    International Nuclear Information System (INIS)

    Kitamura, Y.; Pazsit, I.; Wright, J.; Yamamoto, A.; Yamane, Y.

    2005-01-01

    In previous works, the authors have developed an effective solution technique for calculating the pulsed Feynman- and Rossi-alpha formulae. Through derivation of these formulae, it was shown that the technique can easily handle various pulse shapes of the pulsed neutron source. Furthermore, it was also shown that both the deterministic (i.e., synchronizing with the pulsing of neutron source) and stochastic (non-synchronizing) Feynman-alpha formulae can be obtained with this solution technique. However, for mathematical simplicity and the sake of insight, the formal derivation was performed in a model without delayed neutrons. In this paper, to demonstrate the robustness of the technique, the pulsed Feynman- and Rossi-alpha formulae were re-derived by taking one group of delayed neutrons into account. The results show that the advantages of this technique are retained even by inclusion of the delayed neutrons. Compact explicit formulae are derived for the Feynman- and Rossi-alpha methods for various pulse shapes and pulsing methods

  16. Nondestructive analysis of the natural uranium mass through the measurement of delayed neutrons using the technique of pulsed neutron source; Analise nao destrutiva da massa de uranio natural atraves da medida de neutrons atrasados com o uso da tecnica de fonte pulsada de neutrons rapidos

    Energy Technology Data Exchange (ETDEWEB)

    Coelho, Paulo Rogerio Pinto

    1979-07-01

    This work presents results of non destructive mass analysis of natural uranium by the pulsed source technique. Fissioning is produced by irradiating the test sample with pulses of 14 MeV neutrons and the uranium mass is calculated on a relative scale from the measured emission of delayed neutrons. Individual measurements were normalised against the integral counts of a scintillation detector measuring the 14 MeV neutron intensity. Delayed neutrons were measured using a specially constructed slab detector operated in anti synchronism with the fast pulsed source. The 14 MeV neutrons were produced via the T(d,n) {sup 4}He reaction using a 400 kV Van de Graaff accelerated operated at 200 kV in the pulsed source mode. Three types of sample were analysed, namely: discs of metallic uranium, pellets of sintered uranium oxide and plates of uranium aluminium alloy sandwiched between aluminium. These plates simulated those of Material Testing Reactor fuel elements. Results of measurements were reproducible to within an overall error in the range 1.6 to 3.9%; the specific error depending on the shape, size and mass of the sample. (author)

  17. Energy dependence of average half-life of delayed neutron precursors in fast neutron induced fission of 235U and 236U

    International Nuclear Information System (INIS)

    Isaev, S.G.; Piksaikin, L.E.; Kazakov, L.E.; Tarasko, M.Z.

    2000-01-01

    The measurements of relative abundances and periods of delayed neutrons from fast neutron induced fission of 235 U and 236 U have been made at the electrostatic accelerator CG-2.5 at IPPE. The preliminary results were obtained and discussed in the frame of the systematics of the average half-life of delayed neutron precursors. It was shown that the average half-life value in both reactions depends on the energy of primary neutrons [ru

  18. Delayed neutron kinetic functions for /sup 232/Th and /sup 238/U mixtures

    Energy Technology Data Exchange (ETDEWEB)

    Ganich, P P; Goshovskij, M V; Lendel, A I; Lomonosov, V I; Sikora, D I; Sychev, S I

    1984-11-01

    In order to investigate the applicability of the method based on using kinetic functions, describing the emission of delayed neutrons by samples for determination of the content of fissionable nuclides in binary mixtures, the /sup 232/Th+/sup 238/U mixtures have been analyzed with the M-30 microtron. Fresh samples containing ThO/sub 2/, U/sub 3/O/sub 8/ and their mixtures are irradiated by bremstrahlung at the 15.5 MeV energy of accelerated electrons and 9 ..mu..A average current. The mass of samples is about 6 g. To determine the kinetic functions, temporal distributions of delayed neutron pulses are used, their maximum number for different samples being (1.7-3.0) x 10/sup 4/. In processing the data obtained two methods of normalization of the delayed neutron number in the kinetic functions are used: to the total yield of delayed neutrons and to the yield of /sup 133/I ..gamma..-quanta. The conclusion is drawn that the method investigated permits to determine relative /sup 238/U concentrations in the mixtures considered with 0.06-0.2 errors. Error reduction is achieved during the normalization of the number of delayed neutrons to the yield of /sup 130/I ..gamma..-quanta.

  19. MONSTER: a TOF Spectrometer for beta-delayed Neutron Spetroscopy

    CERN Document Server

    Martinez, T; Castilla, J; Garcia, A R; Marin, J; Martinez, G; Mendoza, E; Santos, C; Tera, F; Jordan, M D; Rubio, B; Tain, J L; Bhattacharya, C; Banerjee, K; Bhattacharya, S; Roy, P; Meena, J K; Kundu, S; Mukherjee, G; Ghosh, T K; Rana, T K; Pandey, R; Saxena, A; Behera, B; Penttila, H; Jokinen, A; Rinta-Antila, S; Guerrero, C; Ovejero, M C; Villamarin, D; Agramunt, J; Algora, A

    2014-01-01

    Beta-delayed neutron (DN) data, including emission probabilities, P-n, and energy spectrum, play an important role in our understanding of nuclear structure, nuclear astrophysics and nuclear technologies. A MOdular Neutron time-of-flight SpectromeTER (MONSTER) is being built for the measurement of the neutron energy spectra and branching ratios. The TOF spectrometer will consist of one hundred liquid scintillator cells covering a significant solid angle. The MONSTER design has been optimized by using Monte Carlo (MC) techniques. The response function of the MONSTER cell has been characterized with mono-energetic neutron beams and compared to dedicated MC simulations.

  20. Atlantic Richfield Hanford Company californium multiplier/delayed neutron counter safety analysis

    International Nuclear Information System (INIS)

    Zimmer, W.H.

    1976-08-01

    The Californium Multiplier (CFX) is a subcritical assembly of uranium surrounding 252 Cf spontaneously fissioning neutron sources; its function is to multiply the neutron flux to a level useful for activation analysis. This document summarizes the safety analysis aspects of the CFX, DNC, pneumatic transfer system, and instrumentation and to detail all the aspects of the total facility as a starting point for the ARHCO Safety Analysis Review. Recognized hazards and steps already taken to neutralize them are itemized

  1. Determination of the effective delayed neutron fraction in the Coral-I Reactor

    International Nuclear Information System (INIS)

    Francisco, J. L. de; Perez-Navarro, A.; Rodriguez-Mayquez, E.

    1973-01-01

    The effective delayed neutron fraction, β eff, has been determined from the measurement of E / β 2 , by means of reactor noise analysis in the time domain, and the neutron detector efficiency, ε. For the ε measurement it is necessary to determine the fission rate in the reactor. This value can be obtained from the absolute measurement of the fission rate per cm 3 , at a certain point of the reactor, and the determination of these two values ratio, which has been calculated by the Monte Cario method and also measured with results in good agreement. (Author)

  2. Investigation of capture reactions far off stability by β-delayed neutron emission

    International Nuclear Information System (INIS)

    Wiescher, M.; Leist, B.; Ziegert, W.; Gabelmann, H.; Steinmueller, B.; Ohm, H.; Kratz, K.h.; Thielemann, F.h.; Hillebrandt, W.

    1985-01-01

    Beta-delayed neutron spectroscopy is applied to determine reaction rates of neutron capture on several neutron rich nuclei. The results of these experiments are presented and discussed in the light of their astrophysical implications. Furthermore, the experimental possibilities and limits of planned measurements are advertised

  3. MODICO, 1-D Time-Dependent 1 Group, 2 Group Neutron Diffusion with Delayed Neutron Precursors

    International Nuclear Information System (INIS)

    Camiciola, P.; Cundari, D.; Montagnini, B.

    1992-01-01

    1 - Description of program or function: The program solves the 1-D time-dependent one and two group coarse-mesh neutron diffusion equations, coupled with the equations for the delayed-neutron precursor, in plane geometry. 2 - Method of solution: The program is based on a simple coarse-mesh cubic approximation formula for the spatial behaviour of the flux inside each interval. An implicit scheme (the time-integrated method) is used for the advancement of the solution. The resulting (block three-diagonal) matrix is inverted at each time step by Thomas' method. 3 - Restrictions on the complexity of the problem: Number of coarse- mesh intervals LE 80; number of material regions LE 10; number of delayed-neutron precursor groups LE 10. Typical mesh sizes range from 5 cm to 20 cm; typical step length (non-prompt critical transients) ranges from 0.005 to 0.1 seconds

  4. One-run Monte Carlo calculation of effective delayed neutron fraction and area-ratio reactivity

    Energy Technology Data Exchange (ETDEWEB)

    Zhaopeng Zhong; Talamo, Alberto; Gohar, Yousry, E-mail: zzhong@anl.gov, E-mail: alby@anl.gov, E-mail: gohar@anl.gov [Nuclear Engineering Division, Argonne National Laboratory, IL (United States)

    2011-07-01

    The Monte Carlo code MCNPX has been utilized to calculate the effective delayed neutron fraction and reactivity by using the area-ratio method. The effective delayed neutron fraction β{sub eff} has been calculated with the fission probability method proposed by Meulekamp and van der Marck. MCNPX was used to calculate separately the fission probability of the delayed and the prompt neutrons by using the TALLYX user subroutine of MCNPX. In this way, β{sub eff} was obtained from the one criticality (k-code) calculation without performing an adjoint calculation. The traditional k-ratio method requires two criticality calculations to calculate β{sub eff}, while this approach utilizes only one MCNPX criticality calculation. Therefore, the approach described here is referred to as a one-run method. In subcritical systems driven by a pulsed neutron source, the area-ratio method is used to calculate reactivity (in dollar units) as the ratio between the prompt and delayed areas. These areas represent the integral of the reaction rates induced from the prompt and delayed neutrons during the pulse period. Traditionally, application of the area-ratio method requires two separate fixed source MCNPX simulations: one with delayed neutrons and the other without. The number of source particles in these two simulations must be extremely high in order to obtain accurate results with low statistical errors because the values of the total and prompt areas are very close. Consequently, this approach is time consuming and suffers from the statistical errors of the two simulations. The present paper introduces a more efficient method for estimating the reactivity calculated with the area method by taking advantage of the TALLYX user subroutine of MCNPX. This subroutine has been developed for separately scoring the reaction rates caused by the delayed and the prompt neutrons during a single simulation. Therefore the method is referred to as a one run calculation. These methodologies have

  5. One-run Monte Carlo calculation of effective delayed neutron fraction and area-ratio reactivity

    International Nuclear Information System (INIS)

    Zhaopeng Zhong; Talamo, Alberto; Gohar, Yousry

    2011-01-01

    The Monte Carlo code MCNPX has been utilized to calculate the effective delayed neutron fraction and reactivity by using the area-ratio method. The effective delayed neutron fraction β_e_f_f has been calculated with the fission probability method proposed by Meulekamp and van der Marck. MCNPX was used to calculate separately the fission probability of the delayed and the prompt neutrons by using the TALLYX user subroutine of MCNPX. In this way, β_e_f_f was obtained from the one criticality (k-code) calculation without performing an adjoint calculation. The traditional k-ratio method requires two criticality calculations to calculate β_e_f_f, while this approach utilizes only one MCNPX criticality calculation. Therefore, the approach described here is referred to as a one-run method. In subcritical systems driven by a pulsed neutron source, the area-ratio method is used to calculate reactivity (in dollar units) as the ratio between the prompt and delayed areas. These areas represent the integral of the reaction rates induced from the prompt and delayed neutrons during the pulse period. Traditionally, application of the area-ratio method requires two separate fixed source MCNPX simulations: one with delayed neutrons and the other without. The number of source particles in these two simulations must be extremely high in order to obtain accurate results with low statistical errors because the values of the total and prompt areas are very close. Consequently, this approach is time consuming and suffers from the statistical errors of the two simulations. The present paper introduces a more efficient method for estimating the reactivity calculated with the area method by taking advantage of the TALLYX user subroutine of MCNPX. This subroutine has been developed for separately scoring the reaction rates caused by the delayed and the prompt neutrons during a single simulation. Therefore the method is referred to as a one run calculation. These methodologies have been

  6. Theory and use of GIRAFFE for analysis of decay characteristics of delayed-neutron precursors in an LMFBR

    International Nuclear Information System (INIS)

    Gross, K.C.

    1980-07-01

    The application of the computer code GIRAFFE (General Isotope Release Analysis For Failed Elements) written in FORTRAN IV is described. GIRAFFE was designed to provide parameter estimates of the nonlinear discrete-measurement models that govern the transport and decay of delayed-neutron precursors in a liquid-metal fast breeder reactor (LMFBR). The code has been organized into a set of small, relatively independent and well-defined modules to facilitate modification and maintenance. The program logic, the numerical techniques, and the methods of solution used by the code are presented, and the functions of the MAIN program and of each subroutine are discussed

  7. A delayed neutron technique for measuring induced fission rates in fresh and burnt LWR fuel

    Energy Technology Data Exchange (ETDEWEB)

    Jordan, K.A., E-mail: kajordan@gmail.co [Paul Scherrer Institut, Laboratory for Reactor Physics and System Behaviour, 5232 Villigen (Switzerland); Perret, G. [Paul Scherrer Institut, Laboratory for Reactor Physics and System Behaviour, 5232 Villigen (Switzerland)

    2011-04-01

    The LIFE-PROTEUS program at the Paul Scherrer Institut is being undertaken to characterize the interfaces between burnt and fresh fuel assemblies in modern LWRs. Techniques are being developed to measure fission rates in burnt fuel following re-irradiation in the zero-power PROTEUS research reactor. One such technique utilizes the measurement of delayed neutrons. To demonstrate the feasibility of the delayed neutron technique, fresh and burnt UO{sub 2} fuel samples were irradiated in different positions in the PROTEUS reactor, and their neutron outputs were recorded shortly after irradiation. Fission rate ratios of the same sample irradiated in two different positions (inter-positional) and of two different samples irradiated in the same position (inter-sample) were derived from the measurements and compared with Monte Carlo predictions. Derivation of fission rate ratios from the delayed neutron measured signal requires correcting the signal for the delayed neutron source properties, the efficiency of the measurement setup, and the time dependency of the signal. In particular, delayed neutron source properties strongly depend on the fissile and fertile isotopes present in the irradiated sample and must be accounted for when deriving inter-sample fission rate ratios. Measured inter-positional fission rate ratios generally agree within 1{sigma} uncertainty (on the order of 1.0%) with the calculation predictions. For a particular irradiation position, however, a bias of about 2% is observed and is currently under investigation. Calculated and measured inter-sample fission rate ratios have C/E values deviating from unity by less than 1% and within 2{sigma} of the statistical uncertainties. Uncertainty arising from delayed neutron data is also assessed, and is found to give an additional 3% uncertainty factor. The measurement data indicate that uncertainty is overestimated.

  8. New Beta-delayed Neutron Measurements in the Light-mass Fission Group

    Energy Technology Data Exchange (ETDEWEB)

    Agramunt, J. [Instituto de Física Corpuscular, CSIC-Univ. Valencia, Apdo. Correos 22085, E-46071 Valencia (Spain); García, A.R. [Centro de Investigaciones Energéticas, Medioambientales y Tecnológicas, E-28040 Madrid (Spain); Algora, A. [Instituto de Física Corpuscular, CSIC-Univ. Valencia, Apdo. Correos 22085, E-46071 Valencia (Spain); Äystö, J. [University of Jyväskylä, FI-40014 Jyväskyä (Finland); Caballero-Folch, R.; Calviño, F. [Secció d' Enginyeria Nuclear, Universitat Politécnica de Catalunya, E-08028 Barcelona (Spain); Cano-Ott, D. [Centro de Investigaciones Energéticas, Medioambientales y Tecnológicas, E-28040 Madrid (Spain); Cortés, G. [Secció d' Enginyeria Nuclear, Universitat Politécnica de Catalunya, E-08028 Barcelona (Spain); Domingo-Pardo, C. [Instituto de Física Corpuscular, CSIC-Univ. Valencia, Apdo. Correos 22085, E-46071 Valencia (Spain); Eronen, T. [University of Jyväskylä, FI-40014 Jyväskyä (Finland); Gelletly, W. [Department of Physics, University of Surrey, Guildford GU2 7XH (United Kingdom); Gómez-Hornillos, M.B. [Secció d' Enginyeria Nuclear, Universitat Politécnica de Catalunya, E-08028 Barcelona (Spain); and others

    2014-06-15

    A new accurate determination of beta-delayed neutron emission probabilities from nuclei in the low mass region of the light fission group has been performed. The measurements were carried out using the BELEN 4π neutron counter at the IGISOL-JYFL mass separator in combination with a Penning trap. The new results significantly improve the uncertainties of neutron emission probabilities for {sup 91}Br, {sup 86}As, {sup 85}As, and {sup 85}Ge nuclei.

  9. Statistical and non statistical models for delayed neutron emission: applications to nuclei near A = 90

    International Nuclear Information System (INIS)

    De Oliveira, Z.M.

    1980-01-01

    A detailed analysis of the simple statistical model description for delayed neutron emission of 87 Br, 137 I, 85 As and 135 Sb has been performed. In agreement with experimental findings, structure in the #betta#-strength function is required to reproduce the envelope of the neutron spectrum from 87 Br. For 85 As and 135 Sb the model is found incapable of simultaneously reproducing envelopes of delayed neutron spectra and neutron branching ratios to excited states in the final nuclei for any choice of #betta#-strength function. The results indicate that partial widths for neutron emission are not compatible with optical-model transmission coefficients. The simple shell model with pairing is shown to qualitatively describe the main features of the #betta#-strength functions for decay of 87 Br and 91 93 95 97 Rb. It is found that the location of apparent resonances in the experimental data are in rough agreement with the location of centroids of strength calculated with this model. An extension of the shell model picture which includes the Gamow-Teller residual interaction is used to investigate decay properties of 84 86 As, 86 92 Br and 88 102 Rb. For a realistic choice of interaction strength, the half lives of these isotopes are fairly well reproduced and semiquantitative agreement with experimental #betta#-strength functions is found. Delayed neutron emission probabilities are reproduced for precursors nearer stability with systematic deviations being observed for the heavier nuclei. Contrary to the assumption of a structureless Gamow-Teller giant resonance as embodied gross theory of #betta#-decay, we find that structures in the tail of the Gamow-Teller giant resonances are expected which strongly influence the decay properties of nuclides in this region

  10. Neutron activation analysis of geochemical samples

    International Nuclear Information System (INIS)

    Rosenberg, R.; Zilliacus, R.; Kaistila, M.

    1983-06-01

    The present paper will describe the work done at the Technical Research Centre of Finland in developing methods for the large-scale activation analysis of samples for the geochemical prospecting of metals. The geochemical prospecting for uranium started in Finland in 1974 and consequently a manually operated device for the delayed neutron activation analysis of uranium was taken into use. During 1974 9000 samples were analyzed. The small capacity of the analyzer made it necessary to develop a completely automated analyzer which was taken into use in August 1975. Since then 20000-30000 samples have been analyzed annually the annual capacity being about 60000 samples when running seven hours per day. Multielemental instrumental neutron activation analysis is used for the analysis of more than 40 elements. Using instrumental epithermal neutron activation analysis 25-27 elements can be analyzed using one irradiation and 20 min measurement. During 1982 12000 samples were analyzed for mining companies and Geological Survey of Finland. The capacity is 600 samples per week. Besides these two analytical methods the analysis of lanthanoids is an important part of the work. 11 lanthanoids have been analyzed using instrumental neutron activation analysis. Radiochemical separation methods have been developed for several elements to improve the sensitivity of the analysis

  11. Activity of the Delayed Neutron Working Group of JNDC and the International Evaluation Cooperation - WPEC/SG6

    International Nuclear Information System (INIS)

    Yoshida, Tadashi

    1999-01-01

    The Delayed Neutron Working Group was established in April 1997 within the Nuclear Data Subcommittee of JNDC. It has two principal missions. One is to coordinate the Japanese activities toward the WPEC/Subgroup-6 efforts, and the other is to recommend the delayed neutron data for JENDL-3.3. The final report of Subgroup-6, which in one of the subgroups of the NEA International Evaluation Cooperation (WPEC) and is in charge of the delayed neutron data, is to be completed in 1999. Here in Japan, JENDL-3.3 is planned to be released in early 2000. Delayed Neutron Working Group is, then, going to finalize its activity by the end of the fiscal year 1999 after recommending appropriate sets of data as coherently as possible with the of Subgroup-6 efforts. (author)

  12. $\\beta$-delayed neutron spectroscopy of $^{130-132}$ Cd isotopes with the ISOLDE decay station and the VANDLE array

    CERN Multimedia

    We propose to use the new ISOLDE decay station and the neutron detector VANDLE to measure the $\\beta$-delayed neutron emission of N=82-84 $^{130-132}$Cd isotopes. The large delayed neutron emission probability observed in a previous ISOLDE measurement is indicative of the Gamow-Teller transitions due to the decay of deep core neutrons. Core Gamow-Teller decay has been experimentally proven in the $^{78}$Ni region for the N>50 nuclei using the VANDLE array. The spectroscopic measurement of delayed neutron emission along the cadmium isotopic chain will allow us to track the evolution of the single particle states and the shell gap.

  13. Applicability of the activation analysis with prompt neutron in medicine

    International Nuclear Information System (INIS)

    Yaghubian-Malhami, R.

    1975-04-01

    The concentrations of boron and cadmium in the human body are of great importance in medicine. The author determined their concentration by prompt neutron activation analysis in aqueous solutions and in urine. The results show that this technique may be used in medical diagnosis. The author discusses the qualities and the applicability of delayed and prompt neutron activation analysis in biology and medicine. (C.R.)

  14. Bioassay method for Uranium in urine by Delay Neutron counting

    International Nuclear Information System (INIS)

    Suratman; Purwanto; Sukarman-Aminjoyo

    1996-01-01

    A bioassay method for uranium in urine by neutron counting has been studied. The aim of this research is to obtain a bioassay method for uranium in urine which is used for the determination of internal dose of radiation workers. The bioassay was applied to the artificially uranium contaminated urine. The weight of the contaminant was varied. The uranium in the urine was irradiated in the Kartini reactor core, through pneumatic system. The delayed neutron was counted by BF3 neutron counter. Recovery of the bioassay was between 69.8-88.8 %, standard deviation was less than 10 % and the minimum detection was 0.387 μg

  15. SOURCES-3A: A code for calculating (α, n), spontaneous fission, and delayed neutron sources and spectra

    International Nuclear Information System (INIS)

    Perry, R.T.; Wilson, W.B.; Charlton, W.S.

    1998-04-01

    In many systems, it is imperative to have accurate knowledge of all significant sources of neutrons due to the decay of radionuclides. These sources can include neutrons resulting from the spontaneous fission of actinides, the interaction of actinide decay α-particles in (α,n) reactions with low- or medium-Z nuclides, and/or delayed neutrons from the fission products of actinides. Numerous systems exist in which these neutron sources could be important. These include, but are not limited to, clean and spent nuclear fuel (UO 2 , ThO 2 , MOX, etc.), enrichment plant operations (UF 6 , PuF 4 , etc.), waste tank studies, waste products in borosilicate glass or glass-ceramic mixtures, and weapons-grade plutonium in storage containers. SOURCES-3A is a computer code that determines neutron production rates and spectra from (α,n) reactions, spontaneous fission, and delayed neutron emission due to the decay of radionuclides in homogeneous media (i.e., a mixture of α-emitting source material and low-Z target material) and in interface problems (i.e., a slab of α-emitting source material in contact with a slab of low-Z target material). The code is also capable of calculating the neutron production rates due to (α,n) reactions induced by a monoenergetic beam of α-particles incident on a slab of target material. Spontaneous fission spectra are calculated with evaluated half-life, spontaneous fission branching, and Watt spectrum parameters for 43 actinides. The (α,n) spectra are calculated using an assumed isotropic angular distribution in the center-of-mass system with a library of 89 nuclide decay α-particle spectra, 24 sets of measured and/or evaluated (α,n) cross sections and product nuclide level branching fractions, and functional α-particle stopping cross sections for Z < 106. The delayed neutron spectra are taken from an evaluated library of 105 precursors. The code outputs the magnitude and spectra of the resultant neutron source. It also provides an

  16. Energy dependence of relative abundances and periods of separate groups of delayed neutrons at neutron induced fission of 239Pu in a range of neutrons energies 0.37 - 5 MeV

    International Nuclear Information System (INIS)

    Roschenko, V.A.; Piksaikin, V.M.; Kazakov, L.E.; Isaev, S.G.; Korolev, G.G.; Tarasko, M.Z.; Tertychnyi, R.G.

    2001-01-01

    The fundamental role of delayed neutrons in behavior, control and safety of reactors is well known today. Delayed neutron data are of great interest not only for reactor physics but also for nuclear fission physics and astrophysics. The purpose of the present work was the measurement of energy dependence of delayed neutrons (DN) group parameters at fission of nuclei 239 Pu in a range of energies of primary neutrons from 0.37 up to 5 MeV. The measurements were executed on installation designed on the basis of the electrostatic accelerator of KG - 2.5 SSC RF IPPE. The data are obtained in 6-group representation. It is shown, that there is a significant energy dependence of DN group parameters in a range of primary neutrons energies from thermal meanings up to 5 MeV, which is expressed in reduction of the average half-life of nuclei of the DN precursors on 10 %. The data, received in the present work, can be used at creation of a set of group constants for reactors with an intermediate spectrum of neutrons. (authors)

  17. Review of experimental methods for evaluating effective delayed neutron fraction

    Energy Technology Data Exchange (ETDEWEB)

    Yamane, Yoshihiro [Nagoya Univ. (Japan). School of Engineering

    1997-03-01

    The International Effective Delayed Neutron Fraction ({beta}{sub eff}) Benchmark Experiments have been carried out at the Fast Critical Assembly of Japan Atomic Energy Research Institute since 1995. Researchers from six countries, namely France, Italy, Russia, U.S.A., Korea, and Japan, participate in this FCA project. Each team makes use of each experimental method, such as Frequency Method, Rossi-{alpha} Method, Nelson Number Method, Cf Neutron Source Method, and Covariance Method. In this report these experimental methods are reviewed. (author)

  18. Calculation of the effective delayed neutron fraction by TRIPOLI-4 code for IPEN/MB-01 research reactor

    International Nuclear Information System (INIS)

    Lee, Y.K.; Hugot, F.X.

    2011-01-01

    The effective delayed neutron fraction βeff is an important reactor physics parameter. Its calculation within the multi-group deterministic transport code can be performed with the aid of adjoint flux weighted integrations. However, in continuous energy Monte Carlo transport code, the adjoint weighted βeff calculation becomes complicated due to the backward treatment of the anisotropy scattering. In TRIPOLI-4 continuous energy Monte Carlo code, the βeff calculation was performed by a two-run method, one run with delayed neutrons and second with only the contribution from prompt fission neutrons. To improve the uncertainty of the βeff two-run calculation for the experimental reactors, two simple and fast one-run methods to estimate the βeff in the continuous energy simulation have been implemented into the TRIPOLI-4 code. First approach is an improved one of the Bretscher's prompt method and second one based on the proposal of Nauchi and Kameyama. In these one-run methods, the prompt and the delayed neutrons are first tagged. Their tracking and statistics are separated performed. The new βeff calculations have been optimized in the power iteration cycles so as to estimate the production of prompt and delayed neutrons from the prompt and delayed neutrons of previous generation. To validate the new βeff calculation by TRIPOLI-4, several benchmarks including fast and thermal systems have been considered. In this paper the recent measurements of βeff in the research reactor IPEN/MB-01 have been benchmarked. The basic components of the βeff and the Keff have been also calculated so as to understand the influences of the cross sections and the delayed neutron yields on the reactor reactivity calculations. Three nuclear data libraries, ENDF/BVI.r4, ENDF/B-VII.0, and JEFF-3.1 were taken into account in this study. (author)

  19. On the combination of delayed neutron and delayed gamma techniques for fission rate measurement in nuclear fuel

    Energy Technology Data Exchange (ETDEWEB)

    Perret, G.; Jordan, K. A. [Paul Scherrer Institut, Villigen, 5232 (Switzerland)

    2011-07-01

    Novel techniques to measure newly induced fissions in spent fuel after re-irradiation at low power have been developed and tested at the Proteus zero-power research reactor. The two techniques are based on the detection of high energy gamma-rays emitted by short-lived fission products and delayed neutrons. The two techniques relate the measured signals to the total fission rate, the isotopic composition of the fuel, and nuclear data. They can be combined to derive better estimates on each of these parameters. This has potential for improvement in many areas. Spent fuel characterisation and safeguard applications can benefit from these techniques for non-destructive assay of plutonium content. Another application of choice is the reduction of uncertainties on nuclear data. As a first application of the combination of the delayed neutron and gamma measurement techniques, this paper shows how to reduce the uncertainties on the relative abundances of the longest delayed neutron group for thermal fissions in {sup 235}U, {sup 239}Pu and fast fissions in {sup 238}U. The proposed experiments are easily achievable in zero-power research reactors using fresh UO{sub 2} and MOX fuel and do not require fast extraction systems. The relative uncertainties (1{sigma}) on the relative abundances are expected to be reduced from 13% to 4%, 16% to 5%, and 38% to 12% for {sup 235}U, {sup 238}U and {sup 239}Pu, respectively. (authors)

  20. Test of statistical models of the ν-delayed neutron emission by application of the Monte Carlo method

    International Nuclear Information System (INIS)

    Ohm, H.

    1982-01-01

    Using the example of the delayed neutron spectrum of 24 s- 137 I the statistical model is tested in view of its applicability. A computer code was developed which simulates delayed neutron spectra by the Monte Carlo method under the assumption that the transition probabilities of the ν and the neutron decays obey the Porter-Thomas distribution while the distances of the neutron emitting levels are Wigner distribution. Gramow-Teller ν-transitions and simply forbidden ν-transitions from the preceding nucleus to the emitting nucleus were regarded. (orig./HSI) [de

  1. Detection of special nuclear material from delayed neutron emission induced by a dual-particle monoenergetic source

    Energy Technology Data Exchange (ETDEWEB)

    Mayer, M. [Department of Mechanical and Nuclear Engineering, The Pennsylvania State University, University Park, Pennsylvania 16802 (United States); Nattress, J.; Jovanovic, I., E-mail: ijov@umich.edu [Department of Nuclear Engineering and Radiological Sciences, University of Michigan, Ann Arbor, Michigan 48109 (United States)

    2016-06-27

    Detection of unique signatures of special nuclear materials is critical for their interdiction in a variety of nuclear security and nonproliferation scenarios. We report on the observation of delayed neutrons from fission of uranium induced in dual-particle active interrogation based on the {sup 11}B(d,n γ){sup 12}C nuclear reaction. Majority of the fissions are attributed to fast fission induced by the incident quasi-monoenergetic neutrons. A Li-doped glass–polymer composite scintillation neutron detector, which displays excellent neutron/γ discrimination at low energies, was used in the measurements, along with a recoil-based liquid scintillation detector. Time-dependent buildup and decay of delayed neutron emission from {sup 238}U were measured between the interrogating beam pulses and after the interrogating beam was turned off, respectively. Characteristic buildup and decay time profiles were compared to the common parametrization into six delayed neutron groups, finding a good agreement between the measurement and nuclear data. This method is promising for detecting fissile and fissionable materials in cargo scanning applications and can be readily integrated with transmission radiography using low-energy nuclear reaction sources.

  2. Delayed neutron spectra from short pulse fission of uranium-235

    International Nuclear Information System (INIS)

    Atwater, H.F.; Goulding, C.A.; Moss, C.E.; Pederson, R.A.; Robba, A.A.; Wimett, T.F.; Reeder, P.; Warner, R.

    1986-01-01

    Delayed neutron spectra from individual short pulse (∼50 μs) fission of small 235 U samples (50 mg) were measured using a small (5 cm OD x 5 cm length) NE 213 neutron spectrometer. The irradiating fast neutron flux (∼10 13 neutrons/cm 2 ) for these measurements was provided by the Godiva fast burst reactor at the Los Alamos Critical Experiment Facility (LACEF). A high speed pneumatic transfer system was used to transfer the 50 mg 235 U samples from the irradiation position near the Godiva assembly to a remote shielded counting room containing the NE 213 spectrometer and associated electronics. Data were acquired in sixty-four 0.5 s time bins and over an energy range 1 to 7 MeV. Comparisons between these measurements and a detailed model calculation performed at Los Alamos is presented

  3. Power measurement in the boiling capsules in R2 using delayed neutron detector

    International Nuclear Information System (INIS)

    Roennberg, G.

    1979-03-01

    LWR fuel testing is performed in the R2 reactor by irradiation in both loops and so-called boiling capsules. The loops have forced cooling, and the power can be measured calorimetrically by conventional instrumentation. The boiling capsules have convection cooling, and it has therefore been necessary to develop a special technique for power measurement, the delayed neutron detector (DND). The DND is a pneumatic rabbit system, which activates small uranium samples in the boiling capsules and counts the delayed neutrons for determination of the fission rate. This report describes the equipment used, the procedure of measurement, and the method of evaluation. (atuhor)

  4. Study of $\\beta$-delayed neutron decay of $^{8}$He

    CERN Multimedia

    The goal of the present proposal is to study $\\beta$-delayed neutron decay branch of $^{8}$He. The energy spectra of the emitted neutrons will be measured in the energy range of 0.1 – 6 MeV using the VANDLE spectrometer. Using coincident $\\gamma$-ray measurement, components of the spectrum corresponding to transitions to the ground- and first- excited states of $^{7}$Li will be disentangled. The new data will allow us to get a more complete picture of the $\\beta$-decay of $^{8}$He and to clarify the discrepancy between the B(GT) distributions derived from the $\\beta$-decay and $^{8}$He(p, n)$^{8}$Li reaction studies.

  5. Beta-delayed gamma and neutron emission near the double shell closure at 78Ni

    International Nuclear Information System (INIS)

    Rykaczewski, Krzysztof Piotr; Mazzocchi, C.; Grzywacz, R.; Batchelder, J. C.; Bingham, C.R.; Fong, D.; Hamilton, J.H.; Hwang, J.K.; Karny, M.; Krolas, W.; Liddick, S. N.; Morton, A. C.; Mantica, P. F.; Mueller, W. F.; Steiner, M.; Stolz, A.; Winger, J.A.

    2005-01-01

    An experiment was performed at the National Superconducting Cyclotron Laboratory at Michigan State University to investigate β decay of very neutron-rich cobalt isotopes. Beta-delayed neutron emission from 71-74 Co has been observed for the first time. Preliminary results are reported

  6. Electron-capture delayed fission properties of neutron-deficient einsteinium nuclei

    International Nuclear Information System (INIS)

    Shaughnessy, Dawn A.

    2000-01-01

    Electron-capture delayed fission (ECDF) properties of neutron-deficient einsteinium isotopes were investigated using a combination of chemical separations and on-line radiation detection methods. 242 Es was produced via the 233 U( 14 N,5n) 242 Es reaction at a beam energy of 87 MeV (on target) in the lab system, and was found to decay with a half-life of 11 ± 3 seconds. The ECDF of 242 Es showed a highly asymmetric mass distribution with an average pre-neutron emission total kinetic energy (TKE) of 183 ± 18 MeV. The probability of delayed fission (P DF ) was measured to be 0.006 ± 0.002. In conjunction with this experiment, the excitation functions of the 233 U( 14 N,xn) 247-x Es and 233 U( 15 N,xn) 248-x Es reactions were measured for 243 Es, 244 Es and 245 Es at projectile energies between 80 MeV and 100 MeV

  7. First Measurement of Several β-Delayed Neutron Emitting Isotopes Beyond N=126.

    Science.gov (United States)

    Caballero-Folch, R; Domingo-Pardo, C; Agramunt, J; Algora, A; Ameil, F; Arcones, A; Ayyad, Y; Benlliure, J; Borzov, I N; Bowry, M; Calviño, F; Cano-Ott, D; Cortés, G; Davinson, T; Dillmann, I; Estrade, A; Evdokimov, A; Faestermann, T; Farinon, F; Galaviz, D; García, A R; Geissel, H; Gelletly, W; Gernhäuser, R; Gómez-Hornillos, M B; Guerrero, C; Heil, M; Hinke, C; Knöbel, R; Kojouharov, I; Kurcewicz, J; Kurz, N; Litvinov, Yu A; Maier, L; Marganiec, J; Marketin, T; Marta, M; Martínez, T; Martínez-Pinedo, G; Montes, F; Mukha, I; Napoli, D R; Nociforo, C; Paradela, C; Pietri, S; Podolyák, Zs; Prochazka, A; Rice, S; Riego, A; Rubio, B; Schaffner, H; Scheidenberger, Ch; Smith, K; Sokol, E; Steiger, K; Sun, B; Taín, J L; Takechi, M; Testov, D; Weick, H; Wilson, E; Winfield, J S; Wood, R; Woods, P; Yeremin, A

    2016-07-01

    The β-delayed neutron emission probabilities of neutron rich Hg and Tl nuclei have been measured together with β-decay half-lives for 20 isotopes of Au, Hg, Tl, Pb, and Bi in the mass region N≳126. These are the heaviest species where neutron emission has been observed so far. These measurements provide key information to evaluate the performance of nuclear microscopic and phenomenological models in reproducing the high-energy part of the β-decay strength distribution. This provides important constraints on global theoretical models currently used in r-process nucleosynthesis.

  8. Sensitivity analysis of the kinetic behaviour of a Gas Cooled Fast Reactor to variations of the delayed neutron parameters

    International Nuclear Information System (INIS)

    Van Rooijen, W. F. G.; Lathouwers, D.

    2007-01-01

    In advanced Generation IV (fast) reactors an integral fuel cycle is envisaged, where all Heavy Metal is recycled in the reactor. This leads to a nuclear fuel with a considerable content of Minor Actinides. For many of these isotopes the nuclear data is not very well known. In this paper the sensitivity of the kinetic behaviour of the reactor to the dynamic parameters λ k , β k and the delayed spectrum χ d,k is studied using first order perturbation theory. In the current study, feedback due to Doppler and/or thermohydraulic effects are not treated. The theoretical framework is applied to a Generation IV Gas Cooled Fast Reactor. The results indicate that the first-order approach is satisfactory for small variations of the data. Sensitivities to delayed neutron data are similar for increasing and decreasing transients. Sensitivities generally increase with reactivity for increasing transients. For decreasing transients, there are less clearly defined trends, although the sensitivity to the delayed neutron spectrum decreases with larger sub-criticality, as expected. For this research, an adjoint capable version of the time-dependent diffusion code DALTON is under development. (authors)

  9. A Neutron Sensitive Microchannel Plate Detector with Cross Delay Line Readout

    International Nuclear Information System (INIS)

    Berry, Kevin D.; Bilheux, Hassina Z.; Crow, Lowell; Diawara, Yacouba; Feller, W. Bruce; Iverson, Erik B.; Martin, Adrian; Robertson, J. Lee

    2012-01-01

    Microchannel plates containing neutron absorbing elements such as boron and gadolinium in the bulk glass are used as the sensing element in high spatial resolution, high rate neutron imaging systems. In this paper we describe one such device, using both 10 B and natural Gd, which employs cross delay line signal readout, with time-of-flight capability. This detector has a measured spatial resolution under 40 m FWHM, thermal neutron efficiency of 19%, and has recorded rates in excess of 500 kHz. A physical and functional description is presented, followed by a discussion of measurements of detector performance and a brief survey of some practical applications.

  10. Development of a photonuclear activation file and measurement of delayed neutron spectra; Creation d'une bibliotheque d'activation photonucleaire et mesures de spectres d'emission de neutrons retardes

    Energy Technology Data Exchange (ETDEWEB)

    Giacri-Mauborgne, M.L

    2005-11-15

    This thesis work consists in two parts. The first part is the description of the creation of a photonuclear activation file which will be used to calculated photonuclear activation. To build this file we have used different data sources: evaluations but also calculations done using several cross sections codes (HMS-ALICE, GNASH, ABLA). This file contains photonuclear activation cross sections for more than 600 nuclides and fission fragments distributions for 30 actinides at tree different Bremsstrahlung energies and the delay neutron spectrum associated. These spectra are not in good agreement with experimental data. That is why we decided to launch measurement of delayed neutrons spectra from photofission. The second part of this thesis consists in demonstrating the possibility to do such measurements at the ELSA accelerator facility. To that purpose, we have developed the detection, the acquisition system and the analysis method of such spectra. These were tested for the measurement of the delayed neutron spectrum of uranium-238 after irradiation in a 2 MeV neutron flux. Finally, we have measured the delayed neutron spectrum of uranium-238 after irradiation in a 15 MeV Bremsstrahlung flux. We compare our results with experimental data. The experiment has allowed us to improve the value of {nu}{sub p}-bar with an absolute uncertainty below 7%, we propose {nu}{sub p}-bar = (3.03 {+-} 0.02) n/100 fissions, and to correct the Nikotin's parameters for the six group representation. Particularly, we have improved the data concerning the sixth group by taking into account results from different irradiation times.

  11. Uranium analysis by neutron induced fissionography method using solid state nuclear track detectors

    International Nuclear Information System (INIS)

    Akyuez, T.; Tretyakova, S. P.; Guezel, T.; Akyuz, S.

    1999-01-01

    In this study total twenty samples (eight reference materials and twelve sediment samples) were analysed for their uranium content which is in the range of 1-17 μg/g, by neutron induced fissionography (NIF) method using solid state nuclear track detectors (SSNTDs) in comparison with the results of neutron activation analysis (NAA), delayed neutron counting (DNC) technique or fluorometric method. It is found that NIF method using SSNTDs is very sensitive for analysis of uranium

  12. Uranium analysis by neutron induced fissionography method using solid state nuclear track detectors

    CERN Document Server

    Akyuez, T; Guezel, T; Akyuz, S

    1999-01-01

    In this study total twenty samples (eight reference materials and twelve sediment samples) were analysed for their uranium content which is in the range of 1-17 mu g/g, by neutron induced fissionography (NIF) method using solid state nuclear track detectors (SSNTDs) in comparison with the results of neutron activation analysis (NAA), delayed neutron counting (DNC) technique or fluorometric method. It is found that NIF method using SSNTDs is very sensitive for analysis of uranium.

  13. High-capacity neutron activation analysis facility

    International Nuclear Information System (INIS)

    Hochel, R.C.

    1979-01-01

    A high-capacity neutron activation analysis facility, the Reactor Activation Facility, was designed and built and has been in operation for about a year at one of the Savannah River Plant's production reactors. The facility determines uranium and about 19 other trace elements in hydrogeochemical samples collected in the National Uranium Resource Evaluation program. The facility has a demonstrated average analysis rate of over 10,000 samples per month, and a peak rate of over 16,000 samples per month. Uranium is determined by cyclic activation and delayed neutron counting of the U-235 fission products; other elements are determined from gamma-ray spectra recorded in subsequent irradiation, decay, and counting steps. The method relies on the absolute activation technique and is highly automated for round-the-clock unattended operation

  14. Beta-delayed fission and neutron emission calculations for the actinide cosmochronometers

    International Nuclear Information System (INIS)

    Meyer, B.S.; Howard, W.M.; Mathews, G.J.; Takahashi, K.; Moeller, P.; Leander, G.A.

    1989-01-01

    The Gamow-Teller beta-strength distributions for 19 neutron-rich nuclei, including ten of interest for the production of the actinide cosmochronometers, are computed microscopically with a code that treats nuclear deformation explicitly. The strength distributions are then used to calculate the beta-delayed fission, neutron emission, and gamma deexcitation probabilities for these nuclei. Fission is treated both in the complete damping and WKB approximations for penetrabilities through the nuclear potential-energy surface. The resulting fission probabilities differ by factors of 2 to 3 or more from the results of previous calculations using microscopically computed beta-strength distributions around the region of greatest interest for production of the cosmochronometers. The indications are that a consistent treatment of nuclear deformation, fission barriers, and beta-strength functions is important in the calculation of delayed fission probabilities and the production of the actinide cosmochronometers. Since we show that the results are very sensitive to relatively small changes in model assumptions, large chronometric ages for the Galaxy based upon high beta-delayed fission probabilities derived from an inconsistent set of nuclear data calculations must be considered quite uncertain

  15. First measurement of several $\\beta$-delayed neutron emitting isotopes beyond N=126

    CERN Document Server

    Caballero-Folch, R.; Agramunt, J.; Algora, A.; Ameil, F.; Arcones, A.; Ayyad, Y.; Benlliure, J.; Borzov, I.N.; Bowry, M.; Calvino, F.; Cano-Ott, D.; Cortés, G.; Davinson, T.; Dillmann, I.; Estrade, A.; Evdokimov, A.; Faestermann, T.; Farinon, F.; Galaviz, D.; García, A.R.; Geissel, H.; Gelletly, W.; Gernhäuser, R.; Gómez-Hornillos, M.B.; Guerrero, C.; Heil, M.; Hinke, C.; Knöbel, R.; Kojouharov, I.; Kurcewicz, J.; Kurz, N.; Litvinov, Y.; Maier, L.; Marganiec, J.; Marketin, T.; Marta, M.; Martínez, T.; Martínez-Pinedo, G.; Montes, F.; Mukha, I.; Napoli, D.R.; Nociforo, C.; Paradela, C.; Pietri, S.; Podolyák, Zs.; Prochazka, A.; Rice, S.; Riego, A.; Rubio, B.; Schaffner, H.; Scheidenberger, Ch.; Smith, K.; Sokol, E.; Steiger, K.; Sun, B.; Taín, J.L.; Takechi, M.; Testov, D.; Weick, H.; Wilson, E.; Winfield, J.S.; Wood, R.; Woods, P.; Yeremin, A.

    2016-01-01

    The $\\beta$-delayed neutron emission probabilities of neutron rich Hg and Tl nuclei have been measured together with $\\beta$-decay half-lives for 20 isotopes of Au, Hg, Tl, Pb and Bi in the mass region N$\\gtrsim$126. These are the heaviest species where neutron emission has been observed so far. These measurements provide key information to evaluate the performance of nuclear microscopic and phenomenological models in reproducing the high-energy part of the $\\beta$-decay strength distribution. In doing so, it provides important constraints to global theoretical models currently used in $r$-process nucleosynthesis.

  16. Determination of the effective delayed neutron fraction in the Coral-I Reactor; Determinacion de la fraccion efectiva de neutrones retardados en el Reactor Coral-1

    Energy Technology Data Exchange (ETDEWEB)

    Francisco, J L. de; Perez-Navarro, A; Rodriguez-Mayquez, E

    1973-07-01

    The effective delayed neutron fraction, {beta} eff, has been determined from the measurement of E / {beta}{sup 2}, by means of reactor noise analysis in the time domain, and the neutron detector efficiency, {epsilon}. For the {epsilon} measurement it is necessary to determine the fission rate in the reactor. This value can be obtained from the absolute measurement of the fission rate per cm{sup 3}, at a certain point of the reactor, and the determination of these two values ratio, which has been calculated by the Monte Cario method and also measured with results in good agreement. (Author)

  17. Study and building of a detection array for delayed neutrons: TONNERRE

    International Nuclear Information System (INIS)

    Martin, Thierry

    1998-01-01

    This work has been undertaken within a French-Romanian collaboration in order to build a high efficiency detector array for delayed neutrons: barrel-shaped TONNERRE. Some neutron-rich nuclei decay through 1, 2 or 3 neutron emission after β - decay. More exotic nuclei will be produced by SPIRAL at GANIL. An array with high efficiency and good resolution is then required. Thirty two BC400 plastic scintillators (160 x 20 x 4 cm 3 ) allow us to get the time of flight neutron spectra. They are bent for uniform flight path and viewed by a photomultiplier tube at both ends. Simulations have allowed to establish scintillator size and to minimize light attenuation. Intrinsic efficiency and crosstalk have been measured with 252 Cf and compared to GEANT. 1 to 5 MeV neutrons are detected with good timing and position properties. Other counters will be built for neutrons from 300 keV to 1 MeV. Planned to run at several particle accelerators (GANIL, CERN, and others), TONNERRE is modular and many geometries are possible. (author)

  18. $\\beta$-delayed neutrons from oriented $^{137,139}$I and $^{87,89}$Br nuclei

    CERN Multimedia

    We propose a world-first measurement of the angular distribution of $\\beta$‐delayed n and $\\gamma$-radiation from oriented $^{137, 139}$I and $^{87,89}$Br nuclei, polarised at low temperature at the NICOLE facility. $\\beta$­-delayed neutron emission is an increasingly important decay mechanism as the drip line is approached and its detailed understanding is essential to phenomena as fundamental as the r‐process and practical as the safe operation of nuclear power reactors. The experiments offer sensitive tests of theoretical input concerning the allowed and first­‐forbidden $\\beta$‐decay strength, the spin-density of neutron emitting states and the partial wave barrier penetration as a function of nuclear deformation. In $^{137}$I and $^{87}$Br the decay feeds predominantly the ground state of the daughters $^{136}$Xe and $^{86}$Kr whereas in $^{139}$I and $^{89}$Br we will explore the use of n-$\\gamma$- coincidence to study neutron transitions to the first and second excited states in the daughters...

  19. Pulse-shape discrimination in radioanalytical methods. Part I. Delayed fission neutron counting

    International Nuclear Information System (INIS)

    Posta, S.; Vacik, J.; Hnatowicz, V.; Cervena, J.

    1999-01-01

    In this study the principle of pulse shape discrimination (PSD) has been employed in delayed fission neutron counting (DNC) method. Effective elimination of unwanted gamma background signals in measured radiation spectra has been proved. (author)

  20. Estimation of delayed neutron emission probability by using the gross theory of nuclear β-decay

    International Nuclear Information System (INIS)

    Tachibana, Takahiro

    1999-01-01

    The delayed neutron emission probabilities (P n -values) of fission products are necessary in the study of reactor physics; e.g. in the calculation of total delayed neutron yields and in the summation calculation of decay heat. In this report, the P n -values estimated by the gross theory for some fission products are compared with experiment, and it is found that, on the average, the semi-gross theory somewhat underestimates the experimental P n -values. A modification of the β-decay strength function is briefly discussed to get more reasonable P n -values. (author)

  1. Data reduction for a high-throughput neutron activation analysis system

    International Nuclear Information System (INIS)

    Bowman, W.W.

    1979-01-01

    To analyze samples collected as part of a geochemical survey for the National Uranium Resource Evaluation program, Savannah River Laboratory has installed a high-throughput neutron activation analysis system. As part of that system, computer programs have been developed to reduce raw data to elemental concentrations in two steps. Program RAGS reduces gamma-ray spectra to lists of photopeak energies, peak areas, and statistical errors. Program RICHES determines the elemental concentrations from photopeak and delayed-neutron data, detector efficiencies, analysis parameters (neutron flux and activation, decay, and counting times), and spectrometric and cross-section data from libraries. Both programs have been streamlined for on-line operation with a minicomputer, each requiring approx. 64 kbytes of core. 3 tables

  2. Population of delayed-neutron granddaughter states and the optical potential

    International Nuclear Information System (INIS)

    Schenter, R.E.; Mann, F.M.; Warner, R.A.; Reeder, P.L.

    1982-08-01

    Using a statistical treatment of beta decay and the Hauser-Feshbach model of nuclear reactions, calculations were made and compared to recent experimental measurements of the population of granddaughter states of several delayed neutron precursors ( 144 145 147 Cs and 96 Rb). Emphasis of this paper is on the sensitivity and interpretation of experimental results to various standard low energy neutron optical model potentials and variations in their forms and parameters. Results for these precursors show qualitative agreement with experiment for all the optical potential models used and good quantitative agreement for two (Moldauer and Becchetti-Greenlees). Questions such as (N-Z) terms, deformation and nonlocality dependence are presented

  3. Development of Pneumatic Transfer Irradiation Facility (PTS no.2) for Neutron Activation Analysis at HANARO Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Y. S.; Moon, J. H.; Kim, S. H.; Sun, G. M.; Baek, S. Y.; Kim, H. R.; Kim, Y. J

    2008-03-15

    A pneumatic transfer irradiation system (PTS) is one of the most important facilities used during neutron irradiation of a target material for instrumental neutron activation analysis (INAA) in a research reactor. In particular, a fast pneumatic transfer system is essential for the measurement of a short half-life nuclide and a delayed neutron counting system. The pneumatic transfer irradiation system (PTS no.2) involving a manual system and an automatic system for delayed neutron activation analysis (DNAA) were reconstructed with new designs of a functional improvement at the HANARO research reactor in 2006. In this technical report, the conception, design, operation and control of PTS no.2 was described. Also the experimental results and the characteristic parameters measured by a mock-up test, a functional operation test and an irradiation test of these systems, such as the transfer time of irradiation capsule, automatic operation control by personal computer, delayed neutron counting system, the different neutron flux, the temperature of the irradiation position with an irradiation time, the radiation dose rate when the rabbit is returned, etc. are reported to provide a user information as well as a reactor's management and safety.

  4. Feasibility study of {sup 235}U and {sup 239}Pu characterization in radioactive waste drums using neutron-induced fission delayed gamma rays

    Energy Technology Data Exchange (ETDEWEB)

    Nicol, T. [CEA, DEN, Cadarache, Nuclear Measurement Laboratory, F-13108 Saint-Paul-lez-Durance (France); FZJ, Institute of Energy and Climate Research – Nuclear Waste Management and Reactor Safety, Wilhelm-Johnen-Straße, d-52425 Jülich (Germany); Pérot, B., E-mail: bertrand.perot@cea.fr [CEA, DEN, Cadarache, Nuclear Measurement Laboratory, F-13108 Saint-Paul-lez-Durance (France); Carasco, C. [CEA, DEN, Cadarache, Nuclear Measurement Laboratory, F-13108 Saint-Paul-lez-Durance (France); Brackx, E. [CEA, DEN, Marcoule, Metallography and Chemical Analysis Laboratory, F-30207 Bagnols-sur-Cèze (France); Mariani, A.; Passard, C. [CEA, DEN, Cadarache, Nuclear Measurement Laboratory, F-13108 Saint-Paul-lez-Durance (France); Mauerhofer, E. [FZJ, Institute of Energy and Climate Research – Nuclear Waste Management and Reactor Safety, Wilhelm-Johnen-Straße, d-52425 Jülich (Germany); Collot, J. [Laboratoire de Physique Subatomique et de Cosmologie, Université Grenoble Alpes, CNRS/IN2P3 Grenoble (France)

    2016-10-01

    This paper reports a feasibility study of {sup 235}U and {sup 239}Pu characterization in 225 L bituminized waste drums or 200 L concrete waste drums, by detecting delayed fission gamma rays between the pulses of a deuterium-tritium neutron generator. The delayed gamma yields were first measured with bare samples of {sup 235}U and {sup 239}Pu in REGAIN, a facility dedicated to the assay of 118 L waste drums by Prompt Gamma Neutron Activation Analysis (PGNAA) at CEA Cadarache, France. Detectability in the waste drums is then assessed using the MCNPX model of MEDINA (Multi Element Detection based on Instrumental Neutron Activation), another PGNAA cell dedicated to 200 L drums at FZJ, Germany. For the bituminized waste drum, performances are severely hampered by the high gamma background due to {sup 137}Cs, which requires the use of collimator and shield to avoid electronics saturation, these elements being very penalizing for the detection of the weak delayed gamma signal. However, for lower activity concrete drums, detection limits range from 10 to 290 g of {sup 235}U or {sup 239}Pu, depending on the delayed gamma rays of interest. These detection limits have been determined by using MCNPX to calculate the delayed gamma useful signal, and by measuring the experimental gamma background in MEDINA with a 200 L concrete drum mock-up. The performances could be significantly improved by using a higher interrogating neutron emission and an optimized experimental setup, which would allow characterizing nuclear materials in a wide range of low and medium activity waste packages.

  5. An examination of the time-dependent background counts of the delayed neutron counting system at the Royal Military College of Canada

    International Nuclear Information System (INIS)

    Sellers, M.T.; Corcoran, E.C.; Kelly, D.G.

    2011-01-01

    A delayed neutron counting (DNC) system for the analysis of special nuclear materials (SNM) has been constructed and calibrated at the Royal Military College of Canada. The polyethylene vials used to transport SNM samples have been found to contribute a time-dependent count rate, B(t), far above the system background. B(t) has been found to be independent of polyethylene mass and shows a dependence on irradiation position in the SLOWPOKE-2 reactor and irradiation time. A comparison of B(t) and the theoretical delayed neutron production from the fission of small amounts of 235 U has indicated that trace amounts of uranium may be present in the DNC system tubing. (author)

  6. Commissioning of the BRIKEN beta-delayed neutron detector for the study of exotic neutron-rich nuclei

    Directory of Open Access Journals (Sweden)

    Tolosa-Delgado A.

    2017-01-01

    Full Text Available The commissioning of a new setup for β-delayed neutron measurements was carried out successfully in November-2016, at the RIKEN Nishina Center in Japan. The β-decay half-lives and Pn branching ratios of several isotopes in the 78Ni region were measured. Details of the experimental setup and the first results are given.

  7. Neutron-neutron probe for uranium exploration

    International Nuclear Information System (INIS)

    Smith, R.C.

    1979-01-01

    A neutron activation probe for assaying the amount of fissionable isotopes in an ore body is described which comprises a casing which is movable through a borehole in the ore body, a neutron source and a number of delayed neutron detectors arranged colinearly in the casing below the neutron source for detecting delayed neutrons

  8. Use of delayed gamma rays for active non-destructive assay of {sup 235}U irradiated by pulsed neutron source (plasma focus)

    Energy Technology Data Exchange (ETDEWEB)

    Andola, Sanjay; Niranjan, Ram [Applied Physics Division, Bhabha Atomic Research Centre, Mumbai 400085 (India); Kaushik, T.C., E-mail: tckk@barc.gov.in [Applied Physics Division, Bhabha Atomic Research Centre, Mumbai 400085 (India); Rout, R.K. [Applied Physics Division, Bhabha Atomic Research Centre, Mumbai 400085 (India); Kumar, Ashwani; Paranjape, D.B.; Kumar, Pradeep; Tomar, B.S.; Ramakumar, K.L. [Radioanalytical Chemistry Division, Bhabha Atomic Research Centre, Mumbai 400085 (India); Gupta, S.C. [Applied Physics Division, Bhabha Atomic Research Centre, Mumbai 400085 (India)

    2014-07-01

    A pulsed neutron source based on plasma focus device has been used for active interrogation and assay of {sup 235}U by monitoring its delayed high energy γ-rays. The method involves irradiation of fissile material by thermal neutrons obtained after moderation of a burst of neutrons emitted upon fusion of deuterium in plasma focus (PF) device. The delayed gamma rays emitted from the fissile material as a consequence of induced fission were detected by a large volume sodium iodide (NaI(Tl)) detector. The detector is coupled to a data acquisition system of 2k input size with 2k ADC conversion gain. Counting was carried out in pulse height analysis mode for time integrated counts up to 100 s while the temporal profile of delayed gamma has been obtained by counting in multichannel scaling mode with dwell time of 50 ms. To avoid the effect of passive (natural) and active (from surrounding materials) backgrounds, counts have been acquired for gamma energy between 3 and 10 MeV. The lower limit of detection of {sup 235}U in the oxide samples with this set-up is estimated to be 14 mg.

  9. High-capacity neutron activation analysis facility

    International Nuclear Information System (INIS)

    Hochel, R.C.; Bowman, W.W.; Zeh, C.W.

    1980-01-01

    A high-capacity neutron activation analysis facility, the Reactor Activation Facility, was designed and built and has been in operation for about a year at one of the Savannah River Plant's production reactors. The facility determines uranium and about 19 other elements in hydrogeochemical samples collected in the National Uranium Resource Evaluation program, which is sponsored and funded by the United States Department of Energy, Grand Junction Office. The facility has a demonstrated average analysis rate of over 10,000 samples per month, and a peak rate of over 16,000 samples per month. Uranium is determined by cyclic activation and delayed neutron counting of the U-235 fission products; other elements are determined from gamma-ray spectra recorded in subsequent irradiation, decay, and counting steps. The method relies on the absolute activation technique and is highly automated for round-the-clock unattended operation

  10. $\\beta$-delayed neutrons from oriented $^{137,139}$I and $^{87,89}$Br nuclei

    CERN Document Server

    Grzywacz, Robert; Stone, Nicholas; Köster, Ulli; Singh, Barlaj; Bingham, Carrol; Gaulard, S; Kolos, Karolina; Madurga, Miguel; Nikolov, J; Otsubo, T; Roccia, S; Veskovic, Miroslav; Walker, Phil; Walters, William

    2013-01-01

    We propose a world-­‐first measurement of the angular distribution of $\\beta$-­‐delayed n and $\\gamma$- radiation from oriented $^{137, 139}$I and $^{87,89}$Br nuclei, polarised at low temperature at the NICOLE facility. $\\beta$-­‐delayed neutron emission is an increasingly important decay mechanism as the drip line is approached and its detailed understanding is essential to phenomena as fundamental as the r‐process and practical as the safe operation of nuclear power reactors. The experiments offer sensitive tests of theoretical input concerning the allowed and first-­‐forbidden $\\beta$‐decay strength, the spin-­‐density of neutron emitting states and the partial wave barrier penetration as a function of nuclear deformation. In $^{137}$I and $^{87}$Br the decay feeds predominantly the ground state of the daughters $^{136}$Xe and $^{86}$Kr whereas in $^{139}$I and $^{89}$Br we will explore the use of n-$\\gamma$- coincidence to study neutron transitions to the first and second excited state...

  11. Electron-capture delayed fission properties of neutron-deficient einsteinium nuclei

    Energy Technology Data Exchange (ETDEWEB)

    Shaughnessy, Dawn A. [Univ. of California, Berkeley, CA (United States)

    2000-01-01

    Electron-capture delayed fission (ECDF) properties of neutron-deficient einsteinium isotopes were investigated using a combination of chemical separations and on-line radiation detection methods. 242Es was produced via the 233U(14N,5n)242Es reaction at a beam energy of 87 MeV (on target) in the lab system, and was found to decay with a half-life of 11 ± 3 seconds. The ECDF of 242Es showed a highly asymmetric mass distribution with an average pre-neutron emission total kinetic energy (TKE) of 183 ± 18 MeV. The probability of delayed fission (PDF) was measured to be 0.006 ± 0.002. In conjunction with this experiment, the excitation functions of the 233U(14N,xn)247-xEs and 233U(15N,xn)248-xEs reactions were measured for 243Es, 244Es and 245Es at projectile energies between 80 MeV and 100 MeV.

  12. Review and comparison of effective delayed neutron fraction calculation methods with Monte Carlo codes

    International Nuclear Information System (INIS)

    Bécares, V.; Pérez-Martín, S.; Vázquez-Antolín, M.; Villamarín, D.; Martín-Fuertes, F.; González-Romero, E.M.; Merino, I.

    2014-01-01

    Highlights: • Review of several Monte Carlo effective delayed neutron fraction calculation methods. • These methods have been implemented with the Monte Carlo code MCNPX. • They have been benchmarked against against some critical and subcritical systems. • Several nuclear data libraries have been used. - Abstract: The calculation of the effective delayed neutron fraction, β eff , with Monte Carlo codes is a complex task due to the requirement of properly considering the adjoint weighting of delayed neutrons. Nevertheless, several techniques have been proposed to circumvent this difficulty and obtain accurate Monte Carlo results for β eff without the need of explicitly determining the adjoint flux. In this paper, we make a review of some of these techniques; namely we have analyzed two variants of what we call the k-eigenvalue technique and other techniques based on different interpretations of the physical meaning of the adjoint weighting. To test the validity of all these techniques we have implemented them with the MCNPX code and we have benchmarked them against a range of critical and subcritical systems for which either experimental or deterministic values of β eff are available. Furthermore, several nuclear data libraries have been used in order to assess the impact of the uncertainty in nuclear data in the calculated value of β eff

  13. JENDL-4.0 benchmarking for effective delayed neutron fraction of fast neutron systems

    International Nuclear Information System (INIS)

    Chiba, Go; Tsuji, Masashi; Sugiyama, Ken-ichiro; Narabayashi, Tadashi

    2011-01-01

    The performance of the latest Japanese evaluated nuclear data library JENDL-4.0 for the prediction of effective delayed neutron fraction β eff is assessed using experimental data of a wide range of fast neutron systems. Covariance data of JENDL-4.0 are used to quantify nuclear-data-induced uncertainties. Calculations with other libraries. JENDL-3.3, ENDF/B-VII.0, and JEFF-3.1, are also carried out for a quantitative comparison. JENDL-4.0 results in good agreement between calculation and experimental values within total uncertainties, and consistency between the differential nuclear data and integral experimental data is confirmed. While the other libraries also show good performance for β eff prediction, there are small differences in the predicted values of β eff among different libraries and ENDF/B-VII.0 gives the most stable results. Furthermore, a simple and convenient procedure to calculate sensitivity profiles of β eff to nuclear data is proposed. (author)

  14. Measurements of periods, relative abundances and absolute yields of delayed neutrons from fast neutron induced fission of {sup 237}Np

    Energy Technology Data Exchange (ETDEWEB)

    Piksaikine, V. [Institute of Physics and Power Engineering, Obninsk (Russian Federation)

    1997-03-01

    The experimental method for measurements of the delayed neutron yields and period is presented. The preliminary results of the total yield, relative abundances and periods are shown comparing with the previously reported values. (J.P.N.)

  15. Safety analysis report for the Neutron Multiplier Facility, 329 Building

    International Nuclear Information System (INIS)

    Rieck, H.G.

    1978-09-01

    Neutron multiplication is a process wherein the flux of a neutron source such as 252 Cf is enhanced by fission reactions that occur in a subcritical assemblage of fissile material. The multiplication factor of the device depends upon the consequences of neutron reactions with matter and is independent of the initial number of neutrons present. Safe utilization of such a device demands that the fissile material assemblage be maintained in a subcritical state throughout all normal and credibly abnormal conditions. Examples of things that can alter the multiplication factor (and degree of subcriticality) are temperature fluctuations, changes in moderator material such as voiding or composition, addition of fissile materials, and change in assembly configuration. The Neutron Multiplier Facility (NMF) utilizes a multiplier- 252 Cf assembly to produce neutrons for activation analysis of organic and inorganic environmental samples and for on-line mass spectrometry analysis of fission products which diffuse from a stationary fissile target (less than or equal to 4 g fissile material) located in the Neutron Multiplier. The NMF annex to the 329 Building provides close proximity to related counting equipment, and delay between sample irradiation and counting is minimized

  16. $\\beta$-decay and $\\beta$-delayed Neutron Emission Measurements at GSI-FRS Beyond N=126, for r-process Nucleosynthesis

    CERN Document Server

    Caballero-Folch, R; Cortès, G; Taín, J L; Agramunt, J; Algora, A; Ameil, F; Ayyad, Y; Benlliure, J; Bowry, M; Calviño, F; Cano-Ott, D; Davinson, T; Dillmann, I; Estrade, A; Evdokimov, A; Faestermann, T; Farinon, F; Galaviz, D; García-Ríos, A; Geissel, H; Gelletly, W; Gernhäuser, R; Gómez-Hornillos, M B; Guerrero, C; Heil, M; Hinke, C; Knöbel, R; Kojouharov, I; Kurcewicz, J; Kurz, N; Litvinov, Y; Maier, L; Marganiec, J; Marta, M; Martínez, T; Montes, F; Mukha, I; Napoli, D R; Nociforo, C; Paradela, C; Pietri, S; Podolyák, Zs; Prochazka, A; Rice, S; Riego, A; Rubio, B; Schaffner, H; Scheidenberger, C; Smith, K; Sokol, E; Steiger, K; Sun, B; Takechi, M; Testov, D; Weick, H; Wilson, E; Winfield, J S; Wood, R; Woods, P J; Yeremin, A

    2014-01-01

    New measurements of very exotic nuclei in the neutron-rich region beyond N=126 have been performed at the GSI facility with the fragment separator (FRS). The aim of the experiment is to determine half-lives and beta-delayed neutron emission branching ratios of isotopes of Hg, Tl and Pb in this region. This contribution summarizes final counting statistics for identification and for implantation, as well as the present status of the data analysis of the half-lives. In summary, isotopes of Pt, Au, Hg, Ti, Pb, Bi, Po, At, Rn and Fr were clearly identified and several of them (Hg208-211, Tl211-215, Pb214-218) were implanted with enough statistics to determine their half-lives. About half of them are expected to be neutron emitters, in such cases it will become possible to obtain the neutron emission probabilities, P-n.

  17. Calculating the effective delayed neutron fraction in the Molten Salt Fast Reactor: Analytical, deterministic and Monte Carlo approaches

    International Nuclear Information System (INIS)

    Aufiero, Manuele; Brovchenko, Mariya; Cammi, Antonio; Clifford, Ivor; Geoffroy, Olivier; Heuer, Daniel; Laureau, Axel; Losa, Mario; Luzzi, Lelio; Merle-Lucotte, Elsa; Ricotti, Marco E.; Rouch, Hervé

    2014-01-01

    Highlights: • Calculation of effective delayed neutron fraction in circulating-fuel reactors. • Extension of the Monte Carlo SERPENT-2 code for delayed neutron precursor tracking. • Forward and adjoint multi-group diffusion eigenvalue problems in OpenFOAM. • Analytical approach for β eff calculation in simple geometries and flow conditions. • Good agreement among the three proposed approaches in the MSFR test-case. - Abstract: This paper deals with the calculation of the effective delayed neutron fraction (β eff ) in circulating-fuel nuclear reactors. The Molten Salt Fast Reactor is adopted as test case for the comparison of the analytical, deterministic and Monte Carlo methods presented. The Monte Carlo code SERPENT-2 has been extended to allow for delayed neutron precursors drift, according to the fuel velocity field. The forward and adjoint eigenvalue multi-group diffusion problems are implemented and solved adopting the multi-physics tool-kit OpenFOAM, by taking into account the convective and turbulent diffusive terms in the precursors balance. These two approaches show good agreement in the whole range of the MSFR operating conditions. An analytical formula for the circulating-to-static conditions β eff correction factor is also derived under simple hypotheses, which explicitly takes into account the spatial dependence of the neutron importance. Its accuracy is assessed against Monte Carlo and deterministic results. The effects of in-core recirculation vortex and turbulent diffusion are finally analysed and discussed

  18. Statistical effects in beta-delayed neutron emission from fission product nuclides

    International Nuclear Information System (INIS)

    McElroy, R.D. Jr.

    1986-01-01

    The delayed neutron spectra for the precursors Rb-93, 94, 95, 96, 97 and Cs-145 were measured by use of the on-line isotope separator facility TRISTAN and a time-of-flight (TOF) spectrometer. Flight paths were used that provided, for energies below 70 keV, a FWHM energy resolution between 2 and 4 percent. Each spectrum showed discrete neutron peaks below 156 keV, with as many as 26 in the Rb-95 spectra. Level densities near the neutron binding energy in the neutron-emitting nuclide were deduced using a missing-level indicator based on a Porter-Thomas distribution of neutron peak intensities. The resulting level density data were compared to the predictions of the Gilbert and Cameron formulism and to those of Dilg, Schantl, Vonach and Uhl. Comparisons were made between the empirically-based level parameter a and the values predicted by each model for Sr-93, 94, 95, 97 and Ba-145. The two models appear, within the uncertainties, to be equally capable of describing these neutron-rich nuclides and equally as capable for them as they are for nuclides in the valley of beta stability. Measurements of the neutron strength function are sometimes possible with the present TOF system for neutron decays with competing neutron branches to levels in the grandchild nucleus. A value for the d-wave strength function of Sr-96 is found to be (4.2 +- 1.1)/10 4 . Improvements in the TOF system, allowing the measurement of the neutron strength function for the more general case, are discussed. 72 refs., 56 figs., 16 tabs

  19. Online failed fuel identification using delayed neutron detector signals in pool type reactors

    International Nuclear Information System (INIS)

    Upadhyay, Chandra Kant; Sivaramakrishna, M.; Nagaraj, C.P.; Madhusoodanan, K.

    2011-01-01

    In todays world, nuclear reactors are at the forefront of modern day innovation and reactor designs are increasingly incorporating cutting edge technology. It is of utmost importance to detect failure or defects in any part of a nuclear reactor for healthy operation of reactor as well as the safety aspects of the environment. Despite careful fabrication and manufacturing of fuel pins, there is a chance of clad failure. After fuel pin clad rupture takes place, it allows fission products to enter in to sodium pool. There are some potential consequences due to this such as Total Instantaneous Blockage (TIB) of coolant and primary component contamination. At present, the failed fuel detection techniques such as cover gas monitoring (alarming the operator), delayed neutron detection (DND-automatic trip) and standalone failed fuel localization module (FFLM) are exercised in various reactors. The first technique is a quantitative measurement of increase in the cover gas activity background whereas DND system causes automatic trip on detecting certain level of activity during clad wet rupture. FFLM is subsequently used to identify the failed fuel subassembly. The later although accurate, but mainly suffers from downtime and reduction in power during identification process. The proposed scheme, reported in this paper, reduces the operation of FFLM by predicting the faulty sector and therefore reducing reactor down time and thermal shocks. The neutron evolution pattern gets modulated because fission products are the delay neutron precursors. When they travel along with coolant to Intermediate heat Exchangers, experienced three effects i.e. delay; decay and dilution which make the neutron pulse frequency vary depending on the location of failed fuel sub assembly. This paper discusses the method that is followed to study the frequency domain properties, so that it is possible to detect exact fuel subassembly failure online, before the reactor automatically trips. (author)

  20. Application of neutron noise analysis to a swimming pool research reactor

    International Nuclear Information System (INIS)

    Behringer, K.; Lescano, V.H.; Meier, F.; Phildius, J.; Winkler, H.

    1982-01-01

    This work is part of a programme of establishing practical applications of neutron noise techniques to a swimming pool research reactor and deals with two different items: (1) The identification of local boiling caused e.g. by a partial blockage of the coolant flow in a fuel element. Local boiling can easily lead to a burn-out situation. The onset of boiling can be detected by neutron noise analysis and a boiling detection system is presently under development. (2) The measurement of the time evolution of the reactivity induced by xenon after reactor shut-down by an on-line reactivity meter based on neutron noise analysis. From the data, the prompt neutron decay constant at delayed critical, the equilibrium xenon reactivity worth, and an estimate of the average steady-state power flux in the core before reactor shut-down were obtained. (author)

  1. Core Power Control of the fast nuclear reactors with estimation of the delayed neutron precursor density using Sliding Mode method

    International Nuclear Information System (INIS)

    Ansarifar, G.R.; Nasrabadi, M.N.; Hassanvand, R.

    2016-01-01

    Highlights: • We present a S.M.C. system based on the S.M.O for control of a fast reactor power. • A S.M.O has been developed to estimate the density of delayed neutron precursor. • The stability analysis has been given by means Lyapunov approach. • The control system is guaranteed to be stable within a large range. • The comparison between S.M.C. and the conventional PID controller has been done. - Abstract: In this paper, a nonlinear controller using sliding mode method which is a robust nonlinear controller is designed to control a fast nuclear reactor. The reactor core is simulated based on the point kinetics equations and one delayed neutron group. Considering the limitations of the delayed neutron precursor density measurement, a sliding mode observer is designed to estimate it and finally a sliding mode control based on the sliding mode observer is presented. The stability analysis is given by means Lyapunov approach, thus the control system is guaranteed to be stable within a large range. Sliding Mode Control (SMC) is one of the robust and nonlinear methods which have several advantages such as robustness against matched external disturbances and parameter uncertainties. The employed method is easy to implement in practical applications and moreover, the sliding mode control exhibits the desired dynamic properties during the entire output-tracking process independent of perturbations. Simulation results are presented to demonstrate the effectiveness of the proposed controller in terms of performance, robustness and stability.

  2. Two specialized delayed-neutron detector designs for assays of fissionable elements in water and sediment samples

    International Nuclear Information System (INIS)

    Balestrini, S.J.; Balagna, J.P.; Menlove, H.O.

    1976-01-01

    Two specialized neutron-sensitive detectors are described which are employed for rapid assays of fissionable elements by sensing for delayed neutrons emitted by samples after they have been irradiated in a nuclear reactor. The more sensitive of the two detectors, designed to assay for uranium in water samples, is 40% efficient; the other, designed for sediment sample assays, is 27% efficient. These detectors are also designed to operate under water as an inexpensive shielding against neutron leakage from the reactor and neutrons from cosmic rays. (Auth.)

  3. Californium-252 neutron activation analysis of high-level processed nuclear tank waste

    International Nuclear Information System (INIS)

    Troyer, G.L.; Purcell, M.A.

    2000-01-01

    The basis for production assessment of the vitrification of Hanford nuclear fuel reprocessing wastes will be high-precision measurements of the elemental sodium content. However, the chemical analysis of both radioactive and nonradioactive components in nuclear waste can be challenged by high radiation dose rates. The dose rates compromise many analytical techniques as well as pose personnel dosimetry risks. In many cases, reduction of dose rates through dilution compromises the precision and sensitivity for certain key components. The use of neutron activation analysis (NAA) provides a method of analysis that avoids the need for dilutions or extensive sample preparation. These waste materials also contain trace quantities of fissionable isotopes, which, through neutron activation, can be estimated by delayed neutron counting of fissioned fragments

  4. Instrumental neutron activation analysis as a routine method for rock analysis

    International Nuclear Information System (INIS)

    Rosenberg, R.J.

    1977-06-01

    Instrumental neutron activation methods for the analysis of geological samples have been developed. Special emphasis has been laid on the improvement of sensitivity and accuracy in order to maximize tha quality of the analyses. Furthermore, the procedures have been automated as far as possible in order to minimize the cost of the analysis. A short review of the basic literature is given followed by a description of the principles of the method. All aspects concerning the sensitivity are discussed thoroughly in view of the analyst's possibility of influencing them. Experimentally determined detection limits for Na, Al, K, Ca, Sc, Cr, Ti, V, Mn, Fe, Ni, Co, Rb, Zr, Sb, Cs, Ba, La, Ce, Nd, Sm, Eu, Gd, Tb, Dy, Yb, Lu, Hf, Ta, Th and U are given. The errors of the method are discussed followed by actions taken to avoid them. The most significant error was caused by flux deviation, but this was avoided by building a rotating sample holder for rotating the samples during irradiation. A scheme for the INAA of 32 elements is proposed. The method has been automated as far as possible and an automatic γ-spectrometer and a computer program for the automatic calculation of the results are described. Furthermore, a completely automated uranium analyzer based on delayed neutron counting is described. The methods are discussed in view of their applicability to rock analysis. It is stated that the sensitivity varies considerably from element to element and instrumental activation analysis is an excellent method for the analysis of some specific elements like lanthanides, thorium and uranium but less so for many other elements. The accuracy is good varying from 2% to 10% for most elements. Instrumental activation analysis for most elements is rather an expensive method there being, however, a few exceptions. The most important of these is uranium. The analysis of uranium by delayed neutron counting is an inexpensive means for the analysis of large numbers of samples needed for

  5. Study of neutron rich nuclei by delayed neutron decay using the Tonnerre multidetector; Etude de la decroissance par neutrons retardes de noyaux legers riches en neutrons avec le multidetecteur tonnerre

    Energy Technology Data Exchange (ETDEWEB)

    Timis, C.N

    2001-07-01

    A new detection array for beta delayed neutrons was built. It includes up to 32 plastic scintillation counters 180 cm long located at 120 cm from the target. Neutron energy spectra are measured by time-of-flight in the 300 keV-15 MeV range with good energy resolution. The device was tested with several known nuclei. Its performances are discussed in comparison with Monte Carlo simulations. They very high overall detection efficiency on the TONNERRE array made it possible to study one and two neutron emission of {sup 11}Li. A complete decay scheme was obtained. The {sup 33}Mg and {sup 35}Al beta decays were investigated for the first time by neutron and gamma spectroscopy. Complete decay schemes were established and compared to large scale shell-model calculations. (authors)

  6. Beta-decay rate and beta-delayed neutron emission probability of improved gross theory

    Science.gov (United States)

    Koura, Hiroyuki

    2014-09-01

    A theoretical study has been carried out on beta-decay rate and beta-delayed neutron emission probability. The gross theory of the beta decay is based on an idea of the sum rule of the beta-decay strength function, and has succeeded in describing beta-decay half-lives of nuclei overall nuclear mass region. The gross theory includes not only the allowed transition as the Fermi and the Gamow-Teller, but also the first-forbidden transition. In this work, some improvements are introduced as the nuclear shell correction on nuclear level densities and the nuclear deformation for nuclear strength functions, those effects were not included in the original gross theory. The shell energy and the nuclear deformation for unmeasured nuclei are adopted from the KTUY nuclear mass formula, which is based on the spherical-basis method. Considering the properties of the integrated Fermi function, we can roughly categorized energy region of excited-state of a daughter nucleus into three regions: a highly-excited energy region, which fully affect a delayed neutron probability, a middle energy region, which is estimated to contribute the decay heat, and a region neighboring the ground-state, which determines the beta-decay rate. Some results will be given in the presentation. A theoretical study has been carried out on beta-decay rate and beta-delayed neutron emission probability. The gross theory of the beta decay is based on an idea of the sum rule of the beta-decay strength function, and has succeeded in describing beta-decay half-lives of nuclei overall nuclear mass region. The gross theory includes not only the allowed transition as the Fermi and the Gamow-Teller, but also the first-forbidden transition. In this work, some improvements are introduced as the nuclear shell correction on nuclear level densities and the nuclear deformation for nuclear strength functions, those effects were not included in the original gross theory. The shell energy and the nuclear deformation for

  7. The universal library of fission products and delayed neutron group yields

    International Nuclear Information System (INIS)

    Koldobskiy, A.B.; Zhivun, V.M.

    1997-01-01

    A new fission product yield library based on the Semiempirical method for the estimation of their mass and charge distribution is described. Contrary to other compilations, this library can be used with all possible excitation energies of fissionable actinides. The library of delayed neutron group yields, based on the fission product yield compilation, is described as well. (author). 15 refs, 4 tabs

  8. Modern Trends in Neutron Activation Analysis. Applications to some African Environmental Samples

    International Nuclear Information System (INIS)

    Hassan, A.M.

    2009-01-01

    This review covers the results of several published articles which deal with the modern trends in neutron activation analysis techniques using some of African research reactors for some environmental samples. The samples used have been collected from different areas in Egypt, South Africa, Ghana, Morocco, Nigeria, and Algeria. The neutron irradiation facilities and the advanced detection systems in each country are outlined. The prompt and delayed gamma-rays emitted due to neutron capture have been applied for investigation of the elemental constituents of such samples. Covered applications include exploration, mining, industrial environment, pollution of air, foodstuffs, soils and irrigation water samples. Some of the developed software programmes as well as the modern methods of data analysis are presented. The thermal and epithermal neutron activation analysis techniques have been applied for estimation of major, minor and trace elements in each material. Some of these data are presented with several comments.

  9. Influence of fission product transport on delayed neutron precursors and decay heat sources in LMFBR accidents

    International Nuclear Information System (INIS)

    Apperson, C.E. Jr.

    1981-01-01

    A method is presented for studying the influence of fission product transpot on delayed neutron precursors and decay heat sources during Liquid Metal Fast Breeder Reactor (LMFBR) unprotected accidents. The model represents the LMFBR core as a closed homogeneous cell. Thermodynamic phase equilibrium theory is used to predict fission product mobility. Reactor kinetics behavior is analyzed by an extension of point kinetics theory. Group dependent delayed neutron precursor and decay heat source retention factors, which represent the fraction of each group retained in the fuel, are developed to link the kinetics and thermodynamics analysis. Application of the method to a highly simplified model of an unprotected loss-of-flow accident shows a time delay on the order of 10 ms is introduced in the predisassembly power history if fission product motion is considered when compared to the traditional transient solution. The post-transient influence of fission product transport calculated by the present model is a 24 percent reduction in the decay heat level in the fuel material which is similar to traditional approximations. Isotopes of the noble gases, Kr and Xe, and the elements I and Br are shown to be very mobile and are responsible for a major part of the observed effects. Isotopes of the elements Cs, Se, Rb, and Te were found to be moderately mobile and contribute to a lesser extent to the observed phenomena. These results obtained from the application of the described model confirm the initial hypothesis that sufficient fission product transport can occur to influence a transient. For these reasons, it is concluded that extension of this model into a multi-cell transient analysis code is warranted

  10. Determination of delayed neutrons source in the frequency domain based on in-pile oscillation measurements

    International Nuclear Information System (INIS)

    Yedvab, Y.; Reiss, I.; Bettan, M.; Harari, R.; Grober, A.; Ettedgui, H.; Caspi, E. N.

    2006-01-01

    A method for determining delayed neutrons source in the frequency domain based on measuring power oscillations in a non-critical reactor is presented. This method is unique in the sense that the delayed neutrons source is derived from the dynamic behavior of the reactor, which serves as the measurement system. An algorithm for analyzing power oscillation measurements was formulated, which avoids the need for a multi-parameter non-linear fit process used by other methods. Using this algorithm results of two sets of measurements performed in IRR-I and IRR-II (Israeli Research Reactors I and II) are presented. The agreement between measured values from both reactors and calculated values based on Keepin (and JENDL-3.3) group parameters is very good. (authors)

  11. Measurement of the most exotic beta-delayed neutron emitters at N=50 and N=126

    Science.gov (United States)

    Dillmann, Iris

    2017-09-01

    Beta-delayed neutron (βn)-emission will be the dominant decay mechanism of neutron-rich nuclei and plays an important role in the stellar nucleosynthesis of heavy elements in the ``r process''. It leads to a detour of the material β-decaying back to stability and the released neutrons increase the neutron-to-seed ratio, and are re-captured during the freeze-out phase and thus influence the final solar r-abundance curve. Thus the neutron branching ratio of very neutron-rich isotopes is a crucial parameter in astrophysical simulations. In addition, β-decay half-lives can be deduced from the time-dependent detection of βn's. I will talk about two recent experimental campaigns. The neutron detector BELEN was used at GSI Darmstadt to measure half-lives and neutron-branching ratios of the heaviest presently accessible βn-emitters at N=126. For isotopes between 204Au and 220Bi nine half-lives and eight neutron-branching ratios were measured for the first time and provide an important input for benchmarking theoretical models in this mass region. Its successor is the BRIKEN detector (``Beta-delayed neutron measurements at RIKEN for nuclear structure, astrophysics, and applications''), the most efficient neutron detector used so far for nuclear structure studies. In conjunction with two clover detectors and the ``Advanced Implantation Detector Array'' (AIDA) the setup has been used a few months ago to measure the most neutron-rich isotopes around 78Ni, 132Sn, and the Rare Earth Region. Some preliminary results are shown from the campaign covering the 78Ni region where the neutron-branching ratio of 78Ni and 28 more isotopes were measured for the first time, as well as the half-lives of 20 isotopes. The BRIKEN campaign aims to (re-)measure almost all βn-emitters between 76Co and 167Eu, many of them for the first time. An extension of the campaign to lighter masses is planned. This work has been supported by the NSERC and NRC in Canada, the US DOE, the Spanish

  12. Measurement of the Effective Delayed Neutron Fraction in Three Different FR0-cores

    Energy Technology Data Exchange (ETDEWEB)

    Moberg, L; Kockum, J

    1972-06-15

    The effective delayed neutron fraction, beta{sub eff}, has been measured in the three cores 3, 5 and 8 of the fast zero-power reactor FR0. The variance-to-mean method, in which the statistical fluctuations of the neutron density in the reactor is studied, was used. A 3He-gas scintillator was placed in the reflector and used as a neutron detector. It was made more sensitive to fast neutrons by surrounding it with polythene. Its efficiency, expressed as the number of counts per fission in the reactor, was determined using fission chambers with known efficiency placed in the core. The space distribution of the fission rate in the core was determined by foil activation technique. The experimental results were compared with theoretical beta{sub eff}-values calculated with perturbation theory. The difference was about 3 % which is of the same order as the accuracy in the experimental values

  13. Measurement of 235U content and flow of UF6 using delayed neutrons or gamma rays following induced fission

    International Nuclear Information System (INIS)

    Stromswold, D.C.; Peurrung, A.J.; Reeder, P.L.; Perkins, R.W.

    1996-06-01

    Feasibility experiments conducted at Pacific Northwest National Laboratory demonstrate that either delayed neutrons or energetic gamma rays from short-lived fission products can be used to monitor the blending of UF 6 gas streams. A 252 Cf neutron source was used to induce 235 U fission in a sample, and delayed neutrons and gamma rays were measured after the sample moved open-quotes down-stream.close quotes The experiments used a UO 2 powder that was transported down the pipe to simulate the flowing UF 6 gas. Computer modeling and analytic calculation extended the test results to a flowing UF 6 gas system. Neutron or gamma-ray measurements made at two downstream positions can be used to indicate both the 235 U content and UF 6 flow rate. Both the neutron and gamma-ray techniques have the benefits of simplicity and long-term reliability, combined with adequate sensitivity for low-intrusion monitoring of the blending process. Alternatively, measuring the neutron emission rate from (a, n) reactions in the UF 6 provides an approximate measure of the 235 U content without using a neutron source to induce fission

  14. Neutron lifetimes behavior analysis considering the two-region kinetic model in the IPEN/MB-01 reactor

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-11-11

    This is a complementary work about the behavior analysis of the neutron lifetimes that was developed in the IPEN/MB-01 nuclear reactor facility. The macroscopic neutron noise technique was experimentally employed using pulse mode detectors for two stages of control rods insertion, where a total of twenty levels of subcriticality have been carried out. It was also considered that the neutron reflector density was treated as an additional group of delayed neutrons, being a sophisticated approach in the two-region kinetic theoretical model.

  15. Neutron lifetimes behavior analysis considering the two-region kinetic model in the IPEN/MB-01 reactor

    International Nuclear Information System (INIS)

    Gonnelli, Eduardo; Diniz, Ricardo

    2014-01-01

    This is a complementary work about the behavior analysis of the neutron lifetimes that was developed in the IPEN/MB-01 nuclear reactor facility. The macroscopic neutron noise technique was experimentally employed using pulse mode detectors for two stages of control rods insertion, where a total of twenty levels of subcriticality have been carried out. It was also considered that the neutron reflector density was treated as an additional group of delayed neutrons, being a sophisticated approach in the two-region kinetic theoretical model

  16. In-situ elemental analysis of coal by neutron activation

    International Nuclear Information System (INIS)

    Mikesell, J.L.; Senftle, F.E.; Tanner, A.B.

    1986-01-01

    The U.S. Geological Survey (USGS) has worked to develop neutron techniques for the borehole measurement of the elemental composition of ores since 1969, and first demonstrated a borehole ultimate analysis of coal in 1977. Borehole measurements such as these permit real-time evaluation of coal quality without the expense of coring or the delays associated with laboratory analyses. Two technological innovations make such measurements possible: the availability, from Savannah River Operations Office, DOE, of small californium-252 (/sup 252/Cf) fission neutron sources, and the development, by USGS and Princeton Gamma-Techn, of the melting-cryogen-cooled high-purity germanium borehole gamma-ray detector. A technique of relating mass fractions to measured gamma-ray intensities, which eliminates the need for detailed knowledge of the geometry of the neutron distribution, is used to calculate elemental compositions without resorting to the test pits or computer borehole modeling. In coal, all of the major constituents (C, H, N, S, Si, Al, Fe, Ti) except oxygen can be determined quantitatively by thermal neutron capture gamma-ray spectroscopy

  17. Perspectives for on-line analysis of bauxite by neutron irradiation

    Science.gov (United States)

    Beurton, Gabriel; Ledru, Bertrand; Letourneur, Philippe

    1995-03-01

    The interest in bauxite as a major source of alumina results in a strong demand for on-line instrumentation suitable for sorting, blending, and processing operations at the bauxite mine and for monitoring instrumentation in the Bayer process. The results of laboratory experiments based on neutron interactions with bauxite are described. The technique was chosen in order to overcome the problem of spatial heterogeneity in bulk mineral analysis. The evaluated elements contributed to approximately 99.5% of the sample weight. In addition, the measurements provide valuable information on physical parameters such as density, hygrometry, and material flow. Using a pulsed generator, the analysis system offers potential for on-line measurements (borehole logging or conveyor belt). An overall description of the experimental set-up is given. The experimental data include measurements of natural radioactivity, delayed radioactivity induced by activation, and prompt gamma rays following neutron reaction. In situ applications of neutron interactions provide continuous analysis and produce results which are more statistically significant. The key factors contributing to advances in industrial applications are the development of high count rate gamma spectroscopy and computational tools to design measurement systems and interpret their results.

  18. Determination of the decay constants and relative abundances of delayed neutrons by noise analysis in zero-power reactors

    International Nuclear Information System (INIS)

    Diniz, Ricardo

    2005-01-01

    A reactor noise approach has been employed at the IPEN/MB-01 research reactor facility in order to determine experimentally the effective delayed neutron parameters β i and λ i in a six group model and assuming the point reactor. The method can be considered a novice one because exploits the very low frequency domain of the spectral densities. The proposed method has some advantages to other in-pile methods since it does not disturb the reactor system and consequently does not 'excite' any sort of harmonic modes. As a byproduct and a consistency check, the β eff parameter was obtained without the need of the Diven factor and the power normalization and it is in excellent agreement with independent measurements. The theory/experiment comparison shows that for the abundances the JENDL 3.3 presents the best performance while for the decay constants the revised version of ENDF/B-VI.8 shows the best agreement. The best performance for the β eff determination is obtained with JENDL3.3. In contrast, ENDF/B-VI.8 and its revised version performed at LANL overestimate β eff by as much as 4%. The β eff results of this work support totally the proposal of reducing the thermal delayed neutron number for 235 U fission as made by Sakurai and Okajima. A new observed effect related to the correlation between the fluctuations of both measurement channels is also presented and discussed. This effect can be considered as an indirect evidence for the use of the point reactor model in this work as well as a possible useful tool in the understanding of reactor dynamics. (author)

  19. A two-dimensional detector with delay line readout for slow neutron fields measurements

    International Nuclear Information System (INIS)

    Cheremukhina, G.A.; Chernenko, S.P.; Ivanov, A.B.

    1992-01-01

    This article presents the description of a two-dimensional detector of slow neutrons together with its readout and data acquisition electronics based on a PC/AT> The detector with a sensitive area of 260x140 mm 2 is based on a high pressure multiwire proportional chamber with delay line readout and gas filling of 3.0 atm. 3 He + propane. 25 refs.; 10 figs.; 2 tabs

  20. Fast neutron activation analysis by means of low voltage neutron generator

    Directory of Open Access Journals (Sweden)

    M.E. Medhat

    Full Text Available A description of D-T neutron generator (NG is presented. This machine can be used for fast neutron activation analysis applied to determine some selected elements, especially light elements, in different materials. Procedure of neutron flux determination and efficiency calculation is described. Examples of testing some Egyptian natural cosmetics are given. Keywords: Neutron generator, Fast neutron activation analysis, Elemental analysis

  1. Study on calculation methods for the effective delayed neutron fraction

    International Nuclear Information System (INIS)

    Irwanto, Dwi; Obara, Toru; Chiba, Go; Nagaya, Yasunobu

    2011-03-01

    The effective delayed neutron fraction β eff is one of the important neutronic parameters from a view point of a reactor kinetics. Several Monte-Carlo-based methods to estimate β eff have been proposed to date. In order to quantify the accuracy of these methods, we study calculation methods for β eff by analyzing various fast neutron systems including the bare spherical systems (Godiva, Jezebel, Skidoo, Jezebel-240), the reflective spherical systems (Popsy, Topsy, Flattop-23), MASURCA-R2 and MASURCA-ZONA2, and FCA XIX-1, XIX-2 and XIX-3. These analyses are performed by using SLAROM-UF and CBG for the deterministic method and MVP-II for the Monte Carlo method. We calculate β eff with various definitions such as the fundamental value β 0 , the standard definition, Nauchi's definition and Meulekamp's definition, and compare these results with each other. Through the present study, we find the following: The largest difference among the standard definition of β eff , Nauchi's β eff and Meulekamp's β eff is approximately 10%. The fundamental value β 0 is quite larger than the others in several cases. For all the cases, Meulekamp's β eff is always higher than Nauchi's β eff . This is because Nauchi's β eff considers the average neutron multiplicity value per fission which is large in the high energy range (1MeV-10MeV), while the definition of Meulekamp's β eff does not include this parameter. Furthermore, we evaluate the multi-generation effect on β eff values and demonstrate that this effect should be considered to obtain the standard definition values of β eff . (author)

  2. Project delay analysis of HRSG

    Science.gov (United States)

    Silvianita; Novega, A. S.; Rosyid, D. M.; Suntoyo

    2017-08-01

    Completion of HRSG (Heat Recovery Steam Generator) fabrication project sometimes is not sufficient with the targeted time written on the contract. The delay on fabrication process can cause some disadvantages for fabricator, including forfeit payment, delay on HRSG construction process up until HRSG trials delay. In this paper, the author is using semi quantitative on HRSG pressure part fabrication delay with configuration plant 1 GT (Gas Turbine) + 1 HRSG + 1 STG (Steam Turbine Generator) using bow-tie analysis method. Bow-tie analysis method is a combination from FTA (Fault tree analysis) and ETA (Event tree analysis) to develop the risk matrix of HRSG. The result from FTA analysis is use as a threat for preventive measure. The result from ETA analysis is use as impact from fabrication delay.

  3. Application of the activation analysis using the method of retarded fission neutrons counting for the determination of some fissionable nuclides

    International Nuclear Information System (INIS)

    Armelin, M.J.A.

    1984-01-01

    A system for the detection and counting of delayed neutrons which allows the analysis of some fissile and fertile nuclides, in samples of milligram size, was developed. This was applied for the analysis of natural uranium and thorium and also for determining the 235 U/ 238 U ratio in non-irradiated samples which contain uranium with different degrees of enrichment in 235 U. The spectrum of activated neutrons was varied in order to discriminate the nuclides, by covering or not the sample with a material (cadmium or boron) able to absorb low energy neutrons. Determination of 235 U/ 238 U ratios, through the number of delayed neutrons, was made by drawing a calibration curve using standards ranging from 0.5% to 93% on 235 U; the accuracy of the method was also examined. In a first step, conditions for a simultaneous and non-destructive analysis of uranium and thorium were developed. The interference between these two nuclides was studied, using simulated samples. Real samples were provided by Nuclemon and IAEA. For samples with uranium concentration in the range of percentages and thorium concentration of some ppm, uranium interferes in the determination of thorium through the non-destructive analytical method. For this case, a fast and quantitative chemical method was studied which allows for the separation of thorium from uranium before the determination of throrium concentration by counting the delayed fission neutrons. It was found that the results obtained by both destructive and non-destructive methods are very consistent and can be considered statistically equivalent within a confidence level of 95%. (Author) [pt

  4. Study and building of a detection array for delayed neutrons: TONNERRE; Etude et realisation d`un ensemble de detection pour neutrons retardes: TONNERRE

    Energy Technology Data Exchange (ETDEWEB)

    Martin, Thierry [Lab. de Physique Corpusculaire, Caen Univ., 14 - Caen (France)

    1998-11-09

    This work has been undertaken within a French-Romanian collaboration in order to build a high efficiency detector array for delayed neutrons: barrel-shaped TONNERRE. Some neutron-rich nuclei decay through 1, 2 or 3 neutron emission after {beta}{sup -} decay. More exotic nuclei will be produced by SPIRAL at GANIL. An array with high efficiency and good resolution is then required. Thirty two BC400 plastic scintillators (160 x 20 x 4 cm{sup 3}) allow us to get the time of flight neutron spectra. They are bent for uniform flight path and viewed by a photomultiplier tube at both ends. Simulations have allowed to establish scintillator size and to minimize light attenuation. Intrinsic efficiency and crosstalk have been measured with {sup 252}Cf and compared to GEANT. 1 to 5 MeV neutrons are detected with good timing and position properties. Other counters will be built for neutrons from 300 keV to 1 MeV. Planned to run at several particle accelerators (GANIL, CERN, and others), TONNERRE is modular and many geometries are possible. (author) 48 refs., 78 figs., 20 tabs.

  5. Assessing neutron generator output using neutron activation of silicon

    International Nuclear Information System (INIS)

    Kehayias, Pauli M.; Kehayias, Joseph J.

    2007-01-01

    D-T neutron generators are used for elemental composition analysis and medical applications. Often composition is determined by examining elemental ratios in which the knowledge of the neutron flux is unnecessary. However, the absolute value of the neutron flux is required when the generator is used for neutron activation analysis, to study radiation damage to materials, to monitor the operation of the generator, and to measure radiation exposure. We describe a method for absolute neutron output and flux measurements of low output D-T neutron generators using delayed activation of silicon. We irradiated a series of silicon oxide samples with 14.1 MeV neutrons and counted the resulting gamma rays of the 28 Al nucleus with an efficiency-calibrated detector. To minimize the photon self-absorption effects within the samples, we used a zero-thickness extrapolation technique by repeating the measurement with samples of different thicknesses. The neutron flux measured 26 cm away from the tritium target of a Thermo Electron A-325 D-T generator (Thermo Electron Corporation, Colorado Springs, CO) was 6.2 x 10 3 n/s/cm 2 ± 5%, which is consistent with the manufacturer's specifications

  6. Assessing neutron generator output using neutron activation of silicon

    Energy Technology Data Exchange (ETDEWEB)

    Kehayias, Pauli M. [Body Composition Laboratory, Jean Mayer United States Department of Agriculture Human Nutrition Research Center on Aging, Tufts University, Boston, MA 02111 (United States); Kehayias, Joseph J. [Body Composition Laboratory, Jean Mayer United States Department of Agriculture Human Nutrition Research Center on Aging, Tufts University, Boston, MA 02111 (United States)]. E-mail: joseph.kehayias@tufts.edu

    2007-08-15

    D-T neutron generators are used for elemental composition analysis and medical applications. Often composition is determined by examining elemental ratios in which the knowledge of the neutron flux is unnecessary. However, the absolute value of the neutron flux is required when the generator is used for neutron activation analysis, to study radiation damage to materials, to monitor the operation of the generator, and to measure radiation exposure. We describe a method for absolute neutron output and flux measurements of low output D-T neutron generators using delayed activation of silicon. We irradiated a series of silicon oxide samples with 14.1 MeV neutrons and counted the resulting gamma rays of the {sup 28}Al nucleus with an efficiency-calibrated detector. To minimize the photon self-absorption effects within the samples, we used a zero-thickness extrapolation technique by repeating the measurement with samples of different thicknesses. The neutron flux measured 26 cm away from the tritium target of a Thermo Electron A-325 D-T generator (Thermo Electron Corporation, Colorado Springs, CO) was 6.2 x 10{sup 3} n/s/cm{sup 2} {+-} 5%, which is consistent with the manufacturer's specifications.

  7. Neutron activation analysis in Romania

    International Nuclear Information System (INIS)

    Apostolescu, St.

    1985-01-01

    The following basic nuclear facilities are used for neutron activation analysis: a 2000 KW VVR-S Nuclear Reactor, a U-200 Cyclotron, a 30 MeV Betatron, several 14 MeV neutron generators and a king size High Voltage tandem Van de'Graaff accelerator. The main domains of application of the thermal neutron activation analysis are: geology and mining, processing of materials, environment and biology, achaeology. Epithermal neutron activation analysis has been used for determination of uranium and thorium in ores with high Th/U ratios or high rare earth contents. One low energy accelerator, used as 14.1 Mev neutron source, is provided with special equipmen for oxigen and low mass elements determination. An useful alternating way to support fast neutron activation analysis is an accurate theoretical description of the fast neutron induced reactions based on the statistical model (Hauser-Feubach STAPRE code) and the preequilibrium decay geometry dependent model. A gravitational sample changer has been installed at the end of a beam line of the Cyclotron, which enables to perform charged particles activation analysis for protein determination in grains

  8. Neutron activation analysis at CDTN/CNEN using the IPR-R1 Triga Mark I reactor

    International Nuclear Information System (INIS)

    Menezes, Maria Angela de B.C.; Maretti Junior, Fausto; Kastner, Geraldo Frederico; Amaral, Angela Maria; Souza, Wagner de

    2009-01-01

    This paper describes in summary the activities developed by the Laboratory for Neutron Activation Analysis since the starting up of the IPR-R1 TRIGA Mark I research reactor in 1960. This Laboratory is located at Centro de Desenvolvimento da Tecnologia Nuclear (Nuclear Technology Development Centre) / Comissao Nacional de Energia Nuclear (Brazilian Commission for Nuclear Energy), CDTN/CNEN. The activities of the Laboratory comprise the delayed fission neutron activation analysis, instrumental (comparative and parametric methods) and radiochemical / chemical methods. These methods are responsible for significant percentage of CDTN's analytical demand, meeting the clients' analytical needs and researches developed by the Laboratory, by CDTN and by other institutions. Over the years the work has been linked to the goals of the country and the institutions. Nowadays the neutron activation analysis is responsible for 70% of the analytical demand and the k 0 - Instrumental method for 80% of this demand answering clients' request and researches. In Brazil, CDTN is the only Institute that fully masters the Instrumental Neutron Activation Analysis k 0 -method using its own nuclear reactor. (author)

  9. Neutronics of the IFMIF neutron source: development and analysis

    International Nuclear Information System (INIS)

    Wilson, P.P.H.

    1999-01-01

    The accurate analysis of this system required the development of a code system and methodology capable of modelling the various physical processes. A generic code system for the neutronics analysis of neutron sources has been created by loosely integrating existing components with new developments: the data processing code NJOY, the Monte Carlo neutron transport code MCNP, and the activation code ALARA were supplemented by a damage data processing program, damChar, and integrated with a number of flexible and extensible modules for the Perl scripting language. Specific advances were required to apply this code system to IFMIF. Based on the ENDF-6 data format requirements of this system, new data evaluations have been implemented for neutron transport and activation. Extensive analysis of the Li(d, xn) reaction has led to a new MCNP source function module, M c DeLi, based on physical reaction models and capable of accurate and flexible modelling of the IFMIF neutron source term. In depth analyses of the neutron flux spectra and spatial distribution throughout the high flux test region permitted a basic validation of the tools and data. The understanding of the features of the neutron flux provided a foundation for the analyses of the other neutron responses. (orig./DGE) [de

  10. Research on activation analysis using short-lived isotopes and a multi-purpose isotopic neutron irradiator. Part of a coordinated programme on on-line X-ray and neutron techniques for industrial process control

    International Nuclear Information System (INIS)

    Ozek, F.

    1981-02-01

    A method of cyclic activation analysis (CA) has been studied and applied. A theoretical comparison between cyclic and conventional neutron activation analysis of gold has been made. The optimum number of cycles in cyclic activation have been investigated and an equation for the rapid calculation of the number of cycles is proposed. The isotopic neutron irradiation system including the 5Ci Pu-Be neutron source was designed and constructed. The system is flexible and transportable and is capable of carrying out prompt and conventional delay gamma-ray analysis and cyclic activation of bulk materials. The advantages as well as the disadvantages of neutron activation analysis with the use of short-lived nuclides were considered, and can be summarized as follows: Advantages: saturation factor approaches unity, speed of analysis, low cost of analysis, increased selectivity, reduced matrix activities. Disadvantages: proximity of neutron source, chemical separation hardly possible or impossible, total number of counts low. Low counting rates can be substantially increased by applying the technique of ''cyclic activation'', which is another reason by the use of short-lived isotopes in neutron activation analysis is steadily becoming more attractive

  11. Analysis of mean lifetime for capture of neutrons in boron-loaded plastic scintillators

    Energy Technology Data Exchange (ETDEWEB)

    Kamykowski, E.A. (Grumman Corp., Bethpage, NY (USA). Research Center)

    1990-12-20

    The commercial availabiltiy of boron-loaded organic scintillators has led to the development of neutron detectors that operate as ''electronically'' black, totally absorbing spectrometers. The key to the enhanced spectroscopy is the delayed capture of nearly thermalized neutrons by {sup 10}B that can occur within a few microseconds after the energy pulse from prompt proton recoils. Accurate information regarding the mean lifetime is important for correct setting of the timing logic of the detection system to obtain good neutron detection efficiency with a low chance coincidence rate. In this paper we present an analysis of the mean lifetime for neutron capture for the boron-loaded plastic BC454. Measurements of the capture time constant obtained with a 7.62 cm diameter, 10.16 cm long detector are compared with values computed with the time-dependent Monte Carlo neutron transport code MCNP. Additional analyses using MCNP examine the dependence of the mean lifetime on the boron concentration, the detector's dimensions and the incident neutron energy. (orig.).

  12. Study of some environmental problem in egypt using neutron activation analysis techniques

    International Nuclear Information System (INIS)

    El-Karim, A.H.M.G.

    2003-01-01

    this thesis deals with the investigation of the possibility of using the new (second) egyptian research reactor (ETRR-2) at Inshas (22 MW) for the neutron activation analysis (ANN) of trace elements, particularly in air dust, collected from cairo and some other cities of egypt. in this concern chapter 1 gives an introduction about the activation methods in general, describing the various techniques used and a comparison of the methods with other instrumental methods of analysis . as a main classification, the neutron activation methods involve prompt γ-ray NAA and delayed γ-ray NAA; cyclic NAA (repeated activation) was also outlined. the methodology of NAA involves the absolute method, the relative method and the mono standard (single comparator) method , which is in between the absolute and relative methods

  13. Uranium borehole logging using delayed or prompt fission neutrons

    International Nuclear Information System (INIS)

    Schulze, G.; Wuerz, H.

    1977-04-01

    The measurement of induced fission neutrons using Cf 252 and 14 MeV neutrons is a sensitive method for an in situ determination of Uranium. Applying this methods requires a unique relation between concentration of Uranium and intensity of induced fission neutrons. A discussion of parameters influencing the determination of concentration is given. A simple method is developed allowing an elemination of the geochemistry of the deposit and of the borehole configuration. Borehole probes using the methods described are of considerable help during the phase of detailed exploration of uranium ore deposits. These on-line tools allow an immediate determination of concentration. Thus avoiding the expensive and time consuming step of core drilling and subsequent chemical analysis. (orig./HP) [de

  14. Developing and investigating a pure Monte-Carlo module for transient neutron transport analysis

    International Nuclear Information System (INIS)

    Mylonakis, Antonios G.; Varvayanni, M.; Grigoriadis, D.G.E.; Catsaros, N.

    2017-01-01

    Highlights: • Development and investigation of a Monte-Carlo module for transient neutronic analysis. • A transient module developed on the open-source Monte-Carlo static code OpenMC. • Treatment of delayed neutrons is inserted. • Simulation of precursors’ decay process is performed. • Transient analysis of simplified test-cases. - Abstract: In the field of computational reactor physics, Monte-Carlo methodology is extensively used in the analysis of static problems while the transient behavior of the reactor core is mostly analyzed using deterministic algorithms. However, deterministic algorithms make use of various approximations mainly in the geometric and energetic domain that may induce inaccuracy. Therefore, Monte-Carlo methodology which generally does not require significant approximations seems to be an attractive candidate tool for the analysis of transient phenomena. One of the most important constraints towards this direction is the significant computational cost; however since nowadays the available computational resources are continuously increasing, the potential use of the Monte-Carlo methodology in the field of reactor core transient analysis seems feasible. So far, very few attempts to employ Monte-Carlo methodology to transient analysis have been reported. Even more, most of those few attempts make use of several approximations, showing the existence of an “open” research field of great interest. It is obvious that comparing to static Monte-Carlo, a straight-forward physical treatment of a transient problem requires the temporal evolution of the simulated neutrons; but this is not adequate. In order to be able to properly analyze transient reactor core phenomena, the proper simulation of delayed neutrons together with other essential extensions and modifications is necessary. This work is actually the first step towards the development of a tool that could serve as a platform for research and development on this interesting but also

  15. Charged particle induced delayed X-rays (DEX) for the analysis of intermediate and heavy elements

    Science.gov (United States)

    Pillay, A. E.; Erasmus, C. S.; Andeweg, A. H.; Sellschop, J. P. F.; Annegarn, H. J.; Dunn, J.

    1988-12-01

    The emission of K X-rays from proton-rich and metastable radionuclides, following proton activation of the stable isotopes of the elements of interest, has not been widely used as a means of analysis. The thrust of this paper proposes a nuclear technique using delayed X-rays for the analysis of low concentrations of intermediate and heavy elements. The method is similar to the delayed gamma-ray technique. Proton bombardment induces mainly (p, n) reactions whereas the delayed X-rays originate largely from e --capture and isomeric transition. Samples of rare earth and platinum group elements (PGE), in the form of compacted powders, were irradiated with an 11 MeV proton beam and delayed X-rays detected with a 100 mm 2 Ge detector. Single element spectra for a range of rare earths and PGEs are presented. Analytical conditions are demonstrated for Pd in the range 0.1-5%. Spectra from actual geological samples of a PGE ore, preconcentrated by fire-assay, and monazite are presented. All six platinum group elements are visible and interference-free in a single spectrum, a marked advance on other nuclear techniques for these elements, including PIXE and neutron activation analysis (NAA).

  16. Assay of fissionable isotopes in aqueous solution by pulsed neutron interrogation

    International Nuclear Information System (INIS)

    Campbell, P.; Gardy, E.M.; Boase, D.G.

    1978-04-01

    Non-destructive assay of uranium-235 and thorium-232 in aqueous nitric acid solutions has been accomplished by irradiation with pulses of neutrons from a 14-MeV Cockcroft-Walton neutron generator, and counting of the delayed neutrons emitted from the fissions induced. Design of the delayed neutron detector assemblies is described, together with the neutron pulse timing and counting systems. The effects of irradiation time, counting time, neutron moderation, detector design and sample geometry on the delayed neutron response from uranium-235 and 238 and thorium-232 are discussed. By using polyethylene to moderate the interrogating neutrons, solutions can be analyzed for both uranium-235 and thorium. Comparative analyses with chemical and γ-spectrometric methods show good agreement. The neutron method is rapid and is shown to be unaffected by the presence in solution of impurities such as iron, nickel, chromium, and aluminum. With the experimental equipment described, detection limits of 0.6 mg of 235 U and 9 mg of 232 Th in a sample volume of 25 mL have been achieved. Analyses of highly radioactive samples may be done easily since the measurements are not affected by the presence of large amounts of βγ radiation. Samples can be enclosed in small lead-shielded flasks during analysis to protect the analyst. The potential of the technique to on-line analysis applications is explored briefly. (author)

  17. The effect of mixed fractionation with X rays and neutrons on tumour growth delay and skin reactions in mice

    International Nuclear Information System (INIS)

    Carl, U.M.

    1987-01-01

    The authors have compared the effects of mixed fractionation schedules with X rays and neutrons on growth delay of a murine tumour and skin reactions in mice. The schedules were five daily fractions of X rays, neutrons or mixtures (NNXXX, XXXNN or NXXXN). For clamped tumours or skin all three mixed schedules had the same effect. In contrast, for unclamped tumours giving the neutrons first (NNXXX) was more effective than the other two mixed schedules. This represented a true therapeutic gain and implies that if neutrons are used clinically as only part of a course of fractionated radiotherapy, they should be given at the beginning rather than at the end of treatment. (author)

  18. Comparison of reactor RA-4 kinetics with simulations with Matlab-Simulink for one group and six groups of delayed neutrons

    International Nuclear Information System (INIS)

    Orso, J A

    2012-01-01

    The critical state of a nuclear reactor is an unstable equilibrium. The nuclear reactor can go from critical to subcritical state or can go from critical to hypercritical state. Although the evolution of the system in these cases is slow, it requires the intervention of an operator to correct deviations. For this reason an automatic control technique was designed, based on the kinetic point to a group of delayed neutrons, which corrects deviations automatically. In this paper we study the point kinetics models in a group and six groups of delayed neutrons for different values of reactivity using the simulations software MATLAB, Simulink. A comparison of two models with the reactor kinetic behavior is made (author)

  19. Easy-to-use application programs for decay heat and delayed neutron calculations on personal computers

    Energy Technology Data Exchange (ETDEWEB)

    Oyamatsu, Kazuhiro [Nagoya Univ. (Japan)

    1998-03-01

    Application programs for personal computers are developed to calculate the decay heat power and delayed neutron activity from fission products. The main programs can be used in any computers from personal computers to main frames because their sources are written in Fortran. These programs have user friendly interfaces to be used easily not only for research activities but also for educational purposes. (author)

  20. Summary Report of Consultants' Meeting on Beta-Delayed Neutron Emission Evaluation

    International Nuclear Information System (INIS)

    Abriola, Daniel; Singh, Balraj; Dillmann, Iris

    2011-12-01

    A summary is given of a Consultants' Meeting assembled to assess the viability of a new IAEA Co-ordinated Research Project (CRP) on Beta-delayed neutron emission evaluation. The current status of the field was reviewed, cases in which new measurements are needed were identified and the current theoretical models were examined. The best known cases were selected as standards and were assessed and preliminary best values of the emission probabilities were obtained. The need of such a CRP was strongly agreed. Both the technical discussions and the expected outcome of such a project are described, along with detailed recommendations for its implementation. (author)

  1. Neutron interrogation of actinides with a 17 MeV electron accelerator and first results from photon and neutron interrogation non-simultaneous measurements combination

    Energy Technology Data Exchange (ETDEWEB)

    Sari, A., E-mail: adrien.sari@cea.fr [CEA, LIST, Laboratoire Capteurs et Architectures Electroniques, 91191 Gif-sur-Yvette Cedex (France); Carrel, F.; Lainé, F. [CEA, LIST, Laboratoire Capteurs et Architectures Electroniques, 91191 Gif-sur-Yvette Cedex (France); Lyoussi, A. [CEA, DEN, 13108 Saint-Paul-Lez-Durance Cedex (France)

    2013-10-01

    In this article, we demonstrate the feasibility of neutron interrogation using the conversion target of a 17 MeV linear electron accelerator as a neutron generator. Signals from prompt neutrons, delayed neutrons, and delayed gamma-rays, emitted by both uranium and plutonium samples were analyzed. First results from photon and neutron interrogation non-simultaneous measurements combination are also reported in this paper. Feasibility of this technique is shown in the frame of the measurement of uranium enrichment. The latter was carried out by combining detection of prompt neutrons from thermal fission and delayed neutrons from photofission, and by combining delayed gamma-rays from thermal fission and delayed gamma-rays from photofission.

  2. Neutron activation analysis at the 'Instituto de Pesquisas Energeticas e Nucleares' (SP, Brazil)

    International Nuclear Information System (INIS)

    Vasconcellos, M.B.A.

    1984-01-01

    A review of the work carried out at IPEN using neutron activation analysis is made. The main characteristics of the technique and general experimental procedures applied for different samples and elements are reported. Geological samples were analysed by using activation with thermal, epithermal and delayed neutrons (for U and Th, specifically). Metallic samples were analysed for several elements in trace amounts (Ta in Nb, Hg in steel, Sn, Sb, As, Cu, Cr and Ag in a tin-lead alloy). Biological materials, such as tomatoes, animal and human viscera, food, hair, nails were also analysed for several components (Hg, Na, K, As, Au and others). (Author) [pt

  3. Measuring delayed part of the current of a self powered neutron detector and comparison with calculations

    International Nuclear Information System (INIS)

    Kophazi, J.; Czifrus, Sz.; Feher, S.; Por, G.

    2001-01-01

    The paper describes the measurement of the delayed signal of a Rh emitter Self Powered Neutron Detector (SPND) separately from other signal components originating from (n-gamma-e), (background gamma-e) and other effects. In order to separate the delayed signal, the detector was removed from the reactor core and placed to an adequately distant location during the measurement, where the radiation from the core was negligible. The experiment was carried out on the 100kW light water tank-type reactor of Technical University of Budapest and the results of the measurement were compared with the results of Monte Carlo calculations.(author)

  4. Isotopic neutron sources for neutron activation analysis

    International Nuclear Information System (INIS)

    Hoste, J.

    1988-06-01

    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

  5. Experiment Design and Analysis Guide - Neutronics & Physics

    Energy Technology Data Exchange (ETDEWEB)

    Misti A Lillo

    2014-06-01

    The purpose of this guide is to provide a consistent, standardized approach to performing neutronics/physics analysis for experiments inserted into the Advanced Test Reactor (ATR). This document provides neutronics/physics analysis guidance to support experiment design and analysis needs for experiments irradiated in the ATR. This guide addresses neutronics/physics analysis in support of experiment design, experiment safety, and experiment program objectives and goals. The intent of this guide is to provide a standardized approach for performing typical neutronics/physics analyses. Deviation from this guide is allowed provided that neutronics/physics analysis details are properly documented in an analysis report.

  6. Application of delayed X-ray spectrometry to the analysis of some rare earth elements

    International Nuclear Information System (INIS)

    Pillay, A.E.; Mboweni, R.C.M.

    1991-01-01

    The capabilities of delayed x-ray spectrometry preceded by isotope-source thermal neutron activation for the specific determination of some rare earth elements (Sm, Eu, Dy, Ho) in small powdered samples was evaluated. The feasibility study relied heavily on the low-energy sensitivity of the detector used. Detection of the delayed x-rays was achieved with a 100-mm 2 Ge detector with the ability to produce optimum photopeak-to-noise ratios. The rare earth elements were chosen on the basis of their inherent favourable nuclear properties for producing a practicable x-ray yield and on the demand for their analysis. Analytical results are presented over a range of concentrations for the elements of interest and the potential of the technique for application to their general routine analysis is discussed. Interferences from the sample matrix can be suppressed to an extent that makes the method almost independent of the matrix. This and other features make the technique a strong rival to conventional activation analysis. (author)

  7. Bioassay method for Uranium in urine by Delay Neutron counting; Metoda Bioassay Uranium dalam urin dengan pencacahan Netron Kasip

    Energy Technology Data Exchange (ETDEWEB)

    Suratman,; Purwanto,; Sukarman-Aminjoyo, [Yogyakarta Nuclear Research Centre, National Atomic Energy Agency, Yogyakarta (Indonesia)

    1996-04-15

    A bioassay method for uranium in urine by neutron counting has been studied. The aim of this research is to obtain a bioassay method for uranium in urine which is used for the determination of internal dose of radiation workers. The bioassay was applied to the artificially uranium contaminated urine. The weight of the contaminant was varied. The uranium in the urine was irradiated in the Kartini reactor core, through pneumatic system. The delayed neutron was counted by BF3 neutron counter. Recovery of the bioassay was between 69.8-88.8 %, standard deviation was less than 10 % and the minimum detection was 0.387 {mu}g.

  8. Limitations for qualitative and quantitative neutron activation analysis using reactor neutrons

    International Nuclear Information System (INIS)

    El-Abbady, W.H.; El-Tanahy, Z.H.; El-Hagg, A.A.; Hassan, A.M.

    1999-01-01

    In this work, the most important limitations for qualitative and quantitative analysis using reactor neutrons for activation are reviewed. Each limitation is discussed using different examples of activated samples. Photopeak estimation, nuclear reactions interference and neutron flux measurements are taken into consideration. Solutions for high accuracy evaluation in neutron activation analysis applications are given. (author)

  9. Benchmark experiments of effective delayed neutron fraction βeff at FCA

    International Nuclear Information System (INIS)

    Sakurai, Takeshi; Okajima, Shigeaki

    1999-01-01

    Benchmark experiments of effective delayed neutron fraction β eff were performed at Fast Critical Assembly (FCA) in the Japan Atomic Energy Research Institute. The experiments were made in three cores providing systematic change of nuclide contribution to the β eff : XIX-1 core fueled with 93% enriched uranium, XIX-2 core fueled with plutonium and uranium (23% enrichment) and XIX-3 core fueled with plutonium (92% fissile Pu). Six organizations from five countries participated in these experiments and measured the β eff by using their own methods and instruments. Target accuracy in the β eff was achieved to be better than ±3% by averaging the β eff values measured using a wide variety of experimental methods. (author)

  10. Delayed Particle Study of Neutron Rich Lithium Isotopes

    CERN Multimedia

    Marechal, F; Perrot, F

    2002-01-01

    We propose to make a systematic complete coincidence study of $\\beta$-delayed particles from the decay of neutron-rich lithium isotopes. The lithium isotopes with A=9,10,11 have proven to contain a vast information on nuclear structure and especially on the formation of halo nuclei. A mapping of the $\\beta$-strength at high energies in the daughter nucleus will make possible a detailed test of our understanding of their structure. An essential step is the comparison of $\\beta$-strength patterns in $^{11}$Li and the core nucleus $^{9}$Li, another is the full characterization of the break-up processes following the $\\beta$-decay. To enable such a measurement of the full decay process we will use a highly segmented detection system where energy and emission angles of both charged and neutral particles are detected in coincidence and with high efficiency and accuracy. We ask for a total of 30 shifts (21 shifts for $^{11}$Li, 9 shifts $^{9}$Li adding 5 shifts for setting up with stable beam) using a Ta-foil target...

  11. Neutron Multiplicity Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Frame, Katherine Chiyoko [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-06-28

    Neutron multiplicity measurements are widely used for nondestructive assay (NDA) of special nuclear material (SNM). When combined with isotopic composition information, neutron multiplicity analysis can be used to estimate the spontaneous fission rate and leakage multiplication of SNM. When combined with isotopic information, the total mass of fissile material can also be determined. This presentation provides an overview of this technique.

  12. Thermal neutron imaging in an active interrogation environment

    International Nuclear Information System (INIS)

    Vanier, P.E.; Forman, L.; Norman, D.R.

    2009-01-01

    We have developed a thermal-neutron coded-aperture imager that reveals the locations of hydrogenous materials from which thermal neutrons are being emitted. This imaging detector can be combined with an accelerator to form an active interrogation system in which fast neutrons are produced in a heavy metal target by means of excitation by high energy photons. The photo-induced neutrons can be either prompt or delayed, depending on whether neutronemitting fission products are generated. Provided that there are hydrogenous materials close to the target, some of the photo-induced neutrons slow down and emerge from the surface at thermal energies. These neutrons can be used to create images that show the location and shape of the thermalizing materials. Analysis of the temporal response of the neutron flux provides information about delayed neutrons from induced fission if there are fissionable materials in the target. The combination of imaging and time-of-flight discrimination helps to improve the signal-to-background ratio. It is also possible to interrogate the target with neutrons, for example using a D-T generator. In this case, an image can be obtained from hydrogenous material in a target without the presence of heavy metal. In addition, if fissionable material is present in the target, probing with fast neutrons can stimulate delayed neutrons from fission, and the imager can detect and locate the object of interest, using appropriate time gating. Operation of this sensitive detection equipment in the vicinity of an accelerator presents a number of challenges, because the accelerator emits electromagnetic interference as well as stray ionizing radiation, which can mask the signals of interest.

  13. On solution to the problem of reactor kinetics with delayed neutrons by Monte Carlo method

    International Nuclear Information System (INIS)

    Kyncl, Jan

    2013-07-01

    The initial value problem is addressed for the neutron transport equation and for the system of equations that describe the behaviour of emitters of delayed neutrons. Examination of the solution to this problem is based on several main assumptions concerning the behaviour of macroscopic effective cross-sections describing the reaction of the neutron with the medium, the temperature of medium and the remaining parameters of the equations. Formulation of these assumptions is adequately general and is in agreement with the properties of all known models of the physical quantities involved. Among others, the assumptions admit dependence of the macroscopic effective cross-sections and temperature on spatial coordinates and time that can be arbitrary to a great extent. The problem starts from a set of integro-differential equations. This problem is first transposed into the equivalent problem of solving a linear integral equation for neutron flux. This integral equation is solved by the method of successive iterations and its uniqueness is demonstrated. Numeric solution to the integral equation by Monte Carlo method consists in finding a functional of the exact solution. For this, a random process is set up and some random variables are proposed. Then it is demonstrated that each of these variables is an unbiased estimator of that functional. (author)

  14. Lifetime Neutron Fluence Analysis of the Ringhals Unit 1 Boiling Water Reactor

    Directory of Open Access Journals (Sweden)

    Kulesza Joel A.

    2016-01-01

    Full Text Available This paper describes a neutron fluence assessment considering the entire commercial operating history (35 cycles or ∼ 25 effective full power years of the Ringhals Unit 1 reactor pressure vessel beltline region. In this assessment, neutron (E >1.0 MeV fluence and iron atom displacement distributions were calculated on the moderator tank and reactor pressure vessel structures. To validate those calculations, five in-vessel surveillance chain dosimetry sets were evaluated as well as material samples taken from the upper core grid and wide range neutron monitor tubes to act as a form of retrospective dosimetry. During the analysis, it was recognized that delays in characterizing the retrospective dosimetry samples reduced the amount of reactions available to be counted and complicated the material composition determination. However, the comparisons between the surveillance chain dosimetry measurements (M and calculated (C results show similar and consistent results with the linear average M/C ratio of 1.13 which is in good agreement with the resultant least squares best estimate (BE/C ratios of 1.10 for both neutron (E >1.0 MeV flux and iron atom displacement rate.

  15. First delayed neutron emission measurements at ALTO with the neutron detector TETRA

    International Nuclear Information System (INIS)

    Testov, D.; Ancelin, S.; Bettane, J.; Ibrahim, F.; Kolos, K.; Mavilla, G.; Niikura, M.; Verney, D.; Wilson, J.; Kuznetsova, E.; Penionzhkevich, Yu.; Smirnov, V.; Sokol, E.

    2013-01-01

    Beta-decay properties are among the easiest and, therefore, the first ones to be measured to study new neutron-rich isotopes. Eventually, a very small number of nuclei could be sufficient to estimate their lifetime and neutron emission probability. With the new radioactive beam facilities which have been commissioned recently (or will be constructed shortly) new areas of neutron-rich isotopes will become reachable. To study beta-decay properties of such nuclei at IPN (Orsay) in the framework of collaboration with JINR (Dubna), a new experimental setup including the neutron detector of high efficiency TETRA was developed and commissioned

  16. Physical basis for prompt-neutron activation analysis

    International Nuclear Information System (INIS)

    Chrien, R.E.

    1982-01-01

    The technique called prompt ν-ray neutron activation analysis has been applied to rapid materials analysis. The radiation following the neutron radiation capture is prompt in the sense that the nuclear decay time is on the order of 10 - 15 second, and thus the technique is not strictly activation, but should be called radiation neutron capture spectroscopy or neutron capture ν-ray spectroscopy. This paper reviews the following: sources and detectors, theory of radiative capture, nonstatistical capture, giant dipole resonance, fast neutron capture, and thermal neutron capture ν-ray spectra. 14 figures

  17. Development of high flux thermal neutron generator for neutron activation analysis

    Energy Technology Data Exchange (ETDEWEB)

    Vainionpaa, Jaakko H., E-mail: hannes@adelphitech.com [Adelphi Technology, 2003 E Bayshore Rd, Redwood City, CA 94063 (United States); Chen, Allan X.; Piestrup, Melvin A.; Gary, Charles K. [Adelphi Technology, 2003 E Bayshore Rd, Redwood City, CA 94063 (United States); Jones, Glenn [G& J Jones Enterprice, 7486 Brighton Ct, Dublin, CA 94568 (United States); Pantell, Richard H. [Department of Electrical Engineering, Stanford University, Stanford, CA (United States)

    2015-05-01

    The new model DD110MB neutron generator from Adelphi Technology produces thermal (<0.5 eV) neutron flux that is normally achieved in a nuclear reactor or larger accelerator based systems. Thermal neutron fluxes of 3–5 · 10{sup 7} n/cm{sup 2}/s are measured. This flux is achieved using four ion beams arranged concentrically around a target chamber containing a compact moderator with a central sample cylinder. Fast neutron yield of ∼2 · 10{sup 10} n/s is created at the titanium surface of the target chamber. The thickness and material of the moderator is selected to maximize the thermal neutron flux at the center. The 2.5 MeV neutrons are quickly thermalized to energies below 0.5 eV and concentrated at the sample cylinder. The maximum flux of thermal neutrons at the target is achieved when approximately half of the neutrons at the sample area are thermalized. In this paper we present simulation results used to characterize performance of the neutron generator. The neutron flux can be used for neutron activation analysis (NAA) prompt gamma neutron activation analysis (PGNAA) for determining the concentrations of elements in many materials. Another envisioned use of the generator is production of radioactive isotopes. DD110MB is small enough for modest-sized laboratories and universities. Compared to nuclear reactors the DD110MB produces comparable thermal flux but provides reduced administrative and safety requirements and it can be run in pulsed mode, which is beneficial in many neutron activation techniques.

  18. Quantitative phase analysis by neutron diffraction

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Chang Hee; Song, Su Ho; Lee, Jin Ho; Shim, Hae Seop [Korea Atomic Energy Research Institute, Taejon (Korea)

    1999-06-01

    This study is to apply quantitative phase analysis (QPA) by neutron diffraction to the round robin samples provided by the International Union of Crystallography(IUCr). We measured neutron diffraction patterns for mixed samples which have several different weight percentages and their unique characteristic features. Neutron diffraction method has been known to be superior to its complementary methods such as X-ray or Synchrotron, but it is still accepted as highly reliable under limited conditions or samples. Neutron diffraction has strong capability especially on oxides due to its scattering cross-section of the oxygen and it can become a more strong tool for analysis on the industrial materials with this quantitative phase analysis techniques. By doing this study, we hope not only to do one of instrument performance tests on our HRPD but also to improve our ability on the analysis of neutron diffraction data by comparing our QPA results with others from any advanced reactor facilities. 14 refs., 4 figs., 6 tabs. (Author)

  19. Neutron activation analysis at the Californium User Facility for Neutron Science

    International Nuclear Information System (INIS)

    Martin, R.C.; Smith, E.H.; Glasgow, D.C.; Jerde, E.A.; Marsh, D.L.; Zhao, L.

    1997-12-01

    The Californium User Facility (CUF) for Neutron Science has been established to provide 252 Cf-based neutron irradiation services and research capabilities including neutron activation analysis (NAA). A major advantage of the CUF is its accessibility and controlled experimental conditions compared with those of a reactor environment The CUF maintains the world's largest inventory of compact 252 Cf neutron sources. Neutron source intensities of ≤ 10 11 neutrons/s are available for irradiations within a contamination-free hot cell, capable of providing thermal and fast neutron fluxes exceeding 10 8 cm -2 s -1 at the sample. Total flux of ≥10 9 cm -2 s -1 is feasible for large-volume irradiation rabbits within the 252 Cf storage pool. Neutron and gamma transport calculations have been performed using the Monte Carlo transport code MCNP to estimate irradiation fluxes available for sample activation within the hot cell and storage pool and to design and optimize a prompt gamma NAA (PGNAA) configuration for large sample volumes. Confirmatory NAA irradiations have been performed within the pool. Gamma spectroscopy capabilities including PGNAA are being established within the CUF for sample analysis

  20. Forensic neutron activation analysis

    International Nuclear Information System (INIS)

    Kishi, T.

    1987-01-01

    The progress of forensic neutron activation analysis (FNAA) in Japan is described. FNAA began in 1965 and during the past 20 years many cases have been handled; these include determination of toxic materials, comparison examination of physical evidences (e.g., paints, metal fragments, plastics and inks) and drug sample differentiation. Neutron activation analysis is applied routinely to the scientific criminal investigation as one of multielement analytical techniques. This paper also discusses these routine works. (author) 14 refs

  1. The structure of the Gamow-Teller giant resonance and consequences for beta-delayed neutron spectra and element synthesis

    International Nuclear Information System (INIS)

    Klapdor, H.V.

    1976-01-01

    Recent results in β-delayed neutron emission are interpreted by structure of the Gamow-Teller giant resonance not included in the 'gross-theory' of β-decay. Inclusion of this structure of the β-decay function is important for calculations of β-decay production rates for heavy nuclides by astrophysical processes and thermonuclear explosions. (Auth.)

  2. Neutron activation analysis of core and drill cutting samples from geothermal well drilling

    International Nuclear Information System (INIS)

    Miller, G.E.

    1986-01-01

    Samples of sandstones and shales were analysed by instrumental neutron activation analysis for a total of 30 elements. Three irradiation and five counting periods were employed. Solutions and National Bureau of Standards Reference Materials were used for comparison. The samples were obtained from drill cuttings (with a few core samples) from drillings in the Salton Sea geothermal field of California. These determinations form part of a major study to establish elemental variation as a function of mineral variation as depth and temperature in the well vary. The overall goal is to examine mineral alteration and/or element migration under typical geothermal conditions. The techniques involve typical compromises between maximizing precision for individual element determinations and the amount of time and effort that can be expended, as it is desired to examine large numbers of samples. With the limitations imposed by the reactor flux available at the U.C.Irvine TRIGA reactor, the detectors available, and time factors, most precisions are acceptable for geological comparison purposes. Some additional measurements were made by delayed-neutron counting methods to compare with uranium determinations made by conventional instrumental neutron activation analysis methods. (author)

  3. Neutron activation analysis with pulsed 14 MeV neutrons for the characterization of heterogeneous radioactive wastes; Neutronenaktivierungsanalyse mit gepulsten 14 MeV Neutronen zur Charakterisierung heterogener radioaktiver Abfaelle

    Energy Technology Data Exchange (ETDEWEB)

    Mildenberger, Frank

    2017-07-01

    For the transport, interim storage and disposal of radioactive waste, it is assumed to have knowledge of the radioactive and non-radioactive inventory. In order to determine the radioactive inventory destructive (e.g. α-, β-, γ-measurements according to wet chemical sample preparation) and non-destructive (e.g. γ-scanning and neutron measurements) measurement methods are used. For the characterization of non-radioactive substances a prototype for the assay of small-volume (50 L) samples was constructed and parameterized using the neutron activation analysis (NAA) with a pulsed 14 MeV neutron source. Subsequently, the non-destructive analytical method called MEDINA (Multi Element Detection Based on Instrumental Neutron Activation) for 200 l waste drums was developed in a cooperation between RWTH Aachen University and Forschungszentrum Juelich GmbH. The aim of this thesis is to investigate and characterize heterogeneous mixed samples regarding their material composition as well as their inhomogeneous distribution. For this purposes, studies were carried out on 200 l steel drums with heterogeneous matrices using the NAA in the MEDINA facility. The samples are composed out of a mixture of concrete and polyethylene (PE) bodies. Due to its high hydrogen content, the PE can have a strong influence on the neutron moderation and neutron absorption and can thereby occur as a possible disturbance variable in the characterization of the non-radioactive inventory. For these studies a pulsed 14 MeV neutron source is used to record the prompt and delayed γ-rays between the neutron pulses, separately. Thus, the performance of the MEDINA method relating to strongly moderating mixed matrices and their characterization is studied. In order to optimize the measurement of delayed γ-rays without any appreciable interference of prompt γ-rays, the decay of thermal neutrons was studied and the thermal neutron die-away time was determined. It ranges between 2 and 5 ms according to

  4. Neutron Activation Analysis with k0-standardisation

    International Nuclear Information System (INIS)

    Pomme, S.

    2001-01-01

    SCK-CEN's programme on Neutron Activation Analysis with k 0 -standardisation concentrates on the improvement of the standardisation method and the characterisation of the neutron field as well as on the improvement of the statistical control on neutron activation analysis. Main achievements in 2000 are reported

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

    International Nuclear Information System (INIS)

    Zhang Dalin; Qiu Suizheng; Su Guanghui; Liu Changliang

    2008-01-01

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

  6. Fundamentals of 3-D Neutron Kinetics and Current Status

    International Nuclear Information System (INIS)

    Aragones, J.M.

    2008-01-01

    This lecture includes the following topics: 1) A summary of the cell and lattice calculations used to generate the neutron reaction data for neutron kinetics, including the spectral and burnup calculations of LWR cells and fuel assembly lattices, and the main nodal kinetics parameters: mean neutron generation time and delayed neutron fraction; 2) the features of the advanced nodal methods for 3-D LWR core physics, including the treatment of partially inserted control rods, fuel assembly grids, fuel burnup and xenon and samarium transients, and excore detector responses, that are essential for core surveillance, axial offset control and operating transient analysis; 3) the advanced nodal methods for 3-D LWR core neutron kinetics (best estimate safety analysis, real-time simulation); and 4) example applications to 3-D neutron kinetics problems in transient analysis of PWR cores, including model, benchmark and operational transients without, or with simple, thermal-hydraulics feedback.

  7. Prompt gamma neutron activation analysis

    International Nuclear Information System (INIS)

    Goswami, A.

    2003-01-01

    Prompt gamma neutron activation analysis (PGNAA) is a technique for the analysis of elements present in solid, liquid and gaseous samples by measuring the capture gamma rays emitted from the sample during neutron irradiation. The technique is complementary to conventional neutron activation analysis (NAA) as it can be used in number of cases where NAA fails. Though the technique was first used in sixties, the advantage of the technique was first highlighted by Lindstrom and Anderson. PGNAA is increasingly being used as a rapid, instrumental, nondestructive and multielement analysis technique. A monograph and several excellent reviews on this topic have appeared recently. In this review, an attempt has been made to bring out the essential aspects of the technique, experimental arrangement and instrumentation involved, and areas of application. Some of the results will also be presented

  8. Optimization of combined delayed neutron and differential die-away prompt neutron signal detection for characterization of spent nuclear fuel assemblies

    International Nuclear Information System (INIS)

    Blanc, Pauline; Tobin, Stephen J.; Croft, Stephen; Menlove, Howard O.; Swinhoe, M.; Lee, T.

    2010-01-01

    The Next Generation Safeguards Initiative (NGSI) of the U.S. Department of Energy (DOE) has funded multiple laboratories and universities to develop a means to accurately quantify the Plutonium (Pu) mass in spent nuclear fuel assemblies and ways to also detect potential diversion of fuel pins. Delayed Neutron (DN) counting provides a signature somewhat more sensitive to 235 U than Pu while Differential Die-Away (DDA) is complementary in that it has greater sensitivity to Pu. The two methods can, with care, be combined into a single instrument which also provides passive neutron information. Individually the techniques cannot robustly quantify the Pu content but coupled together the information content in the signatures enables Pu quantification separate to the total fissile content. The challenge of merging DN and DDA, prompt neutron (PN) signal, capabilities in the same design is the focus of this paper. Other possibilities also suggest themselves, such as a direct measurement of the reactivity (multiplication) by either the boost in signal obtained during the active interrogation itself or by the extension of the die-away profile. In an early study, conceptual designs have been modeled using a neutron detector comprising fission chambers or 3He proportional counters and a ∼14 MeV neutron Deuterium-Tritium (DT) generator as the interrogation source. Modeling was performed using the radiation transport code Monte Carlo N-Particles eXtended (MCNPX). Building on this foundation, the present paper quantifies the capability of a new design using an array of 3 He detectors together with fission chambers to optimize both DN and PN detections and active characterization, respectively. This new design was created in order to minimize fission in 238 U (a nuisance DN emitter), to use a realistic neutron generator, to reduce the cost and to achieve near spatial interrogation and detection of the DN and PN, important for detection of diversion, all within the constraints of

  9. Measurement of the response time of the delay window for the neutron converter of the SPIRAL2 project

    Energy Technology Data Exchange (ETDEWEB)

    Acosta, G. [INFN Laboratori Nazionali di Legnaro, 35020 Legnaro (Italy); Andre, T. [GANIL, Caen (France); Bermudez, J. [INFN Laboratori Nazionali di Legnaro, 35020 Legnaro (Italy); Universidad Simon Bolivar, Caracas (Venezuela, Bolivarian Republic of); Blinov, M.F. [Budker Institute of Nuclear Physics, 630090 Novosibirsk (Russian Federation); Jamet, C. [GANIL, Caen (France); Logatchev, P.V.; Semenov, Y.I.; Starostenko, A.A. [Budker Institute of Nuclear Physics, 630090 Novosibirsk (Russian Federation); Tecchio, L.B., E-mail: tecchio@lnl.infn.it [INFN Laboratori Nazionali di Legnaro, 35020 Legnaro (Italy); Tsyganov, A.S. [Budker Institute of Nuclear Physics, 630090 Novosibirsk (Russian Federation); Udup, E. [INFN Laboratori Nazionali di Legnaro, 35020 Legnaro (Italy); Horia Hulubei National Institute of Physics and Engineering, Bucharest (Romania); University Polytechnic of Bucharest (Romania); Vasquez, J. [INFN Laboratori Nazionali di Legnaro, 35020 Legnaro (Italy); Universidad Simon Bolivar, Caracas (Venezuela, Bolivarian Republic of)

    2014-09-11

    Research and development of a safety system for the SPIRAL2 facility has been conceived to protect the UCx target from a possible interaction with the 200 kW deuteron beam. The system called “delay window” (DW) is designed as an integral part of the neutron converter module and is located in between the neutron converter and the fission target. The device has been designed as a barrier, located directly behind the neutron converter on the axis of the deuteron beam, with the purpose of “delaying” the eventual interaction of the deuteron beam with the UCx target in case of a failure of the neutron converter. The “delay” must be long enough to allow the interlock to react and safely stop the beam operation, before the beam will reach the UCx target. The working concept of the DW is based on the principle of the electrical fuse. Electrically insulated wires placed on the surface of a Tantalum disk assure a so called “free contact”, normally closed to an electronic circuit located on the HV platform, far from the radioactive environment. The melting temperature of the wires is much less than Tantalum. Once the beam is impinging on the disk, one or more wires are melted and the “free contact” is open. A solid state relay is changing its state and a signal is sent to the interlock device. A prototype of the DW has been constructed and tested with an electron beam of power density equivalent to the SPIRAL2 beam. The measured “delay” is 682.5 ms (σ=116 ms), that is rather long in comparison to the intrinsic delays introduced by the detectors itself (2 ms) and by the associated electronic devices (120 ns). The experimental results confirm that, in the case of a failure of the neutron converter, the DW as conceived is enable to withstand the beam power for a period of time sufficiently long to safely shut down the SPIRAL2 accelerator.

  10. Intercomparison of delayed neutron summation calculations among JEF2.2, ENDF/B-VI and JNDC-V2

    Energy Technology Data Exchange (ETDEWEB)

    Sagisaka, Mitsuyuki [Nagoya Univ. (Japan); Oyamatsu, K.; Kukita, Y.

    1998-03-01

    We perform intercomparison of delayed neutron activities calculated with JEF2.2, ENDF/B-VI and JNDC-V2 with a simple new method. Significant differences are found at t < 20 (s) for major fissioning systems. The differences are found to stem from fission yields or decay data of several nuclides. The list of these nuclides are also given for the future experimental determination of these nuclear data. (author)

  11. ORION, a multipurpose detector for neutrons. Some new developments

    International Nuclear Information System (INIS)

    Perier, Y.; Lienard, E.; Lott, B.; Galin, J.; Morjean, M.; Peghaire, A.; Quednau, B.; El Masri, Y.; Keutgen, Th.; Tilquin, I.

    1996-01-01

    Different properties of the four-pi neutron detector ORION have been tested: its efficiency in both modes, fast and delayed, its time resolution and position sensitivity. For the later test, the impact of the neutron beam onto the detector was varied by sliding it, perpendicular to the beam direction. All the presented data are tentative with the analysis still in progress. (K.A.)

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

    Institute of Scientific and Technical Information of China (English)

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

    2008-01-01

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

  13. The determination by irradiation with a pulsed neutron generator and delayed neutron counting of the amount of fissile material present in a sample; Determination de la quantite de matiere fissile presente dans un echantillon par irradiation au moyen d'une source pulsee de neutrons et comptage des neutrons retardes

    Energy Technology Data Exchange (ETDEWEB)

    Beliard, L; Janot, P [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1967-07-01

    A preliminary study was conducted to determine the amount of fissile material present in a sample. The method used consisted in irradiating the sample by means of a pulsed neutron generator and delayed neutron counting. Results show the validity of this method provided some experimental precautions are taken. Checking on the residual proportion of fissile material in leached hulls seems possible. (authors) [French] Ce rapport rend compte d'une etude preliminaire effectuee en vue de determiner la quantite de matiere fissile presente dans un echantillon. La methode utilisee consiste a irradier l'echantillon considere au moyen d'une source puisee de neutrons et a compter les neutrons retardes produits. Les resultats obtenus permettent de conclure a la validite de la methode moyennant certaines precautions. Un controle de la teneur residuelle en matiere fissile des gaines apres traitement semble possible. (auteurs)

  14. Lifetime measurement of prompt neutrons using the neutronic noise analysis

    International Nuclear Information System (INIS)

    Ortiz Servin, J.J.

    1992-01-01

    The purpose of this work is to estimate the life of the prompt neutrons, i, of a nuclear reactor utilizing the neutron noise analysis. This technique carry to development of mathematical model that is valid for lower powers reactor. The equation resulting convey to the observation about power spectrum behaviour respect to the frecquency. In this case, the reactor in study is the Triga Mark III of Nuclear Center of Mexico that it was provided of fission chambers for register the neutron fluxes. These fluxes was digitized and storage in computer disc as signals dependents of time, for later apply the Fourier Transformation and obtain the spectras. The spectras measured to different reactor powers were adjusted to the development equation before, using the method of square minimum and so estimate the parameter i. The analysis of results throw a value of 22.73 +/- 0.92 μs. On the other hand, the calculate value to the resolve the kinetic equation of reactor defer in lower than 4 % about the estimate. Of this, it concludes that the model utilized is trusty with a good mistake margin, moreover of that the technique of Neutron Noise analysis demonstrate be competitive (Author)

  15. Measurement of total body chlorine by prompt gamma in vivo neutron activation analysis

    International Nuclear Information System (INIS)

    Beddoe, A.H.; Streat, S.J.; Hill, G.L.

    1987-01-01

    A method of measuring total body chlorine (TBCl) by prompt gamma in vivo neutron activation analysis is described depending on the same NaI(Tl) spectra used for determinations of total body nitrogen. Ratios of chlorine to hydrogen are derived and TBCl determined using a model of body composition depending on measured body weight, total body water (by tritium dilution) and protein (6.25 x nitrogen) as well as estimated body minerals and glycogen. The precision of the method based on scanning an anthropomorphic phantom is approximately 9% (SD), for a patient dose equivalent of less than 0.30 mSv. Spectra collected from 67 normal volunteers (32 male, 35 female) yielded mean values of TBCl of 72 +- 19 (SD) g in males and 53.6 +- 15 g in females, in broad agreement with values reported by workers using delayed gamma methods. Results are presented for two human cadavers analysed by neutron activation and conventional chemical analysis; the ratios of TBCl (neutron activation) to TBCl (chemical) were 0.980 +- 0.028 (SEM) and 0.91 +- 0.09. It is suggested that an improvement in precision will be achieved by increasing the scanning time (thereby increasing the radiation dose equivalent) and by adding two more detectors. (author)

  16. Proceedings of national seminar neutron activation analysis

    International Nuclear Information System (INIS)

    Agus Taftazani; Muhayatun Santoso; Budi Haryanto; Khatarina Oginawati

    2010-11-01

    Proceedings of national seminar neutron activation analysis in 2010 with the theme of the Role of Nuclear Analytical Techniques in the Field of Environment, Health and Industry. The seminar was organized by Indonesians Neutron Activation Analysis and BATAN Forum. These proceedings contain the result of environmental research in BATAN, universities and institutions associated with the application on neutron activation analysis technique. The purpose of these proceedings was as a useful source of information to spur research and development of activation analysis applications in various fields for the Indonesian welfare. There are 40 articles. (PPIKSN).

  17. Neutron activation probe for measuring the presence of uranium in ore bodies

    International Nuclear Information System (INIS)

    Goldstein, N.P.; Smith, R.C.

    1979-01-01

    A neutron activation proble comprises a pulsed neutron source in series with a plurality of delayed neutron detectors for measuring radioactivity in a well borehole together with a NaI (Tl) counter for measuring the high energy 2.62 MeV gamma line from thorium. The neutron source emits neutrons which produce fission in uranium and thorium in the ore body and the delayed neutron detectors measure the delayed neutrons produced from such fission while the NaI (Tl) counter measures the 2.62 MeV gamma line from the undisturbed thorium in the ore body. The signal from the NaI (Tl) counter is processed and subtracted from the signal from the delayed neutron detectors with the result being indicative of the amount of uranium present in the ore body

  18. Neutron activation analysis of high-purity iron in comparison with chemical analysis

    International Nuclear Information System (INIS)

    Kinomura, Atsushi; Horino, Yuji; Takaki, Seiichi; Abiko, Kenji

    2000-01-01

    Neutron activation analysis of iron samples of three different purity levels has been performed and compared with chemical analysis for 30 metallic and metalloid impurity elements. The concentration of As, Cl, Cu, Sb and V detected by neutron activation analysis was mostly in agreement with that obtained by chemical analysis. The sensitivity limits of neutron activation analysis of three kinds of iron samples were calculated and found to be reasonable compared with measured values or detection limits of chemical analysis; however, most of them were above the detection limits of chemical analysis. Graphite-shielded irradiation to suppress fast neutron reactions was effective for Mn analysis without decreasing sensitivity to the other impurity elements. (author)

  19. Fast neutron activation analysis in metallurgy

    International Nuclear Information System (INIS)

    Sterlinski, S.

    1981-01-01

    Article discusses the usage of a 14 MeV neutron generator for producing fast neutrons of different energies and intensities. A complete instrumental set-up for the neutron activation analysis (NAA) is given. In metallurgy the device is mainly used in the determination of oxygen and silicon in steel and non-ferrous metal, including different alloys

  20. Study of the delayed neutron emission through the time-of-flight method. Application to 49K, 50K and 51K

    International Nuclear Information System (INIS)

    Rachidi, J.

    1983-04-01

    This work is dedicated to the study of the emission of delayed neutrons observed in the decay of 49 K, 50 K and 51 K. Spectroscopic data are non-existent for these 3 isotopes, so we have had to design a specific detection system based on a large-surface scintillation counter. A series of n-γ coincidence measurement has allowed us to determine the energy levels of the non-bound states of 49 Ca, 50 Ca and 51 Ca and to establish the nature of the beta transitions (K → Ca). We have measured the energy of the delayed neutrons through the time-of-flight method. Our results are consistent with the model of the p-n states based on the Bansac-French's works. This model shows that the non-bound states of the calcium isotopes discovered in the experiment are represented by simple configurations of the (sd) -1 (fp) n type. (A.C.)

  1. New generation non-stationary portable neutron generators for biophysical applications of Neutron Activation Analysis.

    Science.gov (United States)

    Marchese, N; Cannuli, A; Caccamo, M T; Pace, C

    2017-01-01

    Neutron sources are increasingly employed in a wide range of research fields. For some specific purposes an alternative to existing large-scale neutron scattering facilities, can be offered by the new generation of portable neutron devices. This review reports an overview for such recently available neutron generators mainly addressed to biophysics applications with specific reference to portable non-stationary neutron generators applied in Neutron Activation Analysis (NAA). The review reports a description of a typical portable neutron generator set-up addressed to biophysics applications. New generation portable neutron devices, for some specific applications, can constitute an alternative to existing large-scale neutron scattering facilities. Deuterium-Deuterium pulsed neutron sources able to generate 2.5MeV neutrons, with a neutron yield of 1.0×10 6 n/s, a pulse rate of 250Hz to 20kHz and a duty factor varying from 5% to 100%, when combined with solid-state photon detectors, show that this kind of compact devices allow rapid and user-friendly elemental analysis. "This article is part of a Special Issue entitled "Science for Life" Guest Editor: Dr. Austen Angell, Dr. Salvatore Magazù and Dr. Federica Migliardo". Copyright © 2016 Elsevier B.V. All rights reserved.

  2. Delayed effects of neutron irradiation on central nervous system microvasculature in the rat

    International Nuclear Information System (INIS)

    Goodman, J.H.; McGregor, J.M.; Clendenon, N.R.; Gordon, W.A.; Yates, A.J.; Gahbauer, R.A.; Barth, R.F.; Fairchild, R.G.

    1988-01-01

    Pathologic examination of a series of 14 patients with malignant gliomas treated with BNCT showed well demarcated zones of radiation damage characterized by coagulation necrosis. Beam attenuation was correlated with edema, loss of parenchymal elements, demyelination, leukocytosis, and peripheral gliosis. Vascular disturbances consisted of endothelial swelling, medial and adventitial proliferation, fibrin impregnation, frequent thrombosis, and perivascular inflammation. Radiation changes appeared to be acute and delayed. The outcome of the patients in this series was not significantly different from the natural course of the disease, even though two of the patients had no residual tumor detected at the time of autopsy. The intensity of the vascular changes raised a suspicion that boron may have sequestered in vessel walls, resulting in selectively high doses of radiation to these structures (Asbury et al., 1972), or that there may have been high blood concentrations of boron at the time of treatment. The potential limiting effects of a vascular ischemic reaction in Boron Neutron Capture Therapy (BNCT) prompted the following study to investigate the delayed response of microvascular structures in a rat model currently being used for pre-clinical investigations. 8 refs., 3 figs., 1 tab

  3. Neutron activation spectrometry and neutron activation analysis in analytical geochemistry

    International Nuclear Information System (INIS)

    Dulski, P.; Moeller, P.

    1975-07-01

    The present report is to show the geochemists who are interested in neutron activation spectrometry (NAS) and neutron activation analysis (NAA) which analytical possibilities these methods offer him. As a review of these analytical possibilities, a lieterature compolation is given which is subdivided into two groups: 1) rock (basic, intermediary, acid, sediments, soils and nuds, diverse minerals, tectites, meteorites and lunar material). 2) ore (Al, Au, Be, Cr, Cu, Mn, Mo, Fe, Pb, Pt, Sn, Ti, W, Zn, Zr, U and phosphate ore, polymetallic ores, fluorite, monazite and diverse ores). The applied methods as well as the determinable elements in the given materials can be got from the tables. On the whole, the literature evaluation carried out makes it clear that neutron activation spectrometry is a very useful multi-element method for the analysis of rocks. The analysis of ores, however, is subjected to great limitations. As rock analysis is very frequently of importance in prospecting for ore deposits, the NAS proves to be extremely useful for this very field of application. (orig./LH) [de

  4. Neutron activation analysis

    International Nuclear Information System (INIS)

    Taure, I.; Riekstina, D.; Veveris, O.

    2004-01-01

    Neutron activation analysis (NAA) in Latvia began to develop after 1961 when nuclear reactor in Salaspils started to work. It provided a powerful neuron source, which is necessary for this analytical method. In 1963 at Institute of Physics of the Latvian Academy of Sciences the Laboratory of Neutron Activation Analysis was formed. At the first stage of development the main tasks were of theoretical and technical aspects of NAA. Later the NAA was used to solve problems in technology, biology, and medicine. In the beginning of the 80-ties more attention was focussed to the use of NAA in the environmental research. Environmental problems stayed the main task till the closing the nuclear reactor in Salaspils in 1998 that ceased the main the existence of the laboratory and of NAA, this significant and powerful analytical method in Latvia and Baltic in general. (authors)

  5. Summary Report of 1st Research Coordination Meeting on Development of Reference Database for Beta-delayed Neutron Emission

    International Nuclear Information System (INIS)

    Dillmann, Iris; Dimitriou, Paraskevi; Singh, Balraj

    2014-03-01

    A summary is given of the 1st Research Coordination Meeting of the new IAEA Coordinated Research Project (CRP) on Development of a Reference Database for Beta-delayed neutron emission data. Participants presented their work, reviewed the current status of the field with regards to individual precursors and aggregate data, and discussed the scope of the work to be undertaken. A list of priorities and task assignments was produced. (author)

  6. Analysis of some Egyptian cosmetic samples by fast neutron activation analysis

    International Nuclear Information System (INIS)

    Medhat, M.E.; Ali, M.A.; Hassan, M.F.

    2001-01-01

    A description of D-T neutron generator (NG) is presented. This generator can be used for fast neutron activation analysis applied to determine some selected elements, especially light elements, in different materials. The concentrations of the elements Na, Mg, Al, Si, K, Cl, Ca and Fe were determined in two domestic brands of face powder by using 14 MeV neutron activation analysis

  7. IPR-RI TRIGA MARK I reactor and the neutron activation analysis at CDTN/CNEN

    International Nuclear Information System (INIS)

    Menezes, Maria Angela de B.C.; Kastner, Geraldo F.; Amaral, Angela M.; Souza, Wagner de; Maretti, Fausto Junior; Leal, Alexandre S.

    2008-01-01

    The IPR-R1 TRIGA Mark I research reactor started up in 1960. It is located at Centro de Desenvolvimento da Tecnologia Nuclear (Nuclear Technology Development Centre) / Comissao Nacional de Energia Nuclear (Brazilian Commission for Nuclear Energy), CDTN/CNEN. Join to the reactor, the Laboratory for Neutron Activation Analysis has been developing its activities since 1960. The activities of the Laboratory comprise the delayed fission neutron activation analysis, instrumental (comparative and parametric methods) and radiochemical / chemical methods. These methods are responsible for relevant percentage of CDTN's analysis demand, meeting the clients' analytical needs and researches developed by the Laboratory, by CDTN and by other institutions. Over the years the work has been linked to the goals of the country and the institutions. Nowadays several elements - Ag, Al, Au, As, Ba, Br, Ca, Cd, Ce, Cl, Co, Cr, Cs, Cu, Dy, Eu, Fe, Ga, Hf, Hg, Ho, K, La, Mg, Mn, Mo, Na, Nd, Rb, Sb, Sc, Se, Sm, Sr, Ta, Tb, Th, Ti, U, V, W, Yb, Zn and Zr - are determined in several matrices and range of concentrations. In Brazil, CDTN is the only Institute that fully masters the instrumental neutron activation analysis k0-method determining short, medium and long half-life radionuclides using its own nuclear reactor. The good performance of the reactor is pointed out in a table with experimental and certified values for Certified Reference Materials. (authors)

  8. Determination of uranium in urine: Comparison of ICP-mass spectrometry and delayed neutron assay

    International Nuclear Information System (INIS)

    Gladney, E.S.; Moss, W.D.; Gautier, M.A.; Bell, M.G.

    1986-01-01

    Los Alamos analytical chemistry group acquired a VG-Plasmaquad ICP-MS in January, 1986 and have applied the technique to a variety of environmental and bioassay analytical problems. The authors report on their experience with the determination of uranium and its isotopics in urine and compare this new method with their current uranium procedure, delayed neutron activation analysis (DNA) at the Los Alamos Omega West Reactor. The authors have utilized DNA for bioassay samples since 1978. They currently analyze approximately 2000 urine samples annually. Quantitative data on uranium concentrations are obtained by concurrent measurement of urine standards of known uranium content and isotopic ratio. Detection of 0.03 μg of normal U in a 25 mL sample (1 μg/L) can be achieved by the DNA system. The NRC has proposed new urine bioassay standards that might require at least an order of magnitude reduction in the authors current DNA detection limits. The authors have fully optimized the reactor, and can forsee no instrumental improvement. They may be forced to resort to time-consuming chemical separations at greatly increased costs. DNA is a mature technology with little prospect for radical change. ICPMS is still in its infancy, and there are several ideas for obtaining drastic improvements in detection limits. Costs and time per analysis for both methods are equal

  9. Analysis of coal by neutron activation

    International Nuclear Information System (INIS)

    Burtner, D.R.

    1983-01-01

    The development of a thermal-neutron activation analysis procedure for determining elemental concentrations in whole coal samples, and the goal of combining this technique with other nuclear methods for determining a total mass balance in these and similar complex materials, is described. Problems of applying a fast-neutron activation analysis method for nitrogen are discussed, as well as an efficient procedure for drying and packaging coal samples. A thermal-neutron activation analysis (TNAA) procedure was developed for determining up to 27 elements in coal samples from the US, China, Nigeria, and Brazil. The comparator form of TNAA was applied, using a unique multielement standard, which contained 48 elements. The difference in net photopeak counts between sample and standard, due to γ-ray attenuation, was reduced by preparing this standard in an organic matrix, which simulates the composition and physical structure of the coal material. The simultaneous irradiation of several aliquots of this standard enabled high precision and accuracy to be attained. An accurate value for oxygen, determined by fast-neutron activation analysis, is used to correct for this effect in the nitrogen determination method

  10. Neutron density fluctuations in point reactor systems with dichotomic reactivity noise

    International Nuclear Information System (INIS)

    Sako, Okitsugu

    1984-01-01

    The exactly solvable stochastic point reactor model systems are analyzed through the stochastic Liouville equation. Three kinds of model systems are treated: (1) linear system without delayed neutrons, (2) linear system with one-group of delayed neutrons, and (3) nonlinear system with direct power feedback. The exact expressions for the fluctuations of neutron density, such as the moments, autocorrelation function and power spectral density, are derived in the case where the colored reactivity noise is described by the dichotomic, or two state, Markov process with arbitrary correlation time and intensity, and the effects of the finite correlation time and intensity of the noise on the neutron density fluctuations are investigated. The influence of presence of delayed neutrons and the effect of nonlinearity of system on the neutron density fluctuations are also elucidated. When the reactivity correlation time is very short, the correlation time has almost no effect on the power spectral density, and the relative fluctuation of neutron density in the stationary state is not affected very much by the presence of delayed neutrons and also by the nonlinearity of system. On the other hand, if the reactivity correlation time is very long, the effect of the reactivity noise on the power spectral density appears at very low frequency, and the presence of delayed neutrons has an effect of reducing the neutron density fluctuations. (author)

  11. Analysis of Some Egyptian Cosmetic Samples by Fast Neutron Activation Analysis

    CERN Document Server

    Medhat, M E; Fayez-Hassan, M

    2001-01-01

    A description of D-T neutron generator (NG) is presented. This generator can be used for fast neutron activation analysis applied to determine some selected elements, especially light elements, in different materials. In our work, the concentration of the elements Na, Mg, Al, Si, K, Cl, Ca and Fe, were determined in two domestic brands of face powder by using 14 MeV neutron activation analysis.

  12. SVIP-N 1.0: An integrated visualization platform for neutronics analysis

    International Nuclear Information System (INIS)

    Luo Yuetong; Long Pengcheng; Wu Guoyong; Zeng Qin; Hu Liqin; Zou Jun

    2010-01-01

    Post-processing is an important part of neutronics analysis, and SVIP-N 1.0 (scientific visualization integrated platform for neutronics analysis) is designed to ease post-processing of neutronics analysis through visualization technologies. Main capabilities of SVIP-N 1.0 include: (1) ability of manage neutronics analysis result; (2) ability to preprocess neutronics analysis result; (3) ability to visualization neutronics analysis result data in different way. The paper describes the system architecture and main features of SVIP-N, some advanced visualization used in SVIP-N 1.0 and some preliminary applications, such as ITER.

  13. Neutronic analysis of fusion tokamak devices by PHITS

    International Nuclear Information System (INIS)

    Sukegawa, Atsuhiko M.; Takiyoshi, Kouji; Amano, Toshio; Kawasaki, Hiromitsu; Okuno, Koichi

    2011-01-01

    A complete 3D neutronic analysis by PHITS (Particle and Heavy Ion Transport code System) has been performed for fusion tokamak devices such as JT-60U device and JT-60 Superconducting tokamak device (JT-60 Super Advanced). The mono-energetic neutrons (E n =2.45 MeV) of the DD fusion devices are used for the neutron source in the analysis. The visual neutron flux distribution for the estimation of the port streaming and the dose rate around the fusion tokamak devices has been calculated by the PHITS. The PHITS analysis makes it clear that the effect of the port streaming of superconducting fusion tokamak device with the cryostat is crucial and the calculated neutron spectrum results by PHITS agree with the MCNP-4C2 results. (author)

  14. Elemental analysis of water and soil environmental samples in Tabuk area by neutron capture gamma-ray spectroscopy techniques

    International Nuclear Information System (INIS)

    Al-Aseery, Sh.M.; Alamoudi, Z.; Hassan, A.M.

    2006-01-01

    The prompt and delayed gamma-rays due to neutron capture in the nuclei of the constituent elements of three soil samples and one drinking water sample have been measured. The 252 Cf and 226 Ra/Be isotopic neutron sources are used for neutron irradiation. Also, the hyper pure germanium detection system is used. The soil samples were from Astra, Tadco and El-Gammaz farms, while the water sample was taken from Tabuk city. In case of prompt gamma-ray analysis, a total of 16 elements were identified and the concentration percentage values by weight were calculated for: C, Na, Mg, Al, Si, S, Cl,, Ca, Ti, Cr, Mn, Fe, Co, Zn, Sr ad Pb elements. A comparative study between the results obtained in this work and the results obtained by ICP-MS and EDX-Ray techniques for the same samples is given

  15. The solution of the neutron point kinetics equation with stochastic extension: an analysis of two moments

    Energy Technology Data Exchange (ETDEWEB)

    Silva, Milena Wollmann da; Vilhena, Marco Tullio M.B.; Bodmann, Bardo Ernst J.; Vasques, Richard, E-mail: milena.wollmann@ufrgs.br, E-mail: vilhena@mat.ufrgs.br, E-mail: bardobodmann@ufrgs.br, E-mail: richard.vasques@fulbrightmail.org [Universidade Federal do Rio Grande do Sul (UFRGS), Porto Alegre, RS (Brazil). Programa de Pos-Graduacao em Engenharia Mecanica

    2015-07-01

    The neutron point kinetics equation, which models the time-dependent behavior of nuclear reactors, is often used to understand the dynamics of nuclear reactor operations. It consists of a system of coupled differential equations that models the interaction between (i) the neutron population; and (II) the concentration of the delayed neutron precursors, which are radioactive isotopes formed in the fission process that decay through neutron emission. These equations are deterministic in nature, and therefore can provide only average values of the modeled populations. However, the actual dynamical process is stochastic: the neutron density and the delayed neutron precursor concentrations vary randomly with time. To address this stochastic behavior, Hayes and Allen have generalized the standard deterministic point kinetics equation. They derived a system of stochastic differential equations that can accurately model the random behavior of the neutron density and the precursor concentrations in a point reactor. Due to the stiffness of these equations, this system was numerically implemented using a stochastic piecewise constant approximation method (Stochastic PCA). Here, we present a study of the influence of stochastic fluctuations on the results of the neutron point kinetics equation. We reproduce the stochastic formulation introduced by Hayes and Allen and compute Monte Carlo numerical results for examples with constant and time-dependent reactivity, comparing these results with stochastic and deterministic methods found in the literature. Moreover, we introduce a modified version of the stochastic method to obtain a non-stiff solution, analogue to a previously derived deterministic approach. (author)

  16. The solution of the neutron point kinetics equation with stochastic extension: an analysis of two moments

    International Nuclear Information System (INIS)

    Silva, Milena Wollmann da; Vilhena, Marco Tullio M.B.; Bodmann, Bardo Ernst J.; Vasques, Richard

    2015-01-01

    The neutron point kinetics equation, which models the time-dependent behavior of nuclear reactors, is often used to understand the dynamics of nuclear reactor operations. It consists of a system of coupled differential equations that models the interaction between (i) the neutron population; and (II) the concentration of the delayed neutron precursors, which are radioactive isotopes formed in the fission process that decay through neutron emission. These equations are deterministic in nature, and therefore can provide only average values of the modeled populations. However, the actual dynamical process is stochastic: the neutron density and the delayed neutron precursor concentrations vary randomly with time. To address this stochastic behavior, Hayes and Allen have generalized the standard deterministic point kinetics equation. They derived a system of stochastic differential equations that can accurately model the random behavior of the neutron density and the precursor concentrations in a point reactor. Due to the stiffness of these equations, this system was numerically implemented using a stochastic piecewise constant approximation method (Stochastic PCA). Here, we present a study of the influence of stochastic fluctuations on the results of the neutron point kinetics equation. We reproduce the stochastic formulation introduced by Hayes and Allen and compute Monte Carlo numerical results for examples with constant and time-dependent reactivity, comparing these results with stochastic and deterministic methods found in the literature. Moreover, we introduce a modified version of the stochastic method to obtain a non-stiff solution, analogue to a previously derived deterministic approach. (author)

  17. Uncertainty analysis of neutron transport calculation

    International Nuclear Information System (INIS)

    Oka, Y.; Furuta, K.; Kondo, S.

    1987-01-01

    A cross section sensitivity-uncertainty analysis code, SUSD was developed. The code calculates sensitivity coefficients for one and two-dimensional transport problems based on the first order perturbation theory. Variance and standard deviation of detector responses or design parameters can be obtained using cross section covariance matrix. The code is able to perform sensitivity-uncertainty analysis for secondary neutron angular distribution(SAD) and secondary neutron energy distribution(SED). Covariances of 6 Li and 7 Li neutron cross sections in JENDL-3PR1 were evaluated including SAD and SED. Covariances of Fe and Be were also evaluated. The uncertainty of tritium breeding ratio, fast neutron leakage flux and neutron heating was analysed on four types of blanket concepts for a commercial tokamak fusion reactor. The uncertainty of tritium breeding ratio was less than 6 percent. Contribution from SAD/SED uncertainties are significant for some parameters. Formulas to estimate the errors of numerical solution of the transport equation were derived based on the perturbation theory. This method enables us to deterministically estimate the numerical errors due to iterative solution, spacial discretization and Legendre polynomial expansion of transfer cross-sections. The calculational errors of the tritium breeding ratio and the fast neutron leakage flux of the fusion blankets were analysed. (author)

  18. Methodological developments and applications of neutron activation analysis

    International Nuclear Information System (INIS)

    Kucera, J.

    2007-01-01

    The paper reviews the author's experience acquired and achievements made in methodological developments of neutron activation analysis (NAA) of mostly biological materials. These involve epithermal neutron activation analysis, radiochemical neutron activation analysis using both single- and multi-element separation procedures, use of various counting modes, and the development and use of the self-verification principle. The role of NAA in the detection of analytical errors is discussed and examples of applications of the procedures developed are given. (author)

  19. Analysis of a first-order delay differential-delay equation containing two delays

    Science.gov (United States)

    Marriott, C.; Vallée, R.; Delisle, C.

    1989-09-01

    An experimental and numerical analysis of the behavior of a two-delay differential equation is presented. It is shown that much of the system's behavior can be related to the stability behavior of the underlying linearized modes. A new phenomenon, mode crossing, is explored.

  20. Improvements of neutron activation techniques for the determination of fissile material concentrations

    International Nuclear Information System (INIS)

    Papadopoulos, N.N.

    1987-01-01

    Certain experimental improvements, as variable sample size and irradiation position, automation and flexibility in radiation detection, broaden the measurable concentration range, increase the possible rate and accuracy of analysis and enlarge the application range of home-made nuclear analyzer for fissile material analysis by delayed fission neutron counting and for short-lived multielement analysis by neutron activation gamma-ray spectrometry. Intercomparisons of results by various methods and laboratories show the need for regular checks of techniques to ensure reliable measurements. (author)

  1. Development of Pneumatic Transfer Irradiation Facility (PTS no.1) for Neutron Activation Analysis at HANARO Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Y. S.; Moon, J. H.; Kim, S. H.; Sun, G. M.; Baek, S. Y.; Kim, H. R.; Kim, Y. J

    2008-03-15

    A pneumatic transfer system (PTS) is one of the most important facilities used during neutron irradiation of a target material for instrumental neutron activation analysis (INAA) in a research reactor. In particular, a fast pneumatic transfer system is essential for the measurement of a short half-life nuclide and a delayed neutron counting system. The pneumatic transfer system (PTS no.1) involving a manual system and an semiautomatic system were reconstructed with new designs of a functional improvement at the HANARO research reactor in 2006. In this technical report, the conception, design, operation and control of these system (PTS no.1) was described. Also the experimental results and the characteristic parameters measured by a mock-up test, a functional operation test and an irradiation test of these systems, such as the transfer time of irradiation capsule, the different neutron flux, the temperature of the irradiation position with an irradiation time, the radiation dose rate when the rabbit is returned, etc. are reported to provide a user information as well as a reactor's management and safety.

  2. Uses of neutron capture gamma-rays in environmental pollution applications

    International Nuclear Information System (INIS)

    AbdAl-Samad, M.A.

    1998-01-01

    As a sensitive and accurate technique, the prompt gamma-rays neutron activation is used with success for elemental analysis. The advantages of this method over the other techniques are rapidity, usage of relatively large sample size and high reliability, beside the detection of the elements which have no gamma activity during the delayed neutron activation analysis or very short lived isotopes. Actually different techniques could be used for estimating the trace, minor and major elements of these environmental samples which are considered as complex samples. In the mean time the neutron activation analysis techniques have been improved and have become an excellent tool for elemental analysis of complex samples (Duffey et al., 1970; Senftle et al., 1971; Henkelmm and Born, 1973 ; Hassan et al., .; 1981, 1982, 1983; Clyton et al., 1983; Zaghloul et al., 1993) and the advantages of the prompt γ- ray neutron activation analysis over the other techniques put this technique in the fore front

  3. COSTANZA, 1-D 2 Group Space-Dependent Reactor Dynamics of Spatial Reactor with 1 Group Delayed Neutrons

    International Nuclear Information System (INIS)

    Agazzi, A.; Gavazzi, C.; Vincenti, E.; Monterosso, R.

    1964-01-01

    1 - Nature of physical problem solved: The programme studies the spatial dynamics of reactor TESI, in the two group and one space dimension approximation. Only one group of delayed neutrons is considered. The programme simulates the vertical movement of the control rods according to any given movement law. The programme calculates the evolution of the fluxes and temperature and precursor concentration in space and time during the power excursion. 2 - Restrictions on the complexity of the problem: The maximum number of lattice points is 100

  4. Steel research using neutron beam techniques. In-situ neutron diffraction, small-angle neutron scattering and residual stress analysis

    International Nuclear Information System (INIS)

    Sueyoshi, Hitoshi; Ishikawa, Nobuyuki; Yamada, Katsumi; Sato, Kaoru; Nakagaito, Tatsuya; Matsuda, Hiroshi; Arakaki, Yu; Tomota, Yo

    2014-01-01

    Recently, the neutron beam techniques have been applied for steel researches and industrial applications. In particular, the neutron diffraction is a powerful non-destructive method that can analyze phase transformation and residual stress inside the steel. The small-angle neutron scattering is also an effective method for the quantitative evaluation of microstructures inside the steel. In this study, in-situ neutron diffraction measurements during tensile test and heat treatment were conducted in order to investigate the deformation and transformation behaviors of TRIP steels. The small-angle neutron scattering measurements of TRIP steels were also conducted. Then, the neutron diffraction analysis was conducted on the high strength steel weld joint in order to investigate the effect of the residual stress distribution on the weld cracking. (author)

  5. Breached fuel location in FFTF by delayed neutron monitor triangulation

    International Nuclear Information System (INIS)

    Bunch, W.L.; Tang, E.L.

    1985-10-01

    The Fast Flux Test Facility (FFTF) features a three-loop, sodium-cooled 400 MWt mixed oxide fueled reactor designed for the irradiation testing of fuels and materials for use in liquid metal cooled fast reactors. To establish the ultimate capability of a particular fuel design and thereby generate information that will lead to improvements, many of the fuel irradiations are continued until a loss of cladding integrity (failure) occurs. When the cladding fails, fission gas escapes from the fuel pin and enters the reactor cover gas system. If the cladding failure permits the primary sodium to come in contact with the fuel, recoil fission products can enter the sodium. The presence of recoil fission products in the sodium can be detected by monitoring for the presence of delayed neutrons in the coolant. It is the present philosophy to not operate FFTF when a failure has occurred that permits fission fragments to enter the sodium. Thus, it is important that the identity and location of the fuel assembly that contains the failed cladding be established in order that it might be removed from the core. This report discusses method of location of fuel element when cladding is breached

  6. An in-beam Compton-suppressed Ge spectrometer for nondestructive neutron activation analysis

    International Nuclear Information System (INIS)

    Zaghloul, R.; Abd El-Haleam, A.; Mostafa, M.; Gantner, E.; Ache, H.J.

    1993-04-01

    A high-efficiency compton background suppressed gamma-ray spectrometer by anti-coincidence counting with a NaI(Tl)-shield around a central HPGe-detector for in-beam prompt gamma-ray neutron activation analysis (AC-PGNAA) using a Cf-252 neutron source has been designed and built to provide simultaneous anti-coincidence spectrometry of natural, industrial and environmental samples. The spectrometer consists of a high-purity germanium detector as the main detector and a large volume cylindrical NaI(Tl) detector as a guard detector. The assembly has the ability to measure instantaneously, simultaneously and nondestructively bulk samples up to about 50 cm 3 . Major constituent elements in several rocks and minerals such as H, B, N, Na, Mg, Al, Si, Cl, K, Ca, P, S, Ti, Fe, Sm, Nd, Mn and Gd can be determined, while oxygen cannot be measured due to its small capture cross section (0.27 mb). Several important minor and trace elements such as B, Cd and Hg beside the low residual activity, rare earths and short-lived isotopes could be detected. The sensitivity of the AC-PGNAA technique is limited by the available neutron flux at the target matrix and the neutron absorption cross section of the elements of interest. PGNAA has the advantage to estimate the constituent elements which are difficult to be measured through the delayed gamm-ray measurements such as B, Bi, C, H, P, Tl, Be, Cl and S in industrial and reference materials and those elements which are transformed into other stable isotopes when undergoing neutron capture. The design of the spectrometer assembly, its properties and performance are described

  7. Applications of neutrons for laboratory and industrial activation analysis problems

    International Nuclear Information System (INIS)

    Szabo, Elek; Bakos, Laszlo

    1986-01-01

    This chapter presents some particular applications and case studies of neutrons in activation analysis for research and industrial development purposes. The reactor neutrons have been applied in Hungarian laboratories for semiconductor research, for analysis of geological (lunar) samples, and for a special comparator measurement of samples. Some industrial applications of neutron generator and sealed sources for analytical problems are presented. Finally, prompt neutron activation analysis is outlined briefly. (R.P.)

  8. Neutron activation analysis: principle and methods

    International Nuclear Information System (INIS)

    Reddy, A.V.R.; Acharya, R.

    2006-01-01

    Neutron activation analysis (NAA) is a powerful isotope specific nuclear analytical technique for simultaneous determination of elemental composition of major, minor and trace elements in diverse matrices. The technique is capable of yielding high analytical sensitivity and low detection limits (ppm to ppb). Due to high penetration power of neutrons and gamma rays, NAA experiences negligible matrix effects in the samples of different origins. Depending on the sample matrix and element of interest NAA technique is used non-destructively, known as instrumental neutron activation analysis (INAA), or through chemical NAA methods. The present article describes principle of NAA, different methods and gives a overview some applications in the fields like environment, biology, geology, material sciences, nuclear technology and forensic sciences. (author)

  9. Delayed photoneutrons of the of the Dalat Nuclear Research Reactor

    International Nuclear Information System (INIS)

    Ngo Quang Huy; Ha Van Thong; Vu Hai Long; Ngo Phu Khang; Nguyen Nhi Dien; Pham Van Lam; Huynh Dong Phuong; Luong Ba Vien; Le Vinh Vinh

    1994-01-01

    Time spectrum of delayed neutrons of the Dalat nuclear research reactor is measured and analyzed. It corresponds to a shut-down neutron fluxes of about 10 5 /10 8 n/cm 2 /sec after 100 hours continuous reactor operation at steady power level of 500 kW. Data processing of experimental time neutron spectrum gives 16 exponents, of which 10, resulting from photoneutrons due to (γ,n) reactions on beryllium used inside the reactor core, are obtained by using successive exponential stripping fitting method. For the Dalat reactor, the effective delayed photoneutron fraction relative to the total effective delayed neutron fraction is β B e eff =0.49%β eff for a beryllium weight relative to U 235 fuel of m B e/m U = 8.5. This result is acceptable in comparison to those obtained for other Be-U 235 media. (author). 5 refs., 2 figs., 4 tabs

  10. In vivo neutron activation facility at Brookhaven National Laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Ma, R.; Yasumura, Seiichi; Dilmanian, F.A.

    1997-11-01

    Seven important body elements, C, N, Ca, P, K, Na, and Cl, can be measured with great precision and accuracy in the in vivo neutron activation facilities at Brookhaven National Laboratory. The facilities include the delayed-gamma neutron activation, the prompt-gamma neutron activation, and the inelastic neutron scattering systems. In conjunction with measurements of total body water by the tritiated-water dilution method several body compartments can be defined from the contents of these elements, also with high precision. In particular, body fat mass is derived from total body carbon together with total body calcium and nitrogen; body protein mass is derived from total body nitrogen; extracellular fluid volume is derived from total body sodium and chlorine; lean body mass and body cell mass are derived from total body potassium; and, skeletal mass is derived from total body calcium. Thus, we suggest that neutron activation analysis may be valuable for calibrating some of the instruments routinely used in clinical studies of body composition. The instruments that would benefit from absolute calibration against neutron activation analysis are bioelectric impedance analysis, infrared interactance, transmission ultrasound, and dual energy x-ray/photon absorptiometry.

  11. Fast neutron activation analysis

    International Nuclear Information System (INIS)

    Pepelnik, R.

    1986-01-01

    Since 1981 numerous 14 MeV neutron activation analyses were performed at Korona. On the basis of that work the advantages of this analysis technique and therewith obtained results are compared with other analytical methods. The procedure of activation analysis, the characteristics of Korona, some analytical investigations in environmental research and material physics, as well as sources of systematic errors in trace analysis are described. (orig.) [de

  12. Measurement of uranium enrichment by 14 MeV neutron irradiation

    International Nuclear Information System (INIS)

    Rezende, H.R.

    1987-01-01

    a non-destructive technique for the determination of uranium in UO 2 samples was developed, making use of the change in the fission cross section of a nuclide with the neutron energy. The active interrogation method was used by irradiating the samples with pulsed 14 MeV neutrons and further detection of delayed fission neutrons. In order to discriminate U-238 from U-235 the neutron energy was tailored by means of two concentric cylinders of lead and paraffin/poliethylene, 11 and 4 cm thick. Between neutron pulses, delayed neutrons from fission were detected by a long counter built with five BF 3 proportional counters. Calibration curves for enrichment and total mass versus delayed neutron response were obtained using available UO 2 pellets of known enrichment. Enrichment detection limit, obtained with 95% confidence level by the Student distribution was estimated to be 0.33%. The minimal detectable mass was estimated to be 4.4 g. (author) [pt

  13. Measure of uranium enrichment by 14 MeV neutron irradiation

    International Nuclear Information System (INIS)

    Rezende, H.R.

    1987-01-01

    A non-destructive technique for the determination of uranium in UO 2 samples was developed, marking use of the change in the fission cross of a nuclide with the neutron energy. The active interrogation method was used by irradiating the samples with pulsed 14 MeV neutrons and furtherdetection of delayed fission neutrons. In order to descriminated U-238 from U-235 the neutron energy was tailored by means of two concentric cylinders of lead and paraffin/poliethylene, 11 and 4 cm thick. Between neutron pulses, delayed neutrons from fission were detected by a long counter built with five BF 3 proportional counters. Calibration curves for enrichment and total mass versus delayed neutron response were obtained using available UO 2 pellets of Known enrichment. Enrichment detection limit, obtained with 95% confidence level by the the Student distribution was estimated to be 0.33%. The minimal detectable mass was estimated to be 4.4 g. (Author) [pt

  14. Man/machine interface algorithm for advanced delayed-neutron signal characterization system

    International Nuclear Information System (INIS)

    Gross, K.C.

    1985-01-01

    The present failed-element rupture detector (FERD) at Experimental Breeder Reactor II (EBR-II) consists of a single bank of delayed-neutron (DN) detectors at a fixed transit time from the core. Plans are currently under way to upgrade the FERD in 1986 and provide advanced DN signal characterization capability that is embodied in an equivalent-recoil-area (ERA) meter. The new configuration will make available to the operator a wealth of quantitative diagnostic information related to the condition and dynamic evolution of a fuel breach. The diagnostic parameters will include a continuous reading of the ERA value for the breach; the transit time, T/sub tr/, for DN emitters traveling from the core to the FERD; and the isotopic holdup time, T/sub h/, for the source. To enhance the processing, interpretation, and display of these parameters to the reactor operator, a man/machine interface (MMI) algorithm has been developed to run in the background on EBR-II's data acquisition system (DAS). The purpose of this paper is to describe the features and implementation of this newly developed MMI algorithm

  15. Principle of neutron activation analysis and its use for determination of trace elements in sediment

    International Nuclear Information System (INIS)

    Verma, Rakesh

    2012-01-01

    Neutron Activation analysis (NAA) is a multi element analysis technique, often non-destructive in nature where approximately 75 elements can be measured with the detection limits ranging from 10 -6 to 10 -12 g of element in a sample. Typical sample sizes range from 1 mg to 1 g, however in principle much larger samples can be activated and the size is only limited by the capacity of the neutron irradiation facility. In NAA, a sample (solid or liquid or gas) is exposed to neutrons and radiations emitted by the radioactive products, formed during the nuclear reaction, are measured using a suitable detector. The energy of the emitted radiation is a characteristic of a radioisotope whereas the intensity of the emitted radiation is proportional to the mass of the analyte. NAA can be carried out by measurement of (i) prompt gamma rays emitted by compound nucleus, called prompt gamma ray NAA (PGNAA) and (ii) β rays emitted from radioactive product or delayed gamma rays emitted subsequent to β decay, called conventional NAA or simply NAA. PGNAA is an online measurement method. PGNAA is complementary to conventional NAA in terms of analyzing low Z elements. Conventional NAA is an offline method and is easy to perform. Depending upon the nature of matrix and analyte to be determined, three approaches are possible in NAA namely, (i) instrumental neutron activation analysis (INAA), (ii) radiochemical neutron activation analysis (RNAA), and (iii) chemical neutron activation analysis (CNAA). Quantification is accomplished by any of the three standardisation methodologies namely (i) absolute method (ii) relative method and (iii) single comparator method. The relative method is most precise and simple to perform. Natural processes responsible for the formation of bottom sediments can be altered by anthropogenic activities. Bottom sediments are a sink as well as a source of contaminants in the aquatic environment. Analysis of-sediments provides environmentally significant

  16. Beta-Delayed Neutron Spectroscopy of 72Co with VANDLE

    Science.gov (United States)

    Keeler, Andrew; Grzywacz, Robert; King, Thomas; Taylor, Steven; Paulauskas, Stanley; Zachary, Christopher; Vandle Collaboration

    2017-09-01

    Measurements of simple, closed-shell isotopes far from stability provide important benchmarks for nuclear models and are a key constraint in r-process calculations. In particular, r-process models are sensitive to beta decay lifetimes and branching ratios of these neutron-rich isotopes. In this experiment, the Versatile Array of Neutron Detectors at Low Energy (VANDLE) was used to observe decays of nuclei produced by the fragmentation of 82Se at the National Superconducting Cyclotron Laboratory (NSCL). The neutron and gamma emissions of 72Co were measured to map the beta strength distribution (S_beta) above the neutron separation energy and infer the size of the Z = 28 shell gap in the 78Ni region. An implantation detector made of a radiation-hardened, inorganic scintillator was used to correlate implanted ions with beta decays as well as provide a start signal for the neutron Time of Flight measurement. Funded by the National Nuclear Security Administration under the Stewardship Science Academic Alliances program through DOE Award No. DE-NA0002132 and by the Office of Nuclear Physics, U.S. Department of Energy under Awards No. DE-FG02-96ER40983 (UTK).

  17. Research on neutron noise analysis stochastic simulation method for α calculation

    International Nuclear Information System (INIS)

    Zhong Bin; Shen Huayun; She Ruogu; Zhu Shengdong; Xiao Gang

    2014-01-01

    The prompt decay constant α has significant application on the physical design and safety analysis in nuclear facilities. To overcome the difficulty of a value calculation with Monte-Carlo method, and improve the precision, a new method based on the neutron noise analysis technology was presented. This method employs the stochastic simulation and the theory of neutron noise analysis technology. Firstly, the evolution of stochastic neutron was simulated by discrete-events Monte-Carlo method based on the theory of generalized Semi-Markov process, then the neutron noise in detectors was solved from neutron signal. Secondly, the neutron noise analysis methods such as Rossia method, Feynman-α method, zero-probability method, and cross-correlation method were used to calculate a value. All of the parameters used in neutron noise analysis method were calculated based on auto-adaptive arithmetic. The a value from these methods accords with each other, the largest relative deviation is 7.9%, which proves the feasibility of a calculation method based on neutron noise analysis stochastic simulation. (authors)

  18. Iterative method for obtaining the prompt and delayed alpha-modes of the diffusion equation

    International Nuclear Information System (INIS)

    Singh, K.P.; Degweker, S.B.; Modak, R.S.; Singh, Kanchhi

    2011-01-01

    Highlights: → A method for obtaining α-modes of the neutron diffusion equation has been developed. → The difference between the prompt and delayed modes is more pronounced for the higher modes. → Prompt and delayed modes differ more in reflector region. - Abstract: Higher modes of the neutron diffusion equation are required in some applications such as second order perturbation theory, and modal kinetics. In an earlier paper we had discussed a method for computing the α-modes of the diffusion equation. The discussion assumed that all neutrons are prompt. The present paper describes an extension of the method for finding the α-modes of diffusion equation with the inclusion of delayed neutrons. Such modes are particularly suitable for expanding the time dependent flux in a reactor for describing transients in a reactor. The method is illustrated by applying it to a three dimensional heavy water reactor model problem. The problem is solved in two and three neutron energy groups and with one and six delayed neutron groups. The results show that while the delayed α-modes are similar to λ-modes they are quite different from prompt modes. The difference gets progressively larger as we go to higher modes.

  19. Electric field strength and plasma delay in silicon surface barrier detector

    International Nuclear Information System (INIS)

    Kanno, I.; Inbe, T.; Kanazawa, S.; Kimura, I.

    1994-01-01

    The resistivity change of a silicon irradiated by high energy neutrons became an interest of study associated with the large scale accelerator projects . The increase of the resistivity of the silicon of a silicon surface barrier detector (SSBD) was studied as a function of neutron fluence. The plasma delay, which was an interesting but not favorite timing property of the SSBD, was reported being dependent on the resistivity of silicon . The neutron irradiation brings the change of timing property as well as the resistivity change on the SSBD. The resistivity dependence of the plasma delay should be studied for the purpose of high energy accelerator experiments. Some empirical formulae of the plasma delay were reported, however, there were no discussions on the physical meanings of the resistivity dependence of the plasma delay. The plasma delay in a SSBD is discussed in the light of electric field strength in the depletion layer of the SSBD. The explanation of the plasma delay is presented taking into account of the competing two electric forces. The resistivity of the silicon affects the plasma delay through the electric forces. 3 figs, 3 refs. (author)

  20. Monte Carlo calculations and neutron spectrometry in quantitative prompt gamma neutron activation analysis (PGNAA) of bulk samples using an isotopic neutron source

    International Nuclear Information System (INIS)

    Spyrou, N.M.; Awotwi-Pratt, J.B.; Williams, A.M.

    2004-01-01

    An activation analysis facility based on an isotopic neutron source (185 GBq 241 Am/Be) which can perform both prompt and cyclic activation analysis on bulk samples, has been used for more than 20 years in many applications including 'in vivo' activation analysis and the determination of the composition of bio-environmental samples, such as, landfill waste and coal. Although the comparator method is often employed, because of the variety in shape, size and elemental composition of these bulk samples, it is often difficult and time consuming to construct appropriate comparator samples for reference. One of the obvious problems is the distribution and energy of the neutron flux in these bulk and comparator samples. In recent years, it was attempted to adopt the absolute method based on a monostandard and to make calculations using a Monte Carlo code (MCNP4C2) to explore this further. In particular, a model of the irradiation facility has been made using the MCNP4C2 code in order to investigate the factors contributing to the quantitative determination of the elemental concentrations through prompt gamma neutron activation analysis (PGNAA) and most importantly, to estimate how the neutron energy spectrum and neutron dose vary with penetration depth into the sample. This simulation is compared against the scattered and transmitted neutron energy spectra that are experimentally and empirically determined using a portable neutron spectrometry system. (author)

  1. Delayed photoneutrons of the of the Dalat Nuclear Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Huy, Ngo Quang [Centre for Nuclear Technique Application, Ho Chi Minh City (Viet Nam); Thong, Ha Van; Long, Vu Hai; Khang, Ngo Phu; Dien, Nguyen Nhi; Lam, Pham Van; Phuong, Huynh Dong; Vien, Luong Ba; Vinh, Le Vinh [Nuclear Research Inst., Da Lat (Viet Nam)

    1994-10-01

    Time spectrum of delayed neutrons of the Dalat nuclear research reactor is measured and analyzed. It corresponds to a shut-down neutron fluxes of about 10{sup 5}/10{sup 8} n/cm{sup 2}/sec after 100 hours continuous reactor operation at steady power level of 500 kW. Data processing of experimental time neutron spectrum gives 16 exponents, of which 10, resulting from photoneutrons due to ({gamma},n) reactions on beryllium used inside the reactor core, are obtained by using successive exponential stripping fitting method. For the Dalat reactor, the effective delayed photoneutron fraction relative to the total effective delayed neutron fraction is {beta}{sup B}e{sub eff}=0.49%{beta}{sub eff} for a beryllium weight relative to U{sup 235} fuel of m{sub B}e/m{sub U} = 8.5. This result is acceptable in comparison to those obtained for other Be-U{sup 235} media. (author). 5 refs., 2 figs., 4 tabs.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-05-06

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

  3. Development of a new deuterium-deuterium (D-D) neutron generator for prompt gamma-ray neutron activation analysis.

    Science.gov (United States)

    Bergaoui, K; Reguigui, N; Gary, C K; Brown, C; Cremer, J T; Vainionpaa, J H; Piestrup, M A

    2014-12-01

    A new deuterium-deuterium (D-D) neutron generator has been developed by Adelphi Technology for prompt gamma neutron activation analysis (PGNAA), neutron activation analysis (NAA), and fast neutron radiography. The generator makes an excellent fast, intermediate, and thermal neutron source for laboratories and industrial applications that require the safe production of neutrons, a small footprint, low cost, and small regulatory burden. The generator has three major components: a Radio Frequency Induction Ion Source, a Secondary Electron Shroud, and a Diode Accelerator Structure and Target. Monoenergetic neutrons (2.5MeV) are produced with a yield of 10(10)n/s using 25-50mA of deuterium ion beam current and 125kV of acceleration voltage. The present study characterizes the performance of the neutron generator with respect to neutron yield, neutron production efficiency, and the ionic current as a function of the acceleration voltage at various RF powers. In addition the Monte Carlo N-Particle Transport (MCNP) simulation code was used to optimize the setup with respect to thermal flux and radiation protection. Copyright © 2014 Elsevier Ltd. All rights reserved.

  4. Neutron activation analysis in Bulgaria

    International Nuclear Information System (INIS)

    Apostolov, D.

    1985-01-01

    The development of instrumental neutron activation analysis (INAA) as a routine method started in 1960 with bringing into use of the experimental nuclear reactor 2 MW -IRT-2000. For the purposes of INAA the vertical channels were used. The neutron flux vary from 1 to 6x10 12 n/cm 2 s, with Cd ratio for gold of about 4,4. In one of the channels the neutron flux is additionally thermalised with grafite, in others - a pneumatic double-tube rabbit system is installed. One of the irradiation positions is equiped with 1 mm Cd shield constantly. With the pressure of the working gas (air) of 2 bar the transport time in one direction is 2,5 sec. Because of lack of special system for uniform irradiation an accuracy of 3% can be reached by use of iron monitors for long irradiations and copper monitors for use in the rabbit system. Two neutron generators are also working but the application of 14 MeV neutrons for INAA is still quite limited. The most developed are the applications of INAA in the fields of geology and paedology, medicine and biology, environment and pollution, archaeology, metallurgy, metrology and hydrology, criminology

  5. Study of the momentum loss achromate and its application to the measurement of the β-delayed neutron radioactivity of 14Be, 17B, and 19C

    International Nuclear Information System (INIS)

    Hanelt, E.

    1992-02-01

    In this thesis it was shown that the projectile fragmentation at relativistic projectile velocities is a production mechanism for exotic nuclei, which is because of its advantageous kinematics especially suited for the fast and efficient separation of the reaction product in an ion optical system. An essential result of these studies is that projectile fragments can be separated in a wide energy range from about 100 MeV/nucleon to 1 GeV/nucleon and over the whole mass range by means of a momentum-loss achromate. In the experiment described in this thesis this method was for the first time applied to the measurement of the β-deLayed neutron radioactivity. The studied isotopes - 1 - 4Be, - 1 - 7B, and - 1 - 9C were produced by the fragmentation of a - 2 - 2Ne beam at 60 MeV/nucleon. A measurement of β half-lifes and neutron branching ratios was performed, the accuracy of which was in other experiments with similarly exotic nuclei hitherto hardly reached. In - 1 - 7B thereby for the first time a β-delayed 4-neutron radioactivity could be detected. The results of these measurements were compared with calculations from different theoretical models. The observed multiplicities of the β-delayed neutrons are consistent with the multiplicities, which are expected by means of a comparison of the Q - β values and the neutron binding energies. The measured neutron branching ratios yield indirect information on distribution of the β strength in the daugther nuclei. At time none of the theories is yet able to reproduce these experimental values in sufficient way. (orig./HSI) [de

  6. Investigation of an egyptian phosphate ore sample by neutron activation analysis technique

    International Nuclear Information System (INIS)

    Eissa, E.A.; Aly, R.A.; Rofail, N.B.; Hassan, A.M.

    1995-01-01

    A domestic phosphate ore sample has been analysed by means of prompt and delayed gamma-ray spectrometry following the activation by thermal neutron capture technique. The rabbit pneumatic transfer system (RPTS), long irradiation facility and two Pu/Be (2,5 Ci each) neutron sources set-Pu for prompt (n,gamma) were applied. The high purity germanium (HPGe) gamma-ray spectrometer with a personal computer analyzer (PCA) system were used for spectrum measurements. Programmes on the VAX computer were utilized for estimating the elemental concentrations of 22 out of 36 elements identified in this work. 2 tabs

  7. The analysis and attribution of the time-dependent neutron background resultant from sample irradiation in a SLOWPOKE-2 reactor

    International Nuclear Information System (INIS)

    Sellers, M.T.; Corcoran, E.C.; Kelly, D.G.

    2013-01-01

    The Royal Military College of Canada (RMCC) has commissioned a Delayed Neutron Counting (DNC) system for the analysis of special nuclear materials. A significant, time-dependent neutron background with an initial maximum count rate, more than 50 times that of the time-independent background, was characterised during the validation of this system. This time-dependent background was found to be dependent on the presence of the polyethylene (PE) vials used to transport the fissile samples, yet was not an activation product of vial impurities. The magnitude of the time-dependent background was found to be irradiation site specific and independent of the mass of PE. The capability of RMCC's DNC system to analyze the neutron count rates in time intervals 235 U contamination was present on each irradiated vial. However, Inductively Coupled Plasma-Mass Spectroscopy measurements of material leached from the outer vial surfaces after their irradiations found only trace amounts of uranium, 0.118 ± 0.048 ng of 235 U derived from natural uranium. These quantities are insufficient to account for the time-independent background, and in fact could not be discriminated from the noise associated with time-independent background. It is suggested that delayed neutron emitters are deposited in the vial surface following fission recoil, leaving the main body of uranium within the irradiation site. This hypothesis is supported by the physical cleaning of the site with materials soaked in distilled water and HNO 3 , which lowered the background from a nominal 235 U mass equivalent of 120 to 50 ng per vial. (author)

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

    International Nuclear Information System (INIS)

    Jervis, R.E.

    1993-01-01

    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)

  9. Detail analysis of fusion neutronics benchmark experiment on beryllium

    International Nuclear Information System (INIS)

    Konno, Chikara; Ochiai, Kentaro; Takakura, Kosuke; Ohnishi, Seiki; Kondo, Keitaro; Wada, Masayuki; Sato, Satoshi

    2010-01-01

    Our previous analysis of the integral experiments (in situ and TOF experiments) on beryllium with DT neutrons at JAEA/FNS pointed out two problems by using MCNP4C and the latest nuclear data libraries; one was a strange larger neutron peak around 12 MeV appearing in the TOF experiment analysis with JEFF-3.1 and the other was an overestimation on law energy neutrons in the in situ experiment analyses with all the nuclear data libraries. We investigated reasons for these problems in detail. It was found out that the official ACE file MCJEFF3.1 of JEFF-3.1 had an inconsistency with the original JEFF-3.1, which caused the strange larger neutron peak around 12 MeV in the TOF experiment analysis. We also found out that the calculated thermal neutron peak was probably too large in the in situ experiment. On trial we examined influence of the thermal neutron scattering law data of beryllium metal in ENDF/B-VI. The result pointed out that the coherent elastic scattering cross-section data in the thermal neutron scattering law data of beryllium metal were probably too large.

  10. Neutron kinetics in moderators and SNM detection through epithermal-neutron-induced fissions

    Energy Technology Data Exchange (ETDEWEB)

    Gozani, Tsahi, E-mail: tgmaven@gmail.com [1050 Harriet St., Palo Alto, CA 94301 (United States); King, Michael J. [Rapiscan Laboratories Inc., 520 Almanor Ave., Sunnyvale, CA 94085 (United States)

    2016-01-01

    Extension of the well-established Differential Die Away Analysis (DDAA) into a faster time domain, where more penetrating epithermal neutrons induce fissions, is proposed and demonstrated via simulations and experiments. In the proposed method the fissions stimulated by thermal, epithermal and even higher-energy neutrons are measured after injection of a narrow pulse of high-energy 14 MeV (d,T) or 2.5 MeV (d,D) source neutrons, appropriately moderated. The ability to measure these fissions stems from the inherent correlation of neutron energy and time (“E–T” correlation) during the process of slowing down of high-energy source neutrons in common moderating materials such as hydrogenous compounds (e.g., polyethylene), heavy water, beryllium and graphite. The kinetic behavior following injection of a delta-function-shaped pulse (in time) of 14 MeV neutrons into such moderators is studied employing MCNPX simulations and, when applicable, some simple “one-group” models. These calculations served as a guide for the design of a source moderator which was used in experiments. Qualitative relationships between slowing-down time after the pulse and the prevailing neutron energy are discussed. A laboratory system consisting of a 14 MeV neutron generator, a polyethylene-reflected Be moderator, a liquid scintillator with pulse-shape discrimination (PSD) and a two-parameter E–T data acquisition system was set up to measure prompt neutron and delayed gamma-ray fission signatures in a 19.5% enriched LEU sample. The measured time behavior of thermal and epithermal neutron fission signals agreed well with the detailed simulations. The laboratory system can readily be redesigned and deployed as a mobile inspection system for SNM in, e.g., cars and vans. A strong pulsed neutron generator with narrow pulse (<75 ns) at a reasonably high pulse frequency could make the high-energy neutron induced fission modality a realizable SNM detection technique.

  11. Quantitative analysis of boron by neutron radiography

    International Nuclear Information System (INIS)

    Bayuelken, A.; Boeck, H.; Schachner, H.; Buchberger, T.

    1990-01-01

    The quantitative determination of boron in ores is a long process with chemical analysis techniques. As nuclear techniques like X-ray fluorescence and activation analysis are not applicable for boron, only the neutron radiography technique, using the high neutron absorption cross section of this element, can be applied for quantitative determinations. This paper describes preliminary tests and calibration experiments carried out at a 250 kW TRIGA reactor. (orig.) [de

  12. The use of computed neutron coincidence counting with time interval analysis for the analysis of Fork-measurements on a fresh MOX-LWR fuel assembly under water

    Energy Technology Data Exchange (ETDEWEB)

    Baeten, P.; Bruggeman, M.; Carchon, R

    1998-06-01

    The objective of this study was to investigate the influence of different important parameters on measurement results for various fork-detectors. Computed Neutron Coincidence Counting (CNCC) with Time Interval Analysis (TIA) was used for this study. The performance of the electronics for the different fork-detectors was studied by investigating the deadtime perturbed zone of the Rossi-alpha distribution in TIA. The measurement revealed anomalies in the performance of the electronics of the IAEA BWR and LANL fork-detector. The IAEA PWR fork-detector functioned well and the deadtime parameter was calculated. The optimal setting for the pre delay was investigated and it was found that a pre delay of 10 micro seconds should be considered as an optimum between excluding from analysis data in the deadtime perturbed zone and keeping a high signal-to-noise ratio. For the shift register electronics used with the fork-detectors, a pre delay of only 4.5 micro seconds was used. The study of the pre delay and the deadtime showed that the calculated triples-rate is strongly dependent on these parameters. An accurate determination of the triple-rate in this type of measurements has proven to be quite difficult and requires proper operation of the electronics, a correct pre delay and an accurate deadtime correction formalism. By varying the boron concentration in water, the change of the decay time of the Rossi-alpha distribution was clearly observed. This change is due to the variation of the thermal multiplication. The variation of this decay time with the boron concentration proves that Boehnel's model for fast neutron multiplication is not valid under these measurement conditions and that a model for fast and thermal multiplication should be used in order to obtain unbiased measurement results. CNCC with TIA has proved to be a valuable tool in which parameter settings can be varied a posterori and the optimal setting can be determined for each measurement. Moreover, the

  13. The use of computed neutron coincidence counting with time interval analysis for the analysis of Fork-measurements on a fresh MOX-LWR fuel assembly under water

    International Nuclear Information System (INIS)

    Baeten, P.; Bruggeman, M.; Carchon, R.

    1998-06-01

    The objective of this study was to investigate the influence of different important parameters on measurement results for various fork-detectors. Computed Neutron Coincidence Counting (CNCC) with Time Interval Analysis (TIA) was used for this study. The performance of the electronics for the different fork-detectors was studied by investigating the deadtime perturbed zone of the Rossi-alpha distribution in TIA. The measurement revealed anomalies in the performance of the electronics of the IAEA BWR and LANL fork-detector. The IAEA PWR fork-detector functioned well and the deadtime parameter was calculated. The optimal setting for the pre delay was investigated and it was found that a pre delay of 10 micro seconds should be considered as an optimum between excluding from analysis data in the deadtime perturbed zone and keeping a high signal-to-noise ratio. For the shift register electronics used with the fork-detectors, a pre delay of only 4.5 micro seconds was used. The study of the pre delay and the deadtime showed that the calculated triples-rate is strongly dependent on these parameters. An accurate determination of the triple-rate in this type of measurements has proven to be quite difficult and requires proper operation of the electronics, a correct pre delay and an accurate deadtime correction formalism. By varying the boron concentration in water, the change of the decay time of the Rossi-alpha distribution was clearly observed. This change is due to the variation of the thermal multiplication. The variation of this decay time with the boron concentration proves that Boehnel's model for fast neutron multiplication is not valid under these measurement conditions and that a model for fast and thermal multiplication should be used in order to obtain unbiased measurement results. CNCC with TIA has proved to be a valuable tool in which parameter settings can be varied a posterori and the optimal setting can be determined for each measurement. Moreover, the

  14. TDTORT: Time-Dependent, 3-D, Discrete Ordinates, Neutron Transport Code System with Delayed Neutrons

    International Nuclear Information System (INIS)

    2002-01-01

    1 - Description of program or function: TDTORT solves the time-dependent, three-dimensional neutron transport equation with explicit representation of delayed neutrons to estimate the fission yield from fissionable material transients. This release includes a modified version of TORT from the C00650MFMWS01 DOORS3.1 code package plus the time-dependent TDTORT code. GIP is also included for cross-section preparation. TORT calculates the flux or fluence of particles due to particles incident upon the system from extraneous sources or generated internally as a result of interaction with the system in two- or three-dimensional geometric systems. The principle application is to the deep-penetration transport of neutrons and photons. Reactor eigenvalue problems can also be solved. Numerous printed edits of the results are available, and results can be transferred to output files for subsequent analysis. TDTORT reads ANISN-format cross-section libraries, which are not included in the package. Users may choose from several available in RSICC's data library collection which can be identified by the keyword 'ANISN FORMAT'. 2 - Methods:The time-dependent spatial flux is expressed as a product of a space-, energy-, and angle-dependent shape function, which is usually slowly varying in time and a purely time-dependent amplitude function. The shape equation is solved for the shape using TORT; and the result is used to calculate the point kinetics parameters (e.g., reactivity) by using their inner product definitions, which are then used to solve the time-dependent amplitude and precursor equations. The amplitude function is calculated by solving the kinetics equations using the LSODE solver. When a new shape calculation is needed, the flux is calculated using the newly computed amplitude function. The Boltzmann transport equation is solved using the method of discrete ordinates to treat the directional variable and weighted finite-difference methods, in addition to Linear Nodal

  15. Neutron capture prompt gamma-ray activation analysis at the NIST cold neutron research facility

    Energy Technology Data Exchange (ETDEWEB)

    Lindstrom, R M; Zeisler, R; Vincent, D H; Greenberg, R R; Stone, C A; Mackey, E A [National Inst. of Standards and Technology, Gaithersburg, MD (United States); Anderson, D L [Food and Drug Administration, Washington, DC (United States); Clark, D D [Cornell Univ., Ithaca, NY (United States)

    1993-01-01

    An instrument for neutron capture prompt gamma-ray activation analysis (PGAA) has been constructed as part of the Cold Neutron Research Facility at the 20 MW National Institute of Standards and Technology Research Reactor. The neutron fluence rate (thermal equivalent) is 1.5*10[sup 8] n*cm[sup -2]*s[sup -] [sup 1], with negligible fast neutrons and gamma-rays. With compact geometry and hydrogen-free construction, the sensitivity is sevenfold better than an existing thermal instrument. Hydrogen background is thirtyfold lower. (author) 17 refs.; 2 figs.

  16. Development of the delyed-neutron triangulation technique for locating failed fuel in LMFBR

    International Nuclear Information System (INIS)

    Kryter, R.C.

    1975-01-01

    Two major accomplishments of the ORNL delayed neutron triangulation program are (1) an analysis of anticipated detector counting rates and sensitivities to unclad fuel and erosion types of pin failure, and (2) an experimental assessment of the accuracy with which the position of failed fuel can be determined in the FFTF (this was performed in a quarter-scale water mockup of realistic outlet plenum geometry using electrolyte injections and conductivity cells to simulate delayed-neutron precursor releases and detections, respectively). The major results and conclusions from these studies are presented, along with plans for further DNT development work at ORNL for the FFTF and CRBR. (author)

  17. Identifying functions for ex-core neutron noise analysis

    International Nuclear Information System (INIS)

    Avila, J.M.; Oliveira, J.C.

    1987-01-01

    A method of performing the phase analysis of signals arising from neutron detectors placed in the periphery of a pressurized water reactor is proposed. It consists in the definition of several identifying functions, based on the phases of cross power spectral densities corresponding to four ex-core neutron detectors. Each of these functions enhances the appearance of different sources of noise. The method, applied to the ex-core neutron fluctuation analysis of a French PWR, proved to be very useful as it allows quick recognition of various patterns in the power spectral densities. (orig.) [de

  18. Applications of neutron activation analysis technique

    International Nuclear Information System (INIS)

    Jonah, S. A.

    2000-07-01

    The technique was developed as far back as 1936 by G. Hevesy and H. Levy for the analysis of Dy using an isotopic source. Approximately 40 elements can be analyzed by instrumental neutron activation analysis (INNA) technique with neutrons from a nuclear reactor. By applying radiochemical separation, the number of elements that can be analysed may be increased to almost 70. Compared with other analytical methods used in environmental and industrial research, NAA has some unique features. These are multi-element capability, rapidity, reproducibility of results, complementarity to other methods, freedom from analytical blank and independency of chemical state of elements. There are several types of neutron sources namely: nuclear reactors, accelerator-based and radioisotope-based sources, but nuclear reactors with high fluxes of neutrons from the fission of 235 U give the most intense irradiation, and hence the highest available sensitivities for NAA. In this paper, the applications of NAA of socio-economic importance are discussed. The benefits of using NAA and related nuclear techniques for on-line applications in industrial process control are highlighted. A brief description of the NAA set-ups at CERT is enumerated. Finally, NAA is compared with other leading analytical techniques

  19. Determination of uranium and thorium contents using a 14 MeV neutron generator and a radiometric method

    International Nuclear Information System (INIS)

    Casagrande, J.A.

    1981-04-01

    A simple method was developed which can determine uranium and thorium in uranium ores, by 14MeV neutron activation and delayed neutron counting. The process can be used in field laboratories to select samples for processing. The method does not require a previous treatment of the samples and the analysis time is below 5 minutes. The detection limit of the method is about 2 ppm, when the yield of the 14MeV source has a value of 2 X 10 11 neutrons/second, and an optimized delayed neutron counter is used. A radiometric method is used determine separately the thorium content of the sample, and this result is combined with the activation one in order to obtain uranium content. The radiometric method in the counting of the 2,6 MeV gamma rays from 208 Tl using a NaI(Tl) detector. Delayed neutron counting is performed with BF 3 detectors inside a paraffin box. The problem of radioactive equilibrium does not affect thorium determination since the biggest activities of thorium daughters are much smaller than the times involved in the displacements of mineral which can give origin to the radioactive desequilibrium. (Author) [pt

  20. Forensic neutron activation analysis - the Japanese scene

    International Nuclear Information System (INIS)

    Kishi, Tohru.

    1986-01-01

    The progress of forensic neutron activation analysis/FNAA/ in Japan is described. FNAA began in 1965 and during the past 20 years many cases have been handled; these include determination of toxic materials, comparison examination of physical evidences /e.g.,paints, metal fragments, plastics and inks/ and drug sample differenciation. Neutron activation analysis is applied routinely to the scientific criminal investigation as one of multielement analytical techniques. This paper also discusses these routine works. (author)

  1. Time-delayed chameleon: Analysis, synchronization and FPGA implementation

    Science.gov (United States)

    Rajagopal, Karthikeyan; Jafari, Sajad; Laarem, Guessas

    2017-12-01

    In this paper we report a time-delayed chameleon-like chaotic system which can belong to different families of chaotic attractors depending on the choices of parameters. Such a characteristic of self-excited and hidden chaotic flows in a simple 3D system with time delay has not been reported earlier. Dynamic analysis of the proposed time-delayed systems are analysed in time-delay space and parameter space. A novel adaptive modified functional projective lag synchronization algorithm is derived for synchronizing identical time-delayed chameleon systems with uncertain parameters. The proposed time-delayed systems and the synchronization algorithm with controllers and parameter estimates are then implemented in FPGA using hardware-software co-simulation and the results are presented.

  2. Application of ray-traced tropospheric slant delays to geodetic VLBI analysis

    Science.gov (United States)

    Hofmeister, Armin; Böhm, Johannes

    2017-08-01

    The correction of tropospheric influences via so-called path delays is critical for the analysis of observations from space geodetic techniques like the very long baseline interferometry (VLBI). In standard VLBI analysis, the a priori slant path delays are determined using the concept of zenith delays, mapping functions and gradients. The a priori use of ray-traced delays, i.e., tropospheric slant path delays determined with the technique of ray-tracing through the meteorological data of numerical weather models (NWM), serves as an alternative way of correcting the influences of the troposphere on the VLBI observations within the analysis. In the presented research, the application of ray-traced delays to the VLBI analysis of sessions in a time span of 16.5 years is investigated. Ray-traced delays have been determined with program RADIATE (see Hofmeister in Ph.D. thesis, Department of Geodesy and Geophysics, Faculty of Mathematics and Geoinformation, Technische Universität Wien. http://resolver.obvsg.at/urn:nbn:at:at-ubtuw:1-3444, 2016) utilizing meteorological data provided by NWM of the European Centre for Medium-Range Weather Forecasts (ECMWF). In comparison with a standard VLBI analysis, which includes the tropospheric gradient estimation, the application of the ray-traced delays to an analysis, which uses the same parameterization except for the a priori slant path delay handling and the used wet mapping factors for the zenith wet delay (ZWD) estimation, improves the baseline length repeatability (BLR) at 55.9% of the baselines at sub-mm level. If no tropospheric gradients are estimated within the compared analyses, 90.6% of all baselines benefit from the application of the ray-traced delays, which leads to an average improvement of the BLR of 1 mm. The effects of the ray-traced delays on the terrestrial reference frame are also investigated. A separate assessment of the RADIATE ray-traced delays is carried out by comparison to the ray-traced delays from the

  3. Particulate matter and neutron activation analysis

    International Nuclear Information System (INIS)

    Otoshi, Tsunehiko

    2003-01-01

    In these years, economy of East Asian region is rapidly growing, and countries in this region are facing serious environmental problems. Neutron activation analysis is known as one of high-sensitive analytical method for multi elements. And it is a useful tool for environmental research, particularly for the study on atmospheric particulate matter that consists of various constituents. Elemental concentration represents status of air, such as emission of heavy metals from industries and municipal incinerators, transportation of soil derived elements more than thousands of kilometers, and so on. These monitoring data obtained by neutron activation analysis can be a cue to evaluate environment problems. Japanese government launched National Air Surveillance Network (NASN) employing neutron activation analysis in 1974, and the data has been accumulated at about twenty sampling sites. As a result of mitigation measure of air pollution sources, concentrations of elements that have anthropogenic sources decreased particularly at the beginning of the monitoring period. However, even now, concentrations of these anthropogenic elements reflect the characteristics of each sampling site, e.g. industrial/urban, rural, and remote. Soil derived elements have a seasonal variation because of the contribution of continental dust transported by strong westerly winds prevailing in winter and spring season. The health effects associated with trace elements in particulate matter have not been well characterized. However, there is increasing evidence that particulate air pollution, especially fine portion of particles in many different cities is associated with acute mortality. Neutron activation analysis is also expected to provide useful information to this new study field related to human exposures and health risk. (author)

  4. Passivity analysis and synthesis for uncertain time-delay systems

    Directory of Open Access Journals (Sweden)

    Magdi S. Mahmoud

    2001-01-01

    Full Text Available In this paper, we investigate the robust passivity analysis and synthesis problems for a class of uncertain time-delay systems. This class of systems arises in the modelling effort of studying water quality constituents in fresh stream. For the analysis problem, we derive a sufficient condition for which the uncertain time-delay system is robustly stable and strictly passive for all admissible uncertainties. The condition is given in terms of a linear matrix inequality. Both the delay-independent and delay-dependent cases are considered. For the synthesis problem, we propose an observer-based design method which guarantees that the closed-loop uncertain time-delay system is stable and strictly passive for all admissible uncertainties. Several examples are worked out to illustrate the developed theory.

  5. Development of educational program for neutron activation analysis

    International Nuclear Information System (INIS)

    Chung, Yong Sam; Moon, Jong Hwa; Kim, Sun Ha; Ryel, Sung; Kang, Young Hwan; Lee, Kil Yong; Yeon, Yeon Yel; Cho, Seung Yeon

    2000-08-01

    This technical report is developed to apply an educational and training program for graduate student and analyst utilizing neutron activation analysis. The contents of guide book consists of five parts as follows; introduction, gamma-ray spectrometry and measurement statistics, its applications, to understand of comprehensive methodology and to utilize a relevant knowledge and information on neutron activation analysis

  6. Development of educational program for neutron activation analysis

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Yong Sam; Moon, Jong Hwa; Kim, Sun Ha; Ryel, Sung; Kang, Young Hwan; Lee, Kil Yong; Yeon, Yeon Yel; Cho, Seung Yeon

    2000-08-01

    This technical report is developed to apply an educational and training program for graduate student and analyst utilizing neutron activation analysis. The contents of guide book consists of five parts as follows; introduction, gamma-ray spectrometry and measurement statistics, its applications, to understand of comprehensive methodology and to utilize a relevant knowledge and information on neutron activation analysis.

  7. Neutron activation analysis: Modelling studies to improve the neutron flux of Americium-Beryllium source

    Energy Technology Data Exchange (ETDEWEB)

    Didi, Abdessamad; Dadouch, Ahmed; Tajmouati, Jaouad; Bekkouri, Hassane [Advanced Technology and Integration System, Dept. of Physics, Faculty of Science Dhar Mehraz, University Sidi Mohamed Ben Abdellah, Fez (Morocco); Jai, Otman [Laboratory of Radiation and Nuclear Systems, Dept. of Physics, Faculty of Sciences, Tetouan (Morocco)

    2017-06-15

    Americium–beryllium (Am-Be; n, γ) is a neutron emitting source used in various research fields such as chemistry, physics, geology, archaeology, medicine, and environmental monitoring, as well as in the forensic sciences. It is a mobile source of neutron activity (20 Ci), yielding a small thermal neutron flux that is water moderated. The aim of this study is to develop a model to increase the neutron thermal flux of a source such as Am-Be. This study achieved multiple advantageous results: primarily, it will help us perform neutron activation analysis. Next, it will give us the opportunity to produce radio-elements with short half-lives. Am-Be single and multisource (5 sources) experiments were performed within an irradiation facility with a paraffin moderator. The resulting models mainly increase the thermal neutron flux compared to the traditional method with water moderator.

  8. Evaluation of response function of moderating-type neutron detector and application to environmental neutron measurement

    International Nuclear Information System (INIS)

    Kosako, Toshiso; Nakamura, Takashi; Iwai, Satoshi; Katsuki, Shinji; Kamata, Masashi.

    1983-08-01

    The energy-dependent response function of a multi-cylinder moderating-type BF 3 counter, so-called Bonner counter, was calculated by the time-dependent multi-group Monte Carlo code, TMMCR. The calculated response function was evaluated experimentally for neutron energy below about 50 keV down to epithermal energy by the time-of-flight method combining with a large lead pile at the Nuclear Engineering Research Laboratory, University of Tokyo and also above 50 keV by using the monoenergetic neutron standard field a t the Electrotechnical Laboratory. The time delay in the polyethylene moderator of the Bonner counter due to multiple collisions with hydrogen was analyzed by the TMMCR code and used for the time-spectrum analysis of the time-of-flight measurement. The response function obtained by these two experiments showed good agreement with the calculated results. This Bonner counter having a response function evaluated from thermal to MeV energy range was used for spectrometry and dosimetry of environmental neutrons around some nuclear facilities. The neutron spectra and dose measured in the environment around a 252 Cf fission source, fast neutron source reactor and electron synchrotron were all in good agreement with the calculated results and the measured results with other neutron detectors. (author)

  9. Neutron cross-sections database for amino acids and proteins analysis

    Energy Technology Data Exchange (ETDEWEB)

    Voi, Dante L.; Ferreira, Francisco de O.; Nunes, Rogerio Chaffin, E-mail: dante@ien.gov.br, E-mail: fferreira@ien.gov.br, E-mail: Chaffin@ien.gov.br [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil); Rocha, Helio F. da, E-mail: hrocha@gbl.com.br [Universidade Federal do Rio de Janeiro (IPPMG/UFRJ), Rio de Janeiro, RJ (Brazil). Instituto de Pediatria

    2015-07-01

    Biological materials may be studied using neutrons as an unconventional tool of analysis. Dynamics and structures data can be obtained for amino acids, protein and others cellular components by neutron cross sections determinations especially for applications in nuclear purity and conformation analysis. The instrument used for this is the crystal spectrometer of the Instituto de Engenharia Nuclear (IEN-CNEN-RJ), the only one in Latin America that uses neutrons for this type of analyzes and it is installed in one of the reactor Argonauta irradiation channels. The experimentally values obtained are compared with calculated values using literature data with a rigorous analysis of the chemical composition, conformation and molecular structure analysis of the materials. A neutron cross-section database was constructed to assist in determining molecular dynamic, structure and formulae of biological materials. The database contains neutron cross-sections values of all amino acids, chemical elements, molecular groups, auxiliary radicals, as well as values of constants and parameters necessary for the analysis. An unprecedented analytical procedure was developed using the neutron cross section parceling and grouping method for data manipulation. This database is a result of measurements obtained from twenty amino acids that were provided by different manufactories and are used in oral administration in hospital individuals for nutritional applications. It was also constructed a small data file of compounds with different molecular groups including carbon, nitrogen, sulfur and oxygen, all linked to hydrogen atoms. A review of global and national scene in the acquisition of neutron cross sections data, the formation of libraries and the application of neutrons for analyzing biological materials is presented. This database has further application in protein analysis and the neutron cross-section from the insulin was estimated. (author)

  10. Neutron cross-sections database for amino acids and proteins analysis

    International Nuclear Information System (INIS)

    Voi, Dante L.; Ferreira, Francisco de O.; Nunes, Rogerio Chaffin; Rocha, Helio F. da

    2015-01-01

    Biological materials may be studied using neutrons as an unconventional tool of analysis. Dynamics and structures data can be obtained for amino acids, protein and others cellular components by neutron cross sections determinations especially for applications in nuclear purity and conformation analysis. The instrument used for this is the crystal spectrometer of the Instituto de Engenharia Nuclear (IEN-CNEN-RJ), the only one in Latin America that uses neutrons for this type of analyzes and it is installed in one of the reactor Argonauta irradiation channels. The experimentally values obtained are compared with calculated values using literature data with a rigorous analysis of the chemical composition, conformation and molecular structure analysis of the materials. A neutron cross-section database was constructed to assist in determining molecular dynamic, structure and formulae of biological materials. The database contains neutron cross-sections values of all amino acids, chemical elements, molecular groups, auxiliary radicals, as well as values of constants and parameters necessary for the analysis. An unprecedented analytical procedure was developed using the neutron cross section parceling and grouping method for data manipulation. This database is a result of measurements obtained from twenty amino acids that were provided by different manufactories and are used in oral administration in hospital individuals for nutritional applications. It was also constructed a small data file of compounds with different molecular groups including carbon, nitrogen, sulfur and oxygen, all linked to hydrogen atoms. A review of global and national scene in the acquisition of neutron cross sections data, the formation of libraries and the application of neutrons for analyzing biological materials is presented. This database has further application in protein analysis and the neutron cross-section from the insulin was estimated. (author)

  11. Delayed neutron fraction and prompt decay constant measurement in the MINERVE reactor using the PSI instrumentation

    Energy Technology Data Exchange (ETDEWEB)

    Perret, Gregory [Paul Scherrer Institute, Villigen, 5232, (Switzerland)

    2015-07-01

    The critical decay constant (B/A), delayed neutron fraction (B) and generation time (A) of the Minerve reactor were measured by the Paul Scherrer Institut (PSI) and the Commissariat a l'Energie Atomique (CEA) in September 2014 using the Feynman-alpha and Power Spectral Density neutron noise measurement techniques. Three slightly subcritical configuration were measured using two 1-g {sup 235}U fission chambers. This paper reports on the results obtained by PSI in the near critical configuration (-2g). The most reliable and precise results were obtained with the Cross-Power Spectral Density technique: B = 708.4±9.2 pcm, B/A = 79.0±0.6 s{sup -1} and A 89.7±1.4 micros. Predictions of the same kinetic parameters were obtained with MCNP5-v1.6 and the JEFF-3.1 and ENDF/B-VII.1 nuclear data libraries. On average the predictions for B and B/A overestimate the experimental results by 5% and 11%, respectively. The discrepancy is suspected to come from either a corruption of the data or from the inadequacy of the point kinetic equations to interpret the measurements in the Minerve driven system. (authors)

  12. Self-Powered Neutron Detector Qualification for Absolute On-Line In-Pile Neutron Flux Measurements in BR2

    Science.gov (United States)

    Vermeeren, L.; Wéber, M.

    2003-06-01

    A set of ten Self-Powered Neutron Detectors with Co, Rh and Ag emitters has been irradiated in several channels of the BR2 research reactor at SCK•CEN aiming at a comparison of their performance as thermal neutron flux detectors under various conditions. To allow for a correct interpretation of their signals, all detector sensitivity contributions (prompt and delayed) were calculated using a dedicated Monte Carlo model. The various contributions were also measured separately; the agreement between calculated and experimental data, including data from activation dosimetry, was excellent. Detailed neutron flux profiles were obtained from the SPND data, after correction for the finite detector lengths and for the slow response of delayed SPNDs.

  13. Application of neutron activation analysis

    International Nuclear Information System (INIS)

    Dybczynski, R.

    2001-01-01

    The physical basis and analytical possibilities of neutron activation analysis have been performed. The number of applications in material engineering, geology, cosmology, oncology, criminology, biology, agriculture, environment protection, archaeology, history of art and especially in chemical analysis have been presented. The place of the method among other methods of inorganic quantitative chemical analysis for trace elements determination has been discussed

  14. High-sensitivity measurements for low-level TRU wastes using advanced passive neutron techniques

    International Nuclear Information System (INIS)

    Menlove, H.O.; Eccleston, G.W.

    1992-01-01

    In recent years, both passive- and active-neutron nondestructive assay (NDA) systems have been used to measure the uranium and plutonium content in 200-ell drums. Because of the heterogeneity of the wastes, representative sampling is not possible and NDA methods are preferred over destructive analysis. Active-neutron assay systems are used to measure the fissile isotopes such as 235 U, 23 Pu, and 241 Pu; the isotopic ratios are used to infer the total plutonium content and thus the specific disintegration rate. The active systems include 14-MeV-neutron (DT) generators with delayed-neutron counting, (D,T) generators with the differential die-away technique, and 252 Cf delayed-neutron shufflers. Passive assay systems (for example, segmented gamma-ray scanners)5 have used gamma-ray sessions, while others (for example, passive drum counters) used passive-neutron signals. We have developed a new passive-neutron measurement technique to improve the accuracy and sensitivity of the NDA of plutonium scrap and waste. This new 200-ell-drum assay system combines the classical NDA method of counting passive-neutron totals and coincidences from plutonium with the new features of ''add-a-source'' (AS) and multiplicity counting to improve the accuracy of matrix corrections and statistical techniques that improve the low-level detectability limits. This paper describes the improvements we have made in passive-neutron assay systems and compares the accuracies and detectability limits of passive- and active-neutron assay systems

  15. Non-destructive bulk analysis of the Buggenum sword by neutron resonance capture analysis and neutron diffraction

    International Nuclear Information System (INIS)

    Postma, H.; Clarijs, M.; Borella, A.; Schillebeeckx, P.; Kamermans, H.

    2010-01-01

    Two neutron based techniques, neutron resonance capture analysis (NRCA) and time-of-flight neutron-diffraction (TOF-ND) have been used to determine the elemental composition and structure of a precious and very well preserved all-metal sword from the Bronze Age. This Buggenum sword was on loan from the National Museum of Antiquities (NMA) in Leiden (NL). NRCA and TOF-ND experiments have been carried out at a number of more or less identical positions of the sword. The tin-bronze ratio and the relative amounts of some minor elements (Sb, As, Ag, In) have been determined. The results of neutron diffraction measurements showed considerable tin-segregation, and clear indications of hardening on the edges of the blade. In addition, radiographs using Bremsstrahlung revealed the construction of the hilt-blade connection. The work was carried out at the EC Joint Research Centre IRMM in Geel (B) and at the ISIS facility of the Rutherford Appleton Laboratory (UK). (author)

  16. Measurement of uranium and plutonium in solid waste by passive photon or neutron counting and isotopic neutron source interrogation

    Energy Technology Data Exchange (ETDEWEB)

    Crane, T.W.

    1980-03-01

    A summary of the status and applicability of nondestructive assay (NDA) techniques for the measurement of uranium and plutonium in 55-gal barrels of solid waste is reported. The NDA techniques reviewed include passive gamma-ray and x-ray counting with scintillator, solid state, and proportional gas photon detectors, passive neutron counting, and active neutron interrogation with neutron and gamma-ray counting. The active neutron interrogation methods are limited to those employing isotopic neutron sources. Three generic neutron sources (alpha-n, photoneutron, and /sup 252/Cf) are considered. The neutron detectors reviewed for both prompt and delayed fission neutron detection with the above sources include thermal (/sup 3/He, /sup 10/BF/sub 3/) and recoil (/sup 4/He, CH/sub 4/) proportional gas detectors and liquid and plastic scintillator detectors. The instrument found to be best suited for low-level measurements (< 10 nCi/g) is the /sup 252/Cf Shuffler. The measurement technique consists of passive neutron counting followed by cyclic activation using a /sup 252/Cf source and delayed neutron counting with the source withdrawn. It is recommended that a waste assay station composed of a /sup 252/Cf Shuffler, a gamma-ray scanner, and a screening station be tested and evaluated at a nuclear waste site. 34 figures, 15 tables.

  17. Capture analysis of element content of a substance with other neutron methods

    International Nuclear Information System (INIS)

    Kurbanov, B.I.

    2004-01-01

    Full text: Neutron analysis method of determining element composition have found wide range of applications in industry thanks to different types of interaction of neutron with substances /1/. With the aim of widening the range of problems to be solved, on the basis of the device /2/ for determining the element content of substance, possibilities of combining the method based on the use of neutron capture gamma-ray spectrometry with other neutron methods, in particular neutron activation analysis and neutron absorption analysis were studied. In this radionuclide source ( 252 Cf) with the yield of 1,5 x 10 7 neutron/sec is used. By means of using neutron capture gamma radiation spectrometry the possibilities of determining some elements (H, B, N, S etc. ), which are not determined by very widely used method, activation analysis. These elements can be determined by both the semiconductor and scintillation detectors with parameters fitting the manufacturing requirements. And for a number of elements ( B, Cl, Cd, Sm, Gd) very high limits of determination ( up to 10- 5 %) are possible using semiconductor Ge (Li) -detectors with high resolution. Possibility of determination of some 'well' activated elements ( K, Al, Fe, Mn, Ti, Sc etc.) in samples of ore and products of their processing using the neutron-activation analysis. For 1 hour of irradiation on the experimental device quite accurate analytical peak, of these elements are obtained, allowing to determine them qualitatively. However, with decreasing neutron yield of radionuclide source it becomes more difficult to achieve the necessary parameters both in neutron capture and activation analysis. Experimental works on determination of some elements with large cross-sections of capture ( B, Cd, Sm ) by absorption of neutrons in the investigated substance, i.e. using the neutron absorption analysis method with absence of other large capture cross section elements in the samples being studied

  18. Prospects for accelerator neutron sources for large volume minerals analysis

    International Nuclear Information System (INIS)

    Clayton, C.G.; Spackman, R.

    1988-01-01

    The electron Linac can be regarded as a practical source of thermal neutrons for activation analysis of large volume mineral samples. With a suitable target and moderator, a neutron flux of about 10 10 n/cm/s over 2-3 kg of rock can be generated. The proton Linac gives the possibility of a high neutron yield (> 10 12 n/s) of fast neutrons at selected energies. For the electron Linac, targets of W-U and W-Be are discussed. The advantages and limitations of the system are demonstrated for the analysis of gold in rocks and ores and for platinum in chromitite. These elements were selected as they are most likely to justify an accelerator installation at the present time. Errors due to self shielding in gold particles for thermal neutrons are discussed. The proton Linac is considered for neutrons generated from a lithium target through the 7 Li(p, n) 7 Be reaction. The analysis of gold by fast neutron activation is considered. This approach avoids particle self-absorption and, by appropriate proton energy selection, avoids potentially dominating interfering reactions. The analysis of 235 U in the presence of 238 U and 232 Th is also considered. (author)

  19. A consistent, differential versus integral, method for measuring the delayed neutron yield in fissions

    International Nuclear Information System (INIS)

    Flip, A.; Pang, H.F.; D'Angelo, A.

    1995-01-01

    Due to the persistent uncertainties: ∼ 5 % (the uncertainty, here and there after, is at 1σ) in the prediction of the 'reactivity scale' (β eff ) for a fast power reactor, an international project was recently initiated in the framework of the OECD/NEA activities for reevaluation, new measurements and integral benchmarking of delayed neutron (DN) data and related kinetic parameters (principally β eff ). Considering that the major part of this uncertainty is due to uncertainties in the DN yields (v d ) and the difficulty for further improvement of the precision in differential (e.g. Keepin's method) measurements, an international cooperative strategy was adopted aiming at extracting and consistently interpreting information from both differential (nuclear) and integral (in reactor) measurements. The main problem arises from the integral side; thus the idea was to realize β eff like measurements (both deterministic and noise) in 'clean' assemblies. The 'clean' calculational context permitted the authors to develop a theory allowing to link explicitly this integral experimental level with the differential one, via a unified 'Master Model' which relates v d and measurables quantities (on both levels) linearly. The combined error analysis is consequently largely simplified and the final uncertainty drastically reduced (theoretically, by a factor √3). On the other hand the same theoretical development leading to the 'Master Model', also resulted in a structured scheme of approximations of the general (stochastic) Boltzmann equation allowing a consistent analysis of the large range of measurements concerned (stochastic, dynamic, static ... ). This paper is focused on the main results of this theoretical development and its application to the analysis of the Preliminary results of the BERENICE program (β eff measurements in MASURCA, the first assembly in CADARACHE-FRANCE)

  20. Preliminary Formulation of Finite Element Solution for the 1-D, 1-G Time Dependent Neutron Diffusion Equation without Consideration about Delay Neutron

    Energy Technology Data Exchange (ETDEWEB)

    Ryu, Eun Hyun; Song, Yong Mann; Park, Joo Hwan [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-05-15

    If time-dependent equation is solved with the FEM, the limitation of the input geometry will disappear. It has often been pointed out that the numerical methods implemented in the RFSP code are not state-of-the-art. Although an acceleration method such as the Coarse Mesh Finite Difference (CMFD) for Finite Difference Method (FDM) does not exist for the FEM, one should keep in mind that the number of time steps for the transient simulation is not large. The rigorous formulation in this study will richen the theoretical basis of the FEM and lead to an extension of the dynamics code to deal with a more complicated problem. In this study, the formulation for the 1-D, 1-G Time Dependent Neutron Diffusion Equation (TDNDE) without consideration of the delay neutron will first be done. A problem including one multiplying medium will be solved. Also several conclusions from a comparison between the numerical and analytic solutions, a comparison between solutions with various element orders, and a comparison between solutions with different time differencing will be made to be certain about the formulation and FEM solution. By investigating various cases with different values of albedo, theta, and the order of elements, it can be concluded that the finite element solution is agree well with the analytic solution. The higher the element order used, the higher the accuracy improvements are obtained.

  1. Automated activation-analysis system

    International Nuclear Information System (INIS)

    Minor, M.M.; Garcia, S.R.; Denton, M.M.

    1982-01-01

    An automated delayed neutron counting and instrumental neutron activation analysis system has been developed at Los Alamos National Laboratory's Omega West Reactor (OWR) to analyze samples for uranium and 31 additional elements with a maximum throughput of 400 samples per day

  2. Monte Carlo criticality analysis for dissolvers with neutron poison

    International Nuclear Information System (INIS)

    Yu, Deshun; Dong, Xiufang; Pu, Fuxiang.

    1987-01-01

    Criticality analysis for dissolvers with neutron poison is given on the basis of Monte Carlo method. In Monte Carlo calculations of thermal neutron group parameters for fuel pieces, neutron transport length is determined in terms of maximum cross section approach. A set of related effective multiplication factors (K eff ) are calculated by Monte Carlo method for the three cases. Related numerical results are quite useful for the design and operation of this kind of dissolver in the criticality safety analysis. (author)

  3. Causal Analysis of Railway Running Delays

    DEFF Research Database (Denmark)

    Cerreto, Fabrizio; Nielsen, Otto Anker; Harrod, Steven

    Operating delays and network propagation are inherent characteristics of railway operations. These are traditionally reduced by provision of time supplements or “slack” in railway timetables and operating plans. Supplement allocation policies must trade off reliability in the service commitments...... Denmark (the Danish infrastructure manager). The statistical analysis of the data identifies the minimum running times and the scheduled running time supplements and investigates the evolution of train delays along given train paths. An improved allocation of time supplements would result in smaller...

  4. Neutron activation analysis in minerals prospecting

    International Nuclear Information System (INIS)

    Gomez, H.; Duque O, J.

    1988-01-01

    One method multielemental analysis in geological samples has been developed by neutron activation analysis without using standards and by eliminating many of the error sources of the absolute method. It uses the ratio of the activities induced by mass unit, between the element in the sample and one cobalt monitor. The detection limits are good for more than thirty elements in many prospecting programs, with a standard deviation less than 7%. The neutron flux used is 2x10 11 nxcm -2 .S -1 and the HPGE detector has a relative efficiency of 20% and an energy resolution of 1.9 KeV in 1332 KeV photopeak

  5. Neutronic analysis of JET external neutron monitor response

    Energy Technology Data Exchange (ETDEWEB)

    Snoj, Luka, E-mail: luka.snoj@ijs.si [Reactor Physics Division, Jožef Stefan Institute, Jamova cesta 39, SI-1000 Ljubljana (Slovenia); Lengar, Igor; Čufar, Aljaž [Reactor Physics Division, Jožef Stefan Institute, Jamova cesta 39, SI-1000 Ljubljana (Slovenia); Syme, Brian; Popovichev, Sergey [Culham Centre for Fusion Energy, Culham Science Centre, Abingdon, OX14 3DB, United Kingdom (United Kingdom); Batistoni, Paola [ENEA C. R. Frascati, via E. Fermi 45, 00044 Frascati, Roma (Italy); Conroy, Sean [VR Association, Uppsala University, Department of Physics and Astronomy, PO Box 516, SE-75120 Uppsala (Sweden)

    2016-11-01

    Highlights: • We model JET tokamak containing JET remote handling system. • We investigate effect of remote handling system on external neutron monitor response. • Remote handling system correction factors are calculated. • Integral correction factors are relatively small, i.e up to 8%. - Abstract: The power output of fusion devices is measured in terms of the neutron yield which relates directly to the fusion yield. JET made a transition from Carbon wall to ITER-Like Wall (Beryllium/Tungsten/Carbon) during 2010–11. Absolutely calibrated measurement of the neutron yield by JET neutron monitors was ensured by direct measurements using a calibrated {sup 252}Cf neutron source (NS) deployed by the in-vessel remote handling system (RHS) inside the JET vacuum vessel. Neutronic calculations were required in order to understand the neutron transport from the source in the vacuum vessel to the fission chamber detectors mounted outside the vessel on the transformer limbs of the tokamak. We developed a simplified computational model of JET and the JET RHS in Monte Carlo neutron transport code MCNP and analyzed the paths and structures through which neutrons reach the detectors and the effect of the JET RHS on the neutron monitor response. In addition we performed several sensitivity studies of the effect of substantial massive structures blocking the ports on the external neutron monitor response. As the simplified model provided a qualitative picture of the process only, some calculations were repeated using a more detailed full 3D model of the JET tokamak.

  6. Nondestructive elemental analysis of coins using accelerator-based thermal neutrons

    International Nuclear Information System (INIS)

    Khairi, F.Z.; Aksoy, A.; Al-Haddad, M.N.

    2007-01-01

    The accelerator-based thermal-neutrons activation analysis setup at KFUPM has an adequate thermal -neutron flux that can be advantageously used for the elemental analysis of a variety of samples including archeological ones. The thermal neutrons are derived from the moderation of fast neutrons from the D (d, n) He reaction which produces fast 2.5 MeV neutrons. A maximum thermals flux of about 2.5x10 n/m-s was achieved. For the purpose of determining the suitability of the set up for the analysis of contemporary and ancient coins, we carried out a feasibility study by irradiating a selected number of Saudi Arabian coins dating from 1958 to 1987 in the thermal-neutron flux. The induced gamma-ray activities were then counted using a HP-GMX detector coupled to a PC-based data acquisition and analysis system. The elements that were determined in the coins were copper (75%), nickel (around 25%) and manganese (<0.5%). Calibration curves were also established for these elements. The determined concentrations are in agreement with the data published by the Standard Catalogue of World Coins. (author)

  7. Safeguards and Physics Measurements: Neutron Activation Analysis with k0-standardisation

    International Nuclear Information System (INIS)

    Pomme, S.

    2000-01-01

    SCK-CEN's programme on Neutron Activation Analysis with k 0 -standardisation concentrates on the improvement of the standardisation method and the characterisation of the neutron field as well as on the improvement of the statistical control on neutron activation analysis. Main achievements in 2000 are reported

  8. Neutron noise analysis for malfunction diagnosis at sodium cooled reactors

    International Nuclear Information System (INIS)

    Hoppe, P.

    1978-09-01

    For the investigation of the potential use of neutron noise analysis at sodium cooled power reactors, measurements have been performed at the KNK I reactor over a period of 18 month under different operational conditions. The signal fluctuations of the following tranducers have been recorded: In-core and Ex-core neutron detectors, temperature-, flow-, pressure-, vibration- and acoustic sensors. These extensive measurements have been analyzed in the frequency range from 0,001 Hz to 1000 Hz with all currently known methods for the identification of noise sources. The following results have been found: - Neutron noise for f 20 Hz the white detection noise prevails. In the region from 1 Hz to 20 Hz the vibrations of core components contribute to neutron noise. - Neutron noise is influenced by the state of the plant. - The contributions to neutron noise due to the fluctuations of coolant flow and inlet temperature are small compared to those produced by the movements of the control rod initiated by the reactor control system. The quantitatively unidentifiable amount of reactivity fluctuations (0,6 time-dependent thermal bowing of the core. With respect to these results and by calculation of the neutron noise patterns to be expected for the SNR 300, the following possible applications for neutron noise analysis have been found: By means of neutron noise analysis only reactivity fluctuations can be identified and supervised which are produced by time dependent changes of the core geometry. Furthermore neutron noise analysis is well suited for a sensitive detection of control rod vibrations and of local sodium boiling. Finally it can be used for the surveillance of the proper functioning of the reactor control system and of the control rod drive mechanism. (orig./HP) 891 HP [de

  9. Thermal neutron self-shielding correction factors for large sample instrumental neutron activation analysis using the MCNP code

    International Nuclear Information System (INIS)

    Tzika, F.; Stamatelatos, I.E.

    2004-01-01

    Thermal neutron self-shielding within large samples was studied using the Monte Carlo neutron transport code MCNP. The code enabled a three-dimensional modeling of the actual source and geometry configuration including reactor core, graphite pile and sample. Neutron flux self-shielding correction factors derived for a set of materials of interest for large sample neutron activation analysis are presented and evaluated. Simulations were experimentally verified by measurements performed using activation foils. The results of this study can be applied in order to determine neutron self-shielding factors of unknown samples from the thermal neutron fluxes measured at the surface of the sample

  10. Recent applications of neutron activation analysis at Lucas Heights

    International Nuclear Information System (INIS)

    Fardy, J.J.

    1978-01-01

    The use of neutron activation analysis to determine key elemental distribution patterns in samples from both the energy industry and health science field is summarised. Instrumental neutron activation analysis has successfully measured simultaneously more than twenty elements in a sample of brown coal from Victoria, black coal from New South Wales and samples from the product stream of ACIRL's batch autoclave, solvent-refined, coal hydrogenation process. Four gallstones removed from the same gallbladder have been examined instrumentally by neutron activation analysis. A total of sixteen trace elements were detected with concentrations in the range 0.8 ng g -1 for gold to 7,800 μg g -1 for calcium

  11. New Results on Passivity Analysis of Stochastic Neural Networks with Time-Varying Delay and Leakage Delay

    Directory of Open Access Journals (Sweden)

    YaJun Li

    2015-01-01

    Full Text Available The passivity problem for a class of stochastic neural networks systems (SNNs with varying delay and leakage delay has been further studied in this paper. By constructing a more effective Lyapunov functional, employing the free-weighting matrix approach, and combining with integral inequality technic and stochastic analysis theory, the delay-dependent conditions have been proposed such that SNNs are asymptotically stable with guaranteed performance. The time-varying delay is divided into several subintervals and two adjustable parameters are introduced; more information about time delay is utilised and less conservative results have been obtained. Examples are provided to illustrate the less conservatism of the proposed method and simulations are given to show the impact of leakage delay on stability of SNNs.

  12. Neutron activation analysis in archaeological chemistry

    Energy Technology Data Exchange (ETDEWEB)

    Harbottle, G [Brookhaven National Lab., Upton, NY (United States)

    1990-01-01

    There is a long history of the application of chemical analysis to archaeological problems, extending to the last years of the 18th century. The nuclear-age technique of neutron activation analysis, permitting the simultaneous, sensitive, non-destructive estimation of many elements in an archaeological specimen, has found wide application. Important advances have been made, using this technique, in locating the origins of archaeological artifacts such as ceramics, metals, obsidian and semiprecious stones, among other articles of ancient ritual and commerce. In addition, the technique of neutron activation analysis has proved to be almost ideal in studies tracing the development of ancient technologies such as glass-making and smelting. In the future, the development of data banks of analyses of archaeological materials should provide an excellent new tool in studies of prehistory.

  13. Neutron activation analysis in archaeological chemistry

    International Nuclear Information System (INIS)

    Harbottle, G.

    1990-01-01

    There is a long history of the application of chemical analysis to archaeological problems, extending to the last years of the 18th century. The nuclear-age technique of neutron activation analysis, permitting the simultaneous, sensitive, non-destructive estimation of many elements in an archaeological specimen, has found wide application. Important advantages have been made, using this technique, in locating the origins of archaeological artifacts such as ceramics, metals, obsidian and semiprecious stones, among other articles of ancient ritual and commerce. In addition, the technique of neutron activation analysis has proved to be almost ideal in studies tracing the development of ancient technologies such as glass-making and smelting. In the future, the development of data banks of analyses of archaeological materials should provide an excellent new tool in studies of prehistory. (orig.)

  14. Accounting for the thermal neutron flux depression in voluminous samples for instrumental neutron activation analysis

    International Nuclear Information System (INIS)

    Overwater, R.M.W.; Hoogenboom, J.E.

    1994-01-01

    At the Delft University of Technology Interfaculty Reactor Institute, a facility has been installed to irradiate cylindrical samples with diameters up to 15 cm and weights up to 50 kg for instrumental neutron activation analysis (INAA) purposes. To be able to do quantitative INAA on voluminous samples, it is necessary to correct for gamma-ray absorption, gamma-ray scattering, neutron absorption, and neutron scattering in the sample. The neutron absorption and the neutron scattering are discussed. An analytical solution is obtained for the diffusion equation in the geometry of the irradiation facility. For samples with known composition, the neutron flux--as a function of position in the sample--can be calculated directly. Those of unknown composition require additional flux measurements on which least-squares fitting must be done to obtain both the thermal neutron diffusion coefficient D s and the diffusion length L s of the sample. Experiments are performed to test the theory

  15. Approximate solution for the reactor neutron probability distribution

    International Nuclear Information System (INIS)

    Ruby, L.; McSwine, T.L.

    1985-01-01

    Several authors have studied the Kolmogorov equation for a fission-driven chain-reacting system, written in terms of the generating function G(x,y,z,t) where x, y, and z are dummy variables referring to the neutron, delayed neutron precursor, and detector-count populations, n, m, and c, respectively. Pal and Zolotukhin and Mogil'ner have shown that if delayed neutrons are neglected, the solution is approximately negative binomial for the neutron population. Wang and Ruby have shown that if the detector effect is neglected, the solution, including the effect of delayed neutrons, is approximately negative binomial. All of the authors assumed prompt-neutron emission not exceeding two neutrons per fission. An approximate method of separating the detector effect from the statistics of the neutron and precursor populations has been proposed by Ruby. In this weak-coupling limit, it is assumed that G(x,y,z,t) = H(x,y)I(z,t). Substitution of this assumption into the Kolmogorov equation separates the latter into two equations, one for H(x,y) and the other for I(z,t). Solution of the latter then gives a generating function, which indicates that in the weak-coupling limit, the detector counts are Poisson distributed. Ruby also showed that if the detector effect is neglected in the equation for H(x,y), i.e., the detector efficiency is set to zero, then the resulting equation is identical with that considered by Wang and Ruby. The authors present here an approximate solution for H(x,y) that does not set the detector efficiency to zero

  16. Relation of photofission cross sections and delayed neutron photoproduction in the range of E1-giant resonance. Sootnoshenie mezhdu secheniyami fotodeleniya i fotoobrazovaniya zapazdyvayushchikh nejtronov v oblasti E1-gigantskogo rezonansa

    Energy Technology Data Exchange (ETDEWEB)

    Ganich, P P; Parlag, O A; Sikora, D I; Sychev, S I

    1989-03-01

    Relation between yields and cross sections of photofission and photoproduction is studied in order to use them in the methods for analysis of fissile nuclides. Total yield of delayed neutrons from the {sup 232}Th target and ratios of total yields from {sup 238}U and {sup 232}Th targets were measured in the M=300 microtron in 6-18 MeV energy range. Efficiency of the suggested method for refining the {sup 238}U photofission cross sections in the range of E1-giant resonance is shown.

  17. Neutron resonance analysis for nuclear safeguards and security applications

    Science.gov (United States)

    Paradela, Carlos; Heyse, Jan; Kopecky, Stefan; Schillebeeckx, Peter; Harada, Hideo; Kitatani, Fumito; Koizumi, Mitsuo; Tsuchiya, Harufumi

    2017-09-01

    Neutron-induced reactions can be used to study the properties of nuclear materials of interest in the fields of nuclear safeguards and security. The elemental and isotopic composition of these materials can be determined by using the presence of resonance structures. This idea is the basis of two non-destructive analysis techniques which have been developed at the GELINA neutron time-of-flight facility at JRC-Geel: Neutron Resonance Capture Analysis (NRCA) and Neutron Resonance Transmission Analysis (NRTA). A combination of NRTA and NRCA has been proposed for the characterisation of particle-like debris of melted fuel formed in severe nuclear accidents. In this work, we present a quantitative validation of the NRTA technique which was used to determine the areal densities of Pu enriched reference samples used for safeguards applications. Less than 2% bias has been obtained for the fissile isotopes, with well-known total cross sections.

  18. Digitizing and analysis of neutron generator waveforms

    International Nuclear Information System (INIS)

    Bryant, T.C.

    1977-11-01

    All neutron generator waveforms from units tested at the SLA neutron generator test site are digitized and the digitized data stored in the CDC 6600 tape library for display and analysis using the CDC 6600 computer. The digitizing equipment consists mainly of seven Biomation Model 8100 transient recorders, Digital Equipment Corporation PDP 11/20 computer, RK05 disk, seven-track magnetic tape transport, and appropriate DEC and SLA controllers and interfaces. The PDP 11/20 computer is programmed in BASIC with assembly language drivers. In addition to digitizing waveforms, this equipment is used for other functions such as the automated testing of multiple-operation electronic neutron generators. Although other types of analysis have been done, the largest use of the digitized data has been for various types of graphical displays using the CDC 6600 and either the SD4020 or DX4460 plotters

  19. Neutron-gamma discrimination by pulse analysis with superheated drop detector

    International Nuclear Information System (INIS)

    Das, Mala; Seth, S.; Saha, S.; Bhattacharya, S.; Bhattacharjee, P.

    2010-01-01

    Superheated drop detector (SDD) consisting of drops of superheated liquid of halocarbon is irradiated to neutrons and gamma-rays from 252 Cf fission neutron source and 137 Cs gamma source, respectively, separately. Analysis of pulse height of signals at the neutron and gamma-ray sensitive temperature provides significant information on the identification of neutron and gamma-ray induced events.

  20. Review of fission product yields and delayed neutron data for the actinides NP-237, PU-242, AM-242M, AM-243, CM-243 and CM-245

    International Nuclear Information System (INIS)

    Mills, R.W.

    1990-07-01

    A review of fission product yields and delayed neutron data for Np-237, Pu-242, Am-242m, Am-243, Cm-243 and Cm-245 has been undertaken. Gaps in understanding and inconsistencies in existing data were identified and priority areas for further experimental, theoretical and evaluation investigation detailed

  1. Neutron Activation analysis of waste water

    International Nuclear Information System (INIS)

    Hernandez H, V.

    1997-01-01

    An instrumental neutron activation analysis for the simultaneous determination of chlorine, bromine, sodium, manganese, cobalt, copper, chromium, zinc, nickel, antimony and iron in waste water is described. They were determined in waste water samples under normal conditions by non-destructive neutron activation simultaneously using a suitable monostandard method. Standardized water samples were used and irradiated in polyethylene ampoules at a neutron flux of 10 13 cm -2 s -1 for periods of 1 minute, 1 and 10 hours. A Ge hyperpure detector was used for your activity determination, with count times of 60, 180, 300 and 600 seconds. The obtained results show than the method can be utilized for the determination of this elements without realize anything previous treatment of the samples. (Author)

  2. High-Sensitivity Fast Neutron Detector KNK-2-8M

    Science.gov (United States)

    Koshelev, A. S.; Dovbysh, L. Ye.; Ovchinnikov, M. A.; Pikulina, G. N.; Drozdov, Yu. M.; Chuklyaev, S. V.; Pepyolyshev, Yu. N.

    2017-12-01

    The design of the fast neutron detector KNK-2-8M is outlined. The results of he detector study in the pulse counting mode with pulses from 238U nuclei fission in the radiator of the neutron-sensitive section and in the current mode with separation of functional section currents are presented. The possibilities of determination of the effective number of 238U nuclei in the radiator of the neutron-sensitive section are considered. The diagnostic capabilities of the detector in the counting mode are demonstrated, as exemplified by the analysis of reference data on characteristics of neutron fields in the BR-1 reactor hall. The diagnostic capabilities of the detector in the current mode are demonstrated, as exemplified by the results of measurements of 238U fission intensity in the power startup of the BR-K1 reactor in the fission pulse generation mode with delayed neutrons and the detector placed in the reactor cavity in conditions of large-scale variation of the reactor radiation fields.

  3. Nondestructive neutron activation analysis of silicon carbide

    Energy Technology Data Exchange (ETDEWEB)

    Vandergraaf, T. T.; Wikjord, A. G.

    1973-10-15

    Instrumentel neutron activation analysis was used to determine trace constituents in silicon carbide. Four commercial powders of different origin, an NBS reference material, and a single crystal were characterized. A total of 36 activation species were identified nondestructively by high resolution gamma spectrometry; quantitative results are given for 12 of the more predominant elements. The limitations of the method for certain elements are discussed. Consideration is given to the depression of the neutron flux by impurities with large neutron absorption cross sections. Radiation fields from the various specimens were estimated assuming all radionuclides have reached their saturation activities. (auth)

  4. A two-solar-mass neutron star measured using Shapiro delay

    NARCIS (Netherlands)

    Demorest, P.B.; Pennucci, T.; Ransom, S.M.; Roberts, M.S.E.; Hessels, J.W.T.

    2010-01-01

    Neutron stars are composed of the densest form of matter known to exist in our Universe, the composition and properties of which are still theoretically uncertain. Measurements of the masses or radii of these objects can strongly constrain the neutron star matter equation of state and rule out

  5. NGI-9 pulsed neutron generator with a fluence to 1010 n/s

    International Nuclear Information System (INIS)

    Allakhverdov, A.Sh.; Ogarkin, V.I.; Silicheva, G.P.; Timofeev, Yu.I.

    1975-01-01

    A neutron pulse generator with 14 MeV energy used for the activation analysis, is described. Its functional diagram and the technical characteristics are presented. The studies of the generator that resulted in determination of the effect of the accelerating voltage amplitude, the delay in the ion source firing with respect to the pulse of the accelerating voltage, the amount of operating ion sources and the energy imparted to them on the neutron flux magnitude are conducted. It is confirmed by the studies that the neutron generator operating in the nominal regime makes it possible to obtain a neutron flux of 5x10 9 -10 10 neutr./s. The dependence of the neutron flux variation on the frequency of pulse sequence for various ion sources is shown

  6. Automated absolute activation analysis with californium-252 sources

    International Nuclear Information System (INIS)

    MacMurdo, K.W.; Bowman, W.W.

    1978-09-01

    A 100-mg 252 Cf neutron activation analysis facility is used routinely at the Savannah River Laboratory for multielement analysis of many solid and liquid samples. An absolute analysis technique converts counting data directly to elemental concentration without the use of classical comparative standards and flux monitors. With the totally automated pneumatic sample transfer system, cyclic irradiation-decay-count regimes can be pre-selected for up to 40 samples, and samples can be analyzed with the facility unattended. An automatic data control system starts and stops a high-resolution gamma-ray spectrometer and/or a delayed-neutron detector; the system also stores data and controls output modes. Gamma ray data are reduced by three main programs in the IBM 360/195 computer: the 4096-channel spectrum and pertinent experimental timing, counting, and sample data are stored on magnetic tape; the spectrum is then reduced to a list of significant photopeak energies, integrated areas, and their associated statistical errors; and the third program assigns gamma ray photopeaks to the appropriate neutron activation product(s) by comparing photopeak energies to tabulated gamma ray energies. Photopeak areas are then converted to elemental concentration by using experimental timing and sample data, calculated elemental neutron capture rates, absolute detector efficiencies, and absolute spectroscopic decay data. Calculational procedures have been developed so that fissile material can be analyzed by cyclic neutron activation and delayed-neutron counting procedures. These calculations are based on a 6 half-life group model of delayed neutron emission; calculations include corrections for delayed neutron interference from 17 O. Detection sensitivities of 239 Pu were demonstrated with 15-g samples at a throughput of up to 140 per day. Over 40 elements can be detected at the sub-ppM level

  7. Computed image analysis of neutron radiographs

    International Nuclear Information System (INIS)

    Dinca, M.; Anghel, E.; Preda, M.; Pavelescu, M.

    2008-01-01

    Similar with X-radiography, using neutron like penetrating particle, there is in practice a nondestructive technique named neutron radiology. When the registration of information is done on a film with the help of a conversion foil (with high cross section for neutrons) that emits secondary radiation (β,γ) that creates a latent image, the technique is named neutron radiography. A radiographic industrial film that contains the image of the internal structure of an object, obtained by neutron radiography, must be subsequently analyzed to obtain qualitative and quantitative information about the structural integrity of that object. There is possible to do a computed analysis of a film using a facility with next main components: an illuminator for film, a CCD video camera and a computer (PC) with suitable software. The qualitative analysis intends to put in evidence possibly anomalies of the structure due to manufacturing processes or induced by working processes (for example, the irradiation activity in the case of the nuclear fuel). The quantitative determination is based on measurements of some image parameters: dimensions, optical densities. The illuminator has been built specially to perform this application but can be used for simple visual observation. The illuminated area is 9x40 cm. The frame of the system is a comparer of Abbe Carl Zeiss Jena type, which has been adapted to achieve this application. The video camera assures the capture of image that is stored and processed by computer. A special program SIMAG-NG has been developed at INR Pitesti that beside of the program SMTV II of the special acquisition module SM 5010 can analyze the images of a film. The major application of the system was the quantitative analysis of a film that contains the images of some nuclear fuel pins beside a dimensional standard. The system was used to measure the length of the pellets of the TRIGA nuclear fuel. (authors)

  8. Calculation of Beta Decay Half-Lives and Delayed Neutron Branching Ratio of Fission Fragments with Skyrme-QRPA

    Directory of Open Access Journals (Sweden)

    Minato Futoshi

    2016-01-01

    Full Text Available Nuclear β-decay and delayed neutron (DN emission is important for the r-process nucleosynthesis after the freeze-out, and stable and safe operation of nuclear reactors. Even though radioactive beam facilities have enabled us to measure β-decay and branching ratio of neutron-rich nuclei apart from the stability line in the nuclear chart, there are still a lot of nuclei which one cannot investigate experimentally. In particular, information on DN is rather scarce than that of T1/2. To predict T1/2 and the branching ratios of DN for next JENDL decay data, we have developed a method which comprises the quasiparticle-random-phase-approximation (QRPA and the Hauser-Feshbach statistical model (HFSM. In this work, we calculate fission fragments with T1/2 ≤ 50 sec. We obtain the rms deviation from experimental half-life of 3:71. Although the result is still worse than GT2 which has been adopted in JENDL decay data, DN spectra are newly calculated. We also discuss further subjects to be done in future for improving the present approach and making next generation of JENDL decay data.

  9. The multielement potential of fast neutron cyclic activation analysis

    International Nuclear Information System (INIS)

    Nonie, S.E.; Randle, K.

    1994-01-01

    Cyclic neutron activation analysis (CNAA) has, in recent years been developed as a useful analytical tool for the assay of short-lived isotopes in single element situations. The work described in this paper investigates the potential of the technique for composite samples having a wide range of elements that produce short-lived and long-lived isotopes on neutron irradiation. Accelerator-derived neutrons with average energies of 3 MeV, 6 MeV and 14 MeV were employed in what has been dubbed 'Fast Neutron Cyclic Neutron Activation Analysis' (FNCAA). The approach to multi-element analysis entailed: determination of cycle parameters in single element samples via the reactions 27 Al(n,p) 27 Mg(9.6 min,E γ =840keV), and 137 Ba(n,n 'γ137m Ba(2.3 min,E γ 137m Ba(2.3 min,E γ =662 keV), a test of the method on a composite rock sample, determination of analytical sensitivities using both powdered kale and rock standards and a comparison of analytical results with other techniques. The results obtained in all these measurements are presented and discussed. (author) 10 refs.; 3 figs.; 5 tabs

  10. Neutron activation analysis: Modelling studies to improve the neutron flux of Americium–Beryllium source

    Directory of Open Access Journals (Sweden)

    Abdessamad Didi

    2017-06-01

    Full Text Available Americium–beryllium (Am-Be; n, γ is a neutron emitting source used in various research fields such as chemistry, physics, geology, archaeology, medicine, and environmental monitoring, as well as in the forensic sciences. It is a mobile source of neutron activity (20 Ci, yielding a small thermal neutron flux that is water moderated. The aim of this study is to develop a model to increase the neutron thermal flux of a source such as Am-Be. This study achieved multiple advantageous results: primarily, it will help us perform neutron activation analysis. Next, it will give us the opportunity to produce radio-elements with short half-lives. Am-Be single and multisource (5 sources experiments were performed within an irradiation facility with a paraffin moderator. The resulting models mainly increase the thermal neutron flux compared to the traditional method with water moderator.

  11. A new facility for rapid neutron activation analysis

    International Nuclear Information System (INIS)

    Zeisler, R.; Makarewicz, M.; Grass, F.; Casta, J.

    1996-01-01

    Many research groups have undertaken efforts on the utilization of short-lived nuclides in a broad spectrum of neutron activation analysis (NAA) applications. The advantages of these approaches are obvious because the information on the sample can be extracted more rapidly. In addition to its other advantages, NAA can become extremely competitive in price and analysis time. Nevertheless, NAA with short-lived nuclides has not gained broad popularity, perhaps because of some difficulties in accuracy and the availability of suitable irradiation facilities. This report discusses the ASTRA reactor for neutron activation analysis capabilities

  12. Neutron activation analysis for antimetabolites. [in food samples

    Science.gov (United States)

    1973-01-01

    Determination of metal ion contaminants in food samples is studied. A weighed quantity of each sample was digested in a concentrated mixture of nitric, hydrochloric and perchloric acids to affect complete solution of the food products. The samples were diluted with water and the pH adjusted according to the specific analysis performed. The samples were analyzed by neutron activation analysis, polarography, and atomic absorption spectrophotometry. The solid food samples were also analyzed by neutron activation analysis for increased sensitivity and lower levels of detectability. The results are presented in tabular form.

  13. Analysis of an Nth-order nonlinear differential-delay equation

    Science.gov (United States)

    Vallée, Réal; Marriott, Christopher

    1989-01-01

    The problem of a nonlinear dynamical system with delay and an overall response time which is distributed among N individual components is analyzed. Such a system can generally be modeled by an Nth-order nonlinear differential delay equation. A linear-stability analysis as well as a numerical simulation of that equation are performed and a comparison is made with the experimental results. Finally, a parallel is established between the first-order differential equation with delay and the Nth-order differential equation without delay.

  14. Optimising polarised neutron scattering measurements--XYZ and polarimetry analysis

    International Nuclear Information System (INIS)

    Cussen, L.D.; Goossens, D.J.

    2002-01-01

    The analytic optimisation of neutron scattering measurements made using XYZ polarisation analysis and neutron polarimetry techniques is discussed. Expressions for the 'quality factor' and the optimum division of counting time for the XYZ technique are presented. For neutron polarimetry the optimisation is identified as analogous to that for measuring the flipping ratio and reference is made to the results already in the literature

  15. Optimising polarised neutron scattering measurements--XYZ and polarimetry analysis

    CERN Document Server

    Cussen, L D

    2002-01-01

    The analytic optimisation of neutron scattering measurements made using XYZ polarisation analysis and neutron polarimetry techniques is discussed. Expressions for the 'quality factor' and the optimum division of counting time for the XYZ technique are presented. For neutron polarimetry the optimisation is identified as analogous to that for measuring the flipping ratio and reference is made to the results already in the literature.

  16. Fast neutron activation analysis of Kalewa (Myanmar) coal

    Energy Technology Data Exchange (ETDEWEB)

    Myint, U; Naing, W [Yangon Univ. (Myanmar). Dept. of Chemistry

    1994-06-01

    Aluminium, silicon, copper, iron, magnesium and sulfur in Kalewa (Myanmar) coal were determined by fast neutron activation analysis. For activation a KAMAN A-710 Neutron Generator was used. Kalewa coal was found to be low in sulfur and relatively rich in iron. (author) 2 refs.; 1 fig.; 1 tab.

  17. Fast neutron activation analysis of Kalewa (Myanmar) coal

    International Nuclear Information System (INIS)

    Myint, U.; Naing, W.

    1994-01-01

    Aluminium, silicon, copper, iron, magnesium and sulfur in Kalewa (Myanmar) coal were determined by fast neutron activation analysis. For activation a KAMAN A-710 Neutron Generator was used. Kalewa coal was found to be low in sulfur and relatively rich in iron. (author) 2 refs.; 1 fig.; 1 tab

  18. Partial neutron capture cross sections of actinides using cold neutron prompt gamma activation analysis

    International Nuclear Information System (INIS)

    Genreith, Christoph

    2015-01-01

    Nuclear waste needs to be characterized for its safe handling and storage. In particular long-lived actinides render the waste characterization challenging. The results described in this thesis demonstrate that Prompt Gamma Neutron Activation Analysis (PGAA) with cold neutrons is a reliable tool for the non-destructive analysis of actinides. Nuclear data required for an accurate identification and quantification of actinides was acquired. Therefore, a sample design suitable for accurate and precise measurements of prompt γ-ray energies and partial cross sections of long-lived actinides at existing PGAA facilities was presented. Using the developed sample design the fundamental prompt γ-ray data on 237 Np, 241 Am and 242 Pu were measured. The data were validated by repetitive analysis of different samples at two individual irradiation and counting facilities - the BRR in Budapest and the FRM II in Garching near Munich. Employing cold neutrons, resonance neutron capture by low energetic resonances was avoided during the experiments. This is an improvement over older neutron activation based works at thermal reactor neutron energies. 152 prompt γ-rays of 237 Np were identified, as well as 19 of 241 Am, and 127 prompt γ-rays of 242 Pu. In all cases, both high and lower energetic prompt γ-rays were identified. The most intense line of 237 Np was observed at an energy of E γ =182.82(10) keV associated with a partial capture cross section of σ γ =22.06(39) b. The most intense prompt γ-ray lines of 241 Am and of 242 Pu were observed at E γ =154.72(7) keV with σ γ =72.80(252) b and E γ =287.69(8) keV with σ γ =7.07(12) b, respectively. The measurements described in this thesis provide the first reported quantifications on partial radiative capture cross sections for 237 Np, 241 Am and 242 Pu measured simultaneously over the large energy range from 45 keV to 12 MeV. Detailed uncertainty assessments were performed and the validity of the given uncertainties was

  19. Self powered neutron detectors

    International Nuclear Information System (INIS)

    Passe, J.; Petitcolas, H.; Verdant, R.

    1975-01-01

    The self-powered neutron detectors (SPND) enable to measure continuously high fluxes of thermal neutrons. They are particularly suitable for power reactor cores because of their robustness. Description of two kinds of SPND's characterized by the electrical current production way is given here: the first SPND's which present a V, Ag or Rh emitter are sensitive enough but they offer a few minute delay time: the second SPND's which are depending on the gamma activation have a short delay time. The emitter is made of Co or Pt. In any case, the signal is linear with reaction rates. Finally, the applications are briefly repeated here: irradiation facility monitor in research reactors, and flux map and space instability control in power reactors [fr

  20. Prompt-gamma neutron activation analysis system design: Effects of D-T versus D-D neutron generator source selection

    Science.gov (United States)

    Prompt-gamma neutron activation (PGNA) analysis is used for the non-invasive measurement of human body composition. Advancements in portable, compact neutron generator design have made those devices attractive as neutron sources. Two distinct generators are available: D-D with 2.5 MeV and D-T with...

  1. Time-of-flight and vector polarization analysis for diffuse neutron scattering

    International Nuclear Information System (INIS)

    Schweika, W.

    2003-01-01

    The potential of pulsed neutron sources for diffuse scattering including time-of-flight (TOF) and polarization analysis is discussed in comparison to the capabilities of the present instrument diffuse neutron scattering at the research center Juelich. We present first results of a new method for full polarization analysis using precessing neutron polarization. A proposal is made for a new type of instrument at pulsed sources, which allows for vector polarization analysis in TOF instruments with multi-detectors

  2. Real time neutron flux monitoring using Rh self powered neutron detector

    Energy Technology Data Exchange (ETDEWEB)

    Juna, Byung Jin; Lee, Byung Chul; Park, Sang Jun; Jung, Hoan Sung [KAERI, Daejeon (Korea, Republic of)

    2012-10-15

    Rhodium (Rh) self powered neutron detectors (SPNDs) are widely used for on line monitoring of local neutron flux. Its signal is slower than the actual variation of neutron flux owing to a delayed {beta} decay of the Rh activation product, but real time monitoring is possible by solving equations between the neutron reaction rate in the detector and its signal. While the measuring system is highly reliable, the accuracy depends on the method solving the equations and accuracy of the parameters in the equations. The uncertain parameters are the contribution of gamma rays to the signal, and the branching ratios of Rh 104 and Rh 104m after the neutron absorption of Rh 103. Real time neutron flux monitoring using Rh SPNDs has been quite successful for neutron transmutation doping (NTD) at HANARO. We revisited the initial data used for the verification of a real time monitoring system, to refine algorithm for a better solution and to check the parameters for correctness. As a result, we suggest an effective way to determine the prompt parameter.

  3. Real time neutron flux monitoring using Rh self powered neutron detector

    International Nuclear Information System (INIS)

    Juna, Byung Jin; Lee, Byung Chul; Park, Sang Jun; Jung, Hoan Sung

    2012-01-01

    Rhodium (Rh) self powered neutron detectors (SPNDs) are widely used for on line monitoring of local neutron flux. Its signal is slower than the actual variation of neutron flux owing to a delayed β decay of the Rh activation product, but real time monitoring is possible by solving equations between the neutron reaction rate in the detector and its signal. While the measuring system is highly reliable, the accuracy depends on the method solving the equations and accuracy of the parameters in the equations. The uncertain parameters are the contribution of gamma rays to the signal, and the branching ratios of Rh 104 and Rh 104m after the neutron absorption of Rh 103. Real time neutron flux monitoring using Rh SPNDs has been quite successful for neutron transmutation doping (NTD) at HANARO. We revisited the initial data used for the verification of a real time monitoring system, to refine algorithm for a better solution and to check the parameters for correctness. As a result, we suggest an effective way to determine the prompt parameter

  4. Stability analysis for cellular neural networks with variable delays

    International Nuclear Information System (INIS)

    Zhang Qiang; Wei Xiaopeng; Xu Jin

    2006-01-01

    Some sufficient conditions for the global exponential stability of cellular neural networks with variable delay are obtained by means of a method based on delay differential inequality. The method, which does not make use of Lyapunov functionals, is simple and effective for the stability analysis of neural networks with delay. Some previously established results in the literature are shown to be special cases of the presented result

  5. Characters of neutron noise in full-size molten salt reactor

    International Nuclear Information System (INIS)

    Wang, Jiangmeng; Cao, Xinrong

    2015-01-01

    Highlights: • The larger system size makes full-size MSR deviate from point kinetic behavior. • The increasing velocity has non-monotonic effect on the effective delayed neutron fraction. • The amplitude of Green’s function at low frequencies is inversely proportional to the effective delayed neutron fraction. • The range of plateau region is smaller due to the more prominent point kinetic effect. - Abstract: In the present paper, the frequency-dependent and space-dependent behavior of the neutron noise in a full-size Molten Salt Reactor (MSR) is investigated. The theoretical models considering the fuel circulation are established based on one-group neutron diffusion theory. Green’s function of the neutron noise induced by a propagating perturbation is introduced with linear noise theory, due to the small perturbation. The equations are numerically calculated by developing a code, in which the eigenfunction expansion method is adopted. The static results show that the effective delayed neutron fraction changes non-monotonically with the increasing fuel velocity. In the dynamic case, the results are compared to those obtained in 1D MSR and a traditional reactor, in order to figure out the effects of both the fuel circulation and the system size. It is found that there is no difference in 1D and 3D MSR systems from the view of fuel circulation, i.e., the fuel circulation enhances the spatial neutronic coupling and leads to the stronger point kinetic effect. The more prominent space-dependent effect founded in 3D traditional reactors is also observed in the MSR, due to the looser neutronic coupling and the unique singularity of Green’s function in the location of the perturbation. Another interesting finding is that Green’s function for low frequencies changes non-monotonically with increasing velocity. The largest magnitude of Green’s function is observed at the velocity where the effective delayed neutron fraction reaches its minimum. Finally, the

  6. Automatization of the neutron activation analysis method in the nuclear analysis laboratory

    International Nuclear Information System (INIS)

    Gonzalez, N.R.; Rivero, D del C.; Gonzalez, M.A.; Larramendi, F.

    1993-01-01

    In the present paper the work done to automatice the Neutron Activation Analysis technic with a neutron generator is described. An interface between an IBM compatible microcomputer and the equipment in use to make this kind of measurement was developed. including the specialized software for this system

  7. Neutron Generators for Spent Fuel Assay

    International Nuclear Information System (INIS)

    Ludewigt, Bernhard A.

    2010-01-01

    The Next Generation Safeguards Initiative (NGSI) of the U.S. DOE has initiated a multi-lab/university collaboration to quantify the plutonium (Pu) mass in, and detect the diversion of pins from, spent nuclear fuel (SNF) assemblies with non-destructive assay (NDA). The 14 NDA techniques being studied include several that require an external neutron source: Delayed Neutrons (DN), Differential Die-Away (DDA), Delayed Gammas (DG), and Lead Slowing-Down Spectroscopy (LSDS). This report provides a survey of currently available neutron sources and their underlying technology that may be suitable for NDA of SNF assemblies. The neutron sources considered here fall into two broad categories. The term 'neutron generator' is commonly used for sealed devices that operate at relatively low acceleration voltages of less than 150 kV. Systems that employ an acceleration structure to produce ion beam energies from hundreds of keV to several MeV, and that are pumped down to vacuum during operation, rather than being sealed units, are usually referred to as 'accelerator-driven neutron sources.' Currently available neutron sources and future options are evaluated within the parameter space of the neutron generator/source requirements as currently understood and summarized in section 2. Applicable neutron source technologies are described in section 3. Commercially available neutron generators and other source options that could be made available in the near future with some further development and customization are discussed in sections 4 and 5, respectively. The pros and cons of the various options and possible ways forward are discussed in section 6. Selection of the best approach must take a number of parameters into account including cost, size, lifetime, and power consumption, as well as neutron flux, neutron energy spectrum, and pulse structure that satisfy the requirements of the NDA instrument to be built.

  8. Neutron activation analysis (NAA), radioisotope production via neutron activation (PNA) and fission product gas-jet (GJA)

    Energy Technology Data Exchange (ETDEWEB)

    Gaeggeler, H W [Paul Scherrer Inst. (PSI), Villigen (Switzerland)

    1996-11-01

    Three different non-diffractive applications of neutrons are outlined, neutron activation analysis, production of radionuclides, mostly for medical applications, and production of short-lived fission nuclides with a so-called gas-jet. It is shown that all three devices may be incorporated into one single insert at SINQ due to their different requests with respect to thermal neutron flux. Some applications of these three facilities are summarized. (author) 3 figs., 1 tab., 8 refs.

  9. Neutron activation analysis (NAA), radioisotope production via neutron activation (PNA) and fission product gas-jet (GJA)

    International Nuclear Information System (INIS)

    Gaeggeler, H.W.

    1996-01-01

    Three different non-diffractive applications of neutrons are outlined, neutron activation analysis, production of radionuclides, mostly for medical applications, and production of short-lived fission nuclides with a so-called gas-jet. It is shown that all three devices may be incorporated into one single insert at SINQ due to their different requests with respect to thermal neutron flux. Some applications of these three facilities are summarized. (author) 3 figs., 1 tab., 8 refs

  10. Delayed neutron detection in canning burst detection studies (1961); Etude sur la detection des neutrons differes en vue de la detection des ruptures de gaines (1961)

    Energy Technology Data Exchange (ETDEWEB)

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

    1961-07-01

    This paper describes a theoretical and experimental study on the detection of neutrons present in the primary cooling circuit of a reactor cooled by heavy or light water, with a view to the installation of a canning burst detection unit. The concentration of background neutrons is first calculated, taking into account the neutrons from nitrogen 17 decay, and the photoneutrons produced by the decay of nitrogen 16 and sodium 24. The emission of delayed fission neutrons, originating at a given crack in the canning, has been estimated. Using the D{sub 2}O circuit of the pile EL-3, three units have been developed by means of which the following three types of detector may be compared: 1) BF{sub 3} proportional counter 2) Boron scintillator 3) Fission chamber Under the present experimental conditions the BF{sub 3} counter gave the best results. The influence on these detectors of the {gamma} flux, which in certain cases reaches 200 R/h, is analysed. Finally a calibration is carried out with an experimental crack of 30 mm{sup 2} of uranium exposed to a flux of 5.8 x 10{sup 13} n.cm{sup -2}.s{sup -1}. The sensitivity obtained with the BF{sub 3} counter during this test is 2 counts/s per mm{sup 2} of exposed uranium. (author) [French] Le present rapport est une etude theorique et experimentale sur la detection des neutrons presents dans le circuit primaire de refroidissement d'un reacteur refrigere par l'eau lourde ou l'eau legere, en vue d'une installation de detection de ruptures de gaines. On fait d'abord un calcul sur la concentration des neutrons de bruit de fond en tenant compte: des neutrons de decroissance de l'azote 17 et des photoneutrons produits par les decroissances de l'azote 16 et du sodium 24. L'emission des neutrons differes de fission, qui ont pour origine une fissure de gaine donnee, a ete evaluee. Utilisant le circuit D{sub 2}O de la pile EL3, trois installations ont ete mises au point permettant de comparer les trois types de detecteurs suivants: 1

  11. Neutron activation analysis for environmental sample in Thailand

    International Nuclear Information System (INIS)

    Busamongkol, Arporn; Nouchpramool, Sunun; Bunprapob, Supamatthree; Sumitra, Tatchai

    2003-01-01

    Neutron Activation Analysis has been applied for the trace elements analysis in environmental samples. Thirty three samples of airborne particulate were collected every week at Ongkharak Nuclear Research Center (ONRC) during the period of June 1998 to March 1999. The Ti, I, Mg, Na, V, K, Cl, Al, Mn, Ca, As, Sm, Sb, Br, La, Ce, Th, Cr, Cs, Sc, Rb, Fe, Zn and Co were analyzed by Neutron Activation Analysis utilizing 2 MW TRIGA MARK III research reactor. The certified reference materials 1632a and 1633a from National Bureau of Standard were select as standard. (author)

  12. Studies on the molten salt reactor. Code development and neutronics analysis of MSRE-type design

    International Nuclear Information System (INIS)

    Zhuang Kun; Cao Liangzhi; Zheng Youqi; Wu Hongchun

    2015-01-01

    The molten salt reactor is characterized by its use of the fluid-fuel, which serves both as a fuel and as a coolant simultaneously. The position of delayed neutron precursors continuously changes both in the core and in the external loop due to the fuel circulation, and the fission products are extracted by an online fuel reprocessing unit, which all lead to the modeling methods for the conventional reactors using solid fuel not applicable. This study establishes suitable calculation models for the neutronics analysis of the molten salt reactor and develops a new code named MOREL based on the three-dimensional diffusion steady and transient calculations. Some numerical tests are chosen to verify the code and the numerical results indicate that MOREL can be used for the analysis of the molten salt reactor. After verification, it is applied to analyze the characteristics of a typical molten salt reactor, including the steady characteristics, the influence of fuel circulation on the kinetic behaviors. Besides, the influence of online fuel reprocessing simulation is also examined. The results show that inherent safety is the character of the molten salt reactor from the aspect of reactivity feedback and the fuel circulation has great influence on the kinetic characteristics of molten salt reactor. (author)

  13. Utilization of neutrons in nuclear data measurements and bulk sample analysis

    International Nuclear Information System (INIS)

    Jonah, S. A.

    1995-01-01

    Experimental investigations were carried out with neutrons in the fields of neutron data measurements and bulk sample analysis based on the interactions of neutron interactions required in the investigations together with some salient features of the sources employed are enumerated. Excitation cross section curves and isomeric cross section ratio of 58 Ni(n,p) 58 Co m , g reaction over the neutron energy range of between 5 and 15 MeV were determined using the activation analysis technique in combination with high-resolution gamma spectroscopy. Characteristics of the incident neutrons produced via the D-T reaction of a neutron generator and D-D reaction of a cyclotron were determined experimentally to account for the contributing effects of background neutrons especially in the 5-13 MeV neutron energy range where existing data are scanty and rather discrepant. The measured data agree well with calculated data using nuclear models but deviate significantly from the recommended data based on existing literature data. The measured δ act and δ m /δ g data made it possible to determine the cross section curve for 58 Ni(n,p) 58 Co m reaction. Furthermore the flux density distributions of thermal and primary fast neutrons in different configurations of bulk samples consisting of water, graphite and coal together with the attenuation characteristics were determined by the activation analysis and pulse height response spectrometry techniques. From the results obtained, an experimental geometry has been proposed for on-line elemental analysis of coal and other minerals. Similarly the total hydrogen content and 0+C/H atomic ratio in household and motor oils as well as crude oil samples of different origins were measured by an improved experimental arrangement based on the thermal neutron reflection technique. A detection limit of 0.12 w % was obtained for hydrogen indicating the possible adaptation of this technique for quality control of petroleum products

  14. Neutron activation analysis-comparative (NAAC)

    International Nuclear Information System (INIS)

    Zimmer, W.H.

    1979-01-01

    A software system for the reduction of comparative neutron activation analysis data is presented. Libraries are constructed to contain the elemental composition and isotopic nuclear data of an unlimited number of standards. Ratios to unknown sample data are performed by standard calibrations. Interfering peak corrections, second-order activation-product corrections, and deconvolution of multiplets are applied automatically. Passive gamma-energy analysis can be performed with the same software. 3 figures

  15. Radioactive waste characterisation by neutron activation

    International Nuclear Information System (INIS)

    Nicol, Tangi

    2016-01-01

    Nuclear activities produce radioactive wastes classified following their radioactive level and decay time. an accurate characterization is necessary for efficient classification and management. Medium and high level wastes containing long lived radioactive isotopes will be stored in deep geological storage for hundreds of thousands years. at the end of this period, it is essential to ensure that the wastes do not represent any risk for humans and environment, not only from radioactive point of view, but also from stable toxic chemicals. This PhD thesis concerns the characterization of toxic chemicals and nuclear material in radioactive waste, by using neutron activation analysis, in the frame of collaboration between the Nuclear Measurement Laboratory of CEA Cadarache, France, and the Institute of Nuclear Waste Management and Reactor Safety of the research center, FZJ (Forschungszentrum Juelich GmbH), Germany. The first study is about the validation of the numerical model of the neutron activation cell MEDINA (FZJ), using MCNP Monte Carlo transport code. Simulations and measurements of prompt capture gamma rays from small samples measured in MEDINA have been compared for a number of elements of interest (beryllium, aluminum, chlorine, copper, selenium, strontium, and tantalum). The comparison was performed using different nuclear databases, resulting in satisfactory agreement and validating simulation in view of following studies. Then, the feasibility of fission delayed gamma-ray measurements of "2"3"9Pu and "2"3"5U in 225 L waste drums has been studied, considering bituminized or concrete matrixes representative of wastes produced in France and Germany. The delayed gamma emission yields were first determined from uranium and plutonium metallic samples measurements in REGAIN, the neutron activation cell of LMN, showing satisfactory consistency with published data. The useful delayed gamma signals of "2"3"9Pu and "2"3"5U, homogeneously distributed in the 225 L

  16. Stability analysis of linear switching systems with time delays

    International Nuclear Information System (INIS)

    Li Ping; Zhong Shouming; Cui Jinzhong

    2009-01-01

    The issue of stability analysis of linear switching system with discrete and distributed time delays is studied in this paper. An appropriate switching rule is applied to guarantee the stability of the whole switching system. Our results use a Riccati-type Lyapunov functional under a condition on the time delay. So, switching systems with mixed delays are developed. A numerical example is given to illustrate the effectiveness of our results.

  17. Analysis of neutron flux measurement systems using statistical functions

    International Nuclear Information System (INIS)

    Pontes, Eduardo Winston

    1997-01-01

    This work develops an integrated analysis for neutron flux measurement systems using the concepts of cumulants and spectra. Its major contribution is the generalization of Campbell's theorem in the form of spectra in the frequency domain, and its application to the analysis of neutron flux measurement systems. Campbell's theorem, in its generalized form, constitutes an important tool, not only to find the nth-order frequency spectra of the radiation detector, but also in the system analysis. The radiation detector, an ionization chamber for neutrons, is modeled for cylindrical, plane and spherical geometries. The detector current pulses are characterized by a vector of random parameters, and the associated charges, statistical moments and frequency spectra of the resulting current are calculated. A computer program is developed for application of the proposed methodology. In order for the analysis to integrate the associated electronics, the signal processor is studied, considering analog and digital configurations. The analysis is unified by developing the concept of equivalent systems that can be used to describe the cumulants and spectra in analog or digital systems. The noise in the signal processor input stage is analysed in terms of second order spectrum. Mathematical expressions are presented for cumulants and spectra up to fourth order, for important cases of filter positioning relative to detector spectra. Unbiased conventional estimators for cumulants are used, and, to evaluate systems precision and response time, expressions are developed for their variances. Finally, some possibilities for obtaining neutron radiation flux as a function of cumulants are discussed. In summary, this work proposes some analysis tools which make possible important decisions in the design of better neutron flux measurement systems. (author)

  18. Neutron activation analysis for certification of standard reference materials

    International Nuclear Information System (INIS)

    Capote Rodriguez, G.; Perez Zayas, G.; Hernandez Rivero, A.; Ribeiro Guevara, S.

    1996-01-01

    Neutron activation analysis is used extensively as one of the analytical techniques in the certification of standard reference materials. Characteristics of neutron activation analysis which make it valuable in this role are: accuracy multielemental capability to asses homogeneity, high sensitivity for many elements, and essentially non-destructive method. This paper report the concentrations of 30 elements (major, minor and trace elements) in four Cuban samples. The samples were irradiated in a thermal neutron flux of 10 12- 10 13 n.cm 2. s -1. The gamma ray spectra were measured by HPGe detectors and were analyzed using ACTAN program development in Center of Applied Studies for Nuclear Development

  19. The comparison of four neutron sources for Prompt Gamma Neutron Activation Analysis (PGNAA) in vivo detections of boron.

    Science.gov (United States)

    Fantidis, J G; Nicolaou, G E; Potolias, C; Vordos, N; Bandekas, D V

    A Prompt Gamma Ray Neutron Activation Analysis (PGNAA) system, incorporating an isotopic neutron source has been simulated using the MCNPX Monte Carlo code. In order to improve the signal to noise ratio different collimators and a filter were placed between the neutron source and the object. The effect of the positioning of the neutron beam and the detector relative to the object has been studied. In this work the optimisation procedure is demonstrated for boron. Monte Carlo calculations were carried out to compare the performance of the proposed PGNAA system using four different neutron sources ( 241 Am/Be, 252 Cf, 241 Am/B, and DT neutron generator). Among the different systems the 252 Cf neutron based PGNAA system has the best performance.

  20. Three dimensional neutronic/thermal-hydraulic coupled simulation of MSR in transient state condition

    International Nuclear Information System (INIS)

    Zhou, Jianjun; Zhang, Daling; Qiu, Suizheng; Su, Guanghui; Tian, Wenxi; Wu, Yingwei

    2015-01-01

    Highlights: • Developed a three dimensional neutronic/thermal-hydraulic coupled transient analysis code for MSR. • Investigated the neutron distribution and thermal-hydraulic characters of the core under transient condition. • Analyzed three different transient conditions of inlet temperature drop, reactivity jump and pump coastdown. - Abstract: MSR (molten salt reactor) use liquid molten salt as coolant and fuel solvent, which was the only one liquid reactor of six Generation IV reactor types. As a liquid reactor the physical property of reactor was significantly influenced by fuel salt flow and the conventional analysis methods applied in solid fuel reactors are not applicable for this type of reactors. The present work developed a three dimensional neutronic/thermal-hydraulic coupled code investigated the neutronics and thermo-hydraulics characteristics of the core in transient condition based on neutron diffusion theory and numerical heat transfer. The code consists of two group neutron diffusion equations for fast and thermal neutron fluxes and six group balance equations for delayed neutron precursors. The code was separately validated by neutron benchmark and flow and heat transfer benchmark. Three different transient conditions was analyzed with inlet temperature drop, reactivity jump and pump coastdown. The results provide some valuable information in design and research this kind of reactor

  1. Three dimensional neutronic/thermal-hydraulic coupled simulation of MSR in transient state condition

    Energy Technology Data Exchange (ETDEWEB)

    Zhou, Jianjun [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xianning Road, 28, Xi’an 710049, Shaanxi (China); College of Mechanical and Power Engineering, China Three Gorges University, No 8, Daxue road, Yichang, Hubei 443002 (China); Zhang, Daling, E-mail: dlzhang@mail.xjtu.edu.cn [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xianning Road, 28, Xi’an 710049, Shaanxi (China); Qiu, Suizheng; Su, Guanghui; Tian, Wenxi; Wu, Yingwei [School of Nuclear Science and Technology, Xi’an Jiaotong University, Xianning Road, 28, Xi’an 710049, Shaanxi (China)

    2015-02-15

    Highlights: • Developed a three dimensional neutronic/thermal-hydraulic coupled transient analysis code for MSR. • Investigated the neutron distribution and thermal-hydraulic characters of the core under transient condition. • Analyzed three different transient conditions of inlet temperature drop, reactivity jump and pump coastdown. - Abstract: MSR (molten salt reactor) use liquid molten salt as coolant and fuel solvent, which was the only one liquid reactor of six Generation IV reactor types. As a liquid reactor the physical property of reactor was significantly influenced by fuel salt flow and the conventional analysis methods applied in solid fuel reactors are not applicable for this type of reactors. The present work developed a three dimensional neutronic/thermal-hydraulic coupled code investigated the neutronics and thermo-hydraulics characteristics of the core in transient condition based on neutron diffusion theory and numerical heat transfer. The code consists of two group neutron diffusion equations for fast and thermal neutron fluxes and six group balance equations for delayed neutron precursors. The code was separately validated by neutron benchmark and flow and heat transfer benchmark. Three different transient conditions was analyzed with inlet temperature drop, reactivity jump and pump coastdown. The results provide some valuable information in design and research this kind of reactor.

  2. Prompt-gamma neutron activation analysis system design. Effects of D-T versus D-D neutron generator source selection

    International Nuclear Information System (INIS)

    Shypailo, R.J.; Ellis, K.J.

    2008-01-01

    Prompt-gamma neutron activation (PGNA) analysis is used for the non-invasive measurement of human body composition. Advancements in portable, compact neutron generator design have made those devices attractive as neutron sources. Two distinct generators are available: D-D with 2.5 MeV and D-T with 14.2 MeV neutrons. To compare the performance of these two units in our present PGNA system, we performed Monte Carlo simulations (MCNP-5; Los Alamos National Laboratory) evaluating the nitrogen reactions produced in tissue-equivalent phantoms and the effects of background interference on the gamma-detectors. Monte Carlo response curves showed increased gamma production per unit dose when using the D-D generator, suggesting that it is the more suitable choice for smaller sized subjects. The increased penetration by higher energy neutrons produced by the D-T generator supports its utility when examining larger, especially obese, subjects. A clinical PGNA analysis design incorporating both neutron generator options may be the best choice for a system required to measure a wide range of subject phenotypes. (author)

  3. Neutron activation analysis applied to energy and environment

    International Nuclear Information System (INIS)

    Lyon, W.S.

    1975-01-01

    Neutron activation analysis was applied to a number of problems concerned with energy production and the environment. Burning of fossil fuel, the search for new sources of uranium, possible presence of toxic elements in food and water, and the relationship of trace elements to cardiovascular disease are some of the problems in which neutron activation was used. (auth)

  4. Analysis and control of issues that delay pharmaceutical projects

    Directory of Open Access Journals (Sweden)

    Nallam Sai Nandeswara Rao

    2015-10-01

    Full Text Available Every project will have certain objectives and service levels to be achieved. The success of a project depends on several dimensions like time, cost/budget, quality, etc. and managing a project involves completing the project within time, within budget and with quality to satisfy the users. Because of the significance of health, pharmaceutical companies realized the importance of project management methods and techniques to make available the life saving drugs in time to the needy patients and hospitals. In literature, there is meager information about pharmaceutical project management oriented towards analysis of issues and factors that contribute to the failure or success of projects. This study attempts to analyse different issues that contribute to time delays in pharmaceutical product-based projects, group them under a finite set of prominent factors and identify remedial measures to control those delays. The feedback of project people of some big pharmaceutical firms of Indian sub-continent was collected for this purpose. Exploratory factor analysis (EFA has been used to reduce the reasons for time delays to a limited number of prominent factors and the EFA model has been further examined by confirmatory factor analysis (CFA for its validation. Remedial measures under each factor of time delays have been gathered and a framework designed to mitigate the time delays in pharmaceutical projects. The derived factors that delay the pharmaceutical projects include resource, monitoring & control, scheduling and planning problems. Important remedial measures like blended resource approach, estimation and forecast of shortage of labour and skills, regular quality training, etc. have been recommended.

  5. Introduction to the problems of uranium analysis

    International Nuclear Information System (INIS)

    Suschny, O.

    1983-08-01

    Brief information is given on different techniques used for recognition, separation and determination of uranium. The analytical methods dealt with are: radiometry, neutron activation analysis, delayed neutron counting, X-ray fluorescence analysis, fluorimetry, spectral photometry, gravimetry, and volumetry. Selection of a suitable method for determining uranium in geochemical surveys is also discussed

  6. Detection of land mines using fast and thermal neutron analysis

    International Nuclear Information System (INIS)

    Bach, P.

    1998-01-01

    The detection of land mines is made possible by using nuclear sensor based on neutron interrogation. Neutron interrogation allows to detect the sensitive elements (C, H, O, N) of the explosives in land mines or in unexploded shells: the evaluation of characteristic ratio N/O and C/O in a volume element gives a signature of high explosives. Fast neutron interrogation has been qualified in our laboratories as a powerful close distance method for identifying the presence of a mine or explosive. This method could be implemented together with a multisensor detection system - for instance IR or microwave - to reduce the false alarm rate by addressing the suspected area. Principle of operation is based on the measurement of gamma rays induced by neutron interaction with irradiated nuclei from the soil and from a possible mine. Specific energy of these gamma rays allows to recognise the elements at the origin of neutron interaction. Several detection methods can be used, depending on nuclei to be identified. Analysis of physical data, computations by simulation codes, and experimentations performed in our laboratory have shown the interest of Fast Neutron Analysis (FNA) combined with Thermal Neutron Analysis (TNA) techniques, especially for detection of nitrogen 14 N, carbon 12 C and oxygen 16 O. The FNA technique can be implemented using a 14 MeV sealed neutron tube, and a set of detectors. The mines detection has been demonstrated from our investigations, using a low power neutron generator working in the 10 8 n/s range, which is reasonable when considering safety rules. A fieldable demonstrator would be made with a detection head including tube and detectors, and with remote electronics, power supplies and computer installed in a vehicle. (author)

  7. Advances in 14 MeV neutron activation analysis by means of a new intense neutron source

    International Nuclear Information System (INIS)

    Pepelnik, R.; Fanger, H.-U.; Michaelis, W.; Anders, B.

    1982-01-01

    A new intense 14 MeV neutron generator with cylindrical acceleration structure has been put in operation at the GKSS Research Center Geesthacht. The sealed neutron tube is combined with a fast pneumatic rabbit system with particular capabilities for neutron activation analysis involving short-lived reaction products. The sample transfer time is less than 140 ms. The maximum neutron flux available for activation is 5.2x10 10 n/cm 2 s. Theoretical sensitivity predictions made in a previous study have been verified for some important trace elements. As a first application, samples of freeze-dried suspended matter and fishes of the Elbe river were analyzed. (author)

  8. Storage and pre-neutron-activation-analysis treatment for trace-element analysis in urine

    International Nuclear Information System (INIS)

    Blotcky, A.J.; Rack, E.P.

    1985-01-01

    The problems regarding storage and pre-neutron-activation-analysis treatment for the elements aluminum, calcium, vanadium, selenium, copper, iodine, zinc, manganese, and magnesium in a urine matrix are reviewed. The type of collection and storage procedure and pre-neutron activation analysis treatment of urine depend on the specific trace element; that is, its inherent physical and chemical properties. Specifically polyethylene in teflon containers are the most suitable for general determinations. Whether any preservative is added would depend upon the stability of the trace element and its tendency for surface adsorption. Preferably, preservatives should contain no radioactivatable elements for maximum efficacy. Freeze drying or packing urine shipments under dry ice needs to be explored on an individual basis. Each pre- or post-neutron activation analysis treatment is specific and optimized for the trace element analyzed

  9. The progress of neutron induced prompt gamma analysis technique in 1988-2002

    International Nuclear Information System (INIS)

    Liu Yuren; Jing Shiwei

    2003-01-01

    The new development of the neutron induced prompt gamma-ray analysis (NIPGA) technology in 1988-2002 are described. The pulse fast-thermal neutron activation analysis method, which utilizes the inelastic reaction and capture reaction jointly is employed to measure the elemental content in the material more efficiently. The lifetime of the neutron generator is more than 10000 h and the capability of HPGe, TeZeCd and MCA (multi-channel analyser) reaches the high level. At the same time, Monte Carlo Library least-square method is used to solve the nonlinearity problem in the PGNAA (Prompt Gamma Neutron Activation Analysis)

  10. Diagnosis of mucoviscidosis by neutron activation analysis. Part 1

    International Nuclear Information System (INIS)

    Bellido, Luis F.; Bellido, Alfredo V.

    1997-02-01

    Symptoms pathology, incidence, and gravity of the inherent syndrome called mucoviscidosis, or cystic fibrosis are described in this Part I. The analytical methods used for its diagnosis, both the conventional chemical ones and by neutron activation analysis are also summarised. Finally, an analytical method to study the incidence of mucoviscidosis in Brazil is presented. This , essentially, consists in bromine determination, in fingernails, by resonance neutron activation analysis. (author)

  11. A Dosimetry Study of Deuterium-Deuterium Neutron Generator-based In Vivo Neutron Activation Analysis.

    Science.gov (United States)

    Sowers, Daniel; Liu, Yingzi; Mostafaei, Farshad; Blake, Scott; Nie, Linda H

    2015-12-01

    A neutron irradiation cavity for in vivo neutron activation analysis (IVNAA) to detect manganese, aluminum, and other potentially toxic elements in human hand bone has been designed and its dosimetric specifications measured. The neutron source is a customized deuterium-deuterium neutron generator that produces neutrons at 2.45 MeV by the fusion reaction 2H(d, n)3He at a calculated flux of 7 × 10(8) ± 30% s(-1). A moderator/reflector/shielding [5 cm high density polyethylene (HDPE), 5.3 cm graphite and 5.7 cm borated (HDPE)] assembly has been designed and built to maximize the thermal neutron flux inside the hand irradiation cavity and to reduce the extremity dose and effective dose to the human subject. Lead sheets are used to attenuate bremsstrahlung x rays and activation gammas. A Monte Carlo simulation (MCNP6) was used to model the system and calculate extremity dose. The extremity dose was measured with neutron and photon sensitive film badges and Fuji electronic pocket dosimeters (EPD). The neutron ambient dose outside the shielding was measured by Fuji NSN3, and the photon dose was measured by a Bicron MicroREM scintillator. Neutron extremity dose was calculated to be 32.3 mSv using MCNP6 simulations given a 10-min IVNAA measurement of manganese. Measurements by EPD and film badge indicate hand dose to be 31.7 ± 0.8 mSv for neutrons and 4.2 ± 0.2 mSv for photons for 10 min; whole body effective dose was calculated conservatively to be 0.052 mSv. Experimental values closely match values obtained from MCNP6 simulations. These are acceptable doses to apply the technology for a manganese toxicity study in a human population.

  12. Instrumental neutron activation analysis - a routine method

    International Nuclear Information System (INIS)

    Bruin, M. de.

    1983-01-01

    This thesis describes the way in which at IRI instrumental neutron activation analysis (INAA) has been developed into an automated system for routine analysis. The basis of this work are 20 publications describing the development of INAA since 1968. (Auth.)

  13. Elementary calculation of the shutdown delay of a pile

    International Nuclear Information System (INIS)

    Yvon, J.

    1949-04-01

    This study analyzes theoretically the progress of the shutdown of a nuclear pile (reactor) when a cadmium rod is introduced instantaneously. For simplification reasons, the environment of the pile is considered as homogenous and only thermal neutrons are considered (delayed neutrons are neglected). Calculation is made first for a plane configuration (plane vessel, plane multiplier without reflector, and plane multiplier with reflector), and then for a cylindrical configuration (multiplier without reflector, multiplier with infinitely thick reflector, finite cylindrical piles without reflector and with reflector). The self-sustain conditions are calculated for each case and the multiplication length and the shutdown delay are deduced. (J.S.)

  14. Spatial neutron kinetic module of ROSA code

    International Nuclear Information System (INIS)

    Cherezov, A.L.; Shchukin, N.V.

    2009-01-01

    A spatial neutron kinetic module was developed for computer code ROSA. The paper describes a numerical scheme used in the module for resolving neutron kinetic equations. Analytical integration for delayed neutrons emitters method and direct numerical integration method (Gear's method) were analyzed. The two methods were compared on their efficiency and accuracy. Both methods were verified with test problems. The results obtained in the verification studies were presented [ru

  15. Current studies of biological materials using instrumental and radiochemical neutron activation analysis

    International Nuclear Information System (INIS)

    Fardy, J.J.; McOrist, G.D.; Farrar, Y.J.

    1985-01-01

    Instrumental neutron activation analysis still remains the preferred option when analysing the trace element distribution in a wide rage of materials by neutron activation analysis. However, when lower limits of detection are required or major interferences reduce the effectiveness of this technique, radiochemical neutron activation analysis is applied. This paper examines the current use of both methods and the development of rapid radiochemical techniques for analysis of the biological materials, hair, cow's milk, human's milk, milk powder, blood and blood serum

  16. 50 mm Diameter digital DC/pulse neutron generator for subcritical reactor test

    International Nuclear Information System (INIS)

    Li Gang; Zhang Zhongshuai; Chi Qian; Liu Linmao

    2012-01-01

    A 50 mm diameter digital DC/pulse neutron generator was developed with 25 mm ceramic drive-in target neutron tube. It was applied in the subcritical reactor test of China Institute of Atomic Energy (CIAE). The generator can produce neutron in three modes: DC, pulse and multiple pulse. The maximum neutron yield of the generator is 1 × 10 8 n/s, while the maximum pulse frequency is 10 kHz, and the minimum pulse width is 10 μs. As a remote controlled generator, it is small in volume, easy to be connected and controlled. The tested results indicate that penning ion source has the feature of delay time in glow discharge, and it is easier for glow discharge to happen when switching the DC voltage of penning ion source into pulse. According to these two characteristics, the generator has been modified. This improved generator can be used in many other areas including Prompt Gamma Neutron Activation Analysis (PGNAA), neutron testing and experiment.

  17. 50 mm Diameter digital DC/pulse neutron generator for subcritical reactor test

    Energy Technology Data Exchange (ETDEWEB)

    Li Gang; Zhang Zhongshuai [Northeast Normal University, Changchun 130024 (China); Chi Qian [Guang Hua College of Chang Chun University, Changchun 130117 (China); Liu Linmao, E-mail: ll888@nenu.edu.cn [Northeast Normal University, Changchun 130024 (China)

    2012-11-01

    A 50 mm diameter digital DC/pulse neutron generator was developed with 25 mm ceramic drive-in target neutron tube. It was applied in the subcritical reactor test of China Institute of Atomic Energy (CIAE). The generator can produce neutron in three modes: DC, pulse and multiple pulse. The maximum neutron yield of the generator is 1 Multiplication-Sign 10{sup 8} n/s, while the maximum pulse frequency is 10 kHz, and the minimum pulse width is 10 {mu}s. As a remote controlled generator, it is small in volume, easy to be connected and controlled. The tested results indicate that penning ion source has the feature of delay time in glow discharge, and it is easier for glow discharge to happen when switching the DC voltage of penning ion source into pulse. According to these two characteristics, the generator has been modified. This improved generator can be used in many other areas including Prompt Gamma Neutron Activation Analysis (PGNAA), neutron testing and experiment.

  18. Solution of the neutron point kinetics equations with temperature feedback effects applying the polynomial approach method

    International Nuclear Information System (INIS)

    Tumelero, Fernanda; Petersen, Claudio Z.; Goncalves, Glenio A.; Lazzari, Luana

    2015-01-01

    In this work, we present a solution of the Neutron Point Kinetics Equations with temperature feedback effects applying the Polynomial Approach Method. For the solution, we consider one and six groups of delayed neutrons precursors with temperature feedback effects and constant reactivity. The main idea is to expand the neutron density, delayed neutron precursors and temperature as a power series considering the reactivity as an arbitrary function of the time in a relatively short time interval around an ordinary point. In the first interval one applies the initial conditions of the problem and the analytical continuation is used to determine the solutions of the next intervals. With the application of the Polynomial Approximation Method it is possible to overcome the stiffness problem of the equations. In such a way, one varies the time step size of the Polynomial Approach Method and performs an analysis about the precision and computational time. Moreover, we compare the method with different types of approaches (linear, quadratic and cubic) of the power series. The answer of neutron density and temperature obtained by numerical simulations with linear approximation are compared with results in the literature. (author)

  19. Solution of the neutron point kinetics equations with temperature feedback effects applying the polynomial approach method

    Energy Technology Data Exchange (ETDEWEB)

    Tumelero, Fernanda, E-mail: fernanda.tumelero@yahoo.com.br [Universidade Federal do Rio Grande do Sul (UFRGS), Porto Alegre, RS (Brazil). Programa de Pos-Graduacao em Engenharia Mecanica; Petersen, Claudio Z.; Goncalves, Glenio A.; Lazzari, Luana, E-mail: claudiopeteren@yahoo.com.br, E-mail: gleniogoncalves@yahoo.com.br, E-mail: luana-lazzari@hotmail.com [Universidade Federal de Pelotas (DME/UFPEL), Capao do Leao, RS (Brazil). Instituto de Fisica e Matematica

    2015-07-01

    In this work, we present a solution of the Neutron Point Kinetics Equations with temperature feedback effects applying the Polynomial Approach Method. For the solution, we consider one and six groups of delayed neutrons precursors with temperature feedback effects and constant reactivity. The main idea is to expand the neutron density, delayed neutron precursors and temperature as a power series considering the reactivity as an arbitrary function of the time in a relatively short time interval around an ordinary point. In the first interval one applies the initial conditions of the problem and the analytical continuation is used to determine the solutions of the next intervals. With the application of the Polynomial Approximation Method it is possible to overcome the stiffness problem of the equations. In such a way, one varies the time step size of the Polynomial Approach Method and performs an analysis about the precision and computational time. Moreover, we compare the method with different types of approaches (linear, quadratic and cubic) of the power series. The answer of neutron density and temperature obtained by numerical simulations with linear approximation are compared with results in the literature. (author)

  20. Current status of neutron scattering in Thailand

    International Nuclear Information System (INIS)

    Ampornrat, Pantip

    2000-01-01

    The neutron scattering experiments in Thailand have been done continuously since the start up of the reactor. In 1977, Thai research reactor was modified into TRIGA MARK III core. After that, the neutron spectrometer was installed again under a development program. Installation of upgrading spectrometer was delayed because of some problems involving the neutron intensity and instruments. However, these problems were solved and the setup is almost completed. The paper reports the current status of neutron spectrometer, the problems and plans for the experiments. (author)

  1. The verification of neutron activation analysis support system (cooperative research)

    Energy Technology Data Exchange (ETDEWEB)

    Sasajima, Fumio; Ichimura, Shigeju; Ohtomo, Akitoshi; Takayanagi, Masaji [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Sawahata, Hiroyuki; Ito, Yasuo [Tokyo Univ. (Japan). Research Center for Nuclear Science and Technology; Onizawa, Kouji [Radiation Application Development Association, Tokai, Ibaraki (Japan)

    2000-12-01

    Neutron activation analysis support system is the system in which even the user who has not much experience in the neutron activation analysis can conveniently and accurately carry out the multi-element analysis of the sample. In this verification test, subjects such functions, usability, precision and accuracy of the analysis and etc. of the neutron activation analysis support system were confirmed. As a method of the verification test, it was carried out using irradiation device, measuring device, automatic sample changer and analyzer equipped in the JRR-3M PN-3 facility, and analysis software KAYZERO/SOLCOI based on the k{sub 0} method. With these equipments, calibration of the germanium detector, measurement of the parameter of the irradiation field and analysis of three kinds of environmental standard sample were carried out. The k{sub 0} method adopted in this system is primarily utilized in Europe recently, and it is the analysis method, which can conveniently and accurately carried out the multi-element analysis of the sample without requiring individual comparison standard sample. By this system, total 28 elements were determined quantitatively, and 16 elements with the value guaranteed as analytical data of the NIST (National Institute of Standards and Technology) environment standard sample were analyzed in the accuracy within 15%. This report describes content and verification result of neutron activation support system. (author)

  2. Dose field research of analysis room for in-hospital neutron irradiator

    International Nuclear Information System (INIS)

    Zhang Zizhu; Song Mingzhe; Li Wei; Chen Jun; Yang Yong; Li Yiguo

    2012-01-01

    Neutron equivalent dose rate and y ray dose rate inside the analysis room of the in-hospital neutron irradiator (IHNI) and outdoor were measured. The results show that γ ray dose rate inside the analysis room exceeds calculation value many times and γ/ ray dose rate outdoor is higher than supervision region dose limit of 7.5 μSv/h. According to the measurement results and the Monte Carlo simulation, the following shielding plan was adopted. Lead shielding with thickness of 16 cm was installed on the wall, which faces the neutron beam, to shield γ ray, and lithium polyethylene plate with thickness of l cm was installed on all the wall (not including ceiling and floor) to shield scattering neutron. After shielding transformation, the highest γ ray dose rate point inside the analysis room decreased 277 times, the neutron equivalent dose rate decreased 5.8 times, and the outdoor γ/ray dose rate decreased nearly 90 times. (authors)

  3. Utilization of the intense pulsed neutron source (IPNS) at Argonne National Laboratory for neutron activation analysis

    International Nuclear Information System (INIS)

    Heinrich, R.R.; Greenwood, L.R.; Popek, R.J.; Schulke, A.W. Jr.

    1983-01-01

    The Intense Pulsed Neutron Source (IPNS) neutron scattering facility (NSF) has been investigated for its applicability to neutron activation analysis. A polyethylene insert has been added to the vertical hole VT3 which enhances the thermal neutron flux by a factor of two. The neutron spectral distribution at this position has been measured by the multiple-foil technique which utilized 28 activation reactions and the STAYSL computer code. The validity of this spectral measurement was tested by two irradiations of National Bureau of Standards SRM-1571 (orchard leaves), SRM-1575 (pine needles), and SRM-1645 (river sediment). The average thermal neutron flux for these irradiations normalized to 10 μamp proton beam is 4.0 x 10 11 n/cm 2 -s. Concentrations of nine trace elements in each of these SRMs have been determined by gamma-ray spectrometry. Agreement of measured values to certified values is demonstrated to be within experiment error

  4. Precision of neutron activation analysis for environmental biological materials

    International Nuclear Information System (INIS)

    Hamaguchi, Hiroshi; Iwata, Shiro; Koyama, Mutsuo; Sasajima, Kazuhisa; Numata, Yuichi.

    1977-01-01

    Between 1973 and 1974 a special committee ''Research on the application of neutron activation analysis to the environmental samples'' had been organized at the Research Reactor Institute, Kyoto University. Eleven research groups composed mainly of the committee members cooperated in the intercomparison programme of the reactor neutron activation analysis of NBS standard reference material, 1571 Orchard Leaves and 1577 Bovine Liver. Five different type of reactors were used for the neutron irradiation; i.e. KUR reactor of the Research Reactor Institute, Kyoto University, TRIGA MARK II reactor of the Institute for Atomic Energy, Rikkyo University, and JRR-2, JRR-3, JRR-4 reactor of Japan Atomic Energy Research Institute. Analyses were performed mainly by instrumental method. Precision of the analysis of 23 elements in Orchard Leaves and 13 elements in Bovine Liver presented by the different research groups was shown in table 4 and 5, respectively. The coefficient of variation for these elements was from several to -- 30 percent. Averages given to these elements agreed well with the NBS certified or reference values. Thus, from the practical point of view for the routine multielement analysis of environmental samples, the validity of the instrumental neutron activation technique for this purpose has been proved. (auth.)

  5. Analysis of Time Delay Simulation in Networked Control System

    OpenAIRE

    Nyan Phyo Aung; Zaw Min Naing; Hla Myo Tun

    2016-01-01

    The paper presents a PD controller for the Networked Control Systems (NCS) with delay. The major challenges in this networked control system (NCS) are the delay of the data transmission throughout the communication network. The comparative performance analysis is carried out for different delays network medium. In this paper, simulation is carried out on Ac servo motor control system using CAN Bus as communication network medium. The True Time toolbox of MATLAB is used for simulation to analy...

  6. Analysis of the Neutron Generator and Target for the LSDTS System

    Energy Technology Data Exchange (ETDEWEB)

    Park, Chang Je; Lee, Yong Deok; Song, Jae Hoon; Song, Kee Chan [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2008-11-15

    A preliminary analysis was performed based on the literatures and the patents for the neutron generators and targets for the lead slowing down time spectrometer (LSDTS) system. It was found that local neutron generator did not exhibit enough neutron intensity such as 1E+12 n/s, which is a minimum requirement for the LSDTS system to overcome curium backgrounds. However, a neutron generator implemented with an electron accelerator may provide a higher intensity around 1E+13 n/s and it is required to investigate further including a detail analysis. In addition to the neutron generator, a study on target was performed with the Monte Carlo simulation. In the study, an optimal design of target was suggested to provide a high neutron yield and a better thermal resistance. The suggested target consists several cylindrical plates with a certain cooling gap, which have increasing thickness and increasing radius.

  7. The role of delay in the dynamics of nuclear reactors

    International Nuclear Information System (INIS)

    Svitra, D.; Bucys, K.

    1999-01-01

    The stability of nuclear reactors based on nonlinear models of reactor dynamics including the action of delayed neutrons is analysed. The point model of reactor dynamics with the system of seven nonlinear simple differential equations was changed to the system of two nonlinear differential equations including the action of delay. The method of the theory of bifurcations for nonlinear differential equations with delay is used. (author)

  8. Experimental investigation of thermal neutron analysis based landmine detection technology

    International Nuclear Information System (INIS)

    Zeng Jun; Chu Chengsheng; Ding Ge; Xiang Qingpei; Hao Fanhua; Luo Xiaobing

    2013-01-01

    Background: Recently, the prompt gamma-rays neutron activation analysis method is wildly used in coal analysis and explosive detection, however there were less application about landmine detection using neutron method especially in the domestic research. Purpose: In order to verify the feasibility of Thermal Neutron Analysis (TNA) method used in landmine detection, and explore the characteristic of this technology. Methods: An experimental system of TNA landmine detection was built based on LaBr 3 (Ce) fast scintillator detector and 252 Cf isotope neutron source. The system is comprised of the thermal neutron transition system, the shield system, and the detector system. Results: On the basis of the TNA, the wide energy area calibration method especially to the high energy area was investigated, and the least detection time for a typical mine was defined. In this study, the 72-type anti-tank mine, the 500 g TNT sample and several interferential objects are tested in loess, red soil, magnetic soil and sand respectively. Conclusions: The experimental results indicate that TNA is a reliable demining method, and it can be used to confirm the existence of Anti-Tank Mines (ATM) and large Anti-Personnel Mines (APM) in complicated condition. (authors)

  9. New neutron-deficient isotopes of barium and rare-earth elements

    CERN Document Server

    Bogdanov, D D; Karnaukhov, V A; Petrov, L A; Plochocki, A; Subbotin, V G; Voboril, J

    1976-01-01

    The authors present an investigation of the short-lived neutron- deficient isotopes of barium and rare-earth elements. By using the BEMS-2 isotope separator on a heavy ion beam, 19 new isotopes were produced with mass numbers ranging from 117 to 138. Five of these (/sup 117/Ba, /sup 129,131/Nd and /sup 133,135/Sm) turned out to be delayed proton emitters. The beta -decay probabilities for the new isotopes have been analyzed in terms of the beta -strength function. An analysis of the proton spectrum shape has been performed using the statistical model for delayed proton emission.

  10. Neutron activation analysis: an emerging technique for conservation/preservation

    International Nuclear Information System (INIS)

    Sayre, E.V.

    1976-01-01

    The diverse applications of neutron activation in analysis, preservation, and documentation of art works and artifacts are described with illustrations for each application. The uses of this technique to solve problems of attribution and authentication, to reveal the inner structure and composition of art objects, and, in some instances to recreate details of the objects are described. A brief discussion of the theory and techniques of neutron activation analysis is also included

  11. Development of neutron diffuse scattering analysis code by thin film and multilayer film

    International Nuclear Information System (INIS)

    Soyama, Kazuhiko

    2004-01-01

    To research surface structure of thin film and multilayer film by neutron, a neutron diffuse scattering analysis code using DWBA (Distorted-Wave Bron Approximation) principle was developed. Subjects using this code contain the surface and interface properties of solid/solid, solid/liquid, liquid/liquid and gas/liquid, and metal, magnetism and polymer thin film and biomembran. The roughness of surface and interface of substance shows fractal self-similarity and its analytical model is based on DWBA theory by Sinha. The surface and interface properties by diffuse scattering are investigated on the basis of the theoretical model. The calculation values are proved to be agreed with the experimental values. On neutron diffuse scattering by thin film, roughness of surface of thin film, correlation function, neutron propagation by thin film, diffuse scattering by DWBA theory, measurement model, SDIFFF (neutron diffuse scattering analysis program by thin film) and simulation results are explained. On neutron diffuse scattering by multilayer film, roughness of multilayer film, principle of diffuse scattering, measurement method and simulation examples by MDIFF (neutron diffuse scattering analysis program by multilayer film) are explained. (S.Y.)To research surface structure of thin film and multilayer film by neutron, a neutron diffuse scattering analysis code using DWBA (Distorted-Wave Bron Approximation) principle was developed. Subjects using this code contain the surface and interface properties of solid/solid, solid/liquid, liquid/liquid and gas/liquid, and metal, magnetism and polymer thin film and biomembran. The roughness of surface and interface of substance shows fractal self-similarity and its analytical model is based on DWBA theory by Sinha. The surface and interface properties by diffuse scattering are investigated on the basis of the theoretical model. The calculation values are proved to be agreed with the experimental values. On neutron diffuse scattering

  12. The new high flux neutron source FRM-2 in Munich

    International Nuclear Information System (INIS)

    Roegler, H.J.; Wierheim, G.

    2002-01-01

    Quite some years ago in 1974 to be exact, the first consideration on a new neutron source started at the technical university of Munich (Germany). 27 years later the new high flux neutron source (FRM-2) was read for hot operation, now delayed by a refused approval for its third partial license by the federal government of Germany despite a wide support from the scientific community. FRM-2 is a tank-type research reactor cooled by water, moderated by heavy water and whose thermal power was limited to 20 MW maximum. The extreme compact core together with the applied inverse flux principle led to a neutron flux design value of 8.10 18 n/m 2 .s at the reflector peak. 10 beam tubes will allow an optimized use of the high neutron flux. A hot neutron source with graphite at about 2200 Celsius degrees and a cold neutron source with liquid D 2 at about 25 K will provide shifted energy spectra. The utilization of FRM-2 is many-fold: neutronography and tomography, medical irradiation, radio-nuclide production, doping of pure silicon, neutron activation analysis. (A.C.)

  13. Method and apparatus for measuring thermal neutron characteristics

    International Nuclear Information System (INIS)

    Johnstone, C.W.

    1983-01-01

    The thermal neutron decay characteristics of an earth formation are measured by detecting indications of the thermal neutron concentration in the formation during a selected set of two measurement intervals following irradiation of the formation with a burst of fast neutrons. These measurement intervals may comprise a sequence of time gates following a delay after the neutron burst. The duration of the neutron bursts, of the delay between the burst and the start of the sequence, and of the individual time gates, may all be adjusted by a common, selected one of a finite number of scale factor values. The set of two measurement intervals is selected from among a number of possible sets as a function of a previously measured value of the decay characteristic. Each measurement interval set is used over only a specific range of decay characteristic values for which it has been determined, in accordance with a previously established relationship between the decay characteristic value and a function of the thermal neutron concentration measurements for the set, to afford enhanced statistical accuracy in the measured value of the decay characteristic. (author)

  14. Neutron Activation Analysis with k0 standardisation

    International Nuclear Information System (INIS)

    Pomme, S.

    1998-01-01

    The objectives of the research are: (1) to develop and implement the k0 standardisation method for neutron activation analysis in close collaboration with scientific partners; (2) to exploit fully the inherent qualities of NAA such as accuracy, traceability, and multi-element offer complete services in health-physics measurements according to international quality standards, (2) to improve continuously these measurement techniques and to follow up international recommendations and legislation concerning the surveillance of workers; (3) to support and advise nuclear and non-nuclear industry on problems of radioactive contamination. Achievements in 1997 related to gamma spectrometry, whole-body counting, beta and alpha spectrometry, dosimetry, radon measurements, calibration, instrumentation, and neutron activation analysis are described

  15. Utilization and facility of neutron activation analysis in HANARO research reactor

    International Nuclear Information System (INIS)

    Chung, Y.S.; Chung, Y.J.; Moon, J.H.

    1998-01-01

    The facilities of neutron activation analysis within a multi-purpose research reactor (HANARO) are described and the main applications of Neutron activation analysis (NAA) in Korea are reviewed. The sample irradiation tube, automatic and manual pneumatic transfer system, are installed at three irradiation holes. One irradiation hole is lined with a cadmium tube for epithermal-nal NAA. The performance of the NAA facility was examined to identify the characteristics of tube transfer system, irradiation sites and polyethylene irradiation capsule. The available thermal neutron flux with each irradiation site are in the range of 3.9x10 13 -1.6x10 14 n/cm 2 ·s and cadmium ratios are 15-250. Neutron activation analysis has been applied in the trace component analysis of nuclear, geological, biological, environmental and high purity materials and various polymers for research and development. Analytical services and the latest analytical results are summarized. (author)

  16. Current status of neutron activation analysis in HANARO Research Reactor

    International Nuclear Information System (INIS)

    Chung, Yong Sam; Moon, Jong Hwa; Sohn, Jae Min

    2003-01-01

    The facilities for neutron activation analysis in the HANARO (Hi-flux Advanced Neutron Application Research Reactor) are described and the main applications of NAA (Neutron Activation Analysis) are reviewed. The sample irradiation tube, automatic and manual pneumatic transfer system were installed at three irradiation holes of HANARO at the end of 1995. The performance of the NAA facility was examined to identify the characteristics of the tube transfer system, irradiation sites and custom-made polyethylene irradiation capsule. The available thermal neutron fluxes at irradiation sites are in the range of 3 x 10 13 - 1 x 10 14 n/cm 2 ·s and cadmium ratios are in 15 - 250. For an automatic sample changer for gamma-ray counting, a domestic product was designed and manufactured. An integrated computer program (Labview) to analyse the content was developed. In 2001, PGNAA (Prompt Gamma Neutron Activation Analysis) facility has been installed using a diffracted neutron beam of ST1. NAA has been applied in the trace component analysis of nuclear, geological, biological, environmental and high purity materials, and various polymers for research and development. The improvement of analytical procedures and establishment of an analytical quality control and assurance system were studied. Applied research and development for the environment, industry and human health by NAA and its standardization were carried out. For the application of the KOLAS (Korea Laboratory Accreditation Scheme), evaluation of measurement uncertainty and proficiency testing of reference materials were performed. Also to verify the reliability and to validate analytical results, intercomparison studies between laboratories were carried out. (author)

  17. Current status of neutron activation analysis in HANARO Research Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Yong Sam; Moon, Jong Hwa; Sohn, Jae Min [Korea Atomic Energy Research Institute, Daejeon (Korea)

    2003-03-01

    The facilities for neutron activation analysis in the HANARO (Hi-flux Advanced Neutron Application Research Reactor) are described and the main applications of NAA (Neutron Activation Analysis) are reviewed. The sample irradiation tube, automatic and manual pneumatic transfer system were installed at three irradiation holes of HANARO at the end of 1995. The performance of the NAA facility was examined to identify the characteristics of the tube transfer system, irradiation sites and custom-made polyethylene irradiation capsule. The available thermal neutron fluxes at irradiation sites are in the range of 3 x 10{sup 13} - 1 x 10{sup 14} n/cm{sup 2}{center_dot}s and cadmium ratios are in 15 - 250. For an automatic sample changer for gamma-ray counting, a domestic product was designed and manufactured. An integrated computer program (Labview) to analyse the content was developed. In 2001, PGNAA (Prompt Gamma Neutron Activation Analysis) facility has been installed using a diffracted neutron beam of ST1. NAA has been applied in the trace component analysis of nuclear, geological, biological, environmental and high purity materials, and various polymers for research and development. The improvement of analytical procedures and establishment of an analytical quality control and assurance system were studied. Applied research and development for the environment, industry and human health by NAA and its standardization were carried out. For the application of the KOLAS (Korea Laboratory Accreditation Scheme), evaluation of measurement uncertainty and proficiency testing of reference materials were performed. Also to verify the reliability and to validate analytical results, intercomparison studies between laboratories were carried out. (author)

  18. Multielement analysis of biological standards by neutron activation analysis

    International Nuclear Information System (INIS)

    Nadkarni, R.A.

    1977-01-01

    Up to 28 elements were determined in two IAEA standards: Animal Muscle H4 and Fish Soluble A 6/74, and three NBS standards: Spinach: SRM-1570, Tomato Leaves: SRM-1573 and Pine Needles: SRM-1575 by instrumental neutron-activation analysis. Seven noble metals were determined in two NBS standards: Coal: SRM-1632 and Coal Fly Ash: SRM-1633 by radiochemical procedure while 11 rare earth elements were determined in NBS standard Orchard Leaves: SRM-1571 by instrumental neutron-activation analysis. The results are in good agreement with the certified and/or literature data where available. The irradiations were performed at the Cornell TRIGA Mark II nuclear reactor at a thermal neutron flux of 1-3x10 12 ncm -2 sec -1 . The short-lived species were determined after a 2-minute irradiation in the pneumatic rabbit tube, and the longer-lived species after an 8-hour irradiation in the central thimble facility. The standards and samples were counted on coaxial 56-cm 3 Ge(Li) detector. The system resolution was 1.96 keV (FWHM) with a peak to Compton ratio of 37:1 and counting efficiency of 13%, all compared to the 1.332 MeV photopeak of Co-60. (T.I.)

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

    International Nuclear Information System (INIS)

    2001-04-01

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

  20. Prospective analysis of delayed colorectal post-polypectomy bleeding.

    Science.gov (United States)

    Park, Soo-Kyung; Seo, Jeong Yeon; Lee, Min-Gu; Yang, Hyo-Joon; Jung, Yoon Suk; Choi, Kyu Yong; Kim, Hungdai; Kim, Hyung Ook; Jung, Kyung Uk; Chun, Ho-Kyung; Park, Dong Il

    2018-01-17

    Although post-polypectomy bleeding is the most frequent complication after colonoscopic polypectomy, only few studies have investigated the incidence of bleeding prospectively. The aim of this study was to investigate the incidence of delayed post-polypectomy bleeding and its associated risk factors prospectively. Patients who underwent colonoscopic polypectomy at Kangbuk Samsung Hospital from January 2013 to December 2014 were prospectively enrolled in this study. Trained nurses contacted patients via telephone 7 and 30 days after polypectomy and completed a standardized questionnaire regarding the development of bleeding. Delayed post-polypectomy bleeding was categorized as minor or major and early or late bleeding. Major delayed bleeding was defined as a > 2-g/dL drop in the hemoglobin level, requiring hospitalization for control of bleeding or blood transfusion; late delayed bleeding was defined as bleeding occurring later than 24 h after polypectomy. A total of 8175 colonoscopic polypectomies were performed in 3887 patients. Overall, 133 (3.4%) patients developed delayed post-polypectomy bleeding. Among them, 90 (2.3%) and 43 (1.1%) patients developed minor and major delayed bleeding, respectively, and 39 (1.0%) patients developed late delayed bleeding. In the polyp-based multivariate analysis, young age ( 10 mm (OR 2.45; 95% CI 1.38-4.36) were significant risk factors for major delayed bleeding, while young age (< 50 years; OR 2.6; 95% CI 1.35-5.12) and immediate bleeding (OR 3.3; 95% CI 1.49-7.30) were significant risk factors for late delayed bleeding. Young age, aspirin use, polyp size, and immediate bleeding were found to be independent risk factors for delayed post-polypectomy bleeding.

  1. Neutron multimonochromator-bipolarizer based on magnetic multilayer Fe/Co and new scheme for the total neutron polarization analysis

    International Nuclear Information System (INIS)

    Syromyatnikov, V.G.; Zaw Lin, Kyaw

    2017-01-01

    In this paper, we present a new neutron-optical element, Neutron Multimonochromator-Bipolarizer (NMB). It consists of a multimultilayer structure made of 12 periodic multilayer Fe/Co magnetic nanostructures whose period increases with distance from the substrate. Results are presented of calculations of the reflection coefficients from the NMB. We propose a new scheme of the total neutron polarization analysis for the time-of-flight method in the reflectometry. In this scheme, double NMB is used as a polarizer and there is no spin-flipper before the sample. NMB can be used in polarized neutron reflectometry, in SESANS, and for research of low-angle and inelastic scattering of polarized neutrons. (paper)

  2. High count problems in elemental analysis using pulsed neutron inelastic scattering

    Energy Technology Data Exchange (ETDEWEB)

    Vartsky, D; Wielopolski, L; Ellis, K J; Cohn, S H [Brookhaven National Lab., Upton, NY (USA). Medical Dept.

    1983-03-01

    Elemental analysis by neutron inelastic scattering using a miniature intense pulsed neutron source ('Zetatron') was evaluated. The particular problems associated with detector pulse-pile-up during the neutron burst and the limited ability of the analyzer to process on average more than one detector pulse per neutron burst were examined. The severity of these problems is described and a solution using a multiple ADC system is proposed.

  3. Description of the CAREM Reactor Neutronic Calculation Codes

    International Nuclear Information System (INIS)

    Villarino, Eduardo; Hergenreder, Daniel

    2000-01-01

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

  4. Automated activation-analysis system

    International Nuclear Information System (INIS)

    Minor, M.M.; Hensley, W.K.; Denton, M.M.; Garcia, S.R.

    1981-01-01

    An automated delayed neutron counting and instrumental neutron activation analysis system has been developed at Los Alamos National Laboratory's Omega West Reactor (OWR) to analyze samples for uranium and 31 additional elements with a maximum throughput of 400 samples per day. The system and its mode of operation for a large reconnaissance survey are described

  5. Numerical Bifurcation Analysis of Delayed Recycle Stream in a Continuously Stirred Tank Reactor

    Science.gov (United States)

    Gangadhar, Nalwala Rohitbabu; Balasubramanian, Periyasamy

    2010-10-01

    In this paper, we present the stability analysis of delay differential equations which arise as a result of transportation lag in the CSTR-mechanical separator recycle system. A first order irreversible elementary reaction is considered to model the system and is governed by the delay differential equations. The DDE-BIFTOOL software package is used to analyze the stability of the delay system. The present analysis reveals that the system exhibits delay independent stability for isothermal operation of the CSTR. In the absence of delay, the system is dynamically unstable for non-isothermal operation of the CSTR, and as a result of delay, the system exhibits delay dependent stability.

  6. Neutron scattering investigation on low-dimensional, quantum and frustrated magnetism and utilization of neutron polarization analysis. My first encounter with neutron research

    International Nuclear Information System (INIS)

    Kakurai, Kazuhisa

    2013-01-01

    My first encounter with neutron scattering research on low-dimensional magnetism at the Hahn-Meitner Institut under the supervision of Prof. H. Dachs and Prof. M. Steiner, were it all began, is accounted for. The polarized neutron analysis research on low-dimensional magnetism at the Institut Laue Langevin under the supervision of Dr. R. Pynn is also reported. I would like to dedicate this article to late Prof. H. Dachs expressing may deepest gratitude for his warm guidance during the early period of my neutron science carrier. (author)

  7. TMCC: a transient three-dimensional neutron transport code by the direct simulation method - 222

    International Nuclear Information System (INIS)

    Shen, H.; Li, Z.; Wang, K.; Yu, G.

    2010-01-01

    A direct simulation method (DSM) is applied to solve the transient three-dimensional neutron transport problems. DSM is based on the Monte Carlo method, and can be considered as an application of the Monte Carlo method in the specific type of problems. In this work, the transient neutronics problem is solved by simulating the dynamic behaviors of neutrons and precursors of delayed neutrons during the transient process. DSM gets rid of various approximations which are always necessary to other methods, so it is precise and flexible in the requirement of geometric configurations, material compositions and energy spectrum. In this paper, the theory of DSM is introduced first, and the numerical results obtained with the new transient analysis code, named TMCC (Transient Monte Carlo Code), are presented. (authors)

  8. Trace element analysis of common salt using neutron activation analysis

    International Nuclear Information System (INIS)

    Usman, K.

    1993-01-01

    Instrumental Fast Neutron Activation Analysis (IFNAA) technique has been used in the qualitative and quantitative determination of the impurity elements in common salt. Samples of the different types of common salt processed in Nigeria and some of those imported into the country were used. The type A711 KAMAN neutron generator and a high-purity Germanium (HpGe) gamma spectrometer available at the Centre for Energy Research and Training, Ahmadu Bello University, Zaria has been used. The ORTEC ADCAM 100 Emulation Software (Maestro) was used in the qualitative measurement of the detected elements. The G.R.G Activation Analysis System by G. R. Gilmore, 1987, was used in the quantitative determination of the elements detected by relative method. Aluminium and arsenic were detected and measured

  9. Coupled neutronics and thermal-hydraulics numerical simulations of a Molten Salt Fast Reactor (MSFR)

    International Nuclear Information System (INIS)

    Laureau, A.; Rubiolo, P.R.; Heuer, D.; Merle-Lucotte, E.; Brovchenko, M.

    2013-01-01

    Coupled neutronics and thermalhydraulic numerical analyses of a molten salt fast reactor (MSFR) are presented. These preliminary numerical simulations are carried-out using the Monte Carlo code MCNP and the Computation Fluid Dynamic code OpenFOAM. The main objectives of this analysis performed at steady-reactor conditions are to confirm the acceptability of the current neutronic and thermalhydraulic designs of the reactor, to study the effects of the reactor operating conditions on some of the key MSFR design parameters such as the temperature peaking factor. The effects of the precursor's motion on the reactor safety parameters such as the effective fraction of delayed neutrons have been evaluated. (authors)

  10. Simultaneous analysis of qualitative parameters of solid fuel using complex neutron gamma method

    International Nuclear Information System (INIS)

    Dombrovskij, V.P.; Ajtsev, N.I.; Ryashchikov, V.I.; Frolov, V.K.

    1983-01-01

    A study was made on complex neutron gamma method for simultaneous analysis of carbon content, ash content and humidity of solid fuel according to gamma radiation of inelastic fast neutron scattering and radiation capture of thermal neutrons. Metrological characteristics of pulse and stationary neutron gamma methods for determination of qualitative solid fuel parameters were analyzed, taking coke breeze as an example. Optimal energy ranges of gamma radiation detection (2-8 MeV) were determined. The advantages of using pulse neutron generator for complex analysis of qualitative parameters of solid fuel in large masses were shown

  11. Neutrons moderation theory; Theorie du ralentissement des neutrons

    Energy Technology Data Exchange (ETDEWEB)

    Vigier, J P

    1949-07-01

    This report gives a summarized presentation of the theory of fast neutrons diffusion and moderation in a given environment as elaborated by M. Langevin, E. Fermi, R. Marshak and others. This statistical theory is based on three assumptions: there is no inelastic diffusion, the elastic diffusion has a spherical symmetry with respect to the center of gravity of the neutron-nucleus system (s-scattering), and the effects of chemical bonds and thermal agitation of nuclei are neglected. The first chapter analyzes the Boltzmann equation of moderation, its first approximate solution (age-velocity equation) and its domain of validity, the extension of the age-velocity theory (general solution) and the boundary conditions, the upper order approximation (spherical harmonics method and Laplace transformation), the asymptotic solutions, and the theory of spatial momenta. The second chapter analyzes the energy distribution of delayed neutrons (stationary and non-stationary cases). (J.S.)

  12. Utilization and facility of neutron activation analysis in HANARO research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Chung, Y S; Chung, Y J; Moon, J H [Korea Atomic Energy Research Institute, P.O.Box 105 Yusong, 305-600, Taejon (Korea, Republic of)

    1998-07-01

    The facilities of neutron activation analysis within a multi-purpose research reactor (HANARO) are described and the main applications of Neutron activation analysis (NAA) in Korea are reviewed. The sample irradiation tube, automatic and manual pneumatic transfer system, are installed at three irradiation holes. One irradiation hole is lined with a cadmium tube for epithermal-nal NAA. The performance of the NAA facility was examined to identify the characteristics of tube transfer system, irradiation sites and polyethylene irradiation capsule. The available thermal neutron flux with each irradiation site are in the range of 3.9x10{sup 13}-1.6x10{sup 14} n/cm{sup 2}{center_dot}s and cadmium ratios are 15-250. Neutron activation analysis has been applied in the trace component analysis of nuclear, geological, biological, environmental and high purity materials and various polymers for research and development. Analytical services and the latest analytical results are summarized. (author)

  13. Practical considerations in instrumental neutron activation analysis

    International Nuclear Information System (INIS)

    Ahmad, N.

    2001-01-01

    Activation analysis is a technique of elemental analysis based on the measurement of characteristics radiation from radionuclides formed directly or indirectly by activation. The activation can be induced by bombarding the material with neutrons or charged particles or gamma rays. This is a well-accepted analytical technique for the determination of composition of complex materials. This technique is also sensitive at trace levels and is almost free from analytical interferences of matrix. It is used for multi-elemental determination in rocks, minerals, alloys, biological materials, geological samples, non-destructive analysis of materials and environmental samples such as water, air particulate matter, plants, soil, sediments and diets. This method is also used for production and measurements of radioisotopes in materials of known composition, for example, when radioactivation is used for nuclear reaction studies, for flux and beam intensity measurements for trace experiments and process quality control. In this article the parameters affecting the sensitivity of instrumental neutron activation analysis are briefly discussed. (author)

  14. Neutron activation analysis of artefacts

    International Nuclear Information System (INIS)

    Mohd Suhaimi Hamzah; Shamsiah Abd Rahman

    2004-01-01

    The paper discussed the utilization of neutron activation analysis in this field. NAA, an analytical technique which analyzing the elements in the sample without any chemical treatment. It is sensitive and accurate. Archaeological objects i.e. ceramics, historical building materials, metals, etc can be analyze with this technique. The analysis results were presented in form of characterization of the artefacts in chemical profiles, which can present the information of the origin of the artefacts as well as it originality. (Author)

  15. General principles of neutron activation analysis

    International Nuclear Information System (INIS)

    Dostal, J.; Elson, C.

    1980-01-01

    Aspects of the principles of atomic and nuclear structure and the processes of radioactivity, nuclear transformation, and the interaction of radiations with matter which are of direct relevance to neutron activation analysis and its application to geologic materials are discussed. (L.L.)

  16. Construction Delay Analysis Techniques—A Review of Application Issues and Improvement Needs

    Directory of Open Access Journals (Sweden)

    Nuhu Braimah

    2013-07-01

    Full Text Available The time for performance of a project is usually of the essence to the employer and the contractor. This has made it quite imperative for contracting parties to analyse project delays for purposes of making right decisions on potential time and/or cost compensation claims. Over the years, existing delay analysis techniques (DATs for aiding this decision-making have been helpful but have not succeeded in curbing the high incidence of disputes associated with delay claims resolutions. A major source of the disputes lies with the limitations and capabilities of the techniques in their practical use. Developing a good knowledge of these aspects of the techniques is of paramount importance in understanding the real problematic issues involved and their improvement needs. This paper seeks to develop such knowledge and understanding (as part of a wider research work via: an evaluation of the most common DATs based on a case study, a review of the key relevant issues often not addressed by the techniques, and the necessary improvements needs. The evaluation confirmed that the various techniques yield different analysis results for the same delay claims scenario, mainly due to their unique application procedures. The issues that are often ignored in the analysis but would also affect delay analysis results are: functionality of the programming software employed for the analysis, resource loading and levelling requirements, resolving concurrent delays, and delay-pacing strategy. Improvement needs by way of incorporating these issues in the analysis and focusing on them in future research work are the key recommendations of the study.

  17. Procedures for multielement analysis using high-flux fast-neutron activation

    International Nuclear Information System (INIS)

    Williams, R.E.; Hopke, P.K.; Meyer, R.A.

    1981-06-01

    Improvements have been made in the rabbit system used for multi-element fast-neutron activation analysis at the Lawrence Livermore National Laboratory Rotating Target Neutron Source, RTNS-I. Procedures have been developed for the analysis of 20 to 25 elements in samples with an inorganic matrix and 10 to 15 elements in biological samples, without the need for prohibitively expensive, long irradiations. Results are presented for the analysis of fly ash, orchard leaves, and bovine liver

  18. Hopf bifurcation analysis of Chen circuit with direct time delay feedback

    International Nuclear Information System (INIS)

    Hai-Peng, Ren; Wen-Chao, Li; Ding, Liu

    2010-01-01

    Direct time delay feedback can make non-chaotic Chen circuit chaotic. The chaotic Chen circuit with direct time delay feedback possesses rich and complex dynamical behaviours. To reach a deep and clear understanding of the dynamics of such circuits described by delay differential equations, Hopf bifurcation in the circuit is analysed using the Hopf bifurcation theory and the central manifold theorem in this paper. Bifurcation points and bifurcation directions are derived in detail, which prove to be consistent with the previous bifurcation diagram. Numerical simulations and experimental results are given to verify the theoretical analysis. Hopf bifurcation analysis can explain and predict the periodical orbit (oscillation) in Chen circuit with direct time delay feedback. Bifurcation boundaries are derived using the Hopf bifurcation analysis, which will be helpful for determining the parameters in the stabilisation of the originally chaotic circuit

  19. Cold neutron prompt gamma activation analysis at NIST; A progress report

    Energy Technology Data Exchange (ETDEWEB)

    Paul, R L; Lindstrom, R M [National Inst. of Standards and Technology, Gaithersburg, MD (United States). Div. of Inorganic Analytical Research; Vincent, D H [Michigan Univ., Ann Arbor, MI (United States). Dept. of Nuclear Engineering

    1994-05-01

    An instrument for prompt gamma-ray activation analysis is now in operation at the NIST Cold Neutron Research Facility (CNRF). The cold neutron beam is relatively free of contamination by fast neutrons and reactor gamma rays, and the neutron fluence rate is 1.5 x 10 [sup 8] cm [sup -2] x s [sup -1] (thermal equivalent). As a result of a compact target-detector geometry the sensitivity is better by a factor of as much as seven than that obtained with an existing thermal instrument, and hydrogen background is a factor of 50 lower. This instrument was applied to multielement analysis of the Allende meteorite and other materials. (author) 14 refs.; 2 figs.; 1 tab.

  20. Neutron activation analysis of high purity substances

    International Nuclear Information System (INIS)

    Gil'bert, Eh.N.

    1987-01-01

    Peculiarities of neutron-activation analysis (NAA) of high purity substances are considered. Simultaneous determination of a wide series of elements, high sensitivity (the lower bound of determined contents 10 -9 -10 -10 %), high selectivity and accuracy (Sr=0.10-0.15, and may be decreased up to 0.001), possibility of analysis of the samples from several micrograms to hundreds of grams, simplicity of calibration may be thought NAA advantages. Questions of accounting of NAA systematic errors associated with the neutron flux screening by the analysed matrix and with production of radionuclides of determined elements from accompanying elements according to concurrent nuclear reactions, as well as accounting of errors due to self-absorption of recorded radiation by compact samples, are considered

  1. Singular Perturbation Analysis and Gene Regulatory Networks with Delay

    Science.gov (United States)

    Shlykova, Irina; Ponosov, Arcady

    2009-09-01

    There are different ways of how to model gene regulatory networks. Differential equations allow for a detailed description of the network's dynamics and provide an explicit model of the gene concentration changes over time. Production and relative degradation rate functions used in such models depend on the vector of steeply sloped threshold functions which characterize the activity of genes. The most popular example of the threshold functions comes from the Boolean network approach, where the threshold functions are given by step functions. The system of differential equations becomes then piecewise linear. The dynamics of this system can be described very easily between the thresholds, but not in the switching domains. For instance this approach fails to analyze stationary points of the system and to define continuous solutions in the switching domains. These problems were studied in [2], [3], but the proposed model did not take into account a time delay in cellular systems. However, analysis of real gene expression data shows a considerable number of time-delayed interactions suggesting that time delay is essential in gene regulation. Therefore, delays may have a great effect on the dynamics of the system presenting one of the critical factors that should be considered in reconstruction of gene regulatory networks. The goal of this work is to apply the singular perturbation analysis to certain systems with delay and to obtain an analog of Tikhonov's theorem, which provides sufficient conditions for constracting the limit system in the delay case.

  2. Development of a neutronic analysis code using data from Monju

    International Nuclear Information System (INIS)

    Rooijen, W.F.G. van; Yamano, N.; Shimazu, Y.

    2015-01-01

    In recent years three major sets of modern evaluated nuclear data have become available: JENDL-4.0, JEFF-3.1.2 and ENDF/B-VII.1. The authors were involved with a research project to establish analysis method for a future commercial-scale LMFBR. This project focused on JENDL-4.0 and conventional Japanese codes. As a cross check, we decided to also apply the fast reactor code ERANOS. This necessitated to produce nuclear data (cross sections, etc) for the ERANOS code system, discussed in this paper. We developed a nuclear data processing system to produce cross sections, probability tables, delayed neutron data, and covariance data from the evaluated nuclear data files for ERANOS. A benchmark calculation on the MZA/MZB benchmark showed very satisfying results. Subsequently, we analyzed the prototype LMFBR Monju with ERANOS and our own sets of nuclear data. The results are very satisfactory. The results from ERANOS indicate that the target accuracies for nuclear data have not been met, although the three sets of evaluated nuclear data all performed very well in our analysis. In the future, the covariance on nuclear data should be reduced to meet the target accuracies on criticality and feedback coefficients. (author)

  3. Kalman filtering of self-powered neutron detectors

    International Nuclear Information System (INIS)

    Kantrowitz, M.L.

    1992-01-01

    Pressurized water reactors employ a wide variety of in-core detectors to determine the neutronic behavior within the core. Among the detectors used are rhodium and vanadium self-powered detectors (SPDs), which are very accurate, but respond slowly to changes in neutron flux. This paper describes a new dynamic compensation algorithm, based on Kalman filtering, which converts delayed-responding rhodium and vanadium SPDs into prompt-responding detectors by reconstructing the dynamic flux signal sensed by the detectors from the prompt and delayed components. This conversion offers the possibility of utilizing current fixed in-core detector systems based on these delayed-responding detectors for core control and/or core protection functions without the need for fixed in-core detectors which are prompt-responding. As a result, the capabilities of current fixed in-core detector systems could be expanded significantly without a major hardware investment

  4. Exploitation of the Fourier chopper in neutron diffractometry at pulsed sources

    International Nuclear Information System (INIS)

    Hiismaeki, P.; Poeyry, H.; Tiitta, A.

    1988-01-01

    The application of the Fourier chopper in the reverse time-of-flight mode to upgrading the performance of neutron powder diffractometry at pulsed sources is considered. Exploitation of a dual-delay-line binary correlator is suggested and a comprehensive analysis of the necessary data acquisition process is given. The best benefit is shown to be achievable at intensity optimized moderators. (orig.)

  5. Delayed Gamma-Ray Spectroscopy for Non-Destructive Assay of Nuclear Materials

    International Nuclear Information System (INIS)

    Ludewigt, Bernhard; Mozin, Vladimir; Campbell, Luke; Favalli, Andrea; Hunt, Alan W.; Reedy, Edward T.E.; Seipel, Heather

    2015-01-01

    High-energy, beta-delayed gamma-ray spectroscopy is a potential, non-destructive assay techniques for the independent verification of declared quantities of special nuclear materials at key stages of the fuel cycle and for directly assaying nuclear material inventories for spent fuel handling, interim storage, reprocessing facilities, repository sites, and final disposal. Other potential applications include determination of MOX fuel composition, characterization of nuclear waste packages, and challenges in homeland security and arms control verification. Experimental measurements were performed to evaluate fission fragment yields, to test methods for determining isotopic fractions, and to benchmark the modeling code package. Experimental measurement campaigns were carried out at the IAC using a photo-neutron source and at OSU using a thermal neutron beam from the TRIGA reactor to characterize the emission of high-energy delayed gamma rays from 235 U, 239 Pu, and 241 Pu targets following neutron induced fission. Data were collected for pure and combined targets for several irradiation/spectroscopy cycle times ranging from 10/10 seconds to 15/30 minutes.The delayed gamma-ray signature of 241 Pu, a significant fissile constituent in spent fuel, was measured and compared to 239 Pu. The 241 Pu/ 239 Pu ratios varied between 0.5 and 1.2 for ten prominent lines in the 2700-3600 keV energy range. Such significant differences in relative peak intensities make it possible to determine relative fractions of these isotopes in a mixed sample. A method for determining fission product yields by fitting the energy and time dependence of the delayed gamma-ray emission was developed and demonstrated on a limited 235 U data set. De-convolution methods for determining fissile fractions were developed and tested on the experimental data. The use of high count-rate LaBr 3 detectors was investigated as a potential alternative to HPGe detectors. Modeling capabilities were added to an

  6. Analysis of the Quasi-Elastic Scattering of Neutrons in Hydrogenous Liquids

    Energy Technology Data Exchange (ETDEWEB)

    Porohit, S N [Nuclear Science and Engineering Dept., Rensselaer Polytechnique Inst., Troy, NY (United States)

    1966-11-15

    A critical discussion of the quasi-elastic scattering of neutrons by incoherent (hydrogenous) liquids is presented. Using the line shape expression a comparative discussion of several phenomenological models has been carried out. Extension of the Singwi-Sjoelander zero phonon expression, for the jump-diffusion model, so as to include the one phonon expression has also been given. For a delayed diffusion model a complete treatment of S(K, {omega}) is presented. Along the lines of the macroscopic diffusion cooling, a microscopic diffusion cooling effect in fluids is speculated.

  7. Analysis of the Quasi-Elastic Scattering of Neutrons in Hydrogenous Liquids

    International Nuclear Information System (INIS)

    Porohit, S.N.

    1966-11-01

    A critical discussion of the quasi-elastic scattering of neutrons by incoherent (hydrogenous) liquids is presented. Using the line shape expression a comparative discussion of several phenomenological models has been carried out. Extension of the Singwi-Sjoelander zero phonon expression, for the jump-diffusion model, so as to include the one phonon expression has also been given. For a delayed diffusion model a complete treatment of S(K, ω) is presented. Along the lines of the macroscopic diffusion cooling, a microscopic diffusion cooling effect in fluids is speculated

  8. General classification and analysis of neutron β-decay experiments

    International Nuclear Information System (INIS)

    Gudkov, V.; Greene, G.L.; Calarco, J.R.

    2006-01-01

    A general analysis of the sensitivities of neutron β-decay experiments to manifestations of possible interaction beyond the standard model is carried out. In a consistent fashion, we take into account all known radiative and recoil corrections arising in the standard model. This provides a description of angular correlations in neutron decay in terms of one parameter, which is accurate to the level of ∼10 -5 . Based on this general expression, we present an analysis of the sensitivities to new physics for selected neutron decay experiments. We emphasize that the usual parametrization of experiments in terms of the tree-level coefficients a,A, and B is inadequate when the experimental sensitivities are at the same or higher level relative to the size of the corrections to the tree-level description

  9. Multiplicity counting from fission detector signals with time delay effects

    Science.gov (United States)

    Nagy, L.; Pázsit, I.; Pál, L.

    2018-03-01

    In recent work, we have developed the theory of using the first three auto- and joint central moments of the currents of up to three fission chambers to extract the singles, doubles and triples count rates of traditional multiplicity counting (Pázsit and Pál, 2016; Pázsit et al., 2016). The objective is to elaborate a method for determining the fissile mass, neutron multiplication, and (α, n) neutron emission rate of an unknown assembly of fissile material from the statistics of the fission chamber signals, analogous to the traditional multiplicity counting methods with detectors in the pulse mode. Such a method would be an alternative to He-3 detector systems, which would be free from the dead time problems that would be encountered in high counting rate applications, for example the assay of spent nuclear fuel. A significant restriction of our previous work was that all neutrons born in a source event (spontaneous fission) were assumed to be detected simultaneously, which is not fulfilled in reality. In the present work, this restriction is eliminated, by assuming an independent, identically distributed random time delay for all neutrons arising from one source event. Expressions are derived for the same auto- and joint central moments of the detector current(s) as in the previous case, expressed with the singles, doubles, and triples (S, D and T) count rates. It is shown that if the time-dispersion of neutron detections is of the same order of magnitude as the detector pulse width, as they typically are in measurements of fast neutrons, the multiplicity rates can still be extracted from the moments of the detector current, although with more involved calibration factors. The presented formulae, and hence also the performance of the proposed method, are tested by both analytical models of the time delay as well as with numerical simulations. Methods are suggested also for the modification of the method for large time delay effects (for thermalised neutrons).

  10. Development of ITER diagnostics: Neutronic analysis and radiation hardness

    Energy Technology Data Exchange (ETDEWEB)

    Vukolov, Konstantin, E-mail: vukolov_KY@nrcki.ru; Borisov, Andrey; Deryabina, Natalya; Orlovskiy, Ilya

    2015-10-15

    Highlights: • Problems of ITER diagnostics caused by neutron radiation from hot DT plasma considered. • Careful neutronic analysis is necessary for ITER diagnostics development. • Effective nuclear shielding for ITER diagnostics in the 11th equatorial port plug proposed. • Requirements for study of radiation hardness of diagnostic elements defined. • Results of optical glasses irradiation tests in a fission reactor given. - Abstract: The paper is dedicated to the problems of ITER diagnostics caused by effects of radiation from hot DT plasma. An effective nuclear shielding must be arranged in diagnostic port plugs to meet the nuclear safety requirements and to provide reliable operation of the diagnostics. This task can be solved with the help of neutronic analysis of the diagnostics environment within the port plugs at the design stage. Problems of neutronic calculations are demonstrated for the 11th equatorial port plug. The numerical simulation includes the calculations of neutron fluxes in the port-plug and in the interspace. Options for nuclear shielding, such as tungsten collimator, boron carbide and water moderators, stainless steel and lead screens are considered. Data on neutron fluxes along diagnostic labyrinths allow to define radiation hardness requirements for the diagnostic components and to specify their materials. Options for windows and lenses materials for optical diagnostics are described. The results of irradiation of flint and silica glasses in nuclear reactor have shown that silica KU-1 and KS-4V retain transparency in visible range after neutron fluence of 10{sup 17} cm{sup −2}. Flints required for achromatic objectives have much less radiation hardness about 5 × 10{sup 14} n/cm{sup 2}.

  11. A device for simultaneous spin analysis of ultracold neutrons

    Energy Technology Data Exchange (ETDEWEB)

    Afach, S. [Institute for Particle Physics, ETH Zuerich, Zuerich (Switzerland); Paul Scherrer Institute, Villigen-PSI (Switzerland); Jena University Hospital, Hans Berger Department of Neurology, Jena (Germany); Ban, G.; Lefort, T.; Lemiere, Y.; Naviliat-Cuncic, O.; Quemener, G. [Universite de Caen, CNRS/IN2P3, LPC Caen ENSICAEN, Caen (France); Bison, G.; Chowdhuri, Z.; Daum, M.; Henneck, R.; Lauss, B.; Mtchedlishvili, A.; Schmidt-Wellenburg, P.; Zsigmond, G. [Paul Scherrer Institute, Villigen-PSI (Switzerland); Bodek, K.; Rawlik, M.; Rozpedzik, D.; Zejma, J. [Jagiellonian University, Marian Smoluchowski Institute of Physics, Cracow (Poland); Fertl, M.; Franke, B.; Kirch, K.; Komposch, S. [Institute for Particle Physics, ETH Zuerich, Zuerich (Switzerland); Paul Scherrer Institute, Villigen-PSI (Switzerland); Geltenbort, P. [Institut Laue-Langevin, Grenoble (France); Grujic, Z.D.; Kasprzak, M.; Weis, A. [University of Fribourg, Physics Department, Fribourg (Switzerland); Hayen, L.; Severijns, N.; Wursten, E. [Katholieke Universiteit Leuven, Instituut voor Kernen Stralingsfysica, Leuven (Belgium); Helaine, V. [Paul Scherrer Institute, Villigen-PSI (Switzerland); Universite de Caen, CNRS/IN2P3, LPC Caen ENSICAEN, Caen (France); Kermaidic, Y.; Pignol, G.; Rebreyend, D. [Universite Grenoble Alpes, CNRS/IN2P3, LPSC, Grenoble (France); Kozela, A. [Henryk Niedwodniczanski Institute for Nuclear Physics, Cracow (Poland); Krempel, J.; Piegsa, F.M. [Institute for Particle Physics, ETH Zuerich, Zuerich (Switzerland); Prashanth, P.N. [Paul Scherrer Institute, Villigen-PSI (Switzerland); Katholieke Universiteit Leuven, Instituut voor Kernen Stralingsfysica, Leuven (Belgium); Ries, D. [Paul Scherrer Institute, Villigen-PSI (Switzerland); Jena University Hospital, Hans Berger Department of Neurology, Jena (Germany); Roccia, S. [Universite Paris Sud, CNRS/IN2P3, CSNSM, Orsay campus (France); Wyszynski, G. [Institute for Particle Physics, ETH Zuerich, Zuerich (Switzerland); Jagiellonian University, Marian Smoluchowski Institute of Physics, Cracow (Poland)

    2015-11-15

    We report on the design and first tests of a device allowing for measurement of ultracold neutrons polarisation by means of the simultaneous analysis of the two spin components. The device was developed in the framework of the neutron electric dipole moment experiment at the Paul Scherrer Institute. Individual parts and the entire newly built system have been characterised with ultracold neutrons. The gain in statistical sensitivity obtained with the simultaneous spin analyser is (18.2 ± 6.1) % relative to the former sequential analyser under nominal running conditions. (orig.)

  12. Activation analysis by filtered neutrons. Preliminary investigation

    International Nuclear Information System (INIS)

    Skarnemark, G.; Rodinson, T.; Skaalberg, M.; Tokay, R.K.

    1986-01-01

    In order to investigate if measuring sensibility and precision by epithermal neutron activation analysis may be improved, different types of geological and biologic test samples were radiated. The test samples were enclosed in an extra filter of tungsten or sodium in order to reduce the flux of those neutrons that otherwise would induce interfering activity in the sample. The geological test samples consist of granites containing lanthanides which had been crushed in tung- sten carbide grinder. Normally such test samples show a interferins 1 87W-activity. By use of a tungsten filter the activity was reduced by up to 60%, which resulted in a considerable improvement of sensibility and precision of the measurement. The biologic test samples consisted of evaporated urine from patients treated with the cell poison cis-platinol. A reliable method to measure the platinum content has not existed so far. This method, however, enables platinum contents as low as about 0.1 ppm to be determined which is quite adequate. To sum up this preliminary study has demonstrated that activation analysis using filtered neutrons, correctly applied, is a satisfactory method of reducing interferences without complicated and time-consuming chemical separation procedures. (O.S.)

  13. SWAN - Detection of explosives by means of fast neutron activation analysis

    International Nuclear Information System (INIS)

    Gierlik, M.; Borsuk, S.; Guzik, Z.; Iwanowska, J.; Kaźmierczak, Ł.; Korolczuk, S.; Kozłowski, T.; Krakowski, T.; Marcinkowski, R.; Swiderski, L.; Szeptycka, M.; Szewiński, J.; Urban, A.

    2016-01-01

    In this work we report on SWAN, the experimental, portable device for explosives detection. The device was created as part of the EU Structural Funds Project “Accelerators & Detectors” (POIG.01.01.02-14-012/08-00), with the goal to increase beneficiary's expertise and competencies in the field of neutron activation analysis. Previous experiences and budged limitations lead toward a less advanced design based on fast neutron interactions and unsophisticated data analysis with the emphasis on the latest gamma detection and spectrometry solutions. The final device has been designed as a portable, fast neutron activation analyzer, with the software optimized for detection of carbon, nitrogen and oxygen. SWAN's performance in the role of explosives detector is elaborated in this paper. We demonstrate that the unique features offered by neutron activation analysis might not be impressive enough when confronted with practical demands and expectations of a generic homeland security customer.

  14. Nuclear data needs for material analysis

    International Nuclear Information System (INIS)

    Molnar, Gabor L.

    2001-01-01

    Nuclear data for material analysis using neutron-based methods are examined. Besides a critical review of the available data, emphasis is given to emerging application areas and new experimental techniques. Neutron scattering and reaction data, as well as decay data for delayed and prompt gamma activation analysis are all discussed in detail. Conclusions are formed concerning the need of new measurement, calculation, evaluation and dissemination activities. (author)

  15. The TENDL neutron data library and the TEND1038 38-group neutron constant system

    International Nuclear Information System (INIS)

    Abramovich, S.N.; Gorelov, V.P.; Gorshikhin, A.A.; Grebennikov, A.N.; Il'in, V.N.; Krut'ko, N.A.; Farafontov, G.G.

    2002-01-01

    The library contains neutron data for 103 nuclei - i.e. for 38 actinide nuclei (from 232 Th to 249 Cm), 26 fission fragment nuclei and 39 nuclei in structural and technological materials. The 38-group constants were obtained from TENDL. The high-energy group boundary is 20 MeV. The energy range below 1.2 eV contains 11 groups. Temperature and resonance effects were taken into account. The delayed neutron parameters for 6 groups and the yields of 40 fission fragments were obtained (light and heavy, stable and non-stable). The fast neutron features of spherical critical assemblies were calculated using constants from TEND1038. (author)

  16. Fast neutron activation analysis at Texas A and M University

    International Nuclear Information System (INIS)

    James, W.D.

    1997-01-01

    Fast neutron generators are used at Texas A and M University to provide a supply of high energy neutrons for nuclear analytical measurements. A series of neutron activation analysis procedures have been developed for determining various major, minor and trace constituents in a variety of materials. These procedures are primarily developed to compliment our reactor based NAA program, thereby expanding the list of determinable elements to include those difficult or impossible to measure using thermal neutrons. A few typical methods are discussed. The unique implementation of the methodologies at Texas A and M are explained. (author)

  17. Non destructive multi elemental analysis using prompt gamma neutron activation analysis techniques: Preliminary results for concrete sample

    Energy Technology Data Exchange (ETDEWEB)

    Dahing, Lahasen Normanshah [School of Applied Physics, Universiti Kebangsaan Malaysia, 43600 Bangi, Selangor, Malaysia and Malaysian Nuclear Agency (Nuklear Malaysia), Bangi 43000, Kajang (Malaysia); Yahya, Redzuan [School of Applied Physics, Universiti Kebangsaan Malaysia, 43600 Bangi, Selangor (Malaysia); Yahya, Roslan; Hassan, Hearie [Malaysian Nuclear Agency (Nuklear Malaysia), Bangi 43000, Kajang (Malaysia)

    2014-09-03

    In this study, principle of prompt gamma neutron activation analysis has been used as a technique to determine the elements in the sample. The system consists of collimated isotopic neutron source, Cf-252 with HPGe detector and Multichannel Analysis (MCA). Concrete with size of 10×10×10 cm{sup 3} and 15×15×15 cm{sup 3} were analysed as sample. When neutrons enter and interact with elements in the concrete, the neutron capture reaction will occur and produce characteristic prompt gamma ray of the elements. The preliminary result of this study demonstrate the major element in the concrete was determined such as Si, Mg, Ca, Al, Fe and H as well as others element, such as Cl by analysis the gamma ray lines respectively. The results obtained were compared with NAA and XRF techniques as a part of reference and validation. The potential and the capability of neutron induced prompt gamma as tool for multi elemental analysis qualitatively to identify the elements present in the concrete sample discussed.

  18. Neutron activation analysis for noble metals in matte leach residues

    International Nuclear Information System (INIS)

    Hart, R.J.

    1978-01-01

    The development of the neutron activation analysis technique as a method for rapid and precise determinations of platinum group metals in matte leach residues depends on obtaining a method for effecting complete and homogeneous sample dilution. A simple method for solid dilution of metal samples is outlined in this study, which provided a basis for the accurate determination of all the noble metals by the Neutron Activation Analysis technique

  19. Neutron activation analysis

    International Nuclear Information System (INIS)

    Borsaru, M.; Eisler, P.L.

    1981-01-01

    A method of simultaneously analysing the aluminium and silicon content of a sample of material is claimed. The method comprises the following steps: (1) irradiating the sample with fast neutrons; (2) monitoring the thermal neutron flux within the sample; (3) monitoring the gamma radiation from the irradiated sample at energies of 1.78 MeV and 1.015 and/or 0.844 MeV; (4) using the monitored gamma radiation at 1.015 and/or 0.844 MeV to estimate the aluminium content of the sample; and (5) using the monitored gamma radiation at 1.78 MeV, compensated by the gamma radiation at 1.78 MeV due to the thermal neutron reaction with the estimated aluminium in the sample to estimate the silicon content

  20. Detection of fast neutrons in a plastic scintillator using digital pulse processing to reject gammas

    International Nuclear Information System (INIS)

    Reeder, P.L.; Peurrung, A.J.; Hansen, R.R.; Stromswold, D.C.; Hensley, W.K.; Hubbard, C.W.

    1999-01-01

    We report on neutron-gamma discrimination in a plastic scintillator based on the time delay inherent in second and third chance neutron scattering. Because of the time delay (∼3 ns) between the first and second scattering of a neutron, calculations of gammas and neutrons in a plastic scintillator predict that a neutron signal should be significantly broader than a pulse from a gamma event. Experimentally, we have used a fast digital oscilloscope coupled to a computer to examine individual pulses from neutron or gamma induced signals in fast scintillators coupled to a fast PMT. Individual neutron-induced signals were consistent with the predictions of our model, but gamma pulses were broader than expected. We present various tests to understand this phenomenon and discuss a way to overcome this problem

  1. Neutron activation analysis of biological substances

    International Nuclear Information System (INIS)

    Ordogh, M.

    1978-08-01

    A Bowen cabbage sample was used as a reference material for the neutron activation studies, and the method was checked by the analysis of other biological substances (blood or serum etc.). For nondestructive measurements also some non-trace elements were determined in order to decide whether the activation analysis is a useful means for such measurements. The new activation analysis procedure was used for biomedical studies as, e.g., for trace element determination in body fluids, and for the analysis of inorganic components in air samples. (R.P.)

  2. Analysis of the neutron generation from a D-Li neutron source

    International Nuclear Information System (INIS)

    Gomes, I.

    1994-02-01

    The study of the neutron generation from the D-Li reaction is an important issue to define the optimum combination of the intervening parameters during the design phase of a D-Li neutron source irradiation facility. The major players in defining the neutron yield from the D-Li reaction are the deuteron incident energy and the beam current, provided that the lithium target is thick enough to stop all incident deuterons. The incident deuteron energy also plays a role on the angular distribution of the generated neutrons, on the energy distribution of the generated neutrons, and on the maximum possible energy of the neutrons. The D-Li reaction produces neutrons with energies ranging from eV's to several MeV's. The angular distribution of these neutrons is dependent on the energy of both, incident deuterons and generated neutrons. The deuterons lose energy interacting with the lithium target material in such a way that the energy of the deuterons inside the lithium target varies from the incident deuteron energy to essentially zero. The first part of this study focuses in analyzing the neutron generation rate from the D-Li reaction as a function of the intervening parameters, in defining the source term, in terms of the energy and angular distributions of the generated neutrons, and finally in providing some insights of the impact of varying input parameters on the generation rate and correlated distributions. In the second part an analytical description of the Monte Carlo sampling procedure of the neutron from the D-Li reaction is provided with the aim at further Monte Carlo transport of the D-Li neutrons

  3. Identification and systematical studies of the electron-capture delayed fission (ECDF) in the lead region

    CERN Multimedia

    Pauwels, D B; Lane, J

    2008-01-01

    In our recent experiment (March 2007) at the velocity filter SHIP(GSI) we observed the electron-capture delayed fission of the odd-odd isotope $^{194}$At. This is the first unambiguous identification of this phenomenon in the very neutron-deficient nuclei in the vicinity of the proton shell closure at Z=82. In addition, the total kinetic energy (TKE) for the daughter nuclide $^{194}$Po was measured, despite the fact that this isotope does not decay via spontaneous fission. Semi-empirical analysis of the electron-capture Q$_{EC}$ values and fission barriers B$_{f}$ shows that a relatively broad island of ECDF must exist in this region of the Nuclide Chart, with some of the nuclei having unusually high ECDF probabilities. Therefore, this Proposal is intended to initiate the systematic identification and study of $\\beta$-delayed fission at ISOLDE in the very neutron-deficient lead region. Our aim is to provide unique low-energy fission data (e.g. probabilities, TKE release, fission barriers and their isospin dep...

  4. A compact neutron beam generator system designed for prompt gamma nuclear activation analysis.

    Science.gov (United States)

    Ghassoun, J; Mostacci, D

    2011-08-01

    In this work a compact system was designed for bulk sample analysis using the technique of PGNAA. The system consists of (252)Cf fission neutron source, a moderator/reflector/filter assembly, and a suitable enclosure to delimit the resulting neutron beam. The moderator/reflector/filter arrangement has been optimised to maximise the thermal neutron component useful for samples analysis with a suitably low level of beam contamination. The neutron beam delivered by this compact system is used to irradiate the sample and the prompt gamma rays produced by neutron reactions within the sample elements are detected by appropriate gamma rays detector. Neutron and gamma rays transport calculations have been performed using the Monte Carlo N-Particle transport code (MCNP5). 2010 Elsevier Ltd. All rights reserved.

  5. Study on neutron activation analysis

    International Nuclear Information System (INIS)

    Chung, Yong Sam; Cho, Seung Yeon

    1993-01-01

    Environmental samples were analyzed quantitatively by neutron activation analysis using high resolution γ-ray spectrometry. The accuracy and precision of the method were checked by the analysis of reference materials, Urban Particulate Matter (NBS SRM 1648) and Coalfly ash (NBS SRM 1633a). Airborne particulates collected for 6 months with low volume air sampler at the outer area of Seoul were analyzed as the start of full scale airborne particulates research. We analyzed 19 trace elements from the samples and the NAA method was confirmed to be utilized for environmental pollution research. (Author)

  6. Materials characterization of radioactive waste forms using a multi-element detection method based on the instrumental neutron activation analysis. MEDINA

    International Nuclear Information System (INIS)

    Havenith, Andreas Wilhelm

    2015-01-01

    the identification and quantification of toxic elements in radioactive waste forms. The physical basis of MEDINA is the Prompt- and Delayed-Gamma-Neutron-Activation-Analysis (P and DGNAA). The neutron activation analysis of material samples in the gram range is state-of-the-art of science and technology under use of thermal or cold neutrons at research reactors. The thereof retrieved nuclear data and the results of the feasibility study for the characterization of large-volume samples up to a volume of 50 l /1-5/ are the scientific basis of the present dissertation. With a newly developed test facility and an innovative algorithms for a rotationally dependent analysis the element quantification of larger inhomogeneous samples can be performed by taking into account the gamma and neutron self-shielding for the first time. A test facility for the chemical characterisation of 200-l-drums was built and several homogeneous and inhomogeneous samples with a waste matrix of concrete were analysed to validate the measurement technique. The conceptual design of the MEDINA test facility is based on stochastic simulations studies with the computer code MCNP. For a measurement the drum of interest is positioned on a turntable inside an irradiation chamber made exclusively of graphite, acting as neutron moderator and reflector. The drum is irradiated with 14 MeV neutrons produced by a deuterium-tritium (D-T) neutron-generator operating in pulse mode. The prompt and delayed gamma rays, induced by neutron reactions occurring at different times after the neutron pulses, are measured with a high-purity germanium (HPGe) detector placed in a wall of the irradiation chamber perpendicular to the neutron generator. The HPGe detector signals are processed through an appropriate nuclear electronics. The gamma rays spectra are recorded for each discrete drum rotation, which allows to investigate the sample homogeneity. The developed algorithm for the element quantification is based on the

  7. Reactor kinetics calculated in the summation method and key delayed-neutron data

    International Nuclear Information System (INIS)

    Oyamatsu, Kazuhiro

    2001-01-01

    The point-reactor kinetics after a step reactivity insertion to a critical condition is solved directly form fission-product (FP) data (fission yields and decay data) for the first time. Numerical calculations are performed with the FP data in ENDF/B-VI. The inhour equation obtained directly from the FP data shows a different behavior at long periods from the one obtained from Tuttle's six-group parameter sets. The behavior is quite similar to the one obtained from the six-group parameter sets in ENDF/B-VI, that were obtained from FP data in a preliminary version of ENDF/B-VI. To identify the erroneous FP data, we examine the asymptotic form of the inhour equation at an infinitely long period. It is found that the most important precursors for long reactor periods are found 137 I, 88 Br and 87 Br. They cover more than 60% of the reactivity. It is remarkable that 137 I alone covers 30-50% depending on the fissioning system. In addition to the three precursors, 136 Te is found a candidate precursor for the peculiarity from the time dependence of the delayed neutron activity. It is recommended that the precision of their Pn values should be improved experimentally. For 137 I, 88 Br, and 87 Br, the relative uncertainty, dPn/Pn, should be decreased down to 2% and for 136 Te to 5%. (author)

  8. Bayesian statistics applied to neutron activation data for reactor flux spectrum analysis

    International Nuclear Information System (INIS)

    Chiesa, Davide; Previtali, Ezio; Sisti, Monica

    2014-01-01

    Highlights: • Bayesian statistics to analyze the neutron flux spectrum from activation data. • Rigorous statistical approach for accurate evaluation of the neutron flux groups. • Cross section and activation data uncertainties included for the problem solution. • Flexible methodology applied to analyze different nuclear reactor flux spectra. • The results are in good agreement with the MCNP simulations of neutron fluxes. - Abstract: In this paper, we present a statistical method, based on Bayesian statistics, to analyze the neutron flux spectrum from the activation data of different isotopes. The experimental data were acquired during a neutron activation experiment performed at the TRIGA Mark II reactor of Pavia University (Italy) in four irradiation positions characterized by different neutron spectra. In order to evaluate the neutron flux spectrum, subdivided in energy groups, a system of linear equations, containing the group effective cross sections and the activation rate data, has to be solved. However, since the system’s coefficients are experimental data affected by uncertainties, a rigorous statistical approach is fundamental for an accurate evaluation of the neutron flux groups. For this purpose, we applied the Bayesian statistical analysis, that allows to include the uncertainties of the coefficients and the a priori information about the neutron flux. A program for the analysis of Bayesian hierarchical models, based on Markov Chain Monte Carlo (MCMC) simulations, was used to define the problem statistical model and solve it. The first analysis involved the determination of the thermal, resonance-intermediate and fast flux components and the dependence of the results on the Prior distribution choice was investigated to confirm the reliability of the Bayesian analysis. After that, the main resonances of the activation cross sections were analyzed to implement multi-group models with finer energy subdivisions that would allow to determine the

  9. Stability Analysis of Fractional-Order Nonlinear Systems with Delay

    Directory of Open Access Journals (Sweden)

    Yu Wang

    2014-01-01

    Full Text Available Stability analysis of fractional-order nonlinear systems with delay is studied. We propose the definition of Mittag-Leffler stability of time-delay system and introduce the fractional Lyapunov direct method by using properties of Mittag-Leffler function and Laplace transform. Then some new sufficient conditions ensuring asymptotical stability of fractional-order nonlinear system with delay are proposed firstly. And the application of Riemann-Liouville fractional-order systems is extended by the fractional comparison principle and the Caputo fractional-order systems. Numerical simulations of an example demonstrate the universality and the effectiveness of the proposed method.

  10. The Aviation System Analysis Capability Airport Capacity and Delay Models

    Science.gov (United States)

    Lee, David A.; Nelson, Caroline; Shapiro, Gerald

    1998-01-01

    The ASAC Airport Capacity Model and the ASAC Airport Delay Model support analyses of technologies addressing airport capacity. NASA's Aviation System Analysis Capability (ASAC) Airport Capacity Model estimates the capacity of an airport as a function of weather, Federal Aviation Administration (FAA) procedures, traffic characteristics, and the level of technology available. Airport capacity is presented as a Pareto frontier of arrivals per hour versus departures per hour. The ASAC Airport Delay Model allows the user to estimate the minutes of arrival delay for an airport, given its (weather dependent) capacity. Historical weather observations and demand patterns are provided by ASAC as inputs to the delay model. The ASAC economic models can translate a reduction in delay minutes into benefit dollars.

  11. Neutron kinetics of fluid-fuel systems by the quasi-static method

    International Nuclear Information System (INIS)

    Dulla, S.; Ravetto, P.; Rostagno, M.M.

    2004-01-01

    The quasi-static method for the neutron kinetics of nuclear reactors is generalized for application to neutron multiplying systems fueled by a fluid multiplying material, typically a mixture of fissile molten salts. The method is derived by the application of factorization formulae for both the neutron density and the delayed precursor concentrations and the projection of the balance equations upon a weighting function. A physically meaningful weight can be assumed as the solution of the adjoint model, which is constructed for the situation considered, including delayed neutrons. The quasi-static scheme is then applied to calculations of some transients for a typical configuration of a molten-salt reactor, in a multigroup diffusion model with a one-dimensional slug-flow velocity field. The physical features associated to the motion of the fissile material are highlighted

  12. SWAN - Detection of explosives by means of fast neutron activation analysis

    Energy Technology Data Exchange (ETDEWEB)

    Gierlik, M., E-mail: m.gierlik@ncbj.gov.pl; Borsuk, S.; Guzik, Z.; Iwanowska, J.; Kaźmierczak, Ł.; Korolczuk, S.; Kozłowski, T.; Krakowski, T.; Marcinkowski, R.; Swiderski, L.; Szeptycka, M.; Szewiński, J.; Urban, A.

    2016-10-21

    In this work we report on SWAN, the experimental, portable device for explosives detection. The device was created as part of the EU Structural Funds Project “Accelerators & Detectors” (POIG.01.01.02-14-012/08-00), with the goal to increase beneficiary's expertise and competencies in the field of neutron activation analysis. Previous experiences and budged limitations lead toward a less advanced design based on fast neutron interactions and unsophisticated data analysis with the emphasis on the latest gamma detection and spectrometry solutions. The final device has been designed as a portable, fast neutron activation analyzer, with the software optimized for detection of carbon, nitrogen and oxygen. SWAN's performance in the role of explosives detector is elaborated in this paper. We demonstrate that the unique features offered by neutron activation analysis might not be impressive enough when confronted with practical demands and expectations of a generic homeland security customer.

  13. High-sensitivity fast neutron detector KNK-2-7M

    Energy Technology Data Exchange (ETDEWEB)

    Koshelev, A. S., E-mail: alexsander.coshelev@yandex.ru; Dovbysh, L. Ye.; Ovchinnikov, M. A.; Pikulina, G. N.; Drozdov, Yu. M. [Russian Federal Nuclear Center All-Russian Research Institute of Experimental Physics (Russian Federation); Chuklyaev, S. V. [Research Institute of Materials Technology (Russian Federation)

    2015-12-15

    The construction of the fast neutron detector KNK-2-7M is briefly described. The results of the study of the detector in the pulse-counting mode are given for the fissions of {sup 237}Np nuclei in the radiator of the neutron-sensitive section and in the current mode with the separation of sectional currents of functional sections. The possibilities of determining the effective number of {sup 237}Np nuclei in the radiator of the neutronsensitive section are considered. The diagnostic possibilities of the detector in the counting mode are shown by example of the analysis of the reference data from the neutron-field characteristics in the working hall of the BR-K1 reactor. The diagnostic possibilities of the detector in the current operating mode are shown by example of the results of measuring the {sup 237}Np-fission intensity in the BR-K1 reactor power start-ups implemented in the mode of fission-pulse generation on delayed neutrons at the detector arrangement inside the reactor core cavity under conditions of a wide variation of the reactor radiation field.

  14. Preliminary study of elemental analysis of hydroxyapatite used neutron activation analysis method

    International Nuclear Information System (INIS)

    Yustinus Purwamargapratala; Rina Mulyaningsih

    2010-01-01

    Preliminary study has been carried out elemental analysis of hydroxyapatite synthesized using the method of neutron activation analysis. Hydroxyapatite is the main component constituent of bones and teeth which can be synthesized from limestone and phosphoric acid. Hydroxyapatite can be used as a bone substitute material and human and animal teeth. Tests on the metal content is necessary to prevent the risk of damage to bones and teeth due to contamination. Results of analysis using neutron activation analysis method with samples irradiated at the neutron flux 10"3 n.det"-"1cm"-"2 for one minute, the impurities of Al (48.60±6.47 mg/kg), CI (38.00±7.47 mg/kg), Mn (1.05±0.19 mg/kg), and Mg (2095.30±203.66 mg/kg), were detected, whereas with irradiation time for 10 minutes and 40 minutes with a time decay of three days there were K (103.89 ± 26.82 mg/kg), Br (1617.06 ± 193.66 mg/kg), and Na (125.10±9.57 mg/kg). These results indicate that there is impurity Al, CI, Mn, Mg, Br, K and Na, although in very small amounts and do not cause damage to bones and teeth. (author)

  15. Neutron activation analysis of Etruscan pottery

    International Nuclear Information System (INIS)

    Whitehead, J.; Silverman, A.; Ouellet, C.G.; Clark, D.D.; Hossain, T.Z.

    1992-01-01

    Neutron activation analysis (NAA) has been widely used in archaeology for compositional analysis of pottery samples taken from sites of archaeological importance. Elemental profiles can determine the place of manufacture. At Cornell, samples from an Etruscan site near Siena, Italy, are being studied. The goal of this study is to compile a trace element concentration profile for a large number of samples. These profiles will be matched with an existing data bank in an attempt to understand the place of origin for these samples. The 500 kW TRIGA reactor at the Ward Laboratory is used to collect NAA data for these samples. Experiments were done to set a procedure for the neutron activation analysis with respect to sample preparation, selection of irradiation container, definition of activation and counting parameters and data reduction. Currently, we are able to analyze some 27 elements in samples of mass 500 mg with a single irradiation of 4 hours and two sequences of counting. Our sensitivity for many of the trace elements is better than 1 ppm by weight under the conditions chosen. In this talk, details of our procedure, including quality assurance as measured by NIST standard reference materials, will be discussed. In addition, preliminary results from data treatment using cluster analysis will be presented. (author)

  16. Heterogeneous analysis of non-uniform neutron field formation

    International Nuclear Information System (INIS)

    Zagrebaev, A.M.; Fedosov, A.M.

    1979-01-01

    Investigated are the specific features of spatial-energy neutron distribution formation in the transient zone between regions, operating at different levels of energy release with accounting for the real structure of fuel element lattice and control elements in the channel reactors of high power. Presented are the calculation results, obtained by heterogeneous method in the two-group monopole approximation by means of the HETLAT code. The analysis, based on the homogeneous model shows, that the efficiency of the transient zone in forming neutron flux qradient can be increased by introducing an additional interlayer of moderator between the layers with extreme multiplying properties. It is stressed, that the most favourable from the point of view of energy release uniformity in zones and width of the transient zone is the variant in which neutron flux gradient is carried out by moving the control elements on the boundaries of regions while the internal rows of control elements create the conditions for flattening the energy release in the zones. The result obtained corresponds to the recommendation on optimal control, coming from the Pontryagin maximum principle. The analysis of neutron field formation using heterogeneous models mainly proves the conclusions following from homogeneous calculations using the maximum principle. At the same time quantitative results for the zones of small dimensions (less than 10 migration lengths) with a vividly expressed heterogeneous structure essentially differ from the forecast, obtained on the basis of the simplified homogeneous one-group model. The heterogeneous analysis shows possibilities for further optimization of the transient zone structure with account of the control element location

  17. Calculation of thermal neutron self-shielding correction factors for aqueous bulk sample prompt gamma neutron activation analysis using the MCNP code

    International Nuclear Information System (INIS)

    Nasrabadi, M.N.; Jalali, M.; Mohammadi, A.

    2007-01-01

    In this work thermal neutron self-shielding in aqueous bulk samples containing neutron absorbing materials is studied using bulk sample prompt gamma neutron activation analysis (BSPGNAA) with the MCNP code. The code was used to perform three dimensional simulations of a neutron source, neutron detector and sample of various material compositions. The MCNP model was validated against experimental measurements of the neutron flux performed using a BF 3 detector. Simulations were performed to predict thermal neutron self-shielding in aqueous bulk samples containing neutron absorbing solutes. In practice, the MCNP calculations are combined with experimental measurements of the relative thermal neutron flux over the sample's surface, with respect to a reference water sample, to derive the thermal neutron self-shielding within the sample. The proposed methodology can be used for the determination of the elemental concentration of unknown aqueous samples by BSPGNAA where knowledge of the average thermal neutron flux within the sample volume is required

  18. Thermogravimetric analysis of reactor-neutrons-irradiated LEXAN polycarbonate film

    International Nuclear Information System (INIS)

    Kalsi, P.C.

    2000-01-01

    The effects of reactor-neutrons irradiation on the thermogravimetric (TG) analysis of LEXAN polycarbonate film in air were studied. Irradiation enhances the degradation rate and the effect increases further with increasing neutron fluence. The kinetics of the different steps of degradation were also evaluated from the TG curves. The activation energy values calculated for all the degradation stages decrease on irradiation. (author)

  19. Neutronic and nuclear post-test analysis of MEGAPIE

    Energy Technology Data Exchange (ETDEWEB)

    Zanini, L.; Aebersold, H. U.; Berg, K.; Eikenberg, J.; Filges, U.; Groeschel, F.; Luethy, M.; Ruethi, M.; Scazzi, S.; Tobler, L.; Wagner, W.; Wernli, B. [Paul Scherrer Institute (PSI), Villigen (Switzerland); Panebianco, S.; David, J.-C.; Dore, D.; Lemaire, S.; Leray, S.; Letourneau, A.; Michel-Sendis, F.; Prevost, A.; Ridikas, D.; Stankunas, G. [CEA, Centre de Saclay, IRFU/Service de Physique Nucleaire, Gif-sur-Yvette (France); Toussaint, J.-C. [CEA, Centre de Saclay, IRFU/Service d' Ingenierie des Systemes, Gif-sur-Yvette (France); Eid, M. [CEA, Centre de Saclay, DEN/DM2S/SERMA, Gif-sur-Yvette (France); Latge, C. [CEA, Centre de Cadarache, DEN/DTN/DIR, Saint Paul Lez, Durance (France); Konobeyev, A. Yu.; Fischer, U. [Institut fuer Reaktorsichereit, Forschungszentrum Karlsruhe Gmbh, Karlsruhe (Germany); Thiolliere, N.; Guertin, A. [SUBATECH Laboratory, CNRS/IN2P3-EMN-University, Nantes (France); Buchillier, T.; Bailat, C. [Institut universitaire de radiophysique appliquee (IRA), Lausanne (Switzerland)

    2008-12-15

    and MEGAPIE has been correctly reproduced. To achieve a good accuracy in the calculation of the neutronic performance of an ADS system, an accurate definition of the geometrical model taking into account the influence of structural materials is of primary importance. The results depend also on the beam profile used in the simulations, at least for the flux calculations close to the target interaction point. Radioactive nuclides produced in liquid metal targets are transported into hot cells, pumps or close to electronics with radiation sensitive components. Besides the considerable amount of decay {gamma} activity in the irradiated liquid metal, a significant amount of the Delayed Neutron (DN) precursor activity accumulates in the target fluid. The transit time of a liquid metal target being as short as a few seconds, DNs may contribute significantly to the activation and dose rates. The importance of the DN issues in liquid metal targets is confirmed. Another problem is the gas production and release in an ADS target, the proton beam generating a large amount of gas by spallation reactions. A large amount of Po isotopes, volatile at relatively high temperatures, are produced in the LBE. The gas production was measured by {gamma} spectroscopy. The release rates of noble gases in MEGAPIE are at the % level after 1-2 days of operation, while the release becomes almost complete weeks after the beginning of operation. Pressure increase in the cover gas could be reproduced with calculations within a factor of 2. The effect of the impurities in the radionuclide inventory of the LBE, using the actual chemical composition of the LBE used in MEGAPIE, is minimal.

  20. Organic Waste Composts, a Serious Rare- Earth Source as Determined by Neutron Activation Analysis

    International Nuclear Information System (INIS)

    Sroor, A.; El-Bahi, S.M.; Abdel-Halieem, A.S.; Abdel-Sabour, M.F.

    1999-01-01

    Delayed Neutron Activation Analysis technique [DNNA] was applied for investigating rare-earth elements and some heavy metals content of some locally organic fertilizers namely cattle manure (CM) , dried sewage sludge [SS] , municipal solid waste [MSW] and mixture for a (SS+MSW). The γ-ray spectrum of each sample was investigated using a HPGe detector equipped with computer unit. Fourteen elements were determined. Some of them were confirmed by the γ-γ cascades using a HPGe-HPGe coincidence spectrometer. The concentration of these elements in each sample was measured in μg/g. Some of these elements may lead to undesirable environmental effects. The undiscriminating use of organic waste as organic fertilizers may result in the increase of toxic elements [Cr, Sc, Sb, Th, etc.) in soil environment which may transfer through food chain to human health

  1. Report of the first United States conference on utility experience with neutron noise analysis

    International Nuclear Information System (INIS)

    Fry, D.N.; Horne, G.P.; Mayo, C.W.

    1984-01-01

    An informal meeting was held in Washington, D.C. on April 3 and 4, 1984, to discuss the current state of the art and experiences with neutron noise analysis in US pressurized water reactors (PWRs). The meeting was attended by 33 persons representing 11 utilities and 3 PWR reactor vendors as well as consultants, universities, and research laboratories. Presentations at the meeting covered several applications of neutron noise for diagnosing such things as vibrations induced by baffle jetting, detection of mechanical degradation of thermal shield supports, and electrical degradation of nuclear instrumentation channels. Twenty-one responses were obtained from a questionnaire circulated to all participants requesting their viewpoints and experiences regarding neutron noise analysis. The meeting participants concluded that a working group on neutron noise analysis should be formed to (1) establish a baseline library of neutron noise data, (2) provide a forum for communicating experiences with neutron noise surveillance, and (3) develop good practices and quality assurance procedures for neutron noise measurement and interpretation

  2. Theory of stochastic space-dependent neutron kinetics with a Gaussian parametric excitation

    International Nuclear Information System (INIS)

    Saito, K.

    1980-01-01

    Neutron kinetics and statics in a multiplying medium with a statistically fluctuating reactivity are unified and systematically studied by applying the Novikov-Furutsu formula. The parametric or multiplicative noise is spatially distributed and of Gaussian nature with an arbitrary spectral profile. It is found that the noise introduces a new definite production term into the conventional balance equation for the mean neutron number. The term is characterized by the magnitude and the correlation function of the random excitation. Its relaxation phenomena bring forth a non-Markoffian or a memory effect, which is conceptualised by introducing 'pseudo-precursors' or 'pseudo-delayed neutrons'. By using the concept, some typical reactor physical problems are solved; they are (1) reactivity and flux perturbation originating from the random dispersal of core materials and (2) analysis of neutron decay mode and it relaxation constant, and derivation of the corresponding new inhour equation. (author)

  3. Dynamic Monte Carlo transient analysis for the Organization for Economic Co-operation and Development Nuclear Energy Agency (OECD/NEA) C5G7-TD benchmark

    Energy Technology Data Exchange (ETDEWEB)

    Shaukat, Nadeem; Ryu, Min; Shim, Hyung Jin [Seoul National University, Seoul (Korea, Republic of)

    2017-08-15

    With ever-advancing computer technology, the Monte Carlo (MC) neutron transport calculation is expanding its application area to nuclear reactor transient analysis. Dynamic MC (DMC) neutron tracking for transient analysis requires efficient algorithms for delayed neutron generation, neutron population control, and initial condition modeling. In this paper, a new MC steady-state simulation method based on time-dependent MC neutron tracking is proposed for steady-state initial condition modeling; during this process, prompt neutron sources and delayed neutron precursors for the DMC transient simulation can easily be sampled. The DMC method, including the proposed time-dependent DMC steady-state simulation method, has been implemented in McCARD and applied for two-dimensional core kinetics problems in the time-dependent neutron transport benchmark C5G7-TD. The McCARD DMC calculation results show good agreement with results of a deterministic transport analysis code, nTRACER.

  4. Neutronic activation analysis of antique ceramics. Groups and differenciation

    International Nuclear Information System (INIS)

    Widemann, F.

    1975-01-01

    Different techniques for clay analysis in view of studying the origin of ceramics are exposed. The element abundance is measured by X-ray fluorescence analysis or by neutron activation analysis. Comparative tables of the results are established [fr

  5. Speciation analysis of cobalt in foods by high-performance liquid chromatography and neutron activation analysis

    International Nuclear Information System (INIS)

    Muto, Toshio; Koyama, Motoko

    1994-01-01

    A combined method by coupling high-performance liquid chromatography (HPLC, as a separation method) with neutron activation analysis (as a detection method) have been applied to the speciation analysis of cobalt in daily foods (e.g. egg, fish and milk). Cobalt species including free cobalt, vitamin B 12 and protein-bound cobalt were separated with a preparative HPLC and a centrifuge. Subsequently, the determination of cobalt in the separated species was made by neutron activation analysis. The results showed that the content of the total cobalt in the foods was found to lie in the range 0.4-11ng/g(0.4-11ppb) based on wet weight. The compositions of free cobalt, vitamin B 12 and protein-bound cobalt were ranged 16-43%, 55-73%, 2.3-17%, respectively. These experimental evidences suggest that the combination of HPLC and neutron activation analysis is expected to be a useful tool for speciation analysis of trace elements in biological as well as environmental materials. (author)

  6. Neutronics analysis of International Fusion Material Irradiation Facility (IFMIF). Japanese contributions

    International Nuclear Information System (INIS)

    Oyama, Yukio; Noda, Kenji; Kosako, Kazuaki.

    1997-10-01

    In fusion reactor development for demonstration reactor, i.e., DEMO, materials tolerable for D-T neutron irradiation are absolutely required for both mechanical and safety point of views. For this requirement, several kinds of low activation materials were proposed. However, experimental data by actual D-T fusion neutron irradiation have not existed so far because of lack of fusion neutron irradiation facility, except fundamental radiation damage studies at very low neutron fluence. Therefore such a facility has been strongly requested. According to agreement of need for such a facility among the international parties, a conceptual design activity (CDA) of International Fusion Material Irradiation Facility (IFMIF) has been carried out under the frame work of the IEA-Implementing Agreement. In the activity, a neutronics analysis on irradiation field optimization in the IFMIF test cell was performed in three parties, Japan, US and EU. As the Japanese contribution, the present paper describes a neutron source term as well as incident deuteron beam angle optimization of two beam geometry, beam shape (foot print) optimization, and dpa, gas production and heating estimation inside various material loading Module, including a sensitivity analysis of source term uncertainty to the estimated irradiation parameters. (author)

  7. Determination of sodium in biological samples by instrumental neutron activation analysis

    International Nuclear Information System (INIS)

    Parwate, D.V.; Garg, A.N.

    1981-01-01

    Sodium is one of the most essential elements needed for metabolic processes amongst human beings. It is consumed in the form of sodium chloride but it is also present in edible plant leaves. Sodium is mostly analyzed by flame photometric method, a destructive and time consuming technique. Sodium has been determined in some green leave vegetables samples-palak, radish, khatta palak (ambat chuka), chaulai leaves, chauli bean covers and its seeds by instrumental neutron activation analysis. The method involves irradiation of samples with thermal neutrons from 241 Am-Be source and counting 24 Na activity (half life 15 hr) from the reaction 23 Na(n,γ) 24 Na. Activity due to 1.37 MeV photopeak was counted with a NaI(Tl) crystal coupled to gamma ray spectrometer. Green leaves of the vegetables were thoroughly washed, dried at constant temperature and powdered. Bowen's Kale powder was used as standard for measuring sodium abundances. About 2g each of samples and the standard were packed in polythene vials. They were irradiated for 24 hrs, delayed by 1 hr and then counted for 20 mts. It is found that radish leaves are most enriched in sodium (14.0 +-0.45%) amongst four leave samples analyzed. For three different parts of chaulai leaves, bean covers and seeds, sodium contents are 1.38%, 1820 and 1010 ppm. Palak contains 2.84 +-0.29% while khatta palak contains only 4210 +- 830 ppm sodium. All values reported here are for dry weight samples and are means of three replicate measurements with standard deviation. (author)

  8. Nondestructive neutron activation analysis of volcanic samples: Hawaii

    International Nuclear Information System (INIS)

    Zoller, W.H.; Finnegan, D.L.; Crowe, B.

    1986-01-01

    Samples of volcanic emissions have been collected between and during eruptions of both Kilauea and Mauna Loa volcanoes during the last three years. Airborne particles have been collected on Teflon filters and acidic gases on base-impregnated cellulose filters. Chemically neutral gas-phase species are collected on charcoal-coated cellulose filters. The primary analytical technique used is nondestructive neutron activation analysis, which has been used to determine the quantities of up to 35 elements on the different filters. The use of neutron activation analysis makes it possible to analyze for a wide range of elements in the different matrices used for the collection and to learn about the distribution between particles and gas phases for each of the elements

  9. Elemental analysis of brazing alloy samples by neutron activation technique

    International Nuclear Information System (INIS)

    Eissa, E.A.; Rofail, N.B.; Hassan, A.M.; El-Shershaby, A.; Walley El-Dine, N.

    1996-01-01

    Two brazing alloy samples (C P 2 and C P 3 ) have been investigated by Neutron activation analysis (NAA) technique in order to identify and estimate their constituent elements. The pneumatic irradiation rabbit system (PIRS), installed at the first egyptian research reactor (ETRR-1) was used for short-time irradiation (30 s) with a thermal neutron flux of 1.6 x 10 1 1 n/cm 2 /s in the reactor reflector, where the thermal to epithermal neutron flux ratio is 106. Long-time irradiation (48 hours) was performed at reactor core periphery with thermal neutron flux of 3.34 x 10 1 2 n/cm 2 /s, and thermal to epithermal neutron flux ratio of 79. Activation by epithermal neutrons was taken into account for the (1/v) and resonance neutron absorption in both methods. A hyper pure germanium detection system was used for gamma-ray acquisitions. The concentration values of Al, Cr, Fe, Co, Cu, Zn, Se, Ag and Sb were estimated as percentages of the sample weight and compared with reported values. 1 tab

  10. Elemental analysis of brazing alloy samples by neutron activation technique

    Energy Technology Data Exchange (ETDEWEB)

    Eissa, E A; Rofail, N B; Hassan, A M [Reactor and Neutron physics Department, Nuclear Research Centre, Atomic Energy Authority, Cairo (Egypt); El-Shershaby, A; Walley El-Dine, N [Physics Department, Faculty of Girls, Ain Shams Universty, Cairo (Egypt)

    1997-12-31

    Two brazing alloy samples (C P{sup 2} and C P{sup 3}) have been investigated by Neutron activation analysis (NAA) technique in order to identify and estimate their constituent elements. The pneumatic irradiation rabbit system (PIRS), installed at the first egyptian research reactor (ETRR-1) was used for short-time irradiation (30 s) with a thermal neutron flux of 1.6 x 10{sup 1}1 n/cm{sup 2}/s in the reactor reflector, where the thermal to epithermal neutron flux ratio is 106. Long-time irradiation (48 hours) was performed at reactor core periphery with thermal neutron flux of 3.34 x 10{sup 1}2 n/cm{sup 2}/s, and thermal to epithermal neutron flux ratio of 79. Activation by epithermal neutrons was taken into account for the (1/v) and resonance neutron absorption in both methods. A hyper pure germanium detection system was used for gamma-ray acquisitions. The concentration values of Al, Cr, Fe, Co, Cu, Zn, Se, Ag and Sb were estimated as percentages of the sample weight and compared with reported values. 1 tab.

  11. Neutron-nucleus interactions and fission. Chapter 1

    International Nuclear Information System (INIS)

    1998-01-01

    The central problem in nuclear-reactor kinetics is to predict the evolution in time of the neutron population in a multiplying medium. Point kinetics allows study of the global behaviour of the neutron population from the average properties of the medium. Before tackling, in the following chapters, the equations governing the time variation of the reactor power (proportional to the total neutron population), the properties of a neutron-multiplying medium shall be discussed briefly. After recalling a number of definitions, a qualitative description shall be given of the principal nuclear reactions at play in a self-sustaining chain reaction, with emphasis on the source of fission neutrons. Since delayed neutrons play a crucial role in reactor kinetics, their production in a reactor shall be described in greater detail. (author)

  12. Time interval approach to the pulsed neutron logging method

    International Nuclear Information System (INIS)

    Zhao Jingwu; Su Weining

    1994-01-01

    The time interval of neighbouring neutrons emitted from a steady state neutron source can be treated as that from a time-dependent neutron source. In the rock space, the neutron flux is given by the neutron diffusion equation and is composed of an infinite terms. Each term s composed of two die-away curves. The delay action is discussed and used to measure the time interval with only one detector in the experiment. Nuclear reactions with the time distribution due to different types of radiations observed in the neutron well-logging methods are presented with a view to getting the rock nuclear parameters from the time interval technique

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

    International Nuclear Information System (INIS)

    Ramirez G, T.

    1976-01-01

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

  14. Uncertainty Assessments in Fast Neutron Activation Analysis

    International Nuclear Information System (INIS)

    W. D. James; R. Zeisler

    2000-01-01

    Fast neutron activation analysis (FNAA) carried out with the use of small accelerator-based neutron generators is routinely used for major/minor element determinations in industry, mineral and petroleum exploration, and to some extent in research. While the method shares many of the operational procedures and therefore errors inherent to conventional thermal neutron activation analysis, its unique implementation gives rise to additional specific concerns that can result in errors or increased uncertainties of measured quantities. The authors were involved in a recent effort to evaluate irreversible incorporation of oxygen into a standard reference material (SRM) by direct measurement of oxygen by FNAA. That project required determination of oxygen in bottles of the SRM stored in varying environmental conditions and a comparison of the results. We recognized the need to accurately describe the total uncertainty of the measurements to accurately characterize any differences in the resulting average concentrations. It is our intent here to discuss the breadth of potential parameters that have the potential to contribute to the random and nonrandom errors of the method and provide estimates of the magnitude of uncertainty introduced. In addition, we will discuss the steps taken in this recent FNAA project to control quality, assess the uncertainty of the measurements, and evaluate results based on the statistical reproducibility

  15. Multi-group transport methods for high-resolution neutron activation analysis

    International Nuclear Information System (INIS)

    Burns, K. A.; Smith, L. E.; Gesh, C. J.; Shaver, M. W.

    2009-01-01

    The accurate and efficient simulation of coupled neutron-photon problems is necessary for several important radiation detection applications. Examples include the detection of nuclear threats concealed in cargo containers and prompt gamma neutron activation analysis for nondestructive determination of elemental composition of unknown samples. In these applications, high-resolution gamma-ray spectrometers are used to preserve as much information as possible about the emitted photon flux, which consists of both continuum and characteristic gamma rays with discrete energies. Monte Carlo transport is the most commonly used modeling tool for this type of problem, but computational times for many problems can be prohibitive. This work explores the use of multi-group deterministic methods for the simulation of neutron activation problems. Central to this work is the development of a method for generating multi-group neutron-photon cross-sections in a way that separates the discrete and continuum photon emissions so that the key signatures in neutron activation analysis (i.e., the characteristic line energies) are preserved. The mechanics of the cross-section preparation method are described and contrasted with standard neutron-gamma cross-section sets. These custom cross-sections are then applied to several benchmark problems. Multi-group results for neutron and photon flux are compared to MCNP results. Finally, calculated responses of high-resolution spectrometers are compared. Preliminary findings show promising results when compared to MCNP. A detailed discussion of the potential benefits and shortcomings of the multi-group-based approach, in terms of accuracy, and computational efficiency, is provided. (authors)

  16. Vanadium determination in pretoleum by neutron activation analysis

    International Nuclear Information System (INIS)

    Lopez, M.; Espinosa, R.

    1983-01-01

    The vanadium concentration in an Peruvian petroleum sample is determined by neutron activation analysis. The samples were irradiated for 20 minutes with a flux of thermal neutrons of 1.75 x 10 7 n/cm 2 -s in a subcritical assembly. The activity of the created samples decreases to half 15 minutes after the irradiation. The result is 28.3 +- 0.8 p.p.m. with a typical deviation of 2.8%. The detection limit of this method is 4 p.p.m

  17. Radiochemical and instrumental neutron activation analysis - recent trends

    International Nuclear Information System (INIS)

    Dams, R.

    1990-01-01

    Recent trends of radiochemical and instrumental neutron activation analysis are discussed. Novel developments include the application of cyclic and pulsed activation, better energy resolution with hyperpure germanium detectors, and use of pulse processing systems allowing extremely high count rates of very short-lived isotopes. Further development is anticipated in the field of speciation in biological and environmental studies. Radiochemical methods have led to accurate determinations at the ng/g level. A promising future is expected for neutron activation techniques. (orig.)

  18. Elemental Study in Soybean and Products by Neutron Activation Analysis

    International Nuclear Information System (INIS)

    Vorapot, Permnamtip; Arporn Busamongkol; Sirinart, Laoharojanaphand

    2009-07-01

    Full text: Elements were analyzed in soybean and products by Instrumental Neutron Activation Analysis (INAA), Pseudo-Cyclic Instrumental Neutron Activation Analysis (PCINAA) and Epithermal Instrumental Neutron Activation Analysis (EINAA). Elements detected in sample were include Al, Br, Ca, Cl, Cu, Fe, I, K, Mg, Mn Na, Se and Zn. The result showed that the nutritional contents changed after food processing. From experiments (n = 2), it was found that after food processing, the concentration of Cl and Na in soy bean curd increased from 0.0045 and 0.0011% to found 0.91 and 0.39 %, respectively. Other elements did not differ from soybean. Limits of detection for Al, Br, Ca, Cl, Cu, Fe, I, K, Mg, Mn Na, Se and Zn were 0.05, 0.2, 50, 6, 10, 15, 0.05, 30, 40, 5, 5, 0.05 and 1 mg.kg - 1, respectively

  19. Design of active-neutron fuel rod scanner

    International Nuclear Information System (INIS)

    Griffith, G.W.; Menlove, H.O.

    1996-01-01

    An active-neutron fuel rod scanner has been designed for the assay of fissile materials in mixed oxide fuel rods. A 252 Cf source is located at the center of the scanner very near the through hole for the fuel rods. Spontaneous fission neutrons from the californium are moderated and induce fissions within the passing fuel rod. The rod continues past a combined gamma-ray and neutron shield where delayed gamma rays above 1 MeV are detected. We used the Monte Carlo code MCNP to design the scanner and review optimum materials and geometries. An inhomogeneous beryllium, graphite, and polyethylene moderator has been designed that uses source neutrons much more efficiently than assay systems using polyethylene moderators. Layers of borated polyethylene and tungsten are used to shield the detectors. Large NaI(Tl) detectors were selected to measure the delayed gamma rays. The enrichment zones of a thermal reactor fuel pin could be measured to within 1% counting statistics for practical rod speeds. Applications of the rod scanner include accountability of fissile material for safeguards applications, quality control of the fissile content in a fuel rod, and the verification of reactivity potential for mixed oxide fuels. (orig.)

  20. Calibration and evaluation of neutron moisturemeter

    International Nuclear Information System (INIS)

    Tang Zhangxiong; Hu Jiangchao; Sun Laiyan; Wang Huaihui; Wu Weixue

    1992-02-01

    Factors influencing the calibration curve of neutron moisture meter, such as soil type, texture, volume weight and depth, were studied. When the soil bulk density water content is between 15% to 45%, the calibration curve is approximately a straight line, and the intercept and slope are only influenced by the above factors. The growing plants also influence the calibration curve slightly. The measuring error for top soil (< 20 cm) is larger. The relative error between neutron method and weighing method is about 8%. The neutron method has many advantages such as non-interfering, simple, fast and non-time-delay

  1. Data analysis for neutron monitoring in an enrichment facility

    International Nuclear Information System (INIS)

    Markin, J.T.; Stewart, J.E.; Goldman, A.S.

    1982-01-01

    Area monitoring of neutron radiation to detect high-enriched uranium production is a potential strategy for inspector verification of operations in the cascade area of a centrifuge enrichment facility. This paper discusses the application of statistical filtering and hypothesis testing procedures to experimental data taken in an enrichment facility. The results demonstrate that these data analysis methods can enhance detection of facility misoperation by neutron monitoring

  2. Reactor neutron activation for multielemental analysis

    International Nuclear Information System (INIS)

    Reddy, A.V.R.

    1999-01-01

    Neutron Activation Analysis using single comparator (K 0 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

  3. Neutron activation analysis of atmospheric aerosol

    International Nuclear Information System (INIS)

    Obrusnik, I.

    1986-01-01

    Neutron activation analysis (NAA) is a modern analytical method well suited for the analysis of atmospheric aerosols. Particular steps of the NAA procedure and especially different types of aerosol sampling and sample preparation for analysis are discussed in detail. Several possible NAA techniques are described and the advantages of a purely instrumental technique with short and long irradiation are pointed out. Important performance characteristics of the NAA method such as precision, accuracy, sensitivity and detection limits are also discussed. Different applications of NAA in environmental studies are reviewed. (author)

  4. Comparison of instrumental neutron activation analysis and instrumental charged-particle activation analysis for determining of Zn-68 abundance

    International Nuclear Information System (INIS)

    Rafii, H.; Mirzaei, M.; Aslani, G.R.; Kamali-Dehghan, M.; Rajamand, A-A.; Rahiminejad, A.; Mirzajani, N.; Sardari, D.; Shahabi, I.; Majidi, F.

    2004-01-01

    Gallium-67 has found important applications in nuclear medicine since last decades. The bombardment of enriched zinc-68 by proton beams in cyclotron is the most suitable method for the carrier-free production of this radionuclide. Any traces and isotopic impurities of the target cause serious radiological hazards because of their associated induced radioactivities. Trace analysis and Zn-68 content determination of the target material before any bombardment and chemical separation provide a valuable assessment of desired product. The elemental abundance evaluation of enriched isotopes is generally carried out by inductively coupled plasma-mass spectrometry method, ICP-Ms Instrumental neutron activation analysis and instrumental charged particle activation analysis. International neutron activation analysis and instrumental charged- particle activation analysis, looks be an alternative nuclear method for determining the abundance evaluation of enriched Zn-68 enrichment in two different samples has been studied by mean of international neutron activation analysis and instrumental charged- particle activation analysis . One sample was purchased from a French company, cortecnet, and the other was separated by an electromagnetic system in the Ions source department of our center, NRCAM. The neutron or proton irradiation was took place respectively in miniature neutron source reactor of Esfahan by flux of (1 to 5) 10 11 n/cm 2 .sec for 30 min and in Cyclon30 by 19 MeV proton beams of 100μA current for 12 min. The produced radioactivity was measured by HpGe detector for determination of trace impurities and evaluation of Zn-68 content in the samples. The result shows a good agreement with the reported ones by their producers and their low derivation of about ± indicates that the international neutron activation analysis and instrumental charged- particle activation analysis are relatively precise and rapid and each one can be used as a supplemental method for analyzing

  5. Applications of neutron activation analysis in environmental science, biology and geoscience

    International Nuclear Information System (INIS)

    1992-01-01

    The applications of neutron activation analysis technique with high sensitivity, good accuracy, multielemental analysis and non-destruction of samples in hydrosphere, soil and lithosphere, atmosphere, cosmosphere and biosphere were introduced in this book. A large amount of research activities in this field during the 20 years and more carried out by Neutron Activation Analysis Laboratory, Institute of High Energy Physics, Academia Sinica, was summarized. A number of the data and information with important scientific significance was provided

  6. Reconfiguration of the NRAD delay loop for proposed 1 MW operations

    International Nuclear Information System (INIS)

    Heidel, C.C.; Richards, W.J.; Pruett, D.P.

    1984-01-01

    Neutron radiography is provided by the NRAD reactor facility, which is located beneath the HFEF hot cell. The NRAD reactor is a TRIGA reactor and is operated at a steady-state power level of 250 kw solely for neutron radiography and the development of radiography techniques. When the NRAD facility was designed and constructed, an operating power level of 250 kw was considered to be adequate for obtaining radiographs of the type of specimens envisaged at that time. Since that time a second radiography station was installed and the thickness of the specimens being radiographed is greater than was initially envisaged. In order to decrease exposure times, the reactor power level is to be increased to 1 Mw. The present delay loop can not to be used at 1 Mw operation, because the passage way where the primary piping exits the reactor room must be maintained less than 1 MR/hr. To obtain the needed delay before the primary water exits the reactor room using the present internal delay loop system would require two more delay loops of the same size to be placed in series with the present delay loop. Because the NRAD reactor tank is small this is not possible; therefore, the delay must take place external to the reactor tank. The delay loop will have to be located in a shielded area to allow the decay of N 16 . The best location for the delay tank will be in the east radiography

  7. Simulations for the neutron detector TETRA with MCNP

    International Nuclear Information System (INIS)

    Testov, D.; Kuznetsova, E.; Wilson, Jh.

    2013-01-01

    To study the nuclear structure of β-delayed neutron precursors at ALTO ISOL-facility at IPN (Orsay), the high efficiency 4π neutron detector TETRA with 3 He filled counters built at JINR (Dubna) was modified. The MCNP simulations to optimize the future configuration were necessary. The details of the calculations and the major results obtained are discussed

  8. Perspectives for online analysis of raw material by pulsed neutron irradiation

    Science.gov (United States)

    Bach, Pierre; Le Tourneur, P.; Poumarede, B.

    1997-02-01

    On-line analysis by pulsed neutron irradiation is an example of an advanced technology application of nuclear techniques, concerning real problems in the cement, mineral and coal industries. The most significant of these nuclear techniques is their capability of continuous measurement without contact and without sampling, which can lead to improved control of processes and resultant large financial savings. Compared to Californium neutron sources, the use of electrical pulsed neutron generators allows to obtain a higher signal/noise ratio for a more sensitive measurement, and allows to overcome a number of safety problems concerning transportation, installation and maintenance. An experiment related to a possible new on-line raw material analyzer is described, using a pulsed neutron generator. The key factors contributing to an accurate measurement are related to a suitable generator, to a high count rate gamma ray spectroscopy electronics, and to computational tools. Calculation and results for the optimization of the neutron irradiation time diagram are reported. One of the operational characteristics of such an equipment is related to neutron flux available: it is possible to adjust it to the requested accuracy, i.e. for a high accuracy during a few hours/day and for a lower accuracy the rest of the time. This feature allows to operate the neutron tube during a longer time, and then to reduce the cost of analysis.

  9. Determination of cadmium in zinc ores by thermal neutron absorption analysis

    International Nuclear Information System (INIS)

    De Norre, L.; Op de Beeck, J.; Hoste, J.

    1983-01-01

    A method has been developed for routine determination of cadmium in zinc ores by thermal neutron absorption analysis, based on the attenuation of a thermal neutron flux passing through a neutron absorbing material. The thermal neutron flux in related to the 52 V activity induced in a vanadium detector, surrounded by pellets pressed from a mixture of powdered material with graphite. Besides cadmium, also the major constituents zinc, iron and sulfur contribute significantly to the total attenuation of the thermal neutron flux. Calibration lines for these elements are worked out. All irradiations are carried out for 200 s in the partially thermalized neutron flux of a 5 Ci 227 Ac-Be isotope neutron source. After a decay of 30 s, the 52 V activity of the vanadium detector is measured for 400 s with a NaI(Tl) scintillation detector. The analysis sequence, including the computation of the results from the counting data, is automated by means of a LSI-11 Microprocessor with 12Kx16 bit memory. Zinc ores, containing 0.02 to 1.45% cadmium, have been analyzed with a precision ranging from 12.6% to 0.54%, resp. As a test for the reliability of the method, two NBS standard reference materials were analyzed in the same way as the zinc ore samples. (author)

  10. Finding determinants of audit delay by pooled OLS regression analysis

    OpenAIRE

    Vuko, Tina; Čular, Marko

    2014-01-01

    The aim of this paper is to investigate determinants of audit delay. Audit delay is measured as the length of time (i.e. the number of calendar days) from the fiscal year-end to the audit report date. It is important to understand factors that influence audit delay since it directly affects the timeliness of financial reporting. The research is conducted on a sample of Croatian listed companies, covering the period of four years (from 2008 to 2011). We use pooled OLS regression analysis, mode...

  11. Analysis of elements present in beers and brewing waters by neutron activation analysis

    International Nuclear Information System (INIS)

    Krausova, Ivana; Kucera, Jan; Dostalek, Pavel; Potesil, Vaclav

    2011-01-01

    Neutron activation analysis (NAA) was used for determination of Si, Na, K, Ca, Sc, V, Cr, Mn, Fe, Co, Zn, Rb, Cs, and La in Czech beers and brewing waters. The Si concentration in beer determined by the reaction 29 Si(n,p) 29 Al with fast neutrons confirmed that beer is an important Si source in human diet. Determination of other trace elements by NAA with the whole spectrum of reactor neutrons aimed at the feasibility of identification of Gambrinus beers brewed in various breweries. The elements Ca and V appeared to be the best candidates for this purpose. The concentrations of elements determined by NAA were also compared with the recommended daily element intake for humans. The accuracy of the method was proved by analysis of reference materials, specifically NIST SRM 2704 Buffalo River Sediment, NIST SRM 1633b Coal Fly Ash, and NIST SRM 1515 Apple Leaves. (author)

  12. Direct integration multiple collision integral transport analysis method for high energy fusion neutronics

    International Nuclear Information System (INIS)

    Koch, K.R.

    1985-01-01

    A new analysis method specially suited for the inherent difficulties of fusion neutronics was developed to provide detailed studies of the fusion neutron transport physics. These studies should provide a better understanding of the limitations and accuracies of typical fusion neutronics calculations. The new analysis method is based on the direct integration of the integral form of the neutron transport equation and employs a continuous energy formulation with the exact treatment of the energy angle kinematics of the scattering process. In addition, the overall solution is analyzed in terms of uncollided, once-collided, and multi-collided solution components based on a multiple collision treatment. Furthermore, the numerical evaluations of integrals use quadrature schemes that are based on the actual dependencies exhibited in the integrands. The new DITRAN computer code was developed on the Cyber 205 vector supercomputer to implement this direct integration multiple-collision fusion neutronics analysis. Three representative fusion reactor models were devised and the solutions to these problems were studied to provide suitable choices for the numerical quadrature orders as well as the discretized solution grid and to understand the limitations of the new analysis method. As further verification and as a first step in assessing the accuracy of existing fusion-neutronics calculations, solutions obtained using the new analysis method were compared to typical multigroup discrete ordinates calculations

  13. Certification of standard reference materials employing neutron activation analysis

    International Nuclear Information System (INIS)

    Capote Rodriguez, G.; Hernandez Rivero, A.; Molina Insfran, J.; Ribeiro Guevara, S.; Santana Encinosa, C.; Perez Zayas, G.

    1997-01-01

    Neutron activation analysis (Naa) is used extensively as one of the analytical techniques in the certification of standard reference materials (Srm). Characteristics of Naa which make it valuable in this role are: accuracy; multielemental capability; ability to assess homogeneity; high sensitivity for many elements, and essentially non-destructive method. This paper reports the concentrations of thirty elements (major, minor and trace elements) in four Cuban Srm's. The samples were irradiated in a thermal neutron flux of 10 12 -10 13 neutrons.cm -2 .s -1 . The gamma-ray spectra were measured by HPGe detectors and were analysed using ACTAN program, developed in CEADEN. (author) [es

  14. Backtracing neutron analysis in the fusion-fission dynamics study

    International Nuclear Information System (INIS)

    Brennand, E. de Goes; Hanappe, F.; Stuttge, L.

    2001-01-01

    A new method for the analysis of multi parametric experimental data is used in the study of the dynamics of the fission process for the compound system 126 Ba. We apply this method to obtain the correlation between thermal energy related to the neutron total multiplicity and the correlation between pre-scission neutron and pos-scission neutron multiplicities. The results obtained are interpreted into the framework of a dynamical model. From this interpretation we have access to the following information: the friction intensity which drives the dynamical evolution of the system; the initial deformation of the compound system; the barrier evolution with temperature and angular momentum, and fission times. (author)

  15. Monte Carlo modeling and analyses of YALINA- booster subcritical assembly Part II: pulsed neutron source

    International Nuclear Information System (INIS)

    Talamo, A.; Gohar, M.Y.A.; Rabiti, C.

    2008-01-01

    One of the most reliable experimental methods for measuring the kinetic parameters of a subcritical assembly is the Sjoestrand method applied to the reaction rate generated from a pulsed neutron source. This study developed a new analytical methodology for characterizing the kinetic parameters of a subcritical assembly using the Sjoestrand method, which allows comparing the analytical and experimental time dependent reaction rates and the reactivity measurements. In this methodology, the reaction rate, detector response, is calculated due to a single neutron pulse using MCNP/MCNPX computer code or any other neutron transport code that explicitly simulates the fission delayed neutrons. The calculation simulates a single neutron pulse over a long time period until the delayed neutron contribution to the reaction is vanished. The obtained reaction rate is superimposed to itself, with respect to the time, to simulate the repeated pulse operation until the asymptotic level of the reaction rate, set by the delayed neutrons, is achieved. The superimposition of the pulse to itself was calculated by a simple C computer program. A parallel version of the C program is used due to the large amount of data being processed, e.g. by the Message Passing Interface (MPI). The new calculation methodology has shown an excellent agreement with the experimental results available from the YALINA-Booster facility of Belarus. The facility has been driven by a Deuterium-Deuterium or Deuterium-Tritium pulsed neutron source and the (n,p) reaction rate has been experimentally measured by a 3 He detector. The MCNP calculation has utilized the weight window and delayed neutron biasing variance reduction techniques since the detector volume is small compared to the assembly volume. Finally, this methodology was used to calculate the IAEA benchmark of the YALINA-Booster experiment

  16. Calculations of neutron spectra after neutron-neutron scattering

    Energy Technology Data Exchange (ETDEWEB)

    Crawford, B E [Gettysburg College, Box 405, Gettysburg, PA 17325 (United States); Stephenson, S L [Gettysburg College, Box 405, Gettysburg, PA 17325 (United States); Howell, C R [Duke University and Triangle Universities Nuclear Laboratory, Durham, NC 27708-0308 (United States); Mitchell, G E [North Carolina State University, Raleigh, NC 27695-8202 (United States); Tornow, W [Duke University and Triangle Universities Nuclear Laboratory, Durham, NC 27708-0308 (United States); Furman, W I [Joint Institute for Nuclear Research, 141980 Dubna (Russian Federation); Lychagin, E V [Joint Institute for Nuclear Research, 141980 Dubna (Russian Federation); Muzichka, A Yu [Joint Institute for Nuclear Research, 141980 Dubna (Russian Federation); Nekhaev, G V [Joint Institute for Nuclear Research, 141980 Dubna (Russian Federation); Strelkov, A V [Joint Institute for Nuclear Research, 141980 Dubna (Russian Federation); Sharapov, E I [Joint Institute for Nuclear Research, 141980 Dubna (Russian Federation); Shvetsov, V N [Joint Institute for Nuclear Research, 141980 Dubna (Russian Federation)

    2004-09-01

    A direct neutron-neutron scattering length, a{sub nn}, measurement with the goal of 3% accuracy (0.5 fm) is under preparation at the aperiodic pulsed reactor YAGUAR. A direct measurement of a{sub nn} will not only help resolve conflicting results of a{sub nn} by indirect means, but also in comparison to the proton-proton scattering length, a{sub pp}, shed light on the charge-symmetry of the nuclear force. We discuss in detail the analysis of the nn-scattering data in terms of a simple analytical expression. We also discuss calibration measurements using the time-of-flight spectra of neutrons scattered on He and Ar gases and the neutron activation technique. In particular, we calculate the neutron velocity and time-of-flight spectra after scattering neutrons on neutrons and after scattering neutrons on He and Ar atoms for the proposed experimental geometry, using a realistic neutron flux spectrum-Maxwellian plus epithermal tail. The shape of the neutron spectrum after scattering is appreciably different from the initial spectrum, due to collisions between thermal-thermal and thermal-epithermal neutrons. At the same time, the integral over the Maxwellian part of the realistic scattering spectrum differs by only about 6 per cent from that of a pure Maxwellian nn-scattering spectrum.

  17. Neutron scattering in Australia

    Energy Technology Data Exchange (ETDEWEB)

    Knott, R.B. [Australian Nuclear Science and Technology Organisation, Menai (Australia)

    1994-12-31

    Neutron scattering techniques have been part of the Australian scientific research community for the past three decades. The High Flux Australian Reactor (HIFAR) is a multi-use facility of modest performance that provides the only neutron source in the country suitable for neutron scattering. The limitations of HIFAR have been recognized and recently a Government initiated inquiry sought to evaluate the future needs of a neutron source. In essence, the inquiry suggested that a delay of several years would enable a number of key issues to be resolved, and therefore a more appropriate decision made. In the meantime, use of the present source is being optimized, and where necessary research is being undertaken at major overseas neutron facilities either on a formal or informal basis. Australia has, at present, a formal agreement with the Rutherford Appleton Laboratory (UK) for access to the spallation source ISIS. Various aspects of neutron scattering have been implemented on HIFAR, including investigations of the structure of biological relevant molecules. One aspect of these investigations will be presented. Preliminary results from a study of the interaction of the immunosuppressant drug, cyclosporin-A, with reconstituted membranes suggest that the hydrophobic drug interdigitated with lipid chains.

  18. Neutron scattering in Australia

    International Nuclear Information System (INIS)

    Knott, R.B.

    1994-01-01

    Neutron scattering techniques have been part of the Australian scientific research community for the past three decades. The High Flux Australian Reactor (HIFAR) is a multi-use facility of modest performance that provides the only neutron source in the country suitable for neutron scattering. The limitations of HIFAR have been recognized and recently a Government initiated inquiry sought to evaluate the future needs of a neutron source. In essence, the inquiry suggested that a delay of several years would enable a number of key issues to be resolved, and therefore a more appropriate decision made. In the meantime, use of the present source is being optimized, and where necessary research is being undertaken at major overseas neutron facilities either on a formal or informal basis. Australia has, at present, a formal agreement with the Rutherford Appleton Laboratory (UK) for access to the spallation source ISIS. Various aspects of neutron scattering have been implemented on HIFAR, including investigations of the structure of biological relevant molecules. One aspect of these investigations will be presented. Preliminary results from a study of the interaction of the immunosuppressant drug, cyclosporin-A, with reconstituted membranes suggest that the hydrophobic drug interdigitated with lipid chains

  19. Development and simulation of various methods for neutron activation analysis

    International Nuclear Information System (INIS)

    Otgooloi, B.

    1993-01-01

    Simple methods for neutron activation analysis have been developed. The results on the studies of installation for determination of fluorine in fluorite ores directly on the lorry by fast neutron activation analysis have been shown. Nitrogen in organic materials was shown by N 14 and N 15 activation. The description of the new equipment 'FLUORITE' for fluorate factory have been shortly given. Pu and Be isotope in organic materials, including in wheat, was measured. 25 figs, 19 tabs. (Author, Translated by J.U)

  20. Polarized neutrons

    International Nuclear Information System (INIS)

    Williams, W.G.

    1988-01-01

    The book on 'polarized neutrons' is intended to inform researchers in condensed matter physics and chemistry of the diversity of scientific problems that can be investigated using polarized neutron beams. The contents include chapters on:- neutron polarizers and instrumentation, polarized neutron scattering, neutron polarization analysis experiments and precessing neutron polarization. (U.K.)