Recent advances in laser-driven neutron sources

Science.gov (United States)

Alejo, A.; Ahmed, H.; Green, A.; Mirfayzi, S. R.; Borghesi, M.; Kar, S.

2016-11-01

Due to the limited number and high cost of large-scale neutron facilities, there has been a growing interest in compact accelerator-driven sources. In this context, several potential schemes of laser-driven neutron sources are being intensively studied employing laser-accelerated electron and ion beams. In addition to the potential of delivering neutron beams with high brilliance, directionality and ultra-short burst duration, a laser-driven neutron source would offer further advantages in terms of cost-effectiveness, compactness and radiation confinement by closed-coupled experiments. Some of the recent advances in this field are discussed, showing improvements in the directionality and flux of the laser-driven neutron beams.

INAA using 252Cf neutron source at University of Pune

International Nuclear Information System (INIS)

Rajurkar, N.S.

2006-01-01

The review presents the work done over last two decades on Instrumental Neutron Activation Analysis (INAA) by our research group at University of Pune using 252 Cf spontaneous fission neutron source. The technique has been applied in different fields viz. numismatics, industry, agriculture, ayurveda, environmental and health sciences and diffusion studies. A brief discussion of the work is presented in this article. (author)

Neutron calibration field of bare {sup 252}Cf source in Vietnam

Energy Technology Data Exchange (ETDEWEB)

Le, Ngoc Thiem; Tran, Hoai Nam; Nguyen, Khai Tuan [Institute for Nuclear Science and Technology, Hanoi (Viet Nam); Trinh, Glap Van [Institute of Research and Development, Duy Tan University, Da Nang (Viet Nam)

2017-02-15

This paper presents the establishment and characterization of a neutron calibration field using a bare {sup 252}Cf source of low neutron source strength in Vietnam. The characterization of the field in terms of neutron flux spectra and neutron ambient dose equivalent rates were performed by Monte Carlo simulations using the MCNP5 code. The anisotropy effect of the source was also investigated. The neutron ambient dose equivalent rates at three reference distances of 75, 125, and 150 cm from the source were calculated and compared with the measurements using the Aloka TPS-451C neutron survey meters. The discrepancy between the calculated and measured values is found to be about 10%. To separate the scattered and the direct components from the total neutron flux spectra, an in-house shadow cone of 10% borated polyethylene was used. The shielding efficiency of the shadow cone was estimated using the MCNP5 code. The results confirmed that the shielding efficiency of the shadow cone is acceptable.

Neutron dose rate for {sup 252} Cf AT source in medical applications

Energy Technology Data Exchange (ETDEWEB)

Paredes, L.; Balcazar, M. [ININ, 52045 Ocoyoacac, Estado de Mexico (Mexico); Azorin, J. [UAM-I, 09340 Mexico D.F. (Mexico); Francois, J.L. [FI-UNAM, 04510 Mexico D.F. (Mexico)

2006-07-01

The AAPM TG-43 modified protocol was used for the calculation of the neutron dose rate of {sup 252}Cf sources for two tissue substitute materials, five normal tissues and six tumours. The {sup 252}Cf AT source model was simulated using the Monte Carlo MCNPX code in spherical geometry for the following factors: a) neutron air kerma strength conversion factor, b) dose rate constant, c) radial dose function, d) geometry factor, e) anisotropy function and f) neutron dose rate. The calculated dose rate in water at 1 cm and 90 degrees from the source long axis, using the Watt fission spectrum, was D{sub n}(r{sub 0}, {theta}{sub 0})= 1.9160 cGy/h-{mu}g. When this value is compared with Rivard et al. calculation using MCNP4B code, 1.8730 cGy/h-{mu}g, a difference of 2.30% is obtained. The results for the reference neutron dose rate in other media show how small variations in the elemental composition between the tissues and malignant tumours, produce variations in the neutron dose rate up to 12.25%. (Author)

Design of a setup for {sup 252}Cf neutron source for storage and analysis purpose

Energy Technology Data Exchange (ETDEWEB)

Hei, Daqian [Department of Nuclear Science and Engineering, College of Materials Science and Engineering, Nanjing University of Aeronautics and Astronautics, Nanjing 211106 (China); Zhuang, Haocheng [Xi’an Middle School of Shanxi Province, Xi’an 710000 (China); Jia, Wenbao, E-mail: jiawenbao@163.com [Department of Nuclear Science and Engineering, College of Materials Science and Engineering, Nanjing University of Aeronautics and Astronautics, Nanjing 211106 (China); Collaborative Innovation Center of Radiation Medicine of Jiangsu Higher Education Institutions, Suzhou 215000 (China); Cheng, Can; Jiang, Zhou; Wang, Hongtao [Department of Nuclear Science and Engineering, College of Materials Science and Engineering, Nanjing University of Aeronautics and Astronautics, Nanjing 211106 (China); Chen, Da [Department of Nuclear Science and Engineering, College of Materials Science and Engineering, Nanjing University of Aeronautics and Astronautics, Nanjing 211106 (China); Collaborative Innovation Center of Radiation Medicine of Jiangsu Higher Education Institutions, Suzhou 215000 (China)

2016-11-01

{sup 252}Cf is a reliable isotopic neutron source and widely used in the prompt gamma ray neutron activation analysis (PGNAA) technique. A cylindrical barrel made by polymethyl methacrylate contained with the boric acid solution was designed for storage and application of a 5 μg {sup 252}Cf neutron source. The size of the setup was optimized with Monte Carlo code. The experiments were performed and the results showed the doses were reduced with the setup and less than the allowable limit. The intensity and collimating radius of the neutron beam could also be adjusted through different collimator.

Frequency spectrum analysis of 252Cf neutron source based on LabVIEW

International Nuclear Information System (INIS)

Mi Deling; Li Pengcheng

2011-01-01

The frequency spectrum analysis of 252 Cf Neutron source is an extremely important method in nuclear stochastic signal processing. Focused on the special '0' and '1' structure of neutron pulse series, this paper proposes a fast-correlation algorithm to improve the computational rate of the spectrum analysis system. And the multi-core processor technology is employed as well as multi-threaded programming techniques of LabVIEW to construct frequency spectrum analysis system of 252 Cf neutron source based on LabVIEW. It not only obtains the auto-correlation and cross correlation results, but also auto-power spectrum,cross-power spectrum and ratio of spectral density. The results show that: analysis tools based on LabVIEW improve the fast auto-correlation and cross correlation code operating efficiency about by 25% to 35%, also verify the feasibility of using LabVIEW for spectrum analysis. (authors)

Source-jerk method for application on ADS neutronics study The ADS is stated for Accelerator Driven sub-critical System

CERN Document Server

Zhu Qing Fu; Li Yi; Xia Pu; Zheng Wu Qing; Zhu Guo Sheng

2003-01-01

The paper is concerned in the source-jerk method used to measure the sub-criticality, and the sub-critical experiment facility, which is used for the study on the neutronics of ADS, driven by external neutron source sup 2 sup 5 sup 2 Cf. The effects of the location of neutron source and material buffer where is at the location of the pipe of proton beam and target of fission-product dispersion on the sub-criticality of reactor are studied by source-jerk method

Transport-level description of the 252Cf-source method using the Langevin technique

International Nuclear Information System (INIS)

Stolle, A.M.; Akcasu, A.Z.

1991-01-01

The fluctuations in the neutron number density and detector outputs in a nuclear reactor can be analyzed conveniently by using the Langevin equation approach. This approach can be implemented at any level of approximation to describe the time evolution of the neutron population, from the most complete transport-level description to the very basic point reactor analysis of neutron number density fluctuations. In this summary, the complete space- and velocity-dependent transport-level formulation of the Langevin equation approach is applied to the analysis of the 252 Cf-source-driven noise analysis (CSDNA) method, an experimental technique developed by J.T. Mihalczo at Oak Ridge National Laboratory, which makes use of noise analysis to determine the reactivity of subcritical media. From this analysis, a theoretical expression for the subcritical multiplication factor is obtained that can then be used to interpret the experimental data. Results at the transport level are in complete agreement with an independent derivation performed by Sutton and Doub, who used the probability density method to interpret the CSDNA experiment, but differed from other expressions that have appeared in the literature

Calculation of dose distribution for 252Cf fission neutron source in tissue equivalent phantoms using Monte Carlo method

International Nuclear Information System (INIS)

Ji Gang; Guo Yong; Luo Yisheng; Zhang Wenzhong

2001-01-01

Objective: To provide useful parameters for neutron radiotherapy, the author presents results of a Monte Carlo simulation study investigating the dosimetric characteristics of linear 252 Cf fission neutron sources. Methods: A 252 Cf fission source and tissue equivalent phantom were modeled. The dose of neutron and gamma radiations were calculated using Monte Carlo Code. Results: The dose of neutron and gamma at several positions for 252 Cf in the phantom made of equivalent materials to water, blood, muscle, skin, bone and lung were calculated. Conclusion: The results by Monte Carlo methods were compared with the data by measurement and references. According to the calculation, the method using water phantom to simulate local tissues such as muscle, blood and skin is reasonable for the calculation and measurements of dose distribution for 252 Cf

Calculated neutron air kerma strength conversion factors for a generically encapsulated Cf-252 brachytherapy source

CERN Document Server

Rivard, M J; D'Errico, F; Tsai, J S; Ulin, K; Engler, M J

2002-01-01

The sup 2 sup 5 sup 2 Cf neutron air kerma strength conversion factor (S sub K sub N /m sub C sub f) is a parameter needed to convert the radionuclide mass (mu g) provided by Oak Ridge National Laboratory into neutron air kerma strength required by modern clinical brachytherapy dosimetry formalisms indicated by Task Group No. 43 of the American Association of Physicists in Medicine (AAPM). The impact of currently used or proposed encapsulating materials for sup 2 sup 5 sup 2 Cf brachytherapy sources (Pt/Ir-10%, 316L stainless steel, nitinol, and Zircaloy-2) on S sub K sub N /m sub C sub f was calculated and results were fit to linear equations. Only for substantial encapsulation thicknesses, did S sub K sub N /m sub C sub f decrease, while the impact of source encapsulation composition is increasingly negligible as Z increases. These findings are explained on the basis of the non-relativistic kinematics governing the majority of sup 2 sup 5 sup 2 Cf neutron interactions. Neutron kerma and energy spectra resul...

Problems in the neutron dynamics of source-driven systems

International Nuclear Information System (INIS)

Ravetto, P.

2001-01-01

The present paper presents some neutronic features of source-driven neutron multiplying systems, with special regards to dynamics, discussing the validity and limitations of classical methods, developed for systems in the vicinity of criticality. Specific characteristics, such as source dominance and the role of delayed neutron emissions are illustrated. Some dynamic peculiarities of innovative concepts proposed for accelerator-driven systems, such as fluid-fuel, are also discussed. The second portion of the work formulates the quasi-static methods for source-driven systems, evidencing its novel features and presenting some numerical results. (author)

Neutron activation analysis detection limits using 252Cf sources

International Nuclear Information System (INIS)

DiPrete, D.P.; Sigg, R.A.

2000-01-01

The Savannah River Technology Center (SRTC) developed a neutron activation analysis (NAA) facility several decades ago using low-flux 252 Cf neutron sources. Through this time, the facility has addressed areas of applied interest in managing the Savannah River Site (SRS). Some applications are unique because of the site's operating history and its chemical-processing facilities. Because sensitivity needs for many applications are not severe, they can be accomplished using an ∼6-mg 252 Cf NAA facility. The SRTC 252 Cf facility continues to support applied research programs at SRTC as well as other SRS programs for environmental and waste management customers. Samples analyzed by NAA include organic compounds, metal alloys, sediments, site process solutions, and many other materials. Numerous radiochemical analyses also rely on the facility for production of short-lived tracers, yielding by activation of carriers and small-scale isotope production for separation methods testing. These applications are more fully reviewed in Ref. 1. Although the flux [approximately2 x 10 7 n/cm 2 ·s] is low relative to reactor facilities, more than 40 elements can be detected at low and sub-part-per-million levels. Detection limits provided by the facility are adequate for many analytical projects. Other multielement analysis methods, particularly inductively coupled plasma atomic emission and inductively coupled plasma mass spectrometry, can now provide sensitivities on dissolved samples that are often better than those available by NAA using low-flux isotopic sources. Because NAA allows analysis of bulk samples, (a) it is a more cost-effective choice when its sensitivity is adequate than methods that require digestion and (b) it eliminates uncertainties that can be introduced by digestion processes

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.

Neutron gauging applications using a small 252Cf source

International Nuclear Information System (INIS)

Helf, S.

1975-01-01

The use of a small 252 Cf source, in the 3 to 4 μg range, for neutron gauging applications is described. Emphasis is placed on determination of low concentrations of moisture in homogeneous media, e.g., solvents, explosives, dried food products, etc. and on measurement of charge or fill weight of hydrogenous materials in sealed items, e.g., propellant in a cartridge case. Both moderation of fast neutrons and attenuation of thermalized neutrons have been explored for these applications. Parameters related to the attainment of optimum sensitivity for each method are discussed. Fast neutron moderation is superior for low level moisture measurement whereas thermal neutron attenuation is more sensitive for ''neutron weighing'' applications. Under optimum conditions, sensitivity for moisture measurement approaches 0.1 weight percent whereas ''neutron weighing'' can detect changes in hydrogeneous material content as little as a fraction of a gram. Examples are given for each technique. A number of different thermal neutron detectors are compared for neutron gauging measurements. A 6 LiI (Eu) scintillation detector is judged to be superior with regard to high thermal neutron detection efficiency and low fast neutron and gamma ray response. In this study, emphasis is placed on the use of simple, portable equipment easily adaptable to field or plant use and for on-line process or quality control. (U.S.)

Design and analysis of nuclear battery driven by the external neutron source

International Nuclear Information System (INIS)

Wang, Sanbing; He, Chaohui

2014-01-01

Highlights: • A new type of space nuclear power called NBDEx is investigated. • NBDEx with 252 Cf has better performance than RTG with similar structure. • Its thermal power gets great improvement with increment of fuel enrichment. • The service life of NBDEx is about 2.96 year. • The launch abortion accident analysis fully demonstrates the advantage of NBDEx. - Abstract: Based on the theory of ADS (Accelerator Driven Subcritical reactor), a new type of nuclear battery was investigated, which was composed of a subcritical fission module and an isotope neutron source, called NBDEx (Nuclear Battery Driven by External neutron source). According to the structure of GPHS-RTG (General Purpose Heat Source Radioisotope Thermoelectric Generator), the fuel cell model and fuel assembly model of NBDEx were set up, and then their performances were analyzed with MCNP code. From these results, it was found that the power and power density of NBDEx were almost six times higher than the RTG’s. For fully demonstrating the advantage of NBDEx, the analysis of its impact factors was performed with MCNP code, and its lifetime was also calculated using the Origen code. These results verified that NBDEx was more suitable for the space missions than RTG

Photon contributions from the 252Cf and 241Am–Be neutron sources at the PSI Calibration Laboratory

International Nuclear Information System (INIS)

Hoedlmoser, H.; Boschung, M.; Meier, K.; Stadtmann, H.; Hranitzky, C.; Figel, M.; Mayer, S.

2012-01-01

At the accredited PSI Calibration Laboratory neutron reference fields traceable to the national standards of the Physikalisch-Technische Bundesanstalt (PTB) in Germany are available for the calibration of ambient and personal dose equivalent (rate) meters and passive dosimeters. The photon contribution to the ambient dose equivalent in the neutron fields of the 252 Cf and 241 Am–Be sources was measured using various photon dose rate meters and active and passive dosimeters. Measuring photons from a neutron source usually involves considerable uncertainties due to the presence of neutron induced photons in the room, due to a non-zero neutron sensitivity of the photon detector, and last but not least due to the energy response of the photon detectors. Therefore eight independent detectors and methods were used to obtain a reliable estimate for the photon contribution of the two sources as an average of the individual methods. For the 241 Am–Be source a photon contribution of approximately 4.9% was determined and for the 252 Cf source a contribution of 3.6%.

Graphite moderated 252Cf source

International Nuclear Information System (INIS)

Sajo B, L.; Barros, H.; Greaves, E. D.; Vega C, H. R.

2014-08-01

The thorium molten salt reactor is an attractive and affordable nuclear power option for developing countries with insufficient infrastructure and limited technological capability. In the aim of personnel training and experience gathering at the Universidad Simon Bolivar there is in progress a project of developing a subcritical thorium liquid fuel reactor. The neutron source to run this subcritical reactor is a 252 Cf source and the reactor will use high-purity graphite as moderator. Using the MCNP5 code the neutron spectra of the 252 Cf in the center of the graphite moderator has been estimated along the channel where the liquid thorium salt will be inserted; also the ambient dose equivalent due to the source has been determined around the moderator. (Author)

Linac-driven spallation-neutron source

International Nuclear Information System (INIS)

Jason, A.J.

1995-01-01

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

Two-quasineutron states in 98248Cf and 98250Cf and the neutron-neutron residual interactions

International Nuclear Information System (INIS)

Katori, K.; Ahmad, I.; Friedman, A. M.

2008-01-01

Two-quasineutron states in 248 Cf and 250 Cf were investigated by single-neutron transfer reactions, 249 Cf(d,t) 248 Cf and 249 Cf(d,p) 250 Cf. The majority of levels observed were assigned to 12 bands in 248 Cf and six bands in 250 Cf, constructed from the single-particle states in neighboring Cf nuclei. The effective two-body interactions between two odd neutrons coupled outside a deformed core were deduced from the differences in the energies of the bandheads formed by the parallel and antiparallel coupling of the intrinsic spins of the two single-particle states

Measurement of Neutron Energy Spectrum Emitted by Cf-252 Source Using Time-of-Flight Method

Energy Technology Data Exchange (ETDEWEB)

Lee, Cheol Ho; Son, Jaebum; Kim, Tae Hoon; Lee, Sangmin; Kim, Yong-Kyun [Hanyang University, Seoul (Korea, Republic of)

2016-10-15

The techniques proposed to detect the neutrons usually require the detection of a secondary recoiling nucleus in a scintillator (or other type of detector) to indicate the rare collision of a neutron with a nucleus. This is the same basic technique, in this case detection of a recoil proton that was used by Chadwick in the 1930 s to discover and identify the neutron and determine its mass. It is primary technique still used today for detection of fast neutron, which typically involves the use of a hydrogen based organic plastic or liquid scintillator coupled to a photo-multiplier tube. The light output from such scintillators is a function of the cross section and nuclear kinematics of the n + nucleus collision. With the exception of deuterated scintillators, the scintillator signal does not necessarily produce a distinct peak in the scintillator spectrum directly related to the incident neutron energy. Instead neutron time-of-flight (TOF) often must be utilized to determine the neutron energy, which requires generation of a prompt start signal from the nuclear source emitting the neutrons. This method takes advantage of the high number of prompt gamma rays. The Time-of-Flight method was used to measure neutron energy spectrum emitted by the Cf-252 neutron source. Plastic scintillator that has a superior discrimination ability of neutron and gamma-ray was used as a stop signal detector and liquid scintillator was used as a stat signal detector. In experiment, neutron and gamma-ray spectrum was firstly measured and discriminated using the TOF method. Secondly, neutron energy spectrum was obtained through spectrum analysis. Equation of neutron energy spectrum that was emitted by Cf-252 source using the Gaussian fitting was obtained.

Fabrication of 50-mg 252Cf neutron sources for the FDA [Food and Drug Administration] activation analysis facility

International Nuclear Information System (INIS)

Bigelow, J.E.; Cagle, E.B.; Knauer, J.B.

1987-01-01

The Transuranium Processing Plant (TPP) at ORNL has been requested by the Food and Drug Administration (FDA) to furnish 200 mg of 252 Cf for use in their new activation analysis facility. This paper discusses the procedure to be employed in fabricating the californium into four neutron sources, each containing a nominal 50-mg of 252 Cf. The ORNL Model LSD (Large, Stainless steel, Doubly encapsulated) neutron source consists of a 6.33-mm-diam aluminum pellet doubly encapsulated in Type 304L stainless steel. The pellet is comprised of an aluminum tube holding Cf 2 O 2 SO 4 microspheres confined by pressed aluminum powder. The microspheres are prepared in a separate vessel and then transferred into the specially designed aluminum tube prior to pressing

Development of the neutron reference calibration field using a 252Cf standard source surrounded with PMMA moderators

International Nuclear Information System (INIS)

Yoshida, T.; Kanai, K.; Tsujimura, N.

2002-01-01

The authors developed the neutron reference calibration fields using a 252 Cf standard source surrounded with PMMA moderators at the Japan Nuclear Cycle Development (JNC), Tokai Works. The moderators are co-axial, hollow cylinders made of lead-contained PMMA with a thickness of 13.5, 35.0, 59.5 and 77.0mm, and the 252 Cf source is guided to the geometric center of moderators by the pneumatic system. These fields can provide the moderated neutron spectra very similar to those encountered around the globe-boxes of the fabrication process of MOX (PuO 2 -UO 2 mixed oxide) fuel. The neutron energy spectrum at the reference calibration point was evaluated from the calculations by MCNP4B and the measurements by the INS-type Bonner multi-sphere spectrometer and the hydrogen-filled proportional counters. The calculated neutron spectra were in good agreements with the measured ones. These fields were characterized in terms of the neutron fluence rate, spectral composition and ambient dose equivalent rate, and have served for the response-characterization of various neutron survey instruments

Neutron total cross section measurements on 249Cf

International Nuclear Information System (INIS)

Carlton, R.F.; Harvey, J.A.; Hill, N.W.; Pandey, M.S.; Benjamin, R.W.

1979-01-01

Neutron total cross section measurements were performed on a sample of 249 Cf (5.65 mg total weight) with the ORELA as a source of pulsed neutrons. The sample, the inverse thickness of which was 1542 barns/atom, consisted of 85.3% 249 Cf and 14.4% 249 Bk, and was cooled to liquid nitrogen temperature. Analyses were also made of data from a thin sample (l/n = 17430) of 65% 249 Cf in the region of the large fission resonance at 0.7 eV. Fifty-five resonances in 249 Cf were observed and analyzed over the energy range 0.1 eV to 90 eV by use of an R-matrix multilevel formalism. The resonance parameters obtained were used to determine the level spacing and the s-wave neutron and fission strength functions. Thermal total cross section measurements were also performed. 5 figures, 3 tables

Importance of the neutrons kerma coefficient in the planning of Brachytherapy treatments with Cf-252 sources

International Nuclear Information System (INIS)

Paredes G, L.; Balcazar G, M.; Azorin N, J.; Francois L, J.L.

2006-01-01

The Cf-252 is a fast neutrons emitting radioisotope by spontaneous fission that can be used as sealed source in medicine applications, industry and research. Commercially its offer sources of different sizes, compact and with a fast neutrons emission of the order of 10 6 n/s-μg and an energy spectra that presents respectively maxim and average energy in 2.1 MeV and 0.7 MeV. In medicine new applications are being developed for the treatment of patient with hypoxic and voluminous tumors, where the therapy with photons has not given positive results, as well as for the protocols of therapy treatment by boron neutron capture, where very small sources of Cf-252 will be used with the interstitial brachytherapy technique of high and low dose rate. In this work an analysis of how the small differences that exist in the elementary composition of 4 wicked tumors, 4 ICRU healthy tissues and 3 substitute materials of ICRU tissue used in dosimetry are presented, its generate changes in the neutrons kerma coefficient in function of the energy and consequently in the absorbed dose in the interval of 11 eV to 29 MeV. These differences can produce maximum variations of the neutron kerma coefficients ratio for E n > 1 keV of the one: 15% tumor/ICRU guest healthy tissue, 12% ICRU tumor/muscle, 12% ICRU healthy tissues ICRU/ICRU muscle, 22% substitutes tissue/tumor and 22% ICRU substitutes tissue/muscle. Also, it was found that the average value of the neutrons kerma coefficient for the 4 wicked tumors is from 6% to 7% smaller that the average value for the soft tissue in the interval energy of interest for therapy with fast neutrons with E n > 1 MeV. These results have a special importance during the planning process of brachytherapy treatments with sources of 252 Cf, to optimize and to individualize the patients treatments. (Author)

Graphite moderated {sup 252}Cf source

Energy Technology Data Exchange (ETDEWEB)

Sajo B, L.; Barros, H.; Greaves, E. D. [Universidad Simon Bolivar, Nuclear Physics Laboratory, Apdo. 89000, 1080A Caracas (Venezuela, Bolivarian Republic of); Vega C, H. R., E-mail: fermineutron@yahoo.com [Universidad Autonoma de Zacatecas, Unidad Academica de Estudios Nucleares, Cipres No. 10, Fracc. La Penuela, 98068 Zacatecas (Mexico)

2014-08-15

The thorium molten salt reactor is an attractive and affordable nuclear power option for developing countries with insufficient infrastructure and limited technological capability. In the aim of personnel training and experience gathering at the Universidad Simon Bolivar there is in progress a project of developing a subcritical thorium liquid fuel reactor. The neutron source to run this subcritical reactor is a {sup 252}Cf source and the reactor will use high-purity graphite as moderator. Using the MCNP5 code the neutron spectra of the {sup 252}Cf in the center of the graphite moderator has been estimated along the channel where the liquid thorium salt will be inserted; also the ambient dose equivalent due to the source has been determined around the moderator. (Author)

Shielding experiments in different materials with 252Cf neutron spectra

International Nuclear Information System (INIS)

Sathian, Deepa; Marathe, P.K.; Pal, Rupali; Jayalakshmi, V.; Chourasiya, G.; Mayya, Y.S.

2008-01-01

Adequate shielding for neutron sources can be determined using analytical method or by actually measuring the attenuation for the target configuration. This paper describes the measurement of Half Value Thickness (HVT), Tenth Value Thickness (TVT), Σ values for four different shielding materials, using a standard 252 Cf neutron source and comparing with the values calculated using an empirical relationship. BF 3 based REM-counter is used for measurement of neutron dose equivalent, against different thickness of the shielding material. The experimental HVT and S values are in good agreement with the calculated values. From this study, it is concluded that, among the four materials studied, high density polyethylene (HDPE) is best suitable for the shielding of a 252 Cf neutron source. (author)

Development of the neutron reference calibration field using a {sup 252}Cf standard source surrounded with PMMA moderators

Energy Technology Data Exchange (ETDEWEB)

Yoshida, T.; Kanai, K.; Tsujimura, N. [Japan Nuclear Cycle Development Institute, Ibaraki-ken (Japan)

2002-07-01

The authors developed the neutron reference calibration fields using a {sup 252} Cf standard source surrounded with PMMA moderators at the Japan Nuclear Cycle Development (JNC), Tokai Works. The moderators are co-axial, hollow cylinders made of lead-contained PMMA with a thickness of 13.5, 35.0, 59.5 and 77.0mm, and the {sup 252} Cf source is guided to the geometric center of moderators by the pneumatic system. These fields can provide the moderated neutron spectra very similar to those encountered around the globe-boxes of the fabrication process of MOX (PuO{sub 2}-UO{sub 2} mixed oxide) fuel. The neutron energy spectrum at the reference calibration point was evaluated from the calculations by MCNP4B and the measurements by the INS-type Bonner multi-sphere spectrometer and the hydrogen-filled proportional counters. The calculated neutron spectra were in good agreements with the measured ones. These fields were characterized in terms of the neutron fluence rate, spectral composition and ambient dose equivalent rate, and have served for the response-characterization of various neutron survey instruments.

Neutron source for a reactor

International Nuclear Information System (INIS)

Kobayashi, Hiromasa.

1975-01-01

Object: To easily increase a start-up power of a reactor without irradiation in other reactors. Structure: A neutron source comprises Cf 252 , a natural antimony rod, a layer of beryllium, and a vessel of neutron source. On upper and lower portion of Cf 252 are arranged natural antimony rods, which are surrounded by the Be layer, the entirety being charged into the vessel. The Cf 252 may emit neutron, has a half life more than a period of operating cycle of the reactor and is less deteriorated even irradiated by radioactive rays while being left within the reactor. The natural antimony rod is radioactivated by neutron from Cf 252 and neutron as reactor power increases to emit γ rays. The Be absorbs γ rays to emit the neutron. The antimony rod is irradiated within the reactor. Further, since the Cf 252 is small in neutron absorption cross section, it is hard to be deteriorated even while being inserted within the reactor. (Kamimura, M.)

Identification of neutron noise sources in a boiling water reactor

International Nuclear Information System (INIS)

Sides, W.H. Jr.; Mathis, M.V.; Smith, C.M.

1977-01-01

Measurements were made at units 2 and 3 of the Browns Ferry Nuclear Power Plant in order to characterize the noise signatures of the neutron and process signals and to determine the usefulness of such signatures for anomaly detection in BWR-4s. Previous measurements and theoretical analyses of BWR noise by others were concerned with the determination of steam velocity and void fraction (using the local component of neutron noise) and with the sources of global noise. The work described is under a five-part program to develop a complete and systematic analysis and representation of BWR neutron and process noise through complementary measurements and stochastic model developments. The parts are: (1) recording as many neutron detector and process noise signals as are available in a BWR-4; (2) reducing these data to noise signatures in order to perform an empirical analysis of these signatures, and documenting the relationships between the signals from spatially separated neutron detectors and between neutron and process variables; (3) developing spatially dependent neutronic models coupled with thermal-hydraulic models to aid in interpreting the observed relationships among the measured noise signatures, (4) comparing measured noise signatures with model predictions to obtain additional insight into BWR-4 dynamic behavior and to validate the models; and (5) using these models to predict the sensitivity of noise monitoring for detection, surveillance, and diagnosis of postulated in-core anomalies in BWRs. The paper describes the procedures used to obtain the noise recordings and presents initial empirical analysis and observations pertaining to the noise signatures and the relationships between several noise variables in the 0.01- to 1-Hz range. The mathematical models have not been developed sufficiently to report theoretical results or to compare measured spectra with model predictions at this time

Sensitivity of various thermoluminescent, radiophotoluminescent and photographic detectors to neutrons emitted by a 252Cf source

International Nuclear Information System (INIS)

Spurny, Frantisek; Medioni, Roger; Chapuis, A.; Portal, Guy.

1975-07-01

The specific sensitivity of various thermoluminescent, radiophotoluminescent and photographic detectors to the neutron spectrum of a 252 Cf source was measured and the effect of the size and composition of the containers in which they might be put was investigated. PB33 radiophotoluminescent glasses, radiothermoluminescent alumina and calcium sulfate were less sensitive to fission neutrons whereas photographic emulsions were more sensitive. The former should be used for γ detection in mixed fields of photons and fission neutrons [fr

New applications of neutron noise theory in power reactor physics

Energy Technology Data Exchange (ETDEWEB)

Arzhanov, Vasiliy

2000-04-01

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

New applications of neutron noise theory in power reactor physics

International Nuclear Information System (INIS)

Arzhanov, Vasiliy

2000-04-01

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

Testing Moderating Detection Systems with 252Cf-Based Reference Neutron Fields

International Nuclear Information System (INIS)

Hertel, Nolan E.; Sweezy, Jeremy; Sauber, Jeremiah S.; Vaughn, David; Cook, Andrew; Tays, Jeff; Ro, Tae-Ik

2001-01-01

Calibration measurements were carried out on a probe designed to measure ambient dose equivalent in accordance with ICRP Pub 60 recommendations. It consists of a cylindrical 3 He proportional counter surrounded by a 25-cm-diameter spherical polyethylene moderator. Its neutron response is optimized for dose rate measurements of neutrons between thermal energies and 20 MeV. The instrument was used to measure the dose rate in four separate neutron fields: unmoderated 252 Cf, D 2 O-moderated 252 Cf, polyethylene-moderated 252 Cf, and WEP neutron howitzer with 252 Cf at its center. Dose equivalent measurements were performed at source-detector centerline distances from 50 to 200 cm. The ratio of air-scatter- and room-return-corrected ambient dose equivalent rates to ambient dose equivalent rates calculated with the code MCNP are tabulated

Monte Carlo design study of a moderated {sup 252}Cf source for in vivo neutron activation analysis of aluminium

Energy Technology Data Exchange (ETDEWEB)

Lewis, D.G.; Natto, S.S.A.; Evans, C.J. [Swansea In Vivo Analysis and Cancer Research Group, Department of Physics, University of Wales, Swansea (United Kingdom); Ryde, S.J.S. [Swansea In Vivo Analysis and Cancer Research Group, Department of Medical Physics and Clinical Engineering, Singleton Hospital, Swansea (United Kingdom)

1997-04-01

The Monte Carlo computer code MCNP has been used to design a moderated 2{sup 52}Cf neutron source for in vivo neutron activation analysis of aluminium (Al) in the bones of the hand. The clinical motivation is the need to monitor l body burden in subjects with renal dysfunction, at risk of Al toxicity. The design involves the source positioned on the central axis at one end of a cylindrical deuterium oxide moderator. The moderator is surrounded by a graphite reflector, with the hand inserted at the end of the moderator opposing the source. For a 1 mg {sup 252}Cf source, 15 cm long x 20 cm radius moderator and 20 cm thick reflector, the estimated minimum detection limit is .5 mg Al for a 20 min irradiation, with an equivalent dose of 16.5 mSv to the hand. Increasing the moderator length and/or introducing a fast neutron filter (for example silicon) further reduces interference from fast-neutron-induced reactions on phosphorus in bone, at the expense of decreased fluence of the thermal neutrons which activate Al. Increased source strengths may be necessary to compensate for this decreased thermal fluence, or allow measurements to be made within an acceptable time limit for the comfort of the patient. (author)

Neutron noise measurements at the Delphi subcritical assembly

International Nuclear Information System (INIS)

Szieberth, M.; Klujber, G.; Kloosterman, J. L.; De Haas, D.

2012-01-01

The paper presents the results and evaluations of a comprehensive set of neutron noise measurements on the Delphi subcritical assembly of the Delft Univ. of Technology. The measurements investigated the effect of different source distributions (inherent spontaneous fission and 252 Cf) and the position of the detectors applied (both radially and vertically). The evaluation of the measured data has been performed by the variance-to-mean ratio (VTMR, Feynman-α), the autocorrelation (ACF, Rossi-α) and the cross-correlation (CCF) methods. The values obtained for the prompt decay constant show a strong bias, which depends both on the detector position and on the source distribution. This is due to the presence of higher modes in the system. It has been observed that the α value fitted is higher when the detector is close to the boundary of the core or to the 252 Cf point-source. The higher alpha-modes have also been observed by fitting functions describing two alpha-modes. The successful set of measurement also provides a good basis for further theoretical investigations including the Monte Carlo simulation of the noise measurements and the calculation of the alpha-modes in the Delphi subcritical assembly. (authors)

Design of a thermal neutron field by 252Cf source for measurement of 10B concentrations in the blood samples for BNCT

International Nuclear Information System (INIS)

Naito, H.; Sakurai, Y.; Maruhashi, A.

2006-01-01

10 B concentrations in the blood samples for BNCT has been estimated due to amounts of prompt gamma rays from 10 B in the fields of thermal neutrons from a special guide tube attached to research reactor. A system using radioisotopes as the source of thermal neutron fields has advantages that are convenient and low cost because it doesn't need running of a reactor or an accelerator. The validity of 252 Cf as a neutron source for 10 B concentrations detection system was investigated. This system is composed of D 2 O moderator, Pb reflector/filter, C reflector, and LiF filter. A thermal neutron field with low background gamma-rays is obtained. A large source of 252 Cf is required to obtain a sufficient flux. (author)

Physics study of D-D/D-T neutron driven experimental subcritical assembly

International Nuclear Information System (INIS)

Sinha, Amar

2015-01-01

An experimental program to design and study external source driven subcritical assembly has been initiated at BARC. This program is aimed at understanding neutronic characteristics of accelerator driven system at low power level. In this series, a zero-power, sub-critical assembly driven by a D-D/D-T neutron generator has been developed. This system is modular in design and it is first in the series of subcritical assemblies being designed. The subcritical core consists of natural uranium fuel with high density polyethylene as moderator and beryllium oxide as reflector. The subcritical core is coupled to Purnima Neutron Generator. Preliminary experiments have been carried out for spatial flux measurement and reactivity estimation using pulsed neutron source (PNS) techniques. Further experiments are being planned to measure the reactivity and other kinetic parameters using noise methods. This facility would also be used for carrying out studies on effect of source importance and measurement of source multiplication factor k s and external neutron source efficiency φ* in great details. Some experiments with D-D and D-T neutrons have been presented. (author)

Importance of the neutrons kerma coefficient in the planning of Brachytherapy treatments with Cf-252 sources; Importancia del coeficiente de kerma de neutrones en la planeacion de tratamientos de Braquiterapia con fuentes de Cf-252

Energy Technology Data Exchange (ETDEWEB)

Paredes G, L.; Balcazar G, M. [ININ, 52045 Ocoyocac, Estado de Mexico (Mexico); Azorin N, J. [Universidad Autonoma Metropolitana, 09000 Mexico D.F. (Mexico); Francois L, J.L. [UNAM, 04500 Mexico D.F. (Mexico)]. e-mail: lpg@nuclear.inin.mx

2006-07-01

The Cf-252 is a fast neutrons emitting radioisotope by spontaneous fission that can be used as sealed source in medicine applications, industry and research. Commercially its offer sources of different sizes, compact and with a fast neutrons emission of the order of 10{sup 6} n/s-{mu}g and an energy spectra that presents respectively maxim and average energy in 2.1 MeV and 0.7 MeV. In medicine new applications are being developed for the treatment of patient with hypoxic and voluminous tumors, where the therapy with photons has not given positive results, as well as for the protocols of therapy treatment by boron neutron capture, where very small sources of Cf-252 will be used with the interstitial brachytherapy technique of high and low dose rate. In this work an analysis of how the small differences that exist in the elementary composition of 4 wicked tumors, 4 ICRU healthy tissues and 3 substitute materials of ICRU tissue used in dosimetry are presented, its generate changes in the neutrons kerma coefficient in function of the energy and consequently in the absorbed dose in the interval of 11 eV to 29 MeV. These differences can produce maximum variations of the neutron kerma coefficients ratio for E{sub n} > 1 keV of the one: 15% tumor/ICRU guest healthy tissue, 12% ICRU tumor/muscle, 12% ICRU healthy tissues ICRU/ICRU muscle, 22% substitutes tissue/tumor and 22% ICRU substitutes tissue/muscle. Also, it was found that the average value of the neutrons kerma coefficient for the 4 wicked tumors is from 6% to 7% smaller that the average value for the soft tissue in the interval energy of interest for therapy with fast neutrons with E{sub n} > 1 MeV. These results have a special importance during the planning process of brachytherapy treatments with sources of {sup 252}Cf, to optimize and to individualize the patients treatments. (Author)

A Monte Carlo Study on the Effect of Various Neutron Capturers on Dose Distribution in Brachytherapy with 252Cf Source

Directory of Open Access Journals (Sweden)

Firoozabadi M. M.

2017-03-01

Full Text Available Background: In neutron interaction with matter and reduction of neutron energy due to multiple scatterings to the thermal energy range, increasing the probability of thermal neutron capture by neutron captures makes dose enhancement in the tumors loaded with these materials. Objective: The purpose of this study is to evaluate dose distribution in the presence of 10B, 157Gd and 33S neutron capturers and to determine the effect of these materials on dose enhancement rate for 252Cf brachytherapy source. Methods: Neutron-ray flux and energy spectra, neutron and gamma dose rates and dose enhancement factor (DEF are determined in the absence and presence of 10B, 157Gd and 33S using Monte Carlo simulation. Results: The difference in the thermal neutron flux rate in the presence of 10B and 157Gd is significant, while the flux changes in the fast and epithermal energy ranges are insensible. The dose enhancement factor has increased with increasing distance from the source and reached its maximum amount equal to 258.3 and 476.1 cGy/h/µg for 157Gd and 10B, respectively at about 8 cm distance from the source center. DEF for 33S is equal to one. Conclusion: Results show that the magnitude of dose augmentation in tumors containing 10B and 157Gd in brachytherapy with 252Cf source will depend not only on the capture product dose level, but also on the tumor distance from the source. 33S makes dose enhancement under specific conditions that these conditions depend on the neutron energy spectra of source, the 33S concentration in tumor and tumor distance from the source.

A hybrid source-driven method to compute fast neutron fluence in reactor pressure vessel - 017

International Nuclear Information System (INIS)

Ren-Tai, Chiang

2010-01-01

A hybrid source-driven method is developed to compute fast neutron fluence with neutron energy greater than 1 MeV in nuclear reactor pressure vessel (RPV). The method determines neutron flux by solving a steady-state neutron transport equation with hybrid neutron sources composed of peripheral fixed fission neutron sources and interior chain-reacted fission neutron sources. The relative rod-by-rod power distribution of the peripheral assemblies in a nuclear reactor obtained from reactor core depletion calculations and subsequent rod-by-rod power reconstruction is employed as the relative rod-by-rod fixed fission neutron source distribution. All fissionable nuclides other than U-238 (such as U-234, U-235, U-236, Pu-239 etc) are replaced with U-238 to avoid counting the fission contribution twice and to preserve fast neutron attenuation for heavy nuclides in the peripheral assemblies. An example is provided to show the feasibility of the method. Since the interior fuels only have a marginal impact on RPV fluence results due to rapid attenuation of interior fast fission neutrons, a generic set or one of several generic sets of interior fuels can be used as the driver and only the neutron sources in the peripheral assemblies will be changed in subsequent hybrid source-driven fluence calculations. Consequently, this hybrid source-driven method can simplify and reduce cost for fast neutron fluence computations. This newly developed hybrid source-driven method should be a useful and simplified tool for computing fast neutron fluence at selected locations of interest in RPV of contemporary nuclear power reactors. (authors)

Effects of neutron spectrum and external neutron source on neutron multiplication parameters in accelerator-driven system

International Nuclear Information System (INIS)

Shahbunder, Hesham; Pyeon, Cheol Ho; Misawa, Tsuyoshi; Lim, Jae-Yong; Shiroya, Seiji

2010-01-01

The neutron multiplication parameters: neutron multiplication M, subcritical multiplication factor k s , external source efficiency φ*, play an important role for numerical assessment and reactor power evaluation of an accelerator-driven system (ADS). Those parameters can be evaluated by using the measured reaction rate distribution in the subcritical system. In this study, the experimental verification of this methodology is performed in various ADS cores; with high-energy (100 MeV) proton-tungsten source in hard and soft neutron spectra cores and 14 MeV D-T neutron source in soft spectrum core. The comparison between measured and calculated multiplication parameters reveals a maximum relative difference in the range of 6.6-13.7% that is attributed to the calculation nuclear libraries uncertainty and accuracy for energies higher than 20 MeV and also dependent on the reaction rate distribution position and count rates. The effects of different core neutron spectra and external neutron sources on the neutron multiplication parameters are discussed.

Application of modular neutron spectrometer to measure neutron spectra from fission of 252Cf

International Nuclear Information System (INIS)

Szeflinski, Z.; Osuch, S.; Popkiewicz, M.; Wilhelmi, Z.; Zelazny, Z.

1996-01-01

The neutron spectrometer MONA (Modular Neutron Array) and its test has been described. The spectrometers consist of eight BC-501A liquid scintillator detectors of BICRON which allow one to distinguish between the pulses from gamma quanta and neutrons using pulse shape discrimination (PSD) method. The electronic equipment for the PSD and the test result using the 252 Cf radioactive source are presented

Experimental research on specific activity of 24Na using Chinese reference man phantom irradiated by 252Cf neutrons source

International Nuclear Information System (INIS)

Wang Yuexing; Yang Yifang; Lu Yongjie; Zhang Jianguo; Xing Hongchuan

2011-01-01

Objective: To investigate the specific activity of '2 4 Na per unit neutron fluence, A B/Φ ,in blood produced for Chinese reference man irradiated by 252 Cf neutron source,and to analyze the effects of scattering neutrons from ground,wall,and ceiling in irradiation site on it.Methods: A 252 Cf neutron source of 3×10 8 n/s and the anthropomorphic phantom were used for experiments. The phantom was made from 4 mm thick of outer covering by perspex and the liquid tissue-equivalent substitute in it. The data of phantom dimensions fit into Chinese reference man.The weight ratios of H, N, O and C in substitute equal from source to long axis of phantom were 1.1, 2.1, 3.1 and 4.1 m, respectively. Both the neutron source and the position of xiphisternum of the phantom were 1.6 m above the floor. Results: The average specific activity of 24 Na per unit neutron fluence was related to the irradiation-distances, d, and its maximum value, A B/ΦM , deduced by experimental data was about 1.85×10 -7 Bq·cm 2 ·g -1 . Conclusions: The A B/ΦM corresponds to that of phantom irradiated by plane-parallel beams, and the value is about more 3% than that by BOMAB phantom reported in literature. It has shown that floor-(wall-)scattered neutrons in irradiation site have significant contribution to the specific activity of 24 Na, but they contributed relatively little to the induced neutron doses. Consequently,using the specific activity of 24 Na for assessing accidental neutron doses received by an individual, the contribution of scattered neutrons in accident site will lead dose to be overestimated, and need to be correct. (authors)

Research opportunities with compact accelerator-driven neutron sources

Energy Technology Data Exchange (ETDEWEB)

Anderson, I.S. [Oak Ridge National Laboratory, Oak Ridge, TN (United States); Andreani, C., E-mail: carla.andreani@uniroma2.it [Università degli Studi di Roma “Tor Vergata”, Physics Department and NAST Centre, Via della Ricerca Scientifica 1, 00133 Roma (Italy); CNR-IPCF Sezione di Messina, Messina (Italy); Museo Storico della Fisica e Centro Studi e Ricerche Enrico Fermi, Roma (Italy); Carpenter, J.M. [Argonne National Laboratory, Argonne, IL (United States); Festa, G., E-mail: giulia.festa@uniroma2.it [Università degli Studi di Roma “Tor Vergata”, Physics Department and NAST Centre, Via della Ricerca Scientifica 1, 00133 Roma (Italy); Museo Storico della Fisica e Centro Studi e Ricerche Enrico Fermi, Roma (Italy); Gorini, G. [Università degli Studi di Milano—Bicocca, Milano (Italy); Loong, C.-K. [Università degli Studi di Roma “Tor Vergata”, Centro NAST, Via della Ricerca Scientifica 1, 00133 Roma (Italy); Senesi, R. [Università degli Studi di Roma “Tor Vergata”, Physics Department and NAST Centre, Via della Ricerca Scientifica 1, 00133 Roma (Italy); CNR-IPCF Sezione di Messina, Messina (Italy); Museo Storico della Fisica e Centro Studi e Ricerche Enrico Fermi, Roma (Italy)

2016-10-13

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

Research opportunities with compact accelerator-driven neutron sources

International Nuclear Information System (INIS)

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

2016-01-01

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

Research opportunities with compact accelerator-driven neutron sources

Science.gov (United States)

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

2016-10-01

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

Nuclear Material Detection by One-Short-Pulse-Laser-Driven Neutron Source

International Nuclear Information System (INIS)

Favalli, Andrea; Aymond, F.; Bridgewater, Jon S.; Croft, Stephen; Deppert, O.; Devlin, Matthew James; Falk, Katerina; Fernandez, Juan Carlos; Gautier, Donald Cort; Gonzales, Manuel A.; Goodsell, Alison Victoria; Guler, Nevzat; Hamilton, Christopher Eric; Hegelich, Bjorn Manuel; Henzlova, Daniela; Ianakiev, Kiril Dimitrov; Iliev, Metodi; Johnson, Randall Philip; Jung, Daniel; Kleinschmidt, Annika; Koehler, Katrina Elizabeth; Pomerantz, Ishay; Roth, Markus; Santi, Peter Angelo; Shimada, Tsutomu; Swinhoe, Martyn Thomas; Taddeucci, Terry Nicholas; Wurden, Glen Anthony; Palaniyappan, Sasikumar; McCary, E.

2015-01-01

Covered in the PowerPoint presentation are the following areas: Motivation and requirements for active interrogation of nuclear material; laser-driven neutron source; neutron diagnostics; active interrogation of nuclear material; and, conclusions, remarks, and future works.

System control for the modulated 252Cf source ''Shuffler''

International Nuclear Information System (INIS)

Stephens, M.M.

1975-06-01

The design and theory of operation of the control chassis for a 252 Cf nondestructive assay system are described. This system repetitively transfers a 252 Cf source from the irradiation region to a shielded position before measuring the delayed neutrons. The design criteria for the system were: rapid movement and precise positioning of the 252 Cf source, precise positioning of the sample, and very accurate timing of the irradiate and count cycles. To achieve these results crystal oscillators were used for timing, and stepping motors were used to position the sample and the source. (U.S.)

Resistive plate chamber neutron and gamma sensitivity measurement with a {sup 252}Cf source

Energy Technology Data Exchange (ETDEWEB)

Abbrescia, M.; Altieri, S.; Baratti, V.; Barnaba, O.; Belli, G.; Bruno, G.; Colaleo, A.; DeVecchi, C.; Guida, R. E-mail: roberto.guida@pv.infn.it; Iaselli, G.; Imbres, E.; Loddo, F.; Maggi, M.; Marangelli, B.; Musitelli, G.; Nardo, R.; Natali, S.; Nuzzo, S.; Pugliese, G.; Ranieri, A.; Ratti, S.; Riccardi, C.; Romano, F.; Torre, P.; Vicini, A.; Vitulo, P.; Volpe, F

2003-06-21

A bakelite double gap Resistive Plate Chamber (RPC), operating in avalanche mode, has been exposed to the radiation emitted from a {sup 252}Cf source to measure its neutron and gamma sensitivity. One of the two gaps underwent the traditional electrodes surface coating with linseed oil. RPC signals were triggered by fission events detected using BaF{sub 2} scintillators. A Monte Carlo code, inside the GEANT 3.21 framework with MICAP interface, has been used to identify the gamma and neutron contributions to the total number of collected RPC signals. A neutron sensitivity of (0.63{+-}0.02)x10{sup -3} (average energy 2 MeV) and a gamma sensitivity of (14.0{+-}0.5)x10{sup -3} (average energy 1.5 MeV) have been measured in double gap mode. Measurements done in single gap mode have shown that both neutron and gamma sensitivity are independent of the oiling treatment.

Detector point of view of reactor internal vibrations under Gaussian coloured random forces - the problem of fitting neutron noise experimental data

International Nuclear Information System (INIS)

Arnal, R.S.; Martin, G.V.; Gonzalez, J.L.M.-C.

1988-01-01

This paper studies the local vibrations of reactor components driven by Gaussian coloured and white forces, when nonlinear vibrations arise. We study also the important problem of noise sources, modelization and the noise propagation through the neutron field using the discrete ordinates transport theory. Finally, we study the effect of the neutron field upon the PSD (power spectral density) of the noise source and we analyse the problem of fitting neutron noise experimental data to perform pattern recognition analysis. (author)

Characterization of a high repetition-rate laser-driven short-pulsed neutron source

Science.gov (United States)

Hah, J.; Nees, J. A.; Hammig, M. D.; Krushelnick, K.; Thomas, A. G. R.

2018-05-01

We demonstrate a repetitive, high flux, short-pulsed laser-driven neutron source using a heavy-water jet target. We measure neutron generation at 1/2 kHz repetition rate using several-mJ pulse energies, yielding a time-averaged neutron flux of 2 × 105 neutrons s‑1 (into 4π steradians). Deuteron spectra are also measured in order to understand source characteristics. Analyses of time-of-flight neutron spectra indicate that two separate populations of neutrons, ‘prompt’ and ‘delayed’, are generated at different locations. Gamma-ray emission from neutron capture 1H(n,γ) is also measured to confirm the neutron flux.

The fast neutron emission spectrum of 252-Cf

International Nuclear Information System (INIS)

Bensch, F.

1979-07-01

The aim of this work was a new measurement of the neutron emission spectrum of 252-Cf neutron standard sources as the IAEA is offering to users. The main feature was the application of gas-filled proton-recoil spectrometers and no TOF technique. The special interest of this document was in the temperature parameter of the Maxwellian distribution and in its relative deviations. In this connection, special measurements with high energy resolution were carried out in a search for fine structure neutron groups, which have been observed in some TOF measurements, but could not be reproduced during this measurement

Study the Possibility of measuring Heavy Metals Arising from Gold Mining Sample using 252Cf Neutron Source

International Nuclear Information System (INIS)

Soliman, N.F.

2012-01-01

Mining activities contribute immensely to trace element pollution .The present work studied the possibility of measuring heavy metals arising from gold mining areas Eastern Desert in Egypt using the isotopic 252 Cf neutron source. 38.5 g of the sample is irradiated in a neutron field of 252 Cf for 36.75 days and 0.3 mg of the sample is irradiated at the Pneumatic Irradiation Rabbit System (PIRS) built in the vertical thermal column of the ET-RR-2 reactor . The induced activities in the samples are measured using two high resolution f ray spectrometry systems (including two calibrated HPGe detectors). The kο standardization neutron activation analysis (kο-NAA) method was used to analyze the sample under investigation. Thermal to epithermal flux ratio fat the two irradiation positions are measured. The obtained results indicated that the values of the suspected heavy metals such as sodium, magnesium, potassium, manganese, arsenic, ruthenium and tantalum were higher than the background values, signifying accumulation of these heavy metals due to gold mining activities .The accuracy and precision of the method has been evaluated

Neutron spectra and dosimetric features of isotopic neutron sources: a review

International Nuclear Information System (INIS)

Vega C, H. R.; Martinez O, S. A.

2015-10-01

A convenient way to produce neutrons is the isotopic neutron source, where the production is through (α, n), (γ, n), and spontaneous fission reactions. Isotopic neutron sources are small, easy to handle, and have a relative low cost. On the other hand the neutron yield is small and mostly of them produces neutrons with a wide energy distribution. In this work, a review is carried out about the the main features of 24 NaBe, 24 NaD 2 O, 116 InBe, 140 LaBe, 238 PuLi, 239 PuBe, 241 AmB, 241 AmBe, 241 AmF, 241 AmLi, 242 CmBe, 210 PoBe, 226 RaBe, 252 Cf and 252 Cf/D 2 O isotopic neutron source. Also, using Monte Carlo methods, the neutron spectra in 31 energy groups, the neutron mean energy; the Ambient dose equivalent, the Personal dose equivalent and the Effective dose were calculated for these isotopic neutron sources. (Author)

Neutron energy spectra of sup 2 sup 5 sup 2 Cf, Am-Be source and of the D(d,n) sup 3 He reaction

CERN Document Server

Sang Tae Park

2003-01-01

The neutron energy spectrum of the following sources were measured using a fast neutron spectrometer with the NE-213 liquid scintillator: sup 2 sup 5 sup 2 Cf, Am-Be and D(d,n) sup 3 He reaction from a 3 MeV Pelletron accelerator in Tokyo Institute of Technology. The measured proton recoil pulse height data of sup 2 sup 5 sup 2 Cf, Am-Be and D(d,n) sup 3 He were unfolded using the mathematical program to obtain the neutron energy spectrum. The sup 2 sup 5 sup 2 Cf and Am-Be neutron energy spectra were measured and the results obtained showed a good agreement with the spectra usually published in the literature. The neutron energy spectrum from D(d,n) sup 3 He was measured and the results obtained also showed a good agreement with the calculation by time of flight (TOF) methods. (author)

Investigation of the characteristics of 252Cf-detectors

International Nuclear Information System (INIS)

Karlsson, Erik

2004-12-01

In the first chapter the characteristic behaviors of two Cf detectors have been investigated by performing pilot measurements. The detector with the stronger source gives an unstable signal with a low signal/noise ratio. Therefore this detector has not been further investigated. The ionization chamber reacts on both fission products and alpha decay. An energy experiment showed that there were large difficulties to separate those decays. A plastic scintillator, which reacts on both photons and neutrons, was used for neutron detection. Energy spectrums were performed and the result showed that it is difficult to set an energy threshold to separate the neutrons and the photons. The discrimination will rather be achieved by time of flight methods which is discussed under the second chapter in this thesis; Experimental results. An other experiment was done in order to investigate whether it is possible to detect any delayed components from the spontaneous fission of Cf. The result showed that delayed components existed. Either they are delayed neutrons from exited fission products, or it is some delay related to the charge collection in the Cf detector. Correlation measurements showed that few events are coincident. Only 50% of the signals from the plastic scintillator are correlated with the Cf source

Synchrotron-driven spallation sources

CERN Document Server

Bryant, P J

1996-01-01

The use of synchrotrons for pulsed neutron spallation sources is an example of scientific and technological spin-off from the accelerator development for particle physics. Accelerator-driven sources provide an alternative to the continuous-flux, nuclear reactors that currently furnish the majority of neutrons for research and development. Although the present demand for neutrons can be adequately met by the existing reactors, this situation is unlikely to continue due to the increasing severity of safety regulations and the declared policies of many countries to close down their reactors within the next decade or so. Since the demand for neutrons as a research tool is, in any case,expected to grow, there has been a corresponding interest in sources that are synchrotron-driven or linac-driven with a pulse compression ring and currently several design studies are being made. These accelerator-driven sources also have the advantage of a time structure with a high peak neutron flux. The basic requirement is for a...

Experimental subcritical facility driven by D-D/D-T neutron generator at BARC, India

Energy Technology Data Exchange (ETDEWEB)

Sinha, Amar, E-mail: image@barc.gov.in; Roy, Tushar; Kashyap, Yogesh; Ray, Nirmal; Shukla, Mayank; Patel, Tarun; Bajpai, Shefali; Sarkar, P.S.; Bishnoi, Saroj

2015-05-01

Highlights: •Experimental subcritical facility BRAHMMA coupled to D-D/D-T neutron generator. •Preliminary results of PNS experiments reported. •Feynman-alpha noise measurements explored with continuous source. -- Abstract: The paper presents design of an experimental subcritical assembly driven by D-D/D-T neutron and preliminary experimental measurements. The system has been developed for investigating the static and dynamic neutronic properties of accelerator driven sub-critical systems. This system is modular in design and it is first in the series of subcritical assemblies being designed. The subcritical core consists of natural uranium fuel with high density polyethylene as moderator and beryllium oxide as reflector. The fuel is embedded in high density polyethylene moderator matrix. Estimated k{sub eff} of the system is ∼0.89. One of the unique features of subcritical core is the use of Beryllium oxide (BeO) as reflector and HDPE as moderator making the assembly a compact modular system. The subcritical core is coupled to Purnima Neutron Generator which works in D-D and D-T mode with both DC and pulsed operation. It has facility for online source strength monitoring using neutron tagging and programmable source modulation. Preliminary experiments have been carried out for spatial flux measurement and reactivity estimation using pulsed neutron source (PNS) techniques with D-D neutrons. Further experiments are being planned to measure the reactivity and other kinetic parameters using noise methods. This facility would also be used for carrying out studies on effect of source importance and measurement of source multiplication factor k{sub s} and external neutron source efficiency φ{sup ∗} in great details. Experiments with D-T neutrons are also underway.

Measurement of multiple α-modes at the Delphi subcritical assembly by neutron noise techniques

International Nuclear Information System (INIS)

Szieberth, Máté; Klujber, Gergely; Kloosterman, Jan Leen; Haas, Dick de

2015-01-01

Highlights: • Neutron noise measurements were performed at the Delphi subcritical assembly. • Bias in the fitted prompt decay constant was observed due to higher modes. • Spatial dependence of the higher mode was surveyed. • Effect of two different source distributions was investigated. • An estimation of the prompt decay constant is given for the Delphi. - Abstract: The paper presents the results and evaluations of a comprehensive set of neutron noise measurements on the Delphi subcritical assembly of the Delft University of Technology. The measurements investigated the effect of different source distributions (inherent spontaneous fission and 252 Cf) and the position of the detectors applied (both radially and vertically). The evaluation of the measured data has been performed by the variance-to-mean (VTM, Feynman-α), the autocorrelation (ACF, Rossi-α) and the cross-correlation (CCF) methods. The values obtained for the prompt decay constant show a strong bias, which depends both on the detector position and on the source distribution. This is due to the presence of higher modes in the system. It has been observed that the α value fitted is higher when the detector is close to the boundary of the core or to the 252 Cf point-source. The higher alpha-modes have also been observed by fitting functions describing two alpha-modes. The successful set of measurements also provides a good basis for further theoretical investigations including the Monte Carlo simulation of the noise measurements and the calculation of the alpha-modes in the Delphi subcritical assembly

Development of neutron induced prompt γ-ray spectroscopy system using 252Cf

International Nuclear Information System (INIS)

Park, Yong-Joon; Song, Byung-Chul; Jee, Kwang-Yong

2003-01-01

For the design and set-up of neutron induced prompt γ-ray spectroscopy system using 252 Cf neutron source, the effects of shielding and moderator materials have been examined. The 252 Cf source being used for TLD badge calibration in Korea Atomic Energy Research Institute was utilized for this preliminary experiment. The γ-ray background and prompt γ-ray spectrum of the sample containing Cl were measured using HPGe (GMX 69% relative efficiency) located at the inside of the system connected to notebook PC at the outside of the system (about 20 meter distance). The background activities of neutron and γ-rays were measured with neutron survey meter as well as γ-ray survey meters, respectively and the system was designed to minimize the activities. Prompt γ-ray spectrum was measured using γ-γ coincident system for reduce the background and the continuum spectrum. The optimum system was designed and set up using the experimental data obtained

Neutron spectra and dosimetric features of isotopic neutron sources: a review

Energy Technology Data Exchange (ETDEWEB)

Vega C, H. R. [Universidad Autonoma de Zacatecas, Unidad Academica de Estudios Nucleares, Cipres No. 10, Fracc. La Penuela, 98060 Zacatecas, Zac. (Mexico); Martinez O, S. A., E-mail: fermineutron@yahoo.com [Universidad Pedagogica y Tecnologica de Colombia, Grupo de Fisica Nuclear Aplicada y Simulacion, Av. Central del Norte 39-115, 150003 Tunja, Boyaca (Colombia)

2015-10-15

A convenient way to produce neutrons is the isotopic neutron source, where the production is through (α, n), (γ, n), and spontaneous fission reactions. Isotopic neutron sources are small, easy to handle, and have a relative low cost. On the other hand the neutron yield is small and mostly of them produces neutrons with a wide energy distribution. In this work, a review is carried out about the the main features of {sup 24}NaBe, {sup 24}NaD{sub 2}O, {sup 116}InBe, {sup 140}LaBe, {sup 238}PuLi, {sup 239}PuBe, {sup 241}AmB, {sup 241}AmBe, {sup 241}AmF, {sup 241}AmLi, {sup 242}CmBe, {sup 210}PoBe, {sup 226}RaBe, {sup 252}Cf and {sup 252}Cf/D{sub 2}O isotopic neutron source. Also, using Monte Carlo methods, the neutron spectra in 31 energy groups, the neutron mean energy; the Ambient dose equivalent, the Personal dose equivalent and the Effective dose were calculated for these isotopic neutron sources. (Author)

New applications of laser-driven neutron sources in the car industry

International Nuclear Information System (INIS)

Kakeno, Mitsutaka

2015-01-01

New applications of LDNS (Laser-Driven Neutron Sources) are described. One of them is ib-DATA (in-beam Double Activation Tracer Analysis) with which we can measure mean drift velocity and mass flow rate in a variety of fluid. In ib-DATA, LDNS with very light and compact beam-head will be constructed to shoot pulsed neutrons into the fluid in pinpoint. (author)

Monte Carlo simulation using MCNP4B for an optimal shielding design of a 252 Cf source

International Nuclear Information System (INIS)

Silva, Ademir X. da; Crispim, Verginia R.

2001-01-01

This study aim to investigate an optimum shielding design against neutrons and gamma-rays from a source of 252 Cf, using Monte Carlo simulation. The shielding materials studied were: borated polyethylene, borated-lead polyethylene and stainless steel. The Monte Carlo code MCNP, version 4B, was used to design shielding for 252 Cf based neutron irradiator systems. By normalizing the dose equivalent rate values presented to the neutron production rate of the source, the resulting calculations are independents of the intensity of actual 252 Cf source. The results shown what the total dose equivalent rates were reduced significantly by the shielding system optimization. (author)

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

International Nuclear Information System (INIS)

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

1999-01-01

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

Fast neutron therapy with high intensity Cf-252 sources by remotely controlled afterloading and clinical experiences in the treatment of gynaecological cancers

International Nuclear Information System (INIS)

Yamashita, H.; Hashimoto, S.; Wada, M.; Dokiya, T.

1986-01-01

Cf-252 fast neutron therapy with high intensity Cf-252 sources was tested for the treatment of advanced gynaecological cancers using a remotely controlled afterloading machine designed by the author and manufactured by Toshiba. Using high intensity sources and short treatment times in a special treatment room, personnel or environment exposure to radiation was at a safe level, i.e. almost nil. During 1978-1983 18 stage III cases of cancer of the uterine cervix were treated with complete response in 78% and 44% 5 year survivals. The types of acute and delayed effects of Cf-252 were the same as Co-60 or Cs-137 but the rectum was found sensitive in this system of brachytherapy. A dose of 1,000-1,500 cGy/6-10 F in 10-22 days of Cf-252 radiation was tolerated and produced tumor cure

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.

Base neutron noise in PWRs

International Nuclear Information System (INIS)

Kosaly, G.; Albrecht, R.W.; Dailey, D.J.; Fry, D.N.

1981-01-01

Considerable activity has been devoted in recent years to the use of neutron noise for investigation of problems in pressurized-water reactors (PWRs). The investigators have found that neutron noise provides an effective way to monitor reactor internal vibrations such as vertical and lateral core motion; core support barrel and thermal shield shell modes, bending modes of fuel assemblies, and control rod vibrations. However, noise analysts have also concluded that diagnosis of a problem is easier if baseline data for normal plant operation is available. Therefore, the authors have obtained ex-core neutron noise signatures from eight PWRs to determine the similarity of signatures between plants and to build a base of data to determine the sources of neutron noise and thus the potential diagnostic information contained in the data. It is concluded that: (1) ex-core neutron noise contains information about the vibration of components in the pressure vessel; (2) baseline signature acquisition can aid understanding of plant specific vibration frequencies and provide a bases for diagnosis of future problems if they occur; and (3) abnormal core support barrel vibration can most likely be detected over and above the plant-to-plant signature variation observed thus far

Neutron beam experiments using nuclear research reactors: honoring the retirement of professor Bernard W. Wehring -II. 5. Testing Moderating Detection Systems with 252Cf-Based Reference Neutron Fields

International Nuclear Information System (INIS)

Hertel, Nolan E.; Sweezy, Jeremy; Sauber, Jeremiah S.; Vaughn, David; Cook, Andrew; Tays, Jeff; Ro, Tae-Ik

2001-01-01

In recent years, Georgia Institute of Technology (Georgia Tech) has been involved in a number of neutron dosimetry research projects. Several reference neutron fields are now available for such projects. They are all based on the use of a 252 Cf source. The source can be used by itself to create a reference un-moderated 252 Cf neutron field, or it can be placed inside several different moderating assemblies. The spectra created by placing the source inside these assemblies and the un-moderated source are employed to investigate detector and dosimeter responses. Currently, the set of moderators available includes a 30-cm diam cadmium-covered D 2 O spherical shell, a 30-cm-thick iron spherical shell, a 30-cm-diam polyethylene spherical shell, an 18.3-cm-thick tungsten spherical shell, a 16-cm-thick lead spherical shell, and a 9-cm-thick tantalum spherical shell. In addition, the 252 Cf source can be placed inside a neutron howitzer recently constructed at Georgia Tech. The howitzer is a WEP cylinder loaded with boron that has a 10.16-cm-diam cylindrical opening. When the source is placed in the cylindrical penetration of the howitzer, a neutron field ∼30 cm in diameter is created at a distance of 50 cm from the californium source. Over the last few years, Bonner sphere spectrometers using LiI(Eu) scintillators and LiF thermoluminescence dosimeters have been calibrated using this facility at Georgia Tech. Recently, the Neely Nuclear Research Center (NNRC) acquired an LB 6411 neutron probe (product of EG and G Berthold). This probe is designed to measure ambient dose equivalent in accordance with International Commission on Radiological Protection Publication 60 recommendations. It consists of a cylindrical 3 He proportional counter surrounded by a 25-cm-diam spherical polyethylene moderator. Its neutron response is optimized for dose rate measurements of neutrons between thermal energies and 20 MeV (Ref. 5). As a test of the instrument's ability to measure ambient

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

CERN Document Server

Matsumoto, T

2003-01-01

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

Fission-neutrons source with fast neutron-emission timing

Energy Technology Data Exchange (ETDEWEB)

Rusev, G., E-mail: rusev@lanl.gov; Baramsai, B.; Bond, E.M.; Jandel, M.

2016-05-01

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

A new facility for non-destructive assay using a 252Cf source

International Nuclear Information System (INIS)

Stevanato, L.; Caldogno, M.; Dima, R.; Fabris, D.; Hao, Xin; Lunardon, M.; Moretto, S.; Nebbia, G.; Pesente, S.; Pino, F.; Sajo-Bohus, L.; Viesti, G.

2013-01-01

A new laboratory facility for non-destructive analysis (NDA) using a time-tagged 252 Cf source is presented. The system is designed to analyze samples having maximum size of about 20×25 cm 2 , the material recognition being obtained by measuring simultaneously total and energy dependent transmission of neutrons and gamma rays. The equipment technical characteristics and performances of the NDA system are presented, exploring also limits due to the sample thickness. Some recent applications in the field of cultural heritage are presented. - Highlights: ► Tagged 252 Cf source setup. ► Material recognition and sample imaging with measurements of gamma ray and neutron transmission. ► Material identification via energy dependent neutron and gamma ray transmission measurements. ► Identification of layered material. ► Effects due to the sample thickness on the above identification techniques

Design, Construction, and Modeling of a 252Cf Neutron Irradiator

Directory of Open Access Journals (Sweden)

Blake C. Anderson

2016-01-01

Full Text Available Neutron production methods are an integral part of research and analysis for an array of applications. This paper examines methods of neutron production, and the advantages of constructing a radioisotopic neutron irradiator assembly using 252Cf. Characteristic neutron behavior and cost-benefit comparative analysis between alternative modes of neutron production are also examined. The irradiator is described from initial conception to the finished design. MCNP modeling shows a total neutron flux of 3 × 105 n/(cm2·s in the irradiation chamber for a 25 μg source. Measurements of the gamma-ray and neutron dose rates near the external surface of the irradiator assembly are 120 μGy/h and 30 μSv/h, respectively, during irradiation. At completion of the project, total material, and labor costs remained below $50,000.

Assessment of Laser-Driven Pulsed Neutron Sources for Poolside Neutron-based Advanced NDE – A Pathway to LANSCE-like Characterization at INL

Energy Technology Data Exchange (ETDEWEB)

Roth, Markus [Technische Univ. Darmstadt (Germany); Vogel, Sven C. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Bourke, Mark Andrew M. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Fernandez, Juan Carlos [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Mocko, Michael Jeffrey [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Glenzer, Siegfried [Stanford Univ., CA (United States); Leemans, Wim [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Siders, Craig [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Haefner, Constantin [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States)

2017-04-19

A variety of opportunities for characterization of fresh nuclear fuels using thermal (~25meV) and epithermal (~10eV) neutrons have been documented at Los Alamos National Laboratory. They include spatially resolved non-destructive characterization of features, isotopic enrichment, chemical heterogeneity and stoichiometry. The LANSCE spallation neutron source is well suited in neutron fluence and temporal characteristics for studies of fuels. However, recent advances in high power short pulse lasers suggest that compact neutron sources might, over the next decade, become viable at a price point that would permit their consideration for poolside characterization on site at irradiation facilities. In a laser-driven neutron source the laser is used to accelerate deuterium ions into a beryllium target where neutrons are produced. At this time, the technology is new and their total neutron production is approximately four orders of magnitude less than a facility like LANSCE. However, recent measurements on a sub-optimized system demonstrated >10¹⁰ neutrons in sub-nanosecond pulses in predominantly forward direction. The compactness of the target system compared to a spallation target may allow exchanging the target during a measurement to e.g. characterize a highly radioactive sample with thermal, epithermal, and fast neutrons as well as hard X-rays, thus avoiding sample handling. At this time several groups are working on laser-driven neutron production and are advancing concepts for lasers, laser targets, and optimized neutron target/moderator systems. Advances in performance sufficient to enable poolside fuels characterization with LANSCE-like fluence on sample within a decade may be possible. This report describes the underlying physics and state-of-the-art of the laser-driven neutron production process from the perspective of the DOE/NE mission. It also discusses the development and understanding that will be necessary to provide customized capability for

Evaluation of Importance of Source Neutrons in Accelerator-Driven System

International Nuclear Information System (INIS)

Kim, Yong Hee; Park, Won Seok

2002-01-01

An importance function of the external spallation neutrons in ADS (Accelerator-Driven System) is defined to characterize the source multiplication in subcritical blanket. For a model ADS problem, the source importance function is evaluated with the TRANSX/TWODANT code system. In order to assess the impact of the power distribution on the importance function, both homogeneous and heterogeneous cores are analyzed and corresponding source multiplications are compared. Also, based on the source importance function, an optimization of the shape of the proton current is performed from the source multiplication point of view. Additionally, the source importance function is compared with the conventional λ-mode adjoint flux, which is used as an importance function of fission neutrons in the critical reactors. Concerning an issue in the ADS design, i.e., difficulty in reducing the fission power unless the proton current is shut off, a study is performed to minimize the source importance, thereby minimizing the fission power, even when the k-eff value of the core is quite high. (authors)

New shielding material development for compact accelerator-driven neutron source

Directory of Open Access Journals (Sweden)

Guang Hu

2017-04-01

Full Text Available The Compact Accelerator-driven Neutron Source (CANS, especially the transportable neutron source is longing for high effectiveness shielding material. For this reason, new shielding material is researched in this investigation. The component of shielding material is designed and many samples are manufactured. Then the attenuation detection experiments were carried out. In the detections, the dead time of the detector appeases when the proton beam is too strong. To grasp the linear range and nonlinear range of the detector, two currents of proton are employed in Pb attenuation detections. The transmission ratio of new shielding material, polyethylene (PE, PE + Pb, BPE + Pb is detected under suitable current of proton. Since the results of experimental neutrons and γ-rays appear as together, the MCNP and PHITS simulations are applied to assisting the analysis. The new shielding material could reduce of the weight and volume compared with BPE + Pb and PE + Pb.

Cf-252 based neutron radiography using real-time image processing system

International Nuclear Information System (INIS)

Mochiki, Koh-ichi; Koiso, Manabu; Yamaji, Akihiro; Iwata, Hideki; Kihara, Yoshitaka; Sano, Shigeru; Murata, Yutaka

2001-01-01

For compact Cf-252 based neutron radiography, a real-time image processing system by particle counting technique has been developed. The electronic imaging system consists of a supersensitive imaging camera, a real-time corrector, a real-time binary converter, a real-time calculator for centroid, a display monitor and a computer. Three types of accumulated NR image; ordinary, binary and centroid images, can be observed during a measurement. Accumulated NR images were taken by the centroid mode, the binary mode and ordinary mode using of Cf-252 neutron source and those images were compared. The centroid mode presented the sharpest image and its statistical characteristics followed the Poisson distribution, while the ordinary mode showed the smoothest image as the averaging effect by particle bright spots with distributed brightness was most dominant. (author)

Neutron dosimetry at SLAC: Neutron sources and instrumentation

International Nuclear Information System (INIS)

Liu, J.C.; Jenkins, T.M.; McCall, R.C.; Ipe, N.E.

1991-10-01

This report summarizes in detail the dosimetric characteristics of the five radioisotopic type neutron sources ( 238 PuBe, 252 Cf, 238 PuB, 238 PuF 4 , and 238 PuLi) and the neutron instrumentation (moderated BF 3 detector, Anderson-Braun (AB) detector, AB remmeter, Victoreen 488 Neutron Survey Meter, Beam Shut-Off Ionization Chamber, 12 C plastic scintillator detector, moderated indium foil detector, and moderated and bare TLDs) that are commonly used for neutron dosimetry at the Stanford Linear Accelerator Center (SLAC). 36 refs,. 19 figs

Application of californium-252 neutron sources for analytical chemistry

International Nuclear Information System (INIS)

Ishii, Daido

1976-01-01

The researches made for the application of Cf-252 neutron sources to analytical chemistry during the period from 1970 to 1974 including partly 1975 are reviewed. The first part is the introduction to the above. The second part deals with general review of symposia, publications and the like. Attention is directed to ERDA publishing the periodical ''Californium-252 Progress'' and to a study group of Cf-252 utilization held by Japanese Radioisotope Association in 1974. The third part deals with its application for radio activation analysis. The automated absolute activation analysis (AAAA) of Savannha River is briefly explained. The joint experiment of Savannha River operation office with New Brunswick laboratory is mentioned. Cf-252 radiation source was used for the non-destructive analysis of elements in river water. East neutrons of Cf-252 were used for the quantitative analysis of lead in paints. Many applications for industrial control processes have been reported. Attention is drawn to the application of Cf-252 neutron sources for the field search of neutral resources. For example, a logging sonde for searching uranium resources was developed. the fourth part deals with the application of the analysis with gamma ray by capturing neutrons. For example, a bore hole sonde and the process control analysis of sulfur in fuel utilized capture gamma ray. The prompt gamma ray by capturing neutrons may be used for the nondestructive analysis of enrivonment. (Iwakiri, K.)

Benchmarking shielding simulations for an accelerator-driven spallation neutron source

Directory of Open Access Journals (Sweden)

Nataliia Cherkashyna

2015-08-01

Full Text Available The shielding at an accelerator-driven spallation neutron facility plays a critical role in the performance of the neutron scattering instruments, the overall safety, and the total cost of the facility. Accurate simulation of shielding components is thus key for the design of upcoming facilities, such as the European Spallation Source (ESS, currently in construction in Lund, Sweden. In this paper, we present a comparative study between the measured and the simulated neutron background at the Swiss Spallation Neutron Source (SINQ, at the Paul Scherrer Institute (PSI, Villigen, Switzerland. The measurements were carried out at several positions along the SINQ monolith wall with the neutron dosimeter WENDI-2, which has a well-characterized response up to 5 GeV. The simulations were performed using the Monte-Carlo radiation transport code geant4, and include a complete transport from the proton beam to the measurement locations in a single calculation. An agreement between measurements and simulations is about a factor of 2 for the points where the measured radiation dose is above the background level, which is a satisfactory result for such simulations spanning many energy regimes, different physics processes and transport through several meters of shielding materials. The neutrons contributing to the radiation field emanating from the monolith were confirmed to originate from neutrons with energies above 1 MeV in the target region. The current work validates geant4 as being well suited for deep-shielding calculations at accelerator-based spallation sources. We also extrapolate what the simulated flux levels might imply for short (several tens of meters instruments at ESS.

Thermal neutron imaging with rare-earth-ion-doped LiCaAlF6 scintillators and a sealed 252Cf source

International Nuclear Information System (INIS)

Kawaguchi, Noriaki; Yanagida, Takayuki; Fujimoto, Yutaka; Yokota, Yuui; Kamada, Kei; Fukuda, Kentaro; Suyama, Toshihisa; Watanabe, Kenichi; Yamazaki, Atsushi; Chani, Valery; Yoshikawa, Akira

2011-01-01

Thermal neutron imaging with Ce-doped LiCaAlF 6 crystals has been performed. The prototype of the neutron imager using a Ce-doped LiCaAlF 6 scintillating crystal and a position sensitive photomultiplier tube (PSPMT) which had 64 multi-channel anode was developed. The Ce-doped LiCaAlF 6 single crystal was grown by the Czochralski method. A plate with dimensions of a diameter of 50x2 mm 2 was cut from the grown crystal, polished, and optically coupled to PSPMT by silicone grease. The 252 Cf source ( 6 .

Neutron dosimetry at SLAC: Neutron sources and instrumentation

Energy Technology Data Exchange (ETDEWEB)

Liu, J.C.; Jenkins, T.M.; McCall, R.C.; Ipe, N.E.

1991-10-01

This report summarizes in detail the dosimetric characteristics of the five radioisotopic type neutron sources ({sup 238}PuBe, {sup 252}Cf, {sup 238}PuB, {sup 238}PuF{sub 4}, and {sup 238}PuLi) and the neutron instrumentation (moderated BF{sub 3} detector, Anderson-Braun (AB) detector, AB remmeter, Victoreen 488 Neutron Survey Meter, Beam Shut-Off Ionization Chamber, {sup 12}C plastic scintillator detector, moderated indium foil detector, and moderated and bare TLDs) that are commonly used for neutron dosimetry at the Stanford Linear Accelerator Center (SLAC). 36 refs,. 19 figs.

Power reactor noise studies and applications

Energy Technology Data Exchange (ETDEWEB)

Arzhanov, V

2002-03-01

The present thesis deals with the neutron noise arising in power reactor systems. Generally, it can be divided into two major parts: first, neutron noise diagnostics, or more specifically, novel methods and algorithms to monitor nuclear industrial reactors; and second, contributions to neutron noise theory as applied to power reactor systems. Neutron noise diagnostics is presented by two topics. The first one is a theoretical study on the possibility to use a newly proposed current-flux (C/F) detector in Pressurised Water Reactors (PWR) for the localisation of anomalies. The second topic concerns various methods to detect guide tube impacting in Boiling Water Reactors (BWR). The significance of these problems comes from the operational experience. The thesis describes a novel method to localise vibrating control rods in a PWR by using only one C/F detector. Another novel method, based on wavelet analysis, is put forward to detect impacting guide tubes in a BWR. Neutron noise theory is developed for both Accelerator Driven Systems (ADS) and traditional reactors. By design the accelerator-driven systems would operate in a subcritical mode with a strong external source. This calls for a revision of many concepts and methods that have been developed for traditional reactors and also it poses a number of new problems. As for the latter, the thesis investigates the space-dependent neutron noise caused by a fluctuating source. It is shown that the frequency-dependent spatial behaviour exhibits some new properties that are different from those known in traditional critical systems. On the other hand, various reactor physics approximations (point kinetic, adiabatic etc.) have not been defined yet for the subcritical systems. In this respect the thesis presents a systematic formulation of the above mentioned approximations as well as investigations of their properties. Another important problem in neutron noise theory is the treatment of moving boundaries. In this case one

Power reactor noise studies and applications

International Nuclear Information System (INIS)

Arzhanov, V.

2002-03-01

The present thesis deals with the neutron noise arising in power reactor systems. Generally, it can be divided into two major parts: first, neutron noise diagnostics, or more specifically, novel methods and algorithms to monitor nuclear industrial reactors; and second, contributions to neutron noise theory as applied to power reactor systems. Neutron noise diagnostics is presented by two topics. The first one is a theoretical study on the possibility to use a newly proposed current-flux (C/F) detector in Pressurised Water Reactors (PWR) for the localisation of anomalies. The second topic concerns various methods to detect guide tube impacting in Boiling Water Reactors (BWR). The significance of these problems comes from the operational experience. The thesis describes a novel method to localise vibrating control rods in a PWR by using only one C/F detector. Another novel method, based on wavelet analysis, is put forward to detect impacting guide tubes in a BWR. Neutron noise theory is developed for both Accelerator Driven Systems (ADS) and traditional reactors. By design the accelerator-driven systems would operate in a subcritical mode with a strong external source. This calls for a revision of many concepts and methods that have been developed for traditional reactors and also it poses a number of new problems. As for the latter, the thesis investigates the space-dependent neutron noise caused by a fluctuating source. It is shown that the frequency-dependent spatial behaviour exhibits some new properties that are different from those known in traditional critical systems. On the other hand, various reactor physics approximations (point kinetic, adiabatic etc.) have not been defined yet for the subcritical systems. In this respect the thesis presents a systematic formulation of the above mentioned approximations as well as investigations of their properties. Another important problem in neutron noise theory is the treatment of moving boundaries. In this case one

Neutron noise in nuclear reactors

International Nuclear Information System (INIS)

Blaquiere, A.; Pachowska, R.

1961-06-01

The power of a nuclear reactor, in the operating conditions, presents fluctuations due to various causes. This random behaviour can be included in the study of 'noises'. Among other sources of noise, we analyse hereafter the fluctuations due: a) to the discontinuous emissions of neutrons from an independent source; b) to the multiplication of neutrons inside the reactor. The method which we present makes use of the analogies between the rules governing a nuclear reactor in operation and a number of radio-electrical systems, in particular the feed-back loops. The reactor can be characterized by its 'passing band' and is described as a system submitted to a sequence of random pulses. In non linear operating condition, the effect of neutron noise is defined by means of a non-linear functional, this theory is thus related to previous works the references of which are given at the end of the present report. This leads us in particular in the case of nuclear reactors to some results given by A. Blaquiere in the case of radio-electrical loops. (author) [fr

Neutron noise analysis of BWR using time series analysis

International Nuclear Information System (INIS)

Fukunishi, Kohyu

1976-01-01

The main purpose of this paper is to give more quantitative understanding of noise source in neutron flux and to provide a useful tool for the detection and diagnosis of reactor. The space dependent effects of distributed neutron flux signals at the axial direction of two different strings are investigated by the power contribution ratio among neutron fluxes and the incoherent noise spectra of neutron fluxes derived from autoregressive spectra. The signals are measured on the medium sized commercial BWR of 460 MWe in Japan. From the obtained results, local and global noise sources in neutron flux are discussed. This method is indicated to be a useful tool for detection and diagnosis of anomalous phenomena in BWR. (orig./RW) [de

Neutron emission during acceleration of 252Cf fission fragments

International Nuclear Information System (INIS)

Batenkov, O.I.; Blinov, M.V.; Blinov, A.B.; Smirnov, S.N.

1991-01-01

We investigate neutron emission during acceleration of fission fragments in the process of spontaneous fission of 252 Cf. Experimental angular and energy distributions of neutrons are compared with the results of calculations of neutron evaporation during fragment acceleration. (author). 8 refs, 3 figs

Microwave Ion Source and Beam Injection for an Accelerator-driven Neutron Source

International Nuclear Information System (INIS)

Vainionpaa, J.H.; Gough, R.; Hoff, M.; Kwan, J.W.; Ludewigt, B.A.; Regis, M.J.; Wallig, J.G.; Wells, R.

2007-01-01

An over-dense microwave driven ion source capable of producing deuterium (or hydrogen) beams at 100-200 mA/cm2 and with atomic fraction >90 percent was designed and tested with an electrostatic low energy beam transport section (LEBT). This ion source was incorporated into the design of an Accelerator Driven Neutron Source (ADNS). The other key components in the ADNS include a 6 MeV RFQ accelerator, a beam bending and scanning system, and a deuterium gas target. In this design a 40 mA D+ beam is produced from a 6 mm diameter aperture using a 60 kV extraction voltage. The LEBT section consists of 5 electrodes arranged to form 2 Einzel lenses that focus the beam into the RFQ entrance. To create the ECR condition, 2 induction coils are used to create ∼ 875 Gauss on axis inside the source chamber. To prevent HV breakdown in the LEBT a magnetic field clamp is necessary to minimize the field in this region. Matching of the microwave power from the waveguide to the plasma is done by an autotuner. We observed significant improvement of the beam quality after installing a boron nitride liner inside the ion source. The measured emittance data are compared with PBGUNS simulations

Microwave Ion Source and Beam Injection for an Accelerator-Driven Neutron Source

International Nuclear Information System (INIS)

Vainionpaa, J.H.; Gough, R.; Hoff, M.; Kwan, J.W.; Ludewigt, B.A.; Regis, M.J.; Wallig, J.G.; Wells, R.

2007-01-01

An over-dense microwave driven ion source capable of producing deuterium (or hydrogen) beams at 100-200 mA/cm 2 and with atomic fraction > 90% was designed and tested with an electrostatic low energy beam transport section (LEBT). This ion source was incorporated into the design of an Accelerator Driven Neutron Source (ADNS). The other key components in the ADNS include a 6 MeV RFQ accelerator, a beam bending and scanning system, and a deuterium gas target. In this design a 40 mA D + beam is produced from a 6 mm diameter aperture using a 60 kV extraction voltage. The LEBT section consists of 5 electrodes arranged to form 2 Einzel lenses that focus the beam into the RFQ entrance. To create the ECR condition, 2 induction coils are used to create ∼ 875 Gauss on axis inside the source chamber. To prevent HV breakdown in the LEBT a magnetic field clamp is necessary to minimize the field in this region. Matching of the microwave power from the waveguide to the plasma is done by an autotuner. They observed significant improvement of the beam quality after installing a boron nitride liner inside the ion source. The measured emittance data are compared with PBGUNS simulations

Calculation of neutron spectra for a 252Cf transport cask using ANISN running on a PC

International Nuclear Information System (INIS)

West, L.; Akin, B.P.; Lemley, E.C.

1995-01-01

Neutron spectra have been calculated using the ANISN one-dimensional discrete ordinates code for the case of a 152 Cf source in a transport cask of a particular design. All computations were done on personal computers (PCs) (mostly 486 models) with the ANISN-ORNL (486 version) computer code. With a source of 252 Cf fission neutrons, the neutron flux spectrum in the cask cannot be characterized as open-quotes moderated.close quotes Concern about an appropriate choice for the cross-section data set has led to a comparison, for this application, of three different cross-section libraries: DABL, HILO, and BUGLE-80. Although the cross-section sets were not originally designed for PC use, the libraries have been successfully employed for PC computations. Work with yet another data library, BUGLE-93, is incomplete at this stage. From neutron flux spectra on the surface of the cask, personnel dosimetric quantities (such as dose equivalent) have been determined for the DABL, HILO, and BUGLE-80 ANISN calculations

A fast neutron and dual-energy gamma-ray absorption method (NEUDEG) for investigating materials using a 252Cf source

International Nuclear Information System (INIS)

Bartle, C. Murray

2014-01-01

DEXA (dual-energy X-ray absorption) is widely used in airport scanners, industrial scanners and bone densitometers. DEXA determines the properties of materials by measuring the absorption differences of X-rays from a bremsstrahlung tube source with and without filtering. Filtering creates a beam with a higher mean energy, which causes lower material absorption. The absorption difference between measurements (those with a filter subtracted from those without a filter) is a positive number that increases with the effective atomic number of the material. In this paper, the concept of using a filter to create a dual beam and an absorption difference in materials is applied to radiation from a 252 Cf source, called NEUDEG (neutron and dual-energy gamma absorption). NEUDEG includes absorptions for fast neutrons as well as the dual photon beams and thus an incentive for developing the method is that, unlike DEXA, it is inherently sensitive to the hydrogen content of materials. In this paper, a model for the absorption difference and absorption sum in NEUDEG is presented using the combined gamma ray and fast neutron mass attenuation coefficients. Absorption differences can be either positive or negative in NEUDEG, increasing with increases in the effective atomic number and decreasing with increases in the hydrogen content. Sample sets of absorption difference curves are calculated for materials with typical gamma-ray and fast neutron mass attenuation coefficients. The model, which uses tabulated mass attenuated coefficients, agrees with experimental data for porcelain tiles and polyethylene sheets. The effects of “beam hardening” are also investigated. - Highlights: • Creation of a dual neutron/gamma beam from 252 Cf is described. • An absorption model is developed using mass attenuation coefficients. • A graphical method is used to show sample results from the model. • The model is successfully compared with experimental results. • The importance of

An absolute measurement of 252Cf prompt fission neutron spectrum at low energy range

International Nuclear Information System (INIS)

Lajtai, A.; Dyachenko, P.P.; Kutzaeva, L.S.; Kononov, V.N.; Androsenko, P.A.; Androsenko, A.A.

1983-01-01

Prompt neutron energy spectrum at low energies (25 keV 252 Cf spontaneous fission has been measured with a time-of-flight technique on a 30 cm flight-path. Ionization chamber and lithium-glass were used as fission fragment and neutron detectors, respectively. Lithium glasses of NE-912 (containing 6 Li) and of NE-913 (containing 7 Li) 45 mm in diameter and 9.5 mm in thickness have been employed alternatively, for the registration of fission neutrons and gammas. For the correct determination of the multiscattering effects - the main difficulty of the low energy neutron spectrum measurements - a special geometry for the neutron detector was used. Special attention was paid also to the determination of the absolute efficiency of the neutron detector. The real response function of the spectrometer was determined by a Monte-Carlo calculation. The scattering material content of the ionization chamber containing a 252 Cf source was minimized. As a result of this measurement a prompt fission neutron spectrum of Maxwell type with a T=1.42 MeV parameter was obtained at this low energy range. We did not find any neutron excess or irregularities over the Maxwellian. (author)

Status of spallation neutron source

Energy Technology Data Exchange (ETDEWEB)

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

1997-03-01

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

Cross-sections of 197Au(n,α)198Au and 63Cu(n,α)64Cu induced by 252Cf neutrons

International Nuclear Information System (INIS)

Mandal, Ranjita; Sengupta, D.; Roy, Malobika; Naik, Mamta; Bhoraskar, V.N.

2014-01-01

Analytical work, employing nuclear techniques, is normally carried out through (n,α) reaction because of the availability of neutrons either from reactors or laboratory sources such as Sb-Be, Am-Be, Ra-Be, Po-Be, 252 Cf, etc. The laboratory neutron sources are though portable and adaptable to a particular experimental arrangement, suffer from the disadvantage of slow neutron yield (except 252 Cf). In this set up since the neutrons available are monoenergetic, it was thought appropriate to initiate a program to measure cross-sections of a few nuclear reactions which have practical applications. Earlier studies on cross-section measurement of the reaction 197 Au(n,α) 198 Au and 63 Cu(n,α) 64 Cu has been carried out using different sources, monitors and techniques

Characterization of noise sources in nuclear power reactors

International Nuclear Information System (INIS)

Andhill, Gustav

2004-03-01

Algorithms for unfolding noise sources in nuclear power reactors are investigated. No preliminary knowledge of the functional form of the space dependence is assumed in contrast to the usual methods. The advantage of this is that the algorithms can be applied to various noise sources and the results can be interpreted without expert knowledge. The results can therefore be directly displayed to the plant operators. The precision will however be lower than that of the traditional methods because of the arbitrariness in the type of the noise source. Two different reactor models are studied. First a simple one-dimensional and homogeneous core is considered. Three methods for finding the noise source from the measured flux noise are investigated here. The first one is based on the inversion of an appropriate pre-calculated noise source-to-measured induced neutron noise transfer function. The second one relies on the use of the measured neutron noise as the solution of the equations giving the neutron noise induced by a given noise source. The advantage of this second method is that the noise source can be determined directly, i.e., without any Inversion of a transfer function. The second method is thus called the direct method. The last method is based on a reconstruction of the noise source by spatial Fourier expansion. The two latter techniques are found usable for different locations of the actual noise source in the 1D core. They are therefore tried on more sophisticated two-dimensional models of cores. The direct method is able both to determine the nature of the noise source and its location in 2D

Characterization of noise sources in nuclear power reactors

Energy Technology Data Exchange (ETDEWEB)

Andhill, Gustav

2004-03-01

Algorithms for unfolding noise sources in nuclear power reactors are investigated. No preliminary knowledge of the functional form of the space dependence is assumed in contrast to the usual methods. The advantage of this is that the algorithms can be applied to various noise sources and the results can be interpreted without expert knowledge. The results can therefore be directly displayed to the plant operators. The precision will however be lower than that of the traditional methods because of the arbitrariness in the type of the noise source. Two different reactor models are studied. First a simple one-dimensional and homogeneous core is considered. Three methods for finding the noise source from the measured flux noise are investigated here. The first one is based on the inversion of an appropriate pre-calculated noise source-to-measured induced neutron noise transfer function. The second one relies on the use of the measured neutron noise as the solution of the equations giving the neutron noise induced by a given noise source. The advantage of this second method is that the noise source can be determined directly, i.e., without any Inversion of a transfer function. The second method is thus called the direct method. The last method is based on a reconstruction of the noise source by spatial Fourier expansion. The two latter techniques are found usable for different locations of the actual noise source in the 1D core. They are therefore tried on more sophisticated two-dimensional models of cores. The direct method is able both to determine the nature of the noise source and its location in 2D.

Analysis of the Photoneutron Yield and Thermal Neutron Flux in an Unreflected Electron Accelerator-Driven Neutron Source

International Nuclear Information System (INIS)

Dale, Gregory E.; Gahl, John M.

2005-01-01

There are several potential uses for a high-flux thermal neutron source in both industrial and clinical applications. The viable commercial implementation of these applications requires a low-cost, high-flux thermal neutron generator suitable for installation in industrial and clinical environments. This paper describes the Monte Carlo for N-Particle modeling results of a high-flux thermal neutron source driven with an electron accelerator. An electron linear accelerator (linac), fitted with a standard X-ray converter, can produce high neutron yields in materials with low photonuclear threshold energies, such as D and 9 Be. Results indicate that a 10-MeV, 10-kW electron linac can produce on the order of 10 12 n/s in a heavy water photoneutron target. The thermal neutron flux in an unreflected heavy water target is calculated to be on the order of 10 10 n.cm -2 .s. The sensitivity of these answers to heavy water purity is also investigated, specifically the dilution of heavy water with light water. It is shown that the peak thermal neutron flux is not adversely effected by dilution up to a light water weight fraction of 35%

Characterization and application of a laser-driven intense pulsed neutron source using Trident

Energy Technology Data Exchange (ETDEWEB)

Vogel, Sven C. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

2016-08-25

A team of Los Alamos researchers supported a final campaign to use the Trident laser to produce neutrons, contributed their multidisciplinary expertise to experimentally assess if laser-driven neutron sources can be useful for MaRIE. MaRIE is the Laboratory’s proposed experimental facility for the study of matter-radiation interactions in extremes. Neutrons provide a radiographic probe that is complementary to x-rays and protons, and can address imaging challenges not amenable to those beams. The team's efforts characterize the Laboratory’s responsiveness, flexibility, and ability to apply diverse expertise where needed to perform successful complex experiments.

Fusion neutron detector calibration using a table-top laser generated plasma neutron source

International Nuclear Information System (INIS)

Hartke, R.; Symes, D.R.; Buersgens, F.; Ruggles, L.E.; Porter, J.L.; Ditmire, T.

2005-01-01

Using a high intensity, femtosecond laser driven neutron source, a high-sensitivity neutron detector was calibrated. This detector is designed for observing fusion neutrons at the Z accelerator in Sandia National Laboratories. Nuclear fusion from laser driven deuterium cluster explosions was used to generate a clean source of nearly monoenergetic 2.45 MeV neutrons at a well-defined time. This source can run at 10 Hz and was used to build up a clean pulse-height spectrum on scintillating neutron detectors giving a very accurate calibration for neutron yields at 2.45 MeV

Energy and angular distributions of neutrons from 252Cf spontaneous fission

International Nuclear Information System (INIS)

Vasil'ev, Yu.A.; Sidorov, L.V.; Vasil'eva, N.K.

1982-01-01

Some results from a first series of measurements of energy and angular distributions of neutrons from 252 Cf spontaneous fission using a spectrometer with high neutron detection efficiency, i.e. a 4π neutron time-of-flight spectrometer, were already presented. Subsequently, a second series of measurements was performed using a more sophisticated technique. For this second series, we used a more intense 252 Cf layer (25,000 spontaneous fissions per second). The angular resolution was improved by a factor of 2-3 by combining the hexahedral counter modules, placed at the same angle with respect to the direction of motion of the fragments, in new panoramic counters. The neutron counters were calibrated against the average 252 Cf neutron spectrum at several positions of the axis of the fragment detector with respect to the neutron counters. In the spectrum measurements and calibration work, the scattered neutron background was not determined theoretically, as in the first series of measurements, but experimentally using four extra scintillation counters with scatter cones; the counters were set up at 60 deg., 80 deg., 100 deg., and 120 deg. to the direction of separation of the fragments

Accelerator-driven neutron sources for materials research

International Nuclear Information System (INIS)

Jameson, R.A.

1990-01-01

Particle accelerators are important tools for materials research and production. Advances in high-intensity linear accelerator technology make it possible to consider enhanced neutron sources for fusion material studies or as a source of spallation neutrons. Energy variability, uniformity of target dose distribution, target bombardment from multiple directions, time-scheduled dose patterns, and other features can be provided, opening new experimental opportunities. New designs have also been used to ensure hands-on maintenance on the accelerator in these factory-type facilities. Designs suitable for proposals such as the Japanese Energy-Selective Intense Neutron Source, and the international Fusion Materials Irradiation Facility are discussed

The D-D Neutron Generator as an Alternative to Am(Li) Isotopic Neutron Source in the Active Well Coincidence Counter

Energy Technology Data Exchange (ETDEWEB)

McElroy, Robert Dennis [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Cleveland, Steven L. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2017-03-01

The ²³⁵U mass assay of bulk uranium items, such as oxide canisters, fuel pellets, and fuel assemblies, is not achievable by traditional gamma-ray assay techniques due to the limited penetration of the item by the characteristic ²³⁵U gamma rays. Instead, fast neutron interrogation methods such as active neutron coincidence counting must be used. For international safeguards applications, the most commonly used active neutron systems, the Active Well Coincidence Counter (AWCC), Uranium Neutron Collar (UNCL) and ²⁵²Cf Shuffler, rely on fast neutron interrogation using an isotopic neutron source [i.e., ²⁵²Cf or Am(Li)] to achieve better measurement accuracies than are possible using gamma-ray techniques for high-mass, high-density items. However, the Am(Li) sources required for the AWCC and UNCL systems are no longer manufactured, and newly produced systems rely on limited supplies of sources salvaged from disused instruments. The ²⁵²Cf shuffler systems rely on the use of high-output ²⁵²Cf sources, which while still available have become extremely costly for use in routine operations and require replacement every five to seven years. Lack of a suitable alternative neutron interrogation source would leave a potentially significant gap in the safeguarding of uranium processing facilities. In this work, we made use of Oak Ridge National Laboratory’s (ORNL’s) Large Volume Active Well Coincidence Counter (LV-AWCC) and a commercially available deuterium-deuterium (D-D) neutron generator to examine the potential of the D-D neutron generator as an alternative to the isotopic sources. We present the performance of the LV-AWCC with D-D generator for the assay of ²³⁵U based on the results of Monte Carlo N-Particle (MCNP) simulations and measurements of depleted uranium (DU), low enriched uranium (LEU), and highly enriched uranium (HEU) items.

Neutron sources and its dosimetric characteristics

International Nuclear Information System (INIS)

Vega C, H.R.; Manzanares A, E.; Hernandez D, V.M.; Mercado S, G.A.; Gallego D, E.; Lorente F, A.

2005-01-01

By means of Monte Carlo methods the spectra of the produced neutrons 252 Cf, 252 Cf/D 2 O, 241 Am Be, 239 Pu Be, 140 La Be, 239 Pu 18 O 2 and 226 Ra Be have been calculated. With the information of the spectrum it was calculated the average energy of the neutrons of each source. By means of the fluence coefficients to dose it was determined, for each one of the studied sources, the fluence factors to dose. The calculated doses were H, H * (10), H p,sIab (10, 0 0 ), E AP and E ISO . During the phase of the calculations the sources were modeled as punctual and their characteristics were determined to 100 cm in the hole. Also, for the case of the sources of 239 Pu Be and 241 Am Be, were carried out calculations modeling the sources with their respective characteristics and the dosimetric properties were determined in a space full with air. The results of this last phase of the calculations were compared with the experimental results obtained for both sources. (Author)

Neutron Sources for Standard-Based Testing

Energy Technology Data Exchange (ETDEWEB)

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

2014-11-10

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

Neutron fluctuations in accelerator driven and power reactors via backward master equations

International Nuclear Information System (INIS)

Zhifeng Kuang

2000-05-01

The transport of neutrons in a reactor is a random process, and thus the number of neutrons in a reactor is a random variable. Fluctuations in the number of neutrons in a reactor can be divided into two categories, namely zero noise and power reactor noise. As the name indicates, they dominate (i.e. are observable) at different power levels. The reasons for their occurrences and utilization are also different. In addition, they are described via different mathematical tools, namely master equations and the Langevin equation, respectively. Zero noise carries information about some nuclear properties such as reactor reactivity. Hence methods such as Feynman- and Rossi-alpha methods have been established to determine the subcritical reactivity of a subcritical system. Such methods received a renewed interest recently with the advent of the so-called accelerator driven systems (ADS). Such systems, intended to be used either for energy production or transuranium transmutation, will use a subcritical core with a strong spallation source. A spallation source has statistical properties that are different from those of the traditionally used radioactive sources which were also assumed in the derivation of the Feynman- and Rossi-alpha formulae. Therefore it is necessary to re-derive the Feynman- and Rossi-alpha formulae. Such formulae for ADS have been derived recently but in simpler neutronic models. One subject of this thesis is the extension of such formulae to a more general case in which six groups of delayed neutron precursors are taken into account, and the full joint statistics of the prompt and all delayed groups is included. The involved complexity problems are solved with a combination of effective analytical techniques and symbolic algebra codes. Power reactor noise carries information about parametric perturbation of the system. Langevin technique has been used to extract such information. In such a treatment, zero noise has been neglected. This is a pragmatic

Spectrum-averaged cross-section measurement of /sup 103/Rh(n,n)/sup 103m/Rh in the /sup 252/Cf fission neutron spectrum

International Nuclear Information System (INIS)

Lamaze, G.P.; Schima, F.J.; Eisenhauer, C.M.; Spiegel, V.

1988-01-01

Because of the similarity in energy dependence of the /sup 103/Rh(n,n') differential cross section to the kerma muscle response function for neutrons, rhodium may be useful as a neutron kerma monitor. In support of its use as a neutron monitor, the spectrum-averaged cross section σ-bar has been measured for a /sup 252/Cf fission neutron spectrum. Pairs of thin rhodium samples were irradiated on opposite sides of a thinly encapsulated /sup 252/Cf neutron source. The neutron emission rate of the /sup 252/Cf source was determined by the manganous sulfate (MnSO/sub 4/) bath technique. In this method, the californium source emission rate is determined by comparison to the known emission rate of NBS-I, a standard radium-beryllium neutron source. The neutron fluence incident on the rhodium samples is determined from the californium source strength, average sample-to-source distance, and the duration of the irradiation. Corrections are made for neutron scattering saturation of activity, and attenuation of the X rays by the sample during counting. The X rays were detected with an intrinsic germanium detector designed specifically for low-energy X-ray detection. The activity was not determined by absolute counting so that the final results depend on the value of P/sub Κx/, to total Κ X-ray emission probability. The results of five separate irradiations yield a value of σ-bar . P/sub Κx/ = 62.3 +- 1.9 mb. Using the most recently published value of P/sub Κx/ gives a value of σ-bar = 739 +- 22 mb. A discussion of systematic uncertainties is given

Thermal neutron imaging with rare-earth-ion-doped LiCaAlF{sub 6} scintillators and a sealed {sup 252}Cf source

Energy Technology Data Exchange (ETDEWEB)

Kawaguchi, Noriaki, E-mail: famicom@mail.tagen.tohoku.ac.jp [Tokuyama Corporation, Shibuya 3-chome, Shibuya-ku, Tokyo 150-8383 (Japan); IMRAM, Tohoku University, 2-1-1 Katahira, Aoba-ku, Sendai, Miyagi 980-8577 (Japan); Yanagida, Takayuki; Fujimoto, Yutaka; Yokota, Yuui; Kamada, Kei [IMRAM, Tohoku University, 2-1-1 Katahira, Aoba-ku, Sendai, Miyagi 980-8577 (Japan); Fukuda, Kentaro; Suyama, Toshihisa [Tokuyama Corporation, Shibuya 3-chome, Shibuya-ku, Tokyo 150-8383 (Japan); Watanabe, Kenichi; Yamazaki, Atsushi [Nagoya University, Furo-cho, Chikusa-ku, Nagoya 464-8603 (Japan); Chani, Valery [IMRAM, Tohoku University, 2-1-1 Katahira, Aoba-ku, Sendai, Miyagi 980-8577 (Japan); Yoshikawa, Akira [IMRAM, Tohoku University, 2-1-1 Katahira, Aoba-ku, Sendai, Miyagi 980-8577 (Japan); New Industry Creation Hatchery Center, Tohoku University, Sendai 980-8579 (Japan)

2011-10-01

Thermal neutron imaging with Ce-doped LiCaAlF{sub 6} crystals has been performed. The prototype of the neutron imager using a Ce-doped LiCaAlF{sub 6} scintillating crystal and a position sensitive photomultiplier tube (PSPMT) which had 64 multi-channel anode was developed. The Ce-doped LiCaAlF{sub 6} single crystal was grown by the Czochralski method. A plate with dimensions of a diameter of 50x2 mm{sup 2} was cut from the grown crystal, polished, and optically coupled to PSPMT by silicone grease. The {sup 252}Cf source (<1 MBq) was sealed with 43 mm of polyethylene for neutron thermalization. Alphabet-shaped Cd pieces with a thickness of 2 mm were used as a mask for the thermal neutrons. After corrections for the pedestals and gain of each pixel, we successfully obtained two-dimensional neutron images using Ce-doped LiCaAlF{sub 6}.

Source characterization of Purnima Neutron Generator (PNG)

International Nuclear Information System (INIS)

Bishnoi, Saroj; Patel, T.; Paul, Ram K.; Sarkar, P.S.; Adhikari, P.S.; Sinha, Amar

2011-01-01

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

Dosimetric analysis of BNCT - Boron Neutron Capture Therapy - coupled to 252Cf brachytherapy

International Nuclear Information System (INIS)

Brandao, Samia F.; Campos, Tarcisio P.R.

2009-01-01

The incidence of brain tumors is increasing in world population; however, the treatments employed in this type of tumor have a high rate of failure and in some cases have been considered palliative, depending on histology and staging of tumor. Its necessary to achieve the control tumor dose without the spread irradiation cause damage in the brain, affecting patient neurological function. Stereotactic radiosurgery is a technique that achieves this; nevertheless, other techniques that can be used on the brain tumor control must be developed, in order to guarantee lower dose on health surroundings tissues other techniques must be developing. The 252 Cf brachytherapy applied to brain tumors has already been suggested, showing promising results in comparison to photon source, since the active source is placed into the tumor, providing greater dose deposition, while more distant regions are spared. BNCT - Boron Neutron Capture Therapy - is another technique that is in developing to brain tumors control, showing theoretical superiority on the rules of conventional treatments, due to a selective irradiation of neoplasics cells, after the patient receives a borate compound infusion and be subjected to a epithermal neutrons beam. This work presents dosimetric studies of the coupling techniques: BNCT with 252 Cf brachytherapy, conducted through computer simulation in MCNP5 code, using a precise and well discretized voxel model of human head, which was incorporated a representative Glioblastoma Multiform tumor. The dosimetric results from MCNP5 code were exported to SISCODES program, which generated isodose curves representing absorbed dose rate in the brain. Isodose curves, neutron fluency, and dose components from BNCT and 252 Cf brachytherapy are presented in this paper. (author)

Analysis of geological material and especially ores by means of a 252Cf source

International Nuclear Information System (INIS)

Barrandon, J.N.; Borderie, B.; Melky, S.; Halfon, J.; Marce, A.

1976-01-01

Tests were made on the possibilities for analysis by 252 Cf activation in the earth sciences and mining research. The results obtained show that while 252 Cf activation can only resolve certain very specific geochemical research problems, it does allow the exact and rapid determination of numerous elements whose ores are of great economic importance such as fluorine, titanium, vanadium, manganese, copper, antimony, barium, and tungsten. The utilization of activation analysis methods in the earth sciences is not a recent phenomenon. It has generally been limited to the analysis of traces in relatively small volumes by means of irradiation in nuclear reactors. Traditional neutron sources were little used and were not very applicable. The development of 252 Cf isotopic sources emitting more intense neutron fluxes make it possible to consider carrying out more sensitive determinations without making use of a nuclear reactor. In addition, this technique can be adapted for in situ analysis in mines and mine borings. Our work which is centered upon the possibilities of instrumental laboratory analyses of geological materials through 252 Cf activation is oriented in two principal directions: the study of the experimental sensitivities of the various elements in different rocks with the usual compositions; and the study of the possibilities for routine ore analyses

Heuristic derivation of the Rossi-alpha formula for a pulsed neutron source

International Nuclear Information System (INIS)

Baeten, P.

2004-01-01

Expressions for the Rossi-alpha distribution for a pulsed neutron source were derived using a heuristic derivation based on the method of joint detection probability. This heuristic technique was chosen over the more rigorous master equation method due to its simplicity and the complementary of both techniques. The derived equations also take into account the presence of delayed neutrons and intrinsic neutron sources which often cannot be neglected in source-driven subcritical cores. The obtained expressions showed that the ratio of the correlated to the uncorrelated signal in the Rossi-Alpha distribution for a Pulsed Source (RAPS) was strongly increased compared to the case for a standard Rossi-alpha distribution for a continuous source. It was also demonstrated that by using this RAPS technique four independent measurement quantities, instead of three with the standard Rossi-alpha technique, can be determined. Hence, it is no longer necessary to combine the Rossi-alpha technique with another method to measure the reactivity expressed in dollars. Both properties, the increased signal-to-noise ratio of the correlated signal and the measurement of a fourth measurement quantity, make that the RAPS technique is an excellent candidate for the measurement of kinetic parameters in source-driven subcritical assemblies

Neutron-neutron angular correlations in spontaneous fission of 252Cf and 240Pu

Science.gov (United States)

Verbeke, J. M.; Nakae, L. F.; Vogt, R.

2018-04-01

Background: Angular anisotropy has been observed between prompt neutrons emitted during the fission process. Such an anisotropy arises because the emitted neutrons are boosted along the direction of the parent fragment. Purpose: To measure the neutron-neutron angular correlations from the spontaneous fission of 252Cf and 240Pu oxide samples using a liquid scintillator array capable of pulse-shape discrimination. To compare these correlations to simulations combining the Monte Carlo radiation transport code MCNPX with the fission event generator FREYA. Method: Two different analysis methods were used to study the neutron-neutron correlations with varying energy thresholds. The first is based on setting a light output threshold while the second imposes a time-of-flight cutoff. The second method has the advantage of being truly detector independent. Results: The neutron-neutron correlation modeled by FREYA depends strongly on the sharing of the excitation energy between the two fragments. The measured asymmetry enabled us to adjust the FREYA parameter x in 240Pu, which controls the energy partition between the fragments and is so far inaccessible in other measurements. The 240Pu data in this analysis was the first available to quantify the energy partition for this isotope. The agreement between data and simulation is overall very good for 252Cf(sf ) and 240Pu(sf ) . Conclusions: The asymmetry in the measured neutron-neutron angular distributions can be predicted by FREYA. The shape of the correlation function depends on how the excitation energy is partitioned between the two fission fragments. Experimental data suggest that the lighter fragment is disproportionately excited.

Evaluation of neutron nuclear data for 249Cf

International Nuclear Information System (INIS)

Zhou Delin; Yu Baosheng; Yuan Hanrong; Liu Tong; Zhang Jin; Su Zongdi; Yan Shiwei; Wang Cuilan; Zhang Jingshang

1992-01-01

A complete set of neutron nuclear data from 10 -5 eV to 20 MeV in ENDF/B-6 format has been evaluated for 249 Cf based on measured data, systematics predications and model theory calculations. The evaluated quantities are the total, elastic and inelastic scattering, fission, capture, (n,2n), (n,3n) reaction cross sections, the resolved and unresolved resonance parameters, the angular and energy distributions of emitted neutrons and the average number of neutrons emitted per fission. The numerical data are available in ENDF/B-6 format

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

Biomedical neutron research at the Californium User Facility for neutron science

International Nuclear Information System (INIS)

Martin, R.C.; Byrne, T.E.; Miller, L.F.

1997-01-01

The Californium User Facility for Neutron Science has been established at Oak Ridge National Laboratory (ORNL). The Californium User Facility (CUF) is a part of the larger Californium Facility, which fabricates and stores compact 252 Cf neutron sources for worldwide distribution. The CUF can provide a cost-effective option for research with 252 Cf sources. Three projects at the CUF that demonstrate the versatility of 252 Cf for biological and biomedical neutron-based research are described: future establishment of a 252 Cf-based neutron activation analysis system, ongoing work to produce miniature high-intensity, remotely afterloaded 252 Cf sources for tumor therapy, and a recent experiment that irradiated living human lung cancer cells impregnated with experimental boron compounds to test their effectiveness for boron neutron capture therapy

Accelerator driven neutron sources in Korea. Current and future

International Nuclear Information System (INIS)

Lee, Young-Ouk; Oh, Byung-Hoon; Hong, Bong-Geun; Chang, Jonghwa; Chang, Moon-Hee; Kim, Guinyun; Kim, Gi-Donng; Choi, Byung-Ho

2008-01-01

The Pohang Neutron Facility, based on a 65 MeV electron linear accelerator, has a neutron-gamma separation circuit, water-moderated tantalum target and 12 m TOF. It produces pulsed photonuclear neutrons with ≅2 μs width, 50 mA peak current and 15 Hz repetition, mainly for the neutron nuclear data production in up to keV energies. The Tandem Van de Graff at Korea Institute of Geoscience and Mineral Resources (KIGAM) is dedicated to measure MeV energy neutron capture and total cross section using TOF and prompt gamma ray detection system. The facility pulsed ≅10 8 mono-energetic neutrons/sec from 3 H(p,n) reaction with 1-2 ns width and 125 ns period. Korea Institute of Radiological and Medical Sciences (KIRAMS) has the MC50 medical cyclotron which accelerates protons up to an energy of 45 MeV and has several beam ports for proton or neutron irradiations. Beam current can be controlled from a few nano amperes to 50 uA. Korea Atomic Energy Research Institute (KAERI) has a plan to develop a neutron source by using 20 MeV electron accelerator. This photo-neutron source will be mainly used for nuclear data measurements based on time-of-flight experiments. A high intensity fast neutron source is also proposed to respond growing demands of fast neutrons, especially for the fusion material test. Throughput will be as high as several 10 13 neutrons/sec from D-T reaction powered by a high current (200 mA) ion source, a drive-in target and cooling systems, and closed circuit tritium ventilation/recovery systems. The Proton Engineering Frontier Project (PEFP) is developing a 100 MeV, 20 mA pulsed proton linear accelerator equipped with 5 target rooms, one of which is dedicated to produce neutrons using tungsten target. PEFP also proposes the 1-2 GeV rapid cycling synchrotron accelerator as an extension of the PEFP linac, which can be used for nuclear and high energy physics experiment, spallation neutron source, radioisotope, medical research, etc. (author)

Evaluation of neutron sources for ISAGE-in-situ-NAA for a future lunar mission

International Nuclear Information System (INIS)

Li, X.; Breitkreutz, H.; Burfeindt, J.; Bernhardt, H.-G.; Trieloff, M.; Hopp, J.; Jessberger, E.K.; Schwarz, W.H.; Hofmann, P.; Hiesinger, H.

2011-01-01

For a future Moon landing, a concept for an in-situ NAA involving age determination using the 40 Ar- 39 Ar method is developed. A neutron source 252 Cf is chosen for sample irradiation on the Moon. A special sample-in-source irradiation geometry is designed to provide a homogeneous distribution of neutron flux at the irradiation position. Using reflector, the neutron flux is likely to increase by almost 200%. Sample age of 1 Ga could be determined. Elemental analysis using INAA is discussed. - Highlights: → We developed a concept for an in-situ age determination on the Moon. → 252 Cf is chosen as the neutron source for the 40 Ar- 39 Ar-method. → A sample-in-source geometry is designed to provide homogeneous neutron flux. → Determination of U, Th, K and Ir using NAA on the Moon is possible.

Reactor physics experiments in PURNIMA sub critical facility coupled with 14 MeV neutron source

International Nuclear Information System (INIS)

Kumar, Rajeev; Degweker, S.B.; Patel, Tarun; Bishnoi, Saroj; Adhikari, P.S.

2011-01-01

Accelerator Driven Sub-critical Systems (ADSS) are attracting increasing worldwide attention due to their superior safety characteristics and their potential for burning actinide and fission product waste and energy production. A number of countries around the world have drawn up roadmaps/programs for development of ADSS. Indian interest in ADSS has an additional dimension, which is related to the planned utilization of our large thorium reserves for future nuclear energy generation. A programme for development of ADSS is taken up at the Bhabha Atomic Research Centre (BARC) in India. This includes R and D activities for high current proton accelerator development, target development and Reactor Physics studies. As part of the ADSS Reactor Physics research programme, a sub-critical facility is coming up in BARC which will be coupled with an existing D-D/D-T neutron generator. Two types of cores are planned. In one of these, the sub-critical reactor assembly consists of natural uranium moderated by high density polyethylene (HDP) and reflected by BeO. The other consists of natural uranium moderated by light water. The maximum neutron yield of the neutron source with tritium target is around 10 10 neutron per sec. Various reactor physics experiments like measurement of the source strength, neutron flux distribution, buckling estimation and sub-critical source multiplication are planned. Apart from this, measurement of the total fission power and neutron spectrum will also be carried out. Mainly activation detectors will be used in all in-core neutron flux measurement. Measurement of the degree of sub-criticality by various deterministic and noise methods is planned. Helium detectors with advanced data acquisition card will be used for the neutron noise experiments. Noise characteristics of ADSS are expected to be different from that of traditional reactors due to the non-Poisson statistical features of the source. A new theory incorporating these features has been

Yields of fission products produced by thermal-neutron fission of 249Cf

International Nuclear Information System (INIS)

Dickens, J.K.; McConnell, J.W.

1981-01-01

Absolute yields have been determined for 107 gamma rays emitted in the decay of 97 fission products representing 54 mass chains created during thermal-neutron fission of 249 Cf. These results include 14 mass chains for which no prior yield data exist. Using a Ge(Li) detector, spectra were obtained of gamma rays emanating from a 0.4 μg sample of 249 Cf between 45 s and 0.4 yr after very short irradiations of the 249 Cf by thermal neutrons. On the basis of measured gamma-ray yields and known nuclear data, total chain mass yields and relative uncertainties were obtained for 51 masses between 89 and 156. The absolute overall normalization uncertainty is approx.8%. The measured A-chain cumulative yields make up 77% of the total light mass (A 249 Cf

Localization of Vibrating Noise Sources in Nuclear Reactor Cores

International Nuclear Information System (INIS)

Hultqvist, Pontus

2004-09-01

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

Design of a portable directional neutron source finder

International Nuclear Information System (INIS)

Yamanishi, Hirokuni

2005-01-01

An instrument that determines the direction of a remote existing neutron source has been designed. This instrument combines a polyethylene block and four 3 He counter tubes. The advantages of the instrument are portability and good angular resolution. The count from the detector was varied with the neutron incident angle due to the moderator. Using this characteristic, the direction of the neutron source can be measured precisely by revising the axis of the instrument so that the difference between the four detectors measurements is minimized. Consequently, the direction of the central axis of the instrument in which the response difference of the four detectors reaches a minimum indicates the direction of the neutron source. The practical use of the instrument was demonstrated by 252 Cf source irradiation experiment and MCNP simulation

Characterization of neutron spectra using sources of {sup 241}AmBe, {sup 238}PuBe e {sup 252}Cf moderated in water; Caracterização de espectros neutrônicos com fontes de {sup 241}AmBe, {sup 238}PuBe e {sup 252}Cf moderados em água

Energy Technology Data Exchange (ETDEWEB)

Gonçalves, A.S.; Silva, F.S.; Patrão, K.C.S.; Fonseca, E.S. da; Pereira, W.W., E-mail: angela.souzagon@gmail.com [Instituto de Radioproteção e Dosimetria (IRD/CNEN-RJ), Rio de Janeiro, RJ (Brazil). Laboratório de Metrologia de Nêutrons

2017-07-01

Recent studies have demonstrated the importance of characterizing the spectrum of neutron sources for various energies in order to make the understanding of neutron interaction closer to reality they work with. This work presents the determination of neutron energy flux from the source of {sup 241}AmBe (0.6 TBq), {sup 238}PuBe (1.8 TBq) and {sup 252}Cf (120 μg) free in the air and inserted into spheres of various diameters containing distilled water. The determination of the spectrum is based on the measurement and simulation by the Monte Carlo computational method, using the sources under study, with the Bonner multisphere spectrometer containing readings with the detector without sphere and covered with polyethylene balls of diameters: 5,08 cm (2 ″), 7.62 cm (3″), 12.70 cm (5 ″), 20.32 cm (8 ″), 25.40 cm (10 ″) and 30.48 cm (12 ″). It is sought to characterize a new moderate spectrum in water using the sources of {sup 241}AmBe, {sup 238}PuBe and {sup 252}Cf that may be useful for testing, calibration and irradiation of individual and area monitors for neutrons.

Regularities in dose field formation for fixation of 252Cf and 60Co sources on the same plane

International Nuclear Information System (INIS)

Ivanov, V.N.; Vtyurin, B.M.; Ivanova, L.F.; Kondzhariya, Yu.R.

1983-01-01

A computerized study was made of a change in the correlation between the sizes of an irradiated volume and the distribution of sources. Criteria have been established for the utilization of 252 Cf and 60 Co sources of different active length and design placed on the same plane. Reference dose rate and its derivatives have been chosen as the main parameters that characterize a dose field. A dosimetric analysis was performed using the data for an adsorbed dose of neutrons in the tissue, taking account of high RBE of 252 Cf fast neutrons that equals 6-7, and a slight change of the local RBE value near the source resulting from a change of the effective spectrum of neutrons in combination with the growth of a contribution of #betta#-radiation into the summary dose with the removal from the source. The irradiated volume was evaluated by introducing 3 linear parameters - length, thickness and width. It has been shown that the ratio of the length of an irradiated volume to the active length of 252 Cf sources with radionuclide regular linear density changes from 0.7 with the isodose value of 85% up to 0.97 with the isodose value of 60%. For 60 Co sources with a higher linear density of the radionuclide on the edges the lower limit of this value equals 1. Intervals of changes in the rest of parameters have also been defined. The results obtained are presented graphically. The results of the study are used for a dosimetric control of the clinical trials of 252 Cf and 60 Co sources in the treatment of patients with tumors of the tongue, oral cavity fundus, lip and other sites

Biomedical neutron research at the Californium User Facility for Neutron Science

International Nuclear Information System (INIS)

Martin, R.C.; Byrne, T.E.; Miller, L.F.

1998-01-01

The Californium User Facility for Neutron Science has been established at Oak Ridge National Laboratory (ORNL). The Californium User Facility (CUF) is a part of the larger Californium Facility, which fabricates and stores compact 252 Cf neutron sources for worldwide distribution. The CUF can provide a cost-effective option for research with 252 Cf sources. Three projects at the CUF that demonstrate the versatility of 252 Cf for biological and biomedical neutron-based research are described: future establishment of a 252 Cf-based neutron activation analysis system, ongoing work to produce miniature high-intensity, remotely afterloaded 252 Cf sources for tumor therapy, and a recent experiment that irradiated living human lung cancer cells impregnated with experimental boron compounds to test their effectiveness for boron neutron capture therapy. (author)

A Detector for 2-D Neutron Imaging for the Spallation Neutron Source

International Nuclear Information System (INIS)

Britton, Charles L. Jr.; Bryan, W.L.; Wintenberg, Alan Lee; Clonts, Lloyd G.; Warmack, Robert J. Bruce; McKnight, Timothy E.; Frank, Steven Shane; Cooper, Ronald G.; Dudney, Nancy J.; Veith, Gabriel M.

2006-01-01

We have designed, built, and tested a 2-D pixellated thermal neutron detector. The detector is modeled after the MicroMegas-type structure previously published for collider-type experiments. The detector consists of a 4X4 square array of 1 cm 2 pixels each of which is connected to an individual preamplifier-shaper-data acquisition system. The neutron converter is a 10B film on an aluminum substrate. We describe the construction of the detector and the test results utilizing 252Cf sources in Lucite to thermalize the neutrons. Drift electrode (Aluminum) Converter (10B) 3 mm Conversion gap neutron (-900 V)

Evaluation of neutron sources for ISAGE-in-situ-NAA for a future lunar mission

Energy Technology Data Exchange (ETDEWEB)

Li, X., E-mail: xiaosong.li@frm2.tum.de [Forschungs-Neutronenquelle Heinz Maier-Leibnitz, FRM II, Technische Universitaet Muenchen, Lichtenbergstr. 1, D-85747 Garching (Germany); Breitkreutz, H. [Forschungs-Neutronenquelle Heinz Maier-Leibnitz, FRM II, Technische Universitaet Muenchen, Lichtenbergstr. 1, D-85747 Garching (Germany); Burfeindt, J.; Bernhardt, H.-G. [Kayser-Threde GmbH, Wolfratshauser Str. 48, D-81379 Muenchen (Germany); Trieloff, M.; Hopp, J. [Institut fuer Geowissenschaften, Ruprecht-Karls-Universitaet Heidelberg, D-69120 Heidelberg (Germany); Jessberger, E.K. [Institut fuer Planetologie, Westfaelische Wilhelms-Universitaet Muenster, D-48149 Muenster (Germany); Schwarz, W.H. [Institut fuer Geowissenschaften, Ruprecht-Karls-Universitaet Heidelberg, D-69120 Heidelberg (Germany); Hofmann, P. [Kayser-Threde GmbH, Wolfratshauser Str. 48, D-81379 Muenchen (Germany); Hiesinger, H. [Institut fuer Planetologie, Westfaelische Wilhelms-Universitaet Muenster, D-48149 Muenster (Germany)

2011-11-15

For a future Moon landing, a concept for an in-situ NAA involving age determination using the {sup 40}Ar-{sup 39}Ar method is developed. A neutron source {sup 252}Cf is chosen for sample irradiation on the Moon. A special sample-in-source irradiation geometry is designed to provide a homogeneous distribution of neutron flux at the irradiation position. Using reflector, the neutron flux is likely to increase by almost 200%. Sample age of 1 Ga could be determined. Elemental analysis using INAA is discussed. - Highlights: > We developed a concept for an in-situ age determination on the Moon. > {sup 252}Cf is chosen as the neutron source for the {sup 40}Ar-{sup 39}Ar-method. > A sample-in-source geometry is designed to provide homogeneous neutron flux. > Determination of U, Th, K and Ir using NAA on the Moon is possible.

The interpretation of neutron noise in boiling water reactors

International Nuclear Information System (INIS)

John, T.M.; Singh, O.P.

1985-01-01

Some qualitative results of neutron noise in a boiling water reactor (BWR) are reported. By using one-group theory, it has been shown that the neutron flux fluctuations caused by a distributed source in space, representative of the coolant boiling noise in BWRs, can be considered as made up of two components: The first one, having a global character, is a quickly varying function of frequency and follows the fundamental mode solution in space; the second, called nonglobal (local), follows the spatial variation of noise-source intensity distribution and is independent of frequency for ω γΣ, this component decreases with increasing frequency. The formulation indicates that the global component is quite sensitive to the neutron multiplication factor of the system and, for the local component, the medium behaves like a nonmultiplying one. The global effect is dominant at lower frequencies in a critical system, and the local effect is dominant at higher fre quencies

Alternative definitions of kinetic parameters for accelerator driven systems

International Nuclear Information System (INIS)

Talamo, Alberto; Gohar, Yousry; Dulla, Sandra; Ravetto, Piero

2012-01-01

Highlights: ► New definition of kinetic parameters for accelerator driven systems. ► Difference between effective and average delayed neutron fraction. ► Difference between effective and average prompt neutron lifetime. ► Effect of the neutron source (Cf, D–D, D–T) on k src . ► Effect of the (n, xn) reactions and source energy-angle distribution on k src . - Abstract: This study introduces a new formulation of kinetic parameters for accelerator driven systems and it is structured into two parts. The first part is dedicated to the classic definition of the kinetic parameters and compares different calculation methodologies. The second part considers a new definition of the kinetic parameters for subcritical assemblies, with particular emphasis on the delayed neutron fraction and the prompt neutron lifetime. This new definition takes into account neutrons from the external neutron source and (n, xn) reactions, which increase the fraction of prompt neutrons. The developed theoretical framework has been applied by Monte Carlo and deterministic calculations to the YALINA Thermal subcritical assembly located in Belarus. This facility can be driven by californium, deuterium–deuterium (D–D), or deuterium–tritium (D–T) external neutron sources. For the D–T neutron source, (n, xn) reactions must be taken into account in order to produce accurate results because the average energy of D–T source neutrons is 14.1 MeV, a value which is much higher than the threshold energy of the (n, 2n) cross section of uranium isotopes.

Neutron sources and its dosimetric characteristics; Fuentes de neutrones y sus caracteristicas dosimetricas

Energy Technology Data Exchange (ETDEWEB)

Vega C, H.R.; Manzanares A, E.; Hernandez D, V.M.; Mercado S, G.A. [Universidad Autonoma de Zacatecas, A.P. 336, 98000 Zacatecas (Mexico); Gallego D, E.; Lorente F, A. [Universidad Politecnica de Madrid, C/Jose Gutierrez Abascal 2, E-28006 Madrid (Spain)

2005-07-01

By means of Monte Carlo methods the spectra of the produced neutrons {sup 252} Cf, {sup 252} Cf/D{sub 2}O, {sup 241} Am Be, {sup 239} Pu Be, {sup 140} La Be, {sup 239} Pu{sup 18}O{sub 2} and {sup 226} Ra Be have been calculated. With the information of the spectrum it was calculated the average energy of the neutrons of each source. By means of the fluence coefficients to dose it was determined, for each one of the studied sources, the fluence factors to dose. The calculated doses were H, H{sup *}(10), H{sub p,sIab} (10, 0{sup 0}), E{sub AP} and E{sub ISO}. During the phase of the calculations the sources were modeled as punctual and their characteristics were determined to 100 cm in the hole. Also, for the case of the sources of {sup 239} Pu Be and {sup 241} Am Be, were carried out calculations modeling the sources with their respective characteristics and the dosimetric properties were determined in a space full with air. The results of this last phase of the calculations were compared with the experimental results obtained for both sources. (Author)

Prospect for application of compact accelerator-based neutron source to neutron engineering diffraction

Energy Technology Data Exchange (ETDEWEB)

Ikeda, Yoshimasa, E-mail: yoshimasa.ikeda@riken.jp [Center for Advanced Photonics, RIKEN, Wako, Saitama 351-0198 (Japan); Taketani, Atsushi; Takamura, Masato; Sunaga, Hideyuki [Center for Advanced Photonics, RIKEN, Wako, Saitama 351-0198 (Japan); Kumagai, Masayoshi [Faculty of Engineering, Tokyo City University, Setagaya, Tokyo 158-8857 (Japan); Oba, Yojiro [Research Reactor Institute, Kyoto University, Kumatori, Osaka 590-0494 (Japan); Otake, Yoshie [Center for Advanced Photonics, RIKEN, Wako, Saitama 351-0198 (Japan); Suzuki, Hiroshi [Materials Sciences Research Center, Japan Atomic Energy Agency, Tokai, Ibaraki 319-1195 (Japan)

2016-10-11

A compact accelerator-based neutron source has been lately discussed on engineering applications such as transmission imaging and small angle scattering as well as reflectometry. However, nobody considers using it for neutron diffraction experiment because of its low neutron flux. In this study, therefore, the neutron diffraction experiments are carried out using Riken Accelerator-driven Compact Neutron Source (RANS), to clarify the capability of the compact neutron source for neutron engineering diffraction. The diffraction pattern from a ferritic steel was successfully measured by suitable arrangement of the optical system to reduce the background noise, and it was confirmed that the recognizable diffraction pattern can be measured by a large sampling volume with 10 mm in cubic for an acceptable measurement time, i.e. 10 min. The minimum resolution of the 110 reflection for RANS is approximately 2.5% at 8 μs of the proton pulse width, which is insufficient to perform the strain measurement by neutron diffraction. The moderation time width at the wavelength corresponding to the 110 reflection is estimated to be approximately 30 μs, which is the most dominant factor to determine the resolution. Therefore, refinements of the moderator system to decrease the moderation time by decreasing a thickness of the moderator or by applying the decoupler system or application of the angular dispersive neutron diffraction technique are important to improve the resolution of the diffraction experiment using the compact neutron source. In contrast, the texture evolution due to plastic deformation was successfully observed by measuring a change in the diffraction peak intensity by RANS. Furthermore, the volume fraction of the austenitic phase in the dual phase mock specimen was also successfully evaluated by fitting the diffraction pattern using a Rietveld code. Consequently, RANS has been proved to be capable for neutron engineering diffraction aiming for the easy access

Simultaneous investigation of fission fragments and neutrons in 252Cf(s,f)

International Nuclear Information System (INIS)

Budtz-Joergensen, C.; Knitter, H.H.

1986-01-01

The gridded twin ion chamber developed at CBNM is used to measure the kinetic energy-, mass- and angular distributions of the fission fragments of 252 Cf in an advantageous 4π-geometry. Together with a neutron time-of-flight detector this experimental arrangement permits to measure the correlation between neutron emission, fragment angle, mass and energy in the spontaneous fission of 252 Cf. With the present experimental set-up a mass resolution for fission fragments of 0.5 a.m.u., an angular resolution of Δcosθ = 0.05 and a timing resolution of 0.7 ns FWHM were observed. Preliminary evaluations of the raw experimental data are presented for the fission fragment mass distribution, the average total kinetic energy and their variance as function of mass, the angular distribution between fragments and neutrons, the number of neutrons emitted per fragment as function of fragment mass, the average neutron emission energies as function of mass, and the prompt fission neutron spectrum averaged over all fragments. (author)

4π-spectrometer technique for measurements of secondary neutron average number in nuclear fission by 252Cf neutrons

International Nuclear Information System (INIS)

Vasil'ev, Yu.A.; Barashkov, Yu.A.; Golovanov, O.A.; Sidorov, L.V.

1977-01-01

A method for determining the average number of secondary neutrons anti ν produced in nuclear fission by the neutrons of the 252 Cf fission spectra by means of a 4π time-of-flight spectrometer is described. Layers of 252 Cf and an isotope studied are placed close to each other; if the isotope layer density is 1 mg/cm 2 probability of its fission is about 10 -5 per one spontaneous fission of californium. Fission fragments of 252 Cf and the isotope investigated have been detected by two surface-barrier counters with an efficiency close to 100%. The layers and the counters are situated in a measuring chamber placed in the center of the 4π time-of-flight spectrometer. The latter is utilized as a neutron counter because of its fast response. The method has been verified by carrying out measurements for 235 U and 239 Pu. A comparison of the experimental and calculated results shows that the method suggested can apply to determine the number of secondary neutrons in fission of isotopes that have not been investigated yet

Feasibility of fissile mass assay of spent nuclear fuel using 252Cf-source-driven frequency-analysis

International Nuclear Information System (INIS)

Mattingly, J.K.; Valentine, T.E.; Mihalczo, J.T.

1996-01-01

The feasibility was evaluated using MCNP-DSP, an analog Monte Carlo transport cod to simulate source-driven measurements. Models of an isolated Westinghouse 17x17 PWR fuel assembly in a 1500-ppM borated water storage pool were used. In the models, the fuel burnup profile was represented using seven axial burnup zones, each with isotopics estimated by the PDQ code. Four different fuel assemblies with average burnups from fresh to 32 GWd/MTU were modeled and analyzed. Analysis of the fuel assemblies was simulated by inducing fission in the fuel using a 252 Cf source adjacent to the assembly and correlating source fissions with the response of a bank of 3 He detectors adjacent to the assembly opposite the source. This analysis was performed at 7 different axial positions on each of the 4 assemblies, and the source-detector cross-spectrum signature was calculated for each of these 28 simulated measurements. The magnitude of the cross-spectrum signature follows a smooth upward trend with increasing fissile material ( 235 U and 239 Pu) content, and the signature is independent of the concentration of spontaneously fissioning isotopes (e.g., 244 Cm) and (α,n) sources. Furthermore, the cross-spectrum signature is highly sensitive to changes in fissile material content. This feasibility study indicated that the signature would increase ∼100% in response to an increase of only 0.1 g/cm 3 of fissile material

Active Interrogation of Sensitive Nuclear Material Using Laser Driven Neutron Beams

International Nuclear Information System (INIS)

Favalli, Andrea; Roth, Markus

2015-01-01

An investigation of the viability of a laser-driven neutron source for active interrogation is reported. The need is for a fast, movable, operationally safe neutron source which is energy tunable and has high-intensity, directional neutron production. Reasons for the choice of neutrons and lasers are set forth. Results from the interrogation of an enriched U sample are shown.

Research program in reactor core diagnostics with neutron noise methods: Stage 3. Final report

International Nuclear Information System (INIS)

Pazsit, I.; Garis, N.S.; Karlsson, J.; Racz, A.

1997-09-01

Stage 3 of the program has been executed 96-04-12. The long term goal is to develop noise methods for identification and localization of perturbations in reactor cores. The main parts of the program consist of modelling the noise source, calculation of the space- and frequency dependent transfer function, calculation of the neutron noise via a convolution of the transfer function of the system and the noise source, i.e. the perturbation, and finally finding an inversion or unfolding procedure to determine noise source parameters from the neutron noise. Most previous work is based on very simple (analytical) reactor models for the calculation of the transfer function as well as analytical unfolding methods. The purpose of this project is to calculate the transfer function in a more realistic model as well as elaborating powerful inversion methods that do not require analytical transfer functions. The work in stage 3 is described under the following headlines: Further investigation of simplified models for the calculation of the neutron noise; Further investigation of methods based on neural networks; Further investigation of methods for detecting the vibrations and impacting of detectors; Application of static codes for determination of the neutron noise using the adiabatic approximation

Source driven breeding thermal power reactors

International Nuclear Information System (INIS)

Greenspan, E.; Ben-Gurion Univ. of the Negev, Beersheba

1978-03-01

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

Instrumentation at pulsed neutron sources

International Nuclear Information System (INIS)

Carpenter, J.M.; Lander, G.H.; Windsor, C.G.

1984-01-01

Scientific investigations involving the use of neutron beams have been centered at reactor sources for the last 35 years. Recently, there has been considerable interest in using the neutrons produced by accelerator driven (pulsed) sources. Such installations are in operation in England, Japan, and the United States. In this article a brief survey is given of how the neutron beams are produced and how they can be optimized for neutron scattering experiments. A detailed description is then given of the various types of instruments that have been, or are planned, at pulsed sources. Numerous examples of the scientific results that are emerging are given. An attempt is made throughout the article to compare the scientific opportunities at pulsed sources with the proven performance of reactor installations, and some familiarity with the latter and the general field of neutron scattering is assumed. New areas are being opened up by pulsed sources, particularly with the intense epithermal neutron beams, which promise to be several orders of magnitude more intense than can be obtained from a thermal reactor

Spectrometers for compact neutron sources

Science.gov (United States)

Voigt, J.; Böhm, S.; Dabruck, J. P.; Rücker, U.; Gutberlet, T.; Brückel, T.

2018-03-01

We discuss the potential for neutron spectrometers at novel accelerator driven compact neutron sources. Such a High Brilliance Source (HBS) relies on low energy nuclear reactions, which enable cryogenic moderators in very close proximity to the target and neutron optics at comparably short distances from the moderator compared to existing sources. While the first effect aims at increasing the phase space density of a moderator, the second allows the extraction of a large phase space volume, which is typically requested for spectrometer applications. We find that competitive spectrometers can be realized if (a) the neutron production rate can be synchronized with the experiment repetition rate and (b) the emission characteristics of the moderator can be matched to the phase space requirements of the experiment. MCNP simulations for protons or deuterons on a Beryllium target with a suitable target/moderator design yield a source brightness, from which we calculate the sample fluxes by phase space considerations for different types of spectrometers. These match closely the figures of todays spectrometers at medium flux sources. Hence we conclude that compact neutron sources might be a viable option for next generation neutron sources.

Identification of new neutron-rich rare-earth nuclei produced in /sup 252/Cf spontaneous fission

CERN Document Server

Greenwood, R C; Gehrke, R J; Meikrantz, D H

1981-01-01

A program of systematic study of the decay properties of neutron-rich rare-earth nuclei with 30 sCf spontaneous fission, is currently underway using the Idaho ESOL (Elemental Separation On Line) Facility. The chemistry system used for the rare-earth elemental separations consists of two high-performance chromatography columns connected in series and coupled to the /sup 252 /Cf fission source via a helium gas-jet transport arrangement. The time delay for separation and initiation of gamma -ray counting with results which have been obtained to date with this system include the identification of a number of new neutron-rich rare-earth isotopes including /sup 155/Pm (t/sub 1/2/=48+or-4 s) and /sup 163/Gd (t/sub 1 /2/=68+or-3 s), in addition to 5.51 min /sup 158/Sm which was identified in an earlier series of experiments. (11 refs).

Prompt neutron energy spectrum for the spontaneous fission of Cf-252

International Nuclear Information System (INIS)

Blinov, M.V.; Boykov, G.S.; Vitenko, V.A.

1985-06-01

The prompt neutron spectrum for the spontaneous fission of Cf-252 has been measured in 0.01-10 MeV region by the time-of-flight technique using a fast ionization chamber with U-235 layers as the neutron detector. Numerical data for the spectrum are presented, with an error file. (author)

Compact D-D Neutron Source-Driven Subcritical Multiplier and Beam-Shaping Assembly for Boron Neutron Capture Therapy

International Nuclear Information System (INIS)

Ganda, Francesco; Vujic, Jasmina; Greenspan, Ehud; Leung, Ka-Ngo

2010-01-01

This work assesses the feasibility of using a small, safe, and inexpensive keff 0.98 subcritical fission assembly [subcritical neutron multiplier (SCM)] to amplify the treatment neutron beam intensity attainable from a compact deuterium-deuterium (D-D) fusion neutron source delivering [approximately]1012 n/s. The objective is to reduce the treatment time for deep-seated brain tumors to [approximately]1 h. The paper describes the optimal SCM design and two optimal beam-shaping assemblies (BSAs) - one designed to maximize the dose rate and the other designed to maximize the total dose that can be delivered to a deep-seated tumor. The neutron beam intensity amplification achieved with the optimized SCM and BSA results in an increase in the treatment dose rate by a factor of 18: from 0.56 Gy/h without the SCM to 10.1 Gy/h. The entire SCM is encased in an aluminum structure. The total amount of 20% enriched uranium required for the SCM is 8.5 kg, and the cost (not including fabrication) is estimated to be less than $60,000. The SCM power level is estimated at 400 W when driven by a 1012 n/s D-D neutron source. This translates into consumption of only [approximately]0.6% of the initially loaded 235U atoms during 50 years of continuous operation and implies that the SCM could operate continuously for the entire lifetime of the facility without refueling. Cooling the SCM does not pose a challenge; it may be accomplished by natural circulation as the maximum heat flux is only 0.034 W/cm2.

A multi-group neutron noise simulator for fast reactors

International Nuclear Information System (INIS)

Tran, Hoai Nam; Zylbersztejn, Florian; Demazière, Christophe; Jammes, Christian; Filliatre, Philippe

2013-01-01

Highlights: • The development of a neutron noise simulator for fast reactors. • The noise equation is solved fully in a frequency-domain. • A good agreement with ERANOS on the static calculations. • Noise calculations induced by a localized perturbation of absorption cross section. - Abstract: A neutron noise simulator has been developed for fast reactors based on diffusion theory with multi-energy groups and several groups of delayed neutron precursors. The tool is expected to be applicable for core monitoring of fast reactors and also for other reactor types with hexagonal fuel assemblies. The noise sources are modeled through small stationary fluctuations of macroscopic cross sections, and the induced first order noise is solved fully in the frequency domain. Numerical algorithms are implemented for solving both the static and noise equations using finite differences for spatial discretization, where a hexagonal assembly is radially divided into finer triangular meshes. A coarse mesh finite difference (CMFD) acceleration has been used for accelerating the convergence of both the static and noise calculations. Numerical calculations have been performed for the ESFR core with 33 energy groups and 8 groups of delayed neutron precursors using the cross section data generated by the ERANOS code. The results of the static state have been compared with those obtained using ERANOS. The results show an adequate agreement between the two calculations. Noise calculations for the ESFR core have also been performed and demonstrated with an assumption of the perturbation of the absorption cross section located at the central fuel ring

Dynamically Polarized Sample for Neutron Scattering At the Spallation Neutron Source

International Nuclear Information System (INIS)

Pierce, Josh; Zhao, J. K.; Crabb, Don

2009-01-01

The recently constructed Spallation Neutron Source at the Oak Ridge National Laboratory is quickly becoming the world's leader in neutron scattering sciences. In addition to the world's most intense pulsed neutron source, we are continuously constructing state of the art neutron scattering instruments as well as sample environments to address today and tomorrow's challenges in materials research. The Dynamically Polarized Sample project at the SNS is aimed at taking maximum advantage of polarized neutron scattering from polarized samples, especially biological samples that are abundant in hydrogen. Polarized neutron scattering will allow us drastically increase the signal to noise ratio in experiments such as neutron protein crystallography. The DPS project is near completion and all key components have been tested. Here we report the current status of the project.

Inspection of CF188 composite flight control surfaces with neutron radiography

International Nuclear Information System (INIS)

Lewis, W.J.; Bennett, L.G.I.; Mullin, S.K.

1996-01-01

At the Royal Military College of Canada's SLOWPOKE-2 Facility, a neutron radiography facility has been designed and installed using a small (20kWth), pool-type research reactor called the SLOWPOKE-2 (Safe Low Power c(K)ritical Experiment) as the neutron source. Since then, the research has continued along two fronts: developing applications and improving the quality of the neutron beam. The most interesting applications investigated to date has been the inspection of various metal ceramic composites and the inspection of the composite flight control surfaces of some of the CF188 Hornet aircraft. As part of the determination of the integrity of the aircraft, it was decided to inspect an aircraft with the highest flight house using both X- and neutron radiography. The neutron radiography and, to a lesser extent, X-radiography inspections completed at McClellan AFB revealed 93 anomalies. After returning to Canada, the component with the greatest structural significance, namely the right hand rudder from the vertical stabilizer, was removed from the aircraft and put through a rigorous program of numerous NDT inspections, including X-radiography (film and real-time), eddy current, ultrasonics (through transmission and pitch-catch), infrared thermography, and neutron radiography. Therefore, of all the techniques investigated, only through transmission ultrasonics and neutron radiography were able to identify large areas of hydration. However, only neutron radiography could identify the small areas of moisture and hydration. Given the structural significance of the flight control surfaces in modern fighter aircraft, even the smallest amounts of hydration could potentially lead to catastrophic results

Future opportunities with pulsed neutron sources

Energy Technology Data Exchange (ETDEWEB)

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

1996-05-01

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

The Los Alamos Neutron Science Center Spallation Neutron Sources

International Nuclear Information System (INIS)

Nowicki, Suzanne F.; Wender, Stephen A.; Mocko, Michael

2017-01-01

The Los Alamos Neutron Science Center (LANSCE) provides the scientific community with intense sources of neutrons, which can be used to perform experiments supporting civilian and national security research. These measurements include nuclear physics experiments for the defense program, basic science, and the radiation effect programs. This paper focuses on the radiation effects program, which involves mostly accelerated testing of semiconductor parts. When cosmic rays strike the earth's atmosphere, they cause nuclear reactions with elements in the air and produce a wide range of energetic particles. Because neutrons are uncharged, they can reach aircraft altitudes and sea level. These neutrons are thought to be the most important threat to semiconductor devices and integrated circuits. The best way to determine the failure rate due to these neutrons is to measure the failure rate in a neutron source that has the same spectrum as those produced by cosmic rays. Los Alamos has a high-energy and a low-energy neutron source for semiconductor testing. Both are driven by the 800-MeV proton beam from the LANSCE accelerator. The high-energy neutron source at the Weapons Neutron Research (WNR) facility uses a bare target that is designed to produce fast neutrons with energies from 100 keV to almost 800 MeV. The measured neutron energy distribution from WNR is very similar to that of the cosmic-ray-induced neutrons in the atmosphere. However, the flux provided at the WNR facility is typically 5×107 times more intense than the flux of the cosmic-ray-induced neutrons. This intense neutron flux allows testing at greatly accelerated rates. An irradiation test of less than an hour is equivalent to many years of neutron exposure due to cosmic-ray neutrons. The low-energy neutron source is located at the Lujan Neutron Scattering Center. It is based on a moderated source that provides useful neutrons from subthermal energies to ~100 keV. The characteristics of these sources

The Los Alamos Neutron Science Center Spallation Neutron Sources

Science.gov (United States)

Nowicki, Suzanne F.; Wender, Stephen A.; Mocko, Michael

The Los Alamos Neutron Science Center (LANSCE) provides the scientific community with intense sources of neutrons, which can be used to perform experiments supporting civilian and national security research. These measurements include nuclear physics experiments for the defense program, basic science, and the radiation effect programs. This paper focuses on the radiation effects program, which involves mostly accelerated testing of semiconductor parts. When cosmic rays strike the earth's atmosphere, they cause nuclear reactions with elements in the air and produce a wide range of energetic particles. Because neutrons are uncharged, they can reach aircraft altitudes and sea level. These neutrons are thought to be the most important threat to semiconductor devices and integrated circuits. The best way to determine the failure rate due to these neutrons is to measure the failure rate in a neutron source that has the same spectrum as those produced by cosmic rays. Los Alamos has a high-energy and a low-energy neutron source for semiconductor testing. Both are driven by the 800-MeV proton beam from the LANSCE accelerator. The high-energy neutron source at the Weapons Neutron Research (WNR) facility uses a bare target that is designed to produce fast neutrons with energies from 100 keV to almost 800 MeV. The measured neutron energy distribution from WNR is very similar to that of the cosmic-ray-induced neutrons in the atmosphere. However, the flux provided at the WNR facility is typically 5×107 times more intense than the flux of the cosmic-ray-induced neutrons. This intense neutron flux allows testing at greatly accelerated rates. An irradiation test of less than an hour is equivalent to many years of neutron exposure due to cosmic-ray neutrons. The low-energy neutron source is located at the Lujan Neutron Scattering Center. It is based on a moderated source that provides useful neutrons from subthermal energies to ∼100 keV. The characteristics of these sources, and

The application of n-γ discrimination in 252Cf spontaneous neutron TOF spectra measurement

International Nuclear Information System (INIS)

Zhou Haojun; Zhang Yi; Li Jiansheng; Jin Yu; Wang Jie; Li Chunyuan

2004-01-01

The BC501 scintillator is used as a fast neutron detector. The effect that the pulse rise time method was used to discriminate γ from 252 Cf spontaneous neutron TOF spectra is studied in the experiment. A pulse rise time separation spectra of γ and 252 Cf spontaneous neuron upon 1 MeV is obtained, the n-γ separation function reaches to 4.6. When the result of pulse rise time separation coincides with the time-of-flight spectra in which the neutron energy is upon 0.5 MeV, 0.8 MeV and 1.0 MeV, comparing with the anticoincidence, γ was eliminated 99.90% at least. (authors)

Theoretical investigation of the neutron noise diagnostics of two-dimensional control rod vibrations in a PWR

International Nuclear Information System (INIS)

Pazsit, I.; Analytis, G.T.

1980-01-01

In order to develop a method for monitoring control rod vibrations by neutron noise measurements, the noise induced by two-dimensional vibrations of control elements is investigated. The two-dimensional Green's function relating the small stochastic cross-section fluctuations to the neutron noise is determined for a rectangular slab reactor in the modified one-group theory, and subsequently, the neutron response to two-dimensional vibrating noise sources is investigated. Two possible diagnostical applications are considered: (a) the reconstruction of the mechanical trajectory of the vibrating element by neutron noise measurements, and (b) the possibility of locating the vibrating element in the core. (author)

252Cf-source-correlated transmission measurements for uranyl fluoride deposit in a 24-in.-OD process pipe

International Nuclear Information System (INIS)

Uckan, T.; Mihalczo, J.T.; Valentine, T.E.; Mullens, J.A.

1998-01-01

Characterization of a hydrated uranyl fluoride (UO 2 F 2 ·nH 2 O) deposit in a 17-ft-long, 24-in.-OD process pipe at the former Oak Ridge Gaseous Diffusion Plant was successfully performed by using 252 Cf-source-correlated time-of-flight (TOF) transmission measurements. These measurements of neutrons and gamma rays through the pipe from an external 2521 Cf fission source were used to measure the deposit profile and its distribution along the pipe, the hydration (or H/U), and the total uranium mass. The measurements were performed with a source in an ionization chamber on one side of the pipe and detectors on the other. Scanning the pipe vertically and horizontally produced a spatial and time-dependent radiograph of the deposit in which transmitted gamma rays and neutrons were separated in time. The cross-correlation function between the source and the detector was measured with the Nuclear Weapons Identification System. After correcting for pipe effects, the deposit thickness was determined from the transmitted neutrons and H/U from the gamma rays. Results were consistent with a later intrusive observation of the shape and the color of the deposit; i.e., the deposit was annular and was on the top of the pipe at some locations, demonstrating the usefulness of this method for deposit characterization

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

Laser-driven x-ray and neutron source development for industrial applications of plasma accelerators

Science.gov (United States)

Brenner, C. M.; Mirfayzi, S. R.; Rusby, D. R.; Armstrong, C.; Alejo, A.; Wilson, L. A.; Clarke, R.; Ahmed, H.; Butler, N. M. H.; Haddock, D.; Higginson, A.; McClymont, A.; Murphy, C.; Notley, M.; Oliver, P.; Allott, R.; Hernandez-Gomez, C.; Kar, S.; McKenna, P.; Neely, D.

2016-01-01

Pulsed beams of energetic x-rays and neutrons from intense laser interactions with solid foils are promising for applications where bright, small emission area sources, capable of multi-modal delivery are ideal. Possible end users of laser-driven multi-modal sources are those requiring advanced non-destructive inspection techniques in industry sectors of high value commerce such as aerospace, nuclear and advanced manufacturing. We report on experimental work that demonstrates multi-modal operation of high power laser-solid interactions for neutron and x-ray beam generation. Measurements and Monte Carlo radiation transport simulations show that neutron yield is increased by a factor ~2 when a 1 mm copper foil is placed behind a 2 mm lithium foil, compared to using a 2 cm block of lithium only. We explore x-ray generation with a 10 picosecond drive pulse in order to tailor the spectral content for radiography with medium density alloy metals. The impact of using  >1 ps pulse duration on laser-accelerated electron beam generation and transport is discussed alongside the optimisation of subsequent bremsstrahlung emission in thin, high atomic number target foils. X-ray spectra are deconvolved from spectrometer measurements and simulation data generated using the GEANT4 Monte Carlo code. We also demonstrate the unique capability of laser-driven x-rays in being able to deliver single pulse high spatial resolution projection imaging of thick metallic objects. Active detector radiographic imaging of industrially relevant sample objects with a 10 ps drive pulse is presented for the first time, demonstrating that features of 200 μm size are resolved when projected at high magnification.

Laser-driven x-ray and neutron source development for industrial applications of plasma accelerators

International Nuclear Information System (INIS)

Brenner, C M; Rusby, D R; Armstrong, C; Wilson, L A; Clarke, R; Haddock, D; McClymont, A; Notley, M; Oliver, P; Allott, R; Hernandez-Gomez, C; Neely, D; Mirfayzi, S R; Alejo, A; Ahmed, H; Kar, S; Butler, N M H; Higginson, A; McKenna, P; Murphy, C

2016-01-01

Pulsed beams of energetic x-rays and neutrons from intense laser interactions with solid foils are promising for applications where bright, small emission area sources, capable of multi-modal delivery are ideal. Possible end users of laser-driven multi-modal sources are those requiring advanced non-destructive inspection techniques in industry sectors of high value commerce such as aerospace, nuclear and advanced manufacturing. We report on experimental work that demonstrates multi-modal operation of high power laser-solid interactions for neutron and x-ray beam generation. Measurements and Monte Carlo radiation transport simulations show that neutron yield is increased by a factor ∼2 when a 1 mm copper foil is placed behind a 2 mm lithium foil, compared to using a 2 cm block of lithium only. We explore x-ray generation with a 10 picosecond drive pulse in order to tailor the spectral content for radiography with medium density alloy metals. The impact of using  >1 ps pulse duration on laser-accelerated electron beam generation and transport is discussed alongside the optimisation of subsequent bremsstrahlung emission in thin, high atomic number target foils. X-ray spectra are deconvolved from spectrometer measurements and simulation data generated using the GEANT4 Monte Carlo code. We also demonstrate the unique capability of laser-driven x-rays in being able to deliver single pulse high spatial resolution projection imaging of thick metallic objects. Active detector radiographic imaging of industrially relevant sample objects with a 10 ps drive pulse is presented for the first time, demonstrating that features of 200 μm size are resolved when projected at high magnification. (paper)

Moderator design studies for a new neutron reference source based on the D–T fusion reaction

International Nuclear Information System (INIS)

Mozhayev, Andrey V.; Piper, Roman K.; Rathbone, Bruce A.; McDonald, Joseph C.

2016-01-01

The radioactive isotope Californium-252 ( 252 Cf) is relied upon internationally as a neutron calibration source for ionizing radiation dosimetry because of its high specific activity. The source may be placed within a heavy-water (D 2 O) moderating sphere to produce a softened spectrum representative of neutron fields common to commercial nuclear power plant environments, among others. Due to termination of the U.S. Department of Energy loan/lease program in 2012, the expense of obtaining 252 Cf sources has undergone a significant increase, rendering high output sources largely unattainable. On the other hand, the use of neutron generators in research and industry applications has increased dramatically in recent years. Neutron generators based on deuteriumtritium (D–T) fusion reaction provide high neutron fluence rates and, therefore, could possibly be used as a replacement for 252 Cf. To be viable, the 14 MeV D–T output spectrum must be significantly moderated to approximate common workplace environments. This paper presents the results of an effort to select appropriate moderating materials and design a configuration to reshape the primary neutron field toward a spectrum approaching that from a nuclear power plant workplace. A series of Monte-Carlo (MCNP) simulations of single layer high- and low-Z materials are used to identify initial candidate moderators. Candidates are refined through a similar series of simulations involving combinations of 2–5 different materials. The simulated energy distribution using these candidate moderators are rated in comparison to a target spectrum. Other properties, such as fluence preservation and/or enhancement, prompt gamma production and other characteristics are also considered. - Highlights: • D–T generator neutron calibration field replacement for D 2 O-moderated 252 Cf. • Determination of representative nuclear power plant workplace neutron spectrum. • Simulations to assess moderating materials to soften 14

Stochastic resonance driven by time-modulated correlated coloured noise sources in a single-mode laser

International Nuclear Information System (INIS)

De-Yi, Chen; Li, Zhang

2009-01-01

This paper investigates the phenomenon of stochastic resonance in a single-mode laser driven by time-modulated correlated coloured noise sources. The power spectrum and signal-to-noise ratio R of the laser intensity are calculated by the linear approximation. The effects caused by noise self-correlation time τ 1 , τ 2 and cross-correlated time τ 3 for stochastic resonance are analysed in two ways: τ 1 , τ 2 and τ 3 are taken to be the independent variables and the parameters respectively. The effects of the gain coefficient Γ and loss coefficient K on the stochastic resonance are also discussed. It is found that besides the presence of the standard form and the broad sense of stochastic resonance, the number of extrema in the curve of R versus K is reduced with the increase of the gain coefficient Γ

Imaging of fast-neutron sources using solid-state track-recorder pinhole radiography

International Nuclear Information System (INIS)

Ruddy, F.H.; Gold, R.; Roberts, J.H.; Kaiser, B.J.; Preston, C.C.

1983-08-01

Pinhole imaging methods are being developed and tested for potential future use in imaging the intense neutron source of the Fusion Materials Irradiation Test (FMIT) Facility. Previously reported, extensive calibration measurements of the proton, neutron, and alpha particle response characteristics of CR-39 polymer solid state track recorders (SSTRs) are being used to interpret the results of imaging experiments using both charged particle and neutron pinhole collimators. High resolution, neutron pinhole images of a 252 Cf source have been obtained in the form of neutron induced proton recoil tracks in CR-39 polymer SSTR. These imaging experiments are described as well as their potential future applications to FMIT

Determination of kinetics parameters using stochastic methods in a 252Cf system

International Nuclear Information System (INIS)

Difilippo, F.C.

1988-01-01

Safety analysis and control system design of nuclear systems require the knowledge of neutron kinetics related parameters like effective delayed neutron fraction, neutron lifetime, time between neutron generations and subcriticality margins. Many methods, deterministic and stochastic, are being used, some since the beginning of nuclear power, to measure these important parameters. The method based on the use of the 252 Cf neutron source has been under intense study at the Oak Ridge National Laboratory, both experimentally and theoretically, during the last years. The increasing demand for this isotope in industrial and medical applications and new designs of advanced high flux reactors to produce it make the isotope available as neutron source (only few micrograms are necessary). A thin layer of 252 Cf is deposited in one of the electrodes of a fission chamber which produces pulses each time the 252 Cf disintegrates via α or spontaneous fission decay; the smaller pulses associated with the α decay can be easily discriminated with the important result that we known the time when v/sub c/ neutrons are injected into the system (number of neutrons per fission of 252 Cf). Thus, a small (few cm 3 ) and nonintrusive device can be used as a random pulsed neutron source with known natural properties that do no depend on biases associated with more complex interrogating devices like accelerators. This paper presents a general formalism that relates the kinetics parameters with stochastic descriptors that naturally appear because of the random nature of the production and transport of neutrons

Design of an electron-accelerator-driven compact neutron source for non-destructive assay

Science.gov (United States)

Murata, A.; Ikeda, S.; Hayashizaki, N.

2017-09-01

The threat of nuclear and radiological terrorism remains one of the greatest challenges to international security, and the threat is constantly evolving. In order to prevent nuclear terrorism, it is important to avoid unlawful import of nuclear materials, such as uranium and plutonium. Development of technologies for non-destructive measurement, detection and recognition of nuclear materials is essential for control at national borders. At Tokyo Institute of Technology, a compact neutron source system driven by an electron-accelerator has been designed for non-destructive assay (NDA). This system is composed of a combination of an S-band (2.856 GHz) RF-gun, a tungsten target to produce photons by bremsstrahlung, a beryllium target, which is suitable for use in generating neutrons because of the low threshold energy of photonuclear reactions, and a moderator to thermalize the fast neutrons. The advantage of this system can accelerate a short pulse beam with a pulse width less than 1 μs which is difficult to produce by neutron generators. The amounts of photons and neutron produced by electron beams were simulated using the Monte Carlo simulation code PHITS 2.82. When the RF-gun is operated with an average electron beam current of 0.1 mA, it is expected that the neutron intensities are 1.19 × 109 n/s and 9.94 × 109 n/s for incident electron beam energies of 5 MeV and 10 MeV, respectively.

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

International Nuclear Information System (INIS)

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

2003-01-01

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

Neutron noise calculations in a hexagonal geometry and comparison with analytical solutions

International Nuclear Information System (INIS)

Tran, H. N.; Demaziere, C.

2012-01-01

This paper presents the development of a neutronic and kinetic solver for hexagonal geometries. The tool is developed based on the diffusion theory with multi-energy groups and multi-groups of delayed neutron precursors allowing the solutions of forward and adjoint problems of static and dynamic states, and is applicable to both thermal and fast systems with hexagonal geometries. In the dynamic problems, the small stationary fluctuations of macroscopic cross sections are considered as noise sources, and then the induced first order noise is calculated fully in the frequency domain. Numerical algorithms for solving the static and noise equations are implemented with a spatial discretization based on finite differences and a power iterative solution. A coarse mesh finite difference method has been adopted for speeding up the convergence. Since no other numerical tool could calculate frequency-dependent noise in hexagonal geometry, validation calculations have been performed and benchmarked to analytical solutions based on a 2-D homogeneous system with two-energy groups and one-group of delayed neutron precursor, in which point-like perturbations of thermal absorption cross section at central and non-central positions are considered as noise sources. (authors)

An evaluation of the spontaneous fission prompt neutron spectrum of 252Cf

International Nuclear Information System (INIS)

Bojkov, G.S.; Yurevich, V.I.

1987-01-01

An evaluation of the spontaneous fission prompt neutron spectrum of 252 Cf from 1 keV to 20 MeV is described. Variance-covariance matrices for a number of recent experimental data sets were constructed and used to evaluate the neutron spectrum following a Bayesian procedure. The evaluated spectrum is compared with various experimental and theoretical representations. (author)

Sealed operation of a rf driven ion source for a compact neutron generator to be used for associated particle imaging.

Science.gov (United States)

Wu, Y; Hurley, J P; Ji, Q; Kwan, J W; Leung, K N

2010-02-01

We present the recent development of a prototype compact neutron generator to be used in conjunction with the method of associated particle imaging for the purpose of active neutron interrogation. In this paper, the performance and device specifications of these compact generators that employ rf driven ion sources will be discussed. Initial measurements of the generator performance include a beam spot size of 1 mm in diameter and a neutron yield of 2x10(5) n/s with air cooling.

Correlation studies of neutron multiplicities in the 252Cf spontaneous fission

International Nuclear Information System (INIS)

Alkhazov, I.D.; Dmitriev, V.D.; Kovalenko, S.S.; Kuznetsov, A.V.; Malkin, L.Z.; Petrzhak, K.A.; Petrov, B.F.; Shpakov, V.I.

1988-01-01

Correlations between the numbers of neutrons emitted by the 252 Cf spontaneous fission fragments have been studied as a function of the fragment mass and total kinetic energy. Behaviour of the neutron number dispersions and covariances was studied for the region of symmetric fission. Parameters of the complementary fragment excitation energy distribution (mean values, dispersions, covariances) were determined. Various factors describing correlations between the complementary fragment excitation energies are considered

Measurements of gamma-ray dose from a moderated 252Cf source

International Nuclear Information System (INIS)

McDonald, J.C.; Griffith, R.V.; Plato, P.; Miklos, J.

1983-06-01

The gamma-ray dose fraction from a moderated 252 Cf source was determined by using three types of dosimetry systems. Measurements were carried out in air at a distance of 35 cm from the surface of the moderating sphere (50 cm from the source which is at the center of the sphere) to the geometrical center of each detector. The moderating sphere is 0.8-mm-thick stainless steel shell filled with D 2 O and covered with 0.5 mm of cadmium. Measurements were also carried out with instruments and dosimeters positioned at the surface of a 40 cm x 40 cm x 15 cm plexiglass irradiation phantom whose front surface was also 35 cm from the surface of the moderating sphere. A-150 tissue-equivalent (TE) plastic ionization chambers and a TE proportional counter (TEPC) were used to measure tissue dose, from which the neutron dose equivalent was computed. The ratio of gamma-ray dose to the neutron dose equivalent was determined by using a relatively neutron-insensitive Geiger-Mueller (GM) counter and thermoluminescent dosimeters (TLD). In addition, the event-size spectrum measured by the TEPC was also used to compute the gamma-ray dose fraction. The average value for the ratio of gamma-ray dose to neutron dose equivalent was found to be 0.18 with an uncertainty of about +-18%

Use of neutron activation and X-ray fluorescence with radioactive sources (Cf-252 and Am-241) for the instrumental qualiquantitative simultaneous analysis of some elements in samples of mineral supplement for animals

International Nuclear Information System (INIS)

Simabuco, S.M.

1984-01-01

To study the possibility of using two non-destructive (neutron activation and X-ray fluorescence) analyses in simultaneous quali-quantitative evaluations of some elements in mineral supplement for animals, a Cf-252 neutron source (11.3 mCi; 21.1 μgrams) and a Am-241 low energy gamma-ray emitter source (59.5 KeV; 100 mCi) were employed. For these sources, shieldings and sample irradiation systems were built. For the neutron activation analysis a reservoir of 72 cm height and 43 cm diameter was filled with paraffine, and the samples and neutron sources were put inside this reservoir using polypropilene and nylon tubes. To detect the gamma-rays emitted by the radioisotopes a well-type solid NaI(Tl) crystal scintillator (3x3') was used, coupled to a multi-channel analyser. For the X-ray fluorescence analysis a lead cylinder of 9.75 cm height and 5.6 cm diameter (with 0.7 cm thickness) was made and internally lined with a 0.36 mm copper and 0.1 mm aluminium foil. (Author) [pt

Standard test method for nondestructive assay of nuclear material in scrap and waste by passive-Active neutron counting using 252Cf shuffler

CERN Document Server

American Society for Testing and Materials. Philadelphia

2008-01-01

1.1 This test method covers the nondestructive assay of scrap and waste items for U, Pu, or both, using a 252Cf shuffler. Shuffler measurements have been applied to a variety of matrix materials in containers of up to several 100 L. Corrections are made for the effects of matrix material. Applications of this test method include measurements for safeguards, accountability, TRU, and U waste segregation, disposal, and process control purposes (1, 2, 3). 1.1.1 This test method uses passive neutron coincidence counting (4) to measure the 240Pu-effective mass. It has been used to assay items with total Pu contents between 0.03 g and 1000 g. It could be used to measure other spontaneously fissioning isotopes such as Cm and Cf. It specifically describes the approach used with shift register electronics; however, it can be adapted to other electronics. 1.1.2 This test method uses neutron irradiation with a moveable Cf source and counting of the delayed neutrons from the induced fissions to measure the 235U equiva...

Prompt neutron fission spectrum mean energies for the fissile nuclides and 252Cf

International Nuclear Information System (INIS)

Holden, N.E.

1985-01-01

The international standard for a neutron spectrum is that produced from the spontaneous fission of 252 Cf, while the thermal neutron induced fission neutron spectra for the four fissile nuclides, 233 U, 235 U, 239 Pu, and 241 Pu are of interest from the standpoint of nuclear reactors. The average neutron energies of these spectra are tabulated. The individual measurements are recorded with the neutron energy range measured, the method of detection as well as the average neutron energy for each author. Also tabulated are the measurements of the ratio of mean energies for pairs of fission neutron spectra. 75 refs., 9 tabs

Neutron Transport Methods for Accelerator-Driven Systems

International Nuclear Information System (INIS)

Nicholas Tsoulfanidis; Elmer Lewis

2005-01-01

The objective of this project has been to develop computational methods that will enable more effective analysis of Accelerator Driven Systems (ADS). The work is centered at the University of Missouri at Rolla, with a subcontract at Northwestern University, and close cooperation with the Nuclear Engineering Division at Argonne National Laboratory. The work has fallen into three categories. First, the treatment of the source for neutrons originating from the spallation target which drives the neutronics calculations of the ADS. Second, the generalization of the nodal variational method to treat the R-Z geometry configurations frequently needed for scoping calculations in Accelerator Driven Systems. Third, the treatment of void regions within variational nodal methods as needed to treat the accelerator beam tube

Source driven breeding thermal power reactors

International Nuclear Information System (INIS)

Greenspan, E.; Schneider, A.; Misulovin, A.; Gilai, D.; Levin, P.; Ben-Gurion Univ. of the Negev, Beersheba

1978-03-01

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

Approaches for the generation of a covariance matrix for the Cf-252 fission-neutron spectrum

International Nuclear Information System (INIS)

Mannhart, W.

1983-01-01

After a brief retrospective glance is cast at the situation, the evaluation of the Cf-252 neutron spectrum with a complete covariance matrix based on the results of integral experiments is proposed. The different steps already taken in such an evaluation and work in progress are reviewed. It is shown that special attention should be given to the normalization of the neutron spectrum which must be reflected in the covariance matrix. The result of the least-squares adjustment procedure applied can easily be combined with the results of direct spectrum measurements and should be regarded as the first step in a new evaluation of the Cf-252 fission-neutron spectrum. (author)

CF3CH(ONO)CF3: Synthesis, IR spectrum, and use as OH radical source for kinetic and mechanistic studies

DEFF Research Database (Denmark)

Andersen, Mads Peter Sulbæk; Hurley, MD; Ball, JC

2003-01-01

The synthesis, IR spectrum, and first-principles characterization of CF3CH(ONO)CF3 as well as its use as an OH radical source in kinetic and mechanistic studies are reported. CF3CH(ONO)CF3 exists in two conformers corresponding to rotation about the RCO-NO bond. The more prevalent trans conformer......C(O)CF3 and, by implication, OH radicals in 100% yield. CF3CH(ONO)CF3 photolysis is a convenient source of OH radicals in the studies of the yields of CO, CO2, HCHO, and HC(O)OH products which can be difficult to measure using more conventional OH radical sources (e.g., CH3ONO photolysis). CF3CH...

1987 Neutron and gamma personnel dosimeter intercomparison study using a D2O-moderated 252Cf source

International Nuclear Information System (INIS)

Swaja, R.E.; West, L.E.; Sims, C.S.; Welty, T.J.

1989-05-01

The thirteenth Personnel Dosimetry Intercomparison Study (i.e., PDIS 13) was conducted during April 1987 as a joint effort by Oak Ridge National Laboratory's (ORNL) Dosimetry Applications Research Group and the Southwest Radiation Calibration Center at the University of Arkansas. A total of 48 organizations (34 from the US and 14 from abroad) participated in PDIS 13. Participants submitted a total of 1,113 neutron and gamma dosimeters for this mixed field study. The dosimeters were transferred by mail and were handled by experimental personnel at ORNL and the University of Arkansas. The type of neutron dosimeter and the percentage of participants submitting that type are as follows: TLD-albedo (49%), direct interaction TLD (31%), CR-39 (17%), film (3%). The type of gamma dosimeter and the percentage of participants submitting that type are as follows: Li 2 B 4 O 7 , alone or in combination with CaSO 4 , (69%), 7 LiF (28%), natural LiF (3%). Radiation exposures in PDIS 13 were limited to 0.5 and 1.5 mSv from 252 Cf moderated by 15-cm of D 2 O. Traditional exposures using the Health Physics Research Reactor (HPRR) were not possible due to the fact that all reactors at ORNL, including the HPRR, were shutdown by order of the Department of Energy at the time the intercomparison was performed. Planned exposures using a 238 PuBe source were negated by a faulty timing mechanism. Based on accuracy and precision, direct interaction TLD dosimeters exhibited the best performance in PDIS 13 neutron measurements. They were followed, in order of best performance, by CR-39, TLD albedo, and film. The Li 2 B 4 O 7 type TLD dosimeters exhibited the best performance in PDIS 13 gamma measurements. They were followed by natural LiF, 7 LiF, and film. 12 refs., 1 fig., 5 tabs

A new facility for non-destructive assay using a 252Cf source.

Science.gov (United States)

Stevanato, L; Caldogno, M; Dima, R; Fabris, D; Hao, Xin; Lunardon, M; Moretto, S; Nebbia, G; Pesente, S; Pino, F; Sajo-Bohus, L; Viesti, G

2013-03-01

A new laboratory facility for non-destructive analysis (NDA) using a time-tagged (252)Cf source is presented. The system is designed to analyze samples having maximum size of about 20 × 25 cm(2), the material recognition being obtained by measuring simultaneously total and energy dependent transmission of neutrons and gamma rays. The equipment technical characteristics and performances of the NDA system are presented, exploring also limits due to the sample thickness. Some recent applications in the field of cultural heritage are presented. Copyright © 2012 Elsevier Ltd. All rights reserved.

Cf-252 neutron brachytherapy: an advance for bulky localized cancer therapy

International Nuclear Information System (INIS)

Maruyama, Y.

1984-01-01

The physical and radiobiogical basis as well as the rationale for neutron brachytherapy, using Cf-252, in human cancer therapy is reviewed. Cf-252 brachytherapy represents an economical and effective form of neutron radiotherapy that is readily and safely applied clinically. It can be used anywhere in the world without unusual personnel, equipment or facilities, or prohibitive expenses or maintenance costs. Used on bulky head and neck, thoracic, abdominal, pelvic, brain and appendage cancers, it overcomes hypoxic radioresistance and produces remarkable rates of tumor clearance. It is easily combined with photon radiotherapy and in proper schedules and doses, it can control advanced but still localized regional cancers to produce tumor cure. It will clear the local manifestations of recurrent or metastatic tumors or advanced stages of primary tumors and therefore in conjunction with other adjuvant therapies offers much more effective tumor control and palliation than present conventional therapy. (Auth.)

Feasibility of subcriticality and NDA measurements for spent fuel by frequency analysis techniques with 252Cf

International Nuclear Information System (INIS)

Mihalczo, J.T.; Valentine, T.E.; Mattingly, J.K.

1996-01-01

The 252 Cf-source-driven frequency analysis method can be used for measuring the subcritical neutron multiplication factor of arrays of LWR fuel and as little as a single PWR fuel assembly. These measurements can be used to verify the criticality safety margins of spent LWR fuel configurations and thus could be a means of obtaining the information to justify burnup credit for spent LWR transportation/storage casks. In addition, the data can be used to validate calculational methods for criticality safety. These measurements provide parameters that have a higher sensitivity to changes in fissile mass than neutron multiplication factor and thus serve as a better test of calculational methods. The analysis have also shown that measurement of the cross power spectral density (CPSD) between detectors on one side of a single fuel assembly and an internal or external 252 Cf source driving the fission chain multiplication process can be used for nondestructive assay of fissile mass along the length of the assembly. This CPSD is a smooth function of fissile mass and does not depend on the varying inherent source in the fuel assembly and thus is ideal for fissile mass assay

'Quantization' of stochastic variables: description and effects on the input noise sources in a BWR

International Nuclear Information System (INIS)

Matthey, M.

1979-01-01

A set of macrostochastic and discrete variables, with Markovian properties, is used to characterize the state of a BWR, whose input noise sources are of interest. The ratio between the auto-power spectral density (APSD) of the neutron noise fluctuations and the square modulus of the transfer function (SMTF) defines 'the total input noise source' (TINS), the components of which are the different noise source corresponding to the relevant variables. A white contribution to TINS arises from the birth and death processes of neutrons in the reactor and corresponds to a 'shot noise' (SN). Non-white contributions arise from fluctuations of the neutron cross-sections caused by fuel temperature and steam content variations. These terms called 'Flicker noises' (FN) are characterized by cut-off frequencies related to time constants of reactivity feedback effects. The respective magnitudes of the shot and flicker noises depend not only on the frequency, the feedback reactivity coefficients or the power of the reactor, but also on the 'quantization' of the continuous variables introduced such as fuel temperature and steam content. The effects of this last 'quantization' on the shapes of the noise sources and their sum are presented in this paper. (author)

Experimental demonstration of a compact epithermal neutron source based on a high power laser

Science.gov (United States)

Mirfayzi, S. R.; Alejo, A.; Ahmed, H.; Raspino, D.; Ansell, S.; Wilson, L. A.; Armstrong, C.; Butler, N. M. H.; Clarke, R. J.; Higginson, A.; Kelleher, J.; Murphy, C. D.; Notley, M.; Rusby, D. R.; Schooneveld, E.; Borghesi, M.; McKenna, P.; Rhodes, N. J.; Neely, D.; Brenner, C. M.; Kar, S.

2017-07-01

Epithermal neutrons from pulsed-spallation sources have revolutionised neutron science allowing scientists to acquire new insight into the structure and properties of matter. Here, we demonstrate that laser driven fast (˜MeV) neutrons can be efficiently moderated to epithermal energies with intrinsically short burst durations. In a proof-of-principle experiment using a 100 TW laser, a significant epithermal neutron flux of the order of 105 n/sr/pulse in the energy range of 0.5-300 eV was measured, produced by a compact moderator deployed downstream of the laser-driven fast neutron source. The moderator used in the campaign was specifically designed, by the help of MCNPX simulations, for an efficient and directional moderation of the fast neutron spectrum produced by a laser driven source.

A new Monte Carlo method for neutron noise calculations in the frequency domain

International Nuclear Information System (INIS)

Rouchon, Amélie; Zoia, Andrea; Sanchez, Richard

2017-01-01

Neutron noise equations, which are obtained by assuming small perturbations of macroscopic cross sections around a steady-state neutron field and by subsequently taking the Fourier transform in the frequency domain, have been usually solved by analytical techniques or by resorting to diffusion theory. A stochastic approach has been recently proposed in the literature by using particles with complex-valued weights and by applying a weight cancellation technique. We develop a new Monte Carlo algorithm that solves the transport neutron noise equations in the frequency domain. The stochastic method presented here relies on a modified collision operator and does not need any weight cancellation technique. In this paper, both Monte Carlo methods are compared with deterministic methods (diffusion in a slab geometry and transport in a simplified rod model) for several noise frequencies and for isotropic and anisotropic noise sources. Our stochastic method shows better performances in the frequency region of interest and is easier to implement because it relies upon the conventional algorithm for fixed-source problems.

Mechanical Design and Analysis of a 200 MHz, Bolt-together RFQ forthe Accelerator Driven Neutron Source

Energy Technology Data Exchange (ETDEWEB)

Virostek, Steve; Hoff, Matt; Li, Derun; Staples, John; Wells,Russell

2007-06-20

A high-yield neutron source to screen sea-land cargocontainers for shielded Special Nuclear Materials (SNM) has been designedat LBNL [1,2]. The Accelerator-Driven Neutron Source (ADNS) uses theD(d,n)3He reaction to create a forward directed neutron beam. Keycomponents are a high-current radio-frequency quadrupole (RFQ)accelerator and a high-power target capable of producing a neutron fluxof>107 n/(cm2 cdot s) at a distance of 2.5 m. The mechanical designand analysis of the four-module, bolt-together RFQ will be presentedhere. Operating at 200 MHz, the 5.1 m long RFQ will accelerate a 40 mAdeuteron beam to 6 MeV. At a 5 percent duty factor, the time-average d+beam current on target is 1.5 mA. Each of the 1.27 m long RFQ moduleswill consist of four solid OFHC copper vanes. A specially designed 3-DO-ring will provide vacuum sealing between both the vanes and themodules. RF connections are made with canted coil spring contacts. Aseries of 60 water-cooled pi-mode rods provides quadrupole modestabilization. A set of 80 evenly spaced fixed slug tuners is used forfinal frequency adjustment and local field perturbationcorrection.

Vacuum treatment of CR-39 for the reduction of background in neutron induced autoradiography of boron

International Nuclear Information System (INIS)

Freyer, K.; Treutler, H.C.; Dietze, K.; Hunyadi, I.; Csige, I.; Somogyi, G.

1991-01-01

The influence of etching time and vacuum processing before, during, and after neutron irradiation on the ''signal/noise'' ratio of the neutron induced autoradiography of boron using CR-39 track detectors has been investigated. The neutron irradiation was carried out in a vacuum chamber using the Cf-252 neutron source of the Central Institute of Isotope and Radiation Research at Leipzig. Hungarian-made CR-39 type track detectors, MA-ND/p and MA-ND/α, produced in different years, are compared. After a few hours of 2 kPa (about 10 -2 Torr) vacuum treatment, the ''signal/noise'' ratio for boron determination is improved remarkably in most cases of the detector and etching-time combinations used. (author)

Calculation of neutron and gamma-ray energy spectra in liquid air and liquid nitrogen due to 14-MeV neutron and californium-252 sources

International Nuclear Information System (INIS)

Straker, E.A.; Gritzner, M.L.; Harris, L. Jr.

1978-01-01

Calculations of neutron and gamma-ray fluences from 14-MeV neutron and 252 Cf sources in liquid air and liquid nitrogen have been performed. These calculations were made specifically for comparison with experimental data measured at Stohl, Federal Republic of Germany. The discrete-ordinates method was utilized with neutron and gamma-ray cross sections from ENDF/B-IV. One-dimensional calculational models were developed for the sources and tank. Limited comparisons are made with experimental data

Multielemental nondestructive neutron activation analysis of Dy, Mn, Eu, Na, Ga, W, La and Sm involving cyclic irradiations with 252Cf

International Nuclear Information System (INIS)

Narkhede, S.S.; Turel, Z.R.

1995-01-01

Dy, Mn, Eu, Na, Ga, W, La and Sm respond very well to INAA technique because of their favourable nuclear properties such as high thermal neutron cross-section or abundance. In the present work a method has been developed for the determination of these elements employing cycle irradiation with 252 Cf thermal neutron source. Radioassaying of the irradiated sample and standard was done employing HPGe detector in conjunction with a PC based MCA units. (author). 2 tabs

Calculation of the neutron noise induced by periodic deformations of a large sodium-cooled fast reactor core

International Nuclear Information System (INIS)

Zylbersztejn, F.; Tran, H.N.; Pazsit, I.; Filliatre, P.; Jammes, C.

2014-01-01

The subject of this paper is the calculation of the neutron noise induced by small-amplitude stationary radial variations of the core size (core expansion/compaction, also called core flowering) of a large sodium-cooled fast reactor. The calculations were performed on a realistic model of the European Sodium Fast Reactor (ESFR) core with a thermal output of 3600 MW(thermal), using a multigroup neutron noise simulator. The multigroup cross sections and their fluctuations that represent the core geometry changes for the neutron noise calculations were generated by the code ERANOS. The space and energy dependences of the noise source represented by the core expansion/compaction and the induced neutron noise are calculated and discussed. (authors)

RBE of Cf-252 neutrons as determined by its lethal, mutagenic, and cytogenetic effects on human cells

International Nuclear Information System (INIS)

Ban, Sadayuki

1989-01-01

To assess the biological effects of neutrons, a man-made spontaneously fissioning isotope, Cf-252, is useful as an experimental model to obtain basic biological data on mixed radiation of gamma-rays and neutrons. The paper describes the lethal effect of Cf-252 radiation on human skin fibroblasts, its lethal and mutagenic effect on HeLa MR cells, and the micronuclei inducing effect on human peripheral lymphocytes. Dose-survival responses of three fibroblast cell strains exposed to Cf-252 radiation are measured. Individual difference is larger than the experimental fluctuation. D 10 values of each strain are obtained from the linear model and linear-quadratic model. Though the dose rate of X-ray is higher than that of Cf-252 radiations, the mean value of RBE(n+γ) is simply obtained as 1.86+0.31 (RBE:relative biological effectiveness). RBE(n) of Cf-252 neutrons to high-dose-rate X-rays is 2.29. After X-ray irradiation, the survival curve of HeLa MR cells gives an extrapolation number of 3.6. It is 1.3 after Cf-252 irradiation. At 50% survival, RBE(n+γ) and RBE(n) are 2.05 and 2.6, respectively. At 10% survival they are 2.05 and 2.6. The mutation frequencies after X-ray irradiation showed a significant non-linear increase with dose. Those after Cf-252 irradiation increase linearly with dose. (N.K.)

Options for the Delft advanced neutron source

International Nuclear Information System (INIS)

Gibcus, H.P.M.; Leege, P.F.A. de; Labohm, F.; Vries, J.W. de; Verkooijen, A.H.M.; Valko, J.; Feltes, W.; Heinecke, J.

2003-01-01

Results of feasibility studies are presented for options for an advanced neutron source for the Delft reactor including upgrading the HOR, a 2 MW pool-type research reactor at the Delft University of Technology. The primary utilisation of the HOR focuses on beam research applications with neutrons and positrons. The aim of being scientifically competitive in that research area requires a thermal neutron flux level of at least 1x10 14 n/cm 2 /s. The feasibility of an accelerator driven neutron source and upgrading the present core to a super compact core for reaching this goal has been investigated at large from a safety and operational point of view. For the upgraded core, a 3x3 fuel assembly arrangement and beryllium reflected at all sides was chosen. Figures on the system performance, including the merits of a cold neutron source application feeding the neutron guide system, are presented. (author)

Investigation and optimisation of mobile NaI(Tl) and 3He-based neutron detectors for finding point sources

International Nuclear Information System (INIS)

Nilsson, Jonas M.C.; Finck, Robert R.; Rääf, Christopher

2015-01-01

Neutron radiation produces high-energy gamma radiation through (n,γ) reactions in matter. This can be used to detect neutron sources indirectly using gamma spectrometers. The sensitivity of a gamma spectrometer to neutrons can be amplified by surrounding it with polyvinyl chloride (PVC). The hydrogen in the PVC acts as a moderator and the chlorine emits prompt gammas when a neutron is captured. A 4.7-l 3 He-based mobile neutron detector was compared to a 4-l NaI(Tl)-detector covered with PVC using this principle. Methods were also developed to optimise the measurement parameters of the systems. The detector systems were compared with regard to their ability to find 241 AmBe, 252 Cf and 238 Pu– 13 C neutron sources. Results from stationary measurements were used to calculate optimal integration times as well as minimum detectable neutron emission rates. It was found that the 3 He-based detector was more sensitive to 252 Cf sources whereas the NaI(Tl) detector was more sensitive to 241 AmBe and 238 Pu– 13 C sources. The results also indicated that the sensitivity of the detectors to sources at known distances could theoretically be improved by 60% by changing from fixed integration times to list mode in mobile surveys

Confusion noise from LISA capture sources

International Nuclear Information System (INIS)

Barack, Leor; Cutler, Curt

2004-01-01

Captures of compact objects (COs) by massive black holes (MBHs) in galactic nuclei will be an important source for LISA, the proposed space-based gravitational wave (GW) detector. However, a large fraction of captures will not be individually resolvable - either because they are too distant, have unfavorable orientation, or have too many years to go before final plunge - and so will constitute a source of 'confusion noise', obscuring other types of sources. In this paper we estimate the shape and overall magnitude of the GW background energy spectrum generated by CO captures. This energy spectrum immediately translates to a spectral density S h capt (f) for the amplitude of capture-generated GWs registered by LISA. The overall magnitude of S h capt (f) is linear in the CO capture rates, which are rather uncertain; therefore we present results for a plausible range of rates. S h capt (f) includes the contributions from both resolvable and unresolvable captures, and thus represents an upper limit on the confusion noise level. We then estimate what fraction of S h capt (f) is due to unresolvable sources and hence constitutes confusion noise. We find that almost all of the contribution to S h capt (f) coming from white dwarf and neutron star captures, and at least ∼30% of the contribution from black hole captures, is from sources that cannot be individually resolved. Nevertheless, we show that the impact of capture confusion noise on the total LISA noise curve ranges from insignificant to modest, depending on the rates. Capture rates at the high end of estimated ranges would raise LISA's overall (effective) noise level [fS h eff (f)] 1/2 by at most a factor ∼2 in the frequency range 1-10 mHz, where LISA is most sensitive. While this slightly elevated noise level would somewhat decrease LISA's sensitivity to other classes of sources, we argue that, overall, this would be a pleasant problem for LISA to have: It would also imply that detection rates for CO captures

Study on development and actual application of scientific crime detection technique using small scale neutron radiation source

International Nuclear Information System (INIS)

Suzuki, Yasuhiro; Kishi, Toru; Tachikawa, Noboru; Ishikawa, Isamu.

1997-01-01

PGA (Prompt γ-ray Analysis) is an analytic method of γ-ray generated from atomic nuclei of elements in the specimen just after irradiation (within 10(exp-14)sec.) of neutron to it. As using neutron with excellent transmission for an exciting source, this method can be used for inspecting the matters in closed containers non-destructively, and can also detect non-destructively light elements such as boron, nitrogen and others difficult by other non-destructive analysis. Especially, it is found that this method can detect such high concentration of nitrogen, chlorine and others which are characteristic elements for the explosives. However, as there are a number of limitations at the nuclear reactor site, development of an analytical apparatus for small scale neutron radiation source was begun, at first. In this fiscal year, analysis of the light elements such as nitrogen, chlorine and others using PGA was attempted by using 252-Cf as the simplest neutron source in its operation. As the 252-Cf neutron flux was considerably lower than that of nuclear reactor, its analytical sensitivity was also investigated. (G.K.)

A 252Cf based nondestructive assay system for fissile material

International Nuclear Information System (INIS)

Menlove, H.O.; Crane, T.W.

1978-01-01

A modulated 252 Cf source assay system 'Shuffler' based on fast-or-thermal-neutron interrogation combined with delayed-neutron counting has been developed for the assay of fissile material. The 252 Cf neutron source is repetitively transferred from the interrogation position to a shielded position while the delayed neutrons are counted in a high efficiency 3 He neutron well-counter. For samples containing plutonium, this well-counter is also used in the passive coincidence mode to assay the effective 240 Pu content. The design of an optimized neutron tailoring assembly for fast-neutron interrogation using a Monte Carlo Neutron Computer Code is described. The Shuffler system has been applied to the assay of fuel pellets, inventory samples, irradiated fuel and plutonium mixed-oxide fuel. The system can assay samples with fissile contents from a few milligrams up to several kilograms using thermal-neutron interrogation for the low mass samples and fast-neutron interrogation for the high mass samples. Samples containing 235 U- 238 U, or 233 U-Th, or UO 2 -PuO 2 fuel mixtures have been assayed with the Shuffler system. (Auth.)

Frequency domain Monte Carlo simulation method for cross power spectral density driven by periodically pulsed spallation neutron source using complex-valued weight Monte Carlo

International Nuclear Information System (INIS)

Yamamoto, Toshihiro

2014-01-01

Highlights: • The cross power spectral density in ADS has correlated and uncorrelated components. • A frequency domain Monte Carlo method to calculate the uncorrelated one is developed. • The method solves the Fourier transformed transport equation. • The method uses complex-valued weights to solve the equation. • The new method reproduces well the CPSDs calculated with time domain MC method. - Abstract: In an accelerator driven system (ADS), pulsed spallation neutrons are injected at a constant frequency. The cross power spectral density (CPSD), which can be used for monitoring the subcriticality of the ADS, is composed of the correlated and uncorrelated components. The uncorrelated component is described by a series of the Dirac delta functions that occur at the integer multiples of the pulse repetition frequency. In the present paper, a Monte Carlo method to solve the Fourier transformed neutron transport equation with a periodically pulsed neutron source term has been developed to obtain the CPSD in ADSs. Since the Fourier transformed flux is a complex-valued quantity, the Monte Carlo method introduces complex-valued weights to solve the Fourier transformed equation. The Monte Carlo algorithm used in this paper is similar to the one that was developed by the author of this paper to calculate the neutron noise caused by cross section perturbations. The newly-developed Monte Carlo algorithm is benchmarked to the conventional time domain Monte Carlo simulation technique. The CPSDs are obtained both with the newly-developed frequency domain Monte Carlo method and the conventional time domain Monte Carlo method for a one-dimensional infinite slab. The CPSDs obtained with the frequency domain Monte Carlo method agree well with those with the time domain method. The higher order mode effects on the CPSD in an ADS with a periodically pulsed neutron source are discussed

Evaluation of neutron sources for ISAGE-in-situ-NAA for a future lunar mission.

Science.gov (United States)

Li, X; Breitkreutz, H; Burfeindt, J; Bernhardt, H-G; Trieloff, M; Hopp, J; Jessberger, E K; Schwarz, W H; Hofmann, P; Hiesinger, H

2011-11-01

For a future Moon landing, a concept for an in-situ NAA involving age determination using the (40)Ar-(39)Ar method is developed. A neutron source (252)Cf is chosen for sample irradiation on the Moon. A special sample-in-source irradiation geometry is designed to provide a homogeneous distribution of neutron flux at the irradiation position. Using reflector, the neutron flux is likely to increase by almost 200%. Sample age of 1Ga could be determined. Elemental analysis using INAA is discussed. Copyright © 2011 Elsevier Ltd. All rights reserved.

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

International Nuclear Information System (INIS)

Kim, S.S.; Levine, S.H.

1985-01-01

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

On the neutron noise diagnostics of pressurized water reactor control rod vibrations. 1. periodic vibrations

International Nuclear Information System (INIS)

Pazsit, I.; Glockler, O.

1983-01-01

Based on the theory of neutron noise arising from the vibration of a localized absorber, the possibility of rod vibration diagnostics is investigated. It is found that noise source characteristics, namely rod position and vibration trajectory and spectra, can be unfolded from measured neutron noise signals. For the localization process, the first and more difficult part of the diagnostics, a procedure is suggested whose novelty is that it is applicable in case of arbitrary vibration trajectories. Applicability of the method is investigated in numerical experiments where effects of background noise are also accounted for

Ion source requirements for pulsed spallation neutron sources

International Nuclear Information System (INIS)

Alonso, J.R.

1995-10-01

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

Mechanical Design and Analysis of a 200 MHz, Bolt-together RFQ for the Accelerator Driven Neutron Source

International Nuclear Information System (INIS)

Virostek, Steve; Hoff, Matt; Li, Derun; Staples, John; Wells, Russell

2007-01-01

A high-yield neutron source to screen sea-land cargo containers for shielded Special Nuclear Materials (SNM) has been designed at LBNL [1,2]. The Accelerator-Driven Neutron Source (ADNS) uses the D(d,n)3He reaction to create a forward directed neutron beam. Key components are a high-current radio-frequency quadrupole (RFQ) accelerator and a high-power target capable of producing a neutron flux of >107 n/(cm2 cdot s) at a distance of 2.5 m. The mechanical design and analysis of the four-module, bolt-together RFQ will be presented here. Operating at 200 MHz, the 5.1 m long RFQ will accelerate a 40 mA deuteron beam to 6 MeV. At a 5 percent duty factor, the time-average d+beam current on target is 1.5 mA. Each of the 1.27 m long RFQ modules will consist of four solid OFHC copper vanes. A specially designed 3-DO-ring will provide vacuum sealing between both the vanes and the modules. RF connections are made with canted coil spring contacts. A series of 60 water-cooled pi-mode rods provides quadrupole mode stabilization. A set of 80 evenly spaced fixed slug tuners is used for final frequency adjustment and local field perturbation correction

Vacuum treatment of CR-39 for the reduction of background in neutron induced autoradiography of boron

Energy Technology Data Exchange (ETDEWEB)

Freyer, K.; Treutler, H.C.; Dietze, K. (Central Inst. of Isotope and Radiation Research, Leipzig (Germany, F.R.)); Hunyadi, I.; Csige, I.; Somogyi, G. (Magyar Tudomanyos Akademia, Debrecen (Hungary). Atommag Kutato Intezete)

1991-01-01

The influence of etching time and vacuum processing before, during, and after neutron irradiation on the ''signal/noise'' ratio of the neutron induced autoradiography of boron using CR-39 track detectors has been investigated. The neutron irradiation was carried out in a vacuum chamber using the Cf-252 neutron source of the Central Institute of Isotope and Radiation Research at Leipzig. Hungarian-made CR-39 type track detectors, MA-ND/p and MA-ND/{alpha}, produced in different years, are compared. After a few hours of 2 kPa (about 10{sup -2} Torr) vacuum treatment, the ''signal/noise'' ratio for boron determination is improved remarkably in most cases of the detector and etching-time combinations used. (author).

Microstructured boron foil scintillating G-GEM detector for neutron imaging

Energy Technology Data Exchange (ETDEWEB)

Fujiwara, Takeshi, E-mail: fujiwara-t@aist.go.jp [Research Institute for Measurement and Analytical Instrumentation, Advanced Industrial Science and Technology (AIST), Tsukuba, Ibaraki (Japan); Center for Advanced Photonics, Neutron Beam Technology Team, RIKEN, Saitama (Japan); Bautista, Unico [Department of Nuclear Engineering and Management, The University of Tokyo, Tokyo (Japan); Philippine Nuclear Research Institute-Department of Science and Technology (PNRI-DOST), Commonwealth Avenue, Diliman, Quezon City (Philippines); Mitsuya, Yuki [Nuclear Professional School, The University of Tokyo, Tokai-mura, Naka-gun, Ibaraki (Japan); Takahashi, Hiroyuki [Department of Nuclear Engineering and Management, The University of Tokyo, Tokyo (Japan); Yamada, Norifumi L. [Neutron Science Laboratory, Institute of Material Structure Science, High Energy Accelerator Research Organization (KEK) (Japan); Otake, Yoshie; Taketani, Atsushi [Center for Advanced Photonics, Neutron Beam Technology Team, RIKEN, Saitama (Japan); Uesaka, Mitsuru [Nuclear Professional School, The University of Tokyo, Tokai-mura, Naka-gun, Ibaraki (Japan); Toyokawa, Hiroyuki [Research Institute for Measurement and Analytical Instrumentation, Advanced Industrial Science and Technology (AIST), Tsukuba, Ibaraki (Japan)

2016-12-01

In this study, a new simple neutron imaging gaseous detector was successfully developed by combining a micro-structured {sup 10}B foil, a glass gas electron multiplier (G-GEM), and a mirror–lens–charge-coupled device (CCD)–camera system. The neutron imaging system consists of a chamber filled with Ar/CF{sub 4} scintillating gas mixture. Inside this system, the G-GEM is mounted for gas multiplication. The neutron detection in this system is based on the reaction between {sup 10}B and neutrons. A micro-structured {sup 10}B is developed to overcome the issue of low detection efficiency. Secondary electrons excite Ar/CF{sub 4} gas molecules, and high-yield visible photons are emitted from those excited gas molecules during the gas electron multiplication process in the G-GEM holes. These photons are easily detected by a mirror–lens–CCD–camera system. A neutron radiograph is then simply formed. We obtain the neutron images of different materials with a compact accelerator-driven neutron source. We confirm that the new scintillating G-GEM-based neutron imager works properly with low gamma ray sensitivity and exhibits a good performance as a new simple digital neutron imaging device.

Cylindrical IEC neutron source design for driven research reactor operation

International Nuclear Information System (INIS)

Miley, G.H.; Ulmen, B.; Amadio, G.; Leon, H.; Hora, H.

2009-01-01

A resurgence in nuclear power use is now underway worldwide. However, due many university research reactors shutdown, they must rely on using subcritical assemblies which employs a cylindrical Inertial Electrostatic Confinement (IEC) device to provide a fusion neutron source. The source is inserted in a fuel element position, with its power input controlled externally at a control panel. This feature opens the way to use of the critical assembly for a number of transient experiments such as sub-critical pulsing and neutron wave propagation. That in turn adds important new insights and excitement for the student teaching laboratory. (author)

Power spectral density measurements with 252Cf for a light water moderated research reactor

International Nuclear Information System (INIS)

King, W.T.; Mihalczo, J.T.

1979-01-01

A method of determining the reactivity of far subcritical systems from neutron noise power spectral density measurements with 252 Cf has previously been tested in fast reactor critical assemblies: a mockup of the Fast Flux Test Facility reactor and a uranium metal sphere. Calculations indicated that this measurement was feasible for a pressurized water reactor (PWR). In order to evaluate the ability to perform these measurements with moderated reactors which have long prompt neutron lifetimes, measurements were performed with a small plate-type research reactor whose neutron lifetime (57 microseconds) was about a factor of three longer than that of a PWR and approx. 50% longer than that of a boiling water reactor. The results of the first measurements of power spectral densities with 252 Cf for a water moderated reactor are presented

A new neutron noise technique for fast reactors

International Nuclear Information System (INIS)

Zhuo Fengguan; Jin Manyi; Yao Shigui; Su Zhuting

1987-12-01

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

Ion source requirements for pulsed spallation neutron sources

International Nuclear Information System (INIS)

Alonso, J.R.

1996-01-01

The neutron scattering community has endorsed the need for a high-power (1 to 5 MW) accelerator-driven source of neutrons for materials research. Properly configured, the accelerator could produce very short (sub-microsecond) bursts of cold neutrons, said time structure offering advantages over the continuous flux from a reactor for a large class of experiments. The recent cancellation of the ANS reactor project has increased the urgency to develop a comprehensive strategy based on the best technological scenarios. Studies to date have built on the experience from ISIS (the 160 kW source in the UK), and call for a high-current (approx. 100 mA peak) H - source-linac combination injecting into one or more accumulator rings in which beam may be further accelerated. The 1 to 5 GeV proton beam is extracted in a single turn and brought to the target-moderator stations. The high current, high duty-factor, high brightness and high reliability required of the ion source present a very large challenge to the ion source community. A workshop held in Berkeley in October 1994, analyzed in detail the source requirements for proposed accelerator scenarios, the present performance capabilities of different H - source technologies, and identified necessary R ampersand D efforts to bridge the gap. copyright 1996 American Institute of Physics

Fission fragment driven neutron source

Science.gov (United States)

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

1976-01-01

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

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

Project 252Cf-D2O. The multisphere system of neutron dosimetry and spectrometry (M.S.-N.D.S.). Studies of applications to health physics

International Nuclear Information System (INIS)

Zaborowski, H.L.

1976-10-01

The project 252 Cf-D 2 O is articulated upon the utilization of a 200μg nominal 252 Cf spontaneous neutron fission source, used bare and under D 2 O spherical moderators, giving leakage neutron spectra experimentally known and/or calculated. This project has for objective the applications of those sources to Health Physics, in dosimetry (calibration of ''rad'' and ''rem-meters'') and in spectrometry, associated with the experimental system of measurements made by the generalization of the BONNER Spheres, known as ''the Multisphere System''. This communication describes the normalization method used and the results obtained leading to the adoption of a reference matrix called ''the Log-Normal Multisphere Matrix'' (LN-MM) giving the energies response functions of the generalized system for all the spheres diameters between 40 and 400 millimeters and for all the energies between 0.4eV and 15MeV [fr

Physical and biological dosimetries of Cf-252 radiation

International Nuclear Information System (INIS)

Yamashita, Hisao; Wada, Tadashi; Dokiya, Takushi; Hashimoto, Shozo

1986-01-01

Recently Cf-252 sources containing 300 μg have become available in a size identical to 1 Ci of Cs-137 and with the use of remotely controlled afterloading apparatus, safe therapy with little exposure to the therapist is now possible. Radiation leakage from the Cf-252 apparatus and from the treatment room was measured with REM-meter and it was possible to reduce the leakage from the treatment room to less than 1 mrem/h (gamma rays) and 0.5 mrem/h (neutrons). Measurement of fast neutrons was made with a twin chamber composed of a tissue equivalent ionization chamber and a carbon ionization chamber. The neutron dose in air and the absorbed dose in tissue equivalent water tank were measured, which showed that in air, neutrons were 70% and photons were 30% of dose. In water, greater distances from the source, neutrons attenuate and gamma rays increase in dose. The results of studies on the skin reaction of mice and sperm cleavage delay time of sea urchins indicated that the RBE ranges from 1.5 to 3.0 using the high dose rate system. Neutrons are remarkably affected by a time factor. With the use of high dose rate sources, the dose rate has become higher, but the overall time has been extended through dose fractionation and it was considered advisable to employ an RBE of 3-4 in these studies. (Auth.)

Activation analysis of stainless steel flux monitors using 252Cf neutron sources

International Nuclear Information System (INIS)

Williams, J.G.; Newton, T.H. Jr.; Cogburn, C.O.

1984-01-01

Activation analysis was performed on stainless steel beads from a chain which is used in reactor pressure vessel surveillance experiments at the Arkansas Power and Light Company reactors. The beads allow monitoring of two fast and three thermal neutron induced reactions: 58 Ni(n,p) 58 Co, 54 Fe(n,p) 54 Mn, 58 Fe(n,γ) 59 Fe, 59 Co(n,γ) 60 Co and 50 Cr(n,γ) 51 Cr. The analysis was performed using 12 beads from various positions along 5 different batches of chain and standard materials in an H 2 O moderator tank using two intense californium sources which had a total neutron emission rate of 3.97 x 10 10 /s. Semiconductor gamma spectrometers were used to count the products of the above reactions in the specimens. The percentage by weight of the iron, chromium and cobalt in the beads were found to be 62.1%, 20.2% and 0.120%, respectively. The excellent uniformity found in the bead compositions demonstrates the reproducibility of the experimental techniques and enhances considerably the value of the beads as neutron flux montitors

Is timing noise important in the gravitational wave detection of neutron stars?

International Nuclear Information System (INIS)

Jones, D.I.

2004-01-01

In this paper we ask whether the phenomenon of timing noise long known in electromagnetic pulsar astronomy is likely to be important in gravitational wave (GW) observations of spinning-down neutron stars. We find that timing noise is strong enough to be of importance only in the young pulsars, which must have larger triaxialities than theory predicts for their GW emission to be detectable. However, assuming that their GW emission is detectable, we list the pulsars for which timing noise is important, either because it is strong enough that its neglect by the observer would render the source undetectable or else because it is a measurable feature of the GW signal. We also find that timing noise places a limit on the observation duration of a coherent blind GW search, and suggest that hierarchical search techniques might be able to cope with this problem. Demonstration of the presence or absence of timing noise in the GW channel would give a new probe of neutron star physics

Calculation of the neutron importance and weighted neutron generation time using MCNIC method in accelerator driven subcritical reactors

Energy Technology Data Exchange (ETDEWEB)

Hassanzadeh, M. [Nuclear Science and Technology Research Institute, AEOI, Tehran, Islamic Republic of Iran (Iran, Islamic Republic of); Feghhi, S.A.H., E-mail: a_feghhi@sbu.ac.ir [Department of Radiation Application, Shahid Beheshti University, G.C., Tehran, Islamic Republic of Iran (Iran, Islamic Republic of); Khalafi, H. [Nuclear Science and Technology Research Institute, AEOI, Tehran, Islamic Republic of Iran (Iran, Islamic Republic of)

2013-09-15

Highlights: • All reactor kinetic parameters are importance weighted quantities. • MCNIC method has been developed for calculating neutron importance in ADSRs. • Mean generation time has been calculated in spallation driven systems. -- Abstract: The difference between non-weighted neutron generation time (Λ) and the weighted one (Λ{sup †}) can be quite significant depending on the type of the system. In the present work, we will focus on developing MCNIC method for calculation of the neutron importance (Φ{sup †}) and importance weighted neutron generation time (Λ{sup †}) in accelerator driven systems (ADS). Two hypothetic bare and graphite reflected spallation source driven system have been considered as illustrative examples for this means. The results of this method have been compared with those obtained by MCNPX code. According to the results, the relative difference between Λ and Λ{sup †} is within 36% and 24,840% in bare and reflected illustrative examples respectively. The difference is quite significant in reflected systems and increases with reflector thickness. In Conclusion, this method may be used for better estimation of kinetic parameters rather than the MCNPX code because of using neutron importance function.

Calculation of the neutron importance and weighted neutron generation time using MCNIC method in accelerator driven subcritical reactors

International Nuclear Information System (INIS)

Hassanzadeh, M.; Feghhi, S.A.H.; Khalafi, H.

2013-01-01

Highlights: • All reactor kinetic parameters are importance weighted quantities. • MCNIC method has been developed for calculating neutron importance in ADSRs. • Mean generation time has been calculated in spallation driven systems. -- Abstract: The difference between non-weighted neutron generation time (Λ) and the weighted one (Λ † ) can be quite significant depending on the type of the system. In the present work, we will focus on developing MCNIC method for calculation of the neutron importance (Φ † ) and importance weighted neutron generation time (Λ † ) in accelerator driven systems (ADS). Two hypothetic bare and graphite reflected spallation source driven system have been considered as illustrative examples for this means. The results of this method have been compared with those obtained by MCNPX code. According to the results, the relative difference between Λ and Λ † is within 36% and 24,840% in bare and reflected illustrative examples respectively. The difference is quite significant in reflected systems and increases with reflector thickness. In Conclusion, this method may be used for better estimation of kinetic parameters rather than the MCNPX code because of using neutron importance function

A survey of uses and users of 252Cf

International Nuclear Information System (INIS)

Bigelow, J.E.; Alexander, C.W.; King, L.J.; Knauer, J.B.; Notz, K.J.

1989-01-01

Californium-252, which is one of the transuranium-element isotopes being produced in the Radiochemical Engineering Development Center (REDC) at Oak Ridge National Laboratory (ORNL), has found many applications in service to industry and medicine. As a neutron source, 252 Cf is unique in providing a highly concentrated and extremely reliable neutron spectrum from a very small assembly. Over the past 22 years, 252 Cf has been applied with great success to cancer therapy, to neutron radiography of objects ranging from flowers to entire aircraft, to startup sources for nuclear reactors, to fission activation assay for quality control and safeguards of all commercial nuclear fuel, and to many other beneficial uses, some of which are now poised for further growth. The extensive exploitation of this highly specialized product has been made possible through 252 Sales/Loan programs sponsored by the US DOE Office of Nuclear Materials Production, initially at the Savannah River Laboratory and now at ORNL

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)

Status of the FRM-II hot neutron source

International Nuclear Information System (INIS)

Mueller, C.; Gutsmiedl, E.

2001-01-01

The new research reactor FRM-II will be equipped with a hot neutron source. This secondary source will shift a part of the thermal neutron energy spectrum in the D 2 O moderator to energies from 0.1 to 1 eV. The hot neutron source consists of a graphite cylinder (200 mm diameter, 300 mm high), which is heated by gamma radiation up to a maximum temperature of about 2400 C. The graphite cylinder is surrounded by a high-temperature insulation of carbon fiber, to achieve this high temperature. We have accomplished mock-up tests of the carbon fiber in a high temperature furnace, to investigate the insulation properties of the material. The graphite cylinder and the insulation are covered with two vessels made out of Zircaloy 4. The space between the vessels is filled with helium. The hot neutron source is permanent under control by pressure and temperature measurements. The temperature inside the graphite cylinder will be measured by a purpose-built noise thermometer due to the extremely harsh environment conditions (temperature and nuclear radiation). The hot neutron source is designed and manufactured according to the general specification basic safety and to the German nuclear atomic rules (KTA). The source will be installed in year 2001. (orig.)

Design and construction of an explosive detection system by Tna methods, using 252Cf radioisotope source

International Nuclear Information System (INIS)

Tavakkoli Farsouli, A.

1999-01-01

Bombs concealed in luggage have threatened human life and property throughout the world's traffic. The plastic explosives could not checked by the X-ray detecting device. Thermal Neutron Activation method has been tested in the present work for non-destructive detection of explosives. A radioisotope neutron source 252 Cf and two gamma spectroscopy systems have been used as a tool to find explosives, regardless of the bomb's shape and the packing materials. The MCNP code has been used to design the neutronic section of the system. The measured thermal neutron fluxes by the gold foils in some location of the system were in good agreement with those data obtained by the MCNP code. Also, detection limits for nitrogen in various counting times were measured. The measurements show that the system is capable to detect 417 gr of HMX explosive material (158 gr nitrogen) by 10 minutes of counting time. To modify the system and to decrease the detection limits some opinions are given

Neutron noise in nuclear reactors; Le bruit neutronique des reacteurs nucleaires

Energy Technology Data Exchange (ETDEWEB)

Blaquiere, A. [Institut National des Sciences et Techniques Nucleaires (France); Pachowska, R. [Universite Technique de Varsovie (Poland)

1961-06-15

The power of a nuclear reactor, in the operating conditions, presents fluctuations due to various causes. This random behaviour can be included in the study of 'noises'. Among other sources of noise, we analyse hereafter the fluctuations due: a) to the discontinuous emissions of neutrons from an independent source; b) to the multiplication of neutrons inside the reactor. The method which we present makes use of the analogies between the rules governing a nuclear reactor in operation and a number of radio-electrical systems, in particular the feed-back loops. The reactor can be characterized by its 'passing band' and is described as a system submitted to a sequence of random pulses. In non linear operating condition, the effect of neutron noise is defined by means of a non-linear functional, this theory is thus related to previous works the references of which are given at the end of the present report. This leads us in particular in the case of nuclear reactors to some results given by A. Blaquiere in the case of radio-electrical loops. (author) [French] La puissance d'un reacteur nucleaire, dans les conditions du regime, est affectee de fluctuations dont les causes sont tres diverses. Ce comportement aleatoire rentre dans le cadre general de l'etude des 'bruits'. Entre autres sources ce bruit, nous analysons ici les fluctuations dues: a) a l'emission discontinue des neutrons provenant d'une source autonome; b) a la multiplication des neutrons au sein du reacteur. La methode que nous introduisons exploite les analogies entre les lois qui regissent un reacteur nucleaire au regime et certains systemes radioelectriques, en particulier les circuits a boucle de reaction. Le reacteur est caracterise par sa 'bande passante' et est decrit comme un systeme soumis a une succession d'impulsions aleatoires. Dans les conditions de fonctionnement non lineaires, l'effet du bruit neutronique est precise en utilisant une fonctionnelle non lineaire, ce qui relie cette theorie a

Group Representation of the Prompt Fission Neutron Spectrum of {sup 252}Cf

Energy Technology Data Exchange (ETDEWEB)

Croft, S.; Miller, K. A. [Safeguards Science and Technology Group (N-1), Nuclear Nonproliferation Division, Los Alamos National Laboratory, Los Alamos(United States)

2011-12-15

We review the spectral representation used for the prompt fission neutron spectrum of 252Cf in the International Organization for Standardization document ISO 8529-1. We find corrections to Table A.2, the discrete group structure form, of this report are needed. We describe the approach to generating replacement values and provide a new tabulation.

Neutron radiography

International Nuclear Information System (INIS)

Hrdlicka, Z.

1977-01-01

Neutron radiography is a radiographic method using a neutron beam of a defined geometry. The neutron source usually consists of a research reactor, a specialized neutron radiography reactor or the 252 Cf radioisotope source. There are two types of the neutron radiography display system, viz., a system producing neutron radiography images by a photographic process or a system allowing a visual display, eg., using a television monitor. The method can be used wherever X-ray radiography is used except applications in the radiography of humans. The neutron radiography unit at UJV uses the WWR-S reactor as the neutron source and both types of the above mentioned display system. (J.P.)

Representation of neutron noise data using neural networks

International Nuclear Information System (INIS)

Korsah, K.; Damiano, B.; Wood, R.T.

1992-01-01

This paper describes a neural network-based method of representing neutron noise spectra using a model developed at the Oak Ridge National Laboratory (ORNL). The backpropagation neural network learned to represent neutron noise data in terms of four descriptors, and the network response matched calculated values to within 3.5 percent. These preliminary results are encouraging, and further research is directed towards the application of neural networks in a diagnostics system for the identification of the causes of changes in structural spectral resonances. This work is part of our current investigation of advanced technologies such as expert systems and neural networks for neutron noise data reduction, analysis, and interpretation. The objective is to improve the state-of-the-art of noise analysis as a diagnostic tool for nuclear power plants and other mechanical systems

Research program in reactor core diagnostics with neutron noise methods: Stage 3. Final report; Forskningsprogram angaaende haerddiagnostik med neutronbrusmetoder. Etapp 3. Slutrapport

Energy Technology Data Exchange (ETDEWEB)

Pazsit, I.; Garis, N.S.; Karlsson, J.; Racz, A. [Chalmers Univ. of Technology, Goeteborg (Sweden). Dept. of Reactor Physics

1997-09-01

Stage 3 of the program has been executed 96-04-12. The long term goal is to develop noise methods for identification and localization of perturbations in reactor cores. The main parts of the program consist of modelling the noise source, calculation of the space- and frequency dependent transfer function, calculation of the neutron noise via a convolution of the transfer function of the system and the noise source, i.e. the perturbation, and finally finding an inversion or unfolding procedure to determine noise source parameters from the neutron noise. Most previous work is based on very simple (analytical) reactor models for the calculation of the transfer function as well as analytical unfolding methods. The purpose of this project is to calculate the transfer function in a more realistic model as well as elaborating powerful inversion methods that do not require analytical transfer functions. The work in stage 3 is described under the following headlines: Further investigation of simplified models for the calculation of the neutron noise; Further investigation of methods based on neural networks; Further investigation of methods for detecting the vibrations and impacting of detectors; Application of static codes for determination of the neutron noise using the adiabatic approximation. 12 refs, 18 figs.

Comparison Of ²⁵²Cf Time Correlated Induced Fisssion With AmLi Induced Fission On Fresh MTR Research Reactor Fuel

Energy Technology Data Exchange (ETDEWEB)

Joshi, Jay Prakash [Los Alamos National Laboratory

2017-03-30

The effective application of international safeguards to research reactors requires verification of spent fuel as well as fresh fuel. To accomplish this goal various nondestructive and destructive assay techniques have been developed in the US and around the world. The Advanced Experimental Fuel Counter (AEFC) is a nondestructive assay (NDA) system developed at Los Alamos National Laboratory (LANL) combining both neutron and gamma measurement capabilities. Since spent fuel assemblies are stored in water, the system was designed to be watertight to facilitate underwater measurements by inspectors. The AEFC is comprised of six ³He detectors as well as a shielded and collimated ion chamber. The ³He detectors are used for active and passive neutron coincidence counting while the ion chamber is used for gross gamma counting. Active coincidence measurement data is used to measure residual fissile mass, whereas the passive coincidence measurement data along with passive gamma measurement can provide information about burnup, cooling time, and initial enrichment. In the past, most of the active interrogation systems along with the AEFC used an AmLi neutron interrogation source. Owing to the difficulty in obtaining an AmLi source, a ²⁵²Cf spontaneous fission (SF) source was used during a 2014 field trail in Uzbekistan as an alternative. In this study, experiments were performed to calibrate the AEFC instrument and compare use of the ²⁵²Cf spontaneous fission source and the AmLi (α,n) neutron emission source. The ²⁵²Cf source spontaneously emits bursts of time-correlated prompt fission neutrons that thermalize in the water and induce fission in the fuel assembly. The induced fission (IF) neutrons are also time correlated resulting in more correlated neutron detections inside the ³He detector, which helps reduce the statistical errors in doubles when using the ²⁵²Cf interrogation source instead of

Neutronics of pulsed spallation neutron sources

International Nuclear Information System (INIS)

Watanabe, Noboru

2003-01-01

Various topics and issues on the neutronics of pulsed spallation neutron sources, mainly for neutron scattering experiments, are reviewed to give a wide circle of readers a better understanding of these sources in order to achieve a high neutronic performance. Starting from what neutrons are needed, what the spallation reaction is and how to produce slow-neutrons more efficiently, the outline of the target and moderator neutronics are explained. Various efforts with some new concepts or ideas have already been devoted to obtaining the highest possible slow-neutron intensity with desired pulse characteristics. This paper also reviews the recent progress of such efforts, mainly focused on moderator neutronics, since moderators are the final devices of a neutron source, which determine the source performance. Various governing parameters for neutron-pulse characteristics such as material issues, geometrical parameters (shape and dimensions), the target-moderator coupling scheme, the ortho-para-hydrogen ratio, poisoning, etc are discussed, aiming at a high performance pulsed spallation source

Average cross sections for the 252Cf neutron spectrum

International Nuclear Information System (INIS)

Dezso, Z.; Csikai, J.

1977-01-01

A number of average cross sections have been measured for 252 Cf neutrons in (n, γ), (n,p), (n,2n), (n,α) reactions by the activation method and for fission by fission chamber. Cross sections have been determined for 19 elements and 45 reactions. The (n,γ) cross section values lie in the interval from 0.3 to 200 mb. The data as a function of target neutron number increases up to about N=60 with minimum near to dosed shells. The values lie between 0.3 mb and 113 mb. These cross sections decrease significantly with increasing the threshold energy. The values are below 20 mb. The data do not exceed 10 mb. Average (n,p) cross sections as a function of the threshold energy and average fission cross sections as a function of Zsup(4/3)/A are shown. The results obtained are summarized in tables

Bulk - Samples gamma-rays activation analysis (PGNAA) with Isotopic Neutron Sources

International Nuclear Information System (INIS)

HASSAN, A.M.

2009-01-01

An overview is given on research towards the Prompt Gamma-ray Neutron Activation Analysis (PGNAA) of bulk-samples. Some aspects in bulk-sample PGNAA are discussed, where irradiation by isotopic neutron sources is used mostly for in-situ or on-line analysis. The research was carried out in a comparative and/or qualitative way or by using a prior knowledge about the sample material. Sometimes we need to use the assumption that the mass fractions of all determined elements add up to 1. The sensitivity curves are also used for some elements in such complex samples, just to estimate the exact percentage concentration values. The uses of 252 Cf, 241 Arn/Be and 239 Pu/Be isotopic neutron sources for elemental investigation of: hematite, ilmenite, coal, petroleum, edible oils, phosphates and pollutant lake water samples have been mentioned.

Calibration of time of flight detectors using laser-driven neutron source

Energy Technology Data Exchange (ETDEWEB)

Mirfayzi, S. R.; Kar, S., E-mail: s.kar@qub.ac.uk; Ahmed, H.; Green, A.; Alejo, A.; Jung, D. [Centre for Plasma Physics, School of Mathematics and Physics, Queen’s University Belfast, Belfast BT7 1NN (United Kingdom); Krygier, A. G.; Freeman, R. R. [Department of Physics, The Ohio State University, Columbus, Ohio 43210 (United States); Clarke, R. [Central Laser Facility, Rutherford Appleton Laboratory, Didcot, Oxfordshire OX11 0QX (United Kingdom); Fuchs, J.; Vassura, L. [LULI, Ecole Polytechnique, CNRS, Route de Saclay, 91128 Palaiseau Cedex (France); Kleinschmidt, A.; Roth, M. [Institut für Kernphysik, Technische Universität Darmstadt, Schloßgartenstrasse 9, D-64289 Darmstadt,Germany (Germany); Morrison, J. T. [Propulsion Systems Directorate, Air Force Research Lab, Wright Patterson Air Force Base, Ohio 45433 (United States); Najmudin, Z.; Nakamura, H. [Blackett Laboratory, Department of Physics, Imperial College, London SW7 2AZ (United Kingdom); Norreys, P. [Central Laser Facility, Rutherford Appleton Laboratory, Didcot, Oxfordshire OX11 0QX (United Kingdom); Department of Physics, University of Oxford, Oxford OX1 3PU (United Kingdom); Oliver, M. [Department of Physics, University of Oxford, Oxford OX1 3PU (United Kingdom); Zepf, M. [Centre for Plasma Physics, School of Mathematics and Physics, Queen’s University Belfast, Belfast BT7 1NN (United Kingdom); Helmholtz Institut Jena, D-07743 Jena (Germany); Borghesi, M. [Centre for Plasma Physics, School of Mathematics and Physics, Queen’s University Belfast, Belfast BT7 1NN (United Kingdom); Institute of Physics of the ASCR, ELI-Beamlines Project, Na Slovance 2, 18221 Prague (Czech Republic)

2015-07-15

Calibration of three scintillators (EJ232Q, BC422Q, and EJ410) in a time-of-flight arrangement using a laser drive-neutron source is presented. The three plastic scintillator detectors were calibrated with gamma insensitive bubble detector spectrometers, which were absolutely calibrated over a wide range of neutron energies ranging from sub-MeV to 20 MeV. A typical set of data obtained simultaneously by the detectors is shown, measuring the neutron spectrum emitted from a petawatt laser irradiated thin foil.

Calibration of time of flight detectors using laser-driven neutron source

Science.gov (United States)

Mirfayzi, S. R.; Kar, S.; Ahmed, H.; Krygier, A. G.; Green, A.; Alejo, A.; Clarke, R.; Freeman, R. R.; Fuchs, J.; Jung, D.; Kleinschmidt, A.; Morrison, J. T.; Najmudin, Z.; Nakamura, H.; Norreys, P.; Oliver, M.; Roth, M.; Vassura, L.; Zepf, M.; Borghesi, M.

2015-07-01

Calibration of three scintillators (EJ232Q, BC422Q, and EJ410) in a time-of-flight arrangement using a laser drive-neutron source is presented. The three plastic scintillator detectors were calibrated with gamma insensitive bubble detector spectrometers, which were absolutely calibrated over a wide range of neutron energies ranging from sub-MeV to 20 MeV. A typical set of data obtained simultaneously by the detectors is shown, measuring the neutron spectrum emitted from a petawatt laser irradiated thin foil.

Calibration of time of flight detectors using laser-driven neutron source

International Nuclear Information System (INIS)

Mirfayzi, S. R.; Kar, S.; Ahmed, H.; Green, A.; Alejo, A.; Jung, D.; Krygier, A. G.; Freeman, R. R.; Clarke, R.; Fuchs, J.; Vassura, L.; Kleinschmidt, A.; Roth, M.; Morrison, J. T.; Najmudin, Z.; Nakamura, H.; Norreys, P.; Oliver, M.; Zepf, M.; Borghesi, M.

2015-01-01

Calibration of three scintillators (EJ232Q, BC422Q, and EJ410) in a time-of-flight arrangement using a laser drive-neutron source is presented. The three plastic scintillator detectors were calibrated with gamma insensitive bubble detector spectrometers, which were absolutely calibrated over a wide range of neutron energies ranging from sub-MeV to 20 MeV. A typical set of data obtained simultaneously by the detectors is shown, measuring the neutron spectrum emitted from a petawatt laser irradiated thin foil

NAA using Cf-252 after preconcentration

International Nuclear Information System (INIS)

Panyo, O.; Moebius, S.; Keller, C.

1988-01-01

Neutron activation analysis (NAA) with thermal neutron using Cf-252 sources was applied to elemental analysis of elements in water samples. A high-resolution Ge(Li) detector was employed for gamma-radiation detection. Both suspended particulate matter and liquid fraction were investigated after filtration. Preconcentration method by co-precipitation using iron (III) hydroxide and oxine were chosen for use. Elements which were considered to be able to detect in the present study are Al, As, Cl, K, Mg, Mn, Na, Sr, Ti, U, V and Zn

Noise-driven phenomena in hysteretic systems

CERN Document Server

Dimian, Mihai

2014-01-01

Noise-Driven Phenomena in Hysteretic Systems provides a general approach to nonlinear systems with hysteresis driven by noisy inputs, which leads to a unitary framework for the analysis of various stochastic aspects of hysteresis. This book includes integral, differential and algebraic models that are used to describe scalar and vector hysteretic nonlinearities originating from various areas of science and engineering. The universality of the authors approach is also reflected by the diversity of the models used to portray the input noise, from the classical Gaussian white noise to its impulsive forms, often encountered in economics and biological systems, and pink noise, ubiquitous in multi-stable electronic systems. The book is accompanied by HysterSoft© - a robust simulation environment designed to perform complex hysteresis modeling – that can be used by the reader to reproduce many of the results presented in the book as well as to research both disruptive and constructive effects of noise in hysteret...

Neutronics of pulsed spallation neutron sources

CERN Document Server

Watanabe, N

2003-01-01

Various topics and issues on the neutronics of pulsed spallation neutron sources, mainly for neutron scattering experiments, are reviewed to give a wide circle of readers a better understanding of these sources in order to achieve a high neutronic performance. Starting from what neutrons are needed, what the spallation reaction is and how to produce slow-neutrons more efficiently, the outline of the target and moderator neutronics are explained. Various efforts with some new concepts or ideas have already been devoted to obtaining the highest possible slow-neutron intensity with desired pulse characteristics. This paper also reviews the recent progress of such efforts, mainly focused on moderator neutronics, since moderators are the final devices of a neutron source, which determine the source performance. Various governing parameters for neutron-pulse characteristics such as material issues, geometrical parameters (shape and dimensions), the target-moderator coupling scheme, the ortho-para-hydrogen ratio, po...

Matrix effects correction on 252Cf shufflers by application of the alternating conditional expectation to neutron flux monitor data

International Nuclear Information System (INIS)

Pickrell, M.M.

1992-01-01

The 252 Cf shuffler assays fissile uranium and plutonium using active neutron interrogation and then counting the induced delayed neutrons. Using the shuffler, we conducted over 1700 assays of 55-gal. drums with 28 different matrices and several different fissionable materials. We measured the drums to diagnose the matrix and position effects on 252 Cf shuffler assays. The matrices incorporated metals, neutron poisons, and hydrogen in densities ranging from 0≤ pH ≤ 0.086 g/cm 3 , a range of cases more extreme than typically found in routine plant use. We used several neutron flux monitors during irradiation and kept statistics on the count rates of individual detector banks. The intent of these measurements was to gauge the effect of the matrix independently from the uranium assay

Research and development program in reactor diagnostics and monitoring with neutron noise methods. Stage 6. Final report

Energy Technology Data Exchange (ETDEWEB)

Pazsit, I.; Demaziere, C.; Avdic, S.; Dahl, B. [Chalmers Univ. of Technology, Goeteborg (Sweden). Dept. of Reactor Physics

2000-07-15

assumptions are fulfilled, because the temperature fluctuations are space dependent. Hence the reactivity effect of the driving source will be smaller than expected, and the induced neutron noise will also be space-dependent. In this Stage investigations were made for the case when the inlet temperature fluctuations are not spatially homogeneous, rather are random in space with a finite correlation length. A simple analytical model was set up to describe the spatially randomly distributed noise source. Then, first the space-dependent neutron noise was calculated and its deviation from point-kinetics was investigated. It was found that if the correlation length of the noise source is much shorter than the core size, appreciable deviations from the point kinetic behaviour occur. These alone however cannot explain the observed experimental error in the MTC measurements. Second, the fact that the reactivity effect of randomly distributed sources is smaller than that of the homogeneous perturbations was also accounted for. This was made first by calculating the reactivity effect alone, and then by calculating the space-dependent neutron fluctuations and the MTC in the model and comparing it with the traditional evaluation formula, to obtain the MTC. From this comparison the bias (error) of the traditional formula, due to the deviation of both the reactivity effect and the induced neutron noise from the expected homogeneous and point-kinetics, respectively, could be assessed. It was found that for a correlation length of the noise source that is comparable with the dimensions of a fuel assembly, the bias of the traditional evaluation formula is in the same range as observed in measurements.

System optimization for continuous on-stream elemental analysis using low-output isotopic neutron sources

International Nuclear Information System (INIS)

Rizk, R.A.M.

1989-01-01

In continuous on-stream neutron activation analysis, the material to be analyzed may be continuously recirculated in a closed loop system between an activation source and a shielded detector. In this paper an analytical formulation of the detector response for such a system is presented. This formulation should be useful in optimizing the system design parameters for specific applications. A study has been made of all parameters that influence the detector response during on-stream analysis. Feasibility applications of the method to solutions of manganese and vanadium using a 5 μg 252 Cf neutron source are demonstrated. (author)

Status of spallation neutron source program in High Intensity Proton Accelerator Project

International Nuclear Information System (INIS)

Oyama, Yukio

2001-01-01

Japan Atomic Energy Research Institute and High Energy Accelerator Organization are jointly designing a 1 MW spallation neutron source as one of the research facilities planned in the High Intensity Proton Accelerator Project. The spallation neutron source is driven by 3 GeV proton beam with a mercury target and liquid hydrogen moderators. The present status of design for these spallation source and relevant facility is overviewed. (author)

The continued development of the Spallation Neutron Source external antenna H- ion source

International Nuclear Information System (INIS)

Welton, R. F.; Carmichael, J.; Fuga, R.; Goulding, R. H.; Han, B.; Kang, Y.; Lee, S. W.; Murray, S. N.; Pennisi, T.; Potter, K. G.; Santana, M.; Stockli, M. P.; Desai, N. J.

2010-01-01

The U.S. Spallation Neutron Source (SNS) is an accelerator-based, pulsed neutron-scattering facility, currently in the process of ramping up neutron production. In order to ensure that the SNS will meet its operational commitments as well as provide for future facility upgrades with high reliability, we are developing a rf-driven, H - ion source based on a water-cooled, ceramic aluminum nitride (AlN) plasma chamber. To date, early versions of this source have delivered up to 42 mA to the SNS front end and unanalyzed beam currents up to ∼100 mA (60 Hz, 1 ms) to the ion source test stand. This source was operated on the SNS accelerator from February to April 2009 and produced ∼35 mA (beam current required by the ramp up plan) with availability of ∼97%. During this run several ion source failures identified reliability issues, which must be addressed before the source re-enters production: plasma ignition, antenna lifetime, magnet cooling, and cooling jacket integrity. This report discusses these issues, details proposed engineering solutions, and notes progress to date.

Core component vibration monitoring in BWRs using neutron noise

International Nuclear Information System (INIS)

Fry, D.N.; Robinson, J.C.; Kryter, R.C.; Cole, O.C.

1975-01-01

Neutron noise from in-core fission detectors in a BWR was investigated to determine its effectiveness as a monitor of mechanical vibrations of core components. In this study the general properties of BWR neutron noise were characterized, and a signal enhancement method was implemented to improve the measurement sensitivity. (auth)

A neutron irradiator applied to cancer treatment

International Nuclear Information System (INIS)

Campos, Tarcisio P.R.; Andrade, Ana P. de

2000-01-01

Cancer and the way of treating it with neutron capture therapy are addressed. This paper discusses also the type of neutron facilities used to treat cancer around the world, as follow: discrete neutron sources, accelerators, and nuclear reactors. The major features of an epithermal neutron irradiation facility applied to BNCT treatment are addressed. The main goal is to give another choice of neutron irradiators to be set in a hospital. The irradiation facility embeds a set of 252 Cf neutron source coupled with a homogeneous mixture of uranium-zirconium hydride alloy containing 8.4 wt % uranium enriched to 20% U 235 . The facility delivers an epithermal neutron beam with low background of fast neutron and gamma rays. The N particle transport code (MCNP-4A) has been used during the simulation in order to achieve the desired configurations and to estimate the multiplication factor, k eff . The present facility loaded with 30 mg of 252 Cf neutron source generates an external beam with an intensity of 10 7 n/cm 2 .s on the spectrum of 4 eV to 40 KeV. The 252 Cf - facility coupled with fissile material was able to amplify the epithermal flux to 10 8 n/cm 2 .s, maintaining the figure-of-merits represented by the ratios of the fast dose and gamma dose in air per epithermal neutron flux closed to those values presented by BMRR, MITR-II and Petten Reactor. The medical irradiation facility loaded with 252 Cf- 235 U can be a choice for BNCT. (author)

The Spallation Neutron Source RF Reference System

CERN Document Server

Piller, Maurice; Crofford, Mark; Doolittle, Lawrence; Ma, Hengjie

2005-01-01

The Spallation Neutron Source (SNS) RF Reference System includes the master oscillator (MO), local oscillator(LO) distribution, and Reference RF distribution systems. Coherent low noise Reference RF signals provide the ability to control the phase relationships between the fields in the front-end and linear accelerator (linac) RF cavity structures. The SNS RF Reference System requirements, implementation details, and performance are discussed.

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

Energy Technology Data Exchange (ETDEWEB)

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

1962-07-01

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

Optimization of in situ prompt gamma-ray analysis using a HPGe-252Cf probe

International Nuclear Information System (INIS)

Chien Chung; Jiunnhsing Chao

1991-01-01

Application of in situ measurements by the neutron-induced prompt gamma-ray activation analysis (PGAA) technique to geochemical analysis and mineral survey have been investigated. An in situ survey of water pollutants by PGAA techniques was first proposed in the authors' previous study, where a 2.7-μg 252 Cf neutron source used in connection with a gamma-ray detecting system to determine water pollutants was described. In this paper the authors describe a modified detection probe designed and constructed to look for the optimum conditions of various-intensity 252 Cf neutron sources in measurement of some elements in lake water. Detecting efficiencies at high-energy regions and detection limits for elements commonly found in polluted lakes were evaluated and predicted to investigate the potential application of the probe for in situ measurements

Information contraction and extraction by multivariate autoregressive (MAR) modelling. Pt. 2. Dominant noise sources in BWRS

International Nuclear Information System (INIS)

Morishima, N.

1996-01-01

The multivariate autoregressive (MAR) modeling of a vector noise process is discussed in terms of the estimation of dominant noise sources in BWRs. The discussion is based on a physical approach: a transfer function model on BWR core dynamics is utilized in developing a noise model; a set of input-output relations between three system variables and twelve different noise sources is obtained. By the least-square fitting of a theoretical PSD on neutron noise to an experimental one, four kinds of dominant noise sources are selected. It is shown that some of dominant noise sources consist of two or more different noise sources and have the spectral properties of being coloured and correlated with each other. By diagonalizing the PSD matrix for dominant noise sources, we may obtain an MAR expression for a vector noise process as a response to the diagonal elements(i.e. residual noises) being white and mutually-independent. (Author)

PHYCAA: Data-driven measurement and removal of physiological noise in BOLD fMRI

DEFF Research Database (Denmark)

Churchill, Nathan W.; Yourganov, Grigori; Spring, Robyn

2012-01-01

, autocorrelated physiological noise sources with reproducible spatial structure, using an adaptation of Canonical Correlation Analysis performed in a split-half resampling framework. The technique is able to identify physiological effects with vascular-linked spatial structure, and an intrinsic dimensionality...... with physiological noise, and real data-driven model prediction and reproducibility, for both block and event-related task designs. This is demonstrated compared to no physiological noise correction, and to the widely used RETROICOR (Glover et al., 2000) physiological denoising algorithm, which uses externally...

Time domain measurements for fast metal assemblies with /sup 252/Cf

Energy Technology Data Exchange (ETDEWEB)

Mihalczo, J T

1975-06-01

Time correlation measurements between the pulses from an ionization counter containing a /sup 252/Cf neutron source, which provided the initiators of fission chains in a neutron multiplying assembly and from sensors that detected particles from the fission chains are reviewed for fast uranium or plutonium metal assemblies. Comparisons are made between the correlated count rate from a /sup 252/Cf measurement and that from both one and two-detector Rossi-..cap alpha.. measurements. The assemblies studied were (1) unmoderated and polyethylene-moderated uranium (93 wt percent /sup 235/U) cylinders with masses from 12 to 160 kg; prompt neutron decay constants from 3 to 10/sup 3/ to 10/sup 8/ sec/sup -1/ and (2) unmoderated plutonium spheres and parts of spheres with plutonium masses from 2.2 to 16 kg with /sup 240/Pu contents of 4.5 to 20.1 at. percent. Measurements with a delayed critical uranium metal sphere determined the effective delayed neutron fraction and served as the basis for verification of the theory of the /sup 252/Cf measurement method in the time domain within a few per cent. (auth)

Neutron detection technique

International Nuclear Information System (INIS)

Oblath, N.S.; Poon, A.W.P.

2000-01-01

The Sudbury Neutrino Observatory (SNO) has the ability to measure the total flux of all active flavors of neutrinos using the neutral current reaction, whose signature is a neutron. By comparing the rates of the neutral current reaction to the charged current reaction, which only detects electron neutrinos, one can test the neutrino oscillation hypothesis independent of solar models. It is necessary to understand the neutron detection efficiency of the detector to make use of the neutral current reaction. This report demonstrates a coincidence technique to identify neutrons emitted from the 252 Cf neutron calibration source. The source releases on average four neutrons when a 252 Cf nucleus spontaneously fissions. Each neutron is detected as a separate event when the neutron is captured by a deuteron, releasing a gamma ray of approximately 6.25 MeV. This gamma ray is in turn detected by the photomultiplier tube (PMT) array. By investigating the time and spatial separation between neutron-like events, it is possible to obtain a pure sample of neutrons for calibration study. Preliminary results of the technique applied to two calibration runs are presented

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)

Hysteretic behavior of spin-crossover noise driven system

Energy Technology Data Exchange (ETDEWEB)

Gudyma, Iurii [Department of General Physics, Chernivtsi National University, Chernivtsi 58012 (Ukraine); Maksymov, Artur, E-mail: maxyartur@gmail.com [Department of General Physics, Chernivtsi National University, Chernivtsi 58012 (Ukraine); Advanced Materials Research Institute, University of New Orleans, LA 70148 (United States); Dimian, Mihai [Department of Electrical and Computer Engineering, Howard University, Washington DC, 20059 (United States); Faculty of Electrical Engineering and Computer Science, Stefan cel Mare University, Suceava 720229 (Romania)

2016-04-01

The influence of white Gaussian noise on hysteretic behavior of spin-crossover system is analyzed in the framework of stochastic Langevin dynamics. Various stochastic simulations are performed and several important properties of spin-transition in spin-crossover system driven by noise are reproduced. The numerical results are tested against the stationary probability function and the associated dynamic potential obtained from Fokker–Planck equation corresponding to spin-crossover Langevin dynamics. The dependence of light-induced optical hysteresis width and non-hysteretic transition curve slope on the noise intensity is illustrated. The role of low-spin and high-spin phase stabilities in the hysteretic behavior of noise-driven spin-crossover system is discussed.

Neutron source

International Nuclear Information System (INIS)

Cason, J.L. Jr.; Shaw, C.B.

1975-01-01

A neutron source which is particularly useful for neutron radiography consists of a vessel containing a moderating media of relatively low moderating ratio, a flux trap including a moderating media of relatively high moderating ratio at the center of the vessel, a shell of depleted uranium dioxide surrounding the moderating media of relatively high moderating ratio, a plurality of guide tubes each containing a movable source of neutrons surrounding the flux trap, a neutron shield surrounding one part of each guide tube, and at least one collimator extending from the flux trap to the exterior of the neutron source. The shell of depleted uranium dioxide has a window provided with depleted uranium dioxide shutters for each collimator. Reflectors are provided above and below the flux trap and on the guide tubes away from the flux trap

Current and noise in driven heterostructures

Energy Technology Data Exchange (ETDEWEB)

Kaiser, Franz

2009-02-18

In this thesis we consider the electron transport in nanoscale systems driven by an external energy source. We introduce a tight-binding Hamiltonian containing an interaction term that describes a very strong Coulomb repulsion between electrons in the system. Since we deal with time-dependent situations, we employ a Floquet theory to take into account the time periodicity induced by different external oscillating fields. For the two-level system, we even provide an analytical solution for the eigenenergies with arbitrary phase shift between the levels for a cosine-shaped driving. To describe time-dependent driven transport, we derive a master equation by tracing out the influence of the surrounding leads in order to obtain the reduced density operator of the system. We generalise the common master equation for the reduced density operator to perform an analysis of the noise characteristics. The concept of Full Counting Statistics in electron transport gained much attention in recent years proven its value as a powerful theoretical technique. Combining its advantages with the master equation approach, we find a hierarchy in the moments of the electron number in one lead that allows us to calculate the first two cumulants. The first cumulant can be identified as the current passing through the system, while the noise of this transmission process is reflected by the second cumulant. Moreover, in combination with our Floquet approach, the formalism is not limited to static situations, which we prove by calculating the current and noise characteristics for the non-adiabatic electron pump. We study the influence of a static energy disorder on the maximal possible current for different realisations. Further, we explore the possibility of non-adiabatically pumping electrons in an initially symmetric system if random fluctuations break this symmetry. Motivated by recent and upcoming experiments, we use our extended Floquet model to properly describe systems driven by

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

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

International Nuclear Information System (INIS)

2005-02-01

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

Spallation neutron sources

International Nuclear Information System (INIS)

Fraser, J.S.; Bartholomew, G.A.

1983-01-01

The principles and theory of spallation neutron sources are outlined and a comparison is given with other types of neutron source. A summary of the available accelerator types for spallation neutron sources and their advantages and disadvantages is presented. Suitable target materials are discussed for specific applications, and typical target assemblies shown. (U.K.)

Determination of neutron interaction effect and subcriticality for light water moderated UO2 lattices

International Nuclear Information System (INIS)

Miyoshi, Y.; Suzaki, T.; Kobayashi, I.

1984-01-01

From the view point of nuclear criticality safety for fuel storage, transport and processing, a series of critical experiments have been performed using a Tank-type Critical Assembly (TCA) at the Japan Atomic Energy Research Institute. The first series of experiments are concerned with the neutron interaction effects between two cores composed of BWR-type fuel rods in water. The reactivity contribution from one core to another have been measured by the water level worth method and a pulsed neutron source method. Two symmetrical rectangular cores were composed in TCA and the water gap between two cores were parametrically changed. The volume ratios of water to fuel are 1.83 and 2.48 of which lattice pitches are 1.96 cm and 2.15 cm respectively. As for the pulsed neutron experiment, Gozani's area ratio method is theoretically extended to a coupled-core system, and the applicability of this method has been studied for determination of the reactivity at a subcritical state and the coupling coefficient that represents reactivity contribution from one core to another. The object of the second series of experiment is development of the technique which determine the reactivity at a high sub-critical state. The CF-252 source driven neutron noise analysis method proposed by Mihalczo has been tested in order to examine whether it could be available for measuring the subcriticality for the light water moderated system. The tested core was water reflected annular type which consisted of 308 UO 2 fuel rods and had a void region at the core center

Use of californium-252 sources in Hungary for teaching and research

International Nuclear Information System (INIS)

Csikai, J.

1976-01-01

An activation facility was designed to accommodate up to 50 mg of 252 Cf; it contains at present a 500 μg source. The absolute values of thermal, epithermal and fast neutron fluxes were determined by the foil activation method using In, Dy, Au, Al and Fe detectors. Cross-sections averaged for unmoderated 252 Cf neutrons were determined for 22 different reactions for elements with atomic weights lying between A=27 and 204. The sensitivity for determination of Al, Ti, Cu, As, Sr, Mo, In, Cd, Ba, Au, Hg and Pb was calculated for NaI(Tl) and Ge(Li) detectors. Average (n,2n) cross-sections for 252 Cf spectrum were calculated for 49 nuclei lying between A=14 and 204. Angular distributions and cross-sections for the fragments from 252 Cf neutron-induced fission of 232 Th and 238 U were measured. Titanium in bauxite and manganese in aluminium alloys were determined with a 252 Cf source. The applicability of solid-state track detectors for neutron dosimetry, radiography and for the determination of fuel burn-up were investigated using 252 Cf neutron and fragment sources. Characteristics of a jumping spark counter for counting fission fragments were studied with 252 Cf sources. (author)

Beyond Californium-A Neutron Generator Alternative for Dosimetry and Instrument Calibration in the U.S.

Science.gov (United States)

Piper, Roman K; Mozhayev, Andrey V; Murphy, Mark K; Thompson, Alan K

2017-09-01

Evaluations of neutron survey instruments, area monitors, and personal dosimeters rely on reference neutron radiations, which have evolved from the heavy reliance on (α,n) sources to a shared reliance on (α,n) and the spontaneous fission neutrons of californium-252 (Cf). Capable of producing high dose equivalent rates from an almost point source geometry, the characteristics of Cf are generally more favorable when compared to the use of (α,n) and (γ,n) sources or reactor-produced reference neutron radiations. Californium-252 is typically used in two standardized configurations: unmoderated, to yield a fission energy spectrum; or with the capsule placed within a heavy-water moderating sphere to produce a softened spectrum that is generally considered more appropriate for evaluating devices used in nuclear power plant work environments. The U.S. Department of Energy Cf Loan/Lease Program, a longtime origin of affordable Cf sources for research, testing and calibration, was terminated in 2009. Since then, high-activity sources have become increasingly cost-prohibitive for laboratories that formerly benefited from that program. Neutron generators, based on the D-T and D-D fusion reactions, have become economically competitive with Cf and are recognized internationally as important calibration and test standards. Researchers from the National Institute of Standards and Technology and the Pacific Northwest National Laboratory are jointly considering the practicality and technical challenges of implementing neutron generators as calibration standards in the U.S. This article reviews the characteristics of isotope-based neutron sources, possible isotope alternatives to Cf, and the rationale behind the increasing favor of electronically generated neutron options. The evaluation of a D-T system at PNNL has revealed characteristics that must be considered in adapting generators to the task of calibration and testing where accurate determination of a dosimetric quantity is

Monte Carlo simulation of a TRIGA source driven core configuration: Preliminary results

International Nuclear Information System (INIS)

Burgio, N.; Ciavola, C.; Santagata, A.

2002-01-01

The different core configurations with a k eff ranging from 0.93 to 0.98, and their response when driven by a pulsed neutron source were simulated with MCNP4C3 (Los Alamos - Monte Carlo N Particles). Simulation results could be considered both as preliminary check for nuclear data and a conceptual design for 'source jerk' experiments on the frame of TRIGA Accelerator Driven Experiment (TRADE) on the reactor facility of Casaccia research center. (author)

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

Time-frequency feature analysis and recognition of fission neutrons signal based on support vector machine

International Nuclear Information System (INIS)

Jin Jing; Wei Biao; Feng Peng; Tang Yuelin; Zhou Mi

2010-01-01

Based on the interdependent relationship between fission neutrons ( 252 Cf) and fission chain ( 235 U system), the paper presents the time-frequency feature analysis and recognition in fission neutron signal based on support vector machine (SVM) through the analysis on signal characteristics and the measuring principle of the 252 Cf fission neutron signal. The time-frequency characteristics and energy features of the fission neutron signal are extracted by using wavelet decomposition and de-noising wavelet packet decomposition, and then applied to training and classification by means of support vector machine based on statistical learning theory. The results show that, it is effective to obtain features of nuclear signal via wavelet decomposition and de-noising wavelet packet decomposition, and the latter can reflect the internal characteristics of the fission neutron system better. With the training accomplished, the SVM classifier achieves an accuracy rate above 70%, overcoming the lack of training samples, and verifying the effectiveness of the algorithm. (authors)

Pulsed neutron sources for epithermal neutrons

International Nuclear Information System (INIS)

Windsor, C.G.

1978-01-01

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

Reactor Neutron Sources

International Nuclear Information System (INIS)

Aksenov, V.L.

1994-01-01

The present status and the prospects for development of reactor neutron sources for neutron scattering research in the world are considered. The fields of application of neutron scattering relative to synchrotron radiation, the creation stages of reactors (steady state and pulsed) and their position in comparison with spallation neutron sources at present and in the foreseen future are discussed. (author). 15 refs.; 8 figs.; 3 tabs

Measurement of β/Λ ratio in IEA-R1 reactor using noise technique

International Nuclear Information System (INIS)

Moreira, J.M.L.; Kassar, E.

1986-01-01

The ratio β/Λ for the IEA-R1 reactor is obtained experimentally through the noise analysis technique. This technique is based on the determination of the power spectral density of the reactor neutron population, with the reactor in a subcritical state driven by a 'white' neutron source. A ratio β/Λ of 43,5 s -1 is estimated from the break frequency of the measured transfer function of the IEA-R1 reactor. (Author) [pt

Validation of IRDFF in 252Cf Standard and IRDF-2002 Reference Neutron Fields

Directory of Open Access Journals (Sweden)

Simakov Stanislav

2016-01-01

Full Text Available The results of validation of the latest release of International Reactor Dosimetry and Fusion File, IRDFF-1.03, in the standard 252Cf(s.f. and reference 235U(nth,f neutron benchmark fields are presented. The spectrum-averaged cross sections were shown to confirm IRDFF-1.03 in the 252Cf standard spontaneous fission spectrum; that was not the case for the current recommended spectra for 235U(nth,f. IRDFF was also validated in the spectra of the research reactor facilities ISNF, Sigma-Sigma and YAYOI, which are available in the IRDF-2002 collection. The ISNF facility was re-simulated to remove unphysical oscillations in the spectrum. IRDFF-1.03 was shown to reproduce reasonably well the spectrum-averaged data measured in these fields except for the case of YAYOI.

Optimizing moderation of He-3 neutron detectors for shielded fission sources

Energy Technology Data Exchange (ETDEWEB)

Rees, Lawrence B., E-mail: Lawrence_Rees@byu.edu [Department of Physics and Astronomy, Brigham Young University, Provo, UT 84602 (United States); Czirr, J. Bart, E-mail: czirr@juno.com [Department of Physics and Astronomy, Brigham Young University, Provo, UT 84602 (United States)

2012-11-01

The response of a {sup 3}He neutron detector is highly dependent on the amount of moderator incorporated into the detector system. If there is too little moderation, neutrons will not react with the {sup 3}He. If there is too much moderation, neutrons will not reach the {sup 3}He. In applications for portal or border monitors where {sup 3}He detectors are used to interdict illicit importation of plutonium, the fission source is always shielded to some extent. Since the energy distribution of neutrons emitted from the source depends on the amount and type of shielding present, the optimum placement of moderating material around {sup 3}He tubes is a function of shielding. In this paper, we use Monte Carlo techniques to model the response of {sup 3}He tubes placed in polyethylene boxes for moderation. To model the shielded fission neutron source, we use a point {sup 252}Cf source placed in the center of polyethylene spheres of varying radius. Detector efficiency as a function of box geometry and shielding is explored. We find that increasing the amount of moderator behind and to the sides of the detector generally improves the detector response, but that incremental benefits are minimal if the thickness of the polyethylene moderator is greater than about 5-7 cm. The thickness of the moderator in front of the {sup 3}He tubes, however, is very important. For bare sources, about 4-5 cm of moderator is optimum, but as the shielding increases, the optimum thickness of this moderator decreases to 0.5-1 cm. Similar conclusions can be applied to polyethylene boxes employing two {sup 3}He tubes. Two-tube boxes with front moderators of non-uniform thickness may be useful for detecting neutrons over a wide energy range.

Neutron sources and applications

Energy Technology Data Exchange (ETDEWEB)

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

1994-01-01

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

Neutron sources and applications

International Nuclear Information System (INIS)

Price, D.L.; Rush, J.J.

1994-01-01

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

Reliability of Monte Carlo simulations in modeling neutron yields from a shielded fission source

Energy Technology Data Exchange (ETDEWEB)

McArthur, Matthew S., E-mail: matthew.s.mcarthur@gmail.com; Rees, Lawrence B., E-mail: Lawrence_Rees@byu.edu; Czirr, J. Bart, E-mail: czirr@juno.com

2016-08-11

Using the combination of a neutron-sensitive {sup 6}Li glass scintillator detector with a neutron-insensitive {sup 7}Li glass scintillator detector, we are able to make an accurate measurement of the capture rate of fission neutrons on {sup 6}Li. We used this detector with a {sup 252}Cf neutron source to measure the effects of both non-borated polyethylene and 5% borated polyethylene shielding on detection rates over a range of shielding thicknesses. Both of these measurements were compared with MCNP calculations to determine how well the calculations reproduced the measurements. When the source is highly shielded, the number of interactions experienced by each neutron prior to arriving at the detector is large, so it is important to compare Monte Carlo modeling with actual experimental measurements. MCNP reproduces the data fairly well, but it does generally underestimate detector efficiency both with and without polyethylene shielding. For non-borated polyethylene it underestimates the measured value by an average of 8%. This increases to an average of 11% for borated polyethylene.

Spectrum shaping assessment of accelerator-based fusion neutron sources to be used in BNCT treatment

Science.gov (United States)

Cerullo, N.; Esposito, J.; Daquino, G. G.

2004-01-01

Monte Carlo modelling of an irradiation facility, for boron neutron capture therapy (BNCT) application, using a set of advanced type, accelerator based, 3H(d,n) 4He (D-T) fusion neutron source device is presented. Some general issues concerning the design of a proper irradiation beam shaping assembly, based on very hard energy neutron source spectrum, are reviewed. The facility here proposed, which represents an interesting solution compared to the much more investigated Li or Be based accelerator driven neutron source could fulfil all the medical and safety requirements to be used by an hospital environment.

Intense fusion neutron sources

International Nuclear Information System (INIS)

Kuteev, B. V.; Goncharov, P. R.; Sergeev, V. Yu.; Khripunov, V. I.

2010-01-01

The review describes physical principles underlying efficient production of free neutrons, up-to-date possibilities and prospects of creating fission and fusion neutron sources with intensities of 10 15 -10 21 neutrons/s, and schemes of production and application of neutrons in fusion-fission hybrid systems. The physical processes and parameters of high-temperature plasmas are considered at which optimal conditions for producing the largest number of fusion neutrons in systems with magnetic and inertial plasma confinement are achieved. The proposed plasma methods for neutron production are compared with other methods based on fusion reactions in nonplasma media, fission reactions, spallation, and muon catalysis. At present, intense neutron fluxes are mainly used in nanotechnology, biotechnology, material science, and military and fundamental research. In the near future (10-20 years), it will be possible to apply high-power neutron sources in fusion-fission hybrid systems for producing hydrogen, electric power, and technological heat, as well as for manufacturing synthetic nuclear fuel and closing the nuclear fuel cycle. Neutron sources with intensities approaching 10 20 neutrons/s may radically change the structure of power industry and considerably influence the fundamental and applied science and innovation technologies. Along with utilizing the energy produced in fusion reactions, the achievement of such high neutron intensities may stimulate wide application of subcritical fast nuclear reactors controlled by neutron sources. Superpower neutron sources will allow one to solve many problems of neutron diagnostics, monitor nano-and biological objects, and carry out radiation testing and modification of volumetric properties of materials at the industrial level. Such sources will considerably (up to 100 times) improve the accuracy of neutron physics experiments and will provide a better understanding of the structure of matter, including that of the neutron itself.

Intense fusion neutron sources

Science.gov (United States)

Kuteev, B. V.; Goncharov, P. R.; Sergeev, V. Yu.; Khripunov, V. I.

2010-04-01

The review describes physical principles underlying efficient production of free neutrons, up-to-date possibilities and prospects of creating fission and fusion neutron sources with intensities of 1015-1021 neutrons/s, and schemes of production and application of neutrons in fusion-fission hybrid systems. The physical processes and parameters of high-temperature plasmas are considered at which optimal conditions for producing the largest number of fusion neutrons in systems with magnetic and inertial plasma confinement are achieved. The proposed plasma methods for neutron production are compared with other methods based on fusion reactions in nonplasma media, fission reactions, spallation, and muon catalysis. At present, intense neutron fluxes are mainly used in nanotechnology, biotechnology, material science, and military and fundamental research. In the near future (10-20 years), it will be possible to apply high-power neutron sources in fusion-fission hybrid systems for producing hydrogen, electric power, and technological heat, as well as for manufacturing synthetic nuclear fuel and closing the nuclear fuel cycle. Neutron sources with intensities approaching 1020 neutrons/s may radically change the structure of power industry and considerably influence the fundamental and applied science and innovation technologies. Along with utilizing the energy produced in fusion reactions, the achievement of such high neutron intensities may stimulate wide application of subcritical fast nuclear reactors controlled by neutron sources. Superpower neutron sources will allow one to solve many problems of neutron diagnostics, monitor nano-and biological objects, and carry out radiation testing and modification of volumetric properties of materials at the industrial level. Such sources will considerably (up to 100 times) improve the accuracy of neutron physics experiments and will provide a better understanding of the structure of matter, including that of the neutron itself.

Accelerator based continuous neutron source.

CERN Document Server

Shapiro, S M; Ruggiero, A G

2003-01-01

Until the last decade, most neutron experiments have been performed at steady-state, reactor-based sources. Recently, however, pulsed spallation sources have been shown to be very useful in a wide range of neutron studies. A major review of neutron sources in the US was conducted by a committee chaired by Nobel laureate Prof. W. Kohn: ''Neutron Sources for America's Future-BESAC Panel on Neutron Sources 1/93''. This distinguished panel concluded that steady state and pulsed sources are complementary and that the nation has need for both to maintain a balanced neutron research program. The report recommended that both a new reactor and a spallation source be built. This complementarity is recognized worldwide. The conclusion of this report is that a new continuous neutron source is needed for the second decade of the 20 year plan to replace aging US research reactors and close the US neutron gap. it is based on spallation production of neutrons using a high power continuous superconducting linac to generate pr...

Modelling of an experiment for the study of neutron spallation source at JINR

International Nuclear Information System (INIS)

Kumawat, Harphool; Goyal, Uttam; Kumar, V.; Barashenkov, V.S.

2002-01-01

Intense neutron spallation source (INSS) is a necessary requirement of accelerator driven sub-critical systems. INSS are proposed to be generated using the high current proton beams. Some studies are conducted for the neutron flux, transmutation rates and energy gains and a larger number of related experiments are being planned

{sup 252}Cf spontaneous prompt fission neutron spectrum measured at 0 degree and 180 degree relative to the fragment motion

Energy Technology Data Exchange (ETDEWEB)

Shanglian, Bao; Jinquan, Liu [Beijing Univ., BJ (China); Batenkov, O I; Blinov, M V; Smirnov, S N [V.G. Khlopin Radium Institute, ST. Petersburg (Russian Federation)

1994-09-01

The {sup 252}Cf spontaneous prompt fission neutron spectrum at 0 degree and 180 degree relative to the motion direction of corresponding fission fragments was measured. High angular resolution for fragment measurements and high energy resolution for neutron measurements were obtained using multi-parameter TOF spectrometer. The results showed that there is a symmetric distribution of `forward` and `backward` for low energy in C.M.S. neutrons, which was an evidence of nonequilibrium neutrons existed in fission process.

Development of a hydrogen analysis using a small neutron source

International Nuclear Information System (INIS)

Ishikawa, I.; Tachikawa, N.; Tominaga, H.

1998-01-01

Most of industrial nuclear gauges are based on the use of radiation transmission through matter. This document presents new techniques to measure hydrogen using a small neutron source. A new technique has been developed for measuring the thickness of a thin layer of 30-200 μm thick plastic, which is sandwiched between two sheets of 0.6-4.2 mm in total thickness. Another technique allows to monitor residual moisture in wet refractory newly coated on the inner surface of a steel vessel from its outside through a thick steel plate. For saving on the use of coke and for strict control of furnace heating in the iron making process a new type moisture gauge was developed using simultaneous measurement of transmission rates of both fast neutrons and gamma rays from 252 Cf

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

Energy Technology Data Exchange (ETDEWEB)

Fhager, V

2000-01-01

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

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

International Nuclear Information System (INIS)

Fhager, V.

2000-01-01

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

Lunar neutron source function

International Nuclear Information System (INIS)

Kornblum, J.J.

1974-01-01

The search for a quantitative neutron source function for the lunar surface region is justified because it contributes to our understanding of the history of the lunar surface and of nuclear process occurring on the moon since its formation. A knowledge of the neutron source function and neutron flux distribution is important for the interpretation of many experimental measurements. This dissertation uses the available pertinent experimental measurements together with theoretical calculations to obtain an estimate of the lunar neutron source function below 15 MeV. Based upon reasonable assumptions a lunar neutron source function having adjustable parameters is assumed for neutrons below 15 MeV. The lunar neutron source function is composed of several components resulting from the action of cosmic rays with lunar material. A comparison with previous neutron calculations is made and significant differences are discussed. Application of the results to the problem of lunar soil histories is examined using the statistical model for soil development proposed by Fireman. The conclusion is drawn that the moon is losing mass

Properties of neutron sources

International Nuclear Information System (INIS)

1987-03-01

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

Transport, shot noise, and topology in AC-driven dimer arrays

Science.gov (United States)

Niklas, Michael; Benito, Mónica; Kohler, Sigmund; Platero, Gloria

2016-11-01

We analyze an AC-driven dimer chain connected to a strongly biased electron source and drain. It turns out that the resulting transport exhibits fingerprints of topology. They are particularly visible in the driving-induced current suppression and the Fano factor. Thus, shot noise measurements provide a topological phase diagram as a function of the driving parameters. The observed phenomena can be explained physically by a mapping to an effective time-independent Hamiltonian and the emergence of edge states. Moreover, by considering quantum dissipation, we determine the requirements for the coherence properties in a possible experimental realization. For the computation of the zero-frequency noise, we develop an efficient method based on matrix-continued fractions.

SU-G-IeP4-04: DD-Neutron Source Collimation for Neutron Stimulated Emission Computed Tomography: A Monte Carlo Simulation Study

Energy Technology Data Exchange (ETDEWEB)

Fong, G; Kapadia, A [Carl E Ravin Advanced Imaging Laboratories, Durham, North Carolina (United States)

2016-06-15

Purpose: To optimize collimation and shielding for a deuterium-deuterium (DD) neutron generator for an inexpensive and compact clinical neutron imaging system. The envisioned application is cancer diagnosis through Neutron Stimulated Emission Computed Tomography (NSECT). Methods: Collimator designs were tested with an isotropic 2.5 MeV neutron source through GEANT4 simulations. The collimator is a 52×52×52 cm{sup 3} polyethylene block coupled with a 1 cm lead sheet in sequence. Composite opening was modeled into the collimator to permit passage of neutrons. The opening varied in shape (cylindrical vs. tapered), size (1–5 cm source-side and target-side openings) and aperture placements (13–39 cm from source-side). Spatial and energy distribution of neutrons and gammas were tracked from each collimator design. Parameters analyzed were primary beam width (FWHM), divergence, and efficiency (percent transmission) for different configurations of the collimator. Select resultant outputs were then used for simulated NSECT imaging of a virtual breast phantom containing a 2.5 cm diameter tumor to assess the effect of the collimator on spatial resolution, noise, and scan time. Finally, composite shielding enclosure made of polyethylene and lead was designed and evaluated to block 99.99% of neutron and gamma radiation generated in the system. Results: Analysis of primary beam indicated the beam-width is linear to the aperture size. Increasing source-side opening allowed at least 20% more neutron throughput for all designs relative to the cylindrical openings. Maximum throughput for all designs was 364% relative to cylindrical openings. Conclusion: The work indicates potential for collimating and shielding a DD neutron generator for use in a clinical NSECT system. The proposed collimator designs produced a well-defined collimated neutron beam that can be used to image samples of interest with millimeter resolution. Balance in output efficiency, noise reduction, and scan

SU-G-IeP4-04: DD-Neutron Source Collimation for Neutron Stimulated Emission Computed Tomography: A Monte Carlo Simulation Study

International Nuclear Information System (INIS)

Fong, G; Kapadia, A

2016-01-01

Purpose: To optimize collimation and shielding for a deuterium-deuterium (DD) neutron generator for an inexpensive and compact clinical neutron imaging system. The envisioned application is cancer diagnosis through Neutron Stimulated Emission Computed Tomography (NSECT). Methods: Collimator designs were tested with an isotropic 2.5 MeV neutron source through GEANT4 simulations. The collimator is a 52×52×52 cm"3 polyethylene block coupled with a 1 cm lead sheet in sequence. Composite opening was modeled into the collimator to permit passage of neutrons. The opening varied in shape (cylindrical vs. tapered), size (1–5 cm source-side and target-side openings) and aperture placements (13–39 cm from source-side). Spatial and energy distribution of neutrons and gammas were tracked from each collimator design. Parameters analyzed were primary beam width (FWHM), divergence, and efficiency (percent transmission) for different configurations of the collimator. Select resultant outputs were then used for simulated NSECT imaging of a virtual breast phantom containing a 2.5 cm diameter tumor to assess the effect of the collimator on spatial resolution, noise, and scan time. Finally, composite shielding enclosure made of polyethylene and lead was designed and evaluated to block 99.99% of neutron and gamma radiation generated in the system. Results: Analysis of primary beam indicated the beam-width is linear to the aperture size. Increasing source-side opening allowed at least 20% more neutron throughput for all designs relative to the cylindrical openings. Maximum throughput for all designs was 364% relative to cylindrical openings. Conclusion: The work indicates potential for collimating and shielding a DD neutron generator for use in a clinical NSECT system. The proposed collimator designs produced a well-defined collimated neutron beam that can be used to image samples of interest with millimeter resolution. Balance in output efficiency, noise reduction, and scan time

Development of moderated neutron calibration fields simulating workplaces of MOX fuel facilities

International Nuclear Information System (INIS)

Tsujimura, Norio; Yoshida, Tadayoshi; Takada, Chie

2005-01-01

It is important for the MOX fuel facilities to control neutrons produced by the spontaneous fission of plutonium isotopes and those from (α,n) reactions between 18 O and α particles emitted by 238 Pu. Neutron dose meters should be calibrated for measuring these neutrons. We have developed moderated-neutron calibration fields employing a 252 Cf neutron source and moderators mainly for the characteristics evaluation and the calibration of neutron detectors used in MOX fuel facilities. Neutron energy spectrum can be adjusted by changing the position of the 252 Cf neutron source and combining different moderators to simulate the neutron field of the MOX fuel facility. This performance is realized owing to using an existing neutron irradiation room. (K. Yoshida)

Spallation Neutron Source (SNS)

Data.gov (United States)

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

Effect of double false pulses in calibrated neutron coincidence collar during measuring time-correlated neutrons from PuBe neutron sources

Energy Technology Data Exchange (ETDEWEB)

Nguyen, Tam Cong, E-mail: tam.nguyen.cong@energia.mta.hu; Huszti, Jozsef; Nguyen, Quan Van

2015-09-01

Effect of double false pulses of preamplifiers in neutron coincidence collar was investigated to explain non-parallel shape of calibrated D/S–M{sub Pu} curves of two commercial neutron coincidence collars, JCC-31 and JCC-13. Two curves, which were constructed from D/S ratio (doubles and singles count rate), and Pu content M{sub Pu}, of the same set of secondary standard PuBe neutron sources, should be parallel. Non-parallelism rises doubt about usability of the method based on this curve for determination of Pu content in PuBe neutron sources. We have shown in three steps that the problem originates from double false pulses of preamplifiers in JCC-13. First we used a pulse train diagram for analyzing the non-parallel shape, second we used Rossi-Alpha distribution measured by pulse train recorder developed in our institute and finally, we investigated the effect of inserted noise pulses. This implies a new type of QA test option in traditional multiplicity shift registers for excluding presence of double false pulses.

Neutron source strength determination for on-line reactivity measurements

Energy Technology Data Exchange (ETDEWEB)

Hoogenboom, J.E.; Sluijs, A.R. van der

1988-01-01

A method is described to determine the effective neutron source strength in a nuclear reactor, which must be known when calculating the time-varying reactivity from inverse reactor kinetics for a reactor at low power. When for an initially subcritical reactor the reactivity is changed and kept constant after the change, the effective source strength can be determined from a linear regression of reactor power to a function proportional to the emission rate of delayed neutrons, which can be calculated from the reactor power history. In view of the relatively strong noise present in the reactor power signal at low power, a grouping method for the regression is preferred over the least-squares method. Experiments with a reactor simulator with known source strength showed good agreement. Application to actual reactor signals gave consistent and satisfactory results.

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

Spallation neutrons pulsed sources

International Nuclear Information System (INIS)

Carpenter, J.

1996-01-01

This article describes the range of scientific applications which can use these pulsed neutrons sources: Studies on super fluids, measures to verify the crawling model for the polymers diffusion; these sources are also useful to study the neutron disintegration, the ultra cold neutrons. In certain applications which were not accessible by neutrons diffusion, for example, radiations damages, radionuclides production and activation analysis, the spallation sources find their use and their improvement will bring new possibilities. Among others contributions, one must notice the place at disposal of pulsed muons sources and neutrinos sources. (N.C.). 3 figs

Gamma-ray emission spectra from spheres with 14 MeV neutron source

International Nuclear Information System (INIS)

Yamamoto, Junji; Kanaoka, Takeshi; Murata, Isao; Takahashi, Akito; Sumita, Kenji

1989-01-01

Energy spectra of neutron-induced gamma-rays emitted from spherical samples were measured using a 14 MeV neutron source. The samples in use were LiF, Teflon:(CF 2 ) n , Si, Cr, Mn, Co, Cu, Nb, Mo, W and Pb. A diameter of the sphere was either 40 or 60 cm. The gamma-ray energy in the emission spectra covered the range from 500 keV to 10 MeV. Measured spectra were compared with transport calculations using the nuclear data files of JENDL-3T and ENDF/B-IV. The agreements between the measurements and the JENDL-3T calculations were good in the emission spectra for the low energy gamma-rays from inelastic scattering. (author)

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)

A proton-driven, intense, subcritical, fission neutron source for radioisotope production

Energy Technology Data Exchange (ETDEWEB)

Jongen, Y. [Chemin du Cyclotron, Louvain-la-Neuve (Belgium)

1995-10-01

{sup 99m}Tc, the most frequently used radioisotope in nuclear medicine, is distributed as {sup 99}Mo=>{sup 99m}Tc generators. {sup 99}Mo is a fission product of {sup 235}U. To replace the aging nuclear reactors used today for this production, the author proposes to use a spallation neutron source, with neutron multiplication by fission. A 150 MeV, H{sup {minus}} cyclotron can produce a 225 kW proton beam with 50% total system energy efficiency. The proton beam would hit a molten lead target, surrounded by a water moderator and a graphite reflector, producing around 0.96 primary neutron per proton. The primary spallation neutrons, moderated, would strike secondary targets containing a subcritical amount of {sup 235}U. The assembly would show a k{sub eff} of 0.8, yielding a fivefold neutron multiplication. The thermal neutron flux at the targets location would be 2 {times} 10{sup 14} n/cm{sup 2}.s, resulting in a fission power of 500 to 750 kW. One such system could supply the world demand in {sup 99}Mo, as well as other radioisotopes. Preliminary indications show that the cost would be lower than the cost of a commercial 10 MW isotope production reactor. The cost of operation, of disposal of radiowaste and of decommissioning should be significantly lower as well. Finally, the non-critical nature of the system would make it more acceptable for the public than a nuclear reactor and should simplify the licensing process.

Measurement and analysis of self-noise in hybrid-driven underwater gliders

Directory of Open Access Journals (Sweden)

LIU Lu

2017-08-01

Full Text Available The Hybrid-driven Underwater Glider (HUG is a new type of submersible vehicle which combines the functions of traditional Autonomous Underwater Vehicles(AUVand Autonomous Underwater Gliders(AUG. In order to study its noise source distribution and basic self-noise characteristics, a self-noise acquisition system based on the HUG was designed and developed, and a noise analysis test carried out in a free-field pool. In August 2016, the sea trial of the Petrel II glider was conducted in the South China Sea, with observation data at a depth range of 1 000 m as the research object. The self-noise data of the glider platform under different working conditions was obtained through the step-by-step operation method. The experimental analysis and results show that the self-noise acquisition system is stable. The contribution of mechanical noise to self-noise is greatest when the glider works in the gliding mode, while the self-noise band above 500 Hz is closely related to the work of the buoyancy adjustment unit, and peaks at 1 kHz. According to the analysis of the basic characteristics of self-noise, this provides some guidance for the implementation of vibration and noise reduction.

Neutron spectra produced by moderating an isotopic neutron source

International Nuclear Information System (INIS)

Carrillo Nunnez, Aureliano; Vega Carrillo, Hector Rene

2001-01-01

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

ADS Neutronic Benchmark A New Approach to the Design of Accelerator Driven Systems

International Nuclear Information System (INIS)

Carminati, F.; Kadi, Y.

1997-01-01

The main parameter characterizing the neutron economy of an accelerator driven subcritical fission device, like the Energy Amplifier (EA), is the factor M by which the source spallation neutrons are multiplied by the fission dominated cascade. A related quantity is the multiplication coefficient K s rc=(M-1)/M, that is the average ratio of the neutron population in two subsequent generations of the source-initiated cascade. Such a factor k s rc, depending on both the properties of the source and of the medium, is in general conceptual and numerically different from the effective criticality factor k e ff, commonly used in reactor theory, which is in fact only relevant to the fundamental mode of the neutron flux distribution, and is independent on the source. The effective criticality factor k e ff is however a meaningful measure of the actual safety characteristics of the device, that is 1-k e ff is a proper gauge of the distance from criticality. In this paper the difference between k e ff is addressed numerically in the case of an externally driven Thorium fuelled and Lead cooled subcritical device representing a simplified version of the Energy Amplifier. It is found that codes or calculations implementing the critical reactor formalism (neutrons are distributed according to a cos-type imposed distribution together with a fission spectrum energy distribution and non-fission multiplication, i. e. n,X n reactions, is not considered explicitly) in order to describe a subcritical device, systematically underestimate the reactivity on the system by about 0.028 in k (∼ 2800 pcm) which implies an error in the estimation of the necessary concentration of ''233U close to 5% which in turn induces an adverse effect on the stability of k during burnup. Finally, the discrepancies arising from the use of different nuclear data libraries are as significant as the effects of using different neutron source approximations and hence also deserve attention. We think that a

Physical model study of neutron noise induced by vibration of reactor internals

International Nuclear Information System (INIS)

Liu Jinhui; Gu Fangyu

1999-01-01

The author presents a physical model of neutron noise induced by reactor internals vibration in frequency domain. Based on system control theory, the reactor dynamic equations are coupled with random vibration equation, and non-linear terms are also taken into accounted while treating the random vibration. Experiments carried out on a zero-power reactor show that the model can be used to describe dynamic character of neutron noise induced by internals' vibration. The model establishes a method to help to determine internals'vibration features, and to diagnosis anomalies through neutron noise

Detector for imaging and dosimetry of laser-driven epithermal neutrons by alpha conversion

Science.gov (United States)

Mirfayzi, S. R.; Alejo, A.; Ahmed, H.; Wilson, L. A.; Ansell, S.; Armstrong, C.; Butler, N. M. H.; Clarke, R. J.; Higginson, A.; Notley, M.; Raspino, D.; Rusby, D. R.; Borghesi, M.; Rhodes, N. J.; McKenna, P.; Neely, D.; Brenner, C. M.; Kar, S.

2016-10-01

An epithermal neutron imager based on detecting alpha particles created via boron neutron capture mechanism is discussed. The diagnostic mainly consists of a mm thick Boron Nitride (BN) sheet (as an alpha converter) in contact with a non-borated cellulose nitride film (LR115 type-II) detector. While the BN absorbs the neutrons in the thermal and epithermal ranges, the fast neutrons register insignificantly on the detector due to their low neutron capture and recoil cross-sections. The use of solid-state nuclear track detectors (SSNTD), unlike image plates, micro-channel plates and scintillators, provide safeguard from the x-rays, gamma-rays and electrons. The diagnostic was tested on a proof-of-principle basis, in front of a laser driven source of moderated neutrons, which suggests the potential of using this diagnostic (BN+SSNTD) for dosimetry and imaging applications.

High-current negative-ion sources for pulsed spallation neutron sources: LBNL workshop, October 1994

International Nuclear Information System (INIS)

Alonso, J.R.

1995-09-01

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

Undergraduate experiments using the neutron radiation from californium-252

International Nuclear Information System (INIS)

Rossel, J.; Golecki, I.

1976-01-01

Three experiments designed to demonstrate and measure several properties of the neutron radiation emitted by a 3μg 252 Cf source are described. The experiments constitute a special project carried out by a third-year undergraduate student at the Institute of Physics of the University of Neuchatel. The 252 Cf source is enclosed in a shield which allows a pencil of fast neutrons to pass through a central tube, while reducing the ambient radiation below the tolerance level. The shield consists of layers of borated paraffin wax, iron and cadmium. The first experiment uses an air-alcohol diffusion cloud chamber for the demonstration of tracks of recoil protons produced by the neutrons. Semi-quantitative measurements of track lengths give the correct order of magnitude of the proton energies. In the second experiment a liquid scintillator detector is used to scan the beam profile across the radiation shield enclosing the source. A pulse-shape-discrimination system discriminates between neutrons and gamma photons. The third experiment makes use of the nuclear emulsion technique to study the neutron energy distribution of 252 Cf. Preliminary results are compared with published values. (author)

The secondary neutron sources for generation of particular neutron fluxes

International Nuclear Information System (INIS)

Tracz, G.

2007-07-01

The foregoing paper presents the doctor's thesis entitled '' The secondary neutron sources for generation of particular neutron fluxes ''. Two secondary neutron sources have been designed, which exploit already existing primary sources emitting neutrons of energies different from the desired ones. The first source is devoted to boron-neutron capture therapy (BNCT). The research reactor MARIA at the Institute of Atomic Energy in Swierk (Poland) is the primary source of the reactor thermal neutrons, while the secondary source should supply epithermal neutrons. The other secondary source is the pulsed source of thermal neutrons that uses fast 14 MeV neutrons from a pulsed generator at the Institute of Nuclear Physics PAN in Krakow (Poland). The physical problems to be solved in the two mentioned cases are different. Namely, in order to devise the BNCT source the initial energy of particles ought to be increased, whilst in the other case the fast neutrons have to be moderated. Slowing down of neutrons is relatively easy since these particles lose energy when they scatter in media; the most effective moderators are the materials which contain light elements (mostly hydrogen). In order to increase the energy of neutrons from thermal to epithermal (the BNCT case) the so-called neutron converter should be exploited. It contains a fissile material, 235 U. The thermal neutrons from the reactor cause fission of uranium and fast neutrons are emitted from the converter. Then fissile neutrons of energy of a few MeV are slowed down to the required epithermal energy range. The design of both secondary sources have been conducted by means of Monte Carlo simulations, which have been carried out using the MCNP code. In the case of the secondary pulsed thermal neutron source, some of the calculated results have been verified experimentally. (author)

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)

Proposed Californium-252 User Facility for Neutron Science at Oak Ridge National Laboratory

International Nuclear Information System (INIS)

Martin, R.C.; Laxson, R.R.; Knauer, J.B.

1996-01-01

The Radiochemical Engineering Development Center (REDC) at ORNL has petitioned to establish a Californium-252 User Facility for Neutron Science for academic, industrial, and governmental researchers. The REDC Californium Facility (CF) stores the national inventory of sealed 252 Cf neutron source for university and research loans. Within the CF, the 252 Cf storage pool and two uncontaminated hot cells currently in service for the Californium Program will form the physical basis for the User Facility. Relevant applications include dosimetry and experiments for neutron tumor therapy; fast and thermal neutron activation analysis of materials; experimental configurations for prompt gamma neutron activation analysis; neutron shielding and material damage studies; and hardness testing of radiation detectors, cameras, and electronics. A formal User Facility simplifies working arrangements and agreements between US DOE facilities, academia, and commercial interests

Dynamic behaviour and neutron noise in molten salt reactors with circulating perturbations

Energy Technology Data Exchange (ETDEWEB)

Pazsit, I.; Dykin, V. [Chalmers Univ. of Tech., Nuclear Engineering, Goteborg (Sweden)

2014-07-01

This paper concerns the calculation of the neutron noise induced in Molten Salt Reactors (MSR) by the random fluctuations in space and time of the molten fuel cross sections which travel together with the fuel and pass the core region. The effect of such fluctuations was already discussed in several publications. The novelty of the present paper is that it takes into account that in addition to the delayed neutron precursors, also the cross section perturbations themselves, whose passing through the core induces the in-core neutron noise, return to the core inlet via the external loop from the core exit. The corresponding theory is developed, and some quantitative investigations are made of the characteristics of the noise, which can be attributed to the recirculation of the perturbation to the core. It is shown that the effect of the returning of the perturbations, even though it is also associated with a temporal decay, has a much stronger effect on the neutron noise spectra than that of the recirculation of the delayed neutron precursors. (author)

Dynamic behaviour and neutron noise in molten salt reactors with circulating perturbations

International Nuclear Information System (INIS)

Pazsit, I.; Dykin, V.

2014-01-01

This paper concerns the calculation of the neutron noise induced in Molten Salt Reactors (MSR) by the random fluctuations in space and time of the molten fuel cross sections which travel together with the fuel and pass the core region. The effect of such fluctuations was already discussed in several publications. The novelty of the present paper is that it takes into account that in addition to the delayed neutron precursors, also the cross section perturbations themselves, whose passing through the core induces the in-core neutron noise, return to the core inlet via the external loop from the core exit. The corresponding theory is developed, and some quantitative investigations are made of the characteristics of the noise, which can be attributed to the recirculation of the perturbation to the core. It is shown that the effect of the returning of the perturbations, even though it is also associated with a temporal decay, has a much stronger effect on the neutron noise spectra than that of the recirculation of the delayed neutron precursors. (author)

Possibilities of activation analysis with 252Cf sources in the area of earth sciences and mining prospection

International Nuclear Information System (INIS)

Melky, Sami.

1975-01-01

The 252 Cf neutron activation experimental set-up is composed of four sources of 225μg each, symmetrically disposed with regard to the sample axis. This installation permits the irradiation of 5 to 40cm 3 of ores, so that the analysis is done on sample weighing about 100g. This method allows, not only the determination of major and minor constituents in the sample, but also the determination of some trace elements. The experimental sensitivity of some important elements are given in mining prospection and it is shown that direct gamma spectrometry after activation permits a fast and precise determination of many elements in ores [fr

Improvements to the internal and external antenna H(-) ion sources at the Spallation Neutron Source.

Science.gov (United States)

Welton, R F; Dudnikov, V G; Han, B X; Murray, S N; Pennisi, T R; Pillar, C; Santana, M; Stockli, M P; Turvey, M W

2014-02-01

The Spallation Neutron Source (SNS), a large scale neutron production facility, routinely operates with 30-40 mA peak current in the linac. Recent measurements have shown that our RF-driven internal antenna, Cs-enhanced, multi-cusp ion sources injects ∼55 mA of H(-) beam current (∼1 ms, 60 Hz) at 65-kV into a Radio Frequency Quadrupole (RFQ) accelerator through a closely coupled electrostatic Low-Energy Beam Transport system. Over the last several years a decrease in RFQ transmission and issues with internal antennas has stimulated source development at the SNS both for the internal and external antenna ion sources. This report discusses progress in improving internal antenna reliability, H(-) yield improvements which resulted from modifications to the outlet aperture assembly (applicable to both internal and external antenna sources) and studies made of the long standing problem of beam persistence with the external antenna source. The current status of the external antenna ion source will also be presented.

Radiation-induced segregation in materials: Implications for accelerator-driven neutron source applications

Energy Technology Data Exchange (ETDEWEB)

Faulkner, R.B.; Song, S. [Loughborough Univ. of Technology (United Kingdom)

1995-10-01

This paper reviews exisiting models for radiation-induced segregation to microstrucural interfaces and surfaces. It indicates how the models have been successfully used in the past in neutron irradiation situations and how they may be modified to account for accelerator-driven RIS. The predictions of the models suggest that any impurity with large misfit will suffer RIS and that the effect is heightened as radiation damage increases. The paper suggests methods to utilise the RIS in transmutation technology by dynamically segregating long life nuclides to preferred sites in the microstructure so that subsequent transmutations occur with maximum efficiency.

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.

New neutron imaging using pulsed sources. Characteristics of a pulsed neutron source and principle of pulsed neutron imaging

International Nuclear Information System (INIS)

Kiyanagi, Yoshiaki

2012-01-01

Neutron beam is one of important tools to obtain the transmission image of an object. Until now, steady state neutron sources such as reactors are mainly used for this imaging purpose. Recently, it has been demonstrated that pulsed neutron imaging based on accelerator neutron sources can provide a real-space distribution of physical information of materials such as crystallographic structure, element, temperature, hydrogen bound state, magnetic field and so on, by analyzing wavelength dependent transmission spectrum, which information cannot be observed or difficult to obtain with a traditional imaging method using steady state neutrons. Here, characteristics of the pulsed neutron source and principle of the pulsed neutron imaging are explained as a basic concept of the new method. (author)

Application of fission track detectors to californium-252 neutron dosimetry in tissue near the radiation source

International Nuclear Information System (INIS)

Oswald, R.A.; Lanzl, L.H.; Rozenfeld, M.

1981-01-01

Fission track detectors were applied to a unique problem in neutron dosimetry. Measurements of neutron doses were required at locations within a tumor of 1 cm diameter implanted on the back of a mouse and surrounded by a square array of four 252 Cf medical sources. Measurements made in a tissue-equivalent mouse phantom showed that the neutron dose rate to the center of the tumor was 2.18 rads mg -1 h -1 +- 8.4%. The spatial variation of neutron dose to the tumor ranged from 1.88 to 2.55 rads mg -1 h -1 . These measurements agree with calculated values of neutron dose to those locations in the phantom. Fission track detectors have been found to be a reliable tool for neutron dosimetry for geometries in which one wishes to know neutron dose values which may vary considerably over distances of 1 cm or less

Application of fission track detectors to californium-252 neutron dosimetry in tissue near the radiation source

International Nuclear Information System (INIS)

Oswald, R.A.; Lanzl, L.H.; Rozenfeld, M.

1981-01-01

Fission track detectors were applied to a unique problem in neutron dosimetry. Measurements of neutron doses were required at locations within a tumor of 1 cm diameter implanted on the back of a mouse and surrounded by a square array of four 252 Cf medical sources. Measurements made in a tissue-equivalent mouse phantom showed that the neutron dose rate to the center of the tumor was 2.18 rads micrograms-1 h-1 +/- 8.4%. The spatial variation of neutron dose to the tumor ranged from 1.88 to 2.55 rads micrograms-1 h-1. These measurements agree with calculated values of neutron dose to those locations in the phantom. Fission track detectors have been found to be a reliable tool for neutron dosimetry for geometries in which one wishes to know neutron dose values which may vary considerably over distances of 1 cm or less

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.

Research and Development Program in Reactor Diagnostics and Monitoring with Neutron Noise Methods. Stage 10. Final Report

International Nuclear Information System (INIS)

Demaziere, C.; Pazsit, I.; Sunde, S.; Wright, J.

2004-12-01

This report constitutes Stage 10 of a long-term research and development program concerning the development of diagnostics and monitoring methods for nuclear reactors. The program executed in Stage 10 consists of three parts and the work performed in each part is summarized below. A so-called 2-D 2-group neutron noise simulator was previously developed at the Department of Reactor Physics. This simulator is able to calculate the space-dependence of the neutron noise induced by localised or spatially-distributed absorbers of variable strength or by vibrating absorbers. These calculations are performed in the 2-group diffusion approximation and can treat an arbitrary heterogeneous 2-D system. The goal of the present investigation is to use the simulator for unfolding purposes, i.e. to reconstruct the noise source from the detector readings. This task is particularly challenging since the number of detectors available in a commercial PWR can be very low. In this study, five detectors are assumed to be present and evenly distributed in the core. Furthermore, only localised absorbers of variable strength are considered as perturbations. Numerical simulations were first carried out to verify that the space-dependent local and global components of the neutron noise were overwhelmingly large compared to the point-kinetic term of the neutron noise whose spatial structure does not depend on the position of the perturbation. The significance of the space-dependent global component is that its relaxation length is large enough, so that several neutron detectors can monitor it. Therefore, the neutron noise induced at the position of the detectors is itself a function of the position of the perturbation. Unfolding procedures could thus be considered. Prior to performing any unfolding, the type of the noise source has to be determined, since the algorithms developed in this report are based on the assumption of an absorber of variable strength. Such a noise source can also be

Developments and application of neutron noise diagnostics of sodium cooled fast reactors

International Nuclear Information System (INIS)

Zylbersztejn, F.

2013-01-01

The Sodium cooled Fast Reactor (SFR) is one of the six reactor types selected by the Generation-IV international forum (GIF), and the building of an industrial prototype is planned in France. The safety standard of the future SFR has to be equivalent to the EPR's. The general improvement of the safety of the new reactor goes through the examination of all the potentially harmful scenarios and both the study and monitoring of early signs. The mechanical deformations of the core can have harmful consequences in sodium fast reactors, such as unexpected power variations due to the reactivity increase in case of core compaction, or the excessive deterioration of the mechanical structures. The monitoring of such phenomena and of their potential early signs is then needed. The monitoring of such phenomena can be done with neutron detectors placed inside and outside the tank. This PhD thesis deals with the study of the neutron noise generated by the periodic deformation of the SFR core, restricted to the so-called core compaction or core flowering phenomenon, a deformation consisting in the variation of the inter-assembly sodium width by a radial bending the assemblies (the assemblies in SFR are held by the base). The PhD thesis has been performed within collaboration between CEA (France) and Chalmers Institute of Technology (Sweden). The work realized during the thesis led to the publication of 3 articles as first author and another as second author. This work has embraced the following topics: A state of the art of the monitoring of the core deformation phenomenon by interpretation of the noise measurements in SFR has been done. The PHENIX reactor multi physics measurements database has been scrutinized to provide an interpretation of the neutron noise bringing out mechanical vibration phenomena. An important conclusion was that the lack of theoretical knowledge about the neutron noise induced by the vibration phenomenon and the ill positioning of the neutron detectors

Evaluation of room-scattered neutrons at the JNC Tokai neutron reference field

Energy Technology Data Exchange (ETDEWEB)

Yoshida, Tadayoshi; Tsujimura, Norio [Japan Nuclear Cycle Development Inst., Tokai, Ibaraki (Japan). Tokai Works; Oyanagi, Katsumi [Japan Radiation Engineering Co., Ltd., Hitachi, Ibaraki (Japan)

2002-09-01

Neutron reference fields for calibrating neutron-measuring devices in JNC Tokai Works are produced by using radionuclide neutron sources, {sup 241}Am-Be and {sup 252}Cf sources. The reference field for calibration includes scattered neutrons from the material surrounding sources, wall, floor and ceiling of the irradiation room. It is, therefore, necessary to evaluate the scattered neutrons contribution and their energy spectra at reference points. Spectral measurements were performed with a set of Bonner multi-sphere spectrometers and the reference fields were characterized in terms of spectral composition and the fractions of room-scattered neutrons. In addition, two techniques stated in ISO 10647, the shadow-cone method and the polynomial fit method, for correcting the contributions from the room-scattered neutrons to the readings of neutron survey instruments were compared. It was found that the two methods gave an equivalent result within a deviation of 3.3% at a source-to-detector distance from 50cm to 500cm. (author)

Evaluation of room-scattered neutrons at the JNC Tokai neutron reference field

International Nuclear Information System (INIS)

Yoshida, Tadayoshi; Tsujimura, Norio

2002-01-01

Neutron reference fields for calibrating neutron-measuring devices in JNC Tokai Works are produced by using radionuclide neutron sources, 241 Am-Be and 252 Cf sources. The reference field for calibration includes scattered neutrons from the material surrounding sources, wall, floor and ceiling of the irradiation room. It is, therefore, necessary to evaluate the scattered neutrons contribution and their energy spectra at reference points. Spectral measurements were performed with a set of Bonner multi-sphere spectrometers and the reference fields were characterized in terms of spectral composition and the fractions of room-scattered neutrons. In addition, two techniques stated in ISO 10647, the shadow-cone method and the polynomial fit method, for correcting the contributions from the room-scattered neutrons to the readings of neutron survey instruments were compared. It was found that the two methods gave an equivalent result within a deviation of 3.3% at a source-to-detector distance from 50cm to 500cm. (author)

Accelerator shield design of KIPT neutron source facility

International Nuclear Information System (INIS)

Zhong, Z.; Gohar, Y.

2013-01-01

Argonne National Laboratory (ANL) of the United States and Kharkov Institute of Physics and Technology (KIPT) of Ukraine have been collaborating on the design development of a neutron source facility at KIPT utilizing an electron-accelerator-driven subcritical assembly. Electron beam power is 100 kW, using 100 MeV electrons. The facility is designed to perform basic and applied nuclear research, produce medical isotopes, and train young nuclear specialists. The biological shield of the accelerator building is designed to reduce the biological dose to less than 0.5-mrem/hr during operation. The main source of the biological dose is the photons and the neutrons generated by interactions of leaked electrons from the electron gun and accelerator sections with the surrounding concrete and accelerator materials. The Monte Carlo code MCNPX serves as the calculation tool for the shield design, due to its capability to transport electrons, photons, and neutrons coupled problems. The direct photon dose can be tallied by MCNPX calculation, starting with the leaked electrons. However, it is difficult to accurately tally the neutron dose directly from the leaked electrons. The neutron yield per electron from the interactions with the surrounding components is less than 0.01 neutron per electron. This causes difficulties for Monte Carlo analyses and consumes tremendous computation time for tallying with acceptable statistics the neutron dose outside the shield boundary. To avoid these difficulties, the SOURCE and TALLYX user subroutines of MCNPX were developed for the study. The generated neutrons are banked, together with all related parameters, for a subsequent MCNPX calculation to obtain the neutron and secondary photon doses. The weight windows variance reduction technique is utilized for both neutron and photon dose calculations. Two shielding materials, i.e., heavy concrete and ordinary concrete, were considered for the shield design. The main goal is to maintain the total

Atmospheric observations for quantifying emissions of point-source synthetic greenhouse gases (CF4, NF3 and HFC-23)

Science.gov (United States)

Arnold, Tim; Manning, Alistair J.; Li, Shanlan; Kim, Jooil; Park, Sunyoung; Fraser, Paul J.; Mitrevski, Blagoj; Steele, L. Paul; Krummel, Paul B.; Mühle, Jens; Weiss, Ray F.

2016-04-01

The fluorinated species carbon tetrafluoride (CF4; PFC-14), nitrogen trifluoride (NF3) and trifluoromethane (CHF3; HFC-23) are potent greenhouse gases with 100-year global warming potentials of 6,630, 16,100 and 12,400, respectively. Unlike the majority of CFC-replacement compounds that are emitted from fugitive and mobile emission sources, these gases are largely emitted from large single point sources - semiconductor manufacturing facilities (all three), aluminium smelting plants (CF4) and chlorodifluoromethane factories (HFC-23). In this work we show the potential for atmospheric measurements to understand regional sources of these gases and to highlight emission 'hotspots'. We target our analysis on measurements from two Advanced Global Atmospheric Gases Experiment (AGAGE) long term monitoring sites that are particularly sensitive to regional emissions of these gases: Gosan on Jeju Island in the Republic of Korea and Cape Grim on Tasmania in Australia. These sites measure CF4, NF3 and HFC-23 alongside a suite of greenhouse and stratospheric ozone depleting gases every two hours using automated in situ gas-chromatography mass-spectrometry instrumentation. We couple each measurement to an analysis of air history using the regional atmospheric transport model NAME (Numerical Atmospheric dispersion Modelling Environment) driven by 3D meteorology from the Met Office's Unified Model, and use a Bayesian inverse method (InTEM - Inversion Technique for Emission Modelling) to calculate yearly emission changes over a decade (2005-2015) at high spatial resolution. At present these gases make a small contribution to global radiative forcing, however, given that their impact could rise significantly and that point sources of such gases can be mitigated, atmospheric monitoring could be an important tool for aiding emissions reduction policy.

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

International Nuclear Information System (INIS)

Stacey, W.M.

2001-01-01

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

Sources of polarized neutrons

International Nuclear Information System (INIS)

Walter, L.

1983-01-01

Various sources of polarized neutrons are reviewed. Monoenergetic source produced with unpolarized or polarized beams, white sources of polarized neutrons, production by transmissions through polarized hydrogen targets and polarized thermal neutronsare discussed, with appropriate applications included. (U.K.)

BEAM-LOSS DRIVEN DESIGN OPTIMIZATION FOR THE SPALLATION NEUTRON SOURCE (SNS) RING.

Energy Technology Data Exchange (ETDEWEB)

WEI,J.; BEEBE-WANG,J.; BLASKIEWICZ,M.; CAMERON,P.; DANBY,G.; GARDNER,C.J.; JACKSON,J.; LEE,Y.Y.; LUDEWIG,H.; MALITSKY,N.; RAPARIA,D.; TSOUPAS,N.; WENG,W.T.; ZHANG,S.Y.

1999-03-29

This paper summarizes three-stage design optimization for the Spallation Neutron Source (SNS) ring: linear machine design (lattice, aperture, injection, magnet field errors and misalignment), beam core manipulation (painting, space charge, instabilities, RF requirements), and beam halo consideration (collimation, envelope variation, e-p issues etc.).

Beam-Loss Driven Design Optimization for the Spallation Neutron Source (SNS) Ring

International Nuclear Information System (INIS)

Wei, J.

1999-01-01

This paper summarizes three-state design optimization for the Spallation Neutron Source (SNS) ring: linear machine design (lattice, aperture, injection, magnet field errors and misalignment), beam core manipulation (painting, space charge, instabilities, RF requirements), and beam halo consideration (collimation, envelope variation, e-p issues etc.)

Joint Neutron Noise Measurements on Metallic Reactor Caliban

International Nuclear Information System (INIS)

Chapelle, Amaury; Authier, Nicolas; Pierre, Casoli; Richard, Benoit; Myers, Will; Hutchinson, Jesson; Sood, Avneet; Rooney, Brian

2013-06-01

The aim of the experiments concerning neutron noise measurements presented in this article is to compare the measured parameters to the simulated ones. The results of these measurements must therefore be very accurate, with controlled uncertainties. To determine the relative contribution of uncertainties to the final result, a table presents the prompt multiplication obtained by a French Team and a U.S. team. The different sources of uncertainties are then explored, distinguishing them between three categories, those linked to the experimental configuration, to the detection process and finally to the analysis process. These experiments improve the safety task of reactivity control far from criticality, with static methods, and the knowledge of the behaviour of a subcritical reactor. (authors)

Linac design study for an intense neutron-source driver

International Nuclear Information System (INIS)

Lynch, M.T.; Browman, A.; DeHaven, R.; Jameson, R.; Jason, A.; Neuschaefer, G.; Tallerico, P.; Regan, A.

1993-01-01

The 1-MW spallation-neutron source under design study at Los Alamos is driven by a linac-compressor-ring scheme that utilizes a large portion of the existing Los Alamos Meson Physics Facility (LAMPF) linac, as well as the facility infrastructure. The project is referred to as the National Center for Neutron Research (NCNR). A second phase of the proposal will upgrade the driver power to 5 MW. A description of the 1-MW scheme is given in this paper. In addition, the upgrade path to the substantial increase of beam power required for the 5 MW scenario is discussed

Intense resonance neutron source (IREN) - new pulsed source for nuclear physical and applied investigations

International Nuclear Information System (INIS)

Anan'ev, V.D.; Furman, W.I.; Kobets, V.V.; Meshkov, I.N.; Pyataev, V.G.; Shirkov, G.D.; Shvets, V.A.; Sumbaev, A.P.; Kuatbekov, R.P.; Tret'yakov, I.T.; Frolov, A.R.; Gurov, S.M.; Logachev, P.V.; Pavlov, V.M.; Skarbo, B.A.

2005-01-01

An accelerator-driven subcritical system (200 MeV electron linac + metallic plutonium subcritical core) IREN is constructed at the Joint Institute for Nuclear Research (JINR). The new pulsed neutron source IREN is optimized for maximal yield of resonance neutrons (1-10 5 eV). The S-band electron linac with a pulse duration near 200 ns, repetition rate up to 150 Hz and the mean beam power 10 kW delivers 200-MeV electrons onto a specially designed tungsten target (an electron-neutron converter) situated in the center of a very compact and fast subcritical assembly with K eff 15 per second. A mean fission power of the multiplying target is planned to be near 15 kW. The current status of the project is presented

Thermal neutron source study

International Nuclear Information System (INIS)

Holden, T.M.

1983-05-01

The value of intense neutron beams for condensed matter research is discussed with emphasis on the complementary nature of steady state and pulsed neutron sources. A large body of information on neutron sources, both existing and planned, is then summarized under four major headings: fission reactors, electron accelerators with heavy metal targets, pulsed spallation sources and 'steady state' spallation sources. Although the cost of a spallation source is expected to exceed that of a fission reactor of the same flux by a factor of two, there are significant advantages for a spallation device such as the proposed Electronuclear Materials Test Facility (EMTF)

Neutron PSDs for the next generation of spallation neutron sources

CERN Document Server

Eijk, C W

2002-01-01

A review of R and D for neutron PSDs to be used at anticipated new spallation neutron sources: the Time-of-Flight system facility, European Spallation Source, Spallation Neutron Source and Neutron Arena, is presented. The gas-filled detectors, scintillation detectors and hybrid systems are emphasized.

Calibration and evaluation of neutron survey meters used at linac facility

Energy Technology Data Exchange (ETDEWEB)

Salgado, A.P. [Instituto de Radioprotecao e Dosimetria - IRD, Av. Salvador Allende s/n, Recreio dos Bandeirantes, CEP 22780-160 Rio de Janeiro (Brazil); Pereira, W.W., E-mail: walsan@ird.gov.b [Instituto de Radioprotecao e Dosimetria - IRD, Av. Salvador Allende s/n, Recreio dos Bandeirantes, CEP 22780-160 Rio de Janeiro (Brazil); Fonseca, E.S. da; Patrao, K.C.S. [Instituto de Radioprotecao e Dosimetria - IRD, Av. Salvador Allende s/n, Recreio dos Bandeirantes, CEP 22780-160 Rio de Janeiro (Brazil); Batista, D.V.S. [Instituto Nacional do Cancer - INCa, Praca Cruz Vermelha, 23 - centro, CEP 20230-130 Rio de Janeiro (Brazil)

2010-12-15

Calibrated survey meters from the Neutron Laboratory of the Instituto de Radioprotecao e Dosimetria (IRD) were used to determine the ambient dose-equivalent rate in a 15 MV linear accelerator treatment room at the Instituto Nacional do Cancer (INCa). Three different models of neutron survey meters were calibrated using four neutron radionuclide neutron sources: {sup 241}AmBe({alpha},n), {sup 252}Cf(f,n), heavy-water moderated {sup 252}Cf(f,n), and {sup 238}PuBe({alpha},n). All neutron sources were standardized in a Manganese Sulphate Bath (MSB) absolute primary system. The response of each of these instruments was compared with reference values of ambient dose-equivalent rate. The results demonstrate the complexity of making measurements in the mixed neutron/photon field produced in electron linear accelerator radiotherapy treatment rooms.

Means and method for controlling the neutron output of a neutron generator tube

International Nuclear Information System (INIS)

1977-01-01

A means and method for energizing and regulating a neutron generator tube is described. It has a target, an ion source and a replenisher. A negative high voltage is applied to the target and the target current monitored. A constant current from a constant current source is divided into a shunt current and a replenisher current in accordance with the target current. The replenisher current is applied to the replenisher in a neutron generator tube so as to control the neutron output in accordance with the target current. (C.F.)

Viability of neutron brachytherapy technique using Cf252 for tumors of salivary glands therapy

International Nuclear Information System (INIS)

Andrade, Lidia M.; Campos, Tarcisio P.R.

2000-01-01

Gland salivary tumors, although uncommon, present significant characteristics such as distant metastasis, recurrence and average survive of five year for treated patients. Those tumors presents low response to a conventional radiotherapy, photon and electron therapy; thus, this modality is associated with surgery. Fast neutron therapy has been presented better results in controlling the local tumor in comparison with conventional radiotherapy. Brachytherapy with Cf-252 presents an alternative treatment for tumors. Those radiotherapy technique are discussed and analyzed. (author)

Analysis of linear energy transfers and quality factors of charged particles produced by spontaneous fission neutrons from 252Cf and 244Pu in the human body

International Nuclear Information System (INIS)

Endo, A.; Sato, T.

2013-01-01

Absorbed doses, linear energy transfers (LETs) and quality factors of secondary charged particles in organs and tissues, generated via the interactions of the spontaneous fission neutrons from. 252 Cf and. 244 Pu within the human body, were studied using the Particle and Heavy Ion Transport Code System (PHITS) coupled with the ICRP Reference Phantom. Both the absorbed doses and the quality factors in target organs generally decrease with increasing distance from the source organ. The analysis of LET distributions of secondary charged particles led to the identification of the relationship between LET spectra and target-source organ locations. A comparison between human body-averaged mean quality factors and fluence-averaged radiation weighting factors showed that the current numerical conventions for the radiation weighting factors of neutrons, updated in ICRP103, and the quality factors for internal exposure are valid. (authors)

Signal and noise analysis in TRION-Time-Resolved Integrative Optical Fast Neutron detector

International Nuclear Information System (INIS)

Vartsky, D; Feldman, G; Mor, I; Goldberg, M B; Bar, D; Dangendorf, V

2009-01-01

TRION is a sub-mm spatial resolution fast neutron imaging detector, which employs an integrative optical time-of-flight technique. The detector was developed for fast neutron resonance radiography, a method capable of detecting a broad range of conventional and improvised explosives. In this study we have analyzed in detail, using Monte-Carlo calculations and experimentally determined parameters, all the processes that influence the signal and noise in the TRION detector. In contrast to event-counting detectors where the signal-to-noise ratio is dependent only on the number of detected events (quantum noise), in an energy-integrating detector additional factors, such as the fluctuations in imparted energy, number of photoelectrons, system gain and other factors will contribute to the noise. The excess noise factor (over the quantum noise) due to these processes was 4.3, 2.7, 2.1, 1.9 and 1.9 for incident neutron energies of 2, 4, 7.5, 10 and 14 MeV, respectively. It is shown that, even under ideal light collection conditions, a fast neutron detection system operating in an integrative mode cannot be quantum-noise-limited due to the relatively large variance in the imparted proton energy and the resulting scintillation light distributions.

Experimental optimisation of a moderated sup 2 sup 5 sup 2 Cf source for land mine detection

CERN Document Server

Zuin, L; Fabris, D; Lunardon, M; Nebbia, G; Viesti, G; Cinausero, M; Fioretto, E; Prete, G; Palomba, M; Pantaleo, A

2000-01-01

A moderated sup 2 sup 5 sup 2 Cf source to be used for the detection of buried land mines has been experimentally characterised. The moderating structure was obtained by using an inner sphere of Pb embedded in a high-density polyethylene (HDPe) brick assembly. The number of capture events in a Cd sample placed outside the structure was used, by varying the dimensions of the brick assembly, to determine the best moderator geometry. Furthermore, the detection of land mines by neutron capture reaction on nitrogen nuclei contained in the explosive was simulated by in field measurements placing the Cd sample at different depths in the soil. The obtained results are compared with predictions from Monte Carlo calculations.

Radiography using californium-252 neutron sources

International Nuclear Information System (INIS)