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Sample records for neutron spectrum flwr

  1. Neutronic characteristics of FLWR in the transition phase changing from high conversion core to breeder core

    International Nuclear Information System (INIS)

    Akie, Hiroshi; Nakano, Yoshihiro; Okubo, Tsutomu

    2009-01-01

    Innovative Water Reactor for Flexible Fuel Cycle (FLWR) is a low moderation type boiling water reactor which can realize plutonium multiple recycling and breeding. For the introduction stage of FLWR, a high conversion (HC) type FLWR is proposed to keep technical continuity from current light water reactors. The HC core of FLWR has a less tight fuel lattice with lower coolant void fraction than the breeder (BR) type core. The HC type FLWR core is to be shifted to the BR core by only replacing the fuel assemblies of the same outer shape and size in the same reactor system. In the HC to BR transition phase of FLWR, there exist both types of fuel assemblies in the same core configuration. In the HC assembly, neutron spectrum is softer than in the BR assembly, and the axial fuel and blanket arrangement is different from the BR assembly. Due to these differences, there might appear a power peaking in the adjacent region between HC and BR assemblies. The power distribution in the HC + BR assemblies mixed core configuration is studied by performing assembly calculations and core calculations on a few assemblies local geometry and the whole core geometry. As a result, although a power peaking can be locally very large in the HC and BR assemblies adjacent regions, such local power peakings are shown to be effectively reduced by considering a rod-wise fuel enrichment distribution. In the whole core calculation, it seems possible to optimize the fuel assembly loading and shuffling pattern to avoid large power level mismatch between the assemblies. It is expected that FLWR can be shifted from HC type to BR type without major neutronic difficulties. (author)

  2. Stepwise evolution of fuel assembly design toward a sustainable fuel cycle with hard neutron spectrum light water reactors

    International Nuclear Information System (INIS)

    Uchikawa, Sadao; Okubo, Tsutomu; Nakano, Yoshihiro

    2011-01-01

    An advanced LWR with hard neutron spectrum, FLWR, aims at efficient and flexible utilization of nuclear resources by evolving its fuel assembly design keeping the same core configuration. A proposed evolution process of the design toward a sustainable fuel cycle is composed of three stages, the first one based on the LWR fuel cycle infrastructures, the second one for transitioning from the LWR fuel cycle to the FR fuel cycle, and the third one based on the FR fuel cycle infrastructures. For the first stage, a fuel assembly design concept named FLWR/MIX has been developed in which enriched UO 2 fuel rods are arranged in the peripheral region of the assembly, surrounding the MOX fuel rods in the central region. The FLWR/MIX design realizes a breeder type operation under the framework of the LWR-MOX technologies and there experience. A modified FLWR/MIX design with low Pu inventory for the second stage has a potential of high Puf conversion ratio of 1.1 and can contribute to smooth and speedy transition from the LWR fuel cycle to the FR fuel cycle. For the third stage, the FLWR/MIX design is extended into a design with natural UO 2 fuel rods to realize multiple Pu recycling keeping a Puf conversion ratio of around 1.0. (author)

  3. Study on high conversion type core of innovative water reactor for flexible fuel cycle (FLWR) for minor actinide (MA) recycling

    International Nuclear Information System (INIS)

    Fukaya, Yuji; Nakano, Yoshihiro; Okubo, Tsutomu

    2009-01-01

    In order to ensure sustainable energy supplies in the future based on the well-established light water reactor (LWR) technologies, conceptual design studies have been performed on the innovative water reactor for flexible fuel cycle (FLWR) with the high conversion ratio core. For early introduction of FLWR without a serious technical gap from the LWR technologies, the conceptual design of the high conversion type one (HC-FLWR) was constructed to recycle reprocessed plutonium. Furthermore, an investigation of minor actinide (MA) recycling based on the HC-FLWR core concept has been performed and is presented in this paper. Because HC-FLWR is a near-term technology, it would be a good option in the future if HC-FLWR can recycle MAs. In order to recycle MAs in HC-FLWR, it has been found that the core design should be changed, because the loaded MA makes the void reactivity coefficient worse and decreases the discharge burn-up. To find a promising core design specification, the investigation on the core characteristics were performed using the results from parameter surveys with core burn-up calculations. The final core designs were established by coupled three dimensional neutronics and thermal-hydraulics core calculations. The major core specifications are as follows. The plutonium fissile (Puf) content is 13 wt%. The discharge burn-up is about 55 GWd/t. Around 2 wt% of Np or Am can be recycled. The MA conversion ratios are around unity. In particular, it has been found that loaded Np can be transmuted effectively in this core concept. Therefore, these concepts would be a good option to reduce environmental burdens.

  4. Neutron spectrum unfolding: Pt. 2

    International Nuclear Information System (INIS)

    Matiullah; Wiyaja, D.S.; Berzonis, M.A.; Bondars, H.; Lapenas, A.A.; Kudo, K.; Majeed, A.; Durrani, S.A.

    1991-01-01

    In Part I of this paper, we described the use of the computer code SAIPS in neutron spectrum unfolding. Here in Part II, we present our experimental work carried out to study the shape of the neutron spectrum in different experimental channels of a 5 MW light-water cooled and moderated research reactor. The spectral neutron flux was determined using various fission foils (placed in close contact with mica track detectors) and activation detectors. From the measured activities, the neutron spectrum was unfolded by SAIPS. (author)

  5. Intermediate neutron spectrum problems and the intermediate neutron spectrum experiment

    International Nuclear Information System (INIS)

    Jaegers, P.J.; Sanchez, R.G.

    1996-01-01

    Criticality benchmark data for intermediate energy spectrum systems does not exist. These systems are dominated by scattering and fission events induced by neutrons with energies between 1 eV and 1 MeV. Nuclear data uncertainties have been reported for such systems which can not be resolved without benchmark critical experiments. Intermediate energy spectrum systems have been proposed for the geological disposition of surplus fissile materials. Without the proper benchmarking of the nuclear data in the intermediate energy spectrum, adequate criticality safety margins can not be guaranteed. The Zeus critical experiment now under construction will provide this necessary benchmark data

  6. Neutron spectrum measurement by TOF

    International Nuclear Information System (INIS)

    Aizawa, Otohiko

    1982-01-01

    The TOF experiments by using various facilities are described. The steady neutron spectra in light water which contains non-1/V absorbing materials were measured by the TOF method at a LINAC facility. The results were compared with the calculations based on the Koppel-Haywood model and two others. The leakage neutron spectra from a heavy-water assembly were measured and compared with model calculations. The time-dependent energy spectra in a small graphite assembly were measured. For this measurement, a chopper system was also used. The two-region calculation explains the spectrum just after the neutron burst. The time-dependent spectra in a small Be assembly and in an assembly of coolant-moderator containing hydrogen were also measured. The calculations based on various models are in progress. The TOF experiments at the reactor-chopper facility were carried out for measuring the total cross sections of crystalline moderators, the thermal neutron total cross section of high temperature beryllium, the thermal neutron total cross sections of granular lead and high temperature liquid lead, and the angle-dependent scattering spectra. A pseudo-chopper was designed and constructed. The spectra of the neutron field for medical use were measured by the chopper-TOF system. The thermal neutron total cross sections of Fe, Zr, Nb and Mg were measured, and the results were compared with the calculations by THRUSH and UNCLE-TOM codes. The random-trigger TOF experiments were made by using Cf-252. (Kato, T.)

  7. Concept of innovative water reactor for flexible fuel cycle (FLWR)

    International Nuclear Information System (INIS)

    Iwamura, T.; Uchikawa, S.; Okubo, T.; Kugo, T.; Akie, H.; Nakatsuka, T.

    2005-01-01

    In order to ensure sustainable energy supply in the future based on the matured Light Water Reactor (LWR) and coming LWR-Mixed Oxide (MOX) technologies, a concept of Innovative Water Reactor for Flexible Fuel Cycle (FLWR) has been investigated in Japan Atomic Energy Research Institute (JAERI). The concept consists of two parts in the chronological sequence. The first part realizes a high conversion type core concept, which is basically intended to keep the smooth technical continuity from current LWR and coming LWR-MOX technologies without significant gaps in technical point of view. The second part represents the Reduced-Moderation Water Reactor (RMWR) core concept, which realizes a high conversion ratio over 1.0 being useful for the long-term sustainable energy supply through plutonium multiple recycling based on the well-experienced LWR technologies. The key point is that the two core concepts utilize the compatible and the same size fuel assemblies, and hence, the former concept can proceed to the latter in the same reactor system, based flexibly on the fuel cycle circumstances during the reactor operation period around 60 years. At present, since the fuel cycle for the plutonium multiple recycling with MOX fuel reprocessing has not been realized yet, reprocessed plutonium from the LWR spent fuel is to be utilized in LWR-MOX. After this stage, the first part of FLWR, i.e. the high conversion type, can be introduced as a replacement of LWR or LWR-MOX. Since the plutonium inventory of FLWR is much larger, the number of the reactor with MOX fuel will be significantly reduced compared to the LWR-MOX utilization. The size of the fuel assembly for the first part is the same as in the RMWR concept, i.e. the hexagonal fuel assembly with the inner face-to-face distance of about 200 mm. Fuel rods are arranged in the triangular lattice with a relatively wide gap size around 3 mm between rods, and the effective MOX length is less than 1.5 m without using the blanket. When

  8. A Wide Spectrum Neutron Polarizer for a Pulsed Neutron Source

    International Nuclear Information System (INIS)

    Nikitenko, Yu.V.

    1994-01-01

    A wide spectrum neutron polarizer for a pulsed neutron source is considered. The polarizer is made in a form of a set of magnetized mirrors placed on a drum. Homogeneous rotation of the polarizer is synchronized with the power pulses of the neutron source. The polarizer may be utilized in a collimated neutron beam with cross section of the order of magnitude of 100 cm 2 within a wavelength from 2 up to 20 A on sources with a pulse repetition frequency up to 50 Hz. (author). 5 refs.; 3 figs

  9. Neutron spectrum unfolding using neural networks

    International Nuclear Information System (INIS)

    Vega C, H.R.; Hernandez D, V.M.; Manzanares A, E.

    2004-01-01

    An artificial neural network has been designed to obtain the neutron spectra from the Bonner spheres spectrometer's count rates. The neural network was trained using a large set of neutron spectra compiled by the International Atomic Energy Agency. These include spectra from iso- topic neutron sources, reference and operational neutron spectra obtained from accelerators and nuclear reactors. The spectra were transformed from lethargy to energy distribution and were re-binned to 31 energy groups using the MCNP 4C code. Re-binned spectra and UTA4 matrix were used to calculate the expected count rates in Bonner spheres spectrometer. These count rates were used as input and correspondent spectrum was used as output during neural network training. The network has 7 input nodes, 56 neurons as hidden layer and 31 neurons in the output layer. After training the network was tested with the Bonner spheres count rates produced by twelve neutron spectra. The network allows unfolding the neutron spectrum from count rates measured with Bonner spheres. Good results are obtained when testing count rates belong to neutron spectra used during training, acceptable results are obtained for count rates obtained from actual neutron fields; however the network fails when count rates belong to monoenergetic neutron sources. (Author)

  10. Measurements of neutron spectrum from uranium converter

    International Nuclear Information System (INIS)

    Ninkovic, M.; Sotic, O.; Marinkovic, S.

    1978-01-01

    The procedure for determination of energetic distribution of neutrons by the multisphere technique is given. The theoretical basis and features of the method are explained. The spectral distribution of neutrons emerging from the neutron converter constructed at the bare reactor assembly RB, has been determined applying the existing computer programme and literature data for the energetic dependence functions of spheres of various diameters. The obtained spectral distribution has a specific maximum in the domain of fast neutrons, justifying thus the reacton for the construction of the converter. The neutron spectrum data obtained and given in this report are very important for the use of the converter in neutron dosimetry and radiation protection, as well as in the radiobiology, shielding, reactor physics etc. (author)

  11. NEUTRON SPECTRUM MEASUREMENTS USING MULTIPLE THRESHOLD DETECTORS

    Energy Technology Data Exchange (ETDEWEB)

    Gerken, William W.; Duffey, Dick

    1963-11-15

    From American Nuclear Society Meeting, New York, Nov. 1963. The use of threshold detectors, which simultaneously undergo reactions with thermal neutrons and two or more fast neutron threshold reactions, was applied to measurements of the neutron spectrum in a reactor. A number of different materials were irradiated to determine the most practical ones for use as multiple threshold detectors. These results, as well as counting techniques and corrections, are presented. Some materials used include aluminum, alloys of Al -Ni, aluminum-- nickel oxides, and magesium orthophosphates. (auth)

  12. Thermal neutron spectrum distribution in TRIGA fuels

    International Nuclear Information System (INIS)

    Gui Ah Auu; Harasawa, Susumu; An, Shigehiro

    1989-01-01

    The dependence of thermal neutron spectrum in TRIGA fuel cell on fuel temperature and TRIGA fuel types were studied using LIBP and THERMOS codes. Some characteristics of the TRIGA fuel including its prompt negative temperature coefficient of reactivity were explained using the results of the study. (author)

  13. Comparison of neutron spectrum unfolding codes

    International Nuclear Information System (INIS)

    Zijp, W.

    1979-02-01

    This final report contains a set of four ECN-reports. The first is dealing with the comparison of the neutron spectrum unfolding codes CRYSTAL BALL, RFSP-JUL, SAND II and STAY'SL. The other three present the results of calculations about the influence of statistical weights in CRYSTAL BALL, SAND II and RFSP-JUL

  14. Approximation for the adjoint neutron spectrum

    International Nuclear Information System (INIS)

    Suster, Luis Carlos; Martinez, Aquilino Senra; Silva, Fernando Carvalho da

    2002-01-01

    The proposal of this work is the determination of an analytical approximation which is capable to reproduce the adjoint neutron flux for the energy range of the narrow resonances (NR). In a previous work we developed a method for the calculation of the adjoint spectrum which was calculated from the adjoint neutron balance equations, that were obtained by the collision probabilities method, this method involved a considerable quantity of numerical calculation. In the analytical method some approximations were done, like the multiplication of the escape probability in the fuel by the adjoint flux in the moderator, and after these approximations, taking into account the case of the narrow resonances, were substituted in the adjoint neutron balance equation for the fuel, resulting in an analytical approximation for the adjoint flux. The results obtained in this work were compared to the results generated with the reference method, which demonstrated a good and precise results for the adjoint neutron flux for the narrow resonances. (author)

  15. ATW neutron spectrum measurements at LAMPF

    Energy Technology Data Exchange (ETDEWEB)

    Butler, G.W.; Littleton, P.E.; Morgan, G.L. [Los Alamos National Laboratory, NM (United States)] [and others

    1995-10-01

    Accelerator transmutation of waste (ATW) is a proposal to use a high flux of accelerator-produced thermalized neutrons to transmute both fission product and higher actinide commercial nuclear waste into stable or short-lived radioactive species in order to avoid long-term storage of nuclear waste. At LAMPF the authors recently performed experiments that were designed to measure the spectrum of neutrons produced per incident proton for full-scale proposed ATW targets of lead and lithium. The neutrons produced in such targets have a spectrum of energies that extends up to the energy of the incident proton beam, but the distribution peaks between 1 and 5 MeV. Transmutation reactions and fission of actinides are most efficient when the neutron energy is below a few eV, so the target must be surrounded by a non-absorbing material (blanket) to produce additional neutrons and reduce the energy of high energy neutrons without loss. The experiments with the lead target, 25 cm diameter by 40 cm long, were conducted with 800 MeV protons, while those with the lithium target, 25 cm diameter by 175 cm long, were conducted with 400 MeV protons. The blanket in both sets of experiments was a 60 cm diameter by 200 cm long annulus of lead that surrounded the target. Surrounding the blanket was a steel water tank with dimensions of 250 cm diameter by 300 cm long that simulated the transmutation region. A small sample pipe penetrated the length of the lead blanket and other sample pipes penetrated the length of the water tank at different radii from the beam axis so that the neutron spectra at different locations could be measured by foil activation. After irradiation the activated foil sets were extracted and counted with calibrated high resolution germanium gamma ray detectors at the Los Alamos nuclear chemistry counting facility.

  16. Simultaneous neutron and gamma spectrum adjustment

    International Nuclear Information System (INIS)

    Remec, I.

    1996-01-01

    The spectrum adjustment procedure was extended to simultaneous neutron and gamma spectrum adjustment, and the feasibility of this technique is demonstrated in the analysis of HFIR dosimetry experiments. Conditions in which gamma rays may contribute considerably to radiation damage in steels are discussed. Beryllium helium accumulation fluence monitors (HAFMs) were found to be good monitors in gamma fields of intensities high enough to contribute to steel embrittlement. Use of 237 Np, 238 U, and 9 Be HAFM as gamma dosimeters is proposed for high-dose irradiations in high-energy, high-intensity gamma fields

  17. Neutron spectrum adjustment. The role of covariances

    International Nuclear Information System (INIS)

    Remec, I.

    1992-01-01

    Neutron spectrum adjustment method is shortly reviewed. Practical example dealing with power reactor pressure vessel exposure rates determination is analysed. Adjusted exposure rates are found only slightly affected by the covariances of measured reaction rates and activation cross sections, while the multigroup spectra covariances were found important. Approximate spectra covariance matrices, as suggested in Astm E944-89, were found useful but care is advised if they are applied in adjustments of spectra at locations without dosimetry. (author) [sl

  18. A technique of measuring neutron spectrum

    International Nuclear Information System (INIS)

    Sarkar, P.K.; Kirthi, K.N.; Ganguly, A.K.

    1975-01-01

    Plastic scintillators have been used to measure fast neutron spectrum from various sources. Gamma background discrimination has been done by selecting thin scintillators and thereby achieving near 100% transmission of Compton-edge electrons. The measured distribution has been unfolded by using an iterative least square technique. This gives minimum variance and maximum likelihood estimate with error minimised. Smoothening of the observed distribution has been done by Fourier and time series analyses. The method developed is applicable in principle for the determination of spectra of high energy neutrons ranging from 1 MeV to 70 MeV and beyond. However, practical application of the method is limited by the non-availability of cross-section data for various neutron induced reactions with carbon and hydrogen present in the polymerised polystyrene scintillator. This procedure has been adopted in the present work for spectral determination up to 14 MeV neutrons using the published value of reaction and scattering cross-sections. The spectra of Po-Be, Pu-Be, Am-Be and Ra-Be arrived at agree well with the published spectra obtained by other methods. Spectrum from spontaneous fission of Cf-252 have also been measured and fitted to the expression N(E)=Esup(1/2)exp(-E/T). The fitted parameter T and spectral details agree well with those in published literature

  19. The Neutron Spectrum in a Uranium Tube

    International Nuclear Information System (INIS)

    Johansson, E.; Jonsson, E.; Lindberg, M.; Mednis, J.

    1963-10-01

    A series of experimental and theoretical investigations on neutron spectra in lattice cells has been started at the reactor R1. This report gives the results from the first one of these cells - one with a tube of natural -uranium surrounded by heavy water. In the measurements the cell was placed in the central, vertical channel of the reactor. The neutron spectrum from a lead scatterer in the uranium tube - outer diameter 49.2 mm, inner diameter 28.3 mm - was measured with a fast chopper in the energy region 0.01 to 100 eV. Subsidiary measurements indicated that the spectrum in the beam from the lead piece corresponds to the spectrum of the angular flux integrated over all angles. This correspondence is important for the interpretation of the experimental data. The thermal part of the spectrum was found to deviate significantly from a Maxwellian. However, the deviation is not very large, and one could use a Maxwellian, at least to give a rough idea of the hardness of the spectrum. For the present tube the temperature of this Maxwellian was estimated as 90 to 100 deg C above the moderator temperature (33 deg C). In the joining region the rise of the spectrum towards the thermal part is slower than for the cell boundary spectrum, measured earlier. In the epithermal region the limited resolution of the chopper has affected the measurements at the energies of the uranium resonances. However, the shape of the spectrum on the flanks of the first resonance in 238 U (6.68 eV) has been obtained accurately. In the theoretical treatment the THERMOS code with a free gas scattering model has been used. The energy region was 3.06 - 0.00025 eV. The agreement with the measurements is good for the thermal part - possibly the theoretical spectrum is a little softer than the experimental one. In the joining region the results from THERMOS are comparatively high - probably due to the scattering model used

  20. The Neutron Spectrum in a Uranium Tube

    Energy Technology Data Exchange (ETDEWEB)

    Johansson, E; Jonsson, E; Lindberg, M; Mednis, J

    1963-10-15

    A series of experimental and theoretical investigations on neutron spectra in lattice cells has been started at the reactor R1. This report gives the results from the first one of these cells - one with a tube of natural -uranium surrounded by heavy water. In the measurements the cell was placed in the central, vertical channel of the reactor. The neutron spectrum from a lead scatterer in the uranium tube - outer diameter 49.2 mm, inner diameter 28.3 mm - was measured with a fast chopper in the energy region 0.01 to 100 eV. Subsidiary measurements indicated that the spectrum in the beam from the lead piece corresponds to the spectrum of the angular flux integrated over all angles. This correspondence is important for the interpretation of the experimental data. The thermal part of the spectrum was found to deviate significantly from a Maxwellian. However, the deviation is not very large, and one could use a Maxwellian, at least to give a rough idea of the hardness of the spectrum. For the present tube the temperature of this Maxwellian was estimated as 90 to 100 deg C above the moderator temperature (33 deg C). In the joining region the rise of the spectrum towards the thermal part is slower than for the cell boundary spectrum, measured earlier. In the epithermal region the limited resolution of the chopper has affected the measurements at the energies of the uranium resonances. However, the shape of the spectrum on the flanks of the first resonance in {sup 238}U (6.68 eV) has been obtained accurately. In the theoretical treatment the THERMOS code with a free gas scattering model has been used. The energy region was 3.06 - 0.00025 eV. The agreement with the measurements is good for the thermal part - possibly the theoretical spectrum is a little softer than the experimental one. In the joining region the results from THERMOS are comparatively high - probably due to the scattering model used.

  1. Fast neutron spectrum measurement in the JMTR

    International Nuclear Information System (INIS)

    Sakurai, K.; Mizuho, M.

    1980-01-01

    Fast neutron spectrum measurement at positions of K-10 (fuel region), J-11 (first beryllium reflector region) and I-12 (second beryllium reflector region) in the JMTRC has been performed with the threshold detectors such as 107 Ag(n,n')sup(107m)Ag, 103 Rh(n,n')sup(103m)Rh, 115 In(n,n')sup(115m)In and 238 U(n,f)F.P. above 0.1 MeV energy region. The activity data have been analyzed by the unfolding code SAND-II using ANISN spectrum for initial spectrum. An effective cross section of 54 Fe(n,p) 54 Mn is calculated with Fabry's cross section data and SAND-II spectrum for K-10, J-11 and I-12. They are 76.7 mb, 69.0 mb and 68.2 mb for K-10, J-11 and I-12 respectively. These values agree with the effective cross sections (calculated by Fabry's cross section data and ANISN spectrum) within +-6%

  2. Neutron spectrum unfolding using computer code SAIPS

    International Nuclear Information System (INIS)

    Karim, S.

    1999-01-01

    The main objective of this project was to study the neutron energy spectrum at rabbit station-1 in Pakistan Research Reactor (PARR-I). To do so, multiple foils activation method was used to get the saturated activities. The computer code SAIPS was used to unfold the neutron spectra from the measured reaction rates. Of the three built in codes in SAIPS, only SANDI and WINDOWS were used. Contribution of thermal part of the spectra was observed to be higher than the fast one. It was found that the WINDOWS gave smooth spectra while SANDII spectra have violet oscillations in the resonance region. The uncertainties in the WINDOWS results are higher than those of SANDII. The results show reasonable agreement with the published results. (author)

  3. Prompt neutron spectrum of the spontaneous fission of californium-252

    International Nuclear Information System (INIS)

    Zamyatnin, Yu.S.; Kroshkin, N.I.; Korostylev, V.A.; Nefedov, V.N.; Ryazanov, D.K.; Starostov, B.I.; Semenov, A.F.

    1976-01-01

    The californium-252 spontaneous fission neutron spectrum was measured in the energy range of 0.01 to 10 MeV by the time-of-flight technique using various neutron detectors. The measurements of 252 Cf neutron spectrum at energies of 0.01 to 5 MeV were performed as a function of fission fragment kinetic energy. The mean neutron spectrum energy in the range of 0.7 to 10 MeV was found from the results of measurements. The irregularity in the 252 Cf neutron spectrum in the neutron energy range of less than 0.7 MeV compared to theoretical values is discussed. The mechanism of 252 Cf neutron emission is also discussed on the basis of neutron yield angle measurements. 12 references

  4. Beryllium phonon spectrum from cold neutron measurements

    International Nuclear Information System (INIS)

    Bulat, I.A.

    1979-01-01

    The inelastic coherent scattering of neutrons with the initial energy E 0 =4.65 MeV on the spectrometer according to the time of flight is studied in polycrystalline beryllium. The measurements are made for the scattering angles THETA=15, 30, 45, 60, 75 and 90 deg at 293 K. The phonon spectrum of beryllium, i-e. g(w) is reestablished from the experimental data. The data obtained are compared with the data of model calculations. It is pointed out that the phonon spectrum of beryllium has a bit excessive state density in the energy range from 10 to 30 MeV. It is caused by the insufficient statistical accuracy of the experiment at low energy transfer

  5. Neutron spectrum measurement in D + Be reaction

    CERN Document Server

    Abbasi-Davani, F; Aslani, G R; Etaati, G R; Koohi-Fayegh, R

    2002-01-01

    In this project the neutron spectra from the reaction of deuteron on beryllium nuclei is measured. The energies of deuterons were 7, 10, 13 and 15 MeV, and these measurements are performed at 10,30 and 50 degrees relative to the beam of deuterons. The detector used is 76 by 76 mm right circular cylinder of N E-213 liquid scintillator. The zero crossing technique is used for gamma discrimination. For the elimination of the background radiation, a Polyethylene block, 40 cm in thickness, with inserted cadmium sheets, and a lead block, 5 cm in thickness, were used. In order to obtain the background radiation spectrum, the latter blocks were placed between the target and the detector to eliminate neutron and gamma radiations reaching the detector directly. sup F ORIST sup c ode is used to unfold the neutron spectra from the measured pulse high t spectra and sup O 5S sup a nd sup R ESPMG sup c odes are used to obtain the detector response matrix.

  6. Effects of silicon cross section and neutron spectrum on the radial uniformity in neutron transmutation doping.

    Science.gov (United States)

    Kim, Haksung; Ho Pyeon, Cheol; Lim, Jae-Yong; Misawa, Tsuyoshi

    2012-01-01

    The effects of silicon cross section and neutron spectrum on the radial uniformity of a Si-ingot are examined experimentally with various neutron spectrum conditions. For the cross section effect, the numerical results using silicon single crystal cross section reveal good agreements with experiments within relative difference of 6%, whereas the discrepancy is approximately 20% in free-gas cross section. For the neutron spectrum effect, the radial uniformity in hard neutron spectrum is found to be more flattening than that in soft spectrum. Copyright © 2011 Elsevier Ltd. All rights reserved.

  7. Activation method for measurement of neutron spectrum parameters

    International Nuclear Information System (INIS)

    Efimov, B.V.; Demidov, A.M.; Ionov, V.S.; Konjaev, S.I.; Marin, S.V.; Bryzgalov, V.I.

    2007-01-01

    Experimental researches of spectrum parameters of neutrons at nuclear installations RRC KI are submitted. The installations have different designs of the cores, reflector, parameters and types of fuel elements. Measurements were carried out with use of the technique developed in RRC KI for irradiation resonance detectors UKD. The arrangement of detectors in the cores ensured possibility of measurement of neutron spectra with distinguished values of parameters. The spectrum parameters which are introduced by parametrical representation of a neutrons spectrum in the form corresponding to formalism Westcott. On experimental data were determinate absolute values of density neutron flux (DNF) in thermal and epithermal area of a spectrum (F t , f epi ), empirical dependence of temperature of neutron gas (Tn) on parameter of a rigidity of a spectrum (z), density neutron flux in transitional energy area of the spectrum. Dependences of spectral indexes of nuclides (UDy/UX), included in UKD, from a rigidity z and-or temperatures of neutron gas Tn are obtained.B Tools of mathematical processing of results are used for activation data and estimation of parameters of a spectrum (F t , f epi , z, Tn, UDy/UX). In the paper are presented some results of researches of neutron spectrum parameters of the nuclear installations (Authors)

  8. Experimental techniques for the consolidation of the neutron spectrum

    International Nuclear Information System (INIS)

    Chiaraviglio, N.; Bazzana, S.

    2013-01-01

    Unfolding techniques are widely known but their use is not widespread due to their complexity. Such procedure consists in the adjustment of calculated quantities to experimental results by the modification of the neutron spectrum, getting correction factors for the calculated quantities. In this work we describe the general procedure that must be executed for a neutron spectrum unfolding. (author) [es

  9. Investigation on innovative water reactor for flexible fuel cycle (FLWR). (1) Conceptual design

    International Nuclear Information System (INIS)

    Uchikawa, Sadao; Okubo, Tsutomu; Kugo, Teruhiko; Akie, Hiroshi; Nakano, Yoshihiko; Ohnuki, Akira; Iwamura, Takamichi

    2005-01-01

    A concept of Innovative Water Reactor for Flexible Fuel Cycle (FLWR) has been investigated in Japan Atomic Energy Research Institute (JAERI) in order to ensure sustainable energy supply in the future based on the well-experienced Light Water Reactor (LWR). The concept aims at effective and flexible utilization of uranium and plutonium resources through plutonium multiple recycling by two stages. In the first stage, the FLWR core realizes a high conversion type core concept, which is basically intended to keep the smooth technical continuity from current LWR and coming LWR-MOX technologies without significant gaps in technical point of view. The core in the second stage represents the Reduced-Moderation Water Reactor (RMWR) core concept, which realizes a high conversion ratio over 1.0 being useful for the long-term sustainable energy supply through plutonium multiple recycling based on the well-experienced LWR technologies. The key point is that the core concepts in both stages utilize the compatible and the same size fuel assemblies, and hence during the reactor operation period, the former concept can proceed to the latter in the same reactor system, corresponding flexibly to the expected change in the future circumstances of natural uranium resource, or establishment of economical reprocessing technology of MOX spent fuel. The FLWR is essentially a BWR-type reactor, and its core design is characterized by use of hexagonal-shaped fuel assemblies with the triangular-lattice fuel rod configuration of highly enriched MOX fuel, control rods with Y-shaped blades, and a short and flat core design. Detailed investigations have been performed on the core design, in conjunction with the other related studies such as on thermal hydraulics in the tight lattice core including experimental activities, and the results obtained so far have shown the proposed concept is feasible and promising. (author)

  10. Spectrum and H(10) of secondary neutrons around Linacs

    International Nuclear Information System (INIS)

    Ortiz H, A.; Hernandez A, B.; Vega C, H. R.; Hernandez D, V. M.; Rivera M, T.

    2009-10-01

    Neutron spectrum and ambient dose equivalent has been measured around two 10 MV linear accelerators. Accelerators are Siemens, one is a Mevatron model while another is the Primus. Main differences between those models are the beam collimator and the vault room. Here, Bonner sphere spectrometer with a passive thermal neutron detector has been utilized to measure the neutron spectrum inside the vault. Using an active detector the neutron spectrum was measured by the vaults door of both accelerators. With a neutron area monitor the dose equivalent was measured by the doors. Neutron strength, total fluence rate and ambient dose equivalent were compared, from this was found that shielding conditions are better in the Primus model. (Author)

  11. Neutron spectrum determination by activation method in fast neutron fields at the RB reactor

    International Nuclear Information System (INIS)

    Sokcic-Kostic, M.; Pesic, M.; Antic, D.

    1994-01-01

    The fast neutron fields of the RB reactor are presented in this paper. The activation method for spectrum determination is described and explained. The obtained results for intermediate and fast spectrum are given and discussed. (author)

  12. Neutron spectrum determination by activation method in fast neutron fields at the RB reactors

    International Nuclear Information System (INIS)

    Sokcic-Kostic, M.S.; Pesic, M.P.; Antic, D.P.

    1994-01-01

    The fast neutron fields of the RB reactor are presented in this paper. The activation method for spectrum determination is described and explained. The obtained results for intermediate and fast spectrum are given and discussed. (authors). 7 refs., 3 tabs

  13. New fission-neutron-spectrum representation for ENDF

    International Nuclear Information System (INIS)

    Madland, D.G.

    1982-04-01

    A new representation of the prompt fission neutron spectrum is proposed for use in the Evaluated Nuclear Data File (ENDF). The proposal is made because a new theory exists by which the spectrum can be accurately predicted as a function of the fissioning nucleus and its excitation energy. Thus, prompt fission neutron spectra can be calculated for cases where no measurements exist or where measurements are not possible. The mathematical formalism necessary for application of the new theory within ENDF is presented and discussed for neutron-induced fission and spontaneous fission. In the case of neutron-induced fission, expressions are given for the first-chance, second-chance, third-chance, and fourth-chance fission components of the spectrum together with that for the total spectrum. An ENDF format is proposed for the new fission spectrum representation, and an example of the use of the format is given

  14. Neutron spectrum measurement using rise-time discrimination method

    International Nuclear Information System (INIS)

    Luo Zhiping; Suzuki, C.; Kosako, T.; Ma Jizeng

    2009-01-01

    PSD method can be used to measure the fast neutron spectrum in n/γ mixed field. A set of assemblies for measuring the pulse height distribution of neutrons is built up,based on a large volume NE213 liquid scintillator and standard NIM circuits,through the rise-time discrimination method. After that,the response matrix is calculated using Monte Carlo method. The energy calibration of the pulse height distribution is accomplished using 60 Co radioisotope. The neutron spectrum of the mono-energetic accelerator neutron source is achieved by unfolding process. Suggestions for further improvement of the system are presented at last. (authors)

  15. Study of the environmental neutron spectrum at Zacatecas city

    International Nuclear Information System (INIS)

    Vega C, H.R.

    2003-01-01

    The environmental neutron spectrum has been measured at Zacatecas City in Mexico. Neutron spectrum was unfolded from count rates obtained with a multisphere neutron spectrometer with a Li I(Eu) scintillator. With the spectrum information the ambient dose equivalent and the isotropic effective dose were calculated. A model based upon the geomagnetic latitude and the altitude above sea level, that allows to estimate the neutron fluence rate is proposed, the model results are compared with total neutron fluences measured at several locations worldwide. Environmental neutron spectrum shows peaks at 1 and 100 MeV as well as a relevant amount of low energy neutrons. The neutron fluence rate was 65 ± 3 cm -2 -h -1 , producing 13.7 ± 0.6 n Sv-h -1 due to ambient dose equivalent rate and an isotropic effective dose rate of 14.1 ± 0.6 n Sv-h -1 . Neutron fluence rates predicted with the model are in agreement with those reported in the literature. (Author)

  16. Study of the environmental neutron spectrum at Zacatecas city

    Energy Technology Data Exchange (ETDEWEB)

    Vega C, H.R. [Universidad Autonoma de Zacatecas, Cuerpo Academico de Radiobiologia, A.P. 336, 98000 Zacatecas (Mexico)

    2003-07-01

    The environmental neutron spectrum has been measured at Zacatecas City in Mexico. Neutron spectrum was unfolded from count rates obtained with a multisphere neutron spectrometer with a Li I(Eu) scintillator. With the spectrum information the ambient dose equivalent and the isotropic effective dose were calculated. A model based upon the geomagnetic latitude and the altitude above sea level, that allows to estimate the neutron fluence rate is proposed, the model results are compared with total neutron fluences measured at several locations worldwide. Environmental neutron spectrum shows peaks at 1 and 100 MeV as well as a relevant amount of low energy neutrons. The neutron fluence rate was 65 {+-} 3 cm{sup -2}-h{sup -1}, producing 13.7 {+-} 0.6 n Sv-h{sup -1} due to ambient dose equivalent rate and an isotropic effective dose rate of 14.1 {+-} 0.6 n Sv-h{sup -1}. Neutron fluence rates predicted with the model are in agreement with those reported in the literature. (Author)

  17. Determining of the intermediate neutron spectrum in fast neutron field at the RB reactor

    International Nuclear Information System (INIS)

    Sokcic-Kostic, M.; Pesic, M.; Antic, D.

    1987-01-01

    The activation method for intermediate neutron spectrum determination is given in this paper. The intermediate neutron spectrum in experimental fuel channel (EFC) at the RB reactor is determined om the basis of this method. The results of measurements are treated with PRAG code and will be treated with KRIFIT and TENET codes that are also developed. (author)

  18. Least-squares adjustment of a 'known' neutron spectrum: The importance of the covariance matrix of the input spectrum

    International Nuclear Information System (INIS)

    Mannhart, W.

    1986-01-01

    Based on the responses of 25 different neutron activation detectors, the neutron spectrum of Cf-252 hs been adjusted with least-squares methods. For a fixed input neutron spectrum, the covariance matrix of this spectrum has been systematically varied to investigate the influence of this matrix on the final result. The investigation showed that the adjusted neutron spectrum is rather sensitive to the structure of the covariance matrix for the input spectrum. (author)

  19. Effects of silicon cross section and neutron spectrum on the radial uniformity in neutron transmutation doping

    International Nuclear Information System (INIS)

    Kim, Haksung; Ho Pyeon, Cheol; Lim, Jae-Yong; Misawa, Tsuyoshi

    2012-01-01

    The effects of silicon cross section and neutron spectrum on the radial uniformity of a Si-ingot are examined experimentally with various neutron spectrum conditions. For the cross section effect, the numerical results using silicon single crystal cross section reveal good agreements with experiments within relative difference of 6%, whereas the discrepancy is approximately 20% in free-gas cross section. For the neutron spectrum effect, the radial uniformity in hard neutron spectrum is found to be more flattening than that in soft spectrum. - Highlights: ► The effects of silicon cross section and neutron spectrum on the radial uniformity in NTD were experimentally investigated. ► The numerical results using silicon single crystal cross section reveal good agreements. ► The radial uniformity in hard neutron spectrum was more flat than that in soft spectrum. ► The silicon single crystal cross section and hard neutron spectrum are recommended for numerical analyses and radial uniformity flattening in NTD, respectively.

  20. A neutron spectrum unfolding code based on iterative procedures

    International Nuclear Information System (INIS)

    Ortiz R, J. M.; Vega C, H. R.

    2012-10-01

    In this work, the version 3.0 of the neutron spectrum unfolding code called Neutron Spectrometry and Dosimetry from Universidad Autonoma de Zacatecas (NSDUAZ), is presented. This code was designed in a graphical interface under the LabVIEW programming environment and it is based on the iterative SPUNIT iterative algorithm, using as entrance data, only the rate counts obtained with 7 Bonner spheres based on a 6 Lil(Eu) neutron detector. The main features of the code are: it is intuitive and friendly to the user; it has a programming routine which automatically selects the initial guess spectrum by using a set of neutron spectra compiled by the International Atomic Energy Agency. Besides the neutron spectrum, this code calculates the total flux, the mean energy, H(10), h(10), 15 dosimetric quantities for radiation protection porpoises and 7 survey meter responses, in four energy grids, based on the International Atomic Energy Agency compilation. This code generates a full report in html format with all relevant information. In this work, the neutron spectrum of a 241 AmBe neutron source on air, located at 150 cm from detector, is unfolded. (Author)

  1. Some neutronics of innovative subcritical assembly with fast neutron spectrum

    International Nuclear Information System (INIS)

    Kiyavitskaya, H.; Fokov, Yu.; Rutkovskaya, Ch.; Sadovich, S.; Kasuk, D.; Gohar, Y.; Bolshinsky, I.

    2013-01-01

    Conclusion: • New assembly can be used to: • develop the experimental techniques and adapt the existing ones for monitoring the sub-criticality level, neutron spectra measurements, etc; • study the spatial kinetics of sub-critical and critical systems with fast neutron spectra; • measure the transmutation reaction rates of minor-actinides etc

  2. Spectrum shaping of accelerator-based neutron beams for BNCT

    CERN Document Server

    Montagnini, B; Esposito, J; Giusti, V; Mattioda, F; Varone, R

    2002-01-01

    We describe Monte Carlo simulations of three facilities for the production of epithermal neutrons for Boron Neutron Capture Therapy (BNCT) and examine general aspects and problems of designing the spectrum-shaping assemblies to be used with these neutron sources. The first facility is based on an accelerator-driven low-power subcritical reactor, operating as a neutron amplifier. The other two facilities have no amplifier and rely entirely on their primary sources, a D-T fusion reaction device and a conventional 2.5 MeV proton accelerator with a Li target, respectively.

  3. Measurement of 235U fission spectrum-averaged cross sections and neutron spectrum adjusted with the activation data

    International Nuclear Information System (INIS)

    Kobayashi, Katsuhei; Kobayashi, Tooru

    1992-01-01

    The 235 U fission spectrum-averaged cross sections for 13 threshold reactions were measured with the fission plate (27 cm in diameter and 1.1 cm thick) at the heavy water thermal neutron facility of the Kyoto University Reactor. The Monte Carlo code MCNP was applied to check the deviation from the 235 U fission neutron spectrum due to the room-scattered neutrons, and it was found that the resultant spectrum was close to that of 235 U fission neutrons. Supplementally, the relations to derive the absorbed dose rates with the fission plate were also given using the calculated neutron spectra and the neutron Kerma factors. Finally, the present values of the fission spectrum-averaged cross sections were employed to adjust the 235 U fission neutron spectrum with the NEUPAC code. The adjusted spectrum showed a good agreement with the Watt-type fission neutron spectrum. (author)

  4. Neutron spectrum for neutron capture therapy in boron

    International Nuclear Information System (INIS)

    Medina C, D.; Soto B, T. G.; Baltazar R, A.; Vega C, H. R.

    2016-10-01

    Glioblastoma multiforme is the most common and aggressive of brain tumors and is difficult to treat by surgery, chemotherapy or conventional radiation therapy. One treatment alternative is the Neutron Capture Therapy in Boron, which requires a beam modulated in neutron energy and a drug with 10 B able to be fixed in the tumor. When the patients head is exposed to the neutron beam, they are captured by the 10 B and produce a nucleus of 7 Li and an alpha particle whose energy is deposited in the cancer cells causing it to be destroyed without damaging the normal tissue. One of the problems associated with this therapy is to have an epithermal neutrons flux of the order of 10 9 n/cm 2 -sec, whereby irradiation channels of a nuclear research reactor are used. In this work using Monte Carlo methods, the neutron spectra obtained in the radial irradiation channel of the TRIGA Mark III reactor are calculated when inserting filters whose position and thickness have been modified. From the arrangements studied, we found that the Fe-Cd-Al-Cd polyethylene filter yielded a ratio between thermal and epithermal neutron fluxes of 0.006 that exceeded the recommended value (<0.05), and the dose due to the capture gamma rays is lower than the dose obtained with the other arrangements studied. (Author)

  5. Pulsed White Spectrum Neutron Generator for Explosive Detection

    International Nuclear Information System (INIS)

    King, Michael J.; Miller, Gill T.; Reijonen, Jani; Ji, Qing; Andresen, Nord; Gicquel, Frederic; Kavlas, Taneli; Leung, Ka-Ngo; Kwan, Joe

    2008-01-01

    Successful explosive material detection in luggage and similar sized containers is a critical issue in securing the safety of all airline passengers. Tensor Technology Inc. has recently developed a methodology that will detect explosive compounds with pulsed fast neutron transmission spectroscopy. In this scheme, tritium beams will be used to generate neutrons with a broad energy spectrum as governed by the T(t,2n)4He fission reaction that produces 0-9 MeV neutrons. Lawrence Berkeley National Laboratory (LBNL), in collaboration with Tensor Technology Inc., has designed and fabricated a pulsed white-spectrum neutron source for this application. The specifications of the neutron source are demanding and stringent due to the requirements of high yield and fast pulsing neutron emission, and sealed tube, tritium operation. In a unique co-axial geometry, the ion source uses ten parallel rf induction antennas to externally couple power into a toroidal discharge chamber. There are 20 ion beam extraction slits and 3 concentric electrode rings to shape and accelerate the ion beam into a titanium cone target. Fast neutron pulses are created by using a set of parallel-plate deflectors switching between +-1500 volts and deflecting the ion beams across a narrow slit. The generator is expected to achieve 5 ns neutron pulses at tritium ion beam energies between 80-120 kV. First experiments demonstrated ion source operation and successful beam pulsing

  6. Experimental study of modification of neutron spectrum using filters

    International Nuclear Information System (INIS)

    Kobayashi, H.; Matsubayashi, M.; Brenizer, J.S. Jr.; Lindsay, J.T.

    1996-01-01

    Filter effects for continuum thermal neutron beams were experimentally studied by means of an effective energy. Be, Bi and Pb were used as filter materials to examine the energy shift of the spectrum. It was found that the effective energy of a thermal neutron beam is easily lowered into a sub-thermal region (down to 10 meV) by use of a filter system without any cooling system. The effectiveness and its applicability will be discussed in this study. (orig.)

  7. 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

  8. Neutron spectrum measurements from a neutron guide tube facility at the ETRR-1 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Maayouf, R M.A.; El-Sayed, L A.A.; El-Kady, A S.I. [Reactor and Neutron Physics Dept., NRC, Atomic Energy Authority, Cairo (Egypt)

    1997-12-31

    The present work deals with measurements of the neutron spectrum emitted from a neutron guide tube (NGT) recently installed at one of the ETRR-1 reactor horizontal channels designed to deliver thermal neutrons, free from fast neutrons and gamma ray background, to a fourier reverse-time-of-flight (RTOF) diffractometer. The measurements were performed using a {sup 6} Li glass scintillation detector combined with a multichannel analyzer set at channel width 4 M sec and installed at 3.4 m from a disc Fermi chopper. Also a theoretical model was specially developed for the neutron spectrum calculations. According to the model developed, the spectrum calculated was found to be in good agreement with the measured one. It was found, both from measurements and calculations, that the spectrum emitted from the NGT covers, after transmission through a fourier chopper, neutron wavelengths from 1-4 A adequate for neutron diffraction measurements at D values between 0.71-2.9 A respectively. 6 FIGS.

  9. Neutron metrology file NMF-90. An integrated database for performing neutron spectrum adjustment calculations

    International Nuclear Information System (INIS)

    Kocherov, N.P.

    1996-01-01

    The Neutron Metrology File NMF-90 is an integrated database for performing neutron spectrum adjustment (unfolding) calculations. It contains 4 different adjustment codes, the dosimetry reaction cross-section library IRDF-90/NMF-G with covariances files, 6 input data sets for reactor benchmark neutron fields and a number of utility codes for processing and plotting the input and output data. The package consists of 9 PC HD diskettes and manuals for the codes. It is distributed by the Nuclear Data Section of the IAEA on request free of charge. About 10 MB of diskspace is needed to install and run a typical reactor neutron dosimetry unfolding problem. (author). 8 refs

  10. Measurements of the energy spectrum of backscattered fast neutrons

    International Nuclear Information System (INIS)

    Segal, Y.

    1976-03-01

    Experimental measurements have been made of the energy spectra of neutrons transmitted through slabs of iron, lead and perspex for incident neutron energies of 0.5, 1.0, 1.5 and 1.8 MeV. The neutron energy measurements were made using a He-3 spectrometer. The dependence of the neutrons energy spectrum as a function of scattering thickness was determined. The neutrons source used was a 3MeV Van de Graaff accelerator with a tritium target using the H 3 (p,n) He 3 reaction. The results obtained by the investigator on energy dependence of transmitted neutrons as a function of thickness of scattering material were compared, where possible, with the results obtained by other workers. The comparisons indicated good agreement. The experiment's results are compared with MORSE Monte Carlo calculated values. It is worthwhile to note that direct comparison between measured cross section values and the recommended ones are very far from satisfactory. In almost all cases the calculated spectrum is harder than the experimental one, a situation common to the penetrating and the back-scattered flux

  11. Neutron Spectrum Measurements from Irradiations at NCERC

    Energy Technology Data Exchange (ETDEWEB)

    Jackman, Kevin Richard [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Mosby, Michelle A. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Bredeweg, Todd Allen [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Hutchens, Gregory Joe [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); White, Morgan Curtis [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2015-04-15

    Several irradiations have been conducted on assemblies (COMET/ZEUS and Flattop) at the National Criticality Experiments Research Center (NCERC) located at the Nevada National Security Site (NNSS). Configurations of the assemblies and irradiated materials changed between experiments. Different metallic foils were analyzed using the radioactivation method by gamma-ray spectrometry to understand/characterize the neutron spectra. Results of MCNP calculations are shown. It was concluded that MCNP simulated spectra agree with experimental measurements, with the caveats that some data are limited by statistics at low-energies and some activation foils have low activities.

  12. Neutron energy spectrum in graphite blankets of fusion reactors

    International Nuclear Information System (INIS)

    Tsechanski, A.

    1981-09-01

    Neutron flux measurements were performed in a graphite stack and compared with calculations made with a two dimensional transport computer code. In the present work it is observed that the calculated spectrum in the elastic and inelastic scattering ranges (the first collision range in both cases), is sensitive to details of the angular distribution of these neutrons. Regarding the discrepancies in the elastic scattering range it is concluded that the microscopic cross section library ENDF/B-IV overestimates the large angle scattering (back scattering) as can be seen from comparison of measured and calculated spectra. The two most important conclusions of the present work are: 1. Inelastic scattering interaction of D-T neutrons in graphite cannot be calculated without a proper account of energy-angle correlation. 2. An experimental setup supplying monoenergetic collimated D-T neutrons constitutes a sensitive although indirect means for measuring angular distributions in inelastic and elastic scattering

  13. DIFFERENTIAL SPECTRUM OF NEUTRONS AT CHACALTAYA-BOLIVIA

    International Nuclear Information System (INIS)

    Mayta, R.; Zanini, A.; Ticona, R.; Velarde, A.

    2009-01-01

    We describe the Neutron Spectrometer Experiment installed at Chacaltaya Cosmic Rays Observatory (68 deg. O, 16.2 deg. S), located in Bolivia, at 5230 m.a.s.l. This experimental system is constituted by passive detectors which register the flux of neutrons, in an energy range of 10 KeV-20 MeV. Using the unfolding code BUNTO a peak around 1 MeV of the characteristic spectrum of neutrons was obtained. Experimental values, observed during April of 2008, are compared with similar ones carried out in 1997 at the same place, in order to look for eventual changes due to local atmosphere. A similar experiment was also carried up at the Laboratory of Testa Grigia-Italy (45.56 deg. N, 7.42 deg. E,. 3480 m.a.l.s). Data of both stations allow us to compare the spectra in order to explain the difference of neutron flux of these two stations.

  14. Updated neutron spectrum characterization of SNL baseline reactor environments

    International Nuclear Information System (INIS)

    Griffin, P.J.; Kelly, J.G.; Vehar, D.W.

    1994-04-01

    This document provides SAND-II and MANIPULATE output listings from calculations used to derive the new spectrum-averaged integral parameters that were reported in volume 1. When used in conjunction with volume 1, this document provides an audit trail for the neutron radiation field characterization and supports current quality assurance initiatives. This document provides detailed information on the neutron spectrum characteristics of the primary Sandia National Laboratories' (SNL) reactor environments. The information in this volume is not intended for the casual user of the SNL reactor facilities. This detailed characterization of the neutron and gamma environments at the Sandia Pulsed Reactor (SPR) and the Annular Core Research Reactor (ACRR) is provided to aid the users who wish to convert the information given in the Radiation Metrology Laboratory (RML) dosimetry reports into other (non-silicon) measures of neutron damage. The spectra provided in these appendices can be used as a source term for Monte Carlo radiation transport calculations to study the impact of experimenter's test package on the neutron environment

  15. Modulation of the neutron spectrum for NCTB; Modulacion del espectro de neutrones para TCNB

    Energy Technology Data Exchange (ETDEWEB)

    Medina C, D.; Letechipia de L, C.; Vega C, H. R., E-mail: dmedina_c@hotmail.com [Universidad Autonoma de Zacatecas, Unidad Academica de Estudios Nucleares, Cipres No.10, Fracc. La Penuela, 98068 Zacatecas, Zac. (Mexico)

    2016-09-15

    Cancer is characterized by the uncontrolled growth of differentiated cells in a part of the organism. Currently in the world there are millions of people living with cancer. Glioblastoma multiform e is the most common and most aggressive of brain tumors and is very difficult to treat by conventional surgery, chemotherapy or radiation. The only viable alternative is its treatment through Neutron Capture Therapy in Boron (NCTB), since is a selective therapy that requires a drug with {sup 10}B (a non-radioactive isotope of boron) and a modulated neutron beam. Thermal neutrons are captured by {sup 10}B, because has a large effective section of thermal neutron absorption, in an exothermic reaction forming the nucleus composed of {sup 11}B in an excited state that induces its cleavage in two nuclei: {sup 7}Li and alpha particle ({sup 4}He). This process causes the destruction of cancer cells by direct DNA damage, without damaging normal tissue. One of the problems associated with this therapy is to have a neutron beam with adequate flow and spectrum. The neutron spectrum must be moderated and filtered from the characteristics of the source. To this end, the main sources of neutrons are nuclear research reactors and particle accelerators. The intensity of the flow should be 2 x 10 E{sup 9} n/cm{sup 2}.s, in order to treat the patient in a reasonable time; thus, is interesting to design filters for a radial beam of a TRIGA reactor, where materials such as Cd, Al, Fe and polyethylene are being implemented in the interest of having a spectrum with which the therapy can be implemented. For this design is being played with the position of the materials, to be able to see the behavior of the spectrum and thus choose some arrangement as indicated, of course taking into account the doses of both neutrons and gammas. (Author)

  16. Little Boy neutron spectrum below 3 MeV

    International Nuclear Information System (INIS)

    Evans, A.E.; Bennett, E.F.; Yule, T.J.

    1984-01-01

    The leakage neutron spectrum from the Little Boy replica has been measured from 12 keV to 3 MeV using a high-resolution 3 He ionization chamber, and from 1 keV to 3 MeV using proton-recoil proportional counters. The 3 He-spectrometer measurements were made at distances of 0.75 and 2.0 m from the active center and at angles of 0 0 , 45 0 , and 90 0 with respect to the axis of the assembly. Proton-recoil measurments were made at 90 0 to the assembly axis at distances of 0.75 and 2.0 m, with a shielded measurement made at 2.0 m to estimate background due to scattering. The 3 He spectrometer was calibrated at Los Alamos using monoenergetic 7 Li(p,n) 7 Be neutrons to generate a family of response functions. The proton-recoil counters were calibrated at Argonne by studying the capture of thermal neutrons by nitrogen in the counters, by observation of the 24-keV neutron resonance in iron, and by relating to the known hydrogen content of the counters. The neutron spectrum from Little Boy was found to be highly structured, with peaks corresponding to minima in the iron total neutron cross section. In particular, influence of the 24-keV iron window was evident in both sets of spectra. The measurements provide information for dosimetry calculations and also a valuable intercomparison of neutron spectrometry using the two different detector types. Spectra measured with both detectors are in essential agreement. 8 references, 7 figures, 2 tables

  17. NRSC, Neutron Resonance Spectrum Calculation System

    International Nuclear Information System (INIS)

    Leszczynski, Francisco

    2004-01-01

    1 - Description of program or function: The NRSC system is a package of four programs for calculating detailed neutron spectra and related quantities, for homogeneous mixtures of isotopes and cylindrical reactor pin cells, in the energy resonance region, using ENDF/B evaluated nuclear data pre-processed with NJOY or Cullen's codes up to the Doppler Broadening and unresolved resonance level. 2 - Methods: NRSC consists of four programs: GEXSCO, RMET21, ALAMBDA and WLUTIL. GEXSCO prepares the nuclear data from ENDF/B evaluated nuclear data pre-processed with NJOY or Cullen's codes up to the Doppler Broadening or unresolved resonance level for RMET21 input. RMET21 calculates spectra and related quantities for homogeneous mixtures of isotopes and cylindrical reactor pin cells, in the energy resonance region, using slowing-down algorithms and, in the case of pin cells, the collision probability method. ALAMBDA obtains lambda factors (Goldstein-Cohen intermediate resonance factors in the formalism of WIMSD code) of different isotopes for including on WIMSD-type multigroup libraries for WIMSD or other cell-codes, from output of RMET21 program. WLUTIL is an auxiliary program for extracting tabulated parameters related with RMET21 program calculations from WIMSD libraries for comparisons, and for producing new WIMSD libraries with parameters calculated with RMET21 and ALAMBDA programs. 3 - Restrictions on the complexity of the problem: GEXSCO program has fixed array dimensions that are suitable for processing all reasonable outputs from nuclear data pre-processing programs. RMET21 program uses variable dimension method from a fixed general array. ALAMBDA and WLUTIL programs have fixed arrays that are adapted to standard WIMSD libraries. All programs can be easily modified to adapt to special requirements

  18. Effect of normalization on the neutron spectrum adjustment procedure

    International Nuclear Information System (INIS)

    Zsolnay, E.M.; Zijp, W.L.; Nolthenius, H.J.

    1983-10-01

    Various computer programs currently applied for neutron spectrum adjustment based on multifoil activation data, use different ways to determine the normalization factor to be applied to an unnormalized input spectrum. The influence is shown of the various definitions of the normalization factor on the adjusted results for the case of the ORR and YAYOI spectra considered in the international REAL-80 exercise. The actual expression for defining the normalization factor is more important than previously assumed. The theory of the generalized least squares principle provides an optimal definition for the normalization factor

  19. Modeling the Complete Gravitational Wave Spectrum of Neutron Star Mergers.

    Science.gov (United States)

    Bernuzzi, Sebastiano; Dietrich, Tim; Nagar, Alessandro

    2015-08-28

    In the context of neutron star mergers, we study the gravitational wave spectrum of the merger remnant using numerical relativity simulations. Postmerger spectra are characterized by a main peak frequency f2 related to the particular structure and dynamics of the remnant hot hypermassive neutron star. We show that f(2) is correlated with the tidal coupling constant κ(2)^T that characterizes the binary tidal interactions during the late-inspiral merger. The relation f(2)(κ(2)^T) depends very weakly on the binary total mass, mass ratio, equation of state, and thermal effects. This observation opens up the possibility of developing a model of the gravitational spectrum of every merger unifying the late-inspiral and postmerger descriptions.

  20. An investigation of TRU recycling with various neutron spectrums

    International Nuclear Information System (INIS)

    Yong-Nam, Kim; Hong-Chul, Kim; Chi-Young, Han; Jong-Kyung, Kim; Won-Seok Park

    2003-01-01

    This study is intended to evaluate the dependency of TRU recycling characteristics on the neutron spectrum shift in a Pb-Bi cooled core. Considering two Pb-Bi cooled cores with the soft and the hard spectrum, respectively, various characteristics of the recycled core are carefully examined and compared with each other. Assuming very simplified fuel cycle management with the homogeneous and single batch fuel loading, the burn-up calculations are performed until the recycled core reached to the (quasi-) equilibrium state. The mechanism of TRU recycling toward the equilibrium is analysed in terms of burn-up reactivity and the isotopic compositions of TRU fuel. In the comparative analyses, the difference in the recycling behaviour between the two cores is clarified. In addition, the basic safety characteristics of the recycled core are also discussed in terms of the Doppler coefficient, the coolant loss reactivity coefficient, and the effective delayed neutron fraction. (author)

  1. Consolidation of the neutron spectrum in the RA-6 reactor

    International Nuclear Information System (INIS)

    Bazzana, S.; Chiaraviglio, N.

    2013-01-01

    Unfolding procedures can be used to determine the neutron or gamma spectrum in a multigroup structure from experimental and calculation results. In this way, it is possible to adjust with high reliability magnitudes that cannot be directly measured. For neutron unfolding it is necessary the use of a set of detectors with different energetic response. In this work we describe two unfolding experiences in different positions of the RA-6 reactor of the Bariloche Atomic Centre. One of them consisted in the unfolding in an incore position and the other one in the BNCT facility beam.Experimental techniques and neutron detectors for each experience are described along with the correction factors that must be taken into account for each experience. In both cases there is good agreement between measured and adjusted quantities. (author) [es

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

    International Nuclear Information System (INIS)

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

    2002-01-01

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

  3. Neutron spectrum in small iron pile surrounded by lead reflector

    International Nuclear Information System (INIS)

    Kimura, Itsuro; Hayashi, S.A.; Kobayashi, Katsuhei; Matsumura, Tetsuo; Nishihara, Hiroshi.

    1978-01-01

    In order to save the quantity of sample material, a possibility to assess group constants of a reactor material through measurement and analysis of neutron spectrum in a small sample pile surrounded by a reflector of heavy moderator, was investigated. As the sample and the reflector, we chose iron and lead, respectively. Although the time dispersion in moderation of neutrons was considerably prolonged by the lead reflector, this hardly interferes with the assessment of group constants. Theoretical calculation revealed that both the neutron flux spectrum and the sensitivity coefficient of group constants in an iron sphere, 35 cm in diameter surrounded by the lead reflector, 25 cm thick, were close to those of the bare iron sphere, 108 cm in diameter. The neutron spectra in a small iron pile surrounded by a lead reflector were experimentally obtained by the time-of-flight method with an electron linear accelerator and the result was compared with the predicted values. It could be confirmed that a small sample pile surrounded by a reflector, such as lead, was as useful as a much larger bulk pile for the assessment of group constants of a reactor material. (auth.)

  4. Neutron spectrum survey around the cyclotron of IEN/Brazilian CNEN: calibration of neutron personnel dosemeter

    International Nuclear Information System (INIS)

    Fajardo, P.W.

    1991-01-01

    The albedo neutron dosimeter is calibrated directly at the work place due to its high energy dependence. This thesis deals with the study, analysis and application of neutron measurement techniques in order to obtain information about the neutron spectrum and neutron dose equivalent at several representative working places of the cyclotron laboratory of the Nuclear Engineering Institute (IEN). These data are employed mainly in the calibration of the brazilian albedo neutron dosimeter. Bonner spheres and foil activation were used in neutron spectra measurements and the neutron dose equivalents were measured with the single sphere albedo technique. BF 3 and 3 He proportional detectors and 6 LiI scintillation detector were also used in these measurements. The single sphere technique turned out to be more appropriate for neutron dosimetry for calibrating the albedo dosimeter in the varying fields of the cyclotron. Calibration the albedo dosimeter in the varying fields of the cyclotron. Calibration factors were found for routine applications, when the workers are protected by shielding and for radiological accident applications, in the case that a worker is exposed inside the cyclotron room. In all situations the performance of the brazilian albedo dosimeter is compared with that of the german albedo dosimeters. (author)

  5. Approximation for the adjoint neutron spectrum; Aproximacao para o espectro adjunto de neutrons

    Energy Technology Data Exchange (ETDEWEB)

    Suster, Luis Carlos; Martinez, Aquilino Senra; Silva, Fernando Carvalho da [Universidade Federal, Rio de Janeiro, RJ (Brazil). Coordenacao dos Programas de Pos-graduacao de Engenharia. Programa de Engenharia Nuclear

    2002-07-01

    The proposal of this work is the determination of an analytical approximation which is capable to reproduce the adjoint neutron flux for the energy range of the narrow resonances (NR). In a previous work we developed a method for the calculation of the adjoint spectrum which was calculated from the adjoint neutron balance equations, that were obtained by the collision probabilities method, this method involved a considerable quantity of numerical calculation. In the analytical method some approximations were done, like the multiplication of the escape probability in the fuel by the adjoint flux in the moderator, and after these approximations, taking into account the case of the narrow resonances, were substituted in the adjoint neutron balance equation for the fuel, resulting in an analytical approximation for the adjoint flux. The results obtained in this work were compared to the results generated with the reference method, which demonstrated a good and precise results for the adjoint neutron flux for the narrow resonances. (author)

  6. Incident spectrum determination for time-of-flight neutron powder diffraction data analysis

    International Nuclear Information System (INIS)

    Hodges, J. P.

    1998-01-01

    Accurate characterization of the incident neutron spectrum is an important requirement for precise Rietveld analysis of time-of-flight powder neutron diffraction data. Without an accurate incident spectrum the calculated model for the measured relative intensities of individual Bragg reflections will possess systematic errors. We describe a method for obtaining an accurate numerical incident spectrum using data from a transmitted beam monitor

  7. 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)

  8. 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.

  9. Gamma spectrum following neutron capture in {sup 167}Er

    Energy Technology Data Exchange (ETDEWEB)

    Visser, D.; Khoo, T.L.; Lister, C.J. [and others

    1995-08-01

    Statistical decay from a highly excited state samples all the lower-lying states and, hence, provides a sensitive measure of the level density. Pairing has a major impact on the level density, e.g. creating a pair gap between the 0- and 2-quasiparticle configurations. Hence the shape of the statistical spectrum contains information on pairing, and can be used to provide information on the reduction of pairing with thermal excitation energy. For this reason, we measured the complete spectrum of {gamma}rays following thermal neutron capture in {sup 167}Er. The experiment was performed at the Brookhaven reactor using Compton-suppressed Ge detectors from TESSA. The spectrum, which was corrected for detector response and efficiency, reveals primary (first-step, high-energy) transitions up to nearly 8 MeV, secondary (last-step, lower-energy) transitions, as we as a continuous statistical component. Effort was expanded to identify all lines from contaminant sources and an upper limit of 5% was tentatively set for their contributions. The spectral shape of the statistical spectrum will be compared with theoretical spectra obtained from a calculation of pairing which accounts for a stepwise reduction of the pair correlations as the number of quasiparticles increases. The primary lines which decay directly to the near-yrast states will also be used to deduce the level densities.

  10. A genetic algorithm based method for neutron spectrum unfolding

    International Nuclear Information System (INIS)

    Suman, Vitisha; Sarkar, P.K.

    2013-03-01

    An approach to neutron spectrum unfolding based on a stochastic evolutionary search mechanism - Genetic Algorithm (GA) is presented. It is tested to unfold a set of simulated spectra, the unfolded spectra is compared to the output of a standard code FERDOR. The method was then applied to a set of measured pulse height spectrum of neutrons from the AmBe source as well as of emitted neutrons from Li(p,n) and Ag(C,n) nuclear reactions carried out in the accelerator environment. The unfolded spectra compared to the output of FERDOR show good agreement in the case of AmBe spectra and Li(p,n) spectra. In the case of Ag(C,n) spectra GA method results in some fluctuations. Necessity of carrying out smoothening of the obtained solution is also studied, which leads to approximation of the solution yielding an appropriate solution finally. Few smoothing techniques like second difference smoothing, Monte Carlo averaging, combination of both and gaussian based smoothing methods are also studied. Unfolded results obtained after inclusion of the smoothening criteria are in close agreement with the output obtained from the FERDOR code. The present method is also tested on a set of underdetermined problems, the outputs of which is compared to the unfolded spectra obtained from the FERDOR applied to a completely determined problem, shows a good match. The distribution of the unfolded spectra is also studied. Uncertainty propagation in the unfolded spectra due to the errors present in the measurement as well as the response function is also carried out. The method appears to be promising for unfolding the completely determined as well as underdetermined problems. It also has provisions to carry out the uncertainty analysis. (author)

  11. Neutron spectrum and dose-equivalent in shuttle flights during solar maximum

    Energy Technology Data Exchange (ETDEWEB)

    Keith, J E; Badhwar, G D; Lindstrom, D J [National Aeronautics and Space Administration, Houston, TX (United States). Lyndon B. Johnson Space Center

    1992-01-01

    This paper presents unambiguous measurements of the spectrum of neutrons found in spacecraft during spaceflight. The neutron spectrum was measured from thermal energies to about 10 MeV using a completely passive system of metal foils as neutron detectors. These foils were exposed to the neutron flux bare, covered by thermal neutron absorbers (Gd) and inside moderators (Bonner spheres). This set of detectors was flown on three U.S. Space Shuttle flights, STS-28, STS-36 and STS-31, during the solar maximum. We show that the measurements of the radioactivity of these foils lead to a differential neutron energy spectrum in all three flights that can be represented by a power law, J(E){approx equal}E{sup -0.765} neutrons cm{sup -2} day {sup -1} MeV{sup -1}. We also show that the measurements are even better represented by a linear combination of the terrestrial neutron albedo and a spectrum of neutrons locally produced in a aluminium by protons, computed by a previous author. We use both approximations to the neutron spectrum to produce a worst case and most probable case for the neutron spectra and the resulting dose-equivalents, computed using ICRP-51 neutron fluence-dose conversion tables. We compare these to the skin dose-equivalents due to charged particles during the same flights. (author).

  12. Neutron energy spectrum flux profile of Ghana's miniature neutron source reactor core

    International Nuclear Information System (INIS)

    Sogbadji, R.B.M.; Abrefah, R.G.; Ampomah-Amoako, E.; Agbemava, S.E.; Nyarko, B.J.B.

    2011-01-01

    Highlights: → The total neutron flux spectrum of the compact core of Ghana's miniature neutron source reactor was studied. → Using 20,484 energy grids, the thermal, slowing down and fast neutron energy regions were studied. - Abstract: The total neutron flux spectrum of the compact core of Ghana's miniature neutron source reactor was understudied using the Monte Carlo method. To create small energy groups, 20,484 energy grids were used for the three neutron energy regions: thermal, slowing down and fast. The moderator, the inner irradiation channels, the annulus beryllium reflector and the outer irradiation channels were the region monitored. The thermal neutrons recorded their highest flux in the inner irradiation channel with a peak flux of (1.2068 ± 0.0008) x 10 12 n/cm 2 s, followed by the outer irradiation channel with a peak flux of (7.9166 ± 0.0055) x 10 11 n/cm 2 s. The beryllium reflector recorded the lowest flux in the thermal region with a peak flux of (2.3288 ± 0.0004) x 10 11 n/cm 2 s. The peak values of the thermal energy range occurred in the energy range (1.8939-3.7880) x 10 -08 MeV. The inner channel again recorded the highest flux of (1.8745 ± 0.0306) x 10 09 n/cm 2 s at the lower energy end of the slowing down region between 8.2491 x 10 -01 MeV and 8.2680 x 10 -01 MeV, but was over taken by the moderator as the neutron energies increased to 2.0465 MeV. The outer irradiation channel recorded the lowest flux in this region. In the fast region, the core, where the moderator is found, the highest flux was recorded as expected, at a peak flux of (2.9110 ± 0.0198) x 10 08 n/cm 2 s at 6.961 MeV. The inner channel recorded the second highest while the outer channel and annulus beryllium recorded very low flux in this region. The flux values in this region reduce asymptotically to 20 MeV.

  13. 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

  14. 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)

  15. Continuous energy Neutron Transport Monte Carlo Simulator Project: Decomposition of the neutron energy spectrum by target nuclei tagging

    Energy Technology Data Exchange (ETDEWEB)

    Barcellos, Luiz Felipe F.C.; Bodmann, Bardo E.J.; Vilhena, Marco T.M.B., E-mail: luizfelipe.fcb@gmail.com, E-mail: bardo.bodmann@ufrgs.br, E-mail: mtmbvilhena@gmail.com [Universidade Federal do Rio Grande do Sul (UFRGS), Porto Alegre, RS (Brazil). Grupo de Estudos Nucleares; Leite, Sergio Q. Bogado, E-mail: sbogado@ibest.com.br [Comissao Nacional de Energia Nuclear (CNEN), Rio de Janeiro, RJ (Brazil)

    2017-07-01

    In this work a Monte Carlo simulator with continuous energy is used. This simulator distinguishes itself by using the sum of three probability distributions to represent the neutron spectrum. Two distributions have known shape, but have varying population of neutrons in time, and these are the fission neutron spectrum (for high energy neutrons) and the Maxwell-Boltzmann distribution (for thermal neutrons). The third distribution has an a priori unknown and possibly variable shape with time and is determined from parametrizations of Monte Carlo simulation. It is common practice in neutron transport calculations, e.g. multi-group transport, to consider that the neutrons only lose energy with each scattering reaction and then to use a thermal group with a Maxwellian distribution. Such an approximation is valid due to the fact that for fast neutrons up-scattering occurrence is irrelevant, being only appreciable at low energies, i.e. in the thermal energy region, in which it can be regarded as a Maxwell-Boltzmann distribution for thermal equilibrium. In this work the possible neutron-matter interactions are simulated with exception of the up-scattering of neutrons. In order to preserve the thermal spectrum, neutrons are selected stochastically as being part of the thermal population and have an energy attributed to them taken from a Maxwellian distribution. It is then shown how this procedure can emulate the up-scattering effect by the increase in the neutron population kinetic energy. Since the simulator uses tags to identify the reactions it is possible not only to plot the distributions by neutron energy, but also by the type of interaction with matter and with the identification of the target nuclei involved in the process. This work contains some preliminary results obtained from a Monte Carlo simulator for neutron transport that is being developed at Federal University of Rio Grande do Sul. (author)

  16. On the proton spectrum in free neutron β-decay

    International Nuclear Information System (INIS)

    Bunatyan, G.G.

    2000-01-01

    We consider the calculations which are appropriate to acquire with a high precision, of ∼ 1% or better, the general characteristics of weak interactions from the experiments on the free neutron β-decay; the principle emphasis is placed on the phenomena associated with the recoil of protons. The part played by electromagnetic interactions in β-decay is visualized, with special attention drawn to the influence of the γ-radiation on the momentum distribution of the particles in the final state. The effect of electromagnetic interactions on the proton recoil spectrum is studied, in the light of the experiments which are carried out and planned for now. The results of the calculations, which are to be confronted with the experimental data, are presented upright in terms of the effective Lagrangian underlying the inquiry. Owing to electromagnetic interactions, the corrections to the energy distribution of protons prove to amount to the value of a few per cent. Nowadays, this is substantial to obtain with a high accuracy the characteristics of weak interactions by processing the data of the experiments on the proton distribution in the free neutron β-decay

  17. 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)

  18. 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.

  19. Analysis of neutron spectrum effects on primary damage in tritium breeding blankets

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Yong Hee, E-mail: cyh871@snu.ac.kr [School of Energy Systems Engineering, Seoul National University, 599 Gwanak-ro, Gwanak-gu, Seoul 151-744 (Korea, Republic of); Joo, Han Gyu [School of Energy Systems Engineering, Seoul National University, 599 Gwanak-ro, Gwanak-gu, Seoul 151-744 (Korea, Republic of)

    2012-07-15

    The effect of neutron spectrum on primary damages in a structural material of a tritium breeding blanket is investigated with a newly established recoil spectrum estimation system. First, a recoil spectrum generation code is developed to obtain the energy spectrum of primary knock-on atoms (PKAs) for a given neutron spectrum utilizing the latest ENDF/B data. Secondly, a method for approximating the high energy tail of the recoil spectrum is introduced to avoid expensive molecular dynamics calculations for high energy PKAs using the concept of recoil energy of the secondary knock-on atoms originated by the INtegration of CAScades (INCAS) model. Thirdly, the modified spectrum is combined with a set of molecular dynamics calculation results to estimate the primary damage parameters such as the number of surviving point defects. Finally, the neutron spectrum is varied by changing the material of the spectral shifter and the result in primary damage parameters is examined.

  20. Analysis of neutron spectrum effects on primary damage in tritium breeding blankets

    Science.gov (United States)

    Choi, Yong Hee; Joo, Han Gyu

    2012-07-01

    The effect of neutron spectrum on primary damages in a structural material of a tritium breeding blanket is investigated with a newly established recoil spectrum estimation system. First, a recoil spectrum generation code is developed to obtain the energy spectrum of primary knock-on atoms (PKAs) for a given neutron spectrum utilizing the latest ENDF/B data. Secondly, a method for approximating the high energy tail of the recoil spectrum is introduced to avoid expensive molecular dynamics calculations for high energy PKAs using the concept of recoil energy of the secondary knock-on atoms originated by the INtegration of CAScades (INCAS) model. Thirdly, the modified spectrum is combined with a set of molecular dynamics calculation results to estimate the primary damage parameters such as the number of surviving point defects. Finally, the neutron spectrum is varied by changing the material of the spectral shifter and the result in primary damage parameters is examined.

  1. Neutron spectrum determination of d(20)+Be source reaction by the dosimetry foils method

    Science.gov (United States)

    Stefanik, Milan; Bem, Pavel; Majerle, Mitja; Novak, Jan; Simeckova, Eva

    2017-11-01

    The cyclotron-based fast neutron generator with the thick beryllium target operated at the NPI Rez Fast Neutron Facility is primarily designed for the fast neutron production in the p+Be source reaction at 35 MeV. Besides the proton beam, the isochronous cyclotron U-120M at the NPI provides the deuterons in the energy range of 10-20 MeV. The experiments for neutron field investigation from the deuteron bombardment of thick beryllium target at 20 MeV were performed just recently. For the neutron spectrum measurement of the d(20)+Be source reaction, the dosimetry foils activation method was utilized. Neutron spectrum reconstruction from resulting reaction rates was performed using the SAND-II unfolding code and neutron cross-sections from the EAF-2010 nuclear data library. Obtained high-flux white neutron field from the d(20)+Be source is useful for the intensive irradiation experiments and cross-section data validation.

  2. Calculating the energy spectrum of neutrons from tritium target of the NG-150 type generator

    International Nuclear Information System (INIS)

    Bortash, A.I.; Kuznetsov, V.S.

    1987-01-01

    Calculation procedure of neutron spectra yielding from the NG-150 generator target chamber with regard to deutron moderation is suggested. Using the suggested procedure, neutron spectra for different escape angles formed in the tritium target are calculated. The spectrum of neutrons scattered in cooling water is calculated. The mean energy of neutrons escaping at the angle of 0 deg equalling 14.5 MeV is obtained

  3. Broad Energy Range Neutron Spectroscopy using a Liquid Scintillator and a Proportional Counter: Application to a Neutron Spectrum Similar to that from an Improvised Nuclear Device.

    Science.gov (United States)

    Xu, Yanping; Randers-Pehrson, Gerhard; Marino, Stephen A; Garty, Guy; Harken, Andrew; Brenner, David J

    2015-09-11

    A novel neutron irradiation facility at the Radiological Research Accelerator Facility (RARAF) has been developed to mimic the neutron radiation from an Improvised Nuclear Device (IND) at relevant distances (e.g. 1.5 km) from the epicenter. The neutron spectrum of this IND-like neutron irradiator was designed according to estimations of the Hiroshima neutron spectrum at 1.5 km. It is significantly different from a standard reactor fission spectrum, because the spectrum changes as the neutrons are transported through air, and it is dominated by neutron energies from 100 keV up to 9 MeV. To verify such wide energy range neutron spectrum, detailed here is the development of a combined spectroscopy system. Both a liquid scintillator detector and a gas proportional counter were used for the recoil spectra measurements, with the individual response functions estimated from a series of Monte Carlo simulations. These normalized individual response functions were formed into a single response matrix for the unfolding process. Several accelerator-based quasi-monoenergetic neutron source spectra were measured and unfolded to test this spectroscopy system. These reference neutrons were produced from two reactions: T(p,n) 3 He and D(d,n) 3 He, generating neutron energies in the range between 0.2 and 8 MeV. The unfolded quasi-monoenergetic neutron spectra indicated that the detection system can provide good neutron spectroscopy results in this energy range. A broad-energy neutron spectrum from the 9 Be(d,n) reaction using a 5 MeV deuteron beam, measured at 60 degrees to the incident beam was measured and unfolded with the evaluated response matrix. The unfolded broad neutron spectrum is comparable with published time-of-flight results. Finally, the pair of detectors were used to measure the neutron spectrum generated at the RARAF IND-like neutron facility and a comparison is made to the neutron spectrum of Hiroshima.

  4. The measurement of prompt neutron spectrum in spontaneous fission of {sup 244}Cm

    Energy Technology Data Exchange (ETDEWEB)

    Batenkov, O.I.; Boykov, G.S.; Drapchinsky, L.V.; Majorov, M.Ju.; Trenkin, V.A. [V.G. Khlopin Radium Inst., Saint Petersburg (Russian Federation)

    1997-03-01

    Under the Program of Measurements of Prompt Fission Neutron Spectra of Minor Actinides for Transmutation Purposes the integral neutron spectrum in spontaneous fission of {sup 244}Cm has been measured by the time-of-flight method in the energy range of 0.1-15 MeV relative to the standard neutron spectrum in {sup 252}Cf spontaneous fission. Essential attention was paid to revealing of possible systematic errors. It is shown, that the {sup 244}Cm spectrum shape may be well described by using Mannhart evaluation with appropriate parameter of Maxwell temperature T{sub M} = 1.37 MeV. (author)

  5. Comparison of Americium-Beryllium neutron spectrum obtained using activation foil detectors and NE-213 spectrometer

    International Nuclear Information System (INIS)

    Sunny, Sunil; Subbaiah, K.V.; Selvakumaran, T.S.

    1999-01-01

    Neutron spectrum of Americium - Beryllium (α,n) source is measured with two different spectrometers vis-a-vis activation foils (foil detectors) and NE-213 organic scintillator. Activity induced in the foils is measured with 4π-β-γ sodium iodide detector by integrating counts under photo peak and the saturation activity is found by correcting to elapsed time before counting. The data on calculated activity is fed into the unfolding code, SAND-II to obtain neutron spectrum. In the case of organic scintillator, the pulse height spectrum is obtained using MCA and this is processed with unfolding code DUST in order to get neutron spectrum. The Americium - Beryllium (α,n) neutron spectrum thus obtained by two different methods is compared. It is inferred that the NE-213 scintillator spectrum is in excellent agreement with the values beyond 1MeV. Neutron spectrum obtained by activation foils depends on initial guess spectrum and is found to be in reasonable agreement with NE-213 spectrum. (author)

  6. 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)

  7. On the Proton Spectrum in Free Neutron beta-decay

    CERN Document Server

    Bunatian, G G

    2000-01-01

    We consider the calculations which are appropriate to acquire with a high precision, of ~1% or better, the general characteristics of weak interactions from the experiments on the free neutron beta-decay; the principle emphasis is placed on the phenomena associated with the recoil of protons. The part played by electromagnetic interactions in beta-decay is visualized, with special attention drawn to the influence of the gamma-radiation on the momentum distribution of the particles in the final state. The effect of electromagnetic interactions on the proton recoil spectrum is studied, in the light of the experiments which are carried out and planned for now. The results of the calculations, which are to be confronted with the experimental data, are presented upright in terms of the effective Lagrangian underlying the inquiry. Owing to electromagnetic interactions, the corrections to the energy distribution of protons prove to amount to the value of a few per cent. Nowadays, this is substantial to obtain with a...

  8. Neutron spectrum for neutron capture therapy in boron; Espectro de neutrones para terapia por captura de neutrones en boro

    Energy Technology Data Exchange (ETDEWEB)

    Medina C, D.; Soto B, T. G. [Universidad Autonoma de Zacatecas, Unidad Academica de Estudios Nucleares, Programa de Doctorado en Ciencias Basicas, 98068 Zacatecas, Zac. (Mexico); Baltazar R, A. [Universidad Autonoma de Zacatecas, Unidad Academica de Ingenieria Electrica, Programa de Doctorado en Ingenieria y Tecnologia Aplicada, 98068 Zacatecas, Zac. (Mexico); Vega C, H. R., E-mail: dmedina_c@hotmail.com [Universidad Autonoma de Zacatecas, Unidad Academica de Estudios Nucleares, Cipres No. 10, Fracc. La Penuela, 98068 Zacatecas, Zac. (Mexico)

    2016-10-15

    Glioblastoma multiforme is the most common and aggressive of brain tumors and is difficult to treat by surgery, chemotherapy or conventional radiation therapy. One treatment alternative is the Neutron Capture Therapy in Boron, which requires a beam modulated in neutron energy and a drug with {sup 10}B able to be fixed in the tumor. When the patients head is exposed to the neutron beam, they are captured by the {sup 10}B and produce a nucleus of {sup 7}Li and an alpha particle whose energy is deposited in the cancer cells causing it to be destroyed without damaging the normal tissue. One of the problems associated with this therapy is to have an epithermal neutrons flux of the order of 10{sup 9} n/cm{sup 2}-sec, whereby irradiation channels of a nuclear research reactor are used. In this work using Monte Carlo methods, the neutron spectra obtained in the radial irradiation channel of the TRIGA Mark III reactor are calculated when inserting filters whose position and thickness have been modified. From the arrangements studied, we found that the Fe-Cd-Al-Cd polyethylene filter yielded a ratio between thermal and epithermal neutron fluxes of 0.006 that exceeded the recommended value (<0.05), and the dose due to the capture gamma rays is lower than the dose obtained with the other arrangements studied. (Author)

  9. Evaluation of the Neutron Detector Response for Cosmic Ray Energy Spectrum by Monte Carlo Transport Simulation

    International Nuclear Information System (INIS)

    Pazianotto, Mauricio T.; Carlson, Brett V.; Federico, Claudio A.; Gonzalez, Odair L.

    2011-01-01

    Neutrons generated by the interaction of cosmic rays with the atmosphere make an important contribution to the dose accumulated in electronic circuits and aircraft crew members at flight altitude. High-energy neutrons are produced in spallation reactions and intranuclear cascade processes by primary cosmic-ray particle interactions with atoms in the atmosphere. These neutrons can produce secondary neutrons and also undergo a moderation process due to atmosphere interactions, resulting in a wider energy spectrum, ranging from thermal energies (0.025 eV) to energies of several hundreds of MeV. The Long-Counter (LC) detector is a widely used neutron detector designed to measure the directional flux of neutrons with about constant response over a wide energy range (thermal to 20 MeV). ). Its calibration process and the determination of its energy response for the wide-energy of cosmic ray induced neutron spectrum is a very difficult process due to the lack of installations with these capabilities. The goal of this study is to assess the behavior of the response of a Long Counter using the Monte Carlo (MC) computational code MCNPX (Monte Carlo N-Particle eXtended). The dependence of the Long Counter response on the angle of incidence, as well as on the neutron energy, will be carefully investigated, compared with the experimental data previously obtained with 241 Am-Be and 252 Cf neutron sources and extended to the neutron spectrum produced by cosmic rays. (Author)

  10. Corrections on energy spectrum and scattering for fast neutron radiography at NECTAR facility

    International Nuclear Information System (INIS)

    Liu Shuquan; Thomas, Boucherl; Li Hang; Zou Yubin; Lu Yuanrong; Guo Zhiyu

    2013-01-01

    Distortions caused by the neutron spectrum and scattered neutrons are major problems in fast neutron radiography and should be considered for improving the image quality. This paper puts emphasis on the removal of these image distortions and deviations for fast neutron radiography performed at the NECTAR facility of the research reactor FRM-Ⅱ in Technische Universitaet Mounchen (TUM), Germany. The NECTAR energy spectrum is analyzed and established to modify the influence caused by the neutron spectrum, and the Point Scattered Function (PScF) simulated by the Monte-Carlo program MCNPX is used to evaluate scattering effects from the object and improve image quality. Good analysis results prove the sound effects of the above two corrections. (authors)

  11. Corrections on energy spectrum and scatterings for fast neutron radiography at NECTAR facility

    Science.gov (United States)

    Liu, Shu-Quan; Bücherl, Thomas; Li, Hang; Zou, Yu-Bin; Lu, Yuan-Rong; Guo, Zhi-Yu

    2013-11-01

    Distortions caused by the neutron spectrum and scattered neutrons are major problems in fast neutron radiography and should be considered for improving the image quality. This paper puts emphasis on the removal of these image distortions and deviations for fast neutron radiography performed at the NECTAR facility of the research reactor FRM- II in Technische Universität München (TUM), Germany. The NECTAR energy spectrum is analyzed and established to modify the influence caused by the neutron spectrum, and the Point Scattered Function (PScF) simulated by the Monte-Carlo program MCNPX is used to evaluate scattering effects from the object and improve image quality. Good analysis results prove the sound effects of the above two corrections.

  12. Measurement of photoneutron spectrum at Pohang Neutron Facility

    Energy Technology Data Exchange (ETDEWEB)

    Kim, G.N.; Kovalchuk, V.; Lee, Y.S.; Skoy, V.; Cho, M.H.; Ko, I.S.; Namkung, W. [POSTECH, Pohang Accelerator Laboratory, Pohang, Kyungbuk (Korea)

    2001-03-01

    Pohang Neutron Facility, which is the pulsed neutron facility based on the 100-MeV electron linear accelerator, was constructed for nuclear data production in Korea. The Pohang Neutron Facility consists of an electron linear accelerator, a water-cooled Ta target with a water moderator and a time-of-flight path with an 11 m length. The neutron energy spectra are measured for different water levels inside the moderator and compared with the MCNP calculation. The optimum size of the water moderator is determined on the base of this result. The time dependent spectra of neutrons in the water moderator are investigated with the MCNP calculation. (author)

  13. BONDI-97 A novel neutron energy spectrum unfolding tool using a genetic algorithm

    CERN Document Server

    Mukherjee, B

    1999-01-01

    The neutron spectrum unfolding procedure using the count rate data obtained from a set of Bonner sphere neutron detectors requires the solution of the Fredholm integral equation of the first kind by using complex mathematical methods. This paper reports a new approach for the unfolding of neutron spectra using the Genetic Algorithm tool BONDI-97 (BOnner sphere Neutron DIfferentiation). The BONDI-97 was used as the input for Genetic Algorithm engine EVOLVER to search for a globally optimised solution vector from a population of randomly generated solutions. This solution vector corresponds to the unfolded neutron energy spectrum. The Genetic Algorithm engine emulates the Darwinian 'Survival of the Fittest' strategy, the key ingredient of the 'Theory of Evolution'. The spectra of sup 2 sup 4 sup 1 Am/Be (alpha,n) and sup 2 sup 3 sup 9 Pu/Be (alpha,n) neutron sources were unfolded using the BONDI-97 tool. (author)

  14. Earth formation porosity log using measurement of neutron energy spectrum

    International Nuclear Information System (INIS)

    1981-01-01

    Methods and apparatus are described for measuring the porosity of subsurface earth formations in the vicinity of a well borehole by means of neutron well logging techniques. All the commercial techniques for measuring porosity currently available are not as accurate as desirable due to variations in the borehole wall diameter, in the borehole fluids (e.g. with chlorine content) in the casings of the borehole etc. This invention seeks to improve accuracy by using a measurement of the epithermal neutron population at one detector and the fast neutron population at a second detector, spaced approximately the same distance from a neutron source. The latter can be detected either by a fast neutron detector or indirectly by an inelastic gamma ray detector. Background correction can be made, and special detectors used, to discriminate against the detection of thermal neutrons or their resultant capture gamma rays. These fluctuations affect the measurement of thermal neutron populations. (U.K.)

  15. Refinements in the Los Alamos model of the prompt fission neutron spectrum

    Energy Technology Data Exchange (ETDEWEB)

    Madland, D.G., E-mail: dgm@lanl.gov; Kahler, A.C.

    2017-01-15

    This paper presents a number of refinements to the original Los Alamos model of the prompt fission neutron spectrum and average prompt neutron multiplicity as derived in 1982. The four refinements are due to new measurements of the spectrum and related fission observables many of which were not available in 1982. They are also due to a number of detailed studies and comparisons of the model with previous and present experimental results including not only the differential spectrum, but also integral cross sections measured in the field of the differential spectrum. The four refinements are (a) separate neutron contributions in binary fission, (b) departure from statistical equilibrium at scission, (c) fission-fragment nuclear level-density models, and (d) center-of-mass anisotropy. With these refinements, for the first time, good agreement has been obtained for both differential and integral measurements using the same Los Alamos model spectrum.

  16. Neutron energy spectrum influence on irradiation hardening and microstructural development of tungsten

    Energy Technology Data Exchange (ETDEWEB)

    Fukuda, Makoto, E-mail: makoto.fukuda@qse.tohoku.ac.jp [Tohoku University, Sendai, 980-8579 (Japan); Kiran Kumar, N.A.P.; Koyanagi, Takaaki; Garrison, Lauren M. [Oak Ridge National Laboratory, Oak Ridge, TN, 37831 (United States); Snead, Lance L. [Massachusetts Institute of Technology, Cambridge, MA, 02139 (United States); Katoh, Yutai [Oak Ridge National Laboratory, Oak Ridge, TN, 37831 (United States); Hasegawa, Akira [Tohoku University, Sendai, 980-8579 (Japan)

    2016-10-15

    Neutron irradiation to single crystal pure tungsten was performed in the mixed spectrum High Flux Isotope Reactor (HFIR). To investigate the influences of neutron energy spectrum, the microstructure and irradiation hardening were compared with previous data obtained from the irradiation campaigns in the mixed spectrum Japan Material Testing Reactor (JMTR) and the sodium-cooled fast reactor Joyo. The irradiation temperatures were in the range of ∼90–∼800 °C and fast neutron fluences were 0.02–9.00 × 10{sup 25} n/m{sup 2} (E > 0.1 MeV). Post irradiation evaluation included Vickers hardness measurements and transmission electron microscopy. The hardness and microstructure changes exhibited a clear dependence on the neutron energy spectrum. The hardness appeared to increase with increasing thermal neutron flux when fast fluence exceeds 1 × 10{sup 25} n/m{sup 2} (E > 0.1 MeV). Irradiation induced precipitates considered to be χ- and σ-phases were observed in samples irradiated to >1 × 10{sup 25} n/m{sup 2} (E > 0.1 MeV), which were pronounced at high dose and due to the very high thermal neutron flux of HFIR. Although the irradiation hardening mainly caused by defects clusters in a low dose regime, the transmutation-induced precipitation appeared to impose additional significant hardening of the tungsten. - Highlights: • The microstructure and irradiation hardening of single crystal pure W irradiated in HFIR was investigated. • The neutron energy spectrum influence was evaluated by comparing the HFIR results with previous work in Joyo and JMTR. • In the dose range up to ∼1 dpa, the neutron energy spectrum influence of irradiation hardening was not clear. • In the dose range above 1 dpa, the neutron energy influence on irradiation hardening and microstructural development was clearly observed. • The irradiation induced precipitates caused significant irradiation hardening of pure W irradiated in HFIR.

  17. Importance of the neutron spectrum for determination of radiation damage

    International Nuclear Information System (INIS)

    Hehn, G.; Stiller, P.; Mattes, M.

    1977-01-01

    Since the radiation effects of neutrons depend strongly on the neutron energy, the correlation between the induced damage and the fluence of the fast neutrons shows appreciable disadvantages. The measured values of changes in material properties resulted in large differences for the same fast neutron fluence, being partly due to different neutron spectra. The uncertainties in damage data led to strong overdesign of important structural components. Different neutron environment at surveillance sample position may give an underestimation of the embrittlement in the reactor pressure vessel, which has to be avoided. The application of damage functions combined with accurately calculated neutron spectra, promise to be a reasonable solution. The damage function has the advantage of a phenomenological quantity that all spectral effects are included. But the correlation quantity has to be determined of high experimental costs. Therefore approximations of its energy distributions are very important. For the keV energy region the kerma function is reasonably good. For the MeV energy region a higher effort is needed to calculate the displacement cross section. The same holds for the low energy part. In all three parts the formation of stable material property levels may vary, so that the final correlation can be determined only by measurements of material properties in different neutron spectra. In material samples the spectra distribution of the displacement production rate was determined at different local positions outside the reactor core of a PWR and a fast breeder showing the most important energy regions of both reactors. (orig.) [de

  18. Translation of selected reports on neutron spectrum unfolding

    International Nuclear Information System (INIS)

    Berzonis, M.; Bondars, Kh.Ya.; Taimina, D.

    1982-05-01

    The paper provides the information needed by users of the SAIPS information system on the neutron cross-section libraries accessible and on the principles upon which they are based. Neutron cross-section integrals in fission and fusion spectra are given. (author)

  19. 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)

  20. Study on neutron spectrum for effective transmutation of minor actinides in thermal reactors

    International Nuclear Information System (INIS)

    Takeda, Toshikazu; Yokoyama, Kenji

    1997-01-01

    The transmutation of minor actinides (MAs) has been investigated in thermal reactor cells using mixed oxide fuel with MAs. The effect of neutron spectra on transmutation is studied by changing the neutron spectra. Five transmutation rates are compared: direct fission incineration rate, capture transmutation rate, consumption rate, overall fission incineration rate and inventory difference transmutation rate. The relations between these transmutation rates and their dependence on the neutron spectrum were investigated. To effectively incinerate MAs it is necessary to maximize the overall fission incineration rate and the inventory difference transmutation rate. These transmutation rates become maximum when the fraction of neutrons below 1 eV is about 8% for the case where the MA addition is 1-3%. When the MA addition is over 5%, the transmutation rates become maximum for very hard neutron spectrum. (Author)

  1. Neutron flux uncertainty and covariances for spectrum adjustment and estimation of WWER-1000 pressure vessel fluences

    International Nuclear Information System (INIS)

    Boehmer, Bertram

    2000-01-01

    Results of estimation of the covariance matrix of the neutron spectrum in the WWER-1000 reactor cavity and pressure vessel positions are presented. Two-dimensional calculations with the discrete ordinates transport code DORT in r-theta and r-z-geometry used to determine the neutron group spectrum covariances including gross-correlations between interesting positions. The new Russian ABBN-93 data set and CONSYST code used to supply all transport calculations with group neutron data. All possible sources of uncertainties namely caused by the neutron gross sections, fission sources, geometrical dimensions and material densities considered, whereas the uncertainty of the calculation method was considered negligible in view of the available precision of Monte Carlo simulation used for more precise evaluation of the neutron fluence. (Authors)

  2. 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.

  3. The 235U prompt fission neutron spectrum measured by the Chi-Nu project at LANSCE

    Directory of Open Access Journals (Sweden)

    Gomez J.A.

    2017-01-01

    Full Text Available The Chi-Nu experiment aims to accurately measure the prompt fission neutron spectrum for the major actinides. At the Los Alamos Neutron Science Center (LANSCE, fission can be induced with neutrons ranging from 0.7 MeV and above. Using a two arm time-of-flight (TOF technique, the fission neutrons are measured in one of two arrays: a 22-6Li glass array for lower energies, or a 54-liquid scintillator array for outgoing energies of 0.5 MeV and greater. Presented here are the collaboration's preliminary efforts at measuring the 235U PFNS.

  4. The Real-time Frequency Spectrum Analysis of Neutron Pulse Signal Series

    International Nuclear Information System (INIS)

    Tang Yuelin; Ren Yong; Wei Biao; Feng Peng; Mi Deling; Pan Yingjun; Li Jiansheng; Ye Cenming

    2009-01-01

    The frequency spectrum analysis of neutron pulse signal is a very important method in nuclear stochastic signal processing Focused on the special '0' and '1' of neutron pulse signal series, this paper proposes new rotation-table and realizes a real-time frequency spectrum algorithm under 1G Hz sample rate based on PC with add, address and SSE. The numerical experimental results show that under the count rate of 3X10 6 s -1 , this algorithm is superior to FFTW in time-consumption and can meet the real-time requirement of frequency spectrum analysis. (authors)

  5. Spectrum of neutrons leaking from an iron sphere with a central 14 MeV neutron source

    International Nuclear Information System (INIS)

    Borisov, A.A.; Zagryadskij, V.A.; Chuvilin, D.Yu.; Kralik, M.; Pulpan, J.; Tichy, M.

    1991-01-01

    Following a review of the present state of nuclear data requisite for the calculation of the transport of 14 MeV neutrons through iron of natural isotopic composition, the results are given of the calculation of the energy spectrum of such neutrons after their passage through an iron sphere 240 mm o.d. and 90 mm i.d., the neutron source being accommodated in the centre of the sphere. The calculations were made using the one-dimensional code BLANK working with the nuclear data libraries ENDL-75, ENDL-83, ENDL/B-IV, JENDL-2 and BROND, and using the three-dimensional code BRAND with the library ENDL-78. The calculated spectra were compared with the experimental spectrum measured at a distance of 3 m from the sphere by means of an NE-213 scintillator, which records reflected protons. The reflected proton spectrum was processed by the matrix method (program FORIST), and the result was normalized to one neutron emitted by the source, as were the calculated spectra. The comparison demonstrates that the experiment is best fitted by the spectrum calculated by using the library JENDL-2, where the integrals of the observed and calculated spectra over the 1-15 MeV range differ as little as approximately 10%. (author). 3 figs., 5 tabs., 16 refs

  6. Cadmium depletion impacts on hardening neutron spectrum for advanced fuel testing in ATR

    International Nuclear Information System (INIS)

    Chang, Gray S.

    2011-01-01

    For transmuting long-lived isotopes contained in spent nuclear fuel into shorter-lived fission products effectively is in a fast neutron spectrum reactor. In the absence of a fast spectrum test reactor in the United States of America (USA), initial irradiation testing of candidate fuels can be performed in a thermal test reactor that has been modified to produce a test region with a hardened neutron spectrum. A test region is achieved with a Cadmium (Cd) filter which can harden the neutron spectrum to a spectrum similar (although still somewhat softer) to that of the liquid metal fast breeder reactor (LMFBR). A fuel test loop with a Cd-filter has been installed within the East Flux Trap (EFT) of the Advanced Test Reactor (ATR) at the Idaho National Laboratory (INL). A detailed comparison analyses between the cadmium (Cd) filter hardened neutron spectrum in the ATR and the LMFBR fast neutron spectrum have been performed using MCWO. MCWO is a set of scripting tools that are used to couple the Monte Carlo transport code MCNP with the isotope depletion and buildup code ORIGEN-2.2. The MCWO-calculated results indicate that the Cd-filter can effectively flatten the Rim-Effect and reduce the linear heat rate (LHGR) to meet the advanced fuel testing project requirements at the beginning of irradiation (BOI). However, the filtering characteristics of Cd as a strong absorber quickly depletes over time, and the Cd-filter must be replaced for every two typical operating cycles within the EFT of the ATR. The designed Cd-filter can effectively depress the LHGR in experimental fuels and harden the neutron spectrum enough to adequately flatten the Rim-Effect in the test region. (author)

  7. A comparison in the reconstruction of neutron spectrums using classical iterative techniques

    International Nuclear Information System (INIS)

    Ortiz R, J. M.; Martinez B, M. R.; Vega C, H. R.; Gallego, E.

    2009-10-01

    One of the key drawbacks to the use of BUNKI code is that the process begins the reconstruction of the spectrum based on a priori knowledge as close as possible to the solution that is sought. The user has to specify the initial spectrum or do it through a subroutine called MAXIET to calculate a Maxwellian and a 1/E spectrum as initial spectrum. Because the application of iterative procedures by to resolve the reconstruction of neutron spectrum needs an initial spectrum, it is necessary to have new proposals for the election of the same. Based on the experience gained with a widely used method of reconstruction, called BUNKI, has developed a new computational tools for neutron spectrometry and dosimetry, which was first introduced, which operates by means of an iterative algorithm for the reconstruction of neutron spectra. The main feature of this tool is that unlike the existing iterative codes, the choice of the initial spectrum is performed automatically by the program, through a neutron spectra catalog. To develop the code, the algorithm was selected as the routine iterative SPUNIT be used in computing tool and response matrix UTA4 for 31 energy groups. (author)

  8. Evaluating the 239Pu Prompt Fission Neutron Spectrum Induced by Thermal to 30 MeV Neutrons

    Directory of Open Access Journals (Sweden)

    Neudecker D.

    2016-01-01

    Full Text Available We present a new evaluation of the 239Pu prompt fission neutron spectrum (PFNS induced by thermal to 30 MeV neutrons. Compared to the ENDF/B-VII.1 evaluation, this one includes recently published experimental data as well as an improved and extended model description to predict PFNS. For instance, the pre-equilibrium neutron emission component to the PFNS is considered and the incident energy dependence of model parameters is parametrized more realistically. Experimental and model parameter uncertainties and covariances are estimated in detail. Also, evaluated covariances are provided between all PFNS at different incident neutron energies. Selected evaluation results and first benchmark calculations using this evaluation are briefly discussed.

  9. Spectrum of ferromagnetic transition metal magnetic excitations and neutron scattering

    International Nuclear Information System (INIS)

    Kuzemskij, A.L.

    1979-01-01

    Quantum statistical models of ferromagnetic transition metals as well as methods of their solutions are reviewed. The correspondence of results on solving these models and the data on scattering thermal neutrons in ferromagnetic is discussed

  10. Sequential measurements of spectrum and dose for cosmic-ray neutrons on the ground

    International Nuclear Information System (INIS)

    Hirabayashi, N.; Nunomiya, T.; Suzuki, H.; Nakamura, T.

    2002-01-01

    The earth is continually bathed in high-energy particles that come from outside the solar system, known as galactic cosmic rays. When these particles penetrate the magnetic fields of the solar system and the Earth and reach the Earth's atmosphere, they collide with atomic nuclei in air and secondary cosmic rays of every kind. On the other hand, levels of accumulation of the semiconductor increase recently, and the soft error that the cosmic-ray neutrons cause has been regarded as questionable. There have been long-term measurements of cosmic-ray neutron fluence at several places in the world, but no systematic study on cosmic-ray neutron spectrum measurements. This study aimed to measure the cosmic-ray neutron spectrum and dose on the ground during the solar maximum period of 2000 to 2002. Measurements have been continuing in a cabin of Tohoku University Kawauchi campus, by using five multi-moderator spectrometers (Bonner sphere), 12.7 cm diam by 12.7 cm long NE213 scintillator, and rem counter. The Bonner sphere uses a 5.08 cm diam spherical 3 He gas proportional counter and the rem counter uses a 12.7 cm diam 3 He gas counter. The neutron spectra were obtained by unfolding from the count rates measured with the Bonner sphere using the SAND code and the pulse height spectra measured with the NE213 scintillator using the FORIST code . The cosmic- ray neutron spectrum and ambient dose rates have been measured sequentially from April 2001. Furthermore, the correlation between ambient dose rate and the atmospheric pressure was investigated with a barometer. We are also very much interested in the variation of neutron spectrum following big solar flares. From the sequential measurements, we found that the cosmic-ray neutron spectrum has two peaks at around 1 MeV and at around 100 MeV, and the higher energy peak increases with a big solar flare

  11. Characterization of a fast to thermal neutron spectrum converter on PROSPERO reactor

    Energy Technology Data Exchange (ETDEWEB)

    Jacquet, X.; Authier, N.; Casoli, P.; Combacon, S. [CEA, Valduc Center, 21120 Is sur Tille (France); Calzavarra, Y. [ILL, Institut Laue Langevin, 38000 Grenoble (France)

    2009-07-01

    The PROSPERO reactor is located at CEA Valduc Center in France. The reactor is composed of an internal core made of High Enriched Uranium metal alloy surrounded by a reflector of depleted uranium. The reactor is used as a fast neutron spectrum source and is operated in delayed critical state with a continuous and steady power for several hours, which can vary from 3 mW to 3 kW, which is the nominal power. The flux at nominal power varies from 5.10{sup +10} n.cm{sup -2}/s at the reflector surface to 10{sup +7} n.cm{sup -2}/s at 5 meters from reactor axis. It has been decided to build a neutron energy converter allowing the production of a neutron thermal spectrum. As the core produces fast neutrons spectrum, we built a hollow cubic box of 50 cm x 50 cm x 50 cm with 10-cm-thick polyethylene bricks and placed one meter away from central reactor axis to moderate as much as possible neutrons to lower energies (E<0.6 eV). Analysis of the moderated flux inside the converter was performed using different activation foils such as indium or gold. We have developed a model of the experiment in the Monte Carlo neutron transport code TRIPOLI-4. A non-analogous transport calculation scheme was necessary to reproduce properly the experimental activities. The results of the calculated activations are within 4% of the experimental measurements given with 10% uncertainty (2 sigma). We show that the converter realizes thermalization of 80 % of the PROSPERO reactor fast neutrons below the cadmium threshold of 0.6 eV. Epithermal neutrons represent 15% of the spectrum and only 5% are in the fast neutron range above 1 MeV. The total flux at the center of the converter is 1.4 10{sup +9} n.cm{sup -2}/s at 3000 W

  12. A test-type hyper-thermal neutron generator for neutron capture therapy - estimation of neutron energy spectrum by simulation calculations and TOF experiments

    International Nuclear Information System (INIS)

    Sakurai, Yoshinori; Kobayashi, Tooru; Kobayashi, Katsuhei

    1999-01-01

    In order to clarify the irradiation characteristics of hyper-thermal neutrons and the feasibility of a hyper-thermal neutron irradiation field for neutron capture therapy, a 'test-type' hyper-thermal neutron generator was designed and made. Graphite of 6 cm thickness and 21 cm diameter was selected as the high temperature scatterer. The scatterer is heated up to 1200 deg. C maximum using molybdenum heaters. The radiation heat is shielded by reflectors of molybdenum and stainless steel. The temperature is measured using three R-type thermo-couples and controlled by a program controller. The total thickness of the generator is designed to be as thin as possible, 20 cm in maximum, in the standing point of the neutron beam intensity. The thermal stability, controllability and safety of the generator at high temperature employment were confirmed by the heating tests. As one of the experiments for the characteristics estimation, the neutron energy spectrum dependent on the scatterer temperature was measured by the TOF (time of flight) method using the LINAC neutron generator. The estimations by simulation calculations were also performed. From the experiment and calculation results, it was confirmed that the neutron temperature shifted higher as the scatterer temperature was higher. The prospect of the feasibility of the 'hyper-thermal neutron irradiation field for NCT' was opened from the estimation results of the generator characteristics by the simulation calculations and experiments

  13. Study of U235 neutron fission spectrum by the knowledge of cross sections average over that spectrum

    International Nuclear Information System (INIS)

    Suarez, P.M.

    1997-01-01

    A literature search of cross sections averaged over the fission neutron spectrum confirms inconsistencies between calculated and experimental values for high threshold reactions. Since, in this case, calculated averaged cross sections are systematically lower than measured values, it is concluded that the representations used to carry out these calculations underestimate the number of neutrons in the high energy region of the spectrum. A careful measurement of the averaged cross section for the 45 Sc(n,2n) 44g Sc and 45 Sc(n,2n) 44m Sc high threshold reactions had been performed in the RA-6 Neutron Activation Analysis Laboratory after carefully checking that the neutron flux at the core position where the samples were being irradiated was indeed an undisturbed fission spectrum. The experimental values are greater than those calculated with either, Watt type representations or the one based on the Madland and Nix model for the prompt fission spectrum. In many areas of nuclear engineering, like validation of nuclear data, reactor calculations, applied nuclear physics, shielding design, etc., it is of great practical importance to have a representation for the neutron flux that can be expressed in a closed analytical form and that agrees with experimental results, specially for the most widely fissile nuclide, 235 U. The results of the calculations mentioned above lead us to propose an analytical form for the 235 U fission neutron spectrum that better agrees with experimental results in the whole energy spectrum. We propose two different forms; both are a modification of the Watt-type form that has been adopted within the ENDF/B-V files. One of the new analytical representations is defined in two regions: below 9.5 MeV it is exactly the same formula as that used within the ENDF/B-V files, above this energy the parameters of this formula are changed. The other proposed analytical representation is expressed by a single formula in the whole energy range. These two new

  14. Experimental Validation of Ex-Vessel Neutron Spectrum by Means of Dosimeter Materials Activation Method

    Directory of Open Access Journals (Sweden)

    S.A. Santa

    2017-06-01

    Full Text Available Neutron spectrum information in reactor core and around of ex-vessel reactor needs to be known with a certain degree of accuracy to support the development of fuels, materials, and other components. The most common method to determine neutron spectra is by utilizing the radioactivation of dosimeter materials. This report presents the evaluation of neutron flux incident on M3dosimeter sets which were irradiated outside the reactor vessel,as well as the validation of  neutron spectrum calculation. Al capsules containing both dosimeter set covered withCd and dosimeter set without Cd cover have been irradiated during the 35th operational cycle in the M3 ex-vessel irradiation hole position207 cmfrom core centerline at the space between the reactor vessel and the safety vessel. The capsules were positioned at Z=0.0 cm of core midplane. Each dosimeter set consists of Co-Al, Sc, Fe, Np, Nb, Ni, B, and Ta. The gamma-ray spectra of irradiated dosimeter materials were measured by 63 cc HPGe solid-state detector and photo-peak spectra were analyzed using BOB75 code. The reaction rates of each dosimeter materials and its uncertainty were analyzed based on 59Co (n,g 60Co, 237Np (n,f 95Zr-103Ru,  45Sc (n,g 46Sc, 58Fe (n,g 59Fe, 181Ta (n,g 182Ta, and 58Ni (n,p58Co reactions. The measured Cd ratios indicate that neutron spectrum at the irradiated dosimeter sets was dominated by low energy neutron. The experimental result shows that the calculated neutron spectra by DORT code at the ex-vessel positions need correction, especially in the fast neutron energy region, so as to obtain reasonable unfolding result consistent with the reaction rate measurement without any exception. Using biased DORT initial spectrum, the neutron spectrum and its integral quantity were unfolded by NEUPAC code. The result shows that total neutron flux, flux above 1.0 MeV, flux above 0.1 MeV, and the displacement rate of the dosimeter set not covered with Cd were 1.75× 1012 n cm2 s-1, 1

  15. The 235U Prompt Fission Neutron Spectrum in the BR1 Reactor at SCK•CEN

    Science.gov (United States)

    Wagemans, Jan; Malambu, Edouard; Borms, Luc; Fiorito, Luca

    2016-02-01

    The BR1 research reactor at SCK•CEN has a spherical cavity in the graphite above the reactor core. In this cavity an accurately characterised Maxwellian thermal neutron field is present. Different converters can be loaded in the cavity in order to obtain other types of neutron (and gamma) irradiation fields. Inside the so-called MARK III converter a fast 235U(n,f) prompt fission neutron field can be obtained. With the support of MCNP calculations, irradiations in MARK III can be directly related to the pure 235U(n,f) prompt fission neutron spectrum. For this purpose MARK III spectrum averaged cross sections for the most relevant fluence dosimetry reactions have been determined. A calibration factor for absolute measurements has been determined applying activation dosimetry following ISO/IEC 17025 standards.

  16. The 235U Prompt Fission Neutron Spectrum in the BR1 Reactor at SCK•CEN

    Directory of Open Access Journals (Sweden)

    Wagemans Jan

    2016-01-01

    Full Text Available The BR1 research reactor at SCK•CEN has a spherical cavity in the graphite above the reactor core. In this cavity an accurately characterised Maxwellian thermal neutron field is present. Different converters can be loaded in the cavity in order to obtain other types of neutron (and gamma irradiation fields. Inside the so-called MARK III converter a fast 235U(n,f prompt fission neutron field can be obtained. With the support of MCNP calculations, irradiations in MARK III can be directly related to the pure 235U(n,f prompt fission neutron spectrum. For this purpose MARK III spectrum averaged cross sections for the most relevant fluence dosimetry reactions have been determined. A calibration factor for absolute measurements has been determined applying activation dosimetry following ISO/IEC 17025 standards.

  17. Nuclear data needs for neutron spectrum tailoring at International Fusion Materials Irradiation Facility (IFMIF)

    International Nuclear Information System (INIS)

    Sugimoto, Masayoshi

    2001-01-01

    International Fusion Materials Irradiation Facility (IFMIF) is a proposal of D-Li intense neutron source to cover all aspects of the fusion materials development in the framework of IEA collaboration. The new activity has been started to qualifying the important technical issues called Key Element technology Phase since 2000. Although the neutron spectrum can be adjusted by changing the incident beam energy, it is favorable to be carried out many irradiation tasks at the same time under the unique beam condition. For designing the tailored neutron spectrum, neutron nuclear data for the moderator-reflector materials up to 50 MeV are required. The data for estimating the induced radioactivity is also required to keep the radiation level low enough at maintenance time. The candidate materials and the required accuracy of nuclear data are summarized. (author)

  18. Nuclear data needs for neutron spectrum tailoring at International Fusion Materials Irradiation Facility (IFMIF)

    Energy Technology Data Exchange (ETDEWEB)

    Sugimoto, Masayoshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2001-03-01

    International Fusion Materials Irradiation Facility (IFMIF) is a proposal of D-Li intense neutron source to cover all aspects of the fusion materials development in the framework of IEA collaboration. The new activity has been started to qualifying the important technical issues called Key Element technology Phase since 2000. Although the neutron spectrum can be adjusted by changing the incident beam energy, it is favorable to be carried out many irradiation tasks at the same time under the unique beam condition. For designing the tailored neutron spectrum, neutron nuclear data for the moderator-reflector materials up to 50 MeV are required. The data for estimating the induced radioactivity is also required to keep the radiation level low enough at maintenance time. The candidate materials and the required accuracy of nuclear data are summarized. (author)

  19. Neutron fluence rate and energy spectrum in SPRR-300 reactor thermal column

    International Nuclear Information System (INIS)

    Dou Haifeng; Dai Junlong

    2006-01-01

    In order to modify the simple one-dimension model, the neutron fluence rate distribution calculated with ANISN code ws checked with that calculated with MCNP code. To modify the error caused by ignoring the neutron landscape orientation leaking, the reflector that can't be modeled in a simple one-dimension model was dealt by extending landscape orientation scale. On this condition the neutron fluence rate distribution and the energy spectrum in the thermal column of SPRR-300 reactor were calculated with one-dimensional code ANISN, and the results of Cd ratio are well accorded with the experimental results. The deviation between them is less than 5% and it isn't above 10% in one or two special positions. It indicates that neutron fluence rate distribution and energy spectrum in the thermal column can be well calculated with one-dimensional code ANISN. (authors)

  20. Using activation method to measure neutron spectrum in an irradiation chamber of a research reactor

    International Nuclear Information System (INIS)

    Zhou Xuemei; Liu Guimin; Wang Xiaohe; Li Da; Meng Lingjie

    2014-01-01

    Neutron spectrum should be measured before test samples are irradiated. Neutron spectrum in an irradiation chamber of a research reactor was measured by using activation method when the reactor is in normal operation under 2 MW. Sixteen kinds of non-fission foils (19 reaction channels) were selected, of which 10 were sensitive to thermal and intermediate energy regions, while the others were of different threshold energy and sensitive to fast energy regions. By measuring the foil radioactivity, the neutron spectrum was unfolded with the iterative methods SAND-II and MSIT. Finally, shielding corrections of group cross-section and main factors affecting the calculation accuracy were studied and the uncertainty of solution was analyzed using the Monte Carlo method in the process of SAND-II. (authors)

  1. Teratogenic and embryolethal effects in mice of fission-spectrum neutrons and γ-rays

    International Nuclear Information System (INIS)

    Cairnie, A.B.; Grahn, D.; Rayburn, H.B.; Williamson, F.S.; Brown, R.J.

    1974-01-01

    Fission-spectrum neutrons from the Janus reactor at Argonne National Laboratory were compared with γ-rays in terms of their relative biological effectiveness (RBE) for embryolethal and teratogenic effects in mice. No evidence was found of any processes that were abnormally sensitive to neutrons. The RBE for killing embryos and producing abnormal embryos or specific abnormalities was between 2 and 3. This is close to the values found in other systems for processes involving cell killing. (U.S.)

  2. Fast neutron spectrum in the exposure room of the TRIGA Mark II reactor in Ljubljana

    International Nuclear Information System (INIS)

    Kristof, E.S.

    2003-01-01

    In this paper a description of the high energy neutrons at a usual position in the dry cell of our reactor is given. Neutrons emerging from the graphite reflector enter the exposure room through the horizontal shaft. At the irradiation position samples of detection materials were irradiated. After irradiation γ-ray spectra were measured and from the saturation activities the spectrum was calculated. (author)

  3. A neutron spectrum unfolding computer code based on artificial neural networks

    International Nuclear Information System (INIS)

    Ortiz-Rodríguez, J.M.; Reyes Alfaro, A.; Reyes Haro, A.; Cervantes Viramontes, J.M.; Vega-Carrillo, H.R.

    2014-01-01

    The Bonner Spheres Spectrometer consists of a thermal neutron sensor placed at the center of a number of moderating polyethylene spheres of different diameters. From the measured readings, information can be derived about the spectrum of the neutron field where measurements were made. Disadvantages of the Bonner system are the weight associated with each sphere and the need to sequentially irradiate the spheres, requiring long exposure periods. Provided a well-established response matrix and adequate irradiation conditions, the most delicate part of neutron spectrometry, is the unfolding process. The derivation of the spectral information is not simple because the unknown is not given directly as a result of the measurements. The drawbacks associated with traditional unfolding procedures have motivated the need of complementary approaches. Novel methods based on Artificial Intelligence, mainly Artificial Neural Networks, have been widely investigated. In this work, a neutron spectrum unfolding code based on neural nets technology is presented. This code is called Neutron Spectrometry and Dosimetry with Artificial Neural networks unfolding code that was designed in a graphical interface. The core of the code is an embedded neural network architecture previously optimized using the robust design of artificial neural networks methodology. The main features of the code are: easy to use, friendly and intuitive to the user. This code was designed for a Bonner Sphere System based on a 6 LiI(Eu) neutron detector and a response matrix expressed in 60 energy bins taken from an International Atomic Energy Agency compilation. The main feature of the code is that as entrance data, for unfolding the neutron spectrum, only seven rate counts measured with seven Bonner spheres are required; simultaneously the code calculates 15 dosimetric quantities as well as the total flux for radiation protection purposes. This code generates a full report with all information of the unfolding

  4. A two-medium thermal neutron spectrum programme

    International Nuclear Information System (INIS)

    Bindon, D.C.

    1960-07-01

    A computer programme is described for computing the thermal neutron spectra and effective cross-sections in a reactor system of two media by the method of H. Takahashi. The programme has been prepared and tested for use with the Ferranti Mercury computer. (author)

  5. Experimental measurement of neutron spectrum in the reflector of a light water reactor

    International Nuclear Information System (INIS)

    Brethe, P.

    1963-09-01

    1. Thermal neutrons: The temperature of the thermal neutron spectrum was calculated using Au-Lu foils. This temperature varies from 300 deg. K (temperature of the moderator) at 30 cm of the core to 350 deg. K in a hole of the core. 2. Slowing down of neutron: Four resonance detectors have been used (Au, In, Co, Mn). We can write a 1/E form of the spectrum. The linking up energy E M between thermal neutron spectrum and slowing down spectrum is about 0.23 eV and is free from the Maxwell spectrum temperature. The decrease of slowing down flux regarding thermal flux, farther from the core, has been showed. 3. Fast neutrons: We used 3 threshold detectors (Ni, Al, Mg). We supposed a E 1/2 e -βE from of the spectrum above 3 MeV. The values of β are in a range from 0.775, at the centre of the core and in a loop-hole, to 0,64 at about 30 cm of the core. 4. Continuous shape of the spectrum: The following interpolations give useful informations between the field where measurements have been made: between 340 eV and 10 keV: 1/E form between 10 keV and 330 keV: 1/(E σ S (E)) form (σ S (E) elastic scattering section on hydrogen) between 330 keV and 3 MeV: calculated form by the moments method (ref. BSR). (author) [fr

  6. Measurements of integral cross sections in the californium-252 fission neutron spectrum

    International Nuclear Information System (INIS)

    Alberts, W.G.; Guenther, E.; Matzke, M.; Rassl, G.

    1977-01-01

    In a low-scattering arrangement cross sections averaged over the californium-252 spontaneous fission neutron spectrum were measured. The reactions 27 Al(n,α) 46 Ti, 47 Ti, 48 Ti(n,p), 54 Fe, 56 Fe(n,p), 58 Ni(n,p), 64 Zn(n,p), 115 In(n,n') were studied in order to obtain a consistent set of threshold detectors used in fast neutron flux density measurements. Overall uncertainties between 2 and 2.5% could be achieved; corrections due to neutron scattering in source and samples are discussed

  7. Bayesian calibration of reactor neutron flux spectrum using activation detectors measurements: Application to CALIBAN reactor

    International Nuclear Information System (INIS)

    Cartier, J.; Casoli, P.; Chappert, F.

    2013-01-01

    In this paper, we present calibration methods in order to estimate reactor neutron flux spectrum and its uncertainties by using integral activation measurements. These techniques are performed using Bayesian and MCMC framework. These methods are applied to integral activation experiments in the cavity of the CALIBAN reactor. We estimate the neutron flux and its related uncertainties. The originality of this work is that these uncertainties take into account measurements uncertainties, cross-sections uncertainties and model error. In particular, our results give a very good approximation of the total flux and indicate that neutron flux from MCNP simulation for energies above about 5 MeV seems to overestimate the 'real flux'. (authors)

  8. Investigation of the influence of the neutron spectrum in determinations of integral cross-section ratios

    Energy Technology Data Exchange (ETDEWEB)

    Smith, D.L.

    1987-11-01

    Ratio measurements are routinely employed in studies of neutron interaction processes in order to generate new differential cross-section data or to test existing differential cross-section information through examination of the corresponding response in integral neutron spectra. Interpretation of such data requires that careful attention be given to details of the neutron spectra involved in these measurements. Two specific tasks are undertaken in the present investigation: (1) Using perturbation theory, a formula is derived which permits one to relate the ratio measured in a realistic quasimonoenergetic spectrum to the desired pure monoenergetic ratio. This expression involves only the lowest-order moments of the neutron energy distribution and corresponding parameters which serve to characterize the energy dependence of the differential cross sections, quantities which can generally be estimated with reasonable precision from the uncorrected data or from auxiliary information. (2) Using covariance methods, a general formalism is developed for calculating the uncertainty of a measured integral cross-section ratio which involves an arbitrary neutron spectrum. This formalism is employed to further examine the conditions which influence the sensitivity of such measured ratios to details of the neutron spectra and to their uncertainties. Several numerical examples are presented in this report in order to illustrate these principles, and some general conclusion are drawn concerning the development and testing of neutron cross-section data by means of ratio experiments. 16 refs., 1 fig., 4 tabs.

  9. Investigation of the influence of the neutron spectrum in determinations of integral cross-section ratios

    International Nuclear Information System (INIS)

    Smith, D.L.

    1987-11-01

    Ratio measurements are routinely employed in studies of neutron interaction processes in order to generate new differential cross-section data or to test existing differential cross-section information through examination of the corresponding response in integral neutron spectra. Interpretation of such data requires that careful attention be given to details of the neutron spectra involved in these measurements. Two specific tasks are undertaken in the present investigation: (1) Using perturbation theory, a formula is derived which permits one to relate the ratio measured in a realistic quasimonoenergetic spectrum to the desired pure monoenergetic ratio. This expression involves only the lowest-order moments of the neutron energy distribution and corresponding parameters which serve to characterize the energy dependence of the differential cross sections, quantities which can generally be estimated with reasonable precision from the uncorrected data or from auxiliary information. (2) Using covariance methods, a general formalism is developed for calculating the uncertainty of a measured integral cross-section ratio which involves an arbitrary neutron spectrum. This formalism is employed to further examine the conditions which influence the sensitivity of such measured ratios to details of the neutron spectra and to their uncertainties. Several numerical examples are presented in this report in order to illustrate these principles, and some general conclusion are drawn concerning the development and testing of neutron cross-section data by means of ratio experiments. 16 refs., 1 fig., 4 tabs

  10. SPECTRUM WEIGHTED RESPONSES OF SEVERAL DETECTORS IN MIXED FIELDS OF FAST AND THERMAL NEUTRONS

    Directory of Open Access Journals (Sweden)

    SANG IN KIM

    2014-04-01

    Full Text Available The spectrum weighted responses of various detectors were calculated to provide guidance on the proper selection and use of survey instruments on the basis of their energy response characteristics on the neutron fields. To yield the spectrum weighted response, the detector response functions of 17 neutron-measuring devices were numerically folded with each of the produced calibration neutron spectra through the in-house developed software ‘K-SWR’. The detectors’ response functions were taken from the IAEA Technical Reports Series No. 403 (TRS-403. The reference neutron fields of 21 kinds with 2 spectra groups with different proportions of thermal and fast neutrons have been produced using neutrons from the 241Am-Be sources held in a graphite pile, a bare 241Am-Be source, and a DT neutron generator. Fluence-average energy (Eave varied from 3.8 MeV to 16.9 MeV, and the ambient-dose-equivalent rate [H*(10/h] varied from 0.99 to 16.5 mSv/h.

  11. 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

  12. Calculation of the fast neutron flux spectrum in the MNSR inner irradiation site using the WIMSD4 code

    International Nuclear Information System (INIS)

    Khattab, K.

    2005-03-01

    The Miniature Neutron Source Reactor (MNSR) in Syria has five inner irradiation sites in the annulus Beryllium reflectors to analyze the unknown samples using the Neutron Activation Analysis technique and to produce medium and short half life isotopes. The fast neutron flux spectrum has a special importance in the MNSR reactor physics where this spectrum is required to measure the fast neutron flux in the MNSR inner irradiation sites. Hence, calculation of the fast neutron flux spectrum in the MNSR inner irradiation site is conducted in this work using the WIMSD4 code. The energy range is divided in the WIMSD4 to 69 energy groups. The first six energy groups represent the fast neutron ranging from 0.5 to 10 MeV. To calculate the fast neutron flux spectrum in the MNSR inner irradiation site using the WIMSD4 code, the MNSR is modeled as a super unit cell. This cell consists of three regions which are: the homogenized core, annulus Beryllium, and water. The fast neutron spectrum is calculated also using the U 235 fission neutron spectrum approximation. The U 235 fission neutron spectrum agrees very good with the WIMSD4 results when neutron energy exceeds 1 MeV, but it fails when the neutron energy ranges from 0.5 to 1 MeV. The WIMSD4 code is used as well to calculate the microscopic fission cross sections for the U 238 using six energy groups where a unit cell of U 238 is used since the U 238 is usually used to measure the fast neutron flux in the reactor. The macroscopic fission cross sections for the U 238 are calculated first then the microscopic fission cross sections are calculated knowing the U 238 atomic density. (Author)

  13. Calculation of the fast neutron flux spectrum in the MNSR inner irradiation site using the WIMSD4 code

    International Nuclear Information System (INIS)

    Khattab, K.

    2006-01-01

    The Miniature Neutron Source Reactor (MNSR) in Syria has five inner irradiation sites in the annulus Beryllium reflectors to analyze the unknown samples using the Neutron Activation Analysis technique and to produce medium and short half life isotopes. The fast neutron flux spectrum has a special importance in the MNSR reactor physics where this spectrum is required to measure the fast neutron flux in the MNSR inner irradiation sites. Hence, calculation of the fast neutron flux spectrum in the MNSR inner irradiation site is conducted in this work using the WIMSD4 code. The energy range is divided in the WIMSD4 to 69 energy groups. The first six energy groups represent the fast neutron ranging from 0.5 to 10 MeV. To calculate the fast neutron flux spectrum in the MNSR inner irradiation site using the WIMSD4 code, the MNSR is modeled as a super unit cell. This cell consists of three regions which are: the homogenized core, annulus Beryllium, and water. The fast neutron spectrum is calculated also using the U 235 fission neutron spectrum approximation. The U 235 fission neutron spectrum agrees very good with the WIMSD4 results when neutron energy exceeds 1 MeV, but it fails when the neutron energy ranges from 0.5 to 1 MeV. The WIMSD4 code is used as well to calculate the microscopic fission cross sections for the U 238 using six energy groups where a unit cell of U 238 is used since the U 238 is usually used to measure the fast neutron flux in the reactor. The macroscopic fission cross sections for the U 238 are calculated first then the microscopic fission cross sections are calculated knowing the U 238 atomic density. (Author)

  14. Electric excitations in liquid He4 and their role in neutron scattering spectrum

    International Nuclear Information System (INIS)

    Poluehktov, Yu.M.; Karnatsevich, L.V.

    2001-01-01

    Data of experiments on excitation spectrum in liquid He 4 by inelastic neutron scattering method are discussed. Exact solution of particle scattering in ideal Bose-gas is given. Influence of inter-particle interactions on the structure of many-particle Bose system is analysed qualitatively. 55 refs., 1 figs

  15. 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.

  16. Impact of statistical uncertainty of the neutron spectrum in the isotopic evolution of fuel

    International Nuclear Information System (INIS)

    Ortega, P.

    2012-01-01

    The results obtained and presented in this study for different calculation conditions (number of stories, number of steps burning, etc.) and their simultaneous impact on neutron spectrum and isotopic composition and a methodology is proposed to determine the minimum parameters for calculation given uncertainty in the results of isotopic composition with high burnup, both UO 2 and MOX fuel.

  17. Measurement of the neutron spectrum of the Big Ten critical assembly by lithium-6 spectrometry

    International Nuclear Information System (INIS)

    De Leeuw-Gierts, G.; De Leeuw, S.; Hansen, G.E.; Helmick, H.H.

    1979-01-01

    The central neutron-flux spectrum of the Los Alamos Scientific Laboratory's critical assembly, Big Ten, was measured with a 6 Li spectrometer and techniques developed at the Centre d'Etude de L'Energie Nucleaire, Mol, as part of an experimental program to establish the characteristics of Big Ten

  18. A novel neutron energy spectrum unfolding code using particle swarm optimization

    International Nuclear Information System (INIS)

    Shahabinejad, H.; Sohrabpour, M.

    2017-01-01

    A novel neutron Spectrum Deconvolution using Particle Swarm Optimization (SDPSO) code has been developed to unfold the neutron spectrum from a pulse height distribution and a response matrix. The Particle Swarm Optimization (PSO) imitates the bird flocks social behavior to solve complex optimization problems. The results of the SDPSO code have been compared with those of the standard spectra and recently published Two-steps Genetic Algorithm Spectrum Unfolding (TGASU) code. The TGASU code have been previously compared with the other codes such as MAXED, GRAVEL, FERDOR and GAMCD and shown to be more accurate than the previous codes. The results of the SDPSO code have been demonstrated to match well with those of the TGASU code for both under determined and over-determined problems. In addition the SDPSO has been shown to be nearly two times faster than the TGASU code. - Highlights: • Introducing a novel method for neutron spectrum unfolding. • Implementation of a particle swarm optimization code for neutron unfolding. • Comparing results of the PSO code with those of recently published TGASU code. • Match results of the PSO code with those of TGASU code. • Greater convergence rate of implemented PSO code than TGASU code.

  19. Iterative code for the reconstruction of the neutrons spectrum using the Bonner spheres

    International Nuclear Information System (INIS)

    Reyes H, A.; Ortiz R, J. M.; Vega C, H. R.

    2012-10-01

    The neutrons are the particles more difficult of detecting for their intrinsic nature. The absence of the neutrons charge makes that an interaction exists with the matter in a different way. The term radiation spectrometry can use to describe the measurement of the intensity of a radiation field with regard to the energy. The intensity distribution with relationship to the energy is commonly known as spectrum. A method to know the neutrons spectrum in the radiation fields to those that people are exposed is the use of the known system as spectrometry system of Bonner spheres, being the more used for the purposes of the radiological protection. The current interest in the electrons spectrometry has stimulated the development of several procedures to carry out the reconstruction of the spectra. During the last decades new codes have been developed such as BUNKIUT, Bums, Fruit, UMG, etc., however, these methods still present several inconveniences as the complexity in their use, the necessity of an expert user and a very near initial spectrum to the spectrum that is wanted to obtain. To solve the mentioned problems it was development the program NSDUAZ (Neutron Spectrometry and Dosimetry from Autonomous University of Zacatecas). The objective of the present work is to prove and to validate the code before mentioned making an analysis of likeness and differences and of advantages and disadvantages with relationship to the codes used at the present time. (Author)

  20. Measurement of the neutron spectrum of the Big Ten critical assembly by lithium-6 spectrometry

    International Nuclear Information System (INIS)

    Leeuw-Gierts, G. de; Leeuw, S. de

    1980-01-01

    The central neutron-flux spectrum of the Los Alamos Scientific Laboratory's critical assembly, Big Ten, was measured with a 6 Li spectrometer and techniques developed at the Centre d'Etude de l'Energie Nucleaire, Mol, as part of an experimental program to establish the characteristics of Big Ten

  1. The gravitational wave spectrum of non-axisymmetric, freely precessing neutron stars

    International Nuclear Information System (INIS)

    Broeck, Chris van den

    2005-01-01

    Evidence for free precession has been observed in the radio signature of several pulsars. Freely precessing pulsars radiate gravitationally at frequencies near the rotation rate and twice the rotation rate, which for rotation frequencies greater than ∼10 Hz is in the LIGO band. In older work, the gravitational wave spectrum of a precessing neutron star has been evaluated to first order in a small precession angle. Here, we calculate the contributions to second order in the wobble angle, and we find that a new spectral line emerges. We show that for reasonable wobble angles, the second-order line may well be observable with the proposed advanced LIGO detectors for precessing neutron stars as far away as the galactic centre. Observation of the full second-order spectrum permits a direct measurement of the star's wobble angle, oblateness and deviation from axisymmetry, with the potential to significantly increase our understanding of neutron star structure

  2. Nodal spectrum method for solving neutron diffusion equation

    International Nuclear Information System (INIS)

    Sanchez, D.; Garcia, C. R.; Barros, R. C. de; Milian, D.E.

    1999-01-01

    Presented here is a new numerical nodal method for solving static multidimensional neutron diffusion equation in rectangular geometry. Our method is based on a spectral analysis of the nodal diffusion equations. These equations are obtained by integrating the diffusion equation in X, Y directions and then considering flat approximations for the current. These flat approximations are the only approximations that are considered in this method, as a result the numerical solutions are completely free from truncation errors. We show numerical results to illustrate the methods accuracy for coarse mesh calculations

  3. A neutron spectrum unfolding computer code based on artificial neural networks

    Science.gov (United States)

    Ortiz-Rodríguez, J. M.; Reyes Alfaro, A.; Reyes Haro, A.; Cervantes Viramontes, J. M.; Vega-Carrillo, H. R.

    2014-02-01

    The Bonner Spheres Spectrometer consists of a thermal neutron sensor placed at the center of a number of moderating polyethylene spheres of different diameters. From the measured readings, information can be derived about the spectrum of the neutron field where measurements were made. Disadvantages of the Bonner system are the weight associated with each sphere and the need to sequentially irradiate the spheres, requiring long exposure periods. Provided a well-established response matrix and adequate irradiation conditions, the most delicate part of neutron spectrometry, is the unfolding process. The derivation of the spectral information is not simple because the unknown is not given directly as a result of the measurements. The drawbacks associated with traditional unfolding procedures have motivated the need of complementary approaches. Novel methods based on Artificial Intelligence, mainly Artificial Neural Networks, have been widely investigated. In this work, a neutron spectrum unfolding code based on neural nets technology is presented. This code is called Neutron Spectrometry and Dosimetry with Artificial Neural networks unfolding code that was designed in a graphical interface. The core of the code is an embedded neural network architecture previously optimized using the robust design of artificial neural networks methodology. The main features of the code are: easy to use, friendly and intuitive to the user. This code was designed for a Bonner Sphere System based on a 6LiI(Eu) neutron detector and a response matrix expressed in 60 energy bins taken from an International Atomic Energy Agency compilation. The main feature of the code is that as entrance data, for unfolding the neutron spectrum, only seven rate counts measured with seven Bonner spheres are required; simultaneously the code calculates 15 dosimetric quantities as well as the total flux for radiation protection purposes. This code generates a full report with all information of the unfolding in

  4. Computational uncertainties in silicon dioxide/plutonium intermediate neutron spectrum systems

    International Nuclear Information System (INIS)

    Jaegers, P.J.

    1997-01-01

    In the past several years, several proposals have been made for the long-term stabilization and storage of surplus fissile materials. Many of these proposed scenarios involve systems that have an intermediate neutron energy spectrum. Such intermediate-energy systems are dominated by scattering and fission events induced by neutrons ranging in energy from 1 eV to 100keV. To ensure adequate safety margins and cost effectiveness, it is necessary to have benchmark data for these intermediate-energy spectrum systems; however, a review of the nuclear criticality benchmarks indicates that no formal benchmarks are available. Nuclear data uncertainties have been reported for some types of intermediate-energy spectrum systems. Using a variety of Monte Carlo computer codes and cross-section sets, reported significant variations in the calculated k ∞ of intermediate-energy spectrum metal/ 235 U systems. We discuss the characteristics of intermediate neutron spectrum systems and some of the computational differences that can occur in calculating the k eff of these systems

  5. New Measurements and Calculations to Characterize the Caliban Pulsed Reactor Cavity Neutron Spectrum by the Foil Activation Method

    Energy Technology Data Exchange (ETDEWEB)

    Jacquet, X.; Casoli, P.; Authier, N.; Rousseau, G. [CEA, Centre de Valduc, 21120 Is-sur-Tille (France); Barsu, C. [Pl. de la fontaine, 25410 Corcelles-Ferrieres (France)

    2011-07-01

    Caliban is a cylindrical metallic core reactor mainly composed of uranium 235. It is operated by the Criticality and Neutron Science Research Laboratory located at the French Atomic Energy Commission research center in Valduc. As with other fast burst reactors, Caliban is used extensively for determining the responses of electronic parts or other objects and materials to neutron-induced displacements. Therefore, Caliban's irradiation characteristics, and especially its central cavity neutron spectrum, have to be very accurately evaluated. The foil activation method has been used in the past by the Criticality and Neutron Science Research Laboratory to evaluate the neutron spectrum of the different facilities it operated, and in particular to characterize the Caliban cavity spectrum. In order to strengthen and to improve our knowledge of the Caliban cavity neutron spectrum and to reduce the uncertainties associated with the available evaluations, new measurements have been performed on the reactor and interpreted by the foil activation method. A sensor set has been selected to sample adequately the studied spectrum. Experimental measured reaction rates have been compared to the results from UMG spectrum unfolding software and to values obtained with the activation code Fispact. Experimental and simulation results are overall in good agreement, although gaps exist for some sensors. UMG software has also been used to rebuild the Caliban cavity neutron spectrum from activation measurements. For this purpose, a default spectrum is needed, and one has been calculated with the Monte-Carlo transport code Tripoli 4 using the benchmarked Caliban description. (authors)

  6. Benchmark experiment on vanadium assembly with D-T neutrons. Leakage neutron spectrum measurement

    Energy Technology Data Exchange (ETDEWEB)

    Kokooo; Murata, I.; Nakano, D.; Takahashi, A. [Osaka Univ., Suita (Japan); Maekawa, F.; Ikeda, Y.

    1998-03-01

    The fusion neutronics benchmark experiments have been done for vanadium and vanadium alloy by using the slab assembly and time-of-flight (TOF) method. The leakage neutron spectra were measured from 50 keV to 15 MeV and comparison were done with MCNP-4A calculations which was made by using evaluated nuclear data of JENDL-3.2, JENDL-Fusion File and FENDL/E-1.0. (author)

  7. {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.

  8. Planetary method to measure the neutrons spectrum in lineal accelerators of medical use

    International Nuclear Information System (INIS)

    Vega C, H. R.; Benites R, J. L.

    2014-08-01

    A novel procedure to measure the neutrons spectrum originated in a lineal accelerator of medical use has been developed. The method uses a passive spectrometer of Bonner spheres. The main advantage of the method is that only requires of a single shot of the accelerator. When this is used around a lineal accelerator is necessary to operate it under the same conditions so many times like the spheres that contain the spectrometer, activity that consumes enough time. The developed procedure consists on situating all the spheres of the spectrometer at the same time and to realize the reading making a single shot. With this method the photo neutrons spectrum produced by a lineal accelerator Varian ix of 15 MV to 100 cm of the isocenter was determined, with the spectrum is determined the total flow and the ambient dose equivalent. (Author)

  9. Neutron energy spectrum from 120 GeV protons on a thick copper target

    Energy Technology Data Exchange (ETDEWEB)

    Shigyo, Nobuhiro; /Kyushu U.; Sanami, Toshiya; /KEK, Tsukuba; Kajimoto, Tsuyoshi; /Kyushu U.; Iwamoto, Yosuke; /JAEA, Ibaraki; Hagiwara, Masayuki; Saito, Kiwamu; /KEK, Tsukuba; Ishibashi, Kenji; /Kyushu U.; Nakashima, Hiroshi; Sakamoto, Yukio; /JAEA, Ibaraki; Lee, Hee-Seock; /Pohang Accelerator Lab.; Ramberg, Erik; /Fermilab

    2010-08-01

    Neutron energy spectrum from 120 GeV protons on a thick copper target was measured at the Meson Test Beam Facility (MTBF) at Fermi National Accelerator Laboratory. The data allows for evaluation of neutron production process implemented in theoretical simulation codes. It also helps exploring the reasons for some disagreement between calculation results and shielding benchmark data taken at high energy accelerator facilities, since it is evaluated separately from neutron transport. The experiment was carried out using a 120 GeV proton beam of 3E5 protons/spill. Since the spill duration was 4 seconds, protoninduced events were counted pulse by pulse. The intensity was maintained using diffusers and collimators installed in the beam line to MTBF. The protons hit a copper block target the size of which is 5cm x 5cm x 60 cm long. The neutrons produced in the target were measured using NE213 liquid scintillator detectors, placed about 5.5 m away from the target at 30{sup o} and 5 m 90{sup o} with respect to the proton beam axis. The neutron energy was determined by time-of-flight technique using timing difference between the NE213 and a plastic scintillator located just before the target. Neutron detection efficiency of NE213 was determined on basis of experimental data from the high energy neutron beam line at Los Alamos National Laboratory. The neutron spectrum was compared with the results of multiparticle transport codes to validate the implemented theoretical models. The apparatus would be applied to future measurements to obtain a systematic data set for secondary particle production on various target materials.

  10. Neutron spectrum perturbations due to scattering materials and their effect on the average neutron energy, the spectral index, and the hardness parameter

    International Nuclear Information System (INIS)

    Wright, H.L.; Meason, J.L.; Wolf, M.; Harvey, J.T.

    1976-01-01

    Measurements have been performed on the perturbing effect of a number of scattering materials by the 'free-field' neutron leakage spectrum from a Godiva Type Critical Assembly (White Sands Missile Range Fast Burst Reactor). The results of these measurements are interpreted in relation to some of the general parameters characterizing a neutron environment, namely, the average neutron energy >10 KeV, the spectral index and the hardness parameter. Three neutron spectrum measurements have been performed, each under different experimental configurations of scattering materials. Results from these measurements show the following with relation to the spectral index: (1) The neutron environment on the core surface and at 12-inches from the core surface (free-field) yield a spectral index of 6.8, (2) The neutron environment behind a 4.75-inch Plexiglas plate yield 4.6 for the spectral index and (3) The neutron environment behind a 2-inch aluminum plate yield 6.7 for the spectral index. It is concluded that the core surface and the 12-inch from core surface neutron environment are identical with the 'free-field' neutron environment at 20-inches when considering only those neutrons with energy >10 KeV. On the other hand, it appears that the 4.75 inches of Plexiglas severely perturbs the 'free-field' neutron environment, i.e., a much harder neutron spectrum >10 KeV. In the situation where 2-inches of aluminum is used as the perturbing medium, essentially no change in the neutron spectrum >10 KeV is noted

  11. Bayesian statistics applied to neutron activation data for reactor flux spectrum analysis

    International Nuclear Information System (INIS)

    Chiesa, Davide; Previtali, Ezio; Sisti, Monica

    2014-01-01

    Highlights: • Bayesian statistics to analyze the neutron flux spectrum from activation data. • Rigorous statistical approach for accurate evaluation of the neutron flux groups. • Cross section and activation data uncertainties included for the problem solution. • Flexible methodology applied to analyze different nuclear reactor flux spectra. • The results are in good agreement with the MCNP simulations of neutron fluxes. - Abstract: In this paper, we present a statistical method, based on Bayesian statistics, to analyze the neutron flux spectrum from the activation data of different isotopes. The experimental data were acquired during a neutron activation experiment performed at the TRIGA Mark II reactor of Pavia University (Italy) in four irradiation positions characterized by different neutron spectra. In order to evaluate the neutron flux spectrum, subdivided in energy groups, a system of linear equations, containing the group effective cross sections and the activation rate data, has to be solved. However, since the system’s coefficients are experimental data affected by uncertainties, a rigorous statistical approach is fundamental for an accurate evaluation of the neutron flux groups. For this purpose, we applied the Bayesian statistical analysis, that allows to include the uncertainties of the coefficients and the a priori information about the neutron flux. A program for the analysis of Bayesian hierarchical models, based on Markov Chain Monte Carlo (MCMC) simulations, was used to define the problem statistical model and solve it. The first analysis involved the determination of the thermal, resonance-intermediate and fast flux components and the dependence of the results on the Prior distribution choice was investigated to confirm the reliability of the Bayesian analysis. After that, the main resonances of the activation cross sections were analyzed to implement multi-group models with finer energy subdivisions that would allow to determine the

  12. Measurements of the neutron spectrum in transit to Mars on the Mars Science Laboratory.

    Science.gov (United States)

    Köhler, J; Ehresmann, B; Zeitlin, C; Wimmer-Schweingruber, R F; Hassler, D M; Reitz, G; Brinza, D E; Appel, J; Böttcher, S; Böhm, E; Burmeister, S; Guo, J; Lohf, H; Martin, C; Posner, A; Rafkin, S

    2015-04-01

    The Mars Science Laboratory spacecraft, containing the Curiosity rover, was launched to Mars on 26 November 2011. Although designed for measuring the radiation on the surface of Mars, the Radiation Assessment Detector (RAD) measured the radiation environment inside the spacecraft during most of the 253-day, 560-million-kilometer cruise to Mars. An important factor for determining the biological impact of the radiation environment inside the spacecraft is the specific contribution of neutrons with their high biological effectiveness. We apply an inversion method (based on a maximum-likelihood estimation) to calculate the neutron and gamma spectra from the RAD neutral particle measurements. The measured neutron spectrum (12-436 MeV) translates into a radiation dose rate of 3.8±1.2 μGy/day and a dose equivalent of 19±5 μSv/day. Extrapolating the measured spectrum (0.1-1000 MeV), we find that the total neutron-induced dose rate is 6±2 μGy/day and the dose equivalent rate is 30±10 μSv/day. For a 360 day round-trip from Earth to Mars with comparable shielding, this translates into a neutron induced dose equivalent of about 11±4 mSv. Copyright © 2015 The Committee on Space Research (COSPAR). Published by Elsevier Ltd. All rights reserved.

  13. Sustainable thorium nuclear fuel cycles: A comparison of intermediate and fast neutron spectrum systems

    Energy Technology Data Exchange (ETDEWEB)

    Brown, N.R., E-mail: nbrown@bnl.gov [Brookhaven National Laboratory, Upton, NY (United States); Powers, J.J. [Oak Ridge National Laboratory, Oak Ridge, TN (United States); Feng, B.; Heidet, F.; Stauff, N.E.; Zhang, G. [Argonne National Laboratory, Argonne, IL (United States); Todosow, M. [Brookhaven National Laboratory, Upton, NY (United States); Worrall, A.; Gehin, J.C. [Oak Ridge National Laboratory, Oak Ridge, TN (United States); Kim, T.K.; Taiwo, T.A. [Argonne National Laboratory, Argonne, IL (United States)

    2015-08-15

    Highlights: • Comparison of intermediate and fast spectrum thorium-fueled reactors. • Variety of reactor technology options enables self-sustaining thorium fuel cycles. • Fuel cycle analyses indicate similar performance for fast and intermediate systems. • Reproduction factor plays a significant role in breeding and burn-up performance. - Abstract: This paper presents analyses of possible reactor representations of a nuclear fuel cycle with continuous recycling of thorium and produced uranium (mostly U-233) with thorium-only feed. The analysis was performed in the context of a U.S. Department of Energy effort to develop a compendium of informative nuclear fuel cycle performance data. The objective of this paper is to determine whether intermediate spectrum systems, having a majority of fission events occurring with incident neutron energies between 1 eV and 10{sup 5} eV, perform as well as fast spectrum systems in this fuel cycle. The intermediate spectrum options analyzed include tight lattice heavy or light water-cooled reactors, continuously refueled molten salt reactors, and a sodium-cooled reactor with hydride fuel. All options were modeled in reactor physics codes to calculate their lattice physics, spectrum characteristics, and fuel compositions over time. Based on these results, detailed metrics were calculated to compare the fuel cycle performance. These metrics include waste management and resource utilization, and are binned to accommodate uncertainties. The performance of the intermediate systems for this self-sustaining thorium fuel cycle was similar to a representative fast spectrum system. However, the number of fission neutrons emitted per neutron absorbed limits performance in intermediate spectrum systems.

  14. Sustainable thorium nuclear fuel cycles: A comparison of intermediate and fast neutron spectrum systems

    International Nuclear Information System (INIS)

    Brown, N.R.; Powers, J.J.; Feng, B.; Heidet, F.; Stauff, N.E.; Zhang, G.; Todosow, M.; Worrall, A.; Gehin, J.C.; Kim, T.K.; Taiwo, T.A.

    2015-01-01

    Highlights: • Comparison of intermediate and fast spectrum thorium-fueled reactors. • Variety of reactor technology options enables self-sustaining thorium fuel cycles. • Fuel cycle analyses indicate similar performance for fast and intermediate systems. • Reproduction factor plays a significant role in breeding and burn-up performance. - Abstract: This paper presents analyses of possible reactor representations of a nuclear fuel cycle with continuous recycling of thorium and produced uranium (mostly U-233) with thorium-only feed. The analysis was performed in the context of a U.S. Department of Energy effort to develop a compendium of informative nuclear fuel cycle performance data. The objective of this paper is to determine whether intermediate spectrum systems, having a majority of fission events occurring with incident neutron energies between 1 eV and 10 5 eV, perform as well as fast spectrum systems in this fuel cycle. The intermediate spectrum options analyzed include tight lattice heavy or light water-cooled reactors, continuously refueled molten salt reactors, and a sodium-cooled reactor with hydride fuel. All options were modeled in reactor physics codes to calculate their lattice physics, spectrum characteristics, and fuel compositions over time. Based on these results, detailed metrics were calculated to compare the fuel cycle performance. These metrics include waste management and resource utilization, and are binned to accommodate uncertainties. The performance of the intermediate systems for this self-sustaining thorium fuel cycle was similar to a representative fast spectrum system. However, the number of fission neutrons emitted per neutron absorbed limits performance in intermediate spectrum systems

  15. The determination of neutron energy spectrum in reactor core C1 of reactor VR-1 Sparrow

    Energy Technology Data Exchange (ETDEWEB)

    Vins, M. [Department of Nuclear Reactors, Faculty of Nuclear Sciences and Physical Engineering, Czech Technical University, V Holesovickach 2, 180 00 Prague 8 (Czech Republic)], E-mail: vinsmiro@seznam.cz

    2008-07-15

    This contribution overviews neutron spectrum measurement, which was done on training reactor VR-1 Sparrow with a new nuclear fuel. Former nuclear fuel IRT-3M was changed for current nuclear fuel IRT-4M with lower enrichment of 235U (enrichment was reduced from former 36% to 20%) in terms of Reduced Enrichment for Research and Test Reactors (RERTR) Program. Neutron spectrum measurement was obtained by irradiation of activation foils at the end of pipe of rabit system and consecutive deconvolution of obtained saturated activities. Deconvolution was performed by computer iterative code SAND-II with 620 groups' structure. All gamma measurements were performed on Canberra HPGe. Activation foils were chosen according physical and nuclear parameters from the set of certificated foils. The Resulting differential flux at the end of pipe of rabit system agreed well with typical spectrum of light water reactor. Measurement of neutron spectrum has brought better knowledge about new reactor core C1 and improved methodology of activation measurement. (author)

  16. The effect of electromagnetic interactions on the proton spectrum in free neutron β-decay

    International Nuclear Information System (INIS)

    Bunatyan, G.G.

    2000-01-01

    In the β decay of an unpolarized free neutron, the effect of electromagnetic interactions on the proton recoil spectrum is studied in the light of the experiments which are carried out and planned for now. The corrections to the energy distribution of protons prove to amount to the value of a few per cent. Nowadays, this is substantial for obtaining with a high accuracy, of ∼ 1% or better, the characteristics of weak interactions by processing the data of the experiments on the proton distribution in the free neutron β-decay

  17. International intercomparison on the neutron flux density spectrum just before the REAL-80 project

    International Nuclear Information System (INIS)

    Ertek, C.

    1981-06-01

    This work briefly presents the results of the international intercomparison on the neutron flux density spectrum just before the REAL-80 intercomparison project. Some of the results of this intercomparison with a smaller number of laboratories will be also reflected in the REAL-80 project, therefore, it has some significant issues. This work is performed within the IAEA programme on standardization of reactor radiation measurements, one of the important objectives of which is the assistance of laboratories in Member States to implement or intercompare the multiple foil activation techniques for different neutron field measurements

  18. Neutron energy spectrum determination near the surface on the JET vacuum vessel using the multifoil activation technique

    Energy Technology Data Exchange (ETDEWEB)

    Pillon, M.; Jarvis, O.N.; Conroy, S. (Associazione EURATOM-ENEA sulla Fusione, Frascati (Italy) JET Joint Undertaking, Abingdonm Oxon (U.K.) Imperial College of Science, Technology and Medicine, London (U.K.))

    1990-03-01

    The activation of foils of zinc, indium, aluminium, copper and magnesium has been used as a means of examining the energy spectrum of neutrons produced by discharges in the Joint European Torus (JET). Several threshold reactions have been used together with a least-squares unfolding code to determine the 2.5 and 14 MeV neutron yields produced by the JET plasma. The analysis shows that the energy spectrum produced by downscattered neutrons is satisfactorily calculated with the MCNP neutron transport code.

  19. Fusion spectrum neutron source computation in "6LiD convertor for HFETR

    International Nuclear Information System (INIS)

    Sun Shouhua; Hu Yifei; Ye Bin

    2014-01-01

    A computation model of 14 MeV neutron from the "6LiD convertor has been established, the 14 MeV neutron sources and flux in the irradiation samples from the "6LiD convertor and the core have been computed separately, the neutron spectrum in the irradiation samples have been computed, too. The results show that the neutron sources that over 13 MeV account for 1 MeV above in the "6LiD convertor is 25.7%, 24.6% respectively, 14 MeV neutron sources get 4.31 × 10"1"3 n_T·s"-"1, 3.34 × 10"1"3 n_T·s"-"1, 14 MeV neutron flux get 2.66 × 10"1"0 n_T·cm"-"2·s"-"1, 3.53 × 10"1"0 n_T·cm"-"2·s"-"1, as He and H_2O charged in the irradiation capsule. (authors)

  20. Measured and Predicted Variations in Fast Neutron Spectrum when Penetrating Laminated Fe-D2O

    International Nuclear Information System (INIS)

    Aalto, E.; Sandlin, R.; Fraeki, R.

    1965-09-01

    Variations of the fast neutron spectrum in thin regions of alternating Fe and D O have been studied using threshold detectors (ln(n, n' ), S(n, p), Al(n, α)). The results have been compared to those calculated by two shielding codes (NRN and RASH D) of multigroup removal-diffusion type. The absolute fast spectrum calculated in our rather complicated configurations was found to agree with measurements within the same accuracy (a factor of two) as did the thermal flux. The calculated spectrum is slightly harder than the measured one, but the detailed variations (covering the range 1:5) in the form of the spectrum when penetrating Fe agree with observations to within 15-20 %. In and Al activities were found to be proportional to the integrated flux over 1 MeV throughout the whole configuration, while S showed the least proportionality

  1. Analysis of the fast-neutron spectrum inside the experimental cavity of the NRU Mk4 FN rod

    International Nuclear Information System (INIS)

    Leung, T.C.

    1995-01-01

    The fast-neutron (FN) rods in the NRU reactor provide a facility to study the effects of irradiation on CANDU reactor materials. The Mark 4 (Mk4) FN rods use natural uranium and supply fast-neutrons for experiments on irradiation creep and growth, and corrosion, for pressure- and calandria-tube materials. The neutron fluxes above 1 MeV are up to 2.7x10 17 n.m -2 .s -1 . This paper describes a calculation of the fast-neutron spectrum inside the NRU Mk4 FN rod cavity. The calculation was performed using the WIMS-AECL code, which is a multi-group transport code with two dimensional capabilities using the collision-probability method. Results for the fast-neutron spectrum above 1 MeV are presented in nine groups. The analysis confirms that the spectrum in the fast-neutron irradiation facility in NRU is representative of the actual irradiation spectrum for fast-neutron damage in a CANDU reactor. The effects of changes in specimen holder size, temperature, coolant density and fuel burnup on the fast neutron spectrum are also presented. (author). 9 refs., 3 tabs., 4 figs

  2. Thick-foils activation technique for neutron spectrum unfolding with the MINUIT routine-Comparison with GEANT4 simulations

    Science.gov (United States)

    Vagena, E.; Theodorou, K.; Stoulos, S.

    2018-04-01

    Neutron activation technique has been applied using a proposed set of twelve thick metal foils (Au, As, Cd, In, Ir, Er, Mn, Ni, Se, Sm, W, Zn) for off-site measurements to obtain the neutron spectrum over a wide energy range (from thermal up to a few MeV) in intense neutron-gamma mixed fields such as around medical Linacs. The unfolding procedure takes into account the activation rates measured using thirteen (n , γ) and two (n , p) reactions without imposing a guess solution-spectrum. The MINUIT minimization routine unfolds a neutron spectrum that is dominated by fast neutrons (70%) peaking at 0.3 MeV, while the thermal peak corresponds to the 15% of the total neutron fluence equal to the epithermal-resonances area. The comparison of the unfolded neutron spectrum against the simulated one with the GEANT4 Monte-Carlo code shows a reasonable agreement within the measurement uncertainties. Therefore, the proposed set of activation thick-foils could be a useful tool in order to determine low flux neutrons spectrum in intense mixed field.

  3. Microscopic integral cross section measurements in the Be(d,n) neutron spectrum for applications in neutron dosimetry, radiation damage and the production of long-lived radionuclides

    International Nuclear Information System (INIS)

    Smith, D.L.; Meadows, J.W.; Greenwood, L.R.

    1990-01-01

    Integral neutron-reaction cross sections have been measured, relative to the U-238 neutron fission cross-section standard, for 27 reactions which are of contemporary interest in various nuclear applications (e.g., fast-neutron dosimetry, neutron radiation damage and the production of long-lived activities which affect nuclear waste disposal). The neutron radiation field employed in this study was produced by bombarding a thick Be-metal target with 7-MeV deuterons from an accelerator. The experimental results are reported along with detailed information on the associated measurement uncertainties and their correlations. These data are also compared with corresponding calculated values, based on contemporary knowledge of the differential cross sections and of the Be(d,n) neutron spectrum. Some conclusions are reached on the utility of this procedure for neutron-reaction data testing

  4. Neutron spectrum effects on TRU recycling in Pb-Bi cooled fast reactor core

    International Nuclear Information System (INIS)

    Kim, Yong Nam; Kim, Jong Kyung; Park, Won Seok

    2003-01-01

    This study is intended to evaluate the dependency of TRU recycling characteristics on the neutron spectrum shift in a Pb-Bi cooled core. Considering two Pb-Bi cooled cores with the soft and the hard spectrum, respectively, various characteristics of the recycled core are carefully examined and compared with each other. Assuming very simplified fuel cycle management with the homogeneous and single region fuel loading, the burnup calculations are performed until the recycled core reached to the (quasi-) equilibrium state. The mechanism of TRU recycling toward the equilibrium is analyzed in terms of burnup reactivity and the isotopic compositions of TRU fuel. In the comparative analyses, the difference in the recycling behavior between the two cores is clarified. In addition, the basic safety characteristics of the recycled core are also discussed in terms of the Doppler coefficient, the coolant loss reactivity coefficient, and the effective delayed neutron fraction

  5. Microstructural evolution of pure tungsten neutron irradiated with a mixed energy spectrum

    Energy Technology Data Exchange (ETDEWEB)

    Koyanagi, Takaaki, E-mail: koyanagit@ornl.gov [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Kumar, N.A.P. Kiran [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Hwang, Taehyun [Tohoku University, Sendai, 980-8579 (Japan); Garrison, Lauren M.; Hu, Xunxiang [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States); Snead, Lance L. [Massachusetts Institute of Technology, Cambridge, MA 02139 (United States); Katoh, Yutai [Oak Ridge National Laboratory, Oak Ridge, TN 37831 (United States)

    2017-07-15

    Microstructures of single-crystal bulk tungsten (W) and polycrystalline W foil with a strong grain texture were investigated using transmission electron microscopy following neutron irradiation at ∼90–800 °C to 0.03–4.6 displacements per atom (dpa) in the High Flux Isotope Reactor with a mixed energy spectrum. The dominant irradiation defects were dislocation loops and small clusters at ∼90 °C. Additional voids were formed in W irradiated at above 460 °C. Voids and precipitates involving transmutation rhenium and osmium were the dominant defects at more than ∼1 dpa. We found a new phenomenon of microstructural evolution in irradiated polycrystalline W: Re- and Os-rich precipitation along grain boundaries. Comparison of results between this study and previous studies using different irradiation facilities revealed that the microstructural evolution of pure W is highly dependent on the neutron energy spectrum in addition to the irradiation temperature and dose.

  6. The magnetic recoil spectrometer for measurements of the absolute neutron spectrum at OMEGA and the NIF

    International Nuclear Information System (INIS)

    Casey, D. T.; Frenje, J. A.; Gatu Johnson, M.; Séguin, F. H.; Li, C. K.; Petrasso, R. D.; Glebov, V. Yu.; Katz, J.; Magoon, J.; Meyerhofer, D. D.; Sangster, T. C.; Shoup, M.; Ulreich, J.; Ashabranner, R. C.; Bionta, R. M.; Carpenter, A. C.; Felker, B.; Khater, H. Y.; LePape, S.; MacKinnon, A.

    2013-01-01

    The neutron spectrum produced by deuterium-tritium (DT) inertial confinement fusion implosions contains a wealth of information about implosion performance including the DT yield, ion-temperature, and areal-density. The Magnetic Recoil Spectrometer (MRS) has been used at both the OMEGA laser facility and the National Ignition Facility (NIF) to measure the absolute neutron spectrum from 3 to 30 MeV at OMEGA and 3 to 36 MeV at the NIF. These measurements have been used to diagnose the performance of cryogenic target implosions to unprecedented accuracy. Interpretation of MRS data requires a detailed understanding of the MRS response and background. This paper describes ab initio characterization of the system involving Monte Carlo simulations of the MRS response in addition to the commission experiments for in situ calibration of the systems on OMEGA and the NIF.

  7. The magnetic recoil spectrometer for measurements of the absolute neutron spectrum at OMEGA and the NIF.

    Science.gov (United States)

    Casey, D T; Frenje, J A; Johnson, M Gatu; Séguin, F H; Li, C K; Petrasso, R D; Glebov, V Yu; Katz, J; Magoon, J; Meyerhofer, D D; Sangster, T C; Shoup, M; Ulreich, J; Ashabranner, R C; Bionta, R M; Carpenter, A C; Felker, B; Khater, H Y; LePape, S; MacKinnon, A; McKernan, M A; Moran, M; Rygg, J R; Yeoman, M F; Zacharias, R; Leeper, R J; Fletcher, K; Farrell, M; Jasion, D; Kilkenny, J; Paguio, R

    2013-04-01

    The neutron spectrum produced by deuterium-tritium (DT) inertial confinement fusion implosions contains a wealth of information about implosion performance including the DT yield, ion-temperature, and areal-density. The Magnetic Recoil Spectrometer (MRS) has been used at both the OMEGA laser facility and the National Ignition Facility (NIF) to measure the absolute neutron spectrum from 3 to 30 MeV at OMEGA and 3 to 36 MeV at the NIF. These measurements have been used to diagnose the performance of cryogenic target implosions to unprecedented accuracy. Interpretation of MRS data requires a detailed understanding of the MRS response and background. This paper describes ab initio characterization of the system involving Monte Carlo simulations of the MRS response in addition to the commission experiments for in situ calibration of the systems on OMEGA and the NIF.

  8. Monte Carlo simulation of fission yields, kinetic energy, fission neutron spectrum and decay γ-ray spectrum for 232Th(n,f) reaction induced by 3H(d,n) 4He neutron source

    International Nuclear Information System (INIS)

    Zheng Wei; Zeen Yao; Changlin Lan; Yan Yan; Yunjian Shi; Siqi Yan; Jie Wang; Junrun Wang; Jingen Chen; Chinese Academy of Sciences, Shanghai

    2015-01-01

    Monte Carlo transport code Geant4 has been successfully utilised to study of neutron-induced fission reaction for 232 Th in the transport neutrons generated from 3 H(d,n) 4 He neutron source. The purpose of this work is to examine the applicability of Monte Carlo simulations for the computation of fission reaction process. For this, Monte Carlo simulates and calculates the characteristics of fission reaction process of 232 Th(n,f), such as the fission yields distribution, kinetic energy distribution, fission neutron spectrum and decay γ-ray spectrum. This is the first time to simulate the process of neutron-induced fission reaction using Geant4 code. Typical computational results of neutron-induced fission reaction of 232 Th(n,f) reaction are presented. The computational results are compared with the previous experimental data and evaluated nuclear data to confirm the certain physical process model in Geant4 of scientific rationality. (author)

  9. Bayesian calibration of reactor neutron flux spectrum using activation detectors measurements: Application to CALIBAN reactor

    Energy Technology Data Exchange (ETDEWEB)

    Cartier, J. [Commissariat a l' Energie Atomique et aux Energies Alternatives CEA, DAM, DIF, F-91297 Arpajon (France); Casoli, P. [Commissariat a l' Energie Atomique et aux Energies Alternatives CEA, DAM, Valduc, F-21120 Is sur Tille (France); Chappert, F. [Commissariat a l' Energie Atomique et aux Energies Alternatives CEA, DAM, DIF, F-91297 Arpajon (France)

    2013-07-01

    In this paper, we present calibration methods in order to estimate reactor neutron flux spectrum and its uncertainties by using integral activation measurements. These techniques are performed using Bayesian and MCMC framework. These methods are applied to integral activation experiments in the cavity of the CALIBAN reactor. We estimate the neutron flux and its related uncertainties. The originality of this work is that these uncertainties take into account measurements uncertainties, cross-sections uncertainties and model error. In particular, our results give a very good approximation of the total flux and indicate that neutron flux from MCNP simulation for energies above about 5 MeV seems to overestimate the 'real flux'. (authors)

  10. Calculated neutron spectrum from 800-MeV protons incident on a copper beam stop

    International Nuclear Information System (INIS)

    Perry, D.G.

    1975-10-01

    A Monte Carlo calculation was performed to obtain the neutron spectrum generated by 800-MeV protons incident on the LAMPF main copper beam stop. The total flux is calculated to be of the order of 10 13 n/cm 2 -sec-mA at full-beam intensity of 1 mA, with flux spectra calculated for angles of 20 0 , 30 0 , 60 0 , 90 0 , 120 0 , and 150 0 . (auth)

  11. Procedure to measure the neutrons spectrum around a lineal accelerator for radiotherapy

    International Nuclear Information System (INIS)

    Vega C, H. R.; Hernandez D, V. M.; Letechipia de L, C.; Benites R, J. L.; Salas L, M. A.

    2013-10-01

    An experimental procedure was developed, by means of Bonner spheres, to measure the neutrons spectrum around Linacs of medical use that only requires of a single shot of the accelerator; to this procedure we denominate Planetary or Isocentric method. One of the problems associated to the neutrons spectrum measurement in a radiotherapy room with lineal accelerator is because inside the room a mixed, intense and pulsed radiation field takes place affecting the detection systems based on active detector; this situation is solved using a passive detector. In the case of the Bonner spheres spectrometer the active detector has been substituted by activation detectors, trace detectors or thermoluminescent dosimeters. This spectrometer uses several spheres that are situated one at a time in the measurement point, this way to have the complete measurements group the accelerator should be operated, under the same conditions, so many times like spheres have the spectrometer, this activity can consume a long time and in occasions due to the work load of Linac to complicate the measurement process too. The procedure developed in this work consisted on to situate all the spectrometer spheres at the same time and to make the reading by means of a single shot, to be able to apply this procedure, is necessary that before the measurements two characteristics are evaluated: the cross-talking of the spheres and the symmetry conditions of the neutron field. This method has been applied to determine the photo-neutrons spectrum produced by a lineal accelerator of medical use Varian ix of 15 MV to 100 cm of the isocenter located to 5 cm of depth of a solid water mannequin of 30 x 30 x 15 cm. The spectrum was used to determine the total flow and the environmental dose equivalent. (Author)

  12. Fusion Neutronic Source deuterium endash tritium neutron spectrum measurements using natural diamond detectors

    International Nuclear Information System (INIS)

    Krasilnikov, A.V.; Kaneko, J.; Isobe, M.; Maekawa, F.; Nishitani, T.

    1997-01-01

    Two natural diamond detectors (NDDs) operating at room temperature were used for Fusion Neutronics Source (FNS) deuterium endash tritium (DT) neutron spectra measurements at different points around the tritium target and for different deuteron beam energies. Energy resolution of both NDDs were measured, with values 1.95% and 2.8%. Due to the higher energy resolution of one of the two NDDs studied it was possible to measure the shape of the DT neutron energy distribution and its broadening due to deuteron scattering inside the target. The influence of pulse pileup on the energy resolution of the combined system (NDD+electronics) at count rates up to 3.8x10 5 counts/s was investigated. A 3.58% energy resolution for the spectrometric system based on NDD and a 0.25 μs shaping time amplifier has been measured at a count rate of 5.7x10 5 counts/s. It is shown that special development of a fast pulse signal processor is necessary for NDD based spectrometry at count rates of approximately 10 6 counts/s. copyright 1997 American Institute of Physics

  13. Reconstruction of the neutron spectrum using an artificial neural network in CPU and GPU

    International Nuclear Information System (INIS)

    Hernandez D, V. M.; Moreno M, A.; Ortiz L, M. A.; Vega C, H. R.; Alonso M, O. E.

    2016-10-01

    The increase in computing power in personal computers has been increasing, computers now have several processors in the CPU and in addition multiple CUDA cores in the graphics processing unit (GPU); both systems can be used individually or combined to perform scientific computation without resorting to processor or supercomputing arrangements. The Bonner sphere spectrometer is the most commonly used multi-element system for neutron detection purposes and its associated spectrum. Each sphere-detector combination gives a particular response that depends on the energy of the neutrons, and the total set of these responses is known like the responses matrix Rφ(E). Thus, the counting rates obtained with each sphere and the neutron spectrum is related to the Fredholm equation in its discrete version. For the reconstruction of the spectrum has a system of poorly conditioned equations with an infinite number of solutions and to find the appropriate solution, it has been proposed the use of artificial intelligence through neural networks with different platforms CPU and GPU. (Author)

  14. Characteristics of SiC neutron sensor spectrum unfolding process based on Bayesian inference

    Energy Technology Data Exchange (ETDEWEB)

    Cetnar, Jerzy; Krolikowski, Igor [Faculty of Energy and Fuels AGH - University of Science and Technology, Al. Mickiewicza 30, 30-059 Krakow (Poland); Ottaviani, L. [IM2NP, UMR CNRS 7334, Aix-Marseille University, Case 231 -13397 Marseille Cedex 20 (France); Lyoussi, A. [CEA, DEN, DER, Instrumentation Sensors and Dosimetry Laboratory, Cadarache, F-13108 St-Paul-Lez-Durance (France)

    2015-07-01

    This paper deals with SiC detector signal interpretation in neutron radiation measurements in mixed neutron gamma radiation fields, which is called the detector inverse problem or the spectrum unfolding, and it aims in finding a representation of the primary radiation, based on the measured detector signals. In our novel methodology we resort to Bayesian inference approach. In the developed procedure the resultant spectra is unfolded form detector channels reading, where the estimated neutron fluence in a group structure is obtained with its statistical characteristic comprising of standard deviation and correlation matrix. In the paper we present results of unfolding process for case of D-T neutron source in neutron moderating environment. Discussions of statistical properties of obtained results are presented as well as of the physical meaning of obtained correlation matrix of estimated group fluence. The presented works has been carried out within the I-SMART project, which is part of the KIC InnoEnergy R and D program. (authors)

  15. The Prompt Fission Neutron Spectrum of 235U for Einc 0.7-5.0 MeV

    Energy Technology Data Exchange (ETDEWEB)

    Gomez, Jaime A. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Devlin, Matthew James [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Haight, Robert Cameron [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); O' Donnell, John M. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Lee, Hye Young [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Mosby, Shea Morgan [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Taddeucci, Terry Nicholas [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Kelly, Keegan John [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Fotiadis, Nikolaos [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Neudecker, Denise [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); White, Morgan Curtis [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Talou, Patrick [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Rising, Michael Evan [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Solomon, Clell Jeffrey Jr. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Wu, Ching-Yen [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Bucher, Brian Michael [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Buckner, Matthew Quinn [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Henderson, Roger Alan [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

    2017-03-23

    The Chi-Nu experiment aims to accurately measure the prompt fission neutron spectrum (PFNS) for the major actinides. At the Los Alamos Neutron Science Center (LANSCE), fission can be induced using the white neutron source. Using a two arm time of flight (T.O.F) technique; Chi-Nu presents a preliminary result of the low energy component of the 235U PFNS measured using an array of 22-Lithium glass scintillators.

  16. Measurement of the energy spectrum of the neutrons inside the neutron flux trap assembled in the center of the reactor core IPEN/MB-01

    Energy Technology Data Exchange (ETDEWEB)

    Bitelli, Ulysses d' Utra; Mura, Luiz Ernesto Credidio; Santos, Diogo Feliciano dos; Jerez, Rogerio; Mura, Luis Felipe Liamos, E-mail: ubitelli@ipen.br, E-mail: credidiomura@gmail.com [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

    2013-07-01

    This paper presents the neutron energy spectrum in the central position of a neutron flux trap assembled in the core center of the research nuclear reactor IPEN/MB-01 obtained by an unfolding method. To this end, have been used several different types of activation foils (Au, Sc, Ti, Ni, and plates) which have been irradiated in the central position of the reactor core (setting number 203) at a reactor power level of 64.57 ±2.91 watts . The activation foils were counted by solid-state detector HPGe (gamma spectrometry). The experimental data of nuclear reaction rates (saturated activity per target nucleus) and a neutron spectrum estimated by a reactor physics computer code are the main input data to get the most suitable neutron spectrum in the irradiation position obtained through SANDBP code: a neutron spectra unfolding code that use an iterative adjustment method. The adjustment resulted in 3.85 ± 0.14 10{sup 9} n cm{sup -2} s{sup -1} for the integral neutron flux, 2.41 ± 0.01 10{sup 9} n cm{sup -2} s{sup -1} for the thermal neutron flux, 1.09 ± 0.02 10{sup 9} n cm{sup -2} s{sup -1} for intermediate neutron flux and 3.41± 0.02 10{sup 8} n cm{sup -2} s{sup -1} for the fast neutrons flux. These results can be used to verify and validate the nuclear reactor codes and its associated nuclear data libraries, besides show how much is effective the use of a neutron flux trap in the nuclear reactor core to increase the thermal neutron flux without increase the operation reactor power level. The thermal neutral flux increased 4.04 ± 0.21 times compared with the standard configuration of the reactor core. (author)

  17. Differential and integral comparisons of three representations of the prompt neutron spectrum for the spontaneous fission of 252Cf

    International Nuclear Information System (INIS)

    Madland, D.G.; LaBauve, R.J.; Nix, J.R.

    1984-01-01

    Because of their importance as neutron standards, we present comparisons of measured and calculated prompt fission neutron spectra N(E) and average prompt neutron multiplicities anti nu/sub p/ for the spontaneous fission of 252 Cf. In particular, we test three representations of N(E) against recent experimental measurements of the differential spectrum and threshold integral cross sections. These representations are the Maxwellian spectrum, the NBS spectrum, and the Los Alamos spectrum of Madland and Nix. For the Maxwellian spectrum, we obtain the value of the Maxwellian temperature T/sub M/ by a least-squares adjustment to the experimental differential spectrum of Poenitz and Tamura. For the Los Alamos spectrum, a similar least-squares adjustment determines the nuclear level-density parameter a, which is the single unknown parameter that appears. The NBS spectrum has been previously constructed by adjustments to eight differential spectra measured during the period 1965 to 1974. Among these three representations, we find that the Los Alamos spectrum best reproduces both the differential and integral measurements, assuming ENDF/B-V cross sections in the calculation of the latter. Although the NBS spectrum reproduces the integral measurements fairly well, it fails to satisfactorily reproduce the new differential measurement, and the Maxwellian spectrum fails to satisfactorily reproduce the integral measurements. Additionally, we calculate a value of anti nu/sub p/ from the Los Alamos theory that is within approximately 1% of experiment. 25 references

  18. Determination of the neutron spectrum at different locations in the Argentine RA-1 Reactor

    International Nuclear Information System (INIS)

    Lerner, A.M.; Madariaga, M.R.

    1998-01-01

    Full text: It is well known that the RA-1 reactor is used to irradiate different types of materials with neutrons. The Radio dosimetry Group (which belongs to the Nuclear Regulatory Authority) uses its fast column for the design, calibration and set up of criticality dosimeters as well as for a quick assessment of the dose to workers in case of an accident. With such purpose, Au(1), Au under Cd and In(2) foils were irradiated to estimate absolute thermal, epithermal and fast neutron fluxes at the irradiation location. The accuracy of this estimation is higher when the response to the present neutron spectrum of the different materials constituting the detectors is better known. This, in turn, requires the previous knowledge of such spectrum (a detailed energy dependence of neutron flux) at the analysed location. In this work a neutronic calculation is presented at the fast irradiation location. The whole calculation was carried out following two different methodologies, and considering a power of 40 kW. The reactor and its surroundings were represented by a simplified one-dimensional model, as a concentric cylindrical set of regions. Figures are drawn representing fast and thermal fluxes (with the cut at 0.4 eV) as a function of the distance to the core centre. The neutron flux (in n/cm 2 sec.eV) as a function of energy is also shown at the fast irradiation location. Values of flux (in n/cm 2 .sec.eV) are also provided as a function of energy in other typical locations, as well as the equivalent integrated flux values (in n/cm 2 .sec). ((1) According to the reaction Au 197 (n,γ)Au 198 , having a cross section of σ 0 =98.8b for thermal neutrons. (2) According to the reaction In 115 (n,n')In 115m , with a cross section of some 70 mb for neutrons with energies above 1.2MeV). (author) [es

  19. Study of a neutronic potential of a modular fast spectrum ADS for radiotoxic waste transmutation

    International Nuclear Information System (INIS)

    Slessarev, I.; Arkhipov, V.

    1999-01-01

    transmutation are the principal criteria for ADS development in nuclear power (NP). It is known that fast spectrum ADS have promising parameters to be used for this goal. However, there is no international consensus yet and many innovative ADS concepts are contradictory. The principal goals have been to evaluate a maximum rate of waste transmutation in a fast spectrum ADS, to assess minimum TRU and MA Specific Fuel Inventory (SFI) and 'critical mass' corresponding to a given K eff , and to evaluate the sensitivity of fast spectrum ADS neutron balance to the external neutron spectrum (this is one of specific features of ADS). This parameter is correlated with the neutron source importance, which in turn affects the accelerator current demand and, hence, the ADS economics

  20. Fission neutron spectrum averaged cross sections for threshold reactions on arsenic

    International Nuclear Information System (INIS)

    Dorval, E.L.; Arribere, M.A.; Kestelman, A.J.; Comision Nacional de Energia Atomica, Cuyo Nacional Univ., Bariloche; Ribeiro Guevara, S.; Cohen, I.M.; Ohaco, R.A.; Segovia, M.S.; Yunes, A.N.; Arrondo, M.; Comision Nacional de Energia Atomica, Buenos Aires

    2006-01-01

    We have measured the cross sections, averaged over a 235 U fission neutron spectrum, for the two high threshold reactions: 75 As(n,p) 75 mGe and 75 As(n,2n) 74 As. The measured averaged cross sections are 0.292±0.022 mb, referred to the 3.95±0.20 mb standard for the 27 Al(n,p) 27 Mg averaged cross section, and 0.371±0.032 mb referred to the 111±3 mb standard for the 58 Ni(n,p) 58m+g Co averaged cross section, respectively. The measured averaged cross sections were also evaluated semi-empirically by numerically integrating experimental differential cross section data extracted for both reactions from the current literature. The calculations were performed for four different representations of the thermal-neutron-induced 235 U fission neutron spectrum. The calculated cross sections, though depending on analytical representation of the flux, agree with the measured values within the estimated uncertainties. (author)

  1. Evaluation of the transmutation of transuranic using neutrons spectrum from the spallation reaction

    Energy Technology Data Exchange (ETDEWEB)

    Gilberti, Mauricio; Pereira, Claubia, E-mail: mgilber@eletronuclear.gov.br [Eletrobras Termonuclear S.A. (ELETRONUCLEAR), Angra dos Reis, RJ (Brazil); Veloso, Maria A. Fortini, E-mail: claubia@nuclear.ufmg.br [Universidade Federal de Minas Gerais (UFMG), Belo Horizante, MG (Brazil). Dept. de Engenharia Nuclear

    2013-07-01

    The transmutation of transuranic was analyzed by simulating the neutron flux from different spallation sources across arrays of fissile material with isotopic composition PWR reprocessing. A simplified model of Accelerator-Driven Systems (ADS) containing target, moderator graphite, lead-bismuth coolant or sodium coolant, is used. The simulation was made using the particles transport code MCNPX 2.6.0 which allowed to evaluate the rate of transmutation of actinides (Np, Pu, Am, and Cm) at different locations in the system. The objective of the study is to evaluate which the behavior and influences the spectrum of the spallation in the transmutation without the contribution or interference of multiplier, medium subcritical, which would add the contribution of fission neutrons generated, thus interfering in the analysis. The arrangement enable to infer the influence of hardened neutron flux from the spallation reaction in the transmutation, the results show that this is independent of the target material chosen, and the spectrum of spallation has a negligible importance compared to the influence of moderation and scattering generated by the coolant or moderator used. (author)

  2. Evaluation of the transmutation of transuranic using neutrons spectrum from the spallation reaction

    International Nuclear Information System (INIS)

    Gilberti, Mauricio; Pereira, Claubia; Veloso, Maria A. Fortini

    2013-01-01

    The transmutation of transuranic was analyzed by simulating the neutron flux from different spallation sources across arrays of fissile material with isotopic composition PWR reprocessing. A simplified model of Accelerator-Driven Systems (ADS) containing target, moderator graphite, lead-bismuth coolant or sodium coolant, is used. The simulation was made using the particles transport code MCNPX 2.6.0 which allowed to evaluate the rate of transmutation of actinides (Np, Pu, Am, and Cm) at different locations in the system. The objective of the study is to evaluate which the behavior and influences the spectrum of the spallation in the transmutation without the contribution or interference of multiplier, medium subcritical, which would add the contribution of fission neutrons generated, thus interfering in the analysis. The arrangement enable to infer the influence of hardened neutron flux from the spallation reaction in the transmutation, the results show that this is independent of the target material chosen, and the spectrum of spallation has a negligible importance compared to the influence of moderation and scattering generated by the coolant or moderator used. (author)

  3. Average cross section measurements in U-235 fission neutron spectrum for some threshold reactions

    International Nuclear Information System (INIS)

    Maidana, N.L.

    1993-01-01

    The average cross section in the 235 U fission spectrum has been measured by the activation technique, for the following thresholds reactions: 115 In(n,n') 115m In, 232 Th(n,f) P.F., 46 , 47 , 48 Ti(n,p) 46,47 , 48 Sc, 55 Mn(n,2 n) 54 Mn, 51 V(n,α) 48 Sc, 90 Zr(n,2 n) 89 Zr, 93 Nb(n,2 n) 92m Nb, 58 Ni(n,2 n) 57 Ni, 24 Mg(n,p) 24 Na, 56 Fe(n,p) 56 Mn, 59 Co(n,α) 56 Mn and 63 Cu(n,α) 60 Co. The activation foils were irradiated close (∼ 4 mm) to the core of the IEA-R1 research reactor in the IPEN-CNEN/SP. The reactor was operated at 2 MW yielding a fast neutron flux around 5 x 10 12 n.cm -2 . s -1 . The neutron flux density was monitored by activation reactions with well known averaged cross sections and with effective thresholds above 1 MeV. The foil activities were measured in a calibrated HPGe spectrometer. The neutron spectrum has been calculated using the SAIPS unfolding system applied to the activation data. A detailed error analysis was performed using the covariance matrix methodology. The results were compared with those from other authors. (author)

  4. Optimization of artificial neural networks for the reconstruction of the neutrons spectrum and their equivalent doses

    International Nuclear Information System (INIS)

    Reyes A, A.; Ortiz R, J. M.; Reyes H, A.; Castaneda M, R.; Solis S, L. O.; Vega C, H. R.

    2014-08-01

    In this work was used the robust design methodology of artificial neural networks to determine a good topology of net able to solve with efficiency the problems of neutrons spectrometry and dosimetry. For the design of the topology of optimized net 36 different net architectures based on an orthogonal arrangement with a configuration L 9 (3 4 ), L 4 (3 2 ) were trained. For the training of the neural networks, was used a computer code developed in the ambient of Mat lab programming, which automates the process and analysis of the information, reducing the time used in this activity considerably for the investigator. For the training of the propagation nets forward was utilized a neutrons spectrum compendium published by the International Atomic Energy Agency, where of the total 80% was used for the training and 20% for the test, it trained with an inverse propagation algorithm being the entrance data the count rates corresponding to the 7 spheres of the spectrometric system of Bonner spheres, as exit data, the neural network obtains the neutrons spectrum expressed in 60 energy groups and are calculated of simultaneous way 15 dosimetric quantities. (Author)

  5. Use of new threshold detector 199Hg(n,n')/sup 199m/Hg for neutron spectrum unfolding

    International Nuclear Information System (INIS)

    Sakurai, K.

    1982-01-01

    The nuclear data for the 199 Hg(n,n')/sup 199m/Hg reaction are reviewed and the data are used for neutron spectrum unfolding. The neutron spectrum of the YAYOI glory-hole is unfolded by SAND II with 10 nuclear reactions including the 199 Hg(n,n')/sup 199m/Hg reaction. The ratio of the measured reaction rate to the calculated reaction rate is about 1:1.1 for the guess spectrum. The 199 Hg(n,n')/sup 199m/Hg, 115 In(n,n')/sup 115m/In, 103 Rh(n,n')/sup 103m/Rh reactions should be useful threshold detectors for the neutron dosimetry with low level fast neutron flux

  6. Research on neutron energy spectrum of the beryllium, iron and polyethylene shells assemblies injected by D-T neutron

    International Nuclear Information System (INIS)

    An, Li; Guo, Haiping; Wang, Xinhua

    2009-04-01

    To test a simulation code, the multi-shell assemblies were established, which were made of beryllium stainless steel and polyethylene from the interior to the outer. The symmetry axes are all in the line of the D + beam. The neutron energy spectra above 1 MeV were obtained in medium by the detector of stilbene crystal of φ18 min x 20 mm. The distance between source and the spherical surface was 30 cm and 50 cm. The measurement channels are in the angle 0 degree and 120 degree relative to D + beam direction. The measurement positions are 0 cm, 9.7 cm, 12.8 cm and 17.3 cm away from the center of the assembly in both directions. The spectrum in different positions of the multi-shell assemblies in medium were compared and analyzed. (authors)

  7. Neutron spectrum at 900 from 800 MeV (p,n) reactions on a Ta target

    International Nuclear Information System (INIS)

    Howe, S.D.; Lisowski, P.W.; King, N.S.P.; Russell, G.J.; Donnert, H.J.

    1979-01-01

    The neutron time-of-flight spectrum produced by a thick tantalum target bombarded by 800-MeV protons was measured at an angle of 90 0 . The data were taken at the Weapons Neutron Research facility by use of a cylindrical Ta target with a radius of 1.27 cm and a length of 15 cm. An NE-213 liquid scintillator was used to detect the neutrons over an energy range of 0.5 to 350 MeV. The neutron yield is presented and compared to a intranuclear-cascade/evaporation model prediction. 3 figures

  8. Possible dark energy imprints in the gravitational wave spectrum of mixed neutron-dark-energy stars

    Energy Technology Data Exchange (ETDEWEB)

    Yazadjiev, Stoytcho S. [Department of Theoretical Physics, Faculty of Physics, St. Kliment Ohridski University of Sofia, James Bourchier Blvd. 5, 1164 Sofia (Bulgaria); Doneva, Daniela D., E-mail: yazad@phys.uni-sofia.bg, E-mail: daniela.doneva@uni-tuebingen.de [Theoretical Astrophysics, IAAT, Eberhard-Karls University of Tübingen, Auf der Morgenstelle 10, 72076 Tübingen (Germany)

    2012-03-01

    In the present paper we study the oscillation spectrum of neutron stars containing both ordinary matter and dark energy in different proportions. Within the model we consider, the equilibrium configurations are numerically constructed and the results show that the properties of the mixed neuron-dark-energy star can differ significantly when the amount of dark energy in the stars is varied. The oscillations of the mixed neuron-dark-energy stars are studied in the Cowling approximation. As a result we find that the frequencies of the fundamental mode and the higher overtones are strongly affected by the dark energy content. This can be used in the future to detect the presence of dark energy in the neutron stars and to constrain the dark-energy models.

  9. Integral test on activation cross section of tag gas nuclides using fast neutron spectrum fields

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-03-01

    Activation cross sections of tag gas nuclides, which will be used for the failed fuel detection and location in FBR plants, were evaluated by the irradiation tests in the fast neutron spectrum fields in JOYO and YAYOI. The comparison of their measured radioactivities and the calculated values using the JENDL-3.2 cross section set showed that the C/E values ranged from 0.8 to 2.8 for the calibration tests in YAYOI and that the present accuracies of these cross sections were confirmed. (author)

  10. Determination of the neutron spectrum in the well Naiade attached to the reactor Nereide

    International Nuclear Information System (INIS)

    Capgras, Andree; Clement, Christophe; Sueur, Maurice.

    1977-11-01

    The spectral distribution of neutrons in the centre of the well Naiade attached to the Fontenay-aux-Roses reactor Nereide is studied. In the thermal, epithermal and over 2.2 MeV regions, activation detectors are used: 197 Au and 55 Mn (bare and under cadmium), and 58 Ni. In the energy band from a few keV to 2.2 MeV two recoil proton proportional counters are employed. Under these conditions the whole spectrum is studied, but some comments are made on the difficulties of interpreting the results obtained by either of these methods [fr

  11. First measurement of the VESUVIO neutron spectrum in the 30–80 MeV energy range using a Proton Recoil Telescope technique

    International Nuclear Information System (INIS)

    Cazzaniga, C; Tardocchi, M; Croci, G; Grosso, G; Rebai, M; Gorini, G; Frost, C; Rhodes, N J; Schooneveld, E M; Giacomelli, L; Hjalmarsson, A

    2013-01-01

    Measurements of the fast neutron energy spectrum at the ISIS spallation source are reported. The measurements were performed with a Proton Recoil Telescope consisting of a thin plastic foil placed in the neutron beam and two scintillator detectors. Results in the neutron energy range 30 MeV n < 80 MeV are in good agreement with Monte Carlo simulations of the neutron spectrum

  12. First measurement of the VESUVIO neutron spectrum in the 30-80 MeV energy range using a Proton Recoil Telescope technique

    Science.gov (United States)

    Cazzaniga, C.; Tardocchi, M.; Croci, G.; Frost, C.; Giacomelli, L.; Grosso, G.; Hjalmarsson, A.; Rebai, M.; Rhodes, N. J.; Schooneveld, E. M.; Gorini, G.

    2013-11-01

    Measurements of the fast neutron energy spectrum at the ISIS spallation source are reported. The measurements were performed with a Proton Recoil Telescope consisting of a thin plastic foil placed in the neutron beam and two scintillator detectors. Results in the neutron energy range 30 MeV < En < 80 MeV are in good agreement with Monte Carlo simulations of the neutron spectrum.

  13. Evaluating the performance of two neutron spectrum unfolding codes based on iterative procedures and artificial neural networks

    Energy Technology Data Exchange (ETDEWEB)

    Ortiz-Rodriguez, J. M.; Reyes Alfaro, A.; Reyes Haro, A.; Solis Sanches, L. O.; Miranda, R. Castaneda; Cervantes Viramontes, J. M. [Universidad Autonoma de Zacatecas, Unidad Academica de Ingenieria Electrica. Av. Ramon Lopez Velarde 801. Col. Centro Zacatecas, Zac (Mexico); Vega-Carrillo, H. R. [Universidad Autonoma de Zacatecas, Unidad Academica de Ingenieria Electrica. Av. Ramon Lopez Velarde 801. Col. Centro Zacatecas, Zac., Mexico. and Unidad Academica de Estudios Nucleares. C. Cip (Mexico)

    2013-07-03

    In this work the performance of two neutron spectrum unfolding codes based on iterative procedures and artificial neural networks is evaluated. The first one code based on traditional iterative procedures and called Neutron spectrometry and dosimetry from the Universidad Autonoma de Zacatecas (NSDUAZ) use the SPUNIT iterative algorithm and was designed to unfold neutron spectrum and calculate 15 dosimetric quantities and 7 IAEA survey meters. The main feature of this code is the automated selection of the initial guess spectrum trough a compendium of neutron spectrum compiled by the IAEA. The second one code known as Neutron spectrometry and dosimetry with artificial neural networks (NDSann) is a code designed using neural nets technology. The artificial intelligence approach of neural net does not solve mathematical equations. By using the knowledge stored at synaptic weights on a neural net properly trained, the code is capable to unfold neutron spectrum and to simultaneously calculate 15 dosimetric quantities, needing as entrance data, only the rate counts measured with a Bonner spheres system. Similarities of both NSDUAZ and NSDann codes are: they follow the same easy and intuitive user's philosophy and were designed in a graphical interface under the LabVIEW programming environment. Both codes unfold the neutron spectrum expressed in 60 energy bins, calculate 15 dosimetric quantities and generate a full report in HTML format. Differences of these codes are: NSDUAZ code was designed using classical iterative approaches and needs an initial guess spectrum in order to initiate the iterative procedure. In NSDUAZ, a programming routine was designed to calculate 7 IAEA instrument survey meters using the fluence-dose conversion coefficients. NSDann code use artificial neural networks for solving the ill-conditioned equation system of neutron spectrometry problem through synaptic weights of a properly trained neural network. Contrary to iterative procedures, in

  14. Evaluating the performance of two neutron spectrum unfolding codes based on iterative procedures and artificial neural networks

    Science.gov (United States)

    Ortiz-Rodríguez, J. M.; Reyes Alfaro, A.; Reyes Haro, A.; Solís Sánches, L. O.; Miranda, R. Castañeda; Cervantes Viramontes, J. M.; Vega-Carrillo, H. R.

    2013-07-01

    In this work the performance of two neutron spectrum unfolding codes based on iterative procedures and artificial neural networks is evaluated. The first one code based on traditional iterative procedures and called Neutron spectrometry and dosimetry from the Universidad Autonoma de Zacatecas (NSDUAZ) use the SPUNIT iterative algorithm and was designed to unfold neutron spectrum and calculate 15 dosimetric quantities and 7 IAEA survey meters. The main feature of this code is the automated selection of the initial guess spectrum trough a compendium of neutron spectrum compiled by the IAEA. The second one code known as Neutron spectrometry and dosimetry with artificial neural networks (NDSann) is a code designed using neural nets technology. The artificial intelligence approach of neural net does not solve mathematical equations. By using the knowledge stored at synaptic weights on a neural net properly trained, the code is capable to unfold neutron spectrum and to simultaneously calculate 15 dosimetric quantities, needing as entrance data, only the rate counts measured with a Bonner spheres system. Similarities of both NSDUAZ and NSDann codes are: they follow the same easy and intuitive user's philosophy and were designed in a graphical interface under the LabVIEW programming environment. Both codes unfold the neutron spectrum expressed in 60 energy bins, calculate 15 dosimetric quantities and generate a full report in HTML format. Differences of these codes are: NSDUAZ code was designed using classical iterative approaches and needs an initial guess spectrum in order to initiate the iterative procedure. In NSDUAZ, a programming routine was designed to calculate 7 IAEA instrument survey meters using the fluence-dose conversion coefficients. NSDann code use artificial neural networks for solving the ill-conditioned equation system of neutron spectrometry problem through synaptic weights of a properly trained neural network. Contrary to iterative procedures, in neural

  15. Evaluating the performance of two neutron spectrum unfolding codes based on iterative procedures and artificial neural networks

    International Nuclear Information System (INIS)

    Ortiz-Rodríguez, J. M.; Reyes Alfaro, A.; Reyes Haro, A.; Solís Sánches, L. O.; Miranda, R. Castañeda; Cervantes Viramontes, J. M.; Vega-Carrillo, H. R.

    2013-01-01

    In this work the performance of two neutron spectrum unfolding codes based on iterative procedures and artificial neural networks is evaluated. The first one code based on traditional iterative procedures and called Neutron spectrometry and dosimetry from the Universidad Autonoma de Zacatecas (NSDUAZ) use the SPUNIT iterative algorithm and was designed to unfold neutron spectrum and calculate 15 dosimetric quantities and 7 IAEA survey meters. The main feature of this code is the automated selection of the initial guess spectrum trough a compendium of neutron spectrum compiled by the IAEA. The second one code known as Neutron spectrometry and dosimetry with artificial neural networks (NDSann) is a code designed using neural nets technology. The artificial intelligence approach of neural net does not solve mathematical equations. By using the knowledge stored at synaptic weights on a neural net properly trained, the code is capable to unfold neutron spectrum and to simultaneously calculate 15 dosimetric quantities, needing as entrance data, only the rate counts measured with a Bonner spheres system. Similarities of both NSDUAZ and NSDann codes are: they follow the same easy and intuitive user's philosophy and were designed in a graphical interface under the LabVIEW programming environment. Both codes unfold the neutron spectrum expressed in 60 energy bins, calculate 15 dosimetric quantities and generate a full report in HTML format. Differences of these codes are: NSDUAZ code was designed using classical iterative approaches and needs an initial guess spectrum in order to initiate the iterative procedure. In NSDUAZ, a programming routine was designed to calculate 7 IAEA instrument survey meters using the fluence-dose conversion coefficients. NSDann code use artificial neural networks for solving the ill-conditioned equation system of neutron spectrometry problem through synaptic weights of a properly trained neural network. Contrary to iterative procedures, in neural

  16. NULIF: neutron spectrum generator, few-group constant calculator, and fuel depletion code

    International Nuclear Information System (INIS)

    Wittkopf, W.A.; Tilford, J.M.; Andrews, J.B. II; Kirschner, G.; Hassan, N.M.; Colpo, P.N.

    1977-02-01

    The NULIF code generates a microgroup neutron spectrum and calculates spectrum-weighted few-group parameters for use in a spatial diffusion code. A wide variety of fuel cells, non-fuel cells, and fuel lattices, typical of PWR (or BWR) lattices, are treated. A fuel depletion routine and change card capability allow a broad range of problems to be studied. Coefficient variation with fuel burnup, fuel temperature change, moderator temperature change, soluble boron concentration change, burnable poison variation, and control rod insertion are readily obtained. Heterogeneous effects, including resonance shielding and thermal flux depressions, are treated. Coefficients are obtained for one thermal group and up to three epithermal groups. A special output routine writes the few-group coefficient data in specified format on an output tape for automated fitting in the PDQ07-HARMONY system of spatial diffusion-depletion codes

  17. Sensitivity studies of the neutron multiplicity spectrum in the spallation of Pb targets

    International Nuclear Information System (INIS)

    Sinha, A.; Garg, S.B.; Srinivasan, M.

    1986-01-01

    The number of neutrons produced per incident proton in the spallation of Pb targets is of direct relevance to the design of accelerator breeders. The nuclear cascade initiated by high-energy protons in spallation targets is usually described by an intranuclear cascade evaporation (INCE) model. Even though this model describes various average nuclear properties of spallation targets fairly well, differential quantities such as energy spectra, angular spectra etc., are not reproduced within the limits of experimental uncertainty. One of the reasons for this is the uncertainty in the magnitude of the parameters involved in the model, notably the level density parameter Bsub(O) whose magnitude is quoted by different workers to be in the range of 8-20 MeV. The accuracy of Bsub(O) could be improved if we could experimentally determine a quantity which is much more sensitive to Bsub(O) than the average neutron yield. In this paper we discuss one such quantity, namely the neutron multiplicity spectrum (MS). We compute the MS due to the spallation of Pb targets of different sizes at proton energies of 1.5, 1.0 and 0.59 GeV using the Monte Carlo code HETC. It is noticed that for the 1.5 GeV proton case the probability P(ν) for leakage of ν neutrons for ν in the range of 60-65, changes by about 70% when Bsub(O) is varied from 8 to 20 MeV. The corresponding change in the average neutron yield is <20%. It is therefore suggested that an accurate measurement of the MS can serve as a useful tool to narrow down the range of uncertainty in the Bsub(O) parameter. (author)

  18. 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

  19. Rigidity spectrum of Forbush decrease calculated by neutron monitors data corrected and uncorrected for geomagnetic disturbances

    International Nuclear Information System (INIS)

    Alania, M V; Wawrzynczak, A; Sdobnov, V E; Kravtsova, M V

    2013-01-01

    Forbush decreases (Fd) of the galactic cosmic ray (GCR) intensity and geomagnetic storms are observed almost at the same time. Geomagnetic storm is a reason of significant disturbances of the magnetic cut off rigidity causing the distortion of the time profile of the Fd of the GCR intensity. We show some differences in the temporal changes of the rigidity spectra of Fd calculated by neutron monitors experimental data corrected and uncorrected for the changes of the geomagnetic cut off rigidity. Nevertheless, the general features of the temporal changes of the rigidity spectrum of Fd maintain as it was found in our previous investigations. Namely, at the beginning phase of Fd rigidity spectrum is relatively soft and gradually becomes hard up to reaching the minimum level of the GCR intensity; then the rigidity spectrum gradually becomes soft during the recovery phase of Fd. We also confirm that for the established temporal profiles of the rigidity spectrum of Fd a structural changes of the interplanetary magnetic field turbulence in the range of frequencies, 10 −-6 ÷10 −-5 Hz are responsible.

  20. Possible error-prone repair of neoplastic transformation induced by fission-spectrum neutrons

    Energy Technology Data Exchange (ETDEWEB)

    Hill, C K; Han, A; Elkind, M M

    1948-01-01

    An examination was made of the effect of fission-spectrum neutrons from the JANUS reactor at Argonne National Laboratory, delivered either as acute or protracted irradiation, on the incidence of neoplastic transformation in the C3H 10T1/2 mouse embryo cell line. Acute exposures were delivered at 10-38 cGy min/sup -1/, protracted exposures at 0.086 or 0.43 cGy min/sup -1/. The total doses for both ranged from 2.4 to 350 cGy. In the low dose region (2.4-80 cGy), there was a large enhancement in transformation frequency when the neutrons were delivered at the low dose rates compared with the high dose rates, but the survival of the cells was not significantly different between the two exposures conditions. Analysis of the intial parts of the curves shows that the regression line for protracted doses is about 9 times steeper than that for single acute exposures. Finally, the possibility is discussed that an ''error-prone'' repair process may be causing the enhanced transformation frequency by protracted neutron exposures.

  1. Neutron spectrum adjustment using reaction rate data acquired with a liquid dosimetry system

    International Nuclear Information System (INIS)

    Smith, D.L.; Ikeda, Y.; Uno, Y.; Maekawa, F.

    1997-01-01

    A dosimetry technique based on neutron activation of circulating water with dissolved salts is discussed. The neutron source was the FNS accelerator at JAERI, Tokai, Japan. Yttrium chloride hexahydrate (YCl 3· 6H 2 O) was the salt (264.9 grams dissolved in 16.094 liters of water). Gamma-ray yields were measured with an intrinsic Ge detector. The following reactions were examined: (1) 16 O(n,p) 16 N (E thresh = 10.245 MeV, t 1/2 = 7.13 sec, E γ = 6.129 MeV); (2) 37 Cl(n,p) 37 S (E thresh = 4.194 MeV, t 1/2 = 5.05 min, E γ = 3.104 MeV); (3) 89 Y(n,n') 89m Y (E thresh = 0.919 MeV, t 1/2 = 16.06 sec, E γ = 0.909 MeV). This paper describes use of the generalized least-squares (GLS) method to adjust the neutron spectrum

  2. An Evaluation of Neutron Energy Spectrum Effects in Iron Based on Molecular Dynamics Displacement Cascade Simulations

    International Nuclear Information System (INIS)

    Greenwood, L.R.; Stoller, R.E.

    1998-01-01

    The results of molecular dynamics (MD) displacement cascade simulations in bcc iron have been used to obtain effective cross sections for two measures of primary damage production: (1) the number of surviving point defects expressed as a fraction of the displacements calculated using the standard secondary displacement model of Norgett, Robinson, and Torrens (NRT), and (2) the fraction of the surviving interstitials contained in clusters that formed during the cascade event. Primary knockon atom spectra for iron obtained from the SPECTER code have been used to weight these MD-based damage production cross sections in order to obtain spectrally-averaged values for several locations in commercial fission reactors and materials test reactors. An evaluation of these results indicates that neutron energy spectrum differences between the various enviromnents do not lead to significant differences between the average primary damage formation parameters. In particular, the defect production cross sections obtained for PWR and BWR neutron spectra were not significantly different. The variation of the defect production cross sections as a function of depth into the reactor pressure vessel wall is used as a sample application of the cross sections. A slight difference between the attenuation behavior of the PWR and BWR was noted; this difference could be explained by a subtle difference in the energy dependence of the neutron spectra. Overall, the simulations support the continued use of dpa as a damage correlation parameter

  3. Methods of neutron spectrum calculation from measured reaction rates in saips. Part 1. Review of mathematical methods

    International Nuclear Information System (INIS)

    Bondars, Kh.Ya.; Lapenas, A.A.

    1981-01-01

    We adapted or used on ES EhVM, operating under the control of OS ES, the currently most common algorithms for calculating neutron spectra from measured reaction rates. These programs, together with the neutron cross-section and spectrum libraries, are part of the computerized information system SAIPS. The present article descibes the basic mathematical concepts used in the algorithms of the SAIPS calculation programs

  4. Evaluation of a new neutron energy spectrum unfolding code based on an Adaptive Neuro-Fuzzy Inference System (ANFIS).

    Science.gov (United States)

    Hosseini, Seyed Abolfazl; Esmaili Paeen Afrakoti, Iman

    2018-01-17

    The purpose of the present study was to reconstruct the energy spectrum of a poly-energetic neutron source using an algorithm developed based on an Adaptive Neuro-Fuzzy Inference System (ANFIS). ANFIS is a kind of artificial neural network based on the Takagi-Sugeno fuzzy inference system. The ANFIS algorithm uses the advantages of both fuzzy inference systems and artificial neural networks to improve the effectiveness of algorithms in various applications such as modeling, control and classification. The neutron pulse height distributions used as input data in the training procedure for the ANFIS algorithm were obtained from the simulations performed by MCNPX-ESUT computational code (MCNPX-Energy engineering of Sharif University of Technology). Taking into account the normalization condition of each energy spectrum, 4300 neutron energy spectra were generated randomly. (The value in each bin was generated randomly, and finally a normalization of each generated energy spectrum was performed). The randomly generated neutron energy spectra were considered as output data of the developed ANFIS computational code in the training step. To calculate the neutron energy spectrum using conventional methods, an inverse problem with an approximately singular response matrix (with the determinant of the matrix close to zero) should be solved. The solution of the inverse problem using the conventional methods unfold neutron energy spectrum with low accuracy. Application of the iterative algorithms in the solution of such a problem, or utilizing the intelligent algorithms (in which there is no need to solve the problem), is usually preferred for unfolding of the energy spectrum. Therefore, the main reason for development of intelligent algorithms like ANFIS for unfolding of neutron energy spectra is to avoid solving the inverse problem. In the present study, the unfolded neutron energy spectra of 252Cf and 241Am-9Be neutron sources using the developed computational code were

  5. Measurements of the absolute neutron fluence spectrum emitted at 00 and 900 from the Little-Boy replica

    International Nuclear Information System (INIS)

    Roberts, J.H.; Gold, R.; Preston, C.C.

    1986-01-01

    Nuclear research emulsions (NRE) have been used to characterize the neutron spectrum emitted by the Little-Boy replica. NRE were irradiated at the Little-Boy surface, as well as approximately 2 m from the center of the Little-Boy replica, using polar angles of 0 0 , 30 0 , 60 0 , and 90 0 . For the NRE exposed at 2 m, neutron background was determined using shadow shields of borated polyethylene. Emulsion scanning to date has concentrated exclusively on the 2-m, 0 0 and 2-m, 90 0 locations. Approximately 5000 proton-recoil tracks have been measured in NRE irradiated at each of these locations. At the 2-m, 90 0 location, the NRE neutron spectrum extends from 0.37 MeV up to 8.2 MeV; whereas the NRE neutron spectrum at the 2-m, 0 0 location is much softer and extends only up to 2.7 MeV. NRE neutron spectrometry results at these two locations are compared with both liquid scintillator neutron spectrometry and Monte Carlo calculations. (author)

  6. An optimum source neutron spectrum and holder shape for extra-corporal treatment of liver cancer by BNCT

    International Nuclear Information System (INIS)

    Nievaart, Sander; Moss, Ray; Sauerwein, Wolfgang; Malago, Massimo; Kloosterman, Jan Leen; Hagen, Tim van der; Dam, Hugo van

    2006-01-01

    In extra-corporal treatment of liver cancer by BNCT, it is desired to have an as homogeneous as possible thermal neutron field throughout the organ. Previous work has shown that when using an epithermal neutron beam, the shape of the holder in which the liver is placed is the critical factor. This study develops the notion further as to what is the optimum neutron spectrum to perform such treatments. In the design calculations, when using Monte Carlo techniques, it is shown that when the expected contributions of the source neutrons in every part of the liver is calculated, a linear optimization scheme such as the Simplex method results in a mix of thermal and epithermal source neutrons to get the highest homogeneity for the thermal neutron field. This optimisation method is demonstrated in 3 holder shapes: cuboid, cylindrical and spherical with each 3 volumes of 2, 4 and 6 litres. A 10 cm thick cuboid model, irradiated from both sides gives the highest homogeneity. The spherical (rotating) holder has the lowest homogeneity but the highest contribution of every source neutron to the thermal neutrons in the liver. This can be advantageous when using a relatively small sized neutron beam with a low strength. (author)

  7. A search for fine structure of the time-of-flight spectrum of the fission neutrons of 252Cf

    International Nuclear Information System (INIS)

    Scobie, J.; Scott, R.D.; Feather, N.; Vass, D.G.

    1977-01-01

    A standard time-of-flight arrangement, in which start pulses were supplied by fission fragments and stop pulses by neutrons, has been employed in an attempt to check recent claims of the existence of fine structures in the time-of-flight spectrum of the fission neutrons of 252 Cf. This structure, in the form of spikes with tails towards longer times, has been attributed to the emission of neutrons of short delay (with half-lives of a few to a hundred or so nanoseconds) in the fission process. It has not been possible to find any convincing evidence for the existence of such structure. (author)

  8. Spallation Neutron Spectrum on a Massive Lead/Paraffin Target Irradiated with 1 GeV Protons

    CERN Document Server

    Adam, J; Barashenkov, V S; Brandt, R; Golovatiouk, V M; Kalinnikov, V G; Katovsky, K; Krivopustov, M I; Kumar, V; Kumawat, H; Odoj, R; Pronskikh, V S; Solnyshkin, A A; Stegailov, V I; Tsoupko-Sitnikov, V M; Westmeier, W

    2004-01-01

    The spectra of gamma-ray emitted by decaying residual nuclei, produced by spallation neutrons with (n, xn), (n,xnyp), (n,p), (n,gamma) reactions in activation threshold detectors - namely, ^{209}Bi, ^{197}Au, ^{59}Co, ^{115}In, ^{232}Th, were measured in the Laboratory of Nuclear Problems (LNP), JINR, Dubna, Russia. Spallation neutrons were generated by bombarding a 20 cm long cylindrical lead target, 8 cm in diameter, surrounded by a 6 cm thick layer of paraffin moderator, with a 1 GeV proton beam from the Nuclotron accelerator. Reaction rates and spallation neutron spectrum were measured and compared with CASCADE code calculations.

  9. Neutron metrology in the L.F.R. Neutron flux density spectrum in the inner graphite reflector of the L.F.R

    International Nuclear Information System (INIS)

    Zsolnay, E.M.

    1979-01-01

    The neutron spectrum in the vertical central plug of the Low Flux Reactor has been determined experimentally. Sets of activation and fission detectors have been irradiated, and the neutron spectrum has been unfolded with aid of 3 special computer programs SAND-II, RFSP-JUEL and CRYSTAL BALL. Using these 3 programs calculations are made on the improvement ratio, which is defined as the ratio of the variance of the input flux density to that of the output flux density. A Monte Carlo error analysis is made to examine the quality of the 3 solution spectra. The results obtained with the different computer codes were compared, and showed a general agreement. The experiment confirmed that the shape of the spectrum in the intermediate energy region is near the 1/E pattern. (author)

  10. Planetary method to measure the neutrons spectrum in lineal accelerators of medical use; Metodo planetario para medir el espectro de neutrones en aceleradores lineales de uso medico

    Energy Technology Data Exchange (ETDEWEB)

    Vega C, H. R. [Universidad Autonoma de Zacatecas, Unidad Academica de Estudios Nucleares, Cipres No. 10, Fracc. La Penuela, 98068 Zacatecas (Mexico); Benites R, J. L., E-mail: fermineutron@yahoo.com [Centro Estatal de Cancerologia de Nayarit, Servicio de Seguridad Radiologica, Calzada de la Cruz 118 Sur, 63000 Tepic, Nayarit (Mexico)

    2014-08-15

    A novel procedure to measure the neutrons spectrum originated in a lineal accelerator of medical use has been developed. The method uses a passive spectrometer of Bonner spheres. The main advantage of the method is that only requires of a single shot of the accelerator. When this is used around a lineal accelerator is necessary to operate it under the same conditions so many times like the spheres that contain the spectrometer, activity that consumes enough time. The developed procedure consists on situating all the spheres of the spectrometer at the same time and to realize the reading making a single shot. With this method the photo neutrons spectrum produced by a lineal accelerator Varian ix of 15 MV to 100 cm of the isocenter was determined, with the spectrum is determined the total flow and the ambient dose equivalent. (Author)

  11. Neutron spectrum unfolding using genetic algorithm in a Monte Carlo simulation

    Energy Technology Data Exchange (ETDEWEB)

    Suman, Vitisha [Health Physics Division, Bhabha Atomic Research Centre, Mumbai 400085 (India); Sarkar, P.K., E-mail: pksarkar02@gmail.com [Manipal Centre for Natural Sciences, Manipal University, Manipal 576104 (India)

    2014-02-11

    A spectrum unfolding technique GAMCD (Genetic Algorithm and Monte Carlo based spectrum Deconvolution) has been developed using the genetic algorithm methodology within the framework of Monte Carlo simulations. Each Monte Carlo history starts with initial solution vectors (population) as randomly generated points in the hyper dimensional solution space that are related to the measured data by the response matrix of the detection system. The transition of the solution points in the solution space from one generation to another are governed by the genetic algorithm methodology using the techniques of cross-over (mating) and mutation in a probabilistic manner adding new solution points to the population. The population size is kept constant by discarding solutions having lesser fitness values (larger differences between measured and calculated results). Solutions having the highest fitness value at the end of each Monte Carlo history are averaged over all histories to obtain the final spectral solution. The present method shows promising results in neutron spectrum unfolding for both under-determined and over-determined problems with simulated test data as well as measured data when compared with some existing unfolding codes. An attractive advantage of the present method is the independence of the final spectra from the initial guess spectra.

  12. Neutron spectrum unfolding using genetic algorithm in a Monte Carlo simulation

    International Nuclear Information System (INIS)

    Suman, Vitisha; Sarkar, P.K.

    2014-01-01

    A spectrum unfolding technique GAMCD (Genetic Algorithm and Monte Carlo based spectrum Deconvolution) has been developed using the genetic algorithm methodology within the framework of Monte Carlo simulations. Each Monte Carlo history starts with initial solution vectors (population) as randomly generated points in the hyper dimensional solution space that are related to the measured data by the response matrix of the detection system. The transition of the solution points in the solution space from one generation to another are governed by the genetic algorithm methodology using the techniques of cross-over (mating) and mutation in a probabilistic manner adding new solution points to the population. The population size is kept constant by discarding solutions having lesser fitness values (larger differences between measured and calculated results). Solutions having the highest fitness value at the end of each Monte Carlo history are averaged over all histories to obtain the final spectral solution. The present method shows promising results in neutron spectrum unfolding for both under-determined and over-determined problems with simulated test data as well as measured data when compared with some existing unfolding codes. An attractive advantage of the present method is the independence of the final spectra from the initial guess spectra

  13. Program HEFEST for calculation of neutron spectrum on the basis of the activity of threshold detectors; Progam HEFEST za obradu neutronskog spektra na osnovu aktivnosti prag detektora

    Energy Technology Data Exchange (ETDEWEB)

    Cupac, S; Sokcic-Kostic, M; Pesic, M [Institute of nuclear sciences Boris Kidric, Vinca, Beograd (Yugoslavia)

    1985-07-01

    Program HEFEST for calculation of neutron spectrum on the basis of the activity of threshold detectors is described in this paper. After testing, program is used for the elaboration of the experimental results in determining the fast neutron spectrum on the coupled fast-thermal system on the reactor RB in IBK. (author)

  14. Kinetic energy spectrum and polarization of neutrons from the reaction 12C(p,n)X at 590 MeV

    International Nuclear Information System (INIS)

    Arnold, J.

    1998-01-01

    The kinetic energy spectrum and the polarization of the PSI neutron beam produced in the reaction 12 C(p,n)X at 0 with 590 MeV polarized protons were investigated. A strong energy dependence of the neutron beam polarization is observed which was not expected at the time the neutron beam was built. (orig.)

  15. Methods of neutron spectrum calculation from measured reaction velocities in SAIPS

    International Nuclear Information System (INIS)

    Berzonis, M.A.; Bondars, Kh.Ya.

    1981-01-01

    When a user (physicist) needs to perform calculations, he faces a number of problems: obtaining or generating calculation programs, comparing these programs, generating a library of reference spectra, study of calculated spectra and so on. This means routine work which is duplicated in many laboratories. To help solve these problems a computerized information system called SAIPS has been developed, some aspects of which are dealt with in references. The present paper gives a short description of data input into SAIPS and the basic principles of its utilization. SAIPS is based on the ES 1022 computer controlled by the operational system OS ES version 4.1. It contains the programs needed for unfolding spectra, neutron cross-section and reference spectrum libraries and the software for the main system and for computerized calculations

  16. Genetic algorithms - A new technique for solving the neutron spectrum unfolding problem

    International Nuclear Information System (INIS)

    Freeman, David W.; Edwards, D. Ray; Bolon, Albert E.

    1999-01-01

    A new technique utilizing genetic algorithms has been applied to the Bonner sphere neutron spectrum unfolding problem. Genetic algorithms are part of a relatively new field of 'evolutionary' solution techniques that mimic living systems with computer-simulated 'chromosome' solutions. Solutions mate and mutate to create better solutions. Several benchmark problems, considered representative of radiation protection environments, have been evaluated using the newly developed UMRGA code which implements the genetic algorithm unfolding technique. The results are compared with results from other well-established unfolding codes. The genetic algorithm technique works remarkably well and produces solutions with relatively high spectral qualities. UMRGA appears to be a superior technique in the absence of a priori data - it does not rely on 'lucky' guesses of input spectra. Calculated personnel doses associated with the unfolded spectra match benchmark values within a few percent

  17. Spectrum of the multigroup neutron transport operator for bounded spatial domains

    International Nuclear Information System (INIS)

    Larsen, E.W.

    1979-01-01

    The spectrum of the multigroup neutron transport operator A is studied for bounded spatial regions D which consist of a finite number of material subregions. Our main results provide simple conditions on the material cross sections which guarantee that (1) A possesses eigenvalues in the finite plane; (2) A possesses a ''leading'' eigenvalue lambda 0 which is real, not less than the real part of any other eigenvalue, and to which there corresponds at least one nonnegative eigenfunction psi/sub lambda/0; and (3) A possesses a ''dominant'' eigenvalue lambda 0 which is real, simple, greater than the real part of any other eigenvalue, and whose eigenfunction psi/sub lambda/0 satisfies psi/sub lambda/0> or =0 and ∫psi/sub lambda/0d 2 Ω>0. We give examples to illustrate the results and to show that a leading eigenvalue need not be simple, nor its eigenfunction(s) positive

  18. Influence of coated particle structure in thermal neutron spectrum energy range

    Energy Technology Data Exchange (ETDEWEB)

    Hansen, U; Teuchert, E

    1971-02-15

    The heterogenity due to lumping the fuel in coated particles affects the thermal neutron spectrum. A calculation model is discussed which, apart from some simplifying assumptions about the statistical distribution, allows a rigorous computation of effective cross sections for all nuclides of the heterogeneous medium. It is based on an exact computation of the neutron penetration probability through coating and kernel. The model is incorporated in a THERMOS-code providing a double heterogeneous cell calculation, which can be repeated automatically at different time steps in the depletion code system MAFIA-V.S.O.P.. A discussion of the effects of the coated particle structure is given by a comparison of calculations for heterogeneous and homogeneous fuel zones in pebble bed reactor elements. This is performed for enriched UO{sub 2} fuel and for a ThO{sub 2}-PuO{sub 2} mixture in the grains. Depending on the energy dependent total sigmas in the kernels the changes of the cross sections are ranging from 0.1% up to 45%. The influence on the spectrum averaged sigmas of the nuclides in the fresh UO{sub 2} fuel is lower than 1%. For the emerging {sup 240}Pu it increases up to 3.3% during irradiation. For the ThO{sub 2}-PuO{sub 2} fuel the averaged sigmas of the isotopes vary from 0.5% to 5.7% depending on the state of irradiation. Correspondingly there is an influence on the plutonium isotopic composition, on breeding ratios, and on the tilt of k{sub eff} during burnup which will be discussed in detail.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-08-21

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

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

    Science.gov (United States)

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

    2015-08-01

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

  1. High-resolution measurements of the DT neutron spectrum using new CD foils in the Magnetic Recoil neutron Spectrometer (MRS) on the National Ignition Facility

    Energy Technology Data Exchange (ETDEWEB)

    Gatu Johnson, M., E-mail: gatu@psfc.mit.edu; Frenje, J. A.; Li, C. K.; Petrasso, R. D.; Séguin, F. H. [Massachusetts Institute of Technology, Cambridge, Massachusetts 02139 (United States); Bionta, R. M.; Casey, D. T.; Eckart, M. J.; Grim, G. P.; Hartouni, E. P.; Hatarik, R.; Sayre, D. B.; Skulina, K.; Yeamans, C. B. [Lawrence Livermore National Laboratory, Livermore, California 94550 (United States); Farrell, M. P.; Hoppe, M.; Kilkenny, J. D.; Reynolds, H. G.; Schoff, M. E. [General Atomics, San Diego, California 92186 (United States)

    2016-11-15

    The Magnetic Recoil neutron Spectrometer (MRS) on the National Ignition Facility measures the DT neutron spectrum from cryogenically layered inertial confinement fusion implosions. Yield, areal density, apparent ion temperature, and directional fluid flow are inferred from the MRS data. This paper describes recent advances in MRS measurements of the primary peak using new, thinner, reduced-area deuterated plastic (CD) conversion foils. The new foils allow operation of MRS at yields 2 orders of magnitude higher than previously possible, at a resolution down to ∼200 keV FWHM.

  2. Sequential measurements of cosmic-ray neutron spectrum and dose rate at sea level in Sendai, Japan

    International Nuclear Information System (INIS)

    Nakamura, Takashi; Nunomiya, Tomoya; Abe, Shigeru; Terunuma, Kazutaka; Suzuki, Hiroyuki

    2005-01-01

    The cosmic-ray neutron energy spectrum and dose rate were measured sequentially for two years from April 2001 up to March 2003 by using three neutron detectors, a 3 He-loaded multi-moderator detector (Bonner ball), 12.7 cm diameter by 12.7 cm long NE213 organic liquid scintillator, and high-sensitivity rem (dose equivalent) counter at the Kawauchi campus of Tohoku University in Sendai, Japan of geomagnetic latitude, 29degN, and cutoff rigidity, 10.43 GV. The neutron spectrum has three major peaks, thermal energy peak, evaporation peak around 1 MeV and cascade peak around 100 MeV. The ambient neutron dose equivalent rates measured by the rem counter, and the Bonner ball keep almost constant values of 4.0 and 6.5 (nSv/h), respectively, throughout this time period, after atmospheric pressure correction, and it often decreased about 30% after a large Solar Flare, that is called as the Forbush decrease. The total neutron flux was also obtained by the Bonner ball measurements to be 7.5x10 -3 (ncm -2 ·s -1 ) in average. The altitude variation of neutron flux and dose was also investigated by comparing the measured results with other results measured at Mt. Fuji area and aboard an airplane, where the cutoff rigidities are similar. (author)

  3. Unfolding neutron spectrum with Markov Chain Monte Carlo at MIT research Reactor with He-3 Neutral Current Detectors

    Science.gov (United States)

    Leder, A.; Anderson, A. J.; Billard, J.; Figueroa-Feliciano, E.; Formaggio, J. A.; Hasselkus, C.; Newman, E.; Palladino, K.; Phuthi, M.; Winslow, L.; Zhang, L.

    2018-02-01

    The Ricochet experiment seeks to measure Coherent (neutral-current) Elastic Neutrino-Nucleus Scattering (CEνNS) using dark-matter-style detectors with sub-keV thresholds placed near a neutrino source, such as the MIT (research) Reactor (MITR), which operates at 5.5 MW generating approximately 2.2 × 1018 ν/second in its core. Currently, Ricochet is characterizing the backgrounds at MITR, the main component of which comes in the form of neutrons emitted from the core simultaneous with the neutrino signal. To characterize this background, we wrapped Bonner cylinders around a 32He thermal neutron detector, whose data was then unfolded via a Markov Chain Monte Carlo (MCMC) to produce a neutron energy spectrum across several orders of magnitude. We discuss the resulting spectrum and its implications for deploying Ricochet at the MITR site as well as the feasibility of reducing this background level via the addition of polyethylene shielding around the detector setup.

  4. Measuring Neutron Spectrum at MIT Research Reactor Utilizing He-3 Bonner Cylinder Approach with an Unfolding Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Leder, A. [MIT; Anderson, A. J. [Chicago U., KICP; Billard, J. [Lyon, IPN; Figueroa-Feliciano, E. [Northwestern U.; Formaggio, J. A. [MIT; Hasselkus, C. [Wisconsin U., Madison; Newman, E. [MIT; Palladino, K. [Wisconsin U., Madison; Phuthi, M. [MIT; Winslow, L. [MIT; Zhang, L. [MIT

    2017-10-02

    The Ricochet experiment seeks to measure Coherent (neutral-current) Elastic Neutrino-Nucleus Scattering using dark-matter-style detectors with sub-keV thresholds placed near a neutrino source, such as the MIT (research) Reactor (MITR), which operates at 5.5 MW generating approximately 2.2e18 neutrinos/second at the core. Currently, Ricochet is characterizing the backgrounds at MITR, the main component of which comes in the form of neutrons emitted from the core simultaneous with the neutrino signal. To characterize this background, we wrapped a Bonner cylinder around a He-3 thermal neutron detector, whose data was then unfolded to produce a neutron energy spectrum across several orders of magnitude. We discuss the resulting spectrum and its implications for deploying Ricochet in the future at the MITR site as well as the feasibility of reducing this background level via the addition of polyethylene shielding around the detector setup.

  5. A comparison of the predicted and observed reaction rates of various neutron detectors in a thermal reactor spectrum

    International Nuclear Information System (INIS)

    Hardiman, J.P.; Maunders, E.J.

    1963-08-01

    A number of the detectors commonly used in integral neutron spectrum measurements have been irradiated in the pitch moderator position of a Calder Hall lattice where the detailed energy spectrum is known from time of flight measurements. Predicted and observed reaction rates are generally in good agreement although they are brought into better agreement by a small modification to the spectrum. The predicted cadmium ratios are quite sensitive to the value adopted for the effective cadmium cut off energy, values of which were determined for various detectors using the Ferranti Mercury computer. The values varied over a wide range, although in every case only 40 mil. cadmium filters were used. (author)

  6. Dependence of the Ratio between the Resonance Integral and Thermal Neutron Cross Section on the Deviation of the Epithermal Neutron Spectrum from the 1/E Law

    International Nuclear Information System (INIS)

    Soliman, N.F.

    2012-01-01

    In k 0 - Neutron Activation Analysis (k 0 -NAA), the conversion from the tabulated Q 0 (ratio of the resonance integral to thermal neutron cross-section)to Q 0 (α) (α is the shape factor of the epithermal neutron flux, indicating the deviation of the epithermal neutron spectrum from the ideal 1/E shape) are calculated using a FORTRAN program. The calculations are done for most elements that can be detected by neutron activation using different values of the parameter (α) ranging from -0.1≤α≤+0.1. The obtained data are used to study the dependence of the values (α) on the irradiation position factor in (k 0 -NAA)equation for some selected isotopes differ in their resonance energy and its Q 0 values. The results show that, the irradiation factor is affective mainly for low thermal tro epithermal flux ratio f especially for Q 0 value greater than 50. so consequently determining the irradiation parameters α value is not needed for irradiation positions that rich with thermal neutron. But for high f values the irradiation position factor should be taken into account. On the other hand the constructed FORTRAN program can be used to calculate the value Q 0 (α) directly for different value of α

  7. Determination of the energy spectrum of the neutrons in the central thimble of the reactor core TRIGA Mark III

    International Nuclear Information System (INIS)

    Parra M, M. A.

    2014-01-01

    This thesis presents the neutron spectrum measurements inside the core of the TRIGA Mark III reactor at 1 MW power in steady-state, with the bridge placed in the center of the swimming pool, using several metallic threshold foils. The activation detectors are inserted in the Central Thimble of the reactor core, all the foils are irradiated in the same position and irradiation conditions (one by one). The threshold detectors are made of different materials such as: Au 197 , Ni 58 , In 115 , Mg 24 , Al 27 , Fe 58 , Co 59 and Cu 63 , they were selected to cover the full range the energies (10 -10 to 20 MeV) of the neutron spectrum in the reactor core. After the irradiation, the activation detectors were measured by means of spectrometry gamma, using a high resolution counting system with a hyper pure Germanium crystal, in order to obtain the saturation activity per target nuclide. The saturation activity is one of the main input data together with the initial spectrum, for the computational code SANDBP (hungarian version of the code SAND-II), which through an iterative adjustment, gives the calculated spectrum. The different saturation activities are necessary for the unfolding method, used by the computational code SANDBP. This research work is very important, since the knowledge of the energetic and spatial distribution of the neutron flux in the irradiation facilities, allows to characterize properly the irradiation facilities, just like, to estimate with a good precision various physics parameters of the reactor such as: neutron fluxes (thermal, intermediate and fast), neutronic dose, neutron activation analysis (NAA), spectral indices (cadmium ratio), buckling, fuel burnup, safety parameters (reactivity, temperature distribution, peak factors). In addition, the knowledge of the already mentioned parameters can give a best use of reactor, optimizing the irradiations requested by the users for their production process or research projects. (Author)

  8. Measured and Predicted Variations in Fast Neutron Spectrum in Massive Shields of Water and Concrete

    Energy Technology Data Exchange (ETDEWEB)

    Aalto, E; Sandlin, R; Fraeki, R

    1965-09-15

    The absolute magnitude, and the variations in form, of the fast neutron spectrum during deep penetration (0.8 - 1.1 metre) in massive shields of water, ordinary and magnetite concrete have been studied by using threshold detectors (In (n, h'), S(n,p), Al(n, {alpha})). The results have been compared with predictions by two rigorous (NIOBE, Moments method) and two non-rigorous (multigroup removal-diffusion) shielding codes (NRN, RASH D). The absolute results predicted were in general within 50% of the measured ones, i. e. showed as good or better accuracy than thermal and epithermal flux predictions in the same small-reactor configurations. No difference in accuracy was found between the rigorous and non-rigorous methods. The changes in the relative form of the spectrum (indicated by variations in the (Al/S) and (In/S) reaction rate ratios and amounting to factors up to 3 - 4 during a one metre penetration in water) were rather accurately (within 10 - 30%) predicted by all of the methods. The photonuclear excitation of the 335 keV level used for detecting the In(n, n') reaction was found to distort completely the In results in water at penetrations > 50 cm.

  9. Measurements of europium-152 depth profile of stone embankments exposed the Nagasaki atomic bomb for neutron spectrum analysis

    International Nuclear Information System (INIS)

    Tatsumi-Miyajima, Junko; Shimasaki, Tatsuya; Okajima, Shunzo; Takada, Jitsuya; Yoshida, Masahiro; Takao, Hideaki; Okumura, Yutaka; Nakazawa, Masaharu.

    1990-01-01

    Quantitative measurement of neutron-induced radionuclide of 152 Eu in rocks near the hypocenter (ground center of the atomic bomb explosion) in Nagasaki was performed to obtain the depth profiles and calculate the neutron energy spectrum. Core samples were drilled and taken from the stone embankments on both sides of river within a radius of 500 m from the hypocenter. After cutting each core into about 27 mm-thick sections, each section was measured its gamma-ray spectrum with a pure germanium semiconductor detector and analyzed a content of natural europium by the activation method. The highest value 8.0 x 10 -2 Bq/μg of 152 Eu at the time of the blast was obtained from the surface plates of rock cores collected near the hypocenter. The surface activity of cores was reduced with increasing the slant distances from the hypocenter. The slopes of the depth profiles were similar among samples taken from the same location. In order to analyze the depth profile of 152 Eu activity in rock andesite, experiments using a fast neutron reactor and thermal neutron reactor were carried out. Comparing the measurements on the A-bomb exposure rock with the simulated results at the reactors, among the experiments, the depth profile using the neutron moderator of 10 mm polyethylene was closed to that obtained from the A-bomb exposed samples. The experiment of thermal neutron incidence only could not reproduce the profiles from the A-bomb exposed samples. This fact indicates that the depth profiles of 152 Eu in rock exposed to the A-bomb include valuable information concerning the neutron spectrum and intensity. (author)

  10. Measurement of the energy spectrum of cosmic-ray induced neutrons aboard an ER-2 high-altitude airplane

    CERN Document Server

    Goldhagen, P E; Kniss, T; Reginatto, M; Singleterry, R C; Van Steveninck, W; Wilson, J W

    2002-01-01

    Crews working on present-day jet aircraft are a large occupationally exposed group with a relatively high average effective dose from galactic cosmic radiation. Crews of future high-speed commercial aircraft flying at higher altitudes would be even more exposed. To help reduce the significant uncertainties in calculations of such exposures, the atmospheric ionizing radiation (AIR) project, an international collaboration of 15 laboratories, made simultaneous radiation measurements with 14 instruments on five flights of a NASA ER-2 high-altitude aircraft. The primary AIR instrument was a highly sensitive extended-energy multisphere neutron spectrometer with lead and steel shells placed within the moderators of two of its 14 detectors to enhance response at high energies. Detector responses were calculated for neutrons and charged hadrons at energies up to 100 GeV using MCNPX. Neutron spectra were unfolded from the measured count rates using the new MAXED code. We have measured the cosmic-ray neutron spectrum (t...

  11. Calculation analysis of Wims/D4-Batan-2DIFF neutron spectrum on RSG-GAS with cadmium ratio

    International Nuclear Information System (INIS)

    Radianti, I.B.; Zuhair; Hamzah, A.

    1998-01-01

    The calculation analysis of WIMS/D4-BATAN-2DIFF neutron spectrum was performed by comparison the calculation result of cadmium ratio with the experiment result on CIP, IP2, IP3 and IP4 irradiation positions of RSG GAS tenth core. The foils of Au, Mn and Co were used for determination of the measured and calculated cadmium ratios. Spectrum calculation was done in 69 energy group with 541 energy group (till 10 MeV) cross section of foil absorption reaction. The difference values between cadmium ratio calculation and experiment result for all cases were in interval of 11.0%-26.3% which are out of measurement deviation range. From these result, it concluded that the use of WIM /D4 in generating group constant is not sufficient to obtain the neutron spectrum, especially for non-fuel region

  12. Neutron dynamics of fast-spectrum dedicated cores for waste transmutation

    International Nuclear Information System (INIS)

    Massara, S.

    2002-04-01

    Among different scenarios achieving minor actinide transmutation, the possibility of double strata scenarios with critical, fast spectrum, dedicated cores must be checked and quantified. In these cores, the waste fraction has to be at the highest level compatible with safety requirements during normal operation and transient conditions. As reactivity coefficients are poor in such critical cores (low delayed neutron fraction and Doppler feed-back, high coolant void coefficient), their dynamic behaviour during transient conditions must be carefully analysed. Three nitride-fuel configurations have been analysed: two liquid metal-cooled (sodium and lead) and a particle-fuel helium-cooled one. A dynamic code, MAT4 DYN, has been developed during the PhD thesis, allowing the study of loss of flow, reactivity insertion and loss of coolant accidents, and taking into account two fuel geometries (cylindrical and spherical) and two thermal-hydraulics models for the coolant (incompressible for liquid metals and compressible for helium). Dynamics calculations have shown that if the fuel nature is appropriately chosen (letting a sufficient margin during transients), this can counterbalance the bad state of reactivity coefficients for liquid metal-cooled cores, thus proving the interest of this kind of concept. On the other side, the gas-cooled core dynamics is very badly affected by the high value of the helium void coefficient (which is a consequence of the choice of a hard spectrum), this effect being amplified by the very low thermal inertia of particle-fuel design. So, a new kind of concept should be considered for a helium-cooled fast-spectrum dedicated core. (authors)

  13. Procedure to measure the neutrons spectrum around a lineal accelerator for radiotherapy; Procedimiento para medir el espectro de los neutrones en torno a un acelerador lineal para radioterapia

    Energy Technology Data Exchange (ETDEWEB)

    Vega C, H. R.; Hernandez D, V. M.; Letechipia de L, C. [Universidad Autonoma de Zacatecas, Unidad Academica de Estudios Nucleares, Cipres No. 10, Fracc. La Penuela, 98060 Zacatecas (Mexico); Benites R, J. L. [Servicios de Salud de Nayarit, Centro Estatal de Cancerologia, Calzada de la Cruz 116 Sur, 63000 Tepic, Nayarit (Mexico); Salas L, M. A., E-mail: fermineutron@yahoo.com [Universidad Autonoma de Zacatecas, Unidad Academica de Agronomia, Apdo. Postal 336, 98000 Zacatecas (Mexico)

    2013-10-15

    An experimental procedure was developed, by means of Bonner spheres, to measure the neutrons spectrum around Linacs of medical use that only requires of a single shot of the accelerator; to this procedure we denominate Planetary or Isocentric method. One of the problems associated to the neutrons spectrum measurement in a radiotherapy room with lineal accelerator is because inside the room a mixed, intense and pulsed radiation field takes place affecting the detection systems based on active detector; this situation is solved using a passive detector. In the case of the Bonner spheres spectrometer the active detector has been substituted by activation detectors, trace detectors or thermoluminescent dosimeters. This spectrometer uses several spheres that are situated one at a time in the measurement point, this way to have the complete measurements group the accelerator should be operated, under the same conditions, so many times like spheres have the spectrometer, this activity can consume a long time and in occasions due to the work load of Linac to complicate the measurement process too. The procedure developed in this work consisted on to situate all the spectrometer spheres at the same time and to make the reading by means of a single shot, to be able to apply this procedure, is necessary that before the measurements two characteristics are evaluated: the cross-talking of the spheres and the symmetry conditions of the neutron field. This method has been applied to determine the photo-neutrons spectrum produced by a lineal accelerator of medical use Varian ix of 15 MV to 100 cm of the isocenter located to 5 cm of depth of a solid water mannequin of 30 x 30 x 15 cm. The spectrum was used to determine the total flow and the environmental dose equivalent. (Author)

  14. The GEANT4 simulation study of the characteristic γ-ray spectrum of TNT under soil induced by DT neutron

    International Nuclear Information System (INIS)

    Qin Xue; Han Jifeng; Yang Chaowen

    2014-01-01

    The characteristic γ-ray spectrum of TNT under soil induced by DT neutron is measured based on the PFTNA demining system. GEANT4 Monte Carlo simulation toolkit is used to simulate the whole experimental procedure. The simulative spectrum is compared with the experimental spectrum. The result shows that they are mainly consistent. It is for the first time to analyze the spectrum by Monte Carlo simulation, the share of the background sources such as neutron, gamma are obtained, the contribution that the experimental apparatus such as shielding, detector sleeve, moderator make to the background is analysed. The study found that the effective gamma signal (from soil and TNT) is only 29% of the full-spectrum signal, and the background signal is more than 68% of the full-spectrum signal, which is mainly produced in the shielding and the detector sleeve. The simulation result shows that by gradually improving the shielding and the cadmium of the detector sleeve, the share of the effective gamma signal can increase to 36% and the background signal can fell 7% eventually. (authors)

  15. Experimental research of neutron yield and spectrum from deuterium gas-puff z-pinch on the GIT-12 generator at current above 2 MA

    Science.gov (United States)

    Cherdizov, R. K.; Fursov, F. I.; Kokshenev, V. A.; Kurmaev, N. E.; Labetsky, A. Yu; Ratakhin, N. A.; Shishlov, A. V.; Cikhardt, J.; Cikhardtova, B.; Klir, D.; Kravarik, J.; Kubes, P.; Rezac, K.; Dudkin, G. N.; Garapatsky, A. A.; Padalko, V. N.; Varlachev, V. A.

    2017-05-01

    The Z-pinch experiments with deuterium gas-puff surrounded by an outer plasma shell were carried out on the GIT-12 generator (Tomsk, Russia) at currents of 2 MA. The plasma shell consisting of hydrogen and carbon ions was formed by 48 plasma guns. The deuterium gas-puff was created by a fast electromagnetic valve. This configuration provides an efficient mode of the neutron production in DD reaction, and the neutron yield reaches a value above 1012 neutrons per shot. Neutron diagnostics included scintillation TOF detectors for determination of the neutron energy spectrum, bubble detectors BD-PND, a silver activation detector, and several activation samples for determination of the neutron yield analysed by a Sodium Iodide (NaI) and a high-purity Germanium (HPGe) detectors. Using this neutron diagnostic complex, we measured the total neutron yield and amount of high-energy neutrons.

  16. Determination of the neutrons energy spectrum in the central thimble of the reactor core TRIGA Mark III

    International Nuclear Information System (INIS)

    Parra M, M. A.; Luis L, M. A.; Raya A, R.; Cruz G, H. S.

    2013-10-01

    This work presents the measurement of the neutrons spectrum in energies in the central thimble of the reactor TRIGA Mark III to a power of 1 MW in stationary state, with the core in the center of the pool. To achieve this objective, several thin sheets were irradiated (one at the time) in the same position of the core. The activation probes were selected in such a way that covered the energy range (1 x 10 -10 to 20 MeV) of the neutrons spectrum in the reactor core, for this purpose thin sheets were used of 197 Au, 58 Ni, 115 In, 24 Mg, 27 Al, 58 Fe, 59 Co and 63 Cu. After the irradiation, the high energy gamma emissions of the activated thin sheets were measured by means of gamma spectrometry, in a counting system of high resolution, with a Hyper pure Germanium detector, obtaining this way the activity induced in the thin sheets whose magnitude is proportional to the intensity of the neutrons flow, this activity together to a theoretical initial spectrum are the main entrance data of the computational code SANDBP (Hungarian version of the code Sand-II) that uses the unfolding method for the calculation of the spectrum. (Author)

  17. Rapid method of calculating the fluence and spectrum of neutrons from a critical assembly and the resulting dose

    International Nuclear Information System (INIS)

    Bessis, J.

    1977-01-01

    The proposed calculation method is unsophisticated but rapid. The first part (computer code CRITIC), which is based on the Fermi age equation, evaluates the number of neutrons per fission emitted from a moderated critical assembly and their energy spectrum. The second part (computer code NARCISSE), which uses the current albedo for concrete, evaluates the product of neutron reflection on the walls and calculates the fluence resulting at any point in the room and its energy distribution by 21 groups. The results obtained are shown to compare satisfactorily with those obtained through the use of a Monte Carlo program

  18. Measurement of the time dependent neutron energy spectrum in the 'DENA' plasma focus device

    Energy Technology Data Exchange (ETDEWEB)

    Abdollahzadeh, M [Department of Physics, Imam Husein University, PO Box 16575-347, Tehran (Iran, Islamic Republic of); Sadat kiai, S M [Nuclear Science and Technology Research Institute (NSTRI), Nuclear Science Research School, A.E.O.I., PO Box 14155-1339, Tehran (Iran, Islamic Republic of); Babazadeh, A R [Physics Department, Qom University, PO Box 37165, Qom (Iran, Islamic Republic of)

    2008-10-15

    An extended time of flight method is used to determine the time dependent neutron energy spectrum in the Filippove type 'Dena' plasma focus (90 kJ, 25 kV, 288 {mu}F), filled with deuterium gas. An array of 5 detectors containing NE-102 plastic scintillators+photomultipliers is used. The number and position of the detectors are determined by a Monte Carlo program and the MCNP code. This paper briefly describes the simulation method and presents the experimental measurements and their results. The mechanisms of neutron production (thermonuclear and non-thermonuclear) and their time variations are discussed.

  19. Methods of neutron spectrum calculation from measured reaction rates in SAIPS. Part 2: Software and data input

    International Nuclear Information System (INIS)

    Berzonis, M.A.; Bondars, H.Ya.

    1981-08-01

    A brief description of the SAIPS software and the basic principles of its application is given. SAIPS contains programs needed to unfold spectra, libraries of neutron cross sections and reference spectra, and software for automatic calculation and for system maintenance. SAIPS offers the possibility of determining the reliability of an unfolded neutron spectrum and of planning measurements and calculations by varying different factors: the errors in the reaction rates, the errors in the cross sections used, the detector assembly, the unfolding programs, etc. SAIPS runs on the ES 1022 computer

  20. ATHENA2D, Simulation Hypothetical Recriticality Accident in a Thermal Neutron Spectrum

    International Nuclear Information System (INIS)

    1999-01-01

    neutronics and CFD equations are solved by successive-over relaxation (SOR) iteration. Chebychev extrapolation is available for the neutronics equations, although this option has not been thoroughly tested. The CFD equations use the semi-implicit method for pressure-linked equations (SIMPLE) iteration to achieve self-consistent solutions for mass, momentum and energy. Asymptotic acceleration of the pressure-correction equation (part of the SIMPLE iteration) is available. The one-dimensional thermal hydraulic fuel pin models employ fast tri-diagonal matrix inversion for a single time step. 3 - Restrictions on the complexity of the problem: ATHENA 2 D assumes moderator boiling occurs before fuel remelt, as there is no mechanism to handle fuel remelt and relocation. However, even the most severe transients simulated during the course of the PhD work showed that this is reasonable. In a thermal neutron spectrum, time constants are longer than those in a fast spectrum. Given typical fuel piece dimensions, moderator boiling always occurred before the peak fuel temperature reached the melting point. Another limitation is the single (liquid) phase CFD model. Boiling is treated, but only insofar as to calculate local void fractions that feed back to the neutronics equations through local cross section interpolation based on reduced moderator density. The reactivity effects of voids being explicitly transported away from their point of origin is not treated. However, these effects are believed to be small as the introduction of voids tends to be a primary shutdown mechanism for these severe transients. Improved two-phase modeling would only affect the details of the shutdown phase of the transient, not the total energy release. Other Limitations: Reactivity feedback effects arising from any potential fluidized bed motion of fuel particles in the debris bed is not treated. No fuel motion is modeled. Blackbody (radiative) heat transfer is not modeled. Radiolytic gas bubble formation and

  1. Measured and Predicted Variations in Fast Neutron Spectrum when Penetrating Laminated Fe-D{sub 2}O

    Energy Technology Data Exchange (ETDEWEB)

    Aalto, E; Sandlin, R; Fraeki, R

    1965-09-15

    Variations of the fast neutron spectrum in thin regions of alternating Fe and D{sub O} have been studied using threshold detectors (ln(n, n' ), S(n, p), Al(n, {alpha})). The results have been compared to those calculated by two shielding codes (NRN and RASH D) of multigroup removal-diffusion type. The absolute fast spectrum calculated in our rather complicated configurations was found to agree with measurements within the same accuracy (a factor of two) as did the thermal flux. The calculated spectrum is slightly harder than the measured one, but the detailed variations (covering the range 1:5) in the form of the spectrum when penetrating Fe agree with observations to within 15-20 %. In and Al activities were found to be proportional to the integrated flux over 1 MeV throughout the whole configuration, while S showed the least proportionality.

  2. Angular dependence of neutron yield and of spectrum of neutrons producted in pA and π-A interactions

    International Nuclear Information System (INIS)

    Bayukov, Yu.D.; Gavrilov, V.B.; Goryainov, N.A.

    1982-01-01

    Neutron spectra are measured in the T kinetic energy range from 6 up to 20a MeV. Neutrons escape from C, Cu, Pb, U nuclei under the angles of THETA=10 deg + 160 deg in p+A → n+x reaction at 7.5 GeV/c and in π - +A → n+x reaction at 5.0 GeV/c. In the 80-200 MeV secondary neutron energy range the obtained data are compared with the results of simultaneous measurements of proton spectra. The effect of itopic symmetry of fast nucleon yield from non-symmetric nuclei are under considereation. Division of contributions of quasi-free and deep inelastic nuclear processes to fast neutron formation is carried out on the basis of the data obtained

  3. Correction Factor Analysis Of Foil Activation And The Effect Of Neglecting The Correction On Neutron Flux And Spectrum Measurement; ANALISIS FAKTOR KOREKSI KEPING AKTIVASI DAN PENGARUH PENGABAIANNYA PADA PENGUKURAN FLUKS DAN SPEKTRUM NEUTRON

    Energy Technology Data Exchange (ETDEWEB)

    Radiyanti, Ita Budi; Hamzah, Amir; Pinem, Surian [Multipurpose Reactor Centre Indonesia, Serpong, (Indonesia)

    1996-04-15

    Foil activation method is commonly used in flux and neutron spectrum measurement in nuclear reactor and other research. The effect of the thickness, type of foil material and neutron spectrum shape on the self shielding correction and activities correction on the edges of the foil have been analyzed. Also the effect of neglecting those correction factors on neutron flux and spectrum measurement were analyzed. The calculation of the correction factor has been done by using the program which had been verified for several foils. The foils used are Au, In. Cu, Co and Dy of 0.00254 cm -0.127 cm thickness and 1.27 cm diameter. The result showed that the correction factor foils were not similar due to the variation of activation cross section and neutron spectrum shape. For the neutron spectrum in RS-2 multi purpose reactor GAS using foils of 0.00254 cm thick. The effect of neglecting correction factor on thermal flux measurement for Au, In, Co and Cu were less than -6%, for Dy was about -25%. On epithermal flux measurement for Au and In were about -60%, Co and Dy was -12% and -6%, for Cu less than -2%. The effect of neglecting correction factor on spectrum measurement was the change on the neutron flux density values along neutron energy region.

  4. Neutron-induced nuclear data for the MYRRHA fast spectrum facility

    Directory of Open Access Journals (Sweden)

    Romojaro Pablo

    2017-01-01

    Full Text Available The MYRRHA (Multi-purpose hYbrid Research Reactor for High-tech Applications concept is a flexible experimental lead-bismuth cooled and mixed-oxide (MOX fueled fast spectrum facility designed to operate both in sub-critical (accelerator driven and critical modes. One of the key issues for the safe operation of the reactor is the uncertainty assessment during the design works. The main objective of the European project CHANDA (solving CHAllenges in Nuclear DAta Work Package 10 is to improve MYRRHA relevant nuclear data in order to reduce the reactor parameter uncertainties derived from them. In order to achieve this goal, several tasks have been undertaken. First, a sensitivity study of MYRRHA integral parameters, such as energy dependent cross sections, fission spectra and neutron multiplicities, to nuclear data has been conducted resulting in a list of MYRRHA relevant quantities (nuclides and reactions. On the second task, an analysis of the existing experimental data and evaluations for the quantities included in the list has been carried out. In this framework, the impact on the multiplication factor of quantities from different nuclear data libraries for different nuclides, reactions and energy regions has been investigated on the MYRRHA MOX critical core model. As the next step, new experiments and evaluations will be performed in order to improve existing nuclear data libraries.

  5. The measurement of tripartition alpha particle low energy spectrum in 235U fission induced by thermal neutrons

    International Nuclear Information System (INIS)

    El Hage Sleiman, F.

    1980-01-01

    The energy spectrum of the α particles emitted in the thermal neutron induced fission of 235 U was measured from 11.5 MeV down to 2 MeV using the parabola mass spectrometer Lohengrin at the ILL high flux reactor. A Monte Carlo program, that simulates the α particle motion to the spectrometer, has been developed. Numerical results of Monte Carlo calculations for differents values of parameter are reported. The overall energy spectrum is slightly asymmetric at low energy. The possible reasons for the existence of this asymmetry are discussed [fr

  6. The magnetic recoil spectrometer (MRSt) for time-resolved measurements of the neutron spectrum at the National Ignition Facility (NIF)

    Energy Technology Data Exchange (ETDEWEB)

    Frenje, J. A., E-mail: jfrenje@psfc.mit.edu; Wink, C. W.; Gatu Johnson, M.; Li, C. K.; Séguin, F. H.; Petrasso, R. D. [Massachusetts Institute of Technology, Cambridge, Massachusetts 02139 (United States); Hilsabeck, T. J.; Kilkenny, J. D. [General Atomics, San Diego, California 92186 (United States); Bell, P.; Bionta, R.; Cerjan, C. [Lawrence Livermore National Laboratory, Livermore, California 94550 (United States)

    2016-11-15

    The next-generation magnetic recoil spectrometer for time-resolved measurements of the neutron spectrum has been conceptually designed for the National Ignition Facility. This spectrometer, called MRSt, represents a paradigm shift in our thinking about neutron spectrometry for inertial confinement fusion applications, as it will provide simultaneously information about the burn history and time evolution of areal density (ρR), apparent ion temperature (T{sub i}), yield (Y{sub n}), and macroscopic flows during burn. From this type of data, an assessment of the evolution of the fuel assembly, hotspot, and alpha heating can be made. According to simulations, the MRSt will provide accurate data with a time resolution of ∼20 ps and energy resolution of ∼100 keV for total neutron yields above ∼10{sup 16}. At lower yields, the diagnostic will be operated at a higher-efficiency, lower-energy-resolution mode to provide a time resolution of ∼20 ps.

  7. The magnetic recoil spectrometer (MRSt) for time-resolved measurements of the neutron spectrum at the National Ignition Facility (NIF).

    Science.gov (United States)

    Frenje, J A; Hilsabeck, T J; Wink, C W; Bell, P; Bionta, R; Cerjan, C; Gatu Johnson, M; Kilkenny, J D; Li, C K; Séguin, F H; Petrasso, R D

    2016-11-01

    The next-generation magnetic recoil spectrometer for time-resolved measurements of the neutron spectrum has been conceptually designed for the National Ignition Facility. This spectrometer, called MRSt, represents a paradigm shift in our thinking about neutron spectrometry for inertial confinement fusion applications, as it will provide simultaneously information about the burn history and time evolution of areal density (ρR), apparent ion temperature (T i ), yield (Y n ), and macroscopic flows during burn. From this type of data, an assessment of the evolution of the fuel assembly, hotspot, and alpha heating can be made. According to simulations, the MRSt will provide accurate data with a time resolution of ∼20 ps and energy resolution of ∼100 keV for total neutron yields above ∼10 16 . At lower yields, the diagnostic will be operated at a higher-efficiency, lower-energy-resolution mode to provide a time resolution of ∼20 ps.

  8. Optimization of Neutron Spectrum in Northwest Beam Tube of Tehran Research Reactor for BNCT, by MCNP Code

    Energy Technology Data Exchange (ETDEWEB)

    Zamani, M. [National Radiation Protection Department - NRPD, Atomic Energy Organization of Iran - AEOI, Tehran (Iran, Islamic Republic of); End of North Kargar st, Atomic Energy Organization of Iran, P.O. Box: 14155-1339, Tehran (Iran, Islamic Republic of); Kasesaz, Y.; Khalafi, H.; Shayesteh, M. [Radiation Application School, Nuclear Science and Technology Research Institute, AEOI, Tehran (Iran, Islamic Republic of)

    2015-07-01

    In order to gain the neutron spectrum with proper components specification for BNCT, it is necessary to design a Beam Shape Assembling (BSA), include of moderator, collimator, reflector, gamma filter and thermal neutrons filter, in front of the initial radiation beam from the source. According to the result of MCNP4C simulation, the Northwest beam tube has the most optimized neuron flux between three north beam tubes of Tehran Research Reactor (TRR). So, it has been chosen for this purpose. Simulation of the BSA has been done in four above mentioned phases. In each stage, ten best configurations of materials with different length and width were selected as the candidates for the next stage. The last BSA configuration includes of: 78 centimeters of air as an empty space, 40 centimeters of Iron plus 52 centimeters of heavy-water as moderator, 30 centimeters of water or 90 centimeters of Aluminum-Oxide as a reflector, 1 millimeters of lithium (Li) as thermal neutrons filter and finally 3 millimeters of Bismuth (Bi) as a filter of gamma radiation. The result of Calculations shows that if we use this BSA configuration for TRR Northwest beam tube, then the best neutron flux and spectrum will be achieved for BNCT. (authors)

  9. Optimization of Neutron Spectrum in Northwest Beam Tube of Tehran Research Reactor for BNCT, by MCNP Code

    International Nuclear Information System (INIS)

    Zamani, M.; Kasesaz, Y.; Khalafi, H.; Shayesteh, M.

    2015-01-01

    In order to gain the neutron spectrum with proper components specification for BNCT, it is necessary to design a Beam Shape Assembling (BSA), include of moderator, collimator, reflector, gamma filter and thermal neutrons filter, in front of the initial radiation beam from the source. According to the result of MCNP4C simulation, the Northwest beam tube has the most optimized neuron flux between three north beam tubes of Tehran Research Reactor (TRR). So, it has been chosen for this purpose. Simulation of the BSA has been done in four above mentioned phases. In each stage, ten best configurations of materials with different length and width were selected as the candidates for the next stage. The last BSA configuration includes of: 78 centimeters of air as an empty space, 40 centimeters of Iron plus 52 centimeters of heavy-water as moderator, 30 centimeters of water or 90 centimeters of Aluminum-Oxide as a reflector, 1 millimeters of lithium (Li) as thermal neutrons filter and finally 3 millimeters of Bismuth (Bi) as a filter of gamma radiation. The result of Calculations shows that if we use this BSA configuration for TRR Northwest beam tube, then the best neutron flux and spectrum will be achieved for BNCT. (authors)

  10. Spectrum

    DEFF Research Database (Denmark)

    Høgfeldt Hansen, Leif

    2016-01-01

    The publication functions as a proces description of the development and construction of an urban furniture SPECTRUM in the city of Gwangju, Republic of Korea. It is used as the cataloque for the exhibition of Spectrum.......The publication functions as a proces description of the development and construction of an urban furniture SPECTRUM in the city of Gwangju, Republic of Korea. It is used as the cataloque for the exhibition of Spectrum....

  11. Contributions to the study of fast neutron spectrum in the 10 keV - 3 MeV range

    International Nuclear Information System (INIS)

    Garlea, I.

    1979-01-01

    The main objective of the work presented in this thesis was to create a fast neutron spectrum corresponding to the conditions required for a reference neutron field. The reference system for the fast neutron dosimetry in reactors, which the author promoted, is referred to as ΣΣ-ITN in the books. The conditions for introducing the ΣΣ systems into the thermal columns have been determined. The original contribution consists in determining the Westcott parameters of reactions 151 Eu(n,γ) 152 Eu and 176 Lu(n,γ) 17 +H7Lu used as thermal spectrum factors. The neutron description of the spectrum in cavity ΣΣ revealed that it is a Maxwell thermal spectrum displaying a temperature of 305+-7 deg C and a very small epithermal component (phisub(epi)/phisub(thermal) =4,5.10 -4 ). Better methods for determining reaction absolute rates resulted in less errors in calculating the microscopic integral sections mediated on the ΣΣ spectrum; there are under 5% errors for the fission cross sections and between 3% and 8% errors for the activating ones. The section values determined by the author have been included into the EXFOR library (IAEA); they are considered as reference measuremtns for the nuclear data improvement program. Testing the proposed method for the TRIGA on the ΣΣ-INT system proved that the multiple foil method provides correct results for both describing the spectral shape and for obtaining absolute values of the flux. Taking into account that the ΣΣ-ITN spectrum is a rapid one, the proposed method could not be tested within the low energy thermal and epithermal domain. For testing the method on an operational reactor, the core of the VVR-S IFIN reactor was employed. Due to the spectral structure of this reactor, it was possible to test the procedure within the whole energy range. In this view, the 5/10 core channel was selected which is similar to the channel required for measurements in the TRIGA-ROMANIA reactor. The absolute spectrum values are given in

  12. Neutron spectrum measurement inside containment vessel at Kori nuclear power plant unit 1

    Energy Technology Data Exchange (ETDEWEB)

    Han, J. M.; Kim, T. W.; Kim, K. D.; Youn, C. H. [Nuclear Environment Technology Institute, Taejon (Korea, Republic of)

    2003-10-01

    There would be a case for the radiation worker have to work inside of the containment vessel to inspect or repair reactor facilities. In this case, the information about distribution of neutron field is needed to estimate neutron exposure dose of worker. Neutron spectra were measured by BMS(Bonner Multisphere Spectrometer) at 4 points of 6 ft and 20 ft, 2 points of 44 ft, 5 points of 70 ft in containment vessel of Kori unit 1. From the calculation, the following results were obtained. Neutron fluxes of 6 ft were between 2.623 x 10{sup 2} and 2.746 x 10{sup 4} neutron/cm{sup 2}{center_dot}sec, average neutron energies were between 9.209 x 10{sup -6} and 3.377 x 10{sup -2} MeV, equivalent doses of neutron were between 0.025 and 2.675 mSv/hr. Neutron fluxes of 20 ft were between 1.771 x 10{sup 1} and 1.682 x 10{sup 3} neutron/cm{sup 2}{center_dot}sec, average neutron energies were between 6.084 x 10{sup -6} and 2.988 x 10{sup -1} MeV, equivalent doses of neutron were between 0.004 and 0.228 mSv/hr. Neutron fluxes of 44 ft were between 3.367 x 10{sup 2} and 3.483 x 10{sup 2} neutron / cm{sup 2}{center_dot}sec, average neutron energies were between 3.962 x 10{sup -2} and 7.360 x 10{sup -2} MeV, equivalent doses of neutron were between 0.069 and 0.089 mSv/hr. Neutron fluxes of 70 ft were between 4.553 x 10{sup 3} and 1.407 x 10{sup 4} neutron/cm{sup 2}{center_dot}sec, average neutron energies were between 3.668 x 10{sup -4} and 6.764 x 10{sup -2} MeV, equivalent doses of neutron were between 0.449 and 2.660 mSv/hr.

  13. The effect of neutron spectrum on the mechanical and physical properties of pure copper and copper alloys

    International Nuclear Information System (INIS)

    Fabritsiev, S.A.; Pokrovsky, A.S.; Sandakov, V.A.; Zinkle, S.J.; Rowcliffe, A.F.; Edwards, D.J.; Garner, F.A.; Singh, B.N.; Barabash, V.R.

    1996-01-01

    The electrical resistivity and tensile properties of copper and oxide dispersion strengthened (DS) copper alloys have been measured before and after fission neutron irradiation to damage levels of 0.5 to 5 displacements per atom (dps) at ∼100 to 400 degrees C. Some of the specimens were irradiated inside a 1.5 mm Cd shroud in order to reduce the thermal neutron flux. The electrical resistivity data could be separated into two components, a solid transmutation component Δρ tr which was proportional to thermal neutron fluence and a radiation defect component Δρ rd which was independent of the displacement dose. The saturation value for Δρ rd was ∼1.2 nanohm-meters for pure copper and ∼1.6 nanohm-meters for the DS copper alloys irradiated at 100 degrees C in positions with a fast-to-thermal neutron flux ratio of 5. Considerable radiation hardening was observed in all specimens at irradiation temperatures below 200 degrees C. The yield strength was relatively insensitive to neutron spectrum in specimens strengthened by dispersoids or cold- working. 17 refs., 7 figs., 1 tab

  14. Study of the influence of the fast neutron spectrum on the production of defects in solids and liquids

    International Nuclear Information System (INIS)

    Mas, P.; Droulers, Y.

    1964-01-01

    In the first part of this work a calculation has been made of the number of defects formed in graphite by a given neutron flux having various spectral distributions. The defect formation function is that of KINCHIN and PEASE; its formulation is briefly given. An efficiency function is then defined for a fast neutron spectrum. This defects produced in a light water reactor and those produced in a graphite reactor. Finally an application of this method is given for comparing the defect forming tendency in graphite in the case of the reactor Melusine and of the reactor G-2 and G-3. In the second part are calculated the integrals for the energy release brought about by fast neutrons in carbon oxygen and hydrogen. In a region of 25 cm around the core of a swimming-pool type reactor these energy release integrals are approximately proportional to the neutron flux above 1 MeV. The determination of the energy released as a result of the passage of neutrons in organic liquids can therefore be reduced to the measurement of the flux above 1 MeV for the real spectral distribution. A calorimetric verification has been carried out in the case of water. (authors) [fr

  15. Fission Product Data Measured at Los Alamos for Fission Spectrum and Thermal Neutrons on 239Pu, 235U, 238U

    International Nuclear Information System (INIS)

    Selby, H.D.; Mac Innes, M.R.; Barr, D.W.; Keksis, A.L.; Meade, R.A.; Burns, C.J.; Chadwick, M.B.; Wallstrom, T.C.

    2010-01-01

    We describe measurements of fission product data at Los Alamos that are important for determining the number of fissions that have occurred when neutrons are incident on plutonium and uranium isotopes. The fission-spectrum measurements were made using a fission chamber designed by the National Institute for Standards and Technology (NIST) in the BIG TEN critical assembly, as part of the Inter-laboratory Liquid Metal Fast Breeder Reactor (LMFBR) Reaction Rate (ILRR) collaboration. The thermal measurements were made at Los Alamos' Omega West Reactor. A related set of measurements were made of fission-product ratios (so-called R-values) in neutron environments provided by a number of Los Alamos critical assemblies that range from having average energies causing fission of 400-600 keV (BIG TEN and the outer regions of the Flattop-25 assembly) to higher energies (1.4-1.9 MeV) in the Jezebel, and in the central regions of the Flattop-25 and Flattop-Pu, critical assemblies. From these data we determine ratios of fission product yields in different fuel and neutron environments (Q-values) and fission product yields in fission spectrum neutron environments for 99 Mo, 95 Zr, 137 Cs, 140 Ba, 141,143 Ce, and 147 Nd. Modest incident-energy dependence exists for the 147 Nd fission product yield; this is discussed in the context of models for fission that include thermal and dynamical effects. The fission product data agree with measurements by Maeck and other authors using mass-spectrometry methods, and with the ILRR collaboration results that used gamma spectroscopy for quantifying fission products. We note that the measurements also contradict earlier 1950s historical Los Alamos estimates by ∼5-7%, most likely owing to self-shielding corrections not made in the early thermal measurements. Our experimental results provide a confirmation of the England-Rider ENDF/B-VI evaluated fission-spectrum fission product yields that were carried over to the ENDF/B-VII.0 library, except

  16. "Measurements of the neutron spectrum in transit to Mars on the Mars Science Laboratory", Köhler et al.

    Science.gov (United States)

    Miller, Jack

    2015-04-01

    The Mars Science Laboratory (MSL) spacecraft carried the Curiosity rover to Mars. While the dramatic, successful landing of Curiosity and its subsequent exploration of the Martian surface have justifiably generated great excitement, from the standpoint of the health of crewmembers on missions to Mars, knowledge of the environment between Earth and Mars is critical. This paper reports data taken during the cruise phase of the MSL by the Radiation Assessment Detector (RAD). The results are of great interest for several reasons. They are a direct measurement of the radiation environment during what will be a significant fraction of the duration of a proposed human mission to Mars; they were made behind the de facto shielding provided by various spacecraft components; and, in particular, they are a measurement of the contribution to radiation dose by neutrons. The neutron environment inside spacecraft is produced primarily by galactic cosmic ray ions interacting in shielding materials, and given the high biological effectiveness of neutrons and the increased contribution of neutrons to dose with increased depth in shielding, accurate knowledge of the neutron energy spectrum behind shielding is vital. The results show a relatively modest contribution from neutrons and gammas compared to that from charged particles, but also a discrepancy in both dose and dose rate between the data and simulations. The failure of the calculations to accurately reproduce the data is significant, given that future manned spacecraft will be more heavily shielded (and thus produce more secondary neutrons) and that spacecraft design will rely on simulations and model calculations of radiation transport. The methodology of risk estimation continues to evolve, and incorporates our knowledge of both the physical and biological effects of radiation. The relatively large uncertainties in the biological data, and the difficulties in reducing those uncertainties, makes it all the more important to

  17. Determination of neutron energy spectrum at a pneumatic rabbit station of a typical swimming pool type material test research reactor

    International Nuclear Information System (INIS)

    Malkawi, S.R.; Ahmad, N.

    2002-01-01

    The method of multiple foil activation was used to measure the neutron energy spectrum, experimentally, at a rabbit station of Pakistan Research Reactor-1 (PARR-1), which is a typical swimming pool type material test research reactor. The computer codes MSITER and SANDBP were used to adjust the spectrum. The pre-information required by the adjustment codes was obtained by modelling the core and its surroundings in three-dimensions by using the one dimensional transport theory code WIMS-D/4 and the multidimensional finite difference diffusion theory code CITATION. The input spectrum covariance information required by MSITER code was also calculated from the CITATION output. A comparison between calculated and adjusted spectra shows a good agreement

  18. 20070607 NATO Advanced Study Institute on the Electromagnetic Spectrum of Neutron Stars Marmaris, Turkey 07 - 18 Jun 2004 2004 marmaris20040607 TR 20040618

    CERN Document Server

    Baykal, Altan; Inam, Sitki C; Grebenev, Sergei

    2005-01-01

    Neutron stars hold a central place in astrophysics, not only because they are made up of the most extreme states of the condensed matter, but also because they are, along with white dwarfs and black holes, one of the stable configurations that stars reach at the end of stellar evolution. Neutron stars posses the highest rotation rates and strongest magnetic fields among all stars. They radiate prolifically, in high energy electromagnetic radiation and in the radio band. This book is devoted to the selected lectures presented in the 6th NATO-ASI series entitled "The Electromagnetic Spectrum of Neutron Stars" in Marmaris, Turkey, on 7-18 June 2004. This ASI is devoted to the spectral properties of neutron stars. Spectral observations of neutron stars help us to understand the magnetospheric emission processes of isolated radio pulsars and the emission processes of accreting neutron stars. This volume includes spectral information from the neutron stars in broadest sense, namely neutrino and gravitational radiat...

  19. Pilot experimental study on continual spectrum thermal neutron in-line phase contrast radiography

    International Nuclear Information System (INIS)

    Tang Bin; Huo Heyong; Wu Yang

    2009-01-01

    The in-line phase contrast radiography is one of phase contrast imaging methods. The neutron in-line phase contrast is developed with X-rays phase contrast radiography. In the paper, the principle of in-line phase contrast is introduced briefly and the experimental result of thermal neutron in-line contrast at SPRR-300 is analysed. It shows that thermal neutron can be used as in-line phase contrast radiography and enhances the edge of some sample in radiography and complements the disadvantage of conventional neutron radiography. (authors)

  20. An investigation on detection and measurement of fusion neutron spectrum and radiation flux in large tokamak

    International Nuclear Information System (INIS)

    Yang Jinwei; Li Wenzhong; Zhang Wei

    2003-01-01

    The detection methods, detectors and spectrometers of D-D and D-T fusion neutron have been overviewed in large tokamaks. Some options are proposed for developing new detection systems of fusion neutrons suitable to the HL-2A tokamak. (authors)

  1. Calibration of LiBaF sub 3 Ce scintillator for fission spectrum neutrons

    CERN Document Server

    Reeder, P L

    2002-01-01

    The scintillator LiBaF sub 3 doped with small amounts of Ce sup + sup 3 has the ability to distinguish heavy charged particles (p, d, t, or alpha) from beta and/or gamma radiation based on the presence or absence of nanosecond components in the scintillation light output. Since the neutron capture reaction on sup 6 Li produces recoil alphas and tritons, this scintillator also discriminates between neutron induced events and beta or gamma interactions. An experimental technique using a time-tagged sup 2 sup 5 sup 2 Cf source has been used to measure the efficiency of this scintillator for neutron capture, the calibration of neutron capture pulse height, and the pulse height resolution--all as a function of incident neutron energy.

  2. Evaluation of the 235U prompt fission neutron spectrum including a detailed analysis of experimental data and improved model information

    Science.gov (United States)

    Neudecker, Denise; Talou, Patrick; Kahler, Albert C.; White, Morgan C.; Kawano, Toshihiko

    2017-09-01

    We present an evaluation of the 235U prompt fission neutron spectrum (PFNS) induced by thermal to 20-MeV neutrons. Experimental data and associated covariances were analyzed in detail. The incident energy dependence of the PFNS was modeled with an extended Los Alamos model combined with the Hauser-Feshbach and the exciton models. These models describe prompt fission, pre-fission compound nucleus and pre-equilibrium neutron emissions. The evaluated PFNS agree well with the experimental data included in this evaluation, preliminary data of the LANL and LLNL Chi-Nu measurement and recent evaluations by Capote et al. and Rising et al. However, they are softer than the ENDF/B-VII.1 (VII.1) and JENDL-4.0 PFNS for incident neutron energies up to 2 MeV. Simulated effective multiplication factors keff of the Godiva and Flattop-25 critical assemblies are further from the measured keff if the current data are used within VII.1 compared to using only VII.1 data. However, if this work is used with ENDF/B-VIII.0β2 data, simulated values of keff agree well with the measured ones.

  3. On the spectrum of the one-speed slab-geometry discrete ordinates operator in neutron transport theory

    International Nuclear Information System (INIS)

    Abreu, Marcos Pimenta de

    1998-01-01

    We describe a numerical method applied to the first-order form of one-speed slab-geometry discrete ordinates equations modelling time-independent neutron transport problems with anisotropic scattering, with no interior source and defined in a nonmultiplying homogeneous host medium. Our numerical method is concerned with the generation of the spectrum and of a vector basis for the null space of the one-speed slab-geometry discrete ordinates operator. Moreover, it allows us to overcome the difficulties introduced in previous methods by anisotropic scattering and by angular quadrature sets of high order. To illustrate the positive features of our numerical method, we present numerical results for one-speed slab-geometry neutron transport model problems with anisotropic scattering

  4. Monte-Carlo study on primary knock-on atom energy spectrum produced by neutron radiation

    International Nuclear Information System (INIS)

    Zhou Wei; Liu Yongkang; Deng Yongjun; Ma Jimin

    2012-01-01

    Computational method on energy distribution of primary knock-on atom (PKA) produced by neutron radiation was built in the paper. Based on the DBCN card in MCNP, reaction position, reaction type and energy transfer between neutrons and atoms were recorded. According to statistic of these data, energy and space distributions of PKAs were obtained. The method resolves preferably randomicity of random number and efficiency of random sampling computation. The results show small statistical fluctuation and well statistical. Three-dimensional figure of energy and space distribution of PKAs were obtained, which would be important to evaluate radiation capability of materials and study radiation damage by neutrons. (authors)

  5. NSPEC - A neutron spectrum code for beam-heated fusion plasmas

    International Nuclear Information System (INIS)

    Scheffel, J.

    1983-06-01

    A 3-dimensional computer code is described, which computes neutron spectra due to beam heating of fusion plasmas. Three types of interactions are considered; thermonuclear of plasma-plasma, beam-plasma and beam-beam interactions. Beam deposition is modelled by the NFREYA code. The applied steady state beam distribution as a function of pitch angle and velocity contains the effects of energy diffusion, friction, angular scattering, charge exchange, electric field and source pitch angle distribution. The neutron spectra, generated by Monte-Carlo methods, are computed with respect to given lines of sight. This enables the code to be used for neutron diagnostics. (author)

  6. The Prompt Fission Neutron Spectrum: From Experiment to the Evaluated Data and its Impact on Critical Assemblies

    Energy Technology Data Exchange (ETDEWEB)

    Rising, Michael Evan [Los Alamos National Laboratory

    2015-06-10

    After a brief introduction concerning nuclear data, prompt fission neutron spectrum (PFNS) evaluations and the limited PFNS covariance data in the ENDF/B-VII library, and the important fact that cross section uncertainties ~ PFNS uncertainties, the author presents background information on the PFNS (experimental data, theoretical models, data evaluation, uncertainty quantification) and discusses the impact on certain well-known critical assemblies with regard to integral quantities, sensitivity analysis, and uncertainty propagation. He sketches recent and ongoing research and concludes with some final thoughts.

  7. The ration/gsub(A)/gsub(V) derived from the proton spectrum in free-neutron decay

    International Nuclear Information System (INIS)

    Stratowa, Ch.; Dobrozemsky, R.; Weinzierl, P.

    1978-08-01

    The electron-neutrino angular correlation coefficient was determined by measuring the shape of the proton recoil spectrum from free-neutron decay. The protons leaving a highly evacuated tangential reactor beam tube were analysed by a spherical condenser spectrometer and counted in an ion-electron converter detector. The design of the apparatus, the possible disturbing influences and the measures to reduce their effects are discussed. The remaining corrections were either calculated or determined by auxiliary measurements and applied to the spectral shape. The sources of systematic errors are considered and included in the final results. We obtained a- is equal to -0.1017+-0.0051 giving

  8. Spectrum measurements in the ZENITH plutonium core 7 using a neutron chopper

    Energy Technology Data Exchange (ETDEWEB)

    Barclay, F R; Cameron, I R; Pitcher, H H.W.; Symons, C R [General Reactor Physics Division, Atomic Energy Establishment, Winfrith, Dorchester, Dorset (United Kingdom)

    1964-05-15

    upwards through the core in the interstitial channels between the elements and then,after mixing with cooler nitrogen in a plenum above the core, flows down through channels in the side reflector. This heating system in principle permits independent variation of the core and side reflector temperatures. Although the plant is capable of operating at core temperatures of up to 800 deg. C, a restriction to 420 deg. C was imposed in the present experiments to avoid possible damage to the fuel. The theoretical treatment of the ZENITH plutonium cores relied on a calculation using the Winfrith DSN program to deal with the spatial and energy distribution of neutrons below 1.5 eV in the fuel cell. The comparison of the measured spectrum with that given by DSN provided a test of the method of calculation, and in particular of the graphite scattering model. The. detailed comparison with theory was complicated by the necessity of accounting adequately for the radial fine structure of the spectrum emergent from the source element, taking into consideration the spatial response function of the chopper and associated collimating system. Some consideration was given to the possibility of providing a source element of simpler geometry than the normal fuel element (e.g. a homogeneous cylindrical spine with fuel, absorber and moderator surrounded by an outer sleeve of graphite) but difficulty then arises in the specification of the fuel content of the spine region in order to match the absorption characteristics of the standard element. It was consequently decided to employ the normal heterogeneous source element and to use the theoretical model to predict the detailed structure of the emergent team. (author)

  9. Spectrum measurements in the ZENITH plutonium core 7 using a neutron chopper

    International Nuclear Information System (INIS)

    Barclay, F.R.; Cameron, I.R.; Pitcher, H.H.W.; Symons, C.R.

    1964-05-01

    upwards through the core in the interstitial channels between the elements and then,after mixing with cooler nitrogen in a plenum above the core, flows down through channels in the side reflector. This heating system in principle permits independent variation of the core and side reflector temperatures. Although the plant is capable of operating at core temperatures of up to 800 deg. C, a restriction to 420 deg. C was imposed in the present experiments to avoid possible damage to the fuel. The theoretical treatment of the ZENITH plutonium cores relied on a calculation using the Winfrith DSN program to deal with the spatial and energy distribution of neutrons below 1.5 eV in the fuel cell. The comparison of the measured spectrum with that given by DSN provided a test of the method of calculation, and in particular of the graphite scattering model. The. detailed comparison with theory was complicated by the necessity of accounting adequately for the radial fine structure of the spectrum emergent from the source element, taking into consideration the spatial response function of the chopper and associated collimating system. Some consideration was given to the possibility of providing a source element of simpler geometry than the normal fuel element (e.g. a homogeneous cylindrical spine with fuel, absorber and moderator surrounded by an outer sleeve of graphite) but difficulty then arises in the specification of the fuel content of the spine region in order to match the absorption characteristics of the standard element. It was consequently decided to employ the normal heterogeneous source element and to use the theoretical model to predict the detailed structure of the emergent team. (author)

  10. Evaluation of the neutrons spectrum near the Venus reactor: use of MCNPX-2.5C

    International Nuclear Information System (INIS)

    Verboomen, B.; Coeck, M.; Baeten, P.

    2003-01-01

    The present study has been justified by the choice of the Venus reactor (SCK-CEN) as a true work environment for the project of the fifth programme - frame E.V.I.D.O.S.. The objective of this programme is the evaluation, in neutron-photon combined field, and in true environment (nuclear industry), of the different methods of measurement used in neutron dosimetry. The project aims to the determination of abilities and limits of dosemeters and to establish methods to get doses equivalents from data gotten by spectrometry, personal and ambient dosimetry. For each environment, reference values have to be determined by spectrometry (energy and angle). The knowledge of the distribution in energy and in angle of neutrons allows then the calculation of the different doses equivalents. The determination of these references values by direct neutron calculation allows the validation of the Monte Carlo model. (N.C.)

  11. Neutron spectrum determination of d(20)+Be source reaction by the dosimetry foils method

    Czech Academy of Sciences Publication Activity Database

    Štefánik, Milan; Bém, Pavel; Majerle, Mitja; Novák, Jan; Šimečková, Eva

    2017-01-01

    Roč. 140, NOV (2017), s. 466-470 ISSN 0969-806X R&D Projects: GA MŠk LM2015056 Institutional support: RVO:61389005 Keywords : multi-foil activation technique * accelerator-based neutron source * neutron spectrometry * Gamma-ray spectrometry * reaction rate * charged particle accelerator Subject RIV: BG - Nuclear, Atomic and Molecular Physics, Colliders OBOR OECD: Nuclear physics Impact factor: 1.315, year: 2016

  12. Measurement of the energy spectrum from the neutron source p lanned for IGISOL

    CERN Document Server

    Mattera, A; Rakopoulos, V; Lantz, M; Pomp, S; Solders, A; Al-Adili, A; Andersson, P; Hjalmarsson, A; Valldor-Blücher, B; Prokofiev, A; Passoth, E; Gentile, A; Bortot, D; Esposito, A; Introini, M V; Pola, A; Penttilä, H; Gorelov, D; Rinta-Antila, S

    2014-01-01

    We report on the characterisation measurements of the energ y spectra from a Be (p,xn) neutron source to be installed at the IGISOL-JYFLTRA P facility for studies of neutron-induced independent fission yields. The measurements were performed at The Svedberg Laboratory (Uppsala, Sweden), during 50 hours of beam-time in June, 2012. A 30 MeV p roton beam impinged on a mock-up of the proton-neutron converter; this was a 5 mm-thick beryllium disc inserted in an aluminium holder, with a 1-cm t hick layer of cool- ing water on the backside. The geometry of the mock-up has bee n chosen to reproduce the one that will be used as the IGISOL-JYFLTRAP so urce. During the experiment, two configurations for the neutron so urce have been used: a fast neutron field, produced using the bare target; an d a moderated field, obtained adding a 10 cm-thick Polyethylene block after the t arget assembly. The neutron fields have been measured using an Extended Range Bonner Sphere Spectrometer (ERBSS), able to simultaneously determine ...

  13. Warhead verification as inverse problem: Applications of neutron spectrum unfolding from organic-scintillator measurements

    Science.gov (United States)

    Lawrence, Chris C.; Febbraro, Michael; Flaska, Marek; Pozzi, Sara A.; Becchetti, F. D.

    2016-08-01

    Verification of future warhead-dismantlement treaties will require detection of certain warhead attributes without the disclosure of sensitive design information, and this presents an unusual measurement challenge. Neutron spectroscopy—commonly eschewed as an ill-posed inverse problem—may hold special advantages for warhead verification by virtue of its insensitivity to certain neutron-source parameters like plutonium isotopics. In this article, we investigate the usefulness of unfolded neutron spectra obtained from organic-scintillator data for verifying a particular treaty-relevant warhead attribute: the presence of high-explosive and neutron-reflecting materials. Toward this end, several improvements on current unfolding capabilities are demonstrated: deuterated detectors are shown to have superior response-matrix condition to that of standard hydrogen-base scintintillators; a novel data-discretization scheme is proposed which removes important detector nonlinearities; and a technique is described for re-parameterizing the unfolding problem in order to constrain the parameter space of solutions sought, sidestepping the inverse problem altogether. These improvements are demonstrated with trial measurements and verified using accelerator-based time-of-flight calculation of reference spectra. Then, a demonstration is presented in which the elemental compositions of low-Z neutron-attenuating materials are estimated to within 10%. These techniques could have direct application in verifying the presence of high-explosive materials in a neutron-emitting test item, as well as other for treaty verification challenges.

  14. Measurement of fast neutron spectrum using CR-39 detectors and a new image analysis program (autoTRAKn)

    International Nuclear Information System (INIS)

    Paul, Sabyasachi; Tripathy, S.P.; Sahoo, G.S.; Bandyopadhyay, T.; Sarkar, P.K.

    2013-01-01

    An attempt is made to estimate the neutron spectrum using the CR-39 (Solid state nuclear track) detector and a new image analyzing program. For this purpose the earlier developed program (autoTRAK) is modified by introducing the required features such as angular correction for the recoil particles, fluence-to-dose conversion coefficient, detection sensitivity of CR-39 detectors, etc. to make it applicable for neutron spectrometry and dosimetry. This upgraded program (autoTRAK n ) is tested with a mono-energetic source (D–T) and two other standard neutron sources, viz. 241 Am–Be and 252 Cf. The program is validated by reproducing these standard spectra, and comparing with the spectra reported by other investigators using different measuring techniques. The ratios of dose equivalent (H ⁎ (10)) to fluence (Φ) are also estimated from the spectra and are compared with the reference values for these neutron sources. An additional feature of this program is explored for counting high density overlapping tracks more precisely and effectively compared to other commonly used image analyzing softwares. This method is found to be simple and promising, which can always be used as a supplementary measuring technique. The details of the modified program, reproduction and comparison of the neutron spectra, reproducibility of the methodology and example of overlapping track counting are presented and discussed. -- Highlights: •A novel image analysis technique (autoTRAK n ) is developed to evaluate CR-39 detectors used for neutron spectrometry and dosimetry. •The methodology is tested to reproduce three standard neutron spectra, (a) D–T, (b) 241 Am–Be, and (c) 252 Cf. •A good matching is observed between dosimetric values obtained by the program and the available reference values. •The program autoTRAK n is also observed to be efficient to distinguish high density overlapping tracks without any segregation procedure. •The methodology seems to be simple, which

  15. COMBINE7.0 - A Portable ENDF/B-VII.0 Based Neutron Spectrum and Cross-Section Generation Program

    Energy Technology Data Exchange (ETDEWEB)

    Woo Y. Yoon; David W. Nigg

    2008-09-01

    COMBINE7.0 is a FORTRAN 90 computer code that generates multigroup neutron constants for use in the deterministic diffusion and transport theory neutronics analysis. The cross-section database used by COMBINE7.0 is derived from the Evaluated Nuclear Data Files (ENDF/B-VII.0). The neutron energy range covered is from 20 MeV to 1.0E-5 eV. The Los Alamos National Laboratory NJOY code is used as the processing code to generate a 167 finegroup cross-section library in MATXS format for Bondarenko self-shielding treatment. Resolved resonance parameters are extracted from ENDF/B-VII.0 File 2 for a separate library to be used in an alternate Nordheim self-shielding treatment in the resolved resonance energy range. The equations solved for energy dependent neutron spectrum in the 167 fine-group structure are the B-3 or B-1 approximations to the transport equation. The fine group cross sections needed for the spectrum calculation are first prepared by Bondarenko selfshielding interpolation in terms of background cross section and temperature. The geometric lump effect, when present, is accounted for by augmenting the background cross section. Nordheim self-shielded fine group cross sections for a material having resolved resonance parameters overwrite correspondingly the existing self-shielded fine group cross sections when this option is used. The fine group cross sections in the thermal energy range are replaced by those selfshielded with the Amouyal/Benoist/Horowitz method in the three region geometry when this option is requested. COMBINE7.0 coalesces fine group cross sections into broad group macroscopic and microscopic constants. The coalescing is performed by utilizing fine-group fluxes and/or currents obtained by spectrum calculation as the weighting functions. The multigroup constant may be output in any of several standard formats including ANISN 14** free format, CCCC ISOTXS format, and AMPX working library format. ANISN-PC, a onedimensional, discrete

  16. COMBINE7.1 - A Portable ENDF/B-VII.0 Based Neutron Spectrum and Cross-Section Generation Program

    Energy Technology Data Exchange (ETDEWEB)

    Woo Y. Yoon; David W. Nigg

    2009-08-01

    COMBINE7.1 is a FORTRAN 90 computer code that generates multigroup neutron constants for use in the deterministic diffusion and transport theory neutronics analysis. The cross-section database used by COMBINE7.1 is derived from the Evaluated Nuclear Data Files (ENDF/B-VII.0). The neutron energy range covered is from 20 MeV to 1.0E-5 eV. The Los Alamos National Laboratory NJOY code is used as the processing code to generate a 167 fine-group cross-section library in MATXS format for Bondarenko self-shielding treatment. Resolved resonance parameters are extracted from ENDF/B-VII.0 File 2 for a separate library to be used in an alternate Nordheim self-shielding treatment in the resolved resonance energy range. The equations solved for energy dependent neutron spectrum in the 167 fine-group structure are the B-3 or B-1 approximations to the transport equation. The fine group cross sections needed for the spectrum calculation are first prepared by Bondarenko self-shielding interpolation in terms of background cross section and temperature. The geometric lump effect, when present, is accounted for by augmenting the background cross section. Nordheim self-shielded fine group cross sections for a material having resolved resonance parameters overwrite correspondingly the existing self-shielded fine group cross sections when this option is used. The fine group cross sections in the thermal energy range are replaced by those self-shielded with the Amouyal/Benoist/Horowitz method in the three region geometry when this option is requested. COMBINE7.1 coalesces fine group cross sections into broad group macroscopic and microscopic constants. The coalescing is performed by utilizing fine-group fluxes and/or currents obtained by spectrum calculation as the weighting functions. The multigroup constant may be output in any of several standard formats including ANISN 14** free format, CCCC ISOTXS format, and AMPX working library format. ANISN-PC, a one-dimensional, discrete

  17. COMBINE7.1 - A Portable ENDF/B-VII.0 Based Neutron Spectrum and Cross-Section Generation Program

    International Nuclear Information System (INIS)

    Yoon, Woo Y.; Nigg, David W.

    2009-01-01

    COMBINE7.1 is a FORTRAN 90 computer code that generates multigroup neutron constants for use in the deterministic diffusion and transport theory neutronics analysis. The cross-section database used by COMBINE7.1 is derived from the Evaluated Nuclear Data Files (ENDF/B-VII.0). The neutron energy range covered is from 20 MeV to 1.0E-5 eV. The Los Alamos National Laboratory NJOY code is used as the processing code to generate a 167 fine-group cross-section library in MATXS format for Bondarenko self-shielding treatment. Resolved resonance parameters are extracted from ENDF/B-VII.0 File 2 for a separate library to be used in an alternate Nordheim self-shielding treatment in the resolved resonance energy range. The equations solved for energy dependent neutron spectrum in the 167 fine-group structure are the B-3 or B-1 approximations to the transport equation. The fine group cross sections needed for the spectrum calculation are first prepared by Bondarenko self-shielding interpolation in terms of background cross section and temperature. The geometric lump effect, when present, is accounted for by augmenting the background cross section. Nordheim self-shielded fine group cross sections for a material having resolved resonance parameters overwrite correspondingly the existing self-shielded fine group cross sections when this option is used. The fine group cross sections in the thermal energy range are replaced by those self-shielded with the Amouyal/Benoist/Horowitz method in the three region geometry when this option is requested. COMBINE7.1 coalesces fine group cross sections into broad group macroscopic and microscopic constants. The coalescing is performed by utilizing fine-group fluxes and/or currents obtained by spectrum calculation as the weighting functions. The multigroup constant may be output in any of several standard formats including ANISN 14** free format, CCCC ISOTXS format, and AMPX working library format. ANISN-PC, a one-dimensional, discrete

  18. COMBINE7.0 - A Portable ENDF/B-VII.0 Based Neutron Spectrum and Cross-Section Generation Program

    International Nuclear Information System (INIS)

    Yoon, Woo Y.; Nigg, David W.

    2008-01-01

    COMBINE7.0 is a FORTRAN 90 computer code that generates multigroup neutron constants for use in the deterministic diffusion and transport theory neutronics analysis. The cross-section database used by COMBINE7.0 is derived from the Evaluated Nuclear Data Files (ENDF/B-VII.0). The neutron energy range covered is from 20 MeV to 1.0E-5 eV. The Los Alamos National Laboratory NJOY code is used as the processing code to generate a 167 finegroup cross-section library in MATXS format for Bondarenko self-shielding treatment. Resolved resonance parameters are extracted from ENDF/B-VII.0 File 2 for a separate library to be used in an alternate Nordheim self-shielding treatment in the resolved resonance energy range. The equations solved for energy dependent neutron spectrum in the 167 fine-group structure are the B-3 or B-1 approximations to the transport equation. The fine group cross sections needed for the spectrum calculation are first prepared by Bondarenko selfshielding interpolation in terms of background cross section and temperature. The geometric lump effect, when present, is accounted for by augmenting the background cross section. Nordheim self-shielded fine group cross sections for a material having resolved resonance parameters overwrite correspondingly the existing self-shielded fine group cross sections when this option is used. The fine group cross sections in the thermal energy range are replaced by those selfshielded with the Amouyal/Benoist/Horowitz method in the three region geometry when this option is requested. COMBINE7.0 coalesces fine group cross sections into broad group macroscopic and microscopic constants. The coalescing is performed by utilizing fine-group fluxes and/or currents obtained by spectrum calculation as the weighting functions. The multigroup constant may be output in any of several standard formats including ANISN 14** free format, CCCC ISOTXS format, and AMPX working library format. ANISN-PC, a onedimensional, discrete

  19. NSDann2BS, a neutron spectrum unfolding code based on neural networks technology and two bonner spheres

    Science.gov (United States)

    Ortiz-Rodríguez, J. M.; Reyes Alfaro, A.; Reyes Haro, A.; Solís Sánches, L. O.; Miranda, R. Castañeda; Cervantes Viramontes, J. M.; Vega-Carrillo, H. R.

    2013-07-01

    In this work a neutron spectrum unfolding code, based on artificial intelligence technology is presented. The code called "Neutron Spectrometry and Dosimetry with Artificial Neural Networks and two Bonner spheres", (NSDann2BS), was designed in a graphical user interface under the LabVIEW programming environment. The main features of this code are to use an embedded artificial neural network architecture optimized with the "Robust design of artificial neural networks methodology" and to use two Bonner spheres as the only piece of information. In order to build the code here presented, once the net topology was optimized and properly trained, knowledge stored at synaptic weights was extracted and using a graphical framework build on the LabVIEW programming environment, the NSDann2BS code was designed. This code is friendly, intuitive and easy to use for the end user. The code is freely available upon request to authors. To demonstrate the use of the neural net embedded in the NSDann2BS code, the rate counts of 252Cf, 241AmBe and 239PuBe neutron sources measured with a Bonner spheres system.

  20. NSDann2BS, a neutron spectrum unfolding code based on neural networks technology and two bonner spheres

    International Nuclear Information System (INIS)

    Ortiz-Rodríguez, J. M.; Reyes Alfaro, A.; Reyes Haro, A.; Solís Sánches, L. O.; Miranda, R. Castañeda; Cervantes Viramontes, J. M.; Vega-Carrillo, H. R.

    2013-01-01

    In this work a neutron spectrum unfolding code, based on artificial intelligence technology is presented. The code called ''Neutron Spectrometry and Dosimetry with Artificial Neural Networks and two Bonner spheres'', (NSDann2BS), was designed in a graphical user interface under the LabVIEW programming environment. The main features of this code are to use an embedded artificial neural network architecture optimized with the ''Robust design of artificial neural networks methodology'' and to use two Bonner spheres as the only piece of information. In order to build the code here presented, once the net topology was optimized and properly trained, knowledge stored at synaptic weights was extracted and using a graphical framework build on the LabVIEW programming environment, the NSDann2BS code was designed. This code is friendly, intuitive and easy to use for the end user. The code is freely available upon request to authors. To demonstrate the use of the neural net embedded in the NSDann2BS code, the rate counts of 252 Cf, 241 AmBe and 239 PuBe neutron sources measured with a Bonner spheres system

  1. Design and spectrum calculation of 4H-SiC thermal neutron detectors using FLUKA and TCAD

    Science.gov (United States)

    Huang, Haili; Tang, Xiaoyan; Guo, Hui; Zhang, Yimen; Zhang, Yimeng; Zhang, Yuming

    2016-10-01

    SiC is a promising material for neutron detection in a harsh environment due to its wide band gap, high displacement threshold energy and high thermal conductivity. To increase the detection efficiency of SiC, a converter such as 6LiF or 10B is introduced. In this paper, pulse-height spectra of a PIN diode with a 6LiF conversion layer exposed to thermal neutrons (0.026 eV) are calculated using TCAD and Monte Carlo simulations. First, the conversion efficiency of a thermal neutron with respect to the thickness of 6LiF was calculated by using a FLUKA code, and a maximal efficiency of approximately 5% was achieved. Next, the energy distributions of both 3H and α induced by the 6LiF reaction according to different ranges of emission angle are analyzed. Subsequently, transient pulses generated by the bombardment of single 3H or α-particles are calculated. Finally, pulse height spectra are obtained with a detector efficiency of 4.53%. Comparisons of the simulated result with the experimental data are also presented, and the calculated spectrum shows an acceptable similarity to the experimental data. This work would be useful for radiation-sensing applications, especially for SiC detector design.

  2. NSDann2BS, a neutron spectrum unfolding code based on neural networks technology and two bonner spheres

    Energy Technology Data Exchange (ETDEWEB)

    Ortiz-Rodriguez, J. M.; Reyes Alfaro, A.; Reyes Haro, A.; Solis Sanches, L. O.; Miranda, R. Castaneda; Cervantes Viramontes, J. M. [Universidad Autonoma de Zacatecas, Unidad Academica de Ingenieria Electrica. Av. Ramon Lopez Velarde 801. Col. Centro Zacatecas, Zac (Mexico); Vega-Carrillo, H. R. [Universidad Autonoma de Zacatecas, Unidad Academica de Ingenieria Electrica. Av. Ramon Lopez Velarde 801. Col. Centro Zacatecas, Zac., Mexico. and Unidad Academica de Estudios Nucleares. C. Cip (Mexico)

    2013-07-03

    In this work a neutron spectrum unfolding code, based on artificial intelligence technology is presented. The code called ''Neutron Spectrometry and Dosimetry with Artificial Neural Networks and two Bonner spheres'', (NSDann2BS), was designed in a graphical user interface under the LabVIEW programming environment. The main features of this code are to use an embedded artificial neural network architecture optimized with the ''Robust design of artificial neural networks methodology'' and to use two Bonner spheres as the only piece of information. In order to build the code here presented, once the net topology was optimized and properly trained, knowledge stored at synaptic weights was extracted and using a graphical framework build on the LabVIEW programming environment, the NSDann2BS code was designed. This code is friendly, intuitive and easy to use for the end user. The code is freely available upon request to authors. To demonstrate the use of the neural net embedded in the NSDann2BS code, the rate counts of {sup 252}Cf, {sup 241}AmBe and {sup 239}PuBe neutron sources measured with a Bonner spheres system.

  3. A neutron spectrum unfolding code based on generalized regression artificial neural networks

    International Nuclear Information System (INIS)

    Ortiz R, J. M.; Martinez B, M. R.; Castaneda M, R.; Solis S, L. O.; Vega C, H. R.

    2015-10-01

    The most delicate part of neutron spectrometry, is the unfolding process. Then derivation of the spectral information is not simple because the unknown is not given directly as result of the measurements. Novel methods based on Artificial Neural Networks have been widely investigated. In prior works, back propagation neural networks (BPNN) have been used to solve the neutron spectrometry problem, however, some drawbacks still exist using this kind of neural nets, as the optimum selection of the network topology and the long training time. Compared to BPNN, is usually much faster to train a generalized regression neural network (GRNN). That is mainly because spread constant is the only parameter used in GRNN. Another feature is that the network will converge to a global minimum. In addition, often are more accurate than BPNN in prediction. These characteristics make GRNN be of great interest in the neutron spectrometry domain. In this work is presented a computational tool based on GRNN, capable to solve the neutron spectrometry problem. This computational code, automates the pre-processing, training and testing stages, the statistical analysis and the post-processing of the information, using 7 Bonner spheres rate counts as only entrance data. The code was designed for a Bonner Spheres System based on a 6 LiI(Eu) neutron detector and a response matrix expressed in 60 energy bins taken from an International Atomic Energy Agency compilation. (Author)

  4. A neutron spectrum unfolding code based on generalized regression artificial neural networks

    Energy Technology Data Exchange (ETDEWEB)

    Ortiz R, J. M.; Martinez B, M. R.; Castaneda M, R.; Solis S, L. O. [Universidad Autonoma de Zacatecas, Unidad Academica de Ingenieria Electrica, Av. Ramon Lopez Velarde 801, Col. Centro, 98000 Zacatecas, Zac. (Mexico); Vega C, H. R., E-mail: morvymm@yahoo.com.mx [Universidad Autonoma de Zacatecas, Unidad Academica de Estudios Nucleares, Cipres No. 10, Fracc. La Penuela, 98068 Zacatecas, Zac. (Mexico)

    2015-10-15

    The most delicate part of neutron spectrometry, is the unfolding process. Then derivation of the spectral information is not simple because the unknown is not given directly as result of the measurements. Novel methods based on Artificial Neural Networks have been widely investigated. In prior works, back propagation neural networks (BPNN) have been used to solve the neutron spectrometry problem, however, some drawbacks still exist using this kind of neural nets, as the optimum selection of the network topology and the long training time. Compared to BPNN, is usually much faster to train a generalized regression neural network (GRNN). That is mainly because spread constant is the only parameter used in GRNN. Another feature is that the network will converge to a global minimum. In addition, often are more accurate than BPNN in prediction. These characteristics make GRNN be of great interest in the neutron spectrometry domain. In this work is presented a computational tool based on GRNN, capable to solve the neutron spectrometry problem. This computational code, automates the pre-processing, training and testing stages, the statistical analysis and the post-processing of the information, using 7 Bonner spheres rate counts as only entrance data. The code was designed for a Bonner Spheres System based on a {sup 6}LiI(Eu) neutron detector and a response matrix expressed in 60 energy bins taken from an International Atomic Energy Agency compilation. (Author)

  5. Use of boron nitride for neutron spectrum characterization and cross-section validation in the epithermal range through integral activation measurements

    Science.gov (United States)

    Radulović, Vladimir; Trkov, Andrej; Jaćimović, Radojko; Gregoire, Gilles; Destouches, Christophe

    2016-12-01

    A recent experimental irradiation and measurement campaign using containers made from boron nitride (BN) at the Jožef Stefan Institute (JSI) TRIGA Mark II reactor in Ljubljana, Slovenia, has shown the applicability of BN for neutron spectrum characterization and cross-section validation in the epithermal range through integral activation measurements. The first part of the paper focuses on the determination of the transmission function of a BN container through Monte Carlo calculations and experimental measurements. The second part presents the process of tayloring the sensitivity of integral activation measurements to specific needs and a selection of suitable radiative capture reactions for neutron spectrum characterization in the epithermal range. A BN container used in our experiments and its qualitative effect on the neutron spectrum in the irradiation position employed is displayed in the Graphical abstract.

  6. The perturbation of backscattered fast neutrons spectrum caused by the resonances of C, N and O for possible use in pyromaterial detection

    Energy Technology Data Exchange (ETDEWEB)

    Abedin, Ahmad Firdaus Zainal, E-mail: firdaus087@gmail.com; Ibrahim, Noorddin; Zabidi, Noriza Ahmad; Abdullah, Abqari Luthfi Albert [Department of Defence Science, Universiti Pertahanan Nasional Malaysia, Kem Sungai Besi, Kuala Lumpur 57000 (Malaysia)

    2015-04-29

    Neutron radiation is able to determine the signature of land mine detection based on backscattering energy spectrum of landmine. In this study, the Monte Carlo simulation of backscattered fast neutrons was performed on four basic elements of land mine; hydrogen, nitrogen, oxygen and carbon. The moderation of fast neutrons to thermal neutrons and their resonances cross-section between 0.01 eV until 14 MeV were analysed. The neutrons energies were divided into 29 groups and ten million neutrons particles histories were used. The geometries consist of four main components: neutrons source, detectors, landmine and soil. The neutrons source was placed at the origin coordinate and shielded with carbon and polyethylene. Americium/Beryllium neutron source was placed inside lead casing of 1 cm thick and 2.5 cm height. Polyethylene was used to absorb and disperse radiation and was placed outside the lead shield of width 10 cm and height 7 cm. Two detectors were placed between source with distance of 8 cm and radius of 1.9 cm. Detectors of Helium-3 was used for neutron detection as it has high absorption cross section for thermal neutrons. For the anomaly, the physical is in cylinder form with radius of 10 cm and 8.9 cm height. The anomaly is buried 5 cm deep in the bed soil measured 80 cm radius and 53.5 cm height. The results show that the energy spectrum for the four basic elements of landmine with specific pattern which can be used as indication for the presence of landmines.

  7. The perturbation of backscattered fast neutrons spectrum caused by the resonances of C, N and O for possible use in pyromaterial detection

    International Nuclear Information System (INIS)

    Abedin, Ahmad Firdaus Zainal; Ibrahim, Noorddin; Zabidi, Noriza Ahmad; Abdullah, Abqari Luthfi Albert

    2015-01-01

    Neutron radiation is able to determine the signature of land mine detection based on backscattering energy spectrum of landmine. In this study, the Monte Carlo simulation of backscattered fast neutrons was performed on four basic elements of land mine; hydrogen, nitrogen, oxygen and carbon. The moderation of fast neutrons to thermal neutrons and their resonances cross-section between 0.01 eV until 14 MeV were analysed. The neutrons energies were divided into 29 groups and ten million neutrons particles histories were used. The geometries consist of four main components: neutrons source, detectors, landmine and soil. The neutrons source was placed at the origin coordinate and shielded with carbon and polyethylene. Americium/Beryllium neutron source was placed inside lead casing of 1 cm thick and 2.5 cm height. Polyethylene was used to absorb and disperse radiation and was placed outside the lead shield of width 10 cm and height 7 cm. Two detectors were placed between source with distance of 8 cm and radius of 1.9 cm. Detectors of Helium-3 was used for neutron detection as it has high absorption cross section for thermal neutrons. For the anomaly, the physical is in cylinder form with radius of 10 cm and 8.9 cm height. The anomaly is buried 5 cm deep in the bed soil measured 80 cm radius and 53.5 cm height. The results show that the energy spectrum for the four basic elements of landmine with specific pattern which can be used as indication for the presence of landmines

  8. A neutron spectrum unfolding code based on generalized regression artificial neural networks

    International Nuclear Information System (INIS)

    Rosario Martinez-Blanco, Ma. del

    2016-01-01

    The most delicate part of neutron spectrometry, is the unfolding process. The derivation of the spectral information is not simple because the unknown is not given directly as a result of the measurements. Novel methods based on Artificial Neural Networks have been widely investigated. In prior works, back propagation neural networks (BPNN) have been used to solve the neutron spectrometry problem, however, some drawbacks still exist using this kind of neural nets, i.e. the optimum selection of the network topology and the long training time. Compared to BPNN, it's usually much faster to train a generalized regression neural network (GRNN). That's mainly because spread constant is the only parameter used in GRNN. Another feature is that the network will converge to a global minimum, provided that the optimal values of spread has been determined and that the dataset adequately represents the problem space. In addition, GRNN are often more accurate than BPNN in the prediction. These characteristics make GRNNs to be of great interest in the neutron spectrometry domain. This work presents a computational tool based on GRNN capable to solve the neutron spectrometry problem. This computational code, automates the pre-processing, training and testing stages using a k-fold cross validation of 3 folds, the statistical analysis and the post-processing of the information, using 7 Bonner spheres rate counts as only entrance data. The code was designed for a Bonner Spheres System based on a "6LiI(Eu) neutron detector and a response matrix expressed in 60 energy bins taken from an International Atomic Energy Agency compilation. - Highlights: • Main drawback of neutron spectrometry with BPNN is network topology optimization. • Compared to BPNN, it’s usually much faster to train a (GRNN). • GRNN are often more accurate than BPNN in the prediction. These characteristics make GRNNs to be of great interest. • This computational code, automates the pre-processing, training

  9. Effect of new cross-section evaluations on criticality and neutron energy spectrum of a typical material test research reactor

    International Nuclear Information System (INIS)

    Ahmad, Siraj-ul-Islam; Ahmad, Nasir; Aslam

    2004-01-01

    Several new WIMSD libraries based on recent cross-section evaluations such as IAEA, ENDFB-VI, JENDL, and JEF have been made available by IAEA. These libraries were used for the computation of multiplication factor and energy spectrum for Pakistan Research Reactor-1 (PARR-1). Methodology was validated for benchmark problems made available by IAEA and comparison with reference results. The value of effective multiplication factors for all newly released libraries are 1.8-3.2% less than that of 1981 WIMSD library. The effect of various cross-section libraries on neutron energy spectrum was also studied. Differences of about -10% to 12.5% were found in thermal flux using the newly released libraries as compared with that obtained using 1981 WIMSD library. From the analysis, it was found that the main source of the difference is the cross-sections of hydrogen bound in water. When these cross-sections of hydrogen (bound in water) from new libraries were used along with all other data in 1981 WIMSD library, the k eff obtained in this way has a difference of only 0.02-0.8% with that obtained from new libraries, while the flux spectrum agreed within 1% below 1 MeV with new libraries

  10. Measurement of the proton recoil spectrum in neutron beta decay with the spectrometer aSPECT. Study of systematic effects

    Energy Technology Data Exchange (ETDEWEB)

    Konrad, Gertrud Emilie

    2012-01-24

    Free neutron decay, n{yields}pe anti {nu}{sub e}, is the simplest nuclear beta decay, well described as a purely left-handed, vector minus axial-vector interaction within the framework of the Standard Model (SM) of elementary particles and fields. Due to its highly precise theoretical description, neutron beta decay data can be used to test certain extensions to the SM. Possible extensions require, e.g., new symmetry concepts like left-right symmetry, new particles, leptoquarks, supersymmetry, or the like. Precision measurements of observables in neutron beta decay address important open questions of particle physics and cosmology, and are generally complementary to direct searches for new physics beyond the SM in high-energy physics. In this doctoral thesis, a measurement of the proton recoil spectrum with the neutron decay spectrometer aSPECT is described. From the proton spectrum the antineutrinoelectron angular correlation coefficient a can be derived. In our first beam time at the Forschungs-Neutronenquelle Heinz Maier-Leibnitz in Munich, Germany (2005-2006), background instabilities due to particle trapping and the electronic noise level of the proton detector prevented us from presenting a new value for a. In the latest beam time at the Institut Laue-Langevin (ILL) in Grenoble, France (2007-2008), the trapped particle background has been reduced sufficiently and the electronic noise problem has essentially been solved. For the first time, a silicon drift detector was used. As a result of the data analysis, we identified and fixed a problem in the detector electronics which caused a significant systematic error. The target figure of the latest beam time was a new value for a with a total relative error well below the present literature value of 4 %. A statistical accuracy of about 1.4% was reached, but we could only set upper limits on the correction of the problem in the detector electronics, which are too high to determine a meaningful result. The present

  11. Measurement of the proton recoil spectrum in neutron beta decay with the spectrometer aSPECT. Study of systematic effects

    International Nuclear Information System (INIS)

    Konrad, Gertrud Emilie

    2012-01-01

    Free neutron decay, n→pe anti ν e , is the simplest nuclear beta decay, well described as a purely left-handed, vector minus axial-vector interaction within the framework of the Standard Model (SM) of elementary particles and fields. Due to its highly precise theoretical description, neutron beta decay data can be used to test certain extensions to the SM. Possible extensions require, e.g., new symmetry concepts like left-right symmetry, new particles, leptoquarks, supersymmetry, or the like. Precision measurements of observables in neutron beta decay address important open questions of particle physics and cosmology, and are generally complementary to direct searches for new physics beyond the SM in high-energy physics. In this doctoral thesis, a measurement of the proton recoil spectrum with the neutron decay spectrometer aSPECT is described. From the proton spectrum the antineutrinoelectron angular correlation coefficient a can be derived. In our first beam time at the Forschungs-Neutronenquelle Heinz Maier-Leibnitz in Munich, Germany (2005-2006), background instabilities due to particle trapping and the electronic noise level of the proton detector prevented us from presenting a new value for a. In the latest beam time at the Institut Laue-Langevin (ILL) in Grenoble, France (2007-2008), the trapped particle background has been reduced sufficiently and the electronic noise problem has essentially been solved. For the first time, a silicon drift detector was used. As a result of the data analysis, we identified and fixed a problem in the detector electronics which caused a significant systematic error. The target figure of the latest beam time was a new value for a with a total relative error well below the present literature value of 4 %. A statistical accuracy of about 1.4% was reached, but we could only set upper limits on the correction of the problem in the detector electronics, which are too high to determine a meaningful result. The present doctoral

  12. Neutron Therapy Facility

    Data.gov (United States)

    Federal Laboratory Consortium — The Neutron Therapy Facility provides a moderate intensity, broad energy spectrum neutron beam that can be used for short term irradiations for radiobiology (cells)...

  13. COMBINE/PC - a portable neutron spectrum and cross-section generation program

    International Nuclear Information System (INIS)

    Nigg, D.W.; Grimesey, R.A.; Curtis, R.L.

    1990-01-01

    Use of personal computers and engineering workstations for complex scientific computations has expanded rapidly in the past few years. This trend is expected to continue in the future with the introduction of increasingly sophisticated microprocessors and microcomputer systems. In response to this, an integrated system of neutronics and radiation transport software suitable for operation in an IBM personal computer (PC)-class environment has been under development at the Idaho National Engineering Laboratory (INEL) for the past 3 years. A key component of this system will be module to produce application-specific multigroup cross-section libraries that can be used in various neutron transport and diffusion theory code modules. This software module, referred to as COMBINE/PC, was recently completed at INEL and is the subject of this paper. COMBINE/PC was developed to provide an ENDF/B-based neutron cross-section generation capability of sufficient sophistication to handle a wide variety of practical fission and fusion-related applications while maintaining a compact machine-independent structure

  14. Evaluation of the {sup 239}Pu prompt fission neutron spectrum induced by neutrons of 500 keV and associated covariances

    Energy Technology Data Exchange (ETDEWEB)

    Neudecker, D., E-mail: dneudecker@lanl.gov [Theoretical Division, Los Alamos National Laboratory, P.O. Box 1663, MS-B283, NM 87545 (United States); Talou, P., E-mail: talou@lanl.gov [Theoretical Division, Los Alamos National Laboratory, P.O. Box 1663, MS-B283, NM 87545 (United States); Kawano, T., E-mail: kawano@lanl.gov [Theoretical Division, Los Alamos National Laboratory, P.O. Box 1663, MS-B283, NM 87545 (United States); Smith, D.L., E-mail: donaldlarnedsmith@gmail.com [Nuclear Engineering Division, Argonne National Laboratory, 1710 Avenida del Mundo #1506, Coronado, CA 92118 (United States); Capote, R., E-mail: r.capotenoy@iaea.org [Nuclear Data Section, International Atomic Energy Agency Vienna, Vienna International Centre, P.O. Box 100, A-1400 Vienna (Austria); Rising, M.E., E-mail: mrising@lanl.gov [X-Division, Los Alamos National Laboratory, P.O. Box 1663, MS-F663, NM 87545 (United States); Kahler, A.C., E-mail: akahler@lanl.gov [Theoretical Division, Los Alamos National Laboratory, P.O. Box 1663, MS-B283, NM 87545 (United States)

    2015-08-11

    We present evaluations of the prompt fission neutron spectrum (PFNS) of {sup 239}Pu induced by 500 keV neutrons, and associated covariances. In a previous evaluation by Talou et al. (2010), surprisingly low evaluated uncertainties were obtained, partly due to simplifying assumptions in the quantification of uncertainties from experiment and model. Therefore, special emphasis is placed here on a thorough uncertainty quantification of experimental data and of the Los Alamos model predicted values entering the evaluation. In addition, the Los Alamos model was extended and an evaluation technique was employed that takes into account the qualitative differences between normalized model predicted values and experimental shape data. These improvements lead to changes in the evaluated PFNS and overall larger evaluated uncertainties than in the previous work. However, these evaluated uncertainties are still smaller than those obtained in a statistical analysis using experimental information only, due to strong model correlations. Hence, suggestions to estimate model defect uncertainties are presented, which lead to more reasonable evaluated uncertainties. The calculated k{sub eff} of selected criticality benchmarks obtained with these new evaluations agree with each other within their uncertainties despite the different approaches to estimate model defect uncertainties. The k{sub eff} one standard deviations overlap with some of those obtained using ENDF/B-VII.1, albeit their mean values are further away from unity. Spectral indexes for the Jezebel critical assembly calculated with the newly evaluated PFNS agree with the experimental data for selected (n,γ) and (n,f) reactions, and show improvements for high-energy threshold (n,2n) reactions compared to ENDF/B-VII.1.

  15. Relative cataractogenic effects of X rays, fission-spectrum neutrons, and 56Fe particles: A comparison with mitotic effects

    International Nuclear Information System (INIS)

    Riley, E.F.; Lindgren, A.L.; Andersen, A.L.; Miller, R.C.; Ainsworth, E.J.

    1991-01-01

    The eyes of Sprague-Dawley rats were irradiated with doses of 2.5-10 Gy 250-kVp X rays, 1.25-2.25 Gy fission-spectrum neutrons (approximately 0.85 MeV), or 0.1-2.0 Gy 600-MeV/A 56Fe particles. Lens opacifications were evaluated for 51-61 weeks following X and neutron irradiations and for 87 weeks following X and 56Fe-particle irradiations. Average stage of opacification was determined relative to time after irradiation, and the time required for 50% of the irradiated lenses to achieve various stages (T50) was determined as a function of radiation dose. Data from two experiments were combined in dose-effect curves as T50 experimental values taken as percentages of the respective T50 control values (T50-% control). Simple exponential curves best describe dose responsiveness for both high-LET radiations. For X rays, a shallow dose-effect relationship (shoulder) up to 4.5 Gy was followed at higher doses by a steeper exponential dose-effect relationship. As a consequence, RBE values for the high-LET radiations are dose dependent. Dose-effect curves for cataracts were compared to those for mitotic abnormalities observed when quiescent lens epithelial cells were stimulated mechanically to proliferate at various intervals after irradiation. Neutrons were about 1.6-1.8 times more effective than 56Fe particles for inducing both cataracts and mitotic abnormalities. For stage 1 and 2 cataracts, the X-ray Dq was 10-fold greater and the D0 was similar to those for mitotic abnormalities initially expressed after irradiation

  16. The Binary Neutron Star Event LIGO/Virgo GW170817 160 Days after Merger: Synchrotron Emission across the Electromagnetic Spectrum

    Science.gov (United States)

    Margutti, R.; Alexander, K. D.; Xie, X.; Sironi, L.; Metzger, B. D.; Kathirgamaraju, A.; Fong, W.; Blanchard, P. K.; Berger, E.; MacFadyen, A.; Giannios, D.; Guidorzi, C.; Hajela, A.; Chornock, R.; Cowperthwaite, P. S.; Eftekhari, T.; Nicholl, M.; Villar, V. A.; Williams, P. K. G.; Zrake, J.

    2018-03-01

    We report deep Chandra X-ray Observatory (CXO), Hubble Space Telescope (HST), and Karl J. Jansky Very Large Array (VLA) observations of the binary neutron star event GW170817 at t spectrum extending for eight orders of magnitude in frequency enables the most precise measurement of the index p of the distribution of non-thermal relativistic electrons N(γ )\\propto {γ }-p accelerated by a shock launched by a neutron star (NS)–NS merger to date. We find p = 2.17 ± 0.01, which indicates that radiation from ejecta with Γ ∼ 3–10 dominates the observed emission. While constraining the nature of the emission process, these observations do not constrain the nature of the relativistic ejecta. We employ simulations of explosive outflows launched in NS ejecta clouds to show that the spectral and temporal evolution of the non-thermal emission from GW170817 is consistent with both emission from radially stratified quasi-spherical ejecta traveling at mildly relativistic speeds, and emission from off-axis collimated ejecta characterized by a narrow cone of ultra-relativistic material with slower wings extending to larger angles. In the latter scenario, GW170817 harbored a normal short gamma-ray burst (SGRB) directed away from our line of sight. Observations at t ≤ 200 days are unlikely to settle the debate, as in both scenarios the observed emission is effectively dominated by radiation from mildly relativistic material.

  17. Impact of statistical uncertainty of the neutron spectrum in the isotopic evolution of fuel; Impacto de la incertidumbre estadistica del espectro neutronico en la evoluacion isotopica del combustible

    Energy Technology Data Exchange (ETDEWEB)

    Ortega, P.

    2012-07-01

    The results obtained and presented in this study for different calculation conditions (number of stories, number of steps burning, etc.) and their simultaneous impact on neutron spectrum and isotopic composition and a methodology is proposed to determine the minimum parameters for calculation given uncertainty in the results of isotopic composition with high burnup, both UO{sub 2} and MOX fuel.

  18. OBJECT KINETIC MONTE CARLO SIMULATIONS OF RADIATION DAMAGE IN TUNGSTEN SUBJECTED TO NEUTRON FLUX WITH PKA SPECTRUM CORRESPONDING TO THE HFIR

    Energy Technology Data Exchange (ETDEWEB)

    Nandipati, Giridhar; Setyawan, Wahyu; Heinisch, Howard L.; Roche, Kenneth J.; Kurtz, Richard J.; Wirth, Brian D.

    2015-12-31

    The objective of this work is to study the damage accumulation in pure tungsten (W) subjected to neutron bombardment with a primary knock-on atom (PKA) spectrum corresponding to the High Flux Isotope Reactor (HFIR), using the object kinetic Monte Carlo (OKMC) method.

  19. Determination of the energy spectrum of the neutrons in the central thimble of the reactor core TRIGA Mark III; Determinacion del espectro de energia de los neutrones en el dedal central del nucleo del reactor Triga Mark III

    Energy Technology Data Exchange (ETDEWEB)

    Parra M, M. A.

    2014-07-01

    This thesis presents the neutron spectrum measurements inside the core of the TRIGA Mark III reactor at 1 MW power in steady-state, with the bridge placed in the center of the swimming pool, using several metallic threshold foils. The activation detectors are inserted in the Central Thimble of the reactor core, all the foils are irradiated in the same position and irradiation conditions (one by one). The threshold detectors are made of different materials such as: Au{sup 197}, Ni{sup 58}, In{sup 115}, Mg{sup 24}, Al{sup 27}, Fe{sup 58}, Co{sup 59} and Cu{sup 63}, they were selected to cover the full range the energies (10{sup -10} to 20 MeV) of the neutron spectrum in the reactor core. After the irradiation, the activation detectors were measured by means of spectrometry gamma, using a high resolution counting system with a hyper pure Germanium crystal, in order to obtain the saturation activity per target nuclide. The saturation activity is one of the main input data together with the initial spectrum, for the computational code SANDBP (hungarian version of the code SAND-II), which through an iterative adjustment, gives the calculated spectrum. The different saturation activities are necessary for the unfolding method, used by the computational code SANDBP. This research work is very important, since the knowledge of the energetic and spatial distribution of the neutron flux in the irradiation facilities, allows to characterize properly the irradiation facilities, just like, to estimate with a good precision various physics parameters of the reactor such as: neutron fluxes (thermal, intermediate and fast), neutronic dose, neutron activation analysis (NAA), spectral indices (cadmium ratio), buckling, fuel burnup, safety parameters (reactivity, temperature distribution, peak factors). In addition, the knowledge of the already mentioned parameters can give a best use of reactor, optimizing the irradiations requested by the users for their production process or

  20. Performance analysis of fusion nuclear-data benchmark experiments for light to heavy materials in MeV energy region with a neutron spectrum shifter

    International Nuclear Information System (INIS)

    Murata, Isao; Ohta, Masayuki; Miyamaru, Hiroyuki; Kondo, Keitaro; Yoshida, Shigeo; Iida, Toshiyuki; Ochiai, Kentaro; Konno, Chikara

    2011-01-01

    Nuclear data are indispensable for development of fusion reactor candidate materials. However, benchmarking of the nuclear data in MeV energy region is not yet adequate. In the present study, benchmark performance in the MeV energy region was investigated theoretically for experiments by using a 14 MeV neutron source. We carried out a systematical analysis for light to heavy materials. As a result, the benchmark performance for the neutron spectrum was confirmed to be acceptable, while for gamma-rays it was not sufficiently accurate. Consequently, a spectrum shifter has to be applied. Beryllium had the best performance as a shifter. Moreover, a preliminary examination of whether it is really acceptable that only the spectrum before the last collision is considered in the benchmark performance analysis. It was pointed out that not only the last collision but also earlier collisions should be considered equally in the benchmark performance analysis.

  1. Nuclear analysis software. Pt. 2: Gamma spectrum analysis, activity calculations and neutron activiation analysis (GANAAS)

    International Nuclear Information System (INIS)

    1991-01-01

    A spectrum acquired with a multichannel analyzer is usually stored with a suitable device (tape, cassette tape, diskette, hard disk). Every manufacturer of multichannel analyzers uses his own method for storage, and records the spectra in his own format. Furthermore, the formats to save the spectra evolve in time: the same manufacturer can have several formats for different generations of multichannel analyzers. A similar situation prevails with the spectrum analysis programmes. They require spectra in a particular format as the input to the analysis. Again, these input formats are many and differ from each other considerably. SPEDAC set of routines was developed to provide the spectroscopist with a tool for converting the spectral formats. They can read the spectra recorded in a number of formats used in different multichannel analyzers, to a number of analysis programmes. In fact, all the major formats are represented. Another serious problem for the user of a stand-alone multichannel analyzer is the transfer of spectra from the MCA to the computer. For several well known types of MCAs, the Version 5.0 of SPEDAC offers a set of routines for spectrum transfer, using the most simple methods of interfacing. All the transfer programmes described in this manual have been carefully tested with the corresponding stand-alone multichannel analyzers

  2. Neutron-induced nuclear data for the MYRRHA fast spectrum facility

    OpenAIRE

    Romojaro Pablo; Žerovnik Gašper; Álvarez-Velarde Francisco; Stankovskiy Alexey; Kodeli Ivan; Fiorito Luca; Díez Carlos Javier; Cabellos Óscar; García-Herranz Nuria; Heyse Jan; Paradela Carlos; Schillebeeckx Peter; Eynde Gert Van den

    2017-01-01

    The MYRRHA (Multi-purpose hYbrid Research Reactor for High-tech Applications) concept is a flexible experimental lead-bismuth cooled and mixed-oxide (MOX) fueled fast spectrum facility designed to operate both in sub-critical (accelerator driven) and critical modes. One of the key issues for the safe operation of the reactor is the uncertainty assessment during the design works. The main objective of the European project CHANDA (solving CHAllenges in Nuclear DAta) Work Package 10 is to improv...

  3. Mutation spectrum in peas following treatment with fast neutrons and NEH

    International Nuclear Information System (INIS)

    Swiecicki, W.K.

    1989-01-01

    Full text: Seeds of Pisum lines 'Wt 3527' and 'Wt 4042' were treated with 200 and 500 rad of N f or 0.014 % NEH (8h, pH 4,5-5) or the combination of the two (irradiation followed by NEH treatment). Among 30,000 M 2 plants 1314 mutants were identified. N f were the least effective in terms of number of mutants, but induced the widest spectrum, alone as well as in combination with NEH. A number of hitherto unknown mutant genes were identified after the combined treatment of the line 'Wt 3527'. (author)

  4. Use of boron nitride for neutron spectrum characterization and cross-section validation in the epithermal range through integral activation measurements

    Energy Technology Data Exchange (ETDEWEB)

    Radulović, Vladimir, E-mail: vladimir.radulovic@ijs.si [Jožef Stefan Institute, Jamova cesta 39, SI-1000 Ljubljana (Slovenia); Trkov, Andrej [Jožef Stefan Institute, Jamova cesta 39, SI-1000 Ljubljana (Slovenia); IAEA, Vienna International Centre, PO Box 100, A-1400 Vienna (Austria); Jaćimović, Radojko [Jožef Stefan Institute, Jamova cesta 39, SI-1000 Ljubljana (Slovenia); Gregoire, Gilles; Destouches, Christophe [CEA, DEN, DER, Instrumentation Sensors and Dosimetry Laboratory, Cadarache, F-13108 St. Paul-Lez-Durance (France)

    2016-12-21

    A recent experimental irradiation and measurement campaign using containers made from boron nitride (BN) at the Jožef Stefan Institute (JSI) TRIGA Mark II reactor in Ljubljana, Slovenia, has shown the applicability of BN for neutron spectrum characterization and cross-section validation in the epithermal range through integral activation measurements. The first part of the paper focuses on the determination of the transmission function of a BN container through Monte Carlo calculations and experimental measurements. The second part presents the process of tayloring the sensitivity of integral activation measurements to specific needs and a selection of suitable radiative capture reactions for neutron spectrum characterization in the epithermal range. A BN container used in our experiments and its qualitative effect on the neutron spectrum in the irradiation position employed is displayed in the Graphical abstract. - Graphical abstract: Neutron spectra inside the JSI TRIGA Mark II PT irradiation position, obtained with a Monte Carlo calculation: blue: unperturbed, green inside a BN container, of wall thickness 4 mm, 13 mm in diameter and 14 mm in height.

  5. Determination of intensity and energy spectrum of neutrons by bombardment of thallium-203 thick target and its copper substrate with 28.5 MeV protons

    International Nuclear Information System (INIS)

    Hajiloo, N.; Raisali, Gh.; Hamidi, S.; Aslani, Gh.

    2007-01-01

    In this research we have determined neutrons spectrum and the intensity that produced from thallium target bombardment. We have applied SRIM and ALICE computer codes to thallium target and its copper substrate for 145 μA of 28.5 MeV incident proton beam from cyclotron Cyclone30. Because of the energy degradation of protons while passing through the thallium target and its copper substrate, the average energy of protons in different depths has been calculated by using SRIM computer code. Then, by applying ALICE computer code for each sub-layer, the neutron production cross sections and their energy spectrum have been calculated to determine the total neutron intensity and spectrum. Using the calculated neutron intensity of 1.22x10 13 n/s as the source, the equivalent dose rate at the distance 6 meters from the target has been calculated by MCNP computer code and the result has been compared with the measured value. The Pb 201 activity has also been calculated as 13.5 Curies. The measured Pb 201 activity by Curie meter CAPINTEC CRC-712 is 13.1 Ci which is in reasonable agreement with the calculated value, bearing in mind the uncertainties in the proposed models and the measurements

  6. Use of boron nitride for neutron spectrum characterization and cross-section validation in the epithermal range through integral activation measurements

    International Nuclear Information System (INIS)

    Radulović, Vladimir; Trkov, Andrej; Jaćimović, Radojko; Gregoire, Gilles; Destouches, Christophe

    2016-01-01

    A recent experimental irradiation and measurement campaign using containers made from boron nitride (BN) at the Jožef Stefan Institute (JSI) TRIGA Mark II reactor in Ljubljana, Slovenia, has shown the applicability of BN for neutron spectrum characterization and cross-section validation in the epithermal range through integral activation measurements. The first part of the paper focuses on the determination of the transmission function of a BN container through Monte Carlo calculations and experimental measurements. The second part presents the process of tayloring the sensitivity of integral activation measurements to specific needs and a selection of suitable radiative capture reactions for neutron spectrum characterization in the epithermal range. A BN container used in our experiments and its qualitative effect on the neutron spectrum in the irradiation position employed is displayed in the Graphical abstract. - Graphical abstract: Neutron spectra inside the JSI TRIGA Mark II PT irradiation position, obtained with a Monte Carlo calculation: blue: unperturbed, green inside a BN container, of wall thickness 4 mm, 13 mm in diameter and 14 mm in height.

  7. Fast Neutron Spectrum Potassium Worth for Space Power Reactor Design Validation

    Energy Technology Data Exchange (ETDEWEB)

    Bess, John D. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Marshall, Margaret A. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Briggs, J. Blair [Idaho National Lab. (INL), Idaho Falls, ID (United States); Tsiboulia, Anatoli [Idaho National Lab. (INL), Idaho Falls, ID (United States); Rozhikhin, Yevgeniy [Idaho National Lab. (INL), Idaho Falls, ID (United States); Mihalczo, John T. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2015-03-01

    A variety of critical experiments were constructed of enriched uranium metal (oralloy ) during the 1960s and 1970s at the Oak Ridge Critical Experiments Facility (ORCEF) in support of criticality safety operations at the Y-12 Plant. The purposes of these experiments included the evaluation of storage, casting, and handling limits for the Y-12 Plant and providing data for verification of calculation methods and cross-sections for nuclear criticality safety applications. These included solid cylinders of various diameters, annuli of various inner and outer diameters, two and three interacting cylinders of various diameters, and graphite and polyethylene reflected cylinders and annuli. Of the hundreds of delayed critical experiments, one was performed that consisted of uranium metal annuli surrounding a potassium-filled, stainless steel can. The outer diameter of the annuli was approximately 13 inches (33.02 cm) with an inner diameter of 7 inches (17.78 cm). The diameter of the stainless steel can was 7 inches (17.78 cm). The critical height of the configurations was approximately 5.6 inches (14.224 cm). The uranium annulus consisted of multiple stacked rings, each with radial thicknesses of 1 inch (2.54 cm) and varying heights. A companion measurement was performed using empty stainless steel cans; the primary purpose of these experiments was to test the fast neutron cross sections of potassium as it was a candidate for coolant in some early space power reactor designs.The experimental measurements were performed on July 11, 1963, by J. T. Mihalczo and M. S. Wyatt (Ref. 1) with additional information in its corresponding logbook. Unreflected and unmoderated experiments with the same set of highly enriched uranium metal parts were performed at the Oak Ridge Critical Experiments Facility in the 1960s and are evaluated in the International Handbook for Evaluated Criticality Safety Benchmark Experiments (ICSBEP Handbook) with the identifier HEU MET FAST 051. Thin

  8. Determination of the neutrons energy spectrum in the central thimble of the reactor core TRIGA Mark III; Determinacion del espectro de energia de los neutrones en el dedal central del nucleo del reactor TRIGA Mark III

    Energy Technology Data Exchange (ETDEWEB)

    Parra M, M. A.; Luis L, M. A. [Universidad Autonoma Metropolitana, Unidad Azcapotzalco, Division de Ciencias Basicas, Av. San Pablo No. 180, Col. Reynosa Tamaulipas, 02200 Mexico D. F. (Mexico); Raya A, R.; Cruz G, H. S., E-mail: roberto.raya@inin.gob.mx [ININ, Departamento del Reactor, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2013-10-15

    This work presents the measurement of the neutrons spectrum in energies in the central thimble of the reactor TRIGA Mark III to a power of 1 MW in stationary state, with the core in the center of the pool. To achieve this objective, several thin sheets were irradiated (one at the time) in the same position of the core. The activation probes were selected in such a way that covered the energy range (1 x 10{sup -10} to 20 MeV) of the neutrons spectrum in the reactor core, for this purpose thin sheets were used of {sup 197}Au, {sup 58}Ni, {sup 115}In, {sup 24}Mg, {sup 27}Al, {sup 58}Fe, {sup 59}Co and {sup 63}Cu. After the irradiation, the high energy gamma emissions of the activated thin sheets were measured by means of gamma spectrometry, in a counting system of high resolution, with a Hyper pure Germanium detector, obtaining this way the activity induced in the thin sheets whose magnitude is proportional to the intensity of the neutrons flow, this activity together to a theoretical initial spectrum are the main entrance data of the computational code SANDBP (Hungarian version of the code Sand-II) that uses the unfolding method for the calculation of the spectrum. (Author)

  9. ESELEM 4: a code for calculating fine neutron spectrum and multi-group cross sections in plate lattice

    International Nuclear Information System (INIS)

    Nakagawa, Masayuki; Katsuragi, Satoru; Narita, Hideo.

    1976-07-01

    The multi-group treatment has been used in the design study of fast reactors and analysis of experiments at fast critical assemblies. The accuracy of the multi-group cross sections therefore affects strongly the results of these analyses. The ESELEM 4 code has been developed to produce multi-group cross sections with an advanced method from the nuclear data libraries used in the JAERI Fast set. ESELEM 4 solves integral transport equation by the collision probability method in plate lattice geometry to obtain the fine neutron spectrum. A typical fine group mesh width is 0.008 in lethargy unit. The multi-group cross sections are calculated by weighting the point data with the fine structure neutron flux. Some devices are applied to reduce computation time and computer core storage required for the calculation. The slowing down sources are calculated with the use of a recurrence formula derived for elastic and inelastic scattering. The broad group treatment is adopted above 2 MeV for dealing with both light any heavy elements. Also the resonance cross sections of heavy elements are represented in a broad group structure, for which we use the values of the JAERI Fast set. The library data are prepared by the PRESM code from ENDF/A type nuclear data files. The cross section data can be compactly stored in the fast computer core memory for saving the core storage and data processing time. The programme uses the variable dimensions to increase its flexibility. The users' guide for ESELEM 4 and PRESM is also presented in this report. (auth.)

  10. Experimental determination of spectral ratios and of neutrons energy spectrum in the fuel of the IPEN/MB-01 nuclear reactor

    International Nuclear Information System (INIS)

    Nunes, Beatriz Guimaraes

    2012-01-01

    This study aims to determine the spectral ratios and the neutron energy spectrum inside the fuel of IPEN/MB-01 Nuclear Reactor. These parameters are of great importance to accurately determine spectral physical parameters of nuclear reactors like reaction rates, fuel lifetime and also security parameters such as reactivity. For the experiment, activation detectors in the form of thin metal foils were introduced in a collapsible fuel rod. Then the rod was placed in the central position of the core which has a standard rectangular configuration of 26 x 28 fuel rods. There were used activation detectors from different elements such Au-197, U-238, Sc-45, Ni-58, Mg-24, Ti-47 and In-115 to cover a large range of the neutron energy spectrum. After the irradiation, the activation detectors were submitted to gamma spectrometry using a counting system with high purity Germanium, to obtain the reaction rates (saturation activity) per target nucleus. The spectral ratios were compared with calculated values obtained by the Monte Carlo method using the MCNP-4C code. The neutron energy spectrum was obtained inside the fuel rod using the SANDBP code with an input spectrum obtained by the MCNP-4C code, based on the saturation activity per target nucleus values of the activation detectors irradiated. (author)

  11. Neutron Spectrum Parameters In Inner Irradiation Channel Of The Nigeria Research Reactor-1 (NIRR-1) For Use In Absolute And KO-NAA Methods

    International Nuclear Information System (INIS)

    Jonah, S.A; Balogun, G.I; Mayaki, M.C.

    2004-01-01

    In Nigeria, the first Nuclear Reactor achieved critically on February 03, 2004 at about 11:35 GMT and has been commissioned or training and research. It is a Miniature Neutron Source Reactor (MNSR), code-named Nigeria Research Reactor-1 (NIRR-1). NIRR-1 has a tan-in-pool structural configuration and a nominal thermal power rating of 30 Kw. With a built-in clean old core excess reactivity of 3.77 mk determined during the on-site zero and critically experimental, the reactor can operate for a n.cm-2 .s-1 in the inner irradiation channels). Under these conditions, the reactor can operate with the same fuel loading for over ten years with a burn-up of <1%. A detailed description of operating characteristics for NIRR-1, measured during the on-site zero-power and criticality experiments has been given elsewhere. In order to extend its utilization to include absolute and ko-NAA methods, the neutron spectrum parameters in the irradiation channels: power and critically experiments has been given elsewhere. In order to extend it's the irradiation channels: thermal-to-epithermal flux ration, F; and epithermal flux shape factor, a in both the inner and outer irradiation channels must be determined experimentally. In this work, we have developed and experimental procedure for monitoring the neutron spectrum parameters in an inner irradiation channel based on irradiation and gamma-ray counting of detector foils via (n,y), (n,p) and (n,a) dosimetry reactions. Results obtained indicate that a thermal neutron flux of (5.14+-0.02) x 1011 n/c m2.s determined by foil activation method in the inner irradiation channel, B2, at a power level of 15.5 kw corresponds to the flux indicators on the control console and the micro-computer control system respectively. Other parameters of the neutron spectrum determined for inner irradiation channel B2, are: a -0.0502+0.003; 18.92+-0.14; F = 3.87=0.23. The method was validated through the comparison of our result with published neutron spectrum

  12. Experimental measurement of neutron spectrum in the reflector of a light water reactor; Determination experimentale du spectre des neutrons dans le reflecteur d'une pile a eau legere

    Energy Technology Data Exchange (ETDEWEB)

    Brethe, P [Commissariat a l' Energie Atomique, Grenoble (France). Centre d' Etudes Nucleaires

    1963-09-15

    1. Thermal neutrons: The temperature of the thermal neutron spectrum was calculated using Au-Lu foils. This temperature varies from 300 deg. K (temperature of the moderator) at 30 cm of the core to 350 deg. K in a hole of the core. 2. Slowing down of neutron: Four resonance detectors have been used (Au, In, Co, Mn). We can write a 1/E form of the spectrum. The linking up energy E{sub M} between thermal neutron spectrum and slowing down spectrum is about 0.23 eV and is free from the Maxwell spectrum temperature. The decrease of slowing down flux regarding thermal flux, farther from the core, has been showed. 3. Fast neutrons: We used 3 threshold detectors (Ni, Al, Mg). We supposed a E{sup 1/2} e{sup -{beta}}{sup E} from of the spectrum above 3 MeV. The values of {beta} are in a range from 0.775, at the centre of the core and in a loop-hole, to 0,64 at about 30 cm of the core. 4. Continuous shape of the spectrum: The following interpolations give useful informations between the field where measurements have been made: between 340 eV and 10 keV: 1/E form between 10 keV and 330 keV: 1/(E {sigma}{sub S}(E)) form ({sigma}{sub S}(E) elastic scattering section on hydrogen) between 330 keV and 3 MeV: calculated form by the moments method (ref. BSR). (author) [French] 1. Neutrons thermiques: La temperature du spectre des neutrons thermiques a ete determinee par la methode (or-lutecium). Cette temperature varie de 300 deg. K (temperature du moderateur) a 30 cm du coeur, a 350 deg. K dans une encoche du coeur. 2. Neutrons en ralentissement: 4 detecteurs resonnants ont ete employes (Au, In, Co, Mn). Le spectre peut etre mis sous la forme 1/E quelle que soit la distance a la limite coeur-reflecteur. L'energie de raccordement E{sub M} entre spectre des neutrons thermiques et spectre en ralentissement est environ 0,23 eV et independante de la temperature du spectre de Maxwell. La diminution relative du flux en ralentissement par rapport au flux thermique quand la distance au coeur

  13. The energy spectrum of cosmic-ray induced neutrons measured on an airplane over a wide range of altitude and latitude

    International Nuclear Information System (INIS)

    Goldhagen, P.; Clem, J. M.; Wilson, J. W.

    2004-01-01

    Crews of high-altitude aircraft are exposed to radiation from galactic cosmic rays (GCRs). To help determine such exposures, the Atmospheric Ionizing Radiation Project, an international collaboration of 15 laboratories, made simultaneous radiation measurements with 14 instruments on a NASA ER-2 high-altitude airplane. The primary instrument was a sensitive extended-energy multisphere neutron spectrometer. Its detector responses were calculated for energies up to 100 GeV using the radiation transport code MCNPX 2.5.d with improved nuclear models and including the effects of the airplane structure. New calculations of GCR-induced particle spectra in the atmosphere were used to correct for spectrometer counts produced by protons, pions and light nuclear ions. Neutron spectra were unfolded from the corrected measured count rates using the deconvolution code MAXED 3.1. The results for the measured cosmic-ray neutron spectrum (thermal to >10 GeV), total neutron fluence rate, and neutron dose equivalent and effective dose rates, and their dependence on altitude and geomagnetic cut-off agree well with results from recent calculations of GCR-induced neutron spectra. (authors)

  14. Reactor AQUILON. The hardening of neutron spectrum in natural uranium rods, with a computation of epithermal fissions (1961); Pile AQUILON. Durcissement du spectre des neutrons dans les barreaux d'uranium et calcul des fissions epithermiques (1961)

    Energy Technology Data Exchange (ETDEWEB)

    Durand -Smet, R; Lourme, P [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1961-07-01

    - Microscopic flux measurements in reactor Aquilon have allowed to investigate the thermal and epithermal flux distribution in natural uranium rods, then to obtain the neutron spectrum variations in uranium, Wescott '{beta}' term of the average spectrum in the rod, and the ratio of epithermal to therma fissions. A new definition for the infinite multiplication factor is proposed in annex, which takes into account epithermal parameters. (authors) [French] - Un certain nombre de mesures effectuees dans la pile Aquilon ont permis d'etablir la distribution fine des flux thermique et epithermique dans les barreaux d'uranium, et d'en deduire les variations du spectre des neutrons dans l'uranium, le terme {beta} du spectre de Wescott moyen dans le barreau et le nombre de fissions epithermiques. En annexe, il est propose une definition nouvelle du coefficient de multiplication infini, qui fait intervenir les parametres epithermiques. (auteurs)

  15. PC based system for absolute neutron flux-spectrum measurements. Final report for the period 1 April 1994 - 15 December 1997

    International Nuclear Information System (INIS)

    Dobrea, D.

    1998-01-01

    When measuring absolute neutron flux-spectra, thin detector foils are irradiated in a neutron field. The absolute activity of isotopes generated by activation or fission reactions (fission products) is then measured, using an efficiency calibrated high resolution gamma-ray spectroscopy system, and the absolute reaction rates for their production is determined. Finally the flux-spectrum is determined based upon the reaction rate values. A general method to obtain flux-spectra from the reaction rate is the unfolding method. The above process involves computations of photopeak area, reaction rate, flux perturbation corrections and flux-spectrum. The PC's are well suited for the data processing system outlined above. Using available interfaces, the PC's can be involved even in the data acquisition. Graphical facilities allow decision upon the data processing flow

  16. SB3. Experiment on secondary gamma-ray production cross sections averaged over a fast-neutron spectrum for each of 13 different elements plus a stainless steel

    International Nuclear Information System (INIS)

    Maerker, R.E.

    1976-01-01

    The experimental and calculational details for a CSEWG integral data testing shielding experiment are presented. This particular experiment measured the secondary gamma-ray production cross sections averaged over a fast-neutron spectrum for iron, oxygen, sodium, aluminum, copper, titanium, calcium, potassium, silicon, nickel, zinc, barium, sulfur, and a type 321 stainless steel. The gamma-ray production cross sections were binned into 0.5-MeV wide gamma-ray energy intervals. 29 tables, 1 figure

  17. Reconstruction of the neutron spectrum using an artificial neural network in CPU and GPU; Reconstruccion del espectro de neutrones usando una red neuronal artificial (RNA) en CPU y GPU

    Energy Technology Data Exchange (ETDEWEB)

    Hernandez D, V. M.; Moreno M, A.; Ortiz L, M. A. [Universidad de Cordoba, 14002 Cordoba (Spain); Vega C, H. R.; Alonso M, O. E., E-mail: vic.mc68010@gmail.com [Universidad Autonoma de Zacatecas, 98000 Zacatecas, Zac. (Mexico)

    2016-10-15

    The increase in computing power in personal computers has been increasing, computers now have several processors in the CPU and in addition multiple CUDA cores in the graphics processing unit (GPU); both systems can be used individually or combined to perform scientific computation without resorting to processor or supercomputing arrangements. The Bonner sphere spectrometer is the most commonly used multi-element system for neutron detection purposes and its associated spectrum. Each sphere-detector combination gives a particular response that depends on the energy of the neutrons, and the total set of these responses is known like the responses matrix Rφ(E). Thus, the counting rates obtained with each sphere and the neutron spectrum is related to the Fredholm equation in its discrete version. For the reconstruction of the spectrum has a system of poorly conditioned equations with an infinite number of solutions and to find the appropriate solution, it has been proposed the use of artificial intelligence through neural networks with different platforms CPU and GPU. (Author)

  18. Nuclear data and measurements series: Ratio of the prompt-fission-neutron spectrum of plutonium 239 to that of uranium 235

    International Nuclear Information System (INIS)

    Sugimoto, M.; Smith, A.B.; Guenther, P.T.

    1986-09-01

    The prompt-fission-neutron spectrum resulting from 239 Pu fission induced by 0.55 MeV incident neutrons is measured from 1.0 to 10.0 MeV relative to that of 235 U fission induced by the same incident-energy neutrons. The measurements employ the time-of-flight technique. Energy-dependent ratios of the two spectra are deduced from the measured values over the energy range 1.0 to 10.0 MeV. The experimentally-derived ratio results are compared with those calculated from ENDF/B-V, revision-2, and with results of recent microscopic measurements. Using the ENDF/B-V 235 U Watt parameters for the 235 U spectrum, the experimental measurements imply a ratio of average fission-spectrum energies of 239 Pu/ 235 U = 1.045 +- 0.003, compared to the value 1.046 calculated from ENDF/B-V, revision 2. 12 refs., 2 figs., 2 tabs

  19. Simulating the spectrum of neutrons produced by a radiation beam of high voltage inside an anthropomorphic phantom; Simulacion de espectro de nuetrones producido por un haz de radioterapia de alto voltaje en el interior de un manique antropomorfico

    Energy Technology Data Exchange (ETDEWEB)

    Gonzalez-Soto, X.; Amgarou, K.; Langares, J. I.; Munez, J. L.; Mendez, R.; Exposito, M. R.; Gomez, F.; Domingo, C.; Sanche-Doblado, F.

    2011-07-01

    Our project aims to provide a universal method to estimate the dose deposited by neutrons in patients, using an anthropomorphic phantom. Both the detector response as relative biological effectiveness have a strong dependence on the energy spectrum of those, for this reason, a series of simulations were performed to calculate the spectrum of the neutron fluence in 16 representative points within the anthropomorphic phantom Standard for a full course of radiotherapy.

  20. Measurement of thermal neutron spectrum by chopper at the RA reactor in the 'Boris Kidric' Institute; Merenje termickog neutronskog spektra iz reaktora RA u Institutu 'Boris Kidric' pomocu copera

    Energy Technology Data Exchange (ETDEWEB)

    Maglic, R [Boris Kidric Institute of Nuclear Sciences Vinca, Belgrade (Yugoslavia)

    1963-04-15

    Measurement of neutron spectrum described in this paper is one of the first measurements of neutron energy distribution at the reactor by time-of-flight method performed in our Institute. Measurement was done by applying the mechanical spectrometer (chopper) designed and constructed in 1961. Spectrometer was calibrated at the end of 1962.

  1. Frequency Spectrum of Liquids and Cold Neutron Scattering; Spectre de frequences des liquides et diffusion de neutrons froids; Chastotnyj spektr zhidkostej i rasseyanie kholodnykh nejtronov; Espectro de frecuencias de los liquidos y dispersion de neutrones frios

    Energy Technology Data Exchange (ETDEWEB)

    Singwi, K S; Sjolander, A; Rahman, A [Argonne National Laboratory, Argonne, IL (United States)

    1963-01-15

    An important question which arises in connection with slow neutron scattering by liquids is: does there exist a frequency spectrum tor liquids analogous to that in solids? The answer to this question is given by showing that the width function {gamma}(t) of the Gaussian space-time self-correlation function G{sub s}(r,t) can be expressed, by using the frequency spectrum of the velocity auto-correlation function, in a form which is formally identical with that of a harmonic solid. Thus a knowledge of the frequency spectrum of the velocity auto-correlation function should enable one to calculate slow neutron scattering by liquids as has been emphasized by Egelstaff and co-workers. Using a stochastic model of a liquid, the frequency spectrum f({omega}) of the velocity auto-correlation function is calculated for water at 300{sup o}K and for liquid lead at 620{sup o}K. In the case of water a maximum in f({omega}) corresponding to {Dirac_h} 75{sup o}K is predicted. For lead the maximum also occurs at nearly the same {omega}-value. There also occurs a minimum in f({omega})in both cases. Larsson and Dahlborg have observed a maximum in f({omega}) at the above predicted {omega}-value. A recent observation in liquid lead of Cotter et al. probably confirms our prediction too. (author) [French] Dans le domaine de la diffusion des neutrons lents par des liquides une question importante se pose: existe-t-il un spectre de frequences pour les liquides analogue au spectre pour les solides? On peut repondre d cette quesnon en montrant que la fonction de largeur {gamma}(t) de la fonction gaussienne spatio-temporelle d'auto-correlation G{sub s}(r,t) peut etre exprimee, en employant le spectre de frequences de la fonction d'auto-correlation des vitesses, sous uune forme qui presente le membe aspect formel que celle d'un solide harmonique. par consequent, la connaissance du spectre de frequences de la fonction d'auto-correlation des vitesses devrait permettre de calculer la diffusion des

  2. Study of the influence of the fast neutron spectrum on the production of defects in solids and liquids; Etude de l'influence du spectre des neutrons rapides sur la creation de defauts dans les solides et les liquides

    Energy Technology Data Exchange (ETDEWEB)

    Mas, P; Droulers, Y [Commissariat a l' Energie Atomique, Grenoble (France). Centre d' Etudes Nucleaires

    1964-07-01

    In the first part of this work a calculation has been made of the number of defects formed in graphite by a given neutron flux having various spectral distributions. The defect formation function is that of KINCHIN and PEASE; its formulation is briefly given. An efficiency function is then defined for a fast neutron spectrum. This defects produced in a light water reactor and those produced in a graphite reactor. Finally an application of this method is given for comparing the defect forming tendency in graphite in the case of the reactor Melusine and of the reactor G-2 and G-3. In the second part are calculated the integrals for the energy release brought about by fast neutrons in carbon oxygen and hydrogen. In a region of 25 cm around the core of a swimming-pool type reactor these energy release integrals are approximately proportional to the neutron flux above 1 MeV. The determination of the energy released as a result of the passage of neutrons in organic liquids can therefore be reduced to the measurement of the flux above 1 MeV for the real spectral distribution. A calorimetric verification has been carried out in the case of water. (authors) [French] Dans la premiere partie de cette etude, on a calcule le nombre de defauts crees dans le graphite par un flux de neutrons donne, pris sous differents spectres. On a ainsi fait les calculs pour 4 spectres de pile. piscine et 2 spectres de pile au graphite. La fonction creation de defauts est celle de KINCHIN et PEASE. On rappelle brievement sa formulation. Puis on definit une fonction efficacite d'un spectre de neutrons rapides. Cette fonction permet alors d'etablir la relation entre defauts crees dans une pile a eau legere et defauts crees dans une pile au graphite. Enfin, une application de cette methode est donnee pour comparer l'aptitude a creer les defauts dans le graphite par la pile MELUSINE et les piles G-2 ou G-3. Dans la deuxieme partie, on a calcule les integrales de degagement d'energie provoque par

  3. A stochastic model for neutron simulation considering the spectrum and nuclear properties with continuous dependence of energy

    International Nuclear Information System (INIS)

    Camargo, Dayana Q. de; Bodmann, Bardo E.J.; Vilhena, Marco T. de; Froehlich, Herberth B.

    2011-01-01

    In this work we developed a stochastic model to simulate neutron transport in a heterogeneous environment, considering continuous neutron spectra and the nuclear properties with its continuous dependence on energy. This model was implemented using the Monte Carlo method for the propagation of neutrons in different environments. Due to restrictions with respect to the number of neutrons that can be simulated in reasonable computational time we introduced a variable control volume together with (pseudo-) periodic boundary conditions in order to overcome this problem. This study allowed a detailed analysis of the influence of energy on the neutron population and its impact on the life cycle of neutrons. From the results, even for a simple geometrical arrangement, we can conclude that there is need to consider the energy dependence and hence defined a spectral effective multiplication factor per Monte Carlo step. (author)

  4. Optimization of artificial neural networks for the reconstruction of the neutrons spectrum and their equivalent doses; Optimizacion de redes neuronales artificiales para la reconstruccion del espectro de neutrones y sus dosis equivalentes

    Energy Technology Data Exchange (ETDEWEB)

    Reyes A, A.; Ortiz R, J. M.; Reyes H, A.; Castaneda M, R.; Solis S, L. O.; Vega C, H. R., E-mail: art8291@hotmail.com [Universidad Autonoma de Zacatecas, Unidad Academica de Ingenieria Electrica, Av. Lopez Velarde No. 801, Col. Centro, 98000 Zacatecas (Mexico)

    2014-08-15

    In this work was used the robust design methodology of artificial neural networks to determine a good topology of net able to solve with efficiency the problems of neutrons spectrometry and dosimetry. For the design of the topology of optimized net 36 different net architectures based on an orthogonal arrangement with a configuration L{sub 9}(3{sup 4}), L{sub 4}(3{sup 2}) were trained. For the training of the neural networks, was used a computer code developed in the ambient of Mat lab programming, which automates the process and analysis of the information, reducing the time used in this activity considerably for the investigator. For the training of the propagation nets forward was utilized a neutrons spectrum compendium published by the International Atomic Energy Agency, where of the total 80% was used for the training and 20% for the test, it trained with an inverse propagation algorithm being the entrance data the count rates corresponding to the 7 spheres of the spectrometric system of Bonner spheres, as exit data, the neural network obtains the neutrons spectrum expressed in 60 energy groups and are calculated of simultaneous way 15 dosimetric quantities. (Author)

  5. Study of U{sup 235} neutron fission spectrum by the knowledge of cross sections average over that spectrum; Estudio del espectro de neutrones de fision del {sup 235}U a traves del conocimiento de secciones eficaces promediadas sobre dicho espectro

    Energy Technology Data Exchange (ETDEWEB)

    Suarez, P M [Comision Nacional de Energia Atomica, San Carlos de Bariloche (Argentina). Centro Atomico Bariloche

    1998-12-31

    A literature search of cross sections averaged over the fission neutron spectrum confirms inconsistencies between calculated and experimental values for high threshold reactions. Since, in this case, calculated averaged cross sections are systematically lower than measured values, it is concluded that the representations used to carry out these calculations underestimate the number of neutrons in the high energy region of the spectrum. A careful measurement of the averaged cross section for the {sup 45}Sc(n,2n) {sup 44g}Sc and {sup 45}Sc(n,2n) {sup 44m}Sc high threshold reactions had been performed in the RA-6 Neutron Activation Analysis Laboratory after carefully checking that the neutron flux at the core position where the samples were being irradiated was indeed an undisturbed fission spectrum. The experimental values are greater than those calculated with either, Watt type representations or the one based on the Madland and Nix model for the prompt fission spectrum. In many areas of nuclear engineering, like validation of nuclear data, reactor calculations, applied nuclear physics, shielding design, etc., it is of great practical importance to have a representation for the neutron flux that can be expressed in a closed analytical form and that agrees with experimental results, specially for the most widely fissile nuclide, {sup 235}U. The results of the calculations mentioned above lead us to propose an analytical form for the {sup 235}U fission neutron spectrum that better agrees with experimental results in the whole energy spectrum. We propose two different forms; both are a modification of the Watt-type form that has been adopted within the ENDF/B-V files. One of the new analytical representations is defined in two regions: below 9.5 MeV it is exactly the same formula as that used within the ENDF/B-V files, above this energy the parameters of this formula are changed. The other proposed analytical representation is expressed by a single formula in the whole

  6. The effect of neutron spectrum on the mechanical and physical properties of pure copper and copper alloys

    DEFF Research Database (Denmark)

    Fabritsiev, S.A.; Pokrovsky, A.S.; Zinkle, S.J.

    1996-01-01

    -irradiated inside a 1.5 mm Cd shroud in order to reduce the thermal neutron flux. The electrical resistivity data could be separated into two components, a solid transmutation component Delta rho(tr) which was proportional to thermal neutron fluence and a radiation defect component Delta rho(rd) which...

  7. Fission neutron spectrum averaged cross sections of some threshold reactions on cadmium: production feasibility of no-carrier-added 103Pd in a nuclear reactor

    International Nuclear Information System (INIS)

    Abbasi, I.A.; Subhani, M.S.; Zaidi, J.H.; Arif, M.

    2006-01-01

    Systematic studies on fission neutron spectrum averaged cross sections of some threshold reactions like (n, p) and (n, α) on cadmium were carried out using the activation technique in combination with radiochemical separations and high-resolution γ-ray spectroscopy. Special attention was paid to the formation of 103 Pd via the 106 Cd(n,a α) 103 Pd reaction since it is an important therapeutic radionuclide. At a fast flux neutron density of 7.5 x 10 13 cm 2 s -1 and an irradiation time of 120 h, using 100% enriched 106 Cd target 340 MBq of no-carrier-added 103 Pd per batch could be produced. The method is thus suitable for medium-scale production of this radionuclide. (orig.)

  8. Concept on coupled spectrum B/T (burning and/or transmutation) reactor for treatment of minor actinides by thermal and fast neutrons

    International Nuclear Information System (INIS)

    Aziz, Ferhat; Kitamoto, Asashi

    1996-01-01

    A conceptual design of B/T (burning and/or transmutation) reactor based on a modified conventional 1150 MWe-PWR system, with core consisted of two concentric regions for thermal and fast neutrons, was proposed herein for B/T treatment of MA (minor actinides). The B/T fuel considered was supposed such that MA discharged from 1 GWe-LWR was blended homogeneously with the composition of LWR fuel. In the outer region 23- Np, 241 Am and 243 Am were loaded and burned by thermal neutron, while in the inner region 244 Cm was loaded and burned mainly by fast neutron. The geometry of B/T fuel and the fuel assembly in the outer region was left in the same condition to those of standard PWR while in the inner region the B/T fuel was arranged in the hexagonal geometry, allowed high fuel to coolant volume ratio (V m /V f ), to keep the harder neutron spectrum. Two cases of the Coupled Spectrum B/T Reactor (CSR) with different (V m 1 f ) ratio in the inner region were studied, and the results for the tight lattice with (V m /V f ) = 0.5 showed that those isotopes approached the equilibrium composition after about 5 recycle period, when the CSR was operated under the reactivity swing of 2.8 % dk/k. The evaluations on the void coefficient of reactivity, the Doppler effect and the reactivity swing showed that the CSR concept has the inherent safety and can burn and/or transmute all kind of MA in a single reactor. This CSR can burn about 808 kg of MA in one recycle period of 3 years, which is equivalent to the discharged fuel from about 12 units of LWR in a year. (author)

  9. Influence of static Jahn-Teller distortion on the magnetic excitation spectrum of PrO2: A synchrotron x-ray and neutron inelastic scattering study

    International Nuclear Information System (INIS)

    Webster, C. H.; Helme, L. M.; Boothroyd, A. T.; McMorrow, D. F.; Wilkins, S. B.; Detlefs, C.; Detlefs, B.; Bewley, R. I.; McKelvy, M. J.

    2007-01-01

    A synchrotron x-ray diffraction study of the crystallographic structure of PrO 2 in the Jahn-Teller distorted phase is reported. The distortion of the oxygen sublattice, which was previously ambiguous, is shown to be a chiral structure in which neighboring oxygen chains have opposite chiralities. A temperature dependent study of the magnetic excitation spectrum, probed by neutron inelastic scattering, is also reported. Changes in the energies and relative intensities of the crystal field transitions provide an insight into the interplay between the static and dynamic Jahn-Teller effects

  10. Many channel spectrum unfolding

    International Nuclear Information System (INIS)

    Najzer, M.; Glumac, B.; Pauko, M.

    1980-01-01

    The principle of the ITER unfolding code as used for the many channel spectrum unfolding is described. Its unfolding ability is tested on seven typical neutron spectra. The effect of the initial spectrum approximation upon the solution is discussed

  11. A CHANGE IN THE QUIESCENT X-RAY SPECTRUM OF THE NEUTRON STAR LOW-MASS X-RAY BINARY MXB 1659-29

    Energy Technology Data Exchange (ETDEWEB)

    Cackett, E. M. [Department of Physics and Astronomy, Wayne State University, 666 W. Hancock Street, Detroit, MI 48201 (United States); Brown, E. F. [Department of Physics and Astronomy, National Superconducting Cyclotron Laboratory, and the Joint Institute for Nuclear Astrophysics, Michigan State University, East Lansing, MI 48824 (United States); Cumming, A. [Department of Physics, McGill University, 3600 rue University, Montreal, QC H3A 2T8 (Canada); Degenaar, N.; Miller, J. M. [Department of Astronomy, University of Michigan, 500 Church Street, Ann Arbor, MI 48109-1042 (United States); Fridriksson, J. K.; Wijnands, R. [Astronomical Institute ' ' Anton Pannekoek' ' , University of Amsterdam, Science Park 904, 1098-XH Amsterdam (Netherlands); Homan, J., E-mail: ecackett@wayne.edu [Kavli Institute for Astrophysics and Space Research, Massachusetts Institute of Technology, 70 Vassar Street, Cambridge, MA 02139 (United States)

    2013-09-10

    The quasi-persistent neutron star low-mass X-ray binary MXB 1659-29 went into quiescence in 2001, and we have followed its quiescent X-ray evolution since. Observations over the first 4 yr showed a rapid drop in flux and temperature of the neutron star atmosphere, interpreted as cooling of the neutron star crust which had been heated during the 2.5 yr outburst. However, observations taken approximately 1400 and 2400 days into quiescence were consistent with each other, suggesting the crust had reached thermal equilibrium with the core. Here we present a new Chandra observation of MXB 1659-29 taken 11 yr into quiescence and 4 yr since the last Chandra observation. This new observation shows an unexpected factor of {approx}3 drop in count rate and change in spectral shape since the last observation, which cannot be explained simply by continued cooling. Two possible scenarios are that either the neutron star temperature has remained unchanged and there has been an increase in the column density, or, alternatively the neutron star temperature has dropped precipitously and the spectrum is now dominated by a power-law component. The first scenario may be possible given that MXB 1659-29 is a near edge-on system, and an increase in column density could be due to build-up of material in, and a thickening of, a truncated accretion disk during quiescence. But, a large change in disk height may not be plausible if standard accretion disk theory holds during quiescence. Alternatively, the disk may be precessing, leading to a higher column density during this latest observation.

  12. A contribution for the problematic of measurements of fast-neutron-energy spectrum in thermal reactor-systems

    International Nuclear Information System (INIS)

    Dederichs, H.

    1978-06-01

    The aims of this work are to check the experimental conditions for using of a 6 Li-semiconductor-spectrometer at thermal reactor-systems and to measure the neutron-energy-spectra at the critical experiment KAHTER comparing with the theory. Using the spectrometer at thermal-neutraon-experiments questions will be attended as resolution, statistic and selection of usable nuclear data. The nuclear data will be gauged by qualified measurements, the statistic will be estimated by simulated calculations and the resolution will be improved by using the Fredholm-equation in the calculations. The calculated spectra show a good agreement with the measured spectra. Only in the energy region of maximum distribution of fission-neutrons there are little difference. The measurements show the using of the spectrometer is recommended at systems with preponderant thermal neutron-spectra, although the resolution and statistic are optimized for the spectrometer by measurements at experiments with fast neutron-spectra. (orig.) 891 RW [de

  13. Summary of fission spectrum workshop held at the National Neutron Cross Section Center, Brookhaven National Laboratory, October 23, 1978

    International Nuclear Information System (INIS)

    Stewart, L.

    1979-03-01

    In response to an action by the Standards Subcommittee of the Cross Section Evaluation Working Group, a workshop was convened to determine the status of available information on prompt fission neutron spectra. The experimental data were reviewed and theoretical models were developed. The current ENDF/B fission neutron spectra files were summarized. Further work is currently under way, especially to provide a better theoretical tool to represent energy-dependent fission spectra. 5 references

  14. Neutron spectra and cross sections for ice and clathrate generated from the synthetic spectrum and synthetic model for molecular solids

    International Nuclear Information System (INIS)

    Petriw, S; Cantargi, F; Granada, R

    2006-01-01

    We present here a Synthetic Model for Molecular Solids, aimed at the description of the interaction of thermal neutrons with this kind of systems.Simple representations of the molecular dynamical modes are used, in order to produce a fair description of neutron scattering kernels and cross sections with a minimum set of input data. Using those spectra, we have generated thermal libraries for M C N P [es

  15. A stochastic model for neutron simulation considering the spectrum and nuclear properties with continuous dependence of energy

    International Nuclear Information System (INIS)

    Camargo, Dayana Queiroz de

    2011-01-01

    This thesis has developed a stochastic model to simulate the neutrons transport in a heterogeneous environment, considering continuous neutron spectra and the nuclear properties with its continuous dependence on energy. This model was implemented using Monte Carlo method for the propagation of neutrons in different environment. Due to restrictions with respect to the number of neutrons that can be simulated in reasonable computational processing time introduced the variable control volume along the (pseudo-) periodic boundary conditions in order to overcome this problem. The choice of class physical Monte Carlo is due to the fact that it can decompose into simpler constituents the problem of solve a transport equation. The components may be treated separately, these are the propagation and interaction while respecting the laws of energy conservation and momentum, and the relationships that determine the probability of their interaction. We are aware of the fact that the problem approached in this thesis is far from being comparable to building a nuclear reactor, but this discussion the main target was to develop the Monte Carlo model, implement the code in a computer language that allows extensions of modular way. This study allowed a detailed analysis of the influence of energy on the neutron population and its impact on the life cycle of neutrons. From the results, even for a simple geometrical arrangement, we can conclude the need to consider the energy dependence, i.e. an spectral effective multiplication factor should be introduced each energy group separately. (author)

  16. Calculations of neutron spectra after neutron-neutron scattering

    Energy Technology Data Exchange (ETDEWEB)

    Crawford, B E [Gettysburg College, Box 405, Gettysburg, PA 17325 (United States); Stephenson, S L [Gettysburg College, Box 405, Gettysburg, PA 17325 (United States); Howell, C R [Duke University and Triangle Universities Nuclear Laboratory, Durham, NC 27708-0308 (United States); Mitchell, G E [North Carolina State University, Raleigh, NC 27695-8202 (United States); Tornow, W [Duke University and Triangle Universities Nuclear Laboratory, Durham, NC 27708-0308 (United States); Furman, W I [Joint Institute for Nuclear Research, 141980 Dubna (Russian Federation); Lychagin, E V [Joint Institute for Nuclear Research, 141980 Dubna (Russian Federation); Muzichka, A Yu [Joint Institute for Nuclear Research, 141980 Dubna (Russian Federation); Nekhaev, G V [Joint Institute for Nuclear Research, 141980 Dubna (Russian Federation); Strelkov, A V [Joint Institute for Nuclear Research, 141980 Dubna (Russian Federation); Sharapov, E I [Joint Institute for Nuclear Research, 141980 Dubna (Russian Federation); Shvetsov, V N [Joint Institute for Nuclear Research, 141980 Dubna (Russian Federation)

    2004-09-01

    A direct neutron-neutron scattering length, a{sub nn}, measurement with the goal of 3% accuracy (0.5 fm) is under preparation at the aperiodic pulsed reactor YAGUAR. A direct measurement of a{sub nn} will not only help resolve conflicting results of a{sub nn} by indirect means, but also in comparison to the proton-proton scattering length, a{sub pp}, shed light on the charge-symmetry of the nuclear force. We discuss in detail the analysis of the nn-scattering data in terms of a simple analytical expression. We also discuss calibration measurements using the time-of-flight spectra of neutrons scattered on He and Ar gases and the neutron activation technique. In particular, we calculate the neutron velocity and time-of-flight spectra after scattering neutrons on neutrons and after scattering neutrons on He and Ar atoms for the proposed experimental geometry, using a realistic neutron flux spectrum-Maxwellian plus epithermal tail. The shape of the neutron spectrum after scattering is appreciably different from the initial spectrum, due to collisions between thermal-thermal and thermal-epithermal neutrons. At the same time, the integral over the Maxwellian part of the realistic scattering spectrum differs by only about 6 per cent from that of a pure Maxwellian nn-scattering spectrum.

  17. Description of spectrum and electromagnetic transitions in 94Mo through the proton-neutron interacting boson model

    Science.gov (United States)

    Mu, ChengFu; Zhang, DaLi

    2018-01-01

    We investigated the properties of low-lying states in 94Mo within the framework of the proton-neutron interacting boson model (IBM-2), with special focus on the characteristics of mixed-symmetry states. We calculated level energies and M1 and E2 transition strengths. The IBM-2 results agree with the available quantitative and qualitative experimental data on 94Mo. The properties of mixed-symmetry states can be well described by IBM-2 given that the energy of the d proton boson is different from that of the neutron boson, especially for the transition of B( M1; 4 2 + → 4 1 + ).

  18. Reactor neutron dosimetry

    International Nuclear Information System (INIS)

    Najzer, M.; Pauko, M.; Glumac, B.; Acquah, I.N.; Moskon, F.

    1977-01-01

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

  19. Optical constants of quartz, vitreous silica and neutron-irradiated vitreous silica. II. Analysis of the infrared spectrum of vitreous silica

    Energy Technology Data Exchange (ETDEWEB)

    Gaskell, P H [Cambridge Univ. (UK). Cavendish Lab.; Johnson, D W [Pilkington Research and Development Laboratories, Lathom, nr. Ormskirk, Lancashire, UK

    1976-03-01

    Optical constant data for vitreous silica and neutron-irradiated vitreous silica, given in part I are examined in an attempt to gain further knowledge of the structure of the glass. Strong features of the spectrum can be described by band broadening parameters, which are calculated using an extension of the Wilson GF matrix method, and are shown to be quantitatively related to the distribution of Si-O distances and oxygen bond angles obtained from X-ray scattering data. The approximation commonly used to generate the optically active vibrational spectra of glasses, namely to form the product of the vibrational density of states function and a weakly frequency-dependent intensity factor gives a relatively poor representation of the experimental spectrum. The magnitude of 'disorder-induced' absorption in regions well away from the major bands is semi-quantitatively estimated by subtracting the contributions of the major bands. Interpretation of some of the features of this difference spectrum is possible in terms of vibrations of nonbridging oxygen atoms, but if this interpretation is correct, it is necessary to postulate clustering of 'dangling' oxygen atoms, which would not be consistent with a random network model for the structure. An alternative explanation, that the vibrations are framework modes, leads to the conclusion that the preferred configuration in vitreous silica resembles the arrangement of silicon-oxygen tetrahedra in cristobalite.

  20. Determination of the neutron spectrum at different locations in the Argentine RA-1 Reactor; Determinacion del espectro neutronico en distintas posiciones del reactor RA-1

    Energy Technology Data Exchange (ETDEWEB)

    Lerner, A M; Madariaga, M R [Autoridad Regulatoria Nuclear, Buenos Aires (Argentina)

    1999-12-31

    Full text: It is well known that the RA-1 reactor is used to irradiate different types of materials with neutrons. The Radio dosimetry Group (which belongs to the Nuclear Regulatory Authority) uses its fast column for the design, calibration and set up of criticality dosimeters as well as for a quick assessment of the dose to workers in case of an accident. With such purpose, Au(1), Au under Cd and In(2) foils were irradiated to estimate absolute thermal, epithermal and fast neutron fluxes at the irradiation location. The accuracy of this estimation is higher when the response to the present neutron spectrum of the different materials constituting the detectors is better known. This, in turn, requires the previous knowledge of such spectrum (a detailed energy dependence of neutron flux) at the analysed location. In this work a neutronic calculation is presented at the fast irradiation location. The whole calculation was carried out following two different methodologies, and considering a power of 40 kW. The reactor and its surroundings were represented by a simplified one-dimensional model, as a concentric cylindrical set of regions. Figures are drawn representing fast and thermal fluxes (with the cut at 0.4 eV) as a function of the distance to the core centre. The neutron flux (in n/cm{sup 2}sec.eV) as a function of energy is also shown at the fast irradiation location. Values of flux (in n/cm{sup 2}.sec.eV) are also provided as a function of energy in other typical locations, as well as the equivalent integrated flux values (in n/cm{sup 2}.sec). ((1) According to the reaction Au{sup 197}(n,{gamma})Au{sup 198}, having a cross section of {sigma}{sub 0}=98.8b for thermal neutrons. (2) According to the reaction In{sup 115}(n,n`)In{sup 115m}, with a cross section of some 70 mb for neutrons with energies above 1.2MeV). (author) [Espanol] Texto completo: Como se sabe, el reactor RA1 se utiliza para irradiar con neutrones distintos tipos de materiales. El grupo de

  1. Neutron/gamma discrimination employing the power spectrum analysis of the signal from the liquid scintillator BC501A

    International Nuclear Information System (INIS)

    Luo, X.L.; Wang, Y.K; Yang, J.; Liu, G.; Lin, C.B.; Hu, Q.Q.; Peng, J.X.

    2013-01-01

    A digital method for the discrimination of neutron and γ-ray events based on analyzing the power spectra of the signals from a BC501A liquid scintillator detector was presented and investigated in this paper. In order to evaluate the feasibility of this novel pulse shape discrimination method, a 5GSample/s 8-bit oscilloscope was used to acquire waveforms for n/γ discrimination. Furthermore, the performance of this novel n/γ discrimination method was compared with that of a widely used method called the reference-pulses method which averaged a large number of neutron and γ-ray pulses to obtain the reference-pulse as the criterion for n/γ discrimination. The results showed that the proposed method performed well over the reference-pulses method, which was verified by the considerable decrease in the error rate of n/γ discrimination and the improvement of the Figure of Merit

  2. Effects of sample and spectrum characteristics on cold and thermal neutron prompt gamma activation analysis in environmental studies of plants

    International Nuclear Information System (INIS)

    Robinson, L.; Zhao, L.

    2009-01-01

    Previous studies including the development of methods for the determination of carbon, nitrogen, and phosphorus in cattail using cold neutron prompt gamma activation (CNPGAA) and thermal neutron prompt gamma activation analysis (TNPGAA); evaluation of the precision and accuracy of these methods through the analysis of Standard Reference Materials (SRMs); and comparison of the sensitivity of CNPGAA to TNPGAA have been done in the CNPGAA and TNPGAA facilities at the National Institute of Standards and Technology (NIST). This paper integrates the findings from all of these prior studies and presents recommendations for the application of CNPGAA and TNPGAA in environmental studies of plants based on synergistic considerations of the effects of neutron energy, matrix factors such as chlorine content, Compton scattering, hydrogen content, sample thickness, and spectral interferences from Cl on the determination of C, N, and P. This paper also provides a new approach that simulates a sensitivity curve for an element of interest (S), which is a function of hydrogen content (X) and sample thickness (Y) as follows: S = aX + bY + c (where a, b, and c are constants). This approach has provided more accurate results from the analysis of SRMs than traditional methods and an opportunity to use models to optimize experimental conditions. (author)

  3. Evaluation of the cross-sections of threshold reactions leading to the production of long-lived radionuclides during irradiation of steels by thermonuclear spectrum neutrons

    CERN Document Server

    Blokhin, A I; Manokhin, V N; Mikhajlyukova, M V; Nasyrova, S M; Skripova, M V

    2001-01-01

    The present paper analyses and evaluates the cross-sections of threshold reactions leading to the production of long-lived radionuclides during the irradiation, by thermonuclear spectrum neutrons, of steels containing V, Ti, Cr, Fe and Ni. On the basis of empirical systematics. a new evaluation of the (n,2n), (n,p), (n,np), (n,alpha) and (n,n alpha) excitation functions is made for all isotopes of V, Ti, Cr, Fe and Ni and for intermediate isotopes produced in the chain from irradiated isotopes up to production of the long-lived radionuclides sup 3 sup 9 Ar, sup 4 sup 2 Ar, sup 4 sup 1 Ca, sup 5 sup 3 Mn, sup 6 sup 0 Fe, sup 6 sup 0 Co, sup 5 sup 9 Ni and sup 6 sup 3 Ni. A comparison is made with the experimental and other evaluated data.

  4. Evaluation of the cross-sections of threshold reactions leading to the production of long-lived radionuclides during irradiation of steels by thermonuclear spectrum neutrons

    International Nuclear Information System (INIS)

    Blokhin, A.I.; Buleeva, N.N.; Manokhin, V.N.; Mikhajlyukova, M.V.; Nasyrova, S.M.; Skripova, M.V.

    2002-01-01

    The present paper analyses and evaluates the cross-sections of threshold reactions leading to the production of long-lived radionuclides during the irradiation, by thermonuclear spectrum neutrons, of steels containing V, Ti, Cr, Fe and Ni. On the basis of empirical systematics. a new evaluation of the (n,2n), (n,p), (n,np), (n,α) and (n,nα) excitation functions is made for all isotopes of V, Ti, Cr, Fe and Ni and for intermediate isotopes produced in the chain from irradiated isotopes up to production of the long-lived radionuclides 39 Ar, 42 Ar, 41 Ca, 53 Mn, 60 Fe, 60 Co, 59 Ni and 63 Ni. A comparison is made with the experimental and other evaluated data. (author)

  5. Measurements of neutron spectrum from stopping-length target irradiated by several tens-MeV/u particles

    Energy Technology Data Exchange (ETDEWEB)

    Meigo, Shin-ichiro; Takada, Hiroshi; Nakashima, Hiroshi; Sasa, Toshinobu [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Tanaka, Susumu; Shin, Kazuo; Ono, Shinji

    1997-03-01

    Using a Time-of-Flight technique, we have measured neutron spectra from stopping-length targets bombarded with 68-MeV protons and 100-MeV {alpha}-particles. The measured spectra were used to validate the results calculated by the Quantum Molecular Dynamics (QMD) plus Statistical Decay Model (SDM). The results of QMD plus SDM code agreed fairly well with the experimental data for the light target. On the other hand, the QMD plus SDM gives a larger value than the experimental for the heavy target. (author)

  6. Thermal neutron polarisation

    International Nuclear Information System (INIS)

    Satya Murthy, N.S.; Madhava Rao, L.

    1984-01-01

    The basic principle for the production of polarised thermal neutrons is discussed and the choice of various crystal monochromators surveyed. Brief mention of broad-spectrum polarisers is made. The application of polarised neutrons to the study of magnetisation density distributions in magnetic crystals, the dynamic concept of polarisation, principle and use of polarisation analysis, the neutron spin-echo technique are discussed. (author)

  7. Influence of dose rate on the transformation of Syrian hamster embryo cells by fission-spectrum neutrons

    International Nuclear Information System (INIS)

    Jones, C.A.; Sedita, B.A.; Hill, C.K.; Elkind, M.M.

    1988-01-01

    Several explanations for this neutron dose-rate effect have been proposed, but further investigation is necessary to determine the mechanisms involved. In all cell transformation studies to date the immortalized, aneuploid 10T1/2 cell-line has been used. These cells may be premalignant; thus their response characteristics and, in particular, the nature of the transformation event, might differ from that in a normal, fibroblast cell. One reason for the present study was to determine whether the low-dose-rate effect of fission neutrons could be demonstrated in normal cells. If so, a normal cell system, which would more closely resemble a normal in vivo system, could be used for mechanistic studies. We chose Syrian hamster embryo (SHE) fibroblasts which are normal, diploid cells with a limited life span in culture. Upon exposure to low doses of ionizing radiation, the fraction of the cells that are transformed can be identified in a standard 8--10 day colony assay by examining their clonal morphology. Transformed cells form colonies with a dense, criss-crossed or piled-up structure. A high percentage of the transformed colonies can be further propagated and will acquire additional neoplastic characteristics; i.e., anchorage independence, immortality, altered proteolytic activity, karyotype alterations, and finally, tumorigenicity

  8. Influence of dose rate on the transformation of Syrian hamster embryo cells by fission-spectrum neutrons

    Energy Technology Data Exchange (ETDEWEB)

    Jones, C.A.; Sedita, B.A.; Hill, C.K.; Elkind, M.M.

    1988-01-01

    Several explanations for this neutron dose-rate effect have been proposed, but further investigation is necessary to determine the mechanisms involved. In all cell transformation studies to date the immortalized, aneuploid 10T1/2 cell-line has been used. These cells may be premalignant; thus their response characteristics and, in particular, the nature of the transformation event, might differ from that in a normal, fibroblast cell. One reason for the present study was to determine whether the low-dose-rate effect of fission neutrons could be demonstrated in normal cells. If so, a normal cell system, which would more closely resemble a normal in vivo system, could be used for mechanistic studies. We chose Syrian hamster embryo (SHE) fibroblasts which are normal, diploid cells with a limited life span in culture. Upon exposure to low doses of ionizing radiation, the fraction of the cells that are transformed can be identified in a standard 8--10 day colony assay by examining their clonal morphology. Transformed cells form colonies with a dense, criss-crossed or piled-up structure. A high percentage of the transformed colonies can be further propagated and will acquire additional neoplastic characteristics; i.e., anchorage independence, immortality, altered proteolytic activity, karyotype alterations, and finally, tumorigenicity.

  9. The energy spectrum of neutrons from 7Li(d,n)8Be reaction at deuteron energy 2.9 MeV

    Science.gov (United States)

    Mitrofanov, Konstantin V.; Piksaikin, Vladimir M.; Zolotarev, Konstantin I.; Egorov, Andrey S.; Gremyachkin, Dmitrii E.

    2017-09-01

    The neutron beams generated at the electrostatic accelerators using nuclear reactions T(p,n)3He, D(d,n)3He, 7Li(p,n)7Be, T(d,n)4He, 7Li(d,n)8Be, 9Be(d,n)10B are widely used in neutron physics and in many practical applications. Among these reactions the least studied reactions are 7Li(d,n)8Be and 9Be(d,n)10B. The present work is devoted to the measurement of the neutron spectrum from 7Li(d,n)8Be reaction at 0∘ angle to the deuteron beam axis on the electrostatic accelerator Tandetron (JSC "SSC RF - IPPE") using activation method and a stilbene crystal scintillation detector. The first time ever 7Li(d,n)8Be reaction was measured by activation method. The target was a thick lithium layer on metallic backing. The energy of the incident deuteron was 2.9 MeV. As activation detectors a wide range of nuclear reactions were used: 27Al(n,p)27Mg, 27Al(n,α)24Na, 113In(n,n')113mIn, 115In(n,n')115mIn, 115In(n,γ)116mIn, 58Ni(n,p)58mCo, 58Ni(n,2n)57Ni, 197Au(n,γ)198Au, 197Au(n,2n)196Au, 59Co(n,p)59Fe, 59Co(n,2n)58m+gCo, 59Co (n,g)60Co. Measurement of the induced gamma-activity was carried out using HPGe detector Canberra GX5019 [1]. The up-to-date evaluations of the cross sections for these reactions were used in processing of the data. The program STAYSL was used to unfold the energy spectra. The neutron spectra obtained by activation detectors is consistent with the corresponding data measured by a stilbene crystal scintillation detector within their uncertainties.

  10. The energy spectrum of neutrons from 7Li(d,n8Be reaction at deuteron energy 2.9 MeV

    Directory of Open Access Journals (Sweden)

    Mitrofanov Konstantin V.

    2017-01-01

    Full Text Available The neutron beams generated at the electrostatic accelerators using nuclear reactions T(p,n3He, D(d,n3He, 7Li(p,n7Be, T(d,n4He, 7Li(d,n8Be, 9Be(d,n10B are widely used in neutron physics and in many practical applications. Among these reactions the least studied reactions are 7Li(d,n8Be and 9Be(d,n10B. The present work is devoted to the measurement of the neutron spectrum from 7Li(d,n8Be reaction at 0∘ angle to the deuteron beam axis on the electrostatic accelerator Tandetron (JSC “SSC RF – IPPE” using activation method and a stilbene crystal scintillation detector. The first time ever 7Li(d,n8Be reaction was measured by activation method. The target was a thick lithium layer on metallic backing. The energy of the incident deuteron was 2.9 MeV. As activation detectors a wide range of nuclear reactions were used: 27Al(n,p27Mg, 27Al(n,α24Na, 113In(n,n'113mIn, 115In(n,n'115mIn, 115In(n,γ116mIn, 58Ni(n,p58mCo, 58Ni(n,2n57Ni, 197Au(n,γ198Au, 197Au(n,2n196Au, 59Co(n,p59Fe, 59Co(n,2n58m+gCo, 59Co (n,g60Co. Measurement of the induced gamma-activity was carried out using HPGe detector Canberra GX5019 [1]. The up-to-date evaluations of the cross sections for these reactions were used in processing of the data. The program STAYSL was used to unfold the energy spectra. The neutron spectra obtained by activation detectors is consistent with the corresponding data measured by a stilbene crystal scintillation detector within their uncertainties.

  11. Neutron dynamics of fast-spectrum dedicated cores for waste transmutation; Etude et amelioration du comportement cinetique de coeurs rapides a la transmutation de dechets a vie longue

    Energy Technology Data Exchange (ETDEWEB)

    Massara, S

    2002-04-01

    Among different scenarios achieving minor actinide transmutation, the possibility of double strata scenarios with critical, fast spectrum, dedicated cores must be checked and quantified. In these cores, the waste fraction has to be at the highest level compatible with safety requirements during normal operation and transient conditions. As reactivity coefficients are poor in such critical cores (low delayed neutron fraction and Doppler feed-back, high coolant void coefficient), their dynamic behaviour during transient conditions must be carefully analysed. Three nitride-fuel configurations have been analysed: two liquid metal-cooled (sodium and lead) and a particle-fuel helium-cooled one. A dynamic code, MAT4 DYN, has been developed during the PhD thesis, allowing the study of loss of flow, reactivity insertion and loss of coolant accidents, and taking into account two fuel geometries (cylindrical and spherical) and two thermal-hydraulics models for the coolant (incompressible for liquid metals and compressible for helium). Dynamics calculations have shown that if the fuel nature is appropriately chosen (letting a sufficient margin during transients), this can counterbalance the bad state of reactivity coefficients for liquid metal-cooled cores, thus proving the interest of this kind of concept. On the other side, the gas-cooled core dynamics is very badly affected by the high value of the helium void coefficient (which is a consequence of the choice of a hard spectrum), this effect being amplified by the very low thermal inertia of particle-fuel design. So, a new kind of concept should be considered for a helium-cooled fast-spectrum dedicated core. (authors)

  12. Effect of updated WIMSD libraries on neutron energy spectrum at irradiation site of Pakistan Research Reactor-1 using 3D modeling

    International Nuclear Information System (INIS)

    Ahmad, Siraj-ul-Islam; Ahmad, Nasir

    2005-01-01

    International Atomic Energy Agency (IAEA) has recently released new WIMSD libraries based on current cross-section evaluations. Using these libraries the effect of different evaluated data sets on effective multiplication factor and neutron energy spectrum was studied with the help of 3D reactor simulation code CITATION. Simulation methodology adopted in this work was validated by analyzing IAEA 10 MW benchmark reactor. The k eff values obtained using all newly released libraries are within 0.45% to the experimental value, while the old library released in 1981 resulted in calculated value 1.05% larger than experimental. The flux spectrum obtained for standard fuel element using 3D modeling is smaller in fast energy range and higher in thermal energy range than is calculated using the 1D model for the standard cell. In the flux trap, differences of about -4% to 13% were found in thermal flux using the newly released libraries as compared to that obtained using 1981 WIMSD library. The major differences in the flux spectra between newly available libraries and the 1981 WIMSD library in thermal energy range are due to the differences in cross-sections of hydrogen bound-in-water. The use of only newly available cross-sections of hydrogen bound-in-water with 1981 WIMSD library resulted in significant improvement in value of k eff as well as in the flux spectrum. Moreover the differences among new libraries in the thermal energy range are also due to these cross-sections. Difference in fission spectra from different libraries is responsible for differences of flux spectra in the fast energy range. These differences in flux are reduced significantly in the fast energy range by only replacement of fission spectra

  13. Measurement of the 235U/238U fission cross section ratio in the 235U fission neutron spectrum

    International Nuclear Information System (INIS)

    Azimi-Garakani, D.; Bagheri-Darbandi, M.

    1983-06-01

    Fission cross section ratio of 235 U to 238 U has been measured in the fast neutron field generated by the 235 U fission plate installed on the thermal column of the Tehran Research Reactor (TRR) with a Makrofol solid state nuclear track detector. The experiments were carried out with a set of total six enriched 235 U and depleted 238 U deposits with different masses and Makrofol films of 0.025mm and 0.060mm thicknesses. The chemically etched tracks were counted by an optical microscope. No significant differences were observed with the thin and the thick films. The results showed that the average fission cross section ratio is 3.83+-0.25. (author)

  14. Application of evaporation model to the calculation of energy spectrum and angular distribution of recoil nuclei from neutron induced reaction

    Energy Technology Data Exchange (ETDEWEB)

    Takeda, Y; Sugimoto, M; Sugiyama, K [Tohoku Univ., Sendai (Japan). Faculty of Engineering

    1978-12-01

    Calculated angular distributions and energy spectra from 14.8 MeV neutron induced (n,2n) reactions based on a simple evaporation model were obtained by means of the Monte Carlo method. It was ascertained that the effects on the spectra of the method of determining the nuclear temperature and the value of the level density parameter are much smaller than those of the reaction Q-value and the nuclear mass. As a check on the calculational procedure, results of similar calculations were compared with the experimental recoil escape efficiency for /sup 27/Al(n,..cap alpha..)/sup 24/Na reaction. Distortions of the energy spectra in thick target materials were also obtained. These results suggest that this model is fully applicable to the calculation of primary knock-on atoms distributions from various nuclear reactions.

  15. Neutron dosimetry; Dosimetria de neutrons

    Energy Technology Data Exchange (ETDEWEB)

    Fratin, Luciano

    1993-12-31

    A neutron irradiation facility was designed and built in order to establish a procedure for calibrating neutron monitors and dosemeters. A 185 GBq {sup 241} Am Be source of known is used as a reference source. The irradiation facility using this source in the air provides neutron dose rates between 9 nSv s{sup -1} and 0,5 {sup {mu}}Sv s{sup -1}. A calibrated 50 nSv s{sup -1} thermal neutron field is obtained by using a specially designed paraffin block in conjunction with the {sup 241} Am Be source. A Bonner multisphere spectrometer was calibrated, using a procedure based on three methods proposed by international standards. The unfold {sup 241} Am Be neutron spectrum was determined from the Bonner spheres data and resulted in a good agreement with expected values for fluence rate, dose rate and mean energy. A dosimetric system based on the electrochemical etching of CR-39 was developed for personal dosimetry. The dosemeter badge using a (n,{alpha}) converter, the etching chamber and high frequency power supply were designed and built specially for this project. The electrochemical etching (ECE) parameters used were: a 6N KOH solution, 59 deg C, 20 kV{sub pp} cm{sup -1}, 2,0 kHz, 3 hours of ECE for thermal and intermediate neutrons and 6 hours for fast neutrons. The calibration factors for thermal, intermediate and fast neutrons were determined for this personal dosemeter. The sensitivities determined for the developed dosimetric system were (1,46{+-} 0,09) 10{sup 4} tracks cm{sup -2} mSv{sup -1} for thermal neutrons, (9{+-}3) 10{sup 2} tracks cm{sup -2} mSV{sup -1} for intermediate neutrons and (26{+-}4) tracks cm{sup -2} mSv{sup -1} for fast neutrons. The lower and upper limits of detection were respectively 0,002 mSv and 0,6 mSv for thermal neutrons, 0,04 mSv and 8 mSv for intermediate neutrons and 1 mSv and 12 mSv for fast neutrons. In view of the 1990`s ICRP recommendations, it is possible to conclude that the personal dosemeter described in this work is

  16. Neutron dosimetry; Dosimetria de neutrons

    Energy Technology Data Exchange (ETDEWEB)

    Fratin, Luciano

    1994-12-31

    A neutron irradiation facility was designed and built in order to establish a procedure for calibrating neutron monitors and dosemeters. A 185 GBq {sup 241} Am Be source of known is used as a reference source. The irradiation facility using this source in the air provides neutron dose rates between 9 nSv s{sup -1} and 0,5 {sup {mu}}Sv s{sup -1}. A calibrated 50 nSv s{sup -1} thermal neutron field is obtained by using a specially designed paraffin block in conjunction with the {sup 241} Am Be source. A Bonner multisphere spectrometer was calibrated, using a procedure based on three methods proposed by international standards. The unfold {sup 241} Am Be neutron spectrum was determined from the Bonner spheres data and resulted in a good agreement with expected values for fluence rate, dose rate and mean energy. A dosimetric system based on the electrochemical etching of CR-39 was developed for personal dosimetry. The dosemeter badge using a (n,{alpha}) converter, the etching chamber and high frequency power supply were designed and built specially for this project. The electrochemical etching (ECE) parameters used were: a 6N KOH solution, 59 deg C, 20 kV{sub pp} cm{sup -1}, 2,0 kHz, 3 hours of ECE for thermal and intermediate neutrons and 6 hours for fast neutrons. The calibration factors for thermal, intermediate and fast neutrons were determined for this personal dosemeter. The sensitivities determined for the developed dosimetric system were (1,46{+-} 0,09) 10{sup 4} tracks cm{sup -2} mSv{sup -1} for thermal neutrons, (9{+-}3) 10{sup 2} tracks cm{sup -2} mSV{sup -1} for intermediate neutrons and (26{+-}4) tracks cm{sup -2} mSv{sup -1} for fast neutrons. The lower and upper limits of detection were respectively 0,002 mSv and 0,6 mSv for thermal neutrons, 0,04 mSv and 8 mSv for intermediate neutrons and 1 mSv and 12 mSv for fast neutrons. In view of the 1990`s ICRP recommendations, it is possible to conclude that the personal dosemeter described in this work is

  17. Neutron Beam Filters

    International Nuclear Information System (INIS)

    Adib, M.

    2011-01-01

    The purpose of filters is to transmit neutrons with selected energy, while remove unwanted ones from the incident neutron beam. This reduces the background, and the number of spurious. The types of commonly used now-a-day neutron filters and their properties are discussed in the present work. There are three major types of neutron filters. The first type is filter of selective thermal neutron. It transmits the main reflected neutrons from a crystal monochromate, while reject the higher order contaminations accompanying the main one. Beams coming from the moderator always contain unwanted radiation like fast neutrons and gamma-rays which contribute to experimental background and to the biological hazard potential. Such filter type is called filter of whole thermal neutron spectrum. The third filter type is it transmits neutrons with energies in the resonance energy range (En . 1 KeV). The main idea of such neutron filter technique is the use of large quantities of a certain material which have the deep interference minima in its total neutron cross-section. By transmitting reactor neutrons through bulk layer of such material, one can obtain the quasimonochromatic neutron lines instead of white reactor spectrum.

  18. Investigation of Response of Several Neutron Surveymeters by a DT Neutron Generator

    International Nuclear Information System (INIS)

    Kim, Sang In; Jang, In Su; Kim, Jang Lyul; Lee, Jung IL; Kim, Bong Hwan

    2012-01-01

    Several neutron measuring devices were tested under the neutron fields characterized with two distinct kinds of thermal and fast neutron spectrum. These neutron fields were constructed by the mixing of both thermal neutron fields and fast neutron fields. The thermal neutron field was constructed using by a graphite pile with eight AmBe neutron sources. The fast neutron field of 14 MeV was made by a DT neutron generator. In order to change the fraction of fast neutron fluence rate in each neutron fields, a neutron generator was placed in the thermal neutron field at 50 cm and 150 cm from the reference position. The polyethylene neutron collimator was used to make moderated 14 MeV neutron field. These neutron spectra were measured by using a Bonner sphere system with an LiI scintillator, and dosimetric quantities delivered to neutron surveymeters were determined from these measurement results.

  19. Neutrons from Antiproton Irradiation

    DEFF Research Database (Denmark)

    Bassler, Niels; Holzscheiter, Michael; Petersen, Jørgen B.B.

    the neutron spectrum. Additionally, we used a cylindrical polystyrene loaded with several pairs of thermoluminescent detectors containing Lithium-6 and Lithium-7, which effectively detects thermalized neutrons. The obtained results are compared with FLUKA imulations. Results: The results obtained...... spectrum is very low, and does not pose a problem for radiation therapy. However, the contribution from fast neutrons is much more significant. The dose equivalent contribution from neutrons originate from the patient alone and reaches levels which are found in passive moderated proton therapy. The exact...

  20. Effect of thermal 3He minorities on knock-on tail formation and the resulting neutron emission spectrum modification in deuterium-tritium plasmas

    International Nuclear Information System (INIS)

    Matsuura, H; Nakao, Y; Nakamura, M; Mitarai, O

    2011-01-01

    The knock-on tail formations in fuel-ion velocity distribution functions by energetic alpha particles (by the T(d,n) 4 He reaction) and protons (by the D(d,p)T and 3 He(d,p) 4 He reactions) are investigated by simultaneously solving the Boltzmann-Fokker-Planck (BFP) equations for deuteron, triton, 3 He, alpha particle and proton in an ITER-like (DT) plasma admixed with a small amount of 3 He. As a result of the 3 He inclusion, a fraction of the transferred energy from energetic ions to thermal deuterons and tritons via nuclear plus interference (NI) scattering is reduced. Owing to the NI scattering of the energetic protons by fuel ions, the latter are knocked up to higher energies. The knocking-up effect of fuel ions is enhanced with increasing 3 He concentration. It is shown that if 3 He with relative concentration of 4.2%, i.e. n 3 He /n e =0.042, is included in T e = 20 keV, n e = 9.5 x 10 19 m -3 plasma, the magnitude of the knock-on tail in deuteron distribution function in 300 keV-3 MeV energy range is reduced by about 15% from the value when 3 He is not externally supplied. Such knock-on tail reduction also results in alternation of the non-Gaussian neutron emission spectrum with energies less than ∼13 MeV and above ∼15 MeV.

  1. Study of neutron moderation using the {sup 241}Am-Be spectrum with hydrogenated materials; Estudo da moderacao de neutrons utilizando o espectro de {sup 241}Am-Be com materiais hidrogenados

    Energy Technology Data Exchange (ETDEWEB)

    Santos, A.R.L.; Silva, F.S.; Martins, M.M.; Pereira, W.W., E-mail: aleiras@ird.gov.br [Instituto de Radioprotecao e Dosimetria (IRD/CNEN-RJ/LNMRI/LN), Rio de Janeiro, RJ (Brazil). Lab. de Neutrons; Freitas, B.M. [Coordenacao dos Programas de Pos-Graduacao em Engenharia (COPPE/UFRJ), Rio de Janeiro, RJ (Brazil). Programa de Engenharia Nuclear; Tavares, D.Y.S. [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

    2014-07-01

    This work intends to assess materials for moderation of neutrons, trying to reduce the rate of H{sub p}(10) and H⁎p(10), reducing the effective dose of Occupationally Exposed Workers (OEW) who handle this source daily. The neutron spectra moderated by different materials was performed with a neutron source of {sup 241}Am-Be in an electronic positioning system, using a neutron spectrometry with Bonner Sphere at 50 cm from the center of source. The materials used for moderation were paraffin, silicone and Polyvinyl Chloride (PVC) resin ball. (author)

  2. Neutron stars

    International Nuclear Information System (INIS)

    Irvine, J.M.

    1978-01-01

    The subject is covered in chapters entitled: introduction (resume of stellar evolution, gross characteristics of neutron stars); pulsars (pulsar characteristics, pulsars as neutron stars); neutron star temperatures (neutron star cooling, superfluidity and superconductivity in neutron stars); the exterior of neutron stars (the magnetosphere, the neutron star 'atmosphere', pulses); neutron star structure; neutron star equations of state. (U.K.)

  3. Neutron Spectrum Measured by Activation Diagnostics in Deuterium Gas-Puff Experiments on the 3 MA GIT-12 Z-Pinch

    Czech Academy of Sciences Publication Activity Database

    Cikhardt, J.; Klír, D.; Řezáč, K.; Shishlov, A. V.; Cherdizov, R. K.; Cikhardtová, B.; Dudkin, G. N.; Fursov, F. I.; Kokshenev, V. A.; Kravařík, J.; Kubeš, P.; Kurmaev, N. E.; Labetsky, A. Yu.; Padalko, V. N.; Ratakhin, N. A.; Šíla, O.; Turek, Karel; Varlachev, V. A.

    2017-01-01

    Roč. 45, č. 12 (2017), s. 3209-3217 ISSN 0093-3813 Institutional support: RVO:61389005 Keywords : neutron detectors * neutron spectroscopy * plasma pinch Subject RIV: BL - Plasma and Gas Discharge Physics OBOR OECD: Fluids and plasma physics (including surface physics) Impact factor: 1.052, year: 2016

  4. 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

  5. Atlas of neutron resonances

    CERN Document Server

    Mughabghab, Said

    2018-01-01

    Atlas of Neutron Resonances: Resonance Properties and Thermal Cross Sections Z= 1-60, Sixth Edition, contains an extensive list of detailed individual neutron resonance parameters for Z=1-60, as well as thermal cross sections, capture resonance integrals, average resonance parameters and a short survey of the physics of thermal and resonance neutrons. The long introduction contains: nuclear physics formulas aimed at neutron physicists; topics of special interest such as valence neutron capture, nuclear level density parameters, and s-, p-, and d-wave neutron strength functions; and various comparisons of measured quantities with the predictions of nuclear models, such as the optical model. As in the last edition, additional features have been added to appeal to a wider spectrum of users. These include: spin-dependent scattering lengths that are of interest to solid-state physicists, nuclear physicists and neutron evaluators; calculated and measured Maxwellian average 5-keV and 30-keV capture cross sections o...

  6. Methods of neutron spectrometry

    International Nuclear Information System (INIS)

    Doerschel, B.

    1981-01-01

    The different methods of neutron spectrometry are based on the direct measurement of neutron velocity or on the use of suitable energy-dependent interaction processes. In the latter case the measuring effect of a detector is connected with the searched neutron spectrum by an integral equation. The solution needs suitable unfolding procedures. The most important methods of neutron spectrometry are the time-of-flight method, the crystal spectrometry, the neutron spectrometry by use of elastic collisions with hydrogen nuclei, and neutron spectrometry with the aid of nuclear reactions, especially of the neutron-induced activation. The advantages and disadvantages of these methods are contrasted considering the resolution, the measurable energy range, the sensitivity, and the experimental and computational efforts. (author)

  7. Neutron response study

    International Nuclear Information System (INIS)

    Endres, G.W.R.; Fix, J.J.; Thorson, M.R.; Nichols, L.L.

    1981-01-01

    Neutron response of the albedo type dosimeter is strongly dependent on the energy of the incident neutrons as well as the moderating material on the backside of the dosimeter. This study characterizes the response of the Hanford dosimeter for a variety of neutron energies for both a water and Rando phantom (a simulated human body consisting of an actual human skeleton with plastic for body muscles and certain organs). The Hanford dosimeter response to neutrons of different energies is typical of albedo type dosimeters. An approximate two orders of magnitude difference in response is observed between neutron energies of 100 keV and 10 MeV. Methods were described to compensate for the difference in dosimeter response between a laboratory neutron spectrum and the different spectra encountered at various facilities in the field. Generally, substantial field support is necessary for accurate neutron dosimetry

  8. Prototype Neutron Energy Spectrometer

    International Nuclear Information System (INIS)

    Mitchell, Stephen; Mukhopadhyay, Sanjoy; Maurer, Richard; Wolff, Ronald

    2010-01-01

    The project goals are: (1) Use three to five pressurized helium tubes with varying polyethylene moderators to build a neutron energy spectrometer that is most sensitive to the incident neutron energy of interest. Neutron energies that are of particular interest are those from the fission neutrons (typically around 1-2 MeV); (2) Neutron Source Identification - Use the neutron energy 'selectivity' property as a tool to discriminate against other competing processes by which neutrons are generated (viz. Cosmic ray induced neutron production (ship effect), (a, n) reactions); (3) Determine the efficiency as a function of neutron energy (response function) of each of the detectors, and thereby obtain the composite neutron energy spectrum from the detector count rates; and (4) Far-field data characterization and effectively discerning shielded fission source. Summary of the presentation is: (1) A light weight simple form factor compact neutron energy spectrometer ready to be used in maritime missions has been built; (2) Under laboratory conditions, individual Single Neutron Source Identification is possible within 30 minutes. (3) Sources belonging to the same type of origin viz., (a, n), fission, cosmic cluster in the same place in the 2-D plot shown; and (4) Isotopes belonging to the same source origin like Cm-Be, Am-Be (a, n) or Pu-239, U-235 (fission) do have some overlap in the 2-D plot.

  9. Prototype Neutron Energy Spectrometer

    Energy Technology Data Exchange (ETDEWEB)

    Stephen Mitchell, Sanjoy Mukhopadhyay, Richard Maurer, Ronald Wolff

    2010-06-16

    The project goals are: (1) Use three to five pressurized helium tubes with varying polyethylene moderators to build a neutron energy spectrometer that is most sensitive to the incident neutron energy of interest. Neutron energies that are of particular interest are those from the fission neutrons (typically around 1-2 MeV); (2) Neutron Source Identification - Use the neutron energy 'selectivity' property as a tool to discriminate against other competing processes by which neutrons are generated (viz. Cosmic ray induced neutron production [ship effect], [a, n] reactions); (3) Determine the efficiency as a function of neutron energy (response function) of each of the detectors, and thereby obtain the composite neutron energy spectrum from the detector count rates; and (4) Far-field data characterization and effectively discerning shielded fission source. Summary of the presentation is: (1) A light weight simple form factor compact neutron energy spectrometer ready to be used in maritime missions has been built; (2) Under laboratory conditions, individual Single Neutron Source Identification is possible within 30 minutes. (3) Sources belonging to the same type of origin viz., (a, n), fission, cosmic cluster in the same place in the 2-D plot shown; and (4) Isotopes belonging to the same source origin like Cm-Be, Am-Be (a, n) or Pu-239, U-235 (fission) do have some overlap in the 2-D plot.

  10. Measurement of the energy spectrum with proportional counters with spherical cathodes between 20 keV and 2.5 MeV with the propagation of 14 MeV neutrons in liquid nitrogen and liquid air

    International Nuclear Information System (INIS)

    Schneider-Kuehnle, P.

    1974-01-01

    This work deals with the measurement of the energy spectrum of a 14 MeV neutron source in liquid nitrogen and liquid air in the energy region of 20 keV to 2.5 MeV as a function of the distance from the source. The measured results together with those of a scintispectrometer which measures the energies between 2.5 MeV and 14 MeV, are to serve as experimentally-supported input data for shielding calculations and are to enable the checking of transport theoretical calculations. (orig./LH) [de

  11. Summary of neutron measurements for the Viking Program

    International Nuclear Information System (INIS)

    Anderson, M.E.

    1975-01-01

    The results of neutron measurements for 238 Pu-fueled, 683-W (thermal) capsules fabricated for the Viking Program (Mars Lander) are presented. These results include, for each capsule, the total neutron emission rate and neutron multiplication and, for one capsule, the neutron energy spectrum. A precision long counter was used for the neutron emission rate measurements and a single stilbene crystal for the neutron spectrum measurement. (U.S.)

  12. X-ray and fast neutron-induced mutations in Arabidopsis thaliana, and the effect of dithiothreitol upon the mutant spectrum

    International Nuclear Information System (INIS)

    Dellaert, L.M.W.

    1980-01-01

    The author discusses the genetic effects of X-ray and fast neutron seed-irradiation of Arabidopsis thaliana (L.) Heynh., and the influence of a pre-irradiation treatment with the radio-protector dithiothreitol (DTT). (Auth.)

  13. Neutron activation studies on JET

    International Nuclear Information System (INIS)

    Loughlin, M.J.; Forrest, R.A.; Edwards, J.E.G.

    2001-01-01

    Extensive neutron transport calculations have been performed to determine the neutron spectrum at a number of points throughout the JET torus hall. The model has been bench-marked against a set of foil activation measurements which were activated during an experimental campaign with deuterium/tritium plasmas. The model can predict the neutron activation of the foils on the torus hall walls to within a factor of three for reactions with little sensitivity to thermal neutrons. The use of scandium foils with and without a cadmium thermal neutron absorber was a useful monitor of the thermal neutron flux. Conclusions regarding the usefulness of other foils for benchmarking the calculations are also given

  14. Characterization and MCNP simulation of neutron energy spectrum shift after transmission through strong absorbing materials and its impact on tomography reconstructed image.

    Science.gov (United States)

    Hachouf, N; Kharfi, F; Boucenna, A

    2012-10-01

    An ideal neutron radiograph, for quantification and 3D tomographic image reconstruction, should be a transmission image which exactly obeys to the exponential attenuation law of a monochromatic neutron beam. There are many reasons for which this assumption does not hold for high neutron absorbing materials. The main deviations from the ideal are due essentially to neutron beam hardening effect. The main challenges of this work are the characterization of neutron transmission through boron enriched steel materials and the observation of beam hardening. Then, in our work, the influence of beam hardening effect on neutron tomographic image, for samples based on these materials, is studied. MCNP and FBP simulation are performed to adjust linear attenuation coefficients data and to perform 2D tomographic image reconstruction with and without beam hardening corrections. A beam hardening correction procedure is developed and applied based on qualitative and quantitative analyses of the projections data. Results from original and corrected 2D reconstructed images obtained shows the efficiency of the proposed correction procedure. Copyright © 2012 Elsevier Ltd. All rights reserved.

  15. 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

  16. In Vivo and Air Dosimetry of Fission-Spectrum Neutrons; Dosimetrie In Vivo et dans l'Air du Spectre des Neutrons de Fission; Dozimetriya v vozdukhe i dozimetriya In Vivo nejtronov spektra deleniya; Dosimetria In Vivo y en el Aire de Neutrones de un Espectro de Fision

    Energy Technology Data Exchange (ETDEWEB)

    Mobley, T. S.; Engel, R. E.; Godden, W. R. [Kirtland Airforce Base, New Mexico (United States); Penikas, V. T. [AFIT, Wright-Patterson Air Force Base, Ohio, with Duty Station at University of Rochester, School of Medicine and Dentistry, Rochester, NY (United States)

    1964-03-15

    In an attempt to estimate the depth-dose pattern during exposure to fission-spectrum neutrons, a method of in vivo neutron dosimetry in the sheep has been developed. A pulsed-reactor assembly similar to the Los Alamos Scientific Laboratory's GODIVA II was used for these studies. The reactor was pulse operated producing a 50-{mu}s pulse width at half maximum pulse height and a fuel-material temperature rise of about 100 Degree-Sign C. A subcutaneous pack contains a cadmium-covered gold foil, a bare gold foil,and three thermoluminescent glass-rod dosimeters in a glass-lined lithium-lead container. The packs were placed in the subcutaneous tissue on both sides of the body to detect the integrated thermal-neutron flux and gamma dose as the neutrons are delivered ''broadside''. Abdominal packs are prepared in a manner similar to the subcutaneous ones and consist of a boron ball containing cadmium covered fission threshold foils (Pu, Np, U) a sulphur tablet, and the three types of dosimeters included in the subcutaneous pack. The dosimetry pack was placed in the abdomen of the experimental animals eighteen hours before exposure. The pack is attached to the anterior pillar of the rumen. A Rumenotomy technique is described. The neutron dose was determined by means of threshold-foil measurements. Gold and cadmium-covered gold were used for determining the integrated thermal flux; plutonium-239, neptunium-237, uranium-238 and sulphur-32 for determining the integrated fast-neutron fluxes; and radiophotoluminescent glass rods for determining the gamma dose. The sheep received lateral exposures at a midline distance of two hundred centimeters or one hundred seventy five centimeters from the centre of the critical assembly. The neutron rad dose measured in air at 200 cm was 161 {+-} 5.5 rad; at 175 cm the neutron dose was 242 {+-} 16.6 rad. The gamma dose at the respective distances were 33 {+-}2.4 and 42 {+-} 2.5 r. The entrance and exit doses as well as the dose to the

  17. Study of the effect of the energy spectrum and of the total flux on the damage produced by neutrons in solids; Contribution a l'etude de l'influence du spectre et du flux integre sur les dommages crees par les neutrons dans les solides

    Energy Technology Data Exchange (ETDEWEB)

    Dulieu, P C [Commissariat a l' Energie Atomique, Fontenay-aux-Roses (France). Centre d' Etudes Nucleaires

    1967-01-01

    In the first part are studied the general relationships between the physical effects produced by neutrons in solids, and the total flux and neutron energy spectrum; some examples are given. The second part, describes the application to a silicon damage detector whose principle is to use the damage produced in a PIN Junction for measuring the neutron flux (intermediate and fast) received by the detector. Chapter I is devoted to the experimental determination of the energy given to the atoms by a primary in the silicon. The results and conclusions drawn from this determination make it possible to consider, in chapter II, the calculation of the detectors response characteristics.Chapter III deals with the measurement of the detectors response function and it is noted that good agreement is obtained between the calculation and experimental results. The whole of the second part constitutes a test of the methods presented in the first part. (author) [French] Dans la premiere partie, nous etudions les relations generales qui lient les effets physiques engendres par les neutrons dans les solides au flux integre et au spectre des neutrons et nous donnons des exemples d'utilisation. La deuxieme partie est une application au detecteur de dommages en silicium, dont le principe est d'utiliser les dommages crees dans une jonction PIN pour mesurer les flux de neutrons (intermediaires et rapides) recus par le detecteur. Le chapitre I est consacre a la determination experimentale de l'energie cedee aux atomes par un primaire dans le silicium. Les resultats et les conclusions que l'on peut tirer de cette determination permettent d'aborder, au chapitre II, le calcul de la fonction de reponse du detecteur. Le chapitre III porte sur la mesure de la fonction de reponse du detecteur et on constate qu'il y a un bon accord entre le calcul et l'experience. L'ensemble de la deuxieme partie constitue un test des methodes exposees dans la premiere partie. (auteur)

  18. ATR neutron spectral characterization

    Energy Technology Data Exchange (ETDEWEB)

    Rogers, J.W.; Anderl, R.A.

    1995-11-01

    The Advanced Test Reactor (ATR) at INEL provides intense neutron fields for irradiation-effects testing of reactor material samples, for production of radionuclides used in industrial and medical applications, and for scientific research. Characterization of the neutron environments in the irradiation locations of the ATR has been done by means of neutronics calculations and by means of neutron dosimetry based on the use of neutron activation monitors that are placed in the various irradiation locations. The primary purpose of this report is to present the results of an extensive characterization of several ATR irradiation locations based on neutron dosimetry measurements and on least-squares-adjustment analyses that utilize both neutron dosimetry measurements and neutronics calculations. This report builds upon the previous publications, especially the reference 4 paper. Section 2 provides a brief description of the ATR and it tabulates neutron spectral information for typical irradiation locations, as derived from the more historical neutron dosimetry measurements. Relevant details that pertain to the multigroup neutron spectral characterization are covered in section 3. This discussion includes a presentation on the dosimeter irradiation and analyses and a development of the least-squares adjustment methodology, along with a summary of the results of these analyses. Spectrum-averaged cross sections for neutron monitoring and for displacement-damage prediction in Fe, Cr, and Ni are given in section 4. In addition, section4 includes estimates of damage generation rates for these materials in selected ATR irradiation locations. In section 5, the authors present a brief discussion of the most significant conclusions of this work and comment on its relevance to the present ATR core configuration. Finally, detailed numerical and graphical results for the spectrum-characterization analyses in each irradiation location are provided in the Appendix.

  19. Pathogenesis of Intrapulmonary Haemorrhage in Dogs Exposed to Pulsed Fission-Spectrum Neutrons; Pathogenese de l'Hemorragie Intrapulmonaire chez des Chiens Exposes a des Neutrons de Fission Pulses; Patogenez vnutrilegochnykh krovoizliyanij u sobak pri obluchenii impul'snymi nejtronami spektra deleniya; Patogenesis de la Hemorragia Intrapulmonar en Perros Expuestos a Neutrones de Fision Pulsados

    Energy Technology Data Exchange (ETDEWEB)

    Jones, R. K.; Engel, R. E.; Godden, W. R. [Kirtland AFB, New Mexico (United States)

    1964-05-15

    Previous comparison in this laboratory of tissue alterations in dogs produced by exposure to 250 kVp X-rays and pulsed fission-spectrum neutrons has shown general similarity of response. However, the lungs of animals exposed to fission-spectrum neutrons are considerably more susceptible to peribronchial and perivascular pulmonary haemorrhage. An attempt has been made to define the pathogenesis of this lesion by exposing five beagles to 400 rad (MAD) fast neutrons from a bare GODIVA-type reactor and serially sacrificing the animals at periodic intervals. Gross pathologic alterations consisting of occasional small peripheral-pulmonary haemorrhages were discernible at +5 d post-irradiation. These became progressively more numerous by the ninth day. The most striking changes occurred at +13 d, consisting of extensive perivascular haemorrhage around pulmonary-arterial branches from the hilum to the pleural suriace. The earliest histologic changes observed in the lungs occurred at +5 d, consisting of focal well-demarcated areas of alveolar congestion, swelling of capillary-endothelial cells and slight alveolar extravasation of red blood cells. These changes became more generalized by + 9 d and in addition there were focal areas of peribronchial extravasation of blood involving the right and left main-stem bronchi. Animals necropsied on the thirteenth day post-irradiation revealed extensive perivascular haemorrhage involving large and small pulmonaryarterial vessels. The blood appeared within the advential-connective tissue and periarterial-lymphatic spaces. Endothelial swelling was extremely marked in affected vessels, but the media appeared essentially unremarkable. Peribronchial haemorrhage was also present adjacent to affected pulmonary-arterial branches. The pathogenesis of these alterations appears to be related to primary endothelial damage produced by pulsed fission-spectrum neutrons. An additional contributing factor is probably thrombocytopenia as the number of

  20. Neutron converter at reactor RB; Konvertor neutrona na reaktoru RB

    Energy Technology Data Exchange (ETDEWEB)

    Strugar, P; Sotic, P; Ninkovic, M; Pesic, M [Boris Kidric Institute of nuclear sciences, Vinca, Belgrade (Yugoslavia)

    1977-07-01

    A neutron converter at Reactor RB in the 'Boris Kidric' Institute of Nuclear Sciences - Vinca has been constructed. Preliminary measurements have been shown that the converted neutron spectrum is very similar to the fission neutron spectrum. For the same integral reactor power, the measured neutron radiation dose has been for about ten times larger with the neutron converter. The neutron converter offers wide possibilities, as in investigations in the reactor physics, where the fission neutron spectra have been required, as well as in the field of neutron dosimetry and biological irradiations (author)

  1. Study on the dose distribution of the mixed field with thermal and epi-thermal neutrons for neutron capture therapy

    International Nuclear Information System (INIS)

    Kobayashi, Tooru; Sakurai, Yoshinori; Kanda, Keiji

    1994-01-01

    Simulation calculations using DOT 3.5 were carried out in order to confirm the characteristics of depth-dependent dose distribution in water phantom dependent on incident neutron energy. The epithermal neutrons mixed to thermal neutron field is effective improving the thermal neutron depth-dose distribution for neutron capture therapy. A feasibility study on the neutron energy spectrum shifter was performed using ANISN-JR for the KUR Heavy Water Facility. The design of the neutron spectrum shifter is feasible, without reducing the performance as a thermal neutron irradiation field. (author)

  2. Polarized neutrons

    International Nuclear Information System (INIS)

    Williams, W.G.

    1988-01-01

    The book on 'polarized neutrons' is intended to inform researchers in condensed matter physics and chemistry of the diversity of scientific problems that can be investigated using polarized neutron beams. The contents include chapters on:- neutron polarizers and instrumentation, polarized neutron scattering, neutron polarization analysis experiments and precessing neutron polarization. (U.K.)

  3. Determination of the 54Fe(n, 2n)53gFe and 54Fe(n, 2n)53mFe cross sections averaged over a 235U fission neutron spectrum

    International Nuclear Information System (INIS)

    Ribeiro Guevara, S.; Arribere, M.; Kestelman, A.J.

    2002-01-01

    The reaction cross sections averaged over a 235 U fission neutron spectrum have been measured for the 54 Fe(n, 2n) 53g Fe and 54 Fe(n, 2n) 53m Fe threshold reactions. The values found are, respectively: (1.14 ± 0.13) μb, and (0.52 ± 0.16) μb. The measured cross sections are referred to the (111± 3) mb standard cross section of the 58 Ni(n, p) 58m+g Co reaction. The (81.7 ± 2.2) mb standard cross section value for the 54 Fe(n, p) 54 Mn reaction, was also used as a monitor to check the results obtained with the Ni standard, leading to an excellent agreement. (author)

  4. Diffuse scattering of neutrons

    International Nuclear Information System (INIS)

    Novion, C.H. de.

    1981-02-01

    The use of neutron scattering to study atomic disorder in metals and alloys is described. The diffuse elastic scattering of neutrons by a perfect crystal lattice leads to a diffraction spectrum with only Bragg spreads. the existence of disorder in the crystal results in intensity and position modifications to these spreads, and above all, to the appearance of a low intensity scatter between Bragg peaks. The elastic scattering of neutrons is treated in this text, i.e. by measuring the number of scattered neutrons having the same energy as the incident neutrons. Such measurements yield information on the static disorder in the crystal and time average fluctuations in composition and atomic displacements [fr

  5. Neutron--neutron logging

    International Nuclear Information System (INIS)

    Allen, L.S.

    1977-01-01

    A borehole logging tool includes a steady-state source of fast neutrons, two epithermal neutron detectors, and two thermal neutron detectors. A count rate meter is connected to each neutron detector. A first ratio detector provides an indication of the porosity of the formation surrounding the borehole by determining the ratio of the outputs of the two count rate meters connected to the two epithermal neutron detectors. A second ratio detector provides an indication of both porosity and macroscopic absorption cross section of the formation surrounding the borehole by determining the ratio of the outputs of the two count rate meters connected to the two thermal neutron detectors. By comparing the signals of the two ratio detectors, oil bearing zones and salt water bearing zones within the formation being logged can be distinguished and the amount of oil saturation can be determined. 6 claims, 2 figures

  6. Burnup calculation with estimated neutron spectrum of JMTR irradiation field. Development of the burnup calculation method for fuel pre-irradiated in the JMTR

    International Nuclear Information System (INIS)

    Okonogi, Kazunari; Nakamura, Takehiko; Yoshinaga, Makio; Hosoyamada, Ryuji

    1999-03-01

    As a series of the pulse irradiation tests with the irradiated fuel, the high-enriched fuel rods pre-irradiated in the JMTR as well as the fuels irradiated in commercial reactors have been irradiated in the NSRR. In the pre-irradiation at the JMTR, the test fuels were placed at the irradiation holes in the reflector region far from the driver core to keep the linear heat generation rate of the test fuel low. Accordingly, neutron energy spectra of the irradiation holes for the test fuels are softened due to the higher moderator ratio than in those of the ordinary LWR core, which causes quite different burnup characteristics. JMTR post irradiation condition corresponds to the pre-test condition in the NSRR. Therefore, proper understanding of the condition is quite important for the precise evaluating the energy deposition and FP generation in the test. Then, neutron spectra at the JMTR irradiation field were evaluated and its effects on the burnup calculation were quantified. Basing on the configuration of the JMTR core in the operation cycle No.85, neutron diffusion calculations of 107 groups were executed in 2-D slab (X-Y) geometry of CITATION of SRAC95 code system, and neutron energy spectra of the irradiation hole for the test fuels were evaluated. Burnup calculations of Test JMN-1 fuel with the estimated neutron energy spectra were performed and the results were compared to both the measurements and calculation results with the PWR and BWR libraries in ORIGEN2 code. SWAT code was used to collapse the 107 groups spectra into 1 group libraries for the ORIGEN2 use. The calculation results for both the generation and depletion of U, Pu and Nd with the JMTR libraries obtained in the present study were in the reasonably good agreement with the measurements, while in the case of calculation with the PWR and BWR libraries in ORIGEN2, the generation of fission products having mass numbers from 105 to 130 and some actinides were overestimated by about 1.5 to 3.5 times

  7. Radionuclide 252Cf neutron source

    International Nuclear Information System (INIS)

    Kolevatov, Yu.I.; Trykov, L.A.

    1979-01-01

    Characteristics of radionuclide neutron sourses of 252 Cf base with the activity from 10 6 to 10 9 n/s have been investigated. Energetic distributions of neutrons and gamma-radiation have been presented. The results obtained have been compared with other data available. The hardness parameter of the neutron spectrum for the energy range from 3 to 15 MeV is 1.4 +- 0.02 MeV

  8. Spallation neutron source moderator design

    International Nuclear Information System (INIS)

    Charlton, L.A.; Barnes, J.M.; Gabriel, T.A.; Johnson, J.O.

    1998-01-01

    This paper describes various aspects of the spallation neutron source (SNS) moderator design. Included are the effects of varying the moderator location, interaction effects between moderators, and the impact on neutron output when various reflector materials are used. Also included is a study of the neutron output from composite moderators, where it is found that a combination of liquid H 2 O and liquid H 2 can produce a spectrum very similar to liquid methane (L-CH 4 ). (orig.)

  9. An Emergency Dosimeter for Neutrons

    Energy Technology Data Exchange (ETDEWEB)

    Braun, J; Nilsson, R

    1960-05-15

    A neutron dosimeter suitable for single emergency exposures is described. The dosimeter is furnished with detectors for thermal, epi-thermal and fast neutrons. This means that three of the constants by which the spectrum of the incident neutron flux is approximated, can be determined. The dose calculated from these approximated spectra is compared to the dose from spectra obtained in different standard spectra of types which may be expected in a radiation accident.

  10. Prompt neutron emission

    International Nuclear Information System (INIS)

    Sher, R.

    1959-01-01

    It is shown that Ramanna and Rao's tentative conclusion that prompt fission neutrons are emitted (in the fragment system) preferentially in the direction of fragment motion is not necessitated by their angular distribution measurements, which are well explained by the usual assumptions of isotropic emission with a Maxwell (or Maxwell-like) emission spectrum. The energy distribution (Watt spectrum) and the angular distribution, both including the effects of anisotropic emission, are given. (author) [fr

  11. Study of the effect of the energy spectrum and of the total flux on the damage produced by neutrons in solids; Contribution a l'etude de l'influence du spectre et du flux integre sur les dommages crees par les neutrons dans les solides

    Energy Technology Data Exchange (ETDEWEB)

    Dulieu, P.C. [Commissariat a l' Energie Atomique, Fontenay-aux-Roses (France). Centre d' Etudes Nucleaires

    1967-01-01

    In the first part are studied the general relationships between the physical effects produced by neutrons in solids, and the total flux and neutron energy spectrum; some examples are given. The second part, describes the application to a silicon damage detector whose principle is to use the damage produced in a PIN Junction for measuring the neutron flux (intermediate and fast) received by the detector. Chapter I is devoted to the experimental determination of the energy given to the atoms by a primary in the silicon. The results and conclusions drawn from this determination make it possible to consider, in chapter II, the calculation of the detectors response characteristics.Chapter III deals with the measurement of the detectors response function and it is noted that good agreement is obtained between the calculation and experimental results. The whole of the second part constitutes a test of the methods presented in the first part. (author) [French] Dans la premiere partie, nous etudions les relations generales qui lient les effets physiques engendres par les neutrons dans les solides au flux integre et au spectre des neutrons et nous donnons des exemples d'utilisation. La deuxieme partie est une application au detecteur de dommages en silicium, dont le principe est d'utiliser les dommages crees dans une jonction PIN pour mesurer les flux de neutrons (intermediaires et rapides) recus par le detecteur. Le chapitre I est consacre a la determination experimentale de l'energie cedee aux atomes par un primaire dans le silicium. Les resultats et les conclusions que l'on peut tirer de cette determination permettent d'aborder, au chapitre II, le calcul de la fonction de reponse du detecteur. Le chapitre III porte sur la mesure de la fonction de reponse du detecteur et on constate qu'il y a un bon accord entre le calcul et l'experience. L'ensemble de la deuxieme partie constitue un test des methodes exposees dans la premiere partie

  12. APPLE-3: improvement of APPLE for neutron and gamma-ray flux, spectrum and reaction rate plotting code, and of its code manual

    International Nuclear Information System (INIS)

    Kawasaki, Hiromitu; Maki, Koichi; Seki, Yasushi.

    1991-03-01

    A code APPLE was produced in 1976 for calculating and plotting tritium breeding ratio and tritium production rate distributions. That code was improved as 'APPLE-2' in 1982, to calculate and plot not only tritium breeding ratio but also distributions of neutron and gamma-ray fluxes, their spectra, nuclear heating rates and other reaction rates, and dose rate distributions during operation and after shutdown in 1982. The code APPLE-2 can calculate and plot these nuclear properties derived from neutron and gamma-ray fluxes by ANISN (one dimensional transport code), DOT3.5 (two dimensional transport code) and MORSE (three dimensional Monte Carlo code). We revised the code APPLE-2 as 'APPLE-3' by adding many functions to the APPLE-2 code in accordance with users' requirements proposed in recent progress of fusion reaction nuclear design. With minor modification of APPLE-2, a number of inconsistencies have been found between the code manual and the input data in the code. In the present report, the new functions added to APPLE-2 and improved users' manual are explained. (author)

  13. Ultracold neutrons

    International Nuclear Information System (INIS)

    Steenstrup, S.

    Briefly surveys recent developments in research work with ultracold neutrons (neutrons of very low velocity, up to 10 m/s at up to 10 -7 eV and 10 -3 K). Slow neutrons can be detected in an ionisation chamber filled with B 10 F 3 . Very slow neutrons can be used for investigations into the dipole moment of neutrons. Neutrons of large wave length have properties similar to those of light. The limit angle for total reflection is governed by the wave length and by the material. Total reflection can be used to filter ultracold neutrons out of the moderator material of a reactor. Total reflection can also be used to store ultracold neutrons but certain problems with storage have not yet been clarified. Slow neutrons can be made to lose speed in a neutron turbine, and come out as ultracold neutrons. A beam of ultracold neutrons could be used in a neutron microscope. (J.S.)

  14. Neutron Investigation of Magnon Spectrum in Haematite; Etude du Spectre de Magnons dans l'Hematite, au Moyen des Neutrons; Nejtronnoe issledovanie spektra magnona v gematite; Estudio, por Metodos Neutronicos, del Espectro de Magnones en la Hematita

    Energy Technology Data Exchange (ETDEWEB)

    Dimitrijevic, Z.; Rzany, H.; Todorovic, J.; Wanic, A. [Institute for Nuclear Physics Cracow (Poland)

    1965-04-15

    Measurements were done in Vinca on the RA reactor using a crystal neutron spectrometer. Monochromatic neutrons ({lambda} = 1.314A) were scattered on a large single crystal of haematite ({alpha}-Fe{sub 2}O{sub 3}). The angular distributions of inelastically scattered neutrons were investigated. For a number of different crystal missetting angles {Delta}{theta}, the width {Gamma} of the scattered neutron beam (so-called scattering cone) was measured. This scattering cone was ascribed to the magnon scattering surface surrounding the (111) reciprocal lattice point. For several magnon velocities the {Gamma} versus missetting angle dependence was calculated and compared with the experimental points. The value of the velocity in the [111] direction was found to be equal to 25.5{+-} 1.0 km/s. Structural anisotropy of the magnon dispersion relation was discovered. The velocity was found to be higher for the directions of propagation parallel to the [111] axis in qualitative agreement with the earlier measurements (Riste and others) giving v = 38 km/s. It was not possible to reveal the existence of an energy gap Eg in the acoustic magnon energy band. It was estimated that the Eg value must be below 1 meV. Using the formalism of Wallace the magnon dispersion relations in haematite were calculated. This was done under the assumption that there exist two non-vanishing exchange integrals J{sub 1} and J{sub 2}. J{sub 1} and J{sub 2} denote the superexchange energy coupling between the spins of neighbouring iron ions bonded via an oxygen ion. The Fe-O-Fe angles of the bonds taken into consideration are correspondingly 132 and 126 Degree-Sign . Satisfactory agreement between the theory and existing experimental data is obtained when SJ{sub 1} = SJ{sub 2} = 5.1 meV. Apart from the acoustic magnon branch, the calculations revealed the existence of an optic branch but this was not accessible to investigation by the neutron diffraction technique. (author) [French] Les mesures ont ete

  15. Neutron metrology in LAMPF, USA

    International Nuclear Information System (INIS)

    Ketema, D.J.; Nolthenius, H.J.

    1990-08-01

    The characterization of appropriate materials for fusion reactors requires a high intensity neutron source which simulates the neutron spectrum and the radiation conditions at the positions of interest in a fusion reactor (first wall). A neutron spectrum of interest is found in the Clinton P. Anderson Los Alamos Meson Physics Facility (LAMPF). Various ceramic materials and some polycrystalline graphites were irradiated in this facility during two intervals of time in 1986 and 1987. The specimens were accompanied by several sets with activation detectors. This report presents the saturation activities per atom obtained from these sets. (author). 3 refs.; 8 figs.; 10 tabs

  16. ATLAS MDT neutron sensitivity measurement and modeling

    International Nuclear Information System (INIS)

    Ahlen, S.; Hu, G.; Osborne, D.; Schulz, A.; Shank, J.; Xu, Q.; Zhou, B.

    2003-01-01

    The sensitivity of the ATLAS precision muon detector element, the Monitored Drift Tube (MDT), to fast neutrons has been measured using a 5.5 MeV Van de Graaff accelerator. The major mechanism of neutron-induced signals in the drift tubes is the elastic collisions between the neutrons and the gas nuclei. The recoil nuclei lose kinetic energy in the gas and produce the signals. By measuring the ATLAS drift tube neutron-induced signal rate and the total neutron flux, the MDT neutron signal sensitivities were determined for different drift gas mixtures and for different neutron beam energies. We also developed a sophisticated simulation model to calculate the neutron-induced signal rate and signal spectrum for ATLAS MDT operation configurations. The calculations agree with the measurements very well. This model can be used to calculate the neutron sensitivities for different gaseous detectors and for neutron energies above those available to this experiment

  17. Different spectra with the same neutron source

    International Nuclear Information System (INIS)

    Vega C, H. R.; Ortiz R, J. M.; Hernandez D, V. M.; Martinez B, M. R.; Hernandez A, B.; Ortiz H, A. A.; Mercado, G. A.

    2010-01-01

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

  18. Neutron flux monitoring device

    International Nuclear Information System (INIS)

    Shimazu, Yoichiro.

    1995-01-01

    In a neutron flux monitoring device, there are disposed a neutron flux measuring means for outputting signals in accordance with the intensity of neutron fluxes, a calculation means for calculating a self power density spectrum at a frequency band suitable to an object to be measured based on the output of the neutron flux measuring means, an alarm set value generation means for outputting an alarm set value as a comparative reference, and an alarm judging means for comparing the alarm set value with the outputted value of the calculation means to judge requirement of generating an alarm and generate an alarm in accordance with the result of the judgement. Namely, the time-series of neutron flux signals is put to fourier transformation for a predetermined period of time by the calculation means, and from each of square sums for real number component and imaginary number component for each of the frequencies, a self power density spectrum in the frequency band suitable to the object to be measured is calculated. Then, when the set reference value is exceeded, an alarm is generated. This can reliably prevent generation of erroneous alarm due to neutron flux noises and can accurately generate an alarm at an appropriate time. (N.H.)

  19. Safety studies dedicated to molten salt reactors with a fast neutron spectrum and operated in the Thorium fuel cycle - Innovative concept of Molten Salt Fast Reactor

    International Nuclear Information System (INIS)

    Brovchenko, Mariya

    2013-01-01

    The nuclear reactors of the 4. generation must allow an optimized use of natural resources, while performing at a high safety level. The framework of this thesis is the deployment study of one of such a system, an innovative and still little studied Molten Salt Fast Reactor. An excellent safety is an ultimate requirement of the nuclear energy deployment, so it is important to raise this question at the current early stage of the MSFR concept development. This concept was the subject of a neutronic tool benchmark within a European project EVOL. Definition, calculations and results analyses were performed during this thesis. Comparisons of static neutronic and burn-up calculations, performed by the project participants, concluded to a good agreement between the different codes and methods used and pointed out the sensibility of the nuclear database choice on the results. With the aim of safety analysis of the MSFR, the decay heat was studied in detail. The tool used for the decay heat calculation was developed and validated, to finally evaluate the decay heat in the reactor. The decay heat source presented in different zones was quantified, concluding to a high importance of the cooling of the fuel salt and the bubbling system enclosing a part of the fission products. The safety analysis methodology was also studied in this thesis. Even if the safety principles are directly transposable to the MSFR, the precise recommendations are not. This is due to the specificity of the design that relies on the liquid state of the fuel, on the reprocessing systems located in the reactor and the embryonic stage of the design. First, a preliminary transposition work of some criteria to the MSFR design was realized, resulting amongst other things in a list of accidental scenarios particular for MSFR. Finally, a preliminary physical study of some types of accidental scenarios was performed, that can be used as a basis for further analyses with more sophisticated tools. (author) [fr

  20. First flux measurement in a SINQ supermirror neutron guide

    Energy Technology Data Exchange (ETDEWEB)

    Janssen, S.; Schlumpf, N.; Bauer, G. [Paul Scherrer Inst. (PSI), Villigen (Switzerland)

    1997-09-01

    On Dec. 3, 1996, the Swiss spallation neutron source SINQ was taken into operation and produced its first neutrons successfully. The neutron spectrum within one of the supermirror guides was estimated by a chopper Time-of-Flight method. The result shows a 30% higher neutron intensity at the flux maximum than expected from previous Monte-Carlo simulations. (author) 1 fig., 4 refs.

  1. Fast Spectrum Reactors

    CERN Document Server

    Todd, Donald; Tsvetkov, Pavel

    2012-01-01

    Fast Spectrum Reactors presents a detailed overview of world-wide technology contributing to the development of fast spectrum reactors. With a unique focus on the capabilities of fast spectrum reactors to address nuclear waste transmutation issues, in addition to the well-known capabilities of breeding new fuel, this volume describes how fast spectrum reactors contribute to the wide application of nuclear power systems to serve the global nuclear renaissance while minimizing nuclear proliferation concerns. Readers will find an introduction to the sustainable development of nuclear energy and the role of fast reactors, in addition to an economic analysis of nuclear reactors. A section devoted to neutronics offers the current trends in nuclear design, such as performance parameters and the optimization of advanced power systems. The latest findings on fuel management, partitioning and transmutation include the physics, efficiency and strategies of transmutation, homogeneous and heterogeneous recycling, in addit...

  2. The 4π neutron detector CARMEN

    Energy Technology Data Exchange (ETDEWEB)

    Ledoux, X., E-mail: Xavier.ledoux@ganil.fr [CEA/DAM/DIF, F-91297 Arpajon (France); GANIL, CEA/DRF-CNRS/IN2P3, Caen, F-14076 France (France); Laborie, J.-M.; Pras, P.; Lantuéjoul-Thfoin, I.; Varignon, C. [CEA/DAM/DIF, F-91297 Arpajon (France)

    2017-02-01

    CARMEN is a 4π neutron detector filled with a gadolinium-loaded liquid scintillator built to measure neutron multiplicity distributions. It is used to study fission and (n,xn) reactions. In addition to neutron multiplicity measurements, CARMEN can be used to measure neutron energy spectra with the time-of-flight technique, thanks to the time properties of the prompt signal. The detector, detection technique and efficiency determination are presented in detail. Two examples are also presented: the measurement of {sup 252}Cf spontaneous fission neutron multiplicity probability distribution and the measurement of the neutron energy spectrum emitted by an Am-Be radioactive source.

  3. Neutron dosimetry in biology

    International Nuclear Information System (INIS)

    Sigurbjoernsson, B.; Smith, H.H.; Gustafsson, A.

    1965-01-01

    To study adequately the biological effects of different energy neutrons it is necessary to have high-intensity sources which are not contaminated by other radiations, the most serious of which are gamma rays. An effective dosimetry must provide an accurate measure of the absorbed dose, in biological materials, of each type of radiation at any reactor facility involved in radiobiological research. A standardized biological dosimetry, in addition to physical and chemical methods, may be desirable. The ideal data needed to achieve a fully documented dosimetry has been compiled by H. Glubrecht: (1) Energy spectrum and intensity of neutrons; (2) Angular distribution of neutrons on the whole surface of the irradiated object; (3) Additional undesired radiation accompanying the neutrons; (4) Physical state and chemical composition of the irradiated object. It is not sufficient to note only an integral dose value (e.g. in 'rad') as the biological effect depends on the above data

  4. Hyper-thermal neutron irradiation field for neutron capture therapy

    International Nuclear Information System (INIS)

    Sakurai, Yoshinori; Kobayashi, Tooru; Kanda, Keiji

    1994-01-01

    The utilization of hyper-thermal neutrons, which have an energy spectrum of a Maxwell distribution higher than the room temperature of 300 K, has been studied in order to improve the thermal neutron flux distribution in a living body for a deep-seated tumor in neutron capture therapy (NCT). Simulation calculations using MCNP-V3 were carried out in order to investigate the characteristics of the hyper-thermal neutron irradiation field. From the results of simulation calculations, the following were confirmed: (i) The irradiation field of the hyper-thermal neutrons is feasible by using some scattering materials with high temperature, such as Be, BeO, C, SiC and ZrH 1.7 . Especially, ZrH 1.7 is thought to be the best material because of good characteristics of up-scattering for thermal neutrons. (ii) The ZrH 1.7 of 1200 K yields the hyper-thermal neutrons of a Maxwell-like distribution at about 2000 K and the treatable depth is about 1.5 cm larger comparing with the irradiation of the thermal neutrons of 300 K. (iii) The contamination by the secondary gamma-rays from the scattering materials can be sufficiently eliminated to the tolerance level for NCT through the bismuth layer, without the larger change of the energy spectrum of hyper-thermal neutrons. ((orig.))

  5. Review of Non-Neutron and Neutron Nuclear Data, 2004

    International Nuclear Information System (INIS)

    Holden, Norman E.

    2005-01-01

    Review articles are in preparation for the 2004 edition of the CRC Handbook of Chemistry and Physics dealing with the evaluation of both non-neutron and neutron nuclear data. Data on the discovery of element 110, Darmstadtium, and element 111 have been officially accepted, while data on element 118 have been withdrawn. Data to be presented include revised values for very short-lived nuclides, long-lived nuclides, and beta-beta decay measurements. There has been a reassessment of the spontaneous fission (sf) half-lives, which distinguishes between sf decay half-lives and cluster decay half-lives, and with cluster-fission decay. New measurements of isotopic abundance values for many elements will be discussed with an emphasis on the minor isotopes of interest for use in neutron activation analysis measurements. Neutron resonance integrals will be discussed emphasizing the differences between the calculated values obtained from the analytical integration over neutron resonances and the measured values in a neutron reactor-spectrum, which does not quite conform to the assumed 1/E neutron energy spectrum. The method used to determine the neutron resonance integral from measurement, using neutron activation analysis, will be discussed

  6. Compilation of neutron flux density spectra and reaction rates in different neutron fields

    International Nuclear Information System (INIS)

    Ertek, C.

    1979-07-01

    Upon the recommendation of International Working Group of Reactor Radiation Measurements (IWGRRM), the compilation of neutron flux density spectra and the reaction rates obtained by activation and fission foils in different neutron fields is presented. The neutron fields considered are as follows: 1/E; iron block; LWR core and pressure vessel; LMFBR core and blanket; CTR first wall and blanket; fission spectrum

  7. Measurement of 237Np fission rate ratio relative to 235U fission rate in cores with various thermal neutron spectrum at the Kyoto University Critical Assembly

    International Nuclear Information System (INIS)

    Unesaki, Hironobu; Shiroya, Seiji; Iwasaki, Tomohiko; Fujiwara, Daisuke; Kitada, Takanori; Kuroda, Mitsuo; Kohashi, Akio; Kato, Takeshi; Ikeuchi, Yoshitaka

    2000-01-01

    Integral measurements of 237 Np fission rate ratio relative to 235 U fission rate have been performed at Kyoto University Citrical Assembly. The fission rates have been measured using the back-to back type double fission chamber at five thermal cores with different H/ 235 U ratio so that the neutron spectra of the cores were systematically varied. The measured fission rate ratio per atom was 0.00439 to 0.0298, with a typical uncertainty of 2 to 3%. The measured data were compared with the calculated results using SRAC/TWOTRAN and MVP based on JENDL-3.2, which gave the averaged C/E values of 0.93 and 0.95, respectively. Obtained results of C/E using 237 Np cross sections from JENDL-3/2, ENDF/B-VI.5 and JEF2.2 show that the latter two gave smaller results than JENDL-3.2 by about 4%, which clearly reflects the discrepancy in the evaluated cross section among the libraries. This difference arises from both fast fission and resonance region. Although further improvement is recommended, 237 Np fission cross section in JENDL-3.2 is considered to be superior to those in the other libraries and can be adopted for use in design calculations for minor actinide transmutation system using thermal reactors with prediction precision of 237 Np fission rate with in 10%. (author)

  8. Facility for fast neutron irradiation tests of electronics at the ISIS spallation neutron source

    International Nuclear Information System (INIS)

    Andreani, C.; Pietropaolo, A.; Salsano, A.; Gorini, G.; Tardocchi, M.; Paccagnella, A.; Gerardin, S.; Frost, C. D.; Ansell, S.; Platt, S. P.

    2008-01-01

    The VESUVIO beam line at the ISIS spallation neutron source was set up for neutron irradiation tests in the neutron energy range above 10 MeV. The neutron flux and energy spectrum were shown, in benchmark activation measurements, to provide a neutron spectrum similar to the ambient one at sea level, but with an enhancement in intensity of a factor of 10 7 . Such conditions are suitable for accelerated testing of electronic components, as was demonstrated here by measurements of soft error rates in recent technology field programable gate arrays

  9. 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.

  10. Compilation of Existing Neutron Screen Technology

    Directory of Open Access Journals (Sweden)

    N. Chrysanthopoulou

    2014-01-01

    Full Text Available The presence of fast neutron spectra in new reactors is expected to induce a strong impact on the contained materials, including structural materials, nuclear fuels, neutron reflecting materials, and tritium breeding materials. Therefore, introduction of these reactors into operation will require extensive testing of their components, which must be performed under neutronic conditions representative of those expected to prevail inside the reactor cores when in operation. Due to limited availability of fast reactors, testing of future reactor materials will mostly take place in water cooled material test reactors (MTRs by tailoring the neutron spectrum via neutron screens. The latter rely on the utilization of materials capable of absorbing neutrons at specific energy. A large but fragmented experience is available on that topic. In this work a comprehensive compilation of the existing neutron screen technology is attempted, focusing on neutron screens developed in order to locally enhance the fast over thermal neutron flux ratio in a reactor core.

  11. Unfolding of neutron spectra from Godiva type critical assemblies

    International Nuclear Information System (INIS)

    Harvey, J.T.; Meason, J.L.; Wright, H.L.

    1976-01-01

    The results from three experiments conducted at the White Sands Missile Range Fast Burst Reactor Facility are discussed. The experiments were designed to measure the ''free-field'' neutron leakage spectrum and the neutron spectra from mildly perturbed environments. SAND-II was used to calculate the neutron spectrum utilizing several different trial input spectra for each experiment. Comparisons are made between the unfolded neutron spectrum for each trial input on the basis of the following parameters: average neutron energy (above 10 KeV), integral fluence (above 10 KeV), spectral index and the hardness parameter, phi/sub eq//phi

  12. Neutron reflectometry

    International Nuclear Information System (INIS)

    Van Well, A.A.

    1999-01-01

    Neutron research where reflection, refraction, and interference play an essential role is generally referred to as 'neutron optics'. The neutron wavelength, the scattering length density and the magnetic properties of the material determine the critical angle for total reflection. The theoretical background of neutron reflection, experimental methods and the interpretation of reflection data are presented. (K.A.)

  13. Neutron spectrometer for improved SNM search.

    Energy Technology Data Exchange (ETDEWEB)

    Vance, Andrew L.; Aigeldinger, Georg

    2007-03-01

    With the exception of large laboratory devices with very low sensitivities, a neutron spectrometer have not been built for fission neutrons such as those emitted by special nuclear materials (SNM). The goal of this work was to use a technique known as Capture Gated Neutron Spectrometry to develop a solid-state device with this functionality. This required modifications to trans-stilbene, a known solid-state scintillator. To provide a neutron capture signal we added lithium to this material. This unique triggering signal allowed identification of neutrons that lose all of their energy in the detector, eliminating uncertainties that arise due to partial energy depositions. We successfully implemented a capture gated neutron spectrometer and were able to distinguish an SNM like fission spectrum from a spectrum stemming from a benign neutron source.

  14. Fast neutron dosimetry

    International Nuclear Information System (INIS)

    DeLuca, P.M. Jr.; Pearson, D.W.

    1992-01-01

    This progress report concentrates on two major areas of dosimetry research: measurement of fast neutron kerma factors for several elements for monochromatic and white spectrum neutron fields and determination of the response of thermoluminescent phosphors to various ultra-soft X-ray energies and beta-rays. Dr. Zhixin Zhou from the Shanghai Institute of Radiation Medicine, People's Republic of China brought with him special expertise in the fabrication and use of ultra-thin TLD materials. Such materials are not available in the USA. The rather unique properties of these materials were investigated during this grant period

  15. Neutron radiography with ultracold neutrons

    International Nuclear Information System (INIS)

    Bates, J.C.

    1981-01-01

    The neutron transmission factor of very thin films may be low if the neutron energy is comparable to the pseudo-potential of the film material. Surprisingly, perhaps, it is relatively easy to obtain neutrons with such low energies in sufficient numbers to produce neutron radiographs. (orig.)

  16. The Thermal Neutron Beam Option for NECTAR at MLZ

    Science.gov (United States)

    Mühlbauer, M. J.; Bücherl, T.; Genreith, C.; Knapp, M.; Schulz, M.; Söllradl, S.; Wagner, F. M.; Ehrenberg, H.

    The beam port SR10 at the neutron source FRM II of Heinz Maier-Leibnitz Zentrum (MLZ) is equipped with a moveable assembly of two uranium plates, which can be placed in front of the entrance window of the beam tube via remote control. With these plates placed in their operating position the thermal neutron spectrum produced by the neutron source FRM II is converted to fission neutrons with 1.9 MeV of mean energy. This fission neutron spectrum is routinely used for medical applications at the irradiation facility MEDAPP, for neutron radiography and tomography experiments at the facility NECTAR and for materials testing. If, however, the uranium plates are in their stand-by position far off the tip of the beam tube and the so-called permanent filter for thermal neutrons is removed, thermal neutrons originating from the moderator tank enter the beam tube and a thermal spectrum becomes available for irradiation or activation of samples. By installing a temporary flight tube the beam may be used for thermal neutron radiography and tomography experiments at NECTAR. The thermal neutron beam option not only adds a pure thermal neutron spectrum to the energy ranges available for neutron imaging at MLZ instruments but it also is an unique possibility to combine two quite different neutron energy ranges at a single instrument including their respective advantages. The thermal neutron beam option for NECTAR is funded by BMBF in frame of research project 05K16VK3.

  17. 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

  18. Neutron Skins and Neutron Stars

    OpenAIRE

    Piekarewicz, J.

    2013-01-01

    The neutron-skin thickness of heavy nuclei provides a fundamental link to the equation of state of neutron-rich matter, and hence to the properties of neutron stars. The Lead Radius Experiment ("PREX") at Jefferson Laboratory has recently provided the first model-independence evidence on the existence of a neutron-rich skin in 208Pb. In this contribution we examine how the increased accuracy in the determination of neutron skins expected from the commissioning of intense polarized electron be...

  19. Feasibility of sealed D-T neutron generator as neutron source for liver BNCT and its beam shaping assembly.

    Science.gov (United States)

    Liu, Zheng; Li, Gang; Liu, Linmao

    2014-04-01

    This paper involves the feasibility of boron neutron capture therapy (BNCT) for liver tumor with four sealed neutron generators as neutron source. Two generators are placed on each side of the liver. The high energy of these emitted neutrons should be reduced by designing a beam shaping assembly (BSA) to make them useable for BNCT. However, the neutron flux decreases as neutrons pass through different materials of BSA. Therefore, it is essential to find ways to increase the neutron flux. In this paper, the feasibility of using low enrichment uranium as a neutron multiplier is investigated to increase the number of neutrons emitted from D-T neutron generators. The neutron spectrum related to our system has a proper epithermal flux, and the fast and thermal neutron fluxes comply with the IAEA recommended values. Copyright © 2014 Elsevier Ltd. All rights reserved.

  20. Neutronic moderator design for the Spallation Neutron Source (SNS)

    International Nuclear Information System (INIS)

    Charlton, L.A.; Barnes, J.M.; Johnson, J.O.; Gabriel, T.A.

    1998-01-01

    Neutronics analyses are now in progress to support the initial selection of moderator design parameters for the Spallation Neutron Source (SNS). The results of the initial optimization studies involving moderator poison plate location, moderator position, and premoderator performance for the target system are presented in this paper. Also presented is an initial study of the use of a composite moderator to produce a liquid methane like spectrum

  1. 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

  2. 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

  3. Neutron Dosimetry

    International Nuclear Information System (INIS)

    Vanhavere, F.

    2001-01-01

    The objective of SCK-CEN's R and D programme on neutron dosimetry is to improve the determination of neutron doses by studying neutron spectra, neutron dosemeters and shielding adaptations. In 2000, R and D focused on the contiued investigation of the bubble detectors type BD-PND and BDT, in particular their sensitivity and temperature dependence; the updating of SCK-CEN's criticality dosemeter, the investigation of the characteristics of new thermoluminescent materials and their use in neutron dosemetry; and the investigation of neutron shielding

  4. Neutron Dosimetry

    Energy Technology Data Exchange (ETDEWEB)

    Vanhavere, F

    2001-04-01

    The objective of SCK-CEN's R and D programme on neutron dosimetry is to improve the determination of neutron doses by studying neutron spectra, neutron dosemeters and shielding adaptations. In 2000, R and D focused on the contiued investigation of the bubble detectors type BD-PND and BDT, in particular their sensitivity and temperature dependence; the updating of SCK-CEN's criticality dosemeter, the investigation of the characteristics of new thermoluminescent materials and their use in neutron dosemetry; and the investigation of neutron shielding.

  5. Constitutive laws for the neutron density current

    International Nuclear Information System (INIS)

    Espinosa-Paredes, Gilberto; Morales-Sandoval, Jaime B.; Vazquez-Rodriguez, Rodolfo; Espinosa-Martinez, Erick-G.

    2008-01-01

    In this technical note, a fractional wave equation for the average neutron motion in nuclear reactor is considered. This representation covers the full spectrum of the average neutron transport behavior, i.e., Fickian and non-Fickian effects. The fractional diffusion model retains the main dynamic characteristics of the neutron motion in which the relaxation time associated with a rapid variation in the neutron flux contains a fractional exponent that can be manipulated to obtain the best representation of the neutron transport phenomena. The detrended fluctuation analysis (DFA) method is presented in this paper to estimate the fractional exponent

  6. CARNAC, Neutron Flux and Neutron Spectra in Criticality Accident

    International Nuclear Information System (INIS)

    Bessis, J.

    1976-01-01

    Nature of physical problem solved: Calculation of flux and neutron spectra in the case of a criticality accident. The method is unsophisticated but fast. The program is divided into two parts: (1) The code CRITIC is based on the Fermi age equation and evaluates the neutron number per fission emitted from a moderate critical system and its energy spectrum. (2) The code NARCISSE uses concrete current albedo, evaluates the product of neutron reflection on walls of the source containment and calculates the resulting flux at any point, and its energy distribution into 21 groups. The results obtained seem satisfactory, if compared with a Monte Carlo program

  7. Neutron absorbed dose in a pacemaker CMOS

    International Nuclear Information System (INIS)

    Borja H, C. G.; Guzman G, K. A.; Valero L, C.; Banuelos F, A.; Hernandez D, V. M.; Vega C, H. R.; Paredes G, L.

    2012-01-01

    The neutron spectrum and the absorbed dose in a Complementary Metal Oxide Semiconductor (CMOS), has been estimated using Monte Carlo methods. Eventually a person with a pacemaker becomes an oncology patient that must be treated in a linear accelerator. Pacemaker has integrated circuits as CMOS that are sensitive to intense and pulsed radiation fields. Above 7 MV therapeutic beam is contaminated with photoneutrons that could damage the CMOS. Here, the neutron spectrum and the absorbed dose in a CMOS cell was calculated, also the spectra were calculated in two point-like detectors in the room. Neutron spectrum in the CMOS cell shows a small peak between 0.1 to 1 MeV and a larger peak in the thermal region, joined by epithermal neutrons, same features were observed in the point-like detectors. The absorbed dose in the CMOS was 1.522 x 10 -17 Gy per neutron emitted by the source. (Author)

  8. Neutron absorbed dose in a pacemaker CMOS

    Energy Technology Data Exchange (ETDEWEB)

    Borja H, C. G.; Guzman G, K. A.; Valero L, C.; Banuelos F, A.; Hernandez D, V. M.; Vega C, H. R. [Universidad Autonoma de Zacatecas, Unidad Academica de Estudios Nucleares, Cipres No. 10, Fracc. La Penuela, 98068 Zacatecas (Mexico); Paredes G, L., E-mail: fermineutron@yahoo.com [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

    2012-06-15

    The neutron spectrum and the absorbed dose in a Complementary Metal Oxide Semiconductor (CMOS), has been estimated using Monte Carlo methods. Eventually a person with a pacemaker becomes an oncology patient that must be treated in a linear accelerator. Pacemaker has integrated circuits as CMOS that are sensitive to intense and pulsed radiation fields. Above 7 MV therapeutic beam is contaminated with photoneutrons that could damage the CMOS. Here, the neutron spectrum and the absorbed dose in a CMOS cell was calculated, also the spectra were calculated in two point-like detectors in the room. Neutron spectrum in the CMOS cell shows a small peak between 0.1 to 1 MeV and a larger peak in the thermal region, joined by epithermal neutrons, same features were observed in the point-like detectors. The absorbed dose in the CMOS was 1.522 x 10{sup -17} Gy per neutron emitted by the source. (Author)

  9. 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.)

  10. Neutron sources and their characteristics

    International Nuclear Information System (INIS)

    McCall, R.C.; Swanson, W.P.

    1979-03-01

    The significant sources of photoneutrons within a linear-accelerator treatment head are identified and absolute estimates of neutron production per treatment dose are given for typical components. It is found that the high-Z materials within the treatment head do not significantly alter the neutron fluence but do substantially reduce the average energy of the transmitted spectrum. Reflection of neutrons from the concrete treatment room contribute to the neutron fluence, but not substantially to the patient integral dose, because of a further reduction in average energy. The ratio of maximum fluence to the treatment dose at the same distance is given as a function of electron energy. This ratio rises with energy to an almost constant value of 2.1 x 10 5 neutrons cm -2 rad -1 at electron energies above about 25 MeV. Measured data obtained at a variety of accelerator installations are presented and compared with these calculations. Reasons for apparent deviations are suggested. Absolute depth-dose and depth-dose-equivalent distributions for realistic neutron spectra that occur at therapy installations are calculated, and a rapid falloff with depth is found. The ratio of neutron integral absorbed dose to leakage photon absorbed dose is estimated to be 0.04 and 0.2 for 14 to 25 MeV incident electron energy, respectively. Possible reasons are given for lesser neutron production from betatrons than from linear accelerators. Possible ways in which neutron production can be reduced are discussed

  11. Neutrons from medical electron accelerators

    International Nuclear Information System (INIS)

    Swanson, W.P.; McCall, R.C.

    1979-06-01

    The significant sources of photoneutrons within a linear-accelerator treatment head are identified and absolute estimates of neutron production per treatment dose are given for typical components. Measured data obtained at a variety of accelerator installations are presented and compared with these calculations. It is found that the high-Z materials within the treatment head do not significantly alter the neutron fluence, but do substantially reduce the average energy of the transmitted spectrum. Reflected neutrons from the concrete treatment room contribute to the neutron fluence, but not substantially to the patient integral dose, because of a further reduction in average energy. Absolute depth-dose distributions for realistic neutron spectra are calculated, and a rapid falloff with depth is found

  12. Neutron fluence spectrometry using disk activation

    International Nuclear Information System (INIS)

    Loevestam, Goeran; Hult, Mikael; Fessler, Andreas; Gasparro, Joel; Kockerols, Pierre; Okkinga, Klaas; Tagziria, Hamid; Vanhavere, Filip; Wieslander, J.S. Elisabeth

    2009-01-01

    A simple and robust detector for spectrometry of environmental neutrons has been developed. The technique is based on neutron activation of a series of different metal disks followed by low-level gamma-ray spectrometry of the activated disks and subsequent neutron spectrum unfolding. The technique is similar to foil activation but here the applied neutron fluence rates are much lower than usually in the case of foil activation. The detector has been tested in quasi mono-energetic neutron fields with fluence rates in the order of 1000-10000 cm -2 s -1 , where the obtained spectra showed good agreement with spectra measured using a Bonner sphere spectrometer. The detector has also been tested using an AmBe source and at a neutron fluence rate of about 40 cm -2 s -1 , again, a good agreement with the assumed spectrum was achieved

  13. Neutron fluence spectrometry using disk activation

    Energy Technology Data Exchange (ETDEWEB)

    Loevestam, Goeran [EC-JRC-Institute for Reference Materials and Measurements (IRMM), Retieseweg 111, B-2440 Geel (Belgium)], E-mail: goeran.loevestam@ec.europa.eu; Hult, Mikael; Fessler, Andreas; Gasparro, Joel; Kockerols, Pierre; Okkinga, Klaas [EC-JRC-Institute for Reference Materials and Measurements (IRMM), Retieseweg 111, B-2440 Geel (Belgium); Tagziria, Hamid [EC-JRC-Institute for the Protection and the Security of the Citizen (IPSC), Via E. Fermi 1, I-21020 Ispra (Vatican City State, Holy See,) (Italy); Vanhavere, Filip [SCK-CEN, Boeretang, 2400 Mol (Belgium); Wieslander, J.S. Elisabeth [EC-JRC-Institute for Reference Materials and Measurements (IRMM), Retieseweg 111, B-2440 Geel (Belgium); Department of Physics, P.O. Box 35 (YFL), FIN-40014, University of Jyvaeskylae (Finland)

    2009-01-15

    A simple and robust detector for spectrometry of environmental neutrons has been developed. The technique is based on neutron activation of a series of different metal disks followed by low-level gamma-ray spectrometry of the activated disks and subsequent neutron spectrum unfolding. The technique is similar to foil activation but here the applied neutron fluence rates are much lower than usually in the case of foil activation. The detector has been tested in quasi mono-energetic neutron fields with fluence rates in the order of 1000-10000 cm{sup -2} s{sup -1}, where the obtained spectra showed good agreement with spectra measured using a Bonner sphere spectrometer. The detector has also been tested using an AmBe source and at a neutron fluence rate of about 40 cm{sup -2} s{sup -1}, again, a good agreement with the assumed spectrum was achieved.

  14. Spectral distortion due to scattered cold neutrons in beryllium filter

    International Nuclear Information System (INIS)

    Sakamoto, Yukio; Inoue, Kazuhiko

    1980-01-01

    Polycrystalline beryllium filters are used to discriminate the cold neutrons from the thermal neutrons with energies above Bragg cut-off energy. The cold neutron scattering cross section is very small, but the remaining cross section is not zero. Then the neutrons scattered once from the filter in the cold neutron energy region have chance of impinging on the outlet of filter. Those neutrons are almost upscattered and develop into thermal neutrons; thus the discriminated cold neutrons include a small spectral distortion due to the thermal neutrons. In the present work we have evaluated the effect on the cold neutron spectrum due to the repeatedly scattered and transmitted neutrons by using a Monte Carlo calculation method. (author)

  15. The neutron

    International Nuclear Information System (INIS)

    Kredov, B.M.

    1979-01-01

    The history of the neutron is displayed on the basis of contributions by scientists who produced outstanding results in neutron research (part 1), of summarizing discoveries and theories which led to the discovery of the neutron and the resulting development of nuclear physics (part 2), and of fundamental papers written by Rutherford, Chadwick, Iwanenko, and others (appendix). Of interest to physicists, historians, and students

  16. Neutron techniques

    International Nuclear Information System (INIS)

    Charlton, J.S.

    1986-01-01

    The way in which neutrons interact with matter such as slowing-down, diffusion, neutron absorption and moderation are described. The use of neutron techniques in industry, in moisture gages, level and interface measurements, the detection of blockages, boron analysis in ore feedstock and industrial radiography are discussed. (author)

  17. Properties of the TRIUMF neutron beam

    International Nuclear Information System (INIS)

    Gan, L.; Berdoz, A.R.; Green, P.W.; Greeniaus, L.G.; Helmer, R.; Korkmaz, E.; Lee, L.; Miller, C.A.; Opper, A.K.; Page, S.A.; Van Oers, W.T.H.; Zhao, J.

    1995-01-01

    Properties of the TRIUMF neutron beam (4A/2) are presented and compared with a Monte Carlo prediction. The beam intensity profile, energy spectrum and polarization are predicted taking into account the beamline geometry, energy losses in the LD 2 production target, the properties of the vector pd→ vector npp reaction, and the scattering of neutrons from the collimator walls. The results allow for improved corrections to systematic errors in a number of TRIUMF neutron experiments. (orig.)

  18. Sound response of superheated drop bubble detectors to neutrons

    International Nuclear Information System (INIS)

    Gao Size; Chen Zhe; Liu Chao; Ni Bangfa; Zhang Guiying; Zhao Changfa; Xiao Caijin; Liu Cunxiong; Nie Peng; Guan Yongjing

    2012-01-01

    The sound response of the bubble detectors to neutrons by using 252 Cf neutron source was described. Sound signals were filtered by sound card and PC. The short-time signal energy. FFT spectrum, power spectrum, and decay time constant were got to determine the authenticity of sound signal for bubbles. (authors)

  19. Neutron source multiplication method

    International Nuclear Information System (INIS)

    Clayton, E.D.

    1985-01-01

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

  20. Testing neutron cross-section files from the BROND-2 and ENDF/B-6 libraries in benchmark experiments on neutron transmission through spherical layers

    International Nuclear Information System (INIS)

    Androsenko, A.A.; Androsenko, P.A.; Blokhin, A.I.; Kulagin, N.T.; Pronyaev, V.G.; Simakov, S.P.

    1997-01-01

    The effect of angular anisotropy in inelastic secondary neutron scattering on neutron leakage spectra from the surface of spherical specimens is investigated. It is shown how inadequate representation of the cross-section structure in the neutron energy resonance region can affect the neutron leakage spectrum. (author). 19 refs, 5 figs, 6 tabs

  1. Testing neutron cross-section files from the BROND-2 and ENDF/B-6 libraries in benchmark experiments on neutron transmission through spherical layers

    Energy Technology Data Exchange (ETDEWEB)

    Androsenko, A A; Androsenko, P A; Blokhin, A I; Kulagin, N T; Pronyaev, V G; Simakov, S P [Institute of Physics and Power Engineering, Obninsk (Russian Federation)

    1997-06-01

    The effect of angular anisotropy in inelastic secondary neutron scattering on neutron leakage spectra from the surface of spherical specimens is investigated. It is shown how inadequate representation of the cross-section structure in the neutron energy resonance region can affect the neutron leakage spectrum. (author). 19 refs, 5 figs, 6 tabs.

  2. Feasibility of the Precise Energy Calibration for Fast Neutron Spectrometers

    Science.gov (United States)

    Gaganov, V. V.; Usenko, P. L.; Kryzhanovskaja, M. A.

    2017-12-01

    Computational studies aimed at improving the accuracy of measurements performed using neutron generators with a tritium target were performed. A measurement design yielding an extremely narrow peak in the energy spectrum of DT neutrons was found. The presence of such a peak establishes the conditions for precise energy calibration of fast-neutron spectrometers.

  3. A study on the utilization of hyper-thermal neutrons for neutron capture therapy

    International Nuclear Information System (INIS)

    Sakurai, Yoshinori; Kobayashi, Tooru; Kanda, Keiji

    1993-01-01

    The utilization of hyper-thermal neutrons, which have an energy spectrum of a Maxwellian distribution of a higher temperature than the room temperature of 300 K, was studied in order to improve the thermal neutron flux distribution at the deeper part in a living body for neutron capture therapy. Simulation calculations were carried out using MCNP-V3 in order to confirm the characteristics of hyper-thermal neutrons, i.e., (1) depth dependence of neutron energy spectrum, and (2) depth distribution of the reaction rate in a water phantom for materials with 1/v neutron absorption. It is confirmed that the hyper-thermal neutron irradiation can improve the thermal neutron flux distribution in the deeper and wider area in a living body compared with the thermal neutron irradiation. Practically, by the incidence of the hyper-thermal neutrons with a 3000 K Maxwellian distribution, the thermal neutron flux at 5 cm depth can be given about four times larger than by the incidence of the thermal neutrons of 300 K. (author)

  4. 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)

  5. A method for prediction of prompt fission neutron spectra

    International Nuclear Information System (INIS)

    Grashin, A.F.; Lepeshkin, M.V.

    1988-01-01

    Three-parameter formula for the prompt-fission-neutron integral spectrum is derived from a thermodynamical model. Two parameters, scission-neutron weight p = 11 % and anisotropy factor for accelerated fragments b = 10 %, are determined from experimental data, the same values being assumed for any type of fission. The thermodynamical theory provides the value of the third parameter, temperature τ, thus prognozing neutron spectrum and average energy with an error about 1 %. (author)

  6. Neutron radiography

    International Nuclear Information System (INIS)

    Hiraoka, Eiichi

    1988-01-01

    The thermal neutron absorption coefficient is essentially different from the X-ray absorption coefficient. Each substance has a characteristic absorption coefficient regardless of its density. Neutron deams have the following features: (1) neutrons are not transmitted efficiently by low molecular weight substances, (2) they are transmitted efficiently by heavy metals, and (3) the transmittance differs among isotopes. Thus, neutron beams are suitable for cheking for foreign matters in heavy metals and testing of composites consisting of both heavy and light materials. A neutron source generates fast neutrons, which should be converted into thermal neutrons by reducing their energy. Major neutron souces include nuclear reactors, radioisotopes and particle accelerators. Photographic films and television systems are mainly used to observe neutron transmission images. Computers are employed for image processing, computerized tomography and three-dimensional analysis. The major applications of neutron radiography include inspection of neclear fuel; evaluation of material for airplane; observation of fuel in the engine and oil in the hydraulic systems in airplanes; testing of composite materials; etc. (Nogami, K.)

  7. Calculations to support JET neutron yield calibration: Modelling of neutron emission from a compact DT neutron generator

    Science.gov (United States)

    Čufar, Aljaž; Batistoni, Paola; Conroy, Sean; Ghani, Zamir; Lengar, Igor; Milocco, Alberto; Packer, Lee; Pillon, Mario; Popovichev, Sergey; Snoj, Luka; JET Contributors

    2017-03-01

    At the Joint European Torus (JET) the ex-vessel fission chambers and in-vessel activation detectors are used as the neutron production rate and neutron yield monitors respectively. In order to ensure that these detectors produce accurate measurements they need to be experimentally calibrated. A new calibration of neutron detectors to 14 MeV neutrons, resulting from deuterium-tritium (DT) plasmas, is planned at JET using a compact accelerator based neutron generator (NG) in which a D/T beam impinges on a solid target containing T/D, producing neutrons by DT fusion reactions. This paper presents the analysis that was performed to model the neutron source characteristics in terms of energy spectrum, angle-energy distribution and the effect of the neutron generator geometry. Different codes capable of simulating the accelerator based DT neutron sources are compared and sensitivities to uncertainties in the generator's internal structure analysed. The analysis was performed to support preparation to the experimental measurements performed to characterize the NG as a calibration source. Further extensive neutronics analyses, performed with this model of the NG, will be needed to support the neutron calibration experiments and take into account various differences between the calibration experiment and experiments using the plasma as a source of neutrons.

  8. Calculations to support JET neutron yield calibration: Modelling of neutron emission from a compact DT neutron generator

    Energy Technology Data Exchange (ETDEWEB)

    Čufar, Aljaž, E-mail: aljaz.cufar@ijs.si [Reactor Physics Department, Jožef Stefan Institute, Jamova cesta 39, SI-1000 Ljubljana (Slovenia); EUROfusion Consortium, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Batistoni, Paola [ENEA, Department of Fusion and Nuclear Safety Technology, I-00044 Frascati, Rome (Italy); EUROfusion Consortium, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Conroy, Sean [Uppsala University, Department of Physics and Astronomy, PO Box 516, SE-75120 Uppsala (Sweden); EUROfusion Consortium, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Ghani, Zamir [Culham Centre for Fusion Energy, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); EUROfusion Consortium, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Lengar, Igor [Reactor Physics Department, Jožef Stefan Institute, Jamova cesta 39, SI-1000 Ljubljana (Slovenia); EUROfusion Consortium, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Milocco, Alberto; Packer, Lee [Culham Centre for Fusion Energy, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); EUROfusion Consortium, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Pillon, Mario [ENEA, Department of Fusion and Nuclear Safety Technology, I-00044 Frascati, Rome (Italy); EUROfusion Consortium, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Popovichev, Sergey [Culham Centre for Fusion Energy, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); EUROfusion Consortium, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Snoj, Luka [Reactor Physics Department, Jožef Stefan Institute, Jamova cesta 39, SI-1000 Ljubljana (Slovenia); EUROfusion Consortium, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom)

    2017-03-01

    At the Joint European Torus (JET) the ex-vessel fission chambers and in-vessel activation detectors are used as the neutron production rate and neutron yield monitors respectively. In order to ensure that these detectors produce accurate measurements they need to be experimentally calibrated. A new calibration of neutron detectors to 14 MeV neutrons, resulting from deuterium–tritium (DT) plasmas, is planned at JET using a compact accelerator based neutron generator (NG) in which a D/T beam impinges on a solid target containing T/D, producing neutrons by DT fusion reactions. This paper presents the analysis that was performed to model the neutron source characteristics in terms of energy spectrum, angle–energy distribution and the effect of the neutron generator geometry. Different codes capable of simulating the accelerator based DT neutron sources are compared and sensitivities to uncertainties in the generator's internal structure analysed. The analysis was performed to support preparation to the experimental measurements performed to characterize the NG as a calibration source. Further extensive neutronics analyses, performed with this model of the NG, will be needed to support the neutron calibration experiments and take into account various differences between the calibration experiment and experiments using the plasma as a source of neutrons.

  9. Improved Bonner sphere neutron spectrometry measurements for the nuclear industry

    Science.gov (United States)

    Roberts, N. J.; Thomas, D. J.; Visser, T. P. P.

    2017-11-01

    A novel, two-stage approach has been developed for producing the a priori spectrum for Bonner sphere unfolding in a case where neutrons are produced by spontaneous fission and (α,n) reactions, e.g. in UF6. The code SOURCES 4C is first used to obtain the energy spectrum of the neutrons inside the material, which is then fed into a MCNP model of the entire geometry to derive the neutron spectrum at the location of the Bonner sphere. Using this as the a priori spectrum produces a much more detailed unfolded Bonner sphere spectrum retaining fine structure from the calculation that would not be present if a simple estimated spectrum had been used as the a priori spectrum. This is illustrated using a Bonner sphere measurement of the neutron energy spectrum produced by a 48Y cylinder of UF6. From the unfolded spectrum an estimate has been made of the neutron ambient dose equivalent, i.e. the quantity which a neutron survey instrument should measure. The difference in the ambient dose equivalent of the unfolded spectrum is over 10% when using the novel approach instead of using a simpler estimate consisting of a single high energy peak, 1/E continuum, and thermal peak.

  10. Measuring neutron spectra in radiotherapy using the nested neutron spectrometer

    Energy Technology Data Exchange (ETDEWEB)

    Maglieri, Robert, E-mail: robert.maglieri@mail.mcgill.ca; Evans, Michael; Seuntjens, Jan; Kildea, John [Medical Physics Unit, McGill University, Montreal, Quebec H4A 3J1 (Canada); Licea, Angel [Canadian Nuclear Safety Commission, Ottawa, Ontario K1P 5S9 (Canada)

    2015-11-15

    Purpose: Out-of-field neutron doses resulting from photonuclear interactions in the head of a linear accelerator pose an iatrogenic risk to patients and an occupational risk to personnel during radiotherapy. To quantify neutron production, in-room measurements have traditionally been carried out using Bonner sphere systems (BSS) with activation foils and TLDs. In this work, a recently developed active detector, the nested neutron spectrometer (NNS), was tested in radiotherapy bunkers. Methods: The NNS is designed for easy handling and is more practical than the traditional BSS. Operated in current-mode, the problem of pulse pileup due to high dose-rates is overcome by measuring current, similar to an ionization chamber. In a bunker housing a Varian Clinac 21EX, the performance of the NNS was evaluated in terms of reproducibility, linearity, and dose-rate effects. Using a custom maximum-likelihood expectation–maximization algorithm, measured neutron spectra at various locations inside the bunker were then compared to Monte Carlo simulations of an identical setup. In terms of dose, neutron ambient dose equivalents were calculated from the measured spectra and compared to bubble detector neutron dose equivalent measurements. Results: The NNS-measured spectra for neutrons at various locations in a treatment room were found to be consistent with expectations for both relative shape and absolute magnitude. Neutron fluence-rate decreased with distance from the source and the shape of the spectrum changed from a dominant fast neutron peak near the Linac head to a dominant thermal neutron peak in the moderating conditions of the maze. Monte Carlo data and NNS-measured spectra agreed within 30% at all locations except in the maze where the deviation was a maximum of 40%. Neutron ambient dose equivalents calculated from the authors’ measured spectra were consistent (one standard deviation) with bubble detector measurements in the treatment room. Conclusions: The NNS may

  11. Neutron-gamma discrimination of boron loaded plastic scintillator

    International Nuclear Information System (INIS)

    Wang Dong; He Bin; Zhang Quanhu; Wu Chuangxin; Luo Zhonghui

    2010-01-01

    Boron loaded plastic scintillator could detect both fast neutrons thanks to hydrogen and thermal neutrons thanks to 10B. Both reactions have large cross sections, and results in high detection efficiency of incident neutrons. However, similar with other organic scintillators, boron loaded plastic scintillator is sensitive to gamma rays and neutrons. So gamma rays must be rejected from neutrons using their different behavior in the scintillator. In the present research zero crossing method was used to test neutron-gamma discrimination of BC454 boron loaded plastic scintillator. There are three Gaussian peaks in the time spectrum, they are corresponding to gamma rays, fast neutrons and flow neutrons respectively. Conclusion could be made that BC454 could clear discriminate slow neutrons and gamma, but the discrimination performance turns poor as the neutrons' energy becomes larger. (authors)

  12. On the calculation of multi-group fission spectrum vectors

    International Nuclear Information System (INIS)

    Mueller, E.Z.

    1984-05-01

    In this report, the problem of calculating fission spectrum vectors in a consistent manner is formulated. The practical implications of using fission spectrum vectors in multi-group transport calculations are also addressed. The significance of the weighting spectra used for the calculation of fission spectrum vectors is illustrated for the case of a simple neutronic assembly

  13. Neutron detector

    Science.gov (United States)

    Stephan, Andrew C [Knoxville, TN; Jardret,; Vincent, D [Powell, TN

    2011-04-05

    A neutron detector has a volume of neutron moderating material and a plurality of individual neutron sensing elements dispersed at selected locations throughout the moderator, and particularly arranged so that some of the detecting elements are closer to the surface of the moderator assembly and others are more deeply embedded. The arrangement captures some thermalized neutrons that might otherwise be scattered away from a single, centrally located detector element. Different geometrical arrangements may be used while preserving its fundamental characteristics. Different types of neutron sensing elements may be used, which may operate on any of a number of physical principles to perform the function of sensing a neutron, either by a capture or a scattering reaction, and converting that reaction to a detectable signal. High detection efficiency, an ability to acquire spectral information, and directional sensitivity may be obtained.

  14. Fusion neutronics

    CERN Document Server

    Wu, Yican

    2017-01-01

    This book provides a systematic and comprehensive introduction to fusion neutronics, covering all key topics from the fundamental theories and methodologies, as well as a wide range of fusion system designs and experiments. It is the first-ever book focusing on the subject of fusion neutronics research. Compared with other nuclear devices such as fission reactors and accelerators, fusion systems are normally characterized by their complex geometry and nuclear physics, which entail new challenges for neutronics such as complicated modeling, deep penetration, low simulation efficiency, multi-physics coupling, etc. The book focuses on the neutronics characteristics of fusion systems and introduces a series of theories and methodologies that were developed to address the challenges of fusion neutronics, and which have since been widely applied all over the world. Further, it introduces readers to neutronics design’s unique principles and procedures, experimental methodologies and technologies for fusion systems...

  15. Neutron spectometers

    International Nuclear Information System (INIS)

    Poortmans, F.

    1977-01-01

    Experimental work in the field of low-energy neutron physics can be subdivided into two classes: 1)Study of the decay process of the compound-nucleus state as for example the study of the capture gamma rays and of the neutron induced fission process; 2)Study of the reaction mechanism, mainly by measuring the reaction cross-sections and resonance parameters. These neutron cross-sections and resonance parameters are also important data required for many technological applications especially for reactor development programmes. In general, the second class of experiments impose other requirements on the neutron spectrometer than the first class. In most cases, a better neutron energy resolution and a broader neutron energy range are required for the study of the reaction mechanism than for the study of various aspects of the decay process. (author)

  16. A prototype detector using the neutron image intensifier and multi-anode type photomultiplier tube for pulsed neutron imaging

    International Nuclear Information System (INIS)

    Ishikawa, Hirotaku; Sato, Hirotaka; Hara, Kaoru Y.; Kamiyama, Takashi

    2016-01-01

    We developed a neutron two-dimensional (2-D) detector for pulsed neutron imaging as a prototype detector, which was composed of a neutron image intensifier and a multi-anode type photomultiplier tube. A neutron transmission spectrum of α-Fe plate was measured by the prototype detector, and compared with the one measured by a typical neutron 2-D detector. The spectrum was in reasonable agreement with the one measured by the typical detector in the neutron wavelength region above 0.15 nm. In addition, a neutron transmission image of a cadmium indicator was obtained by the prototype detector. The usefulness of the prototype detector for pulsed neutron imaging was demonstrated. (author)

  17. Neutron exposure

    International Nuclear Information System (INIS)

    Prillinger, G.; Konynenburg, R.A. van

    1998-01-01

    As a result of the popularity of the Agencies report 'Neutron Irradiation Embrittlement of Reactor Pressure Vessel Steels' of 1975, it was decided that another report on this broad subject would be of use. In this report, background and contemporary views on specially identified areas of the subject are considered as self-contained chapters, written by experts. In chapter 6, LWR-PV neutron transport calculations and dosimetry methods and how they are combined to evaluate the neutron exposure of the steel of pressure vessels are discussed. An effort to correlate neutron exposure parameters with damage is made

  18. Atmospheric neutrons

    International Nuclear Information System (INIS)

    Preszler, A.M.; Moon, S.; White, R.S.

    1976-01-01

    Additional calibrations of the University of California double-scatter neutron and additional analysis corrections lead to the slightly changed neutron fluxes reported here. The theoretical angular distributions of Merker (1975) are in general agreement with our experimental fluxes but do not give the peaks for vertical upward and downward moving neutrons. The theoretical neutron escape current J 2 /sub pi/ (Merker, 1972; Armstrong et al., 1973) is in agreement with the experimental values from 10 to 100 MeV. Our experimental fluxes agree with those of the Kanbach et al. (1974) in the overlap region from 70 to 100 MeV

  19. Neutron Albedo

    CERN Document Server

    Ignatovich, V K

    2005-01-01

    A new, algebraic, method is applied to calculation of neutron albedo from substance to check the claim that use of ultradispersive fuel and moderator of an active core can help to gain in size and mass of the reactor. In a model of isotropic distribution of incident and reflected neutrons it is shown that coherent scattering on separate grains in the case of thermal neutrons increases transport cross section negligibly, however it decreases albedo from a wall of finite thickness because of decrease of substance density. A visible increase of albedo takes place only for neutrons with wave length of the order of the size of a single grain.

  20. Spectrum Recombination.

    Science.gov (United States)

    Greenslade, Thomas B., Jr.

    1984-01-01

    Describes several methods of executing lecture demonstrations involving the recombination of the spectrum. Groups the techniques into two general classes: bringing selected portions of the spectrum together using lenses or mirrors and blurring the colors by rapid movement or foreshortening. (JM)

  1. Average radiation weighting factors for specific distributed neutron spectra

    International Nuclear Information System (INIS)

    Ninkovic, M.M.; Raicevic, J.J.

    1993-01-01

    Spectrum averaged radiation weighting factors for 6 specific neutron fields in the environment of 3 categories of the neutron sources (fission, spontaneous fission and (α,n)) are determined in this paper. Obtained values of these factors are greater 1.5 to 2 times than the corresponding quality factors used for the same purpose until a few years ago. This fact is very important to have in mind in the conversion of the neutron fluence into the neutron dose equivalent. (author)

  2. The effect of fast neutrons, as compared with X-rays upon mutation spectrum and mutation frequency in Arabidopsis thaliana (L.) Heynh. and Hordeum vulgare L. in relation to evaluation of the BARN-reactor

    International Nuclear Information System (INIS)

    Dellaert, L.M.W.

    1980-01-01

    Explanations were sought for the 'saturation' in mutant frequency, observed after relatively high irradiation doses (fast neutrons and X-rays) in Arabidopsis thaliana (L.) Heynh, when scoring for mutants is done in the siliques (Mueller's embryotest) of the 'main' inflorescence of M 1 -plants. Studies have been carried out on the effect of the presence of dithiothreitol (DTT) during irradiation, on fast neutron and X-ray induced M 1 -ovule sterility, M 2 -embryonic lethals, M 2 -chlorophyll mutants and M 2 -viable mutants in Arabidopsis thaliana. It was found that DTT provides considerable protection against both fast neutron and X-ray induced genetic damage. (Auth.)

  3. Spectral correction factors for conventional neutron dosemeters used in high-energy neutron environments

    International Nuclear Information System (INIS)

    Lee, K.W.; Sheu, R.J.

    2015-01-01

    High-energy neutrons (>10 MeV) contribute substantially to the dose fraction but result in only a small or negligible response in most conventional moderated-type neutron detectors. Neutron dosemeters used for radiation protection purpose are commonly calibrated with 252 Cf neutron sources and are used in various workplace. A workplace-specific correction factor is suggested. In this study, the effect of the neutron spectrum on the accuracy of dose measurements was investigated. A set of neutron spectra representing various neutron environments was selected to study the dose responses of a series of Bonner spheres, including standard and extended-range spheres. By comparing 252 Cf-calibrated dose responses with reference values based on fluence-to-dose conversion coefficients, this paper presents recommendations for neutron field characterisation and appropriate correction factors for responses of conventional neutron dosemeters used in environments with high-energy neutrons. The correction depends on the estimated percentage of high-energy neutrons in the spectrum or the ratio between the measured responses of two Bonner spheres (the 4P6-8 extended-range sphere versus the 6'' standard sphere). (authors)

  4. The study of prompt neutron spectra of 238U fission induced by fast neutron

    International Nuclear Information System (INIS)

    Li Anli; Bai Xixiang; Wang Yufeng; Wang Xiaozhong; Men Jiangchen; Huang Shengnian

    1990-01-01

    The measurements of prompt neutron time-of-flight spectra of U fission induced by 11 MeV neutrons were carried out at HI-13 Tandem Van de Graaff Accelerator Laboratory in 1989. The block diagram of the electronics is shown. A fission neutron TOF spectrum for the sixth section of the fission plates and the left detector at low bias is given. The data accumulation time is 60 h

  5. Fast neutron dosimetry in research reactors

    International Nuclear Information System (INIS)

    Eckert, R.

    1960-01-01

    This work chiefly concerns the measurement of fast neutron fluxes by means of threshold detectors. It is shown first that the cross sections to use for measurements by threshold detectors depend largely on the neutron spectrum, that is the position in which the measurement is performed. The spectrum is determined by calculation for several positions in the piles EL2 and EL3; from this can be deduced the cross-sections to be used for the measurements carried out in these positions. In the last part of the report, possible methods for the experimental determination of the spectrum are indicated. (author) [fr

  6. New neutron detector based on micromegas technology for ADS projects

    International Nuclear Information System (INIS)

    Andriamonje, Samuel; Andriamonje, Gregory; Aune, Stephan; Ban, Gilles; Breaud, Stephane; Blandin, Christophe; Ferrer, Esther; Geslot, Benoit; Giganon, Arnaud; Giomataris, Ioannis; Jammes, Christian; Kadi, Yacine; Laborie, Philippe; Lecolley, Jean Francois; Pancin, Julien; Riallot, Marc; Rosa, Roberto; Sarchiapone, Lucia; Steckmeyer, Jean Claude; Tillier, Joel

    2006-01-01

    A new neutron detector based on Micromegas technology has been developed for the measurement of the simulated neutron spectrum in the ADS project. After the presentation of simulated neutron spectra obtained in the interaction of 140 MeV protons with the spallation target inside the TRIGA core, a full description of the new detector configuration is given. The advantage of this detector compared to conventional neutron flux detectors and the results obtained with the first prototype at the CELINA 14 MeV neutron source facility at CEA-Cadarache are presented. The future developments of operational Piccolo-Micromegas for fast neutron reactors are also described

  7. New neutron detector based on micromegas technology for ADS projects

    Energy Technology Data Exchange (ETDEWEB)

    Andriamonje, Samuel [CEA-Saclay, DSM/DAPNIA, F-91191 Gif-sur-Yvette (France)]. E-mail: sandriamonje@cea.fr; Andriamonje, Gregory [IXL-Universite Bordeaux 1-BAT. A31-351 cours de la Liberation-F-33405 Talence Cedex (France); Aune, Stephan [CEA-Saclay, DSM/DAPNIA, F-91191 Gif-sur-Yvette (France); Ban, Gilles [CNRS/IN2P3 LPC Caen, 6 Boulevard Marechal Juin, F-14050 Caen Cedex (France); Breaud, Stephane [CEA/DEN/Cadarache, 13108 Saint-Paul Lez Durance (France); Blandin, Christophe [CEA/DEN/Cadarache, 13108 Saint-Paul Lez Durance (France); Ferrer, Esther [CEA-Saclay, DSM/DAPNIA, F-91191 Gif-sur-Yvette (France); Geslot, Benoit [CEA/DEN/Cadarache, 13108 Saint-Paul Lez Durance (France); Giganon, Arnaud [CEA-Saclay, DSM/DAPNIA, F-91191 Gif-sur-Yvette (France); Giomataris, Ioannis [CEA-Saclay, DSM/DAPNIA, F-91191 Gif-sur-Yvette (France); Jammes, Christian [CEA/DEN/Cadarache, 13108 Saint-Paul Lez Durance (France); Kadi, Yacine [CERN CH 1211 Geneva (Switzerland); Laborie, Philippe [CNRS/IN2P3 LPC Caen, 6 Boulevard Marechal Juin, F-14050 Caen Cedex (France); Lecolley, Jean Francois [CNRS/IN2P3 LPC Caen, 6 Boulevard Marechal Juin, F-14050 Caen Cedex (France); Pancin, Julien [CEA-Saclay, DSM/DAPNIA, F-91191 Gif-sur-Yvette (France); Riallot, Marc [CEA-Saclay, DSM/DAPNIA, F-91191 Gif-sur-Yvette (France); Rosa, Roberto [ENEA-Casaccia, Via Anguillarese, 00060 Rome (Italy); Sarchiapone, Lucia [CERN CH 1211 Geneva (Switzerland); Steckmeyer, Jean Claude [CNRS/IN2P3 LPC Caen, 6 Boulevard Marechal Juin, F-14050 Caen Cedex (France); Tillier, Joel [CNRS/IN2P3 LPC Caen, 6 Boulevard Marechal Juin, F-14050 Caen Cedex (France)

    2006-06-23

    A new neutron detector based on Micromegas technology has been developed for the measurement of the simulated neutron spectrum in the ADS project. After the presentation of simulated neutron spectra obtained in the interaction of 140 MeV protons with the spallation target inside the TRIGA core, a full description of the new detector configuration is given. The advantage of this detector compared to conventional neutron flux detectors and the results obtained with the first prototype at the CELINA 14 MeV neutron source facility at CEA-Cadarache are presented. The future developments of operational Piccolo-Micromegas for fast neutron reactors are also described.

  8. Microstructure-tensile property correlation of 316SS in low dose neutron irradiations

    International Nuclear Information System (INIS)

    Yoshida, N.; Muroga, T.; Araki, K.; Heinisch, H.L.; Kiritani, M.

    1990-05-01

    The objective of this work is to determine the effects of the neutron spectrum on radiation-induced changes in mechanical properties for metals irradiated with fission and fusion neutrons. 10 refs., 6 figs

  9. Tests of Neutron Spectrum Calculations with the Help of Foil Measurements in a D{sub 2}O and in an H{sub 2}O-Moderated Reactor and in Reactor Shields of Concrete an Iron

    Energy Technology Data Exchange (ETDEWEB)

    Nilsson, R; Aalto, E

    1964-09-15

    Foil measurements covering the fast, epithermal and thermal neutron energy regions have been made in the centre of the Swedish D{sub 2}O-moderated reactor R1, in the pool reactor R2-0, and in different positions in reactor shields of iron, magnetite concrete and ordinary concrete. Neutron spectra have also been calculated for most of these positions, often with the help of a numerical integration of the Boltzmann equation. The measurements and the calculated spectra are presented.

  10. Neutron dosimetry in boron neutron capture therapy

    International Nuclear Information System (INIS)

    Fairchild, R.G.; Miola, U.J.; Ettinger, K.V.

    1981-01-01

    The recent development of various borated compounds and the utilization of one of these (Na 2 B 12 H 11 SH) to treat brain tumors in clinical studies in Japan has renewed interest in neutron capture therapy. In these procedures thermal neutrons interact with 10 B in boron containing cells through the 10 B(n,α) 7 Li reaction producing charged particles with a maximum range of approx. 10μm in tissue. Borated analogs of chlorpromazine, porphyrin, thiouracil and deoxyuridine promise improved tumor uptake and blood clearance. The therapy beam from the Medical Research Reactor in Brookhaven contains neutrons from a modified and filtered fission spectrum and dosimetric consequences of the use of the above mentioned compounds in conjunction with thermal and epithermal fluxes are discussed in the paper. One of the important problems of radiation dosimetry in capture therapy is determination of the flux profile and, hence, the dose profile in the brain. This has been achieved by constructing a brain phantom made of TE plastic. The lyoluminescence technique provides a convenient way of monitoring the neutron flux distributions; the detectors for this purpose utilize 6 Li and 10 B compounds. Such compounds have been synthesized specially for the purpose of dosimetry of thermal and epithermal beams. In addition, standard lyoluminescent phosphors, like glutamine, could be used to determine the collisional component of the dose as well as the contribution of the 14 N(n,p) 14 C reaction. Measurements of thermal flux were compared with calculations and with measurements done with activation foils

  11. Neutronic reactor

    International Nuclear Information System (INIS)

    Wende, C.W.J.

    1976-01-01

    A safety rod for a nuclear reactor has an inner end portion having a gamma absorption coefficient and neutron capture cross section approximately equal to those of the adjacent shield, a central portion containing materials of high neutron capture cross section and an outer end portion having a gamma absorption coefficient at least equal to that of the adjacent shield

  12. Neutron polarization

    International Nuclear Information System (INIS)

    Firk, F.W.K.

    1976-01-01

    Some recent experiments involving polarized neutrons are discussed; they demonstrate how polarization studies provide information on fundamental aspects of nuclear structure that cannot be obtained from more traditional neutron studies. Until recently, neutron polarization studies tended to be limited either to very low energies or to restricted regions at higher energies, determined by the kinematics of favorable (p, vector n) and (d, vector n) reactions. With the advent of high intensity pulsed electron and proton accelerators and of beams of vector polarized deuterons, this is no longer the case. One has entered an era in which neutron polarization experiments are now being carried out, in a routine way, throughout the entire range from thermal energies to tens-of-MeV. The significance of neutron polarization studies is illustrated in discussions of a wide variety of experiments that include the measurement of T-invariance in the β-decay of polarized neutrons, a search for the effects of meson exchange currents in the photo-disintegration of the deuteron, the determination of quantum numbers of states in the fission of aligned 235 U and 237 Np induced by polarized neutrons, and the double- and triple-scattering of fast neutrons by light nuclei

  13. Neutron holography

    International Nuclear Information System (INIS)

    Beynon, T.D.

    1986-01-01

    the paper concerns neutron holography, which allows an image to be constructed of the surfaces, as well as the interiors, of objects. The technique of neutron holography and its applications are described. Present and future use of the method is briefly outlined. (U.K.)

  14. A stochastic model for neutron simulation considering the spectrum and nuclear properties with continuous dependence of energy; Um modelo estocastico de simulacao neutronica considerando o espectro e propriedades nucleares com dependencia continua de energia

    Energy Technology Data Exchange (ETDEWEB)

    Camargo, Dayana Queiroz de

    2011-01-15

    This thesis has developed a stochastic model to simulate the neutrons transport in a heterogeneous environment, considering continuous neutron spectra and the nuclear properties with its continuous dependence on energy. This model was implemented using Monte Carlo method for the propagation of neutrons in different environment. Due to restrictions with respect to the number of neutrons that can be simulated in reasonable computational processing time introduced the variable control volume along the (pseudo-) periodic boundary conditions in order to overcome this problem. The choice of class physical Monte Carlo is due to the fact that it can decompose into simpler constituents the problem of solve a transport equation. The components may be treated separately, these are the propagation and interaction while respecting the laws of energy conservation and momentum, and the relationships that determine the probability of their interaction. We are aware of the fact that the problem approached in this thesis is far from being comparable to building a nuclear reactor, but this discussion the main target was to develop the Monte Carlo model, implement the code in a computer language that allows extensions of modular way. This study allowed a detailed analysis of the influence of energy on the neutron population and its impact on the life cycle of neutrons. From the results, even for a simple geometrical arrangement, we can conclude the need to consider the energy dependence, i.e. an spectral effective multiplication factor should be introduced each energy group separately. (author)

  15. 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

  16. Neutron transport

    International Nuclear Information System (INIS)

    Berthoud, Georges; Ducros, Gerard; Feron, Damien; Guerin, Yannick; Latge, Christian; Limoge, Yves; Santarini, Gerard; Seiler, Jean-Marie; Vernaz, Etienne; Coste-Delclaux, Mireille; M'Backe Diop, Cheikh; Nicolas, Anne; Andrieux, Catherine; Archier, Pascal; Baudron, Anne-Marie; Bernard, David; Biaise, Patrick; Blanc-Tranchant, Patrick; Bonin, Bernard; Bouland, Olivier; Bourganel, Stephane; Calvin, Christophe; Chiron, Maurice; Damian, Frederic; Dumonteil, Eric; Fausser, Clement; Fougeras, Philippe; Gabriel, Franck; Gagnier, Emmanuel; Gallo, Daniele; Hudelot, Jean-Pascal; Hugot, Francois-Xavier; Dat Huynh, Tan; Jouanne, Cedric; Lautard, Jean-Jacques; Laye, Frederic; Lee, Yi-Kang; Lenain, Richard; Leray, Sylvie; Litaize, Olivier; Magnaud, Christine; Malvagi, Fausto; Mijuin, Dominique; Mounier, Claude; Naury, Sylvie; Nicolas, Anne; Noguere, Gilles; Palau, Jean-Marc; Le Pallec, Jean-Charles; Peneliau, Yannick; Petit, Odile; Poinot-Salanon, Christine; Raepsaet, Xavier; Reuss, Paul; Richebois, Edwige; Roque, Benedicte; Royer, Eric; Saint-Jean, Cyrille de; Santamarina, Alain; Serot, Olivier; Soldevila, Michel; Tommasi, Jean; Trama, Jean-Christophe; Tsilanizara, Aime; Behar, Christophe; Provitina, Olivier; Lecomte, Michael; Forestier, Alain; Bender, Alexandra; Parisot, Jean-Francois; Finot, Pierre

    2013-10-01

    This bibliographical note presents a reference book which addresses the study of neutron transport in matter, the study of conditions for a chain reaction and the study of modifications of matter composition due to nuclear reactions. This book presents the main nuclear data, their measurement, assessment and processing, and the spallation. It proposes an overview of methods applied for the study of neutron transport: basic equations and their derived forms, deterministic methods and Monte Carlo method of resolution of the Boltzmann equation, methods of resolution of generalized Bateman equations, methods of time resolution of space kinetics coupled equations. It presents the main calculation codes, discusses the qualification and experimental aspects, and gives an overview of neutron transport applications: neutron transport calculation of reactors, neutron transport coupled with other disciplines, physics of fuel cycle, criticality

  17. On the mechanism of fission neutron emission

    International Nuclear Information System (INIS)

    Maerten, H.; Richter, D.; Seeliger, D.

    1986-01-01

    This review represents the present knowledge of the mechanism of prompt fission neutron emission. Starting with a brief fission process characterization related with neutron emission, possible emission mechanisms are discussed. It is emphasized that the experimental study of special mechanisms, i.e. scission neutron emission processes, requires a sufficiently correct description of emission probabilities on the base of the main mechanism, i.e. the evaporation from fully accelerated fragments. Adequate statistical-model approaches have to account for the complexity of nuclear fission reflected by an intricate fragment distribution. The present picture of scission neutron emission is not clarified neither experimentally nor theoretically. Deduced data are contradictory and depend on the used analysis procedures often involving rough discriptions of evaporated-neutron distributions. The contribution of two secondary mechanisms of fission neutron emission, i.e. the neutron evaporation during fragment acceleration and neutron emission due to the decay of 5 He after ternary fission, is estimated. We summarize the recent progress of the theoretical description of fission neutron spectra in the framework of statistical models considering the standard spectrum of 252 Cf(sf) neutrons especially. The main experimental basis for the study of fission neutron emission is the accurate measurement of emission probabilities as a function of emission energy and angle (at least) as well as fragment parameters (mass number ratio and kinetic energy). The present status is evaluated. (author)

  18. Neutron and gamma-ray spectra of 239PuBe and 241AmBe

    International Nuclear Information System (INIS)

    Vega-Carrillo, H.R.; Manzanares-Acuna, Eduardo; Becerra-Ferreiro, A.M.; Carrillo-Nunez, Aureliano

    2002-01-01

    Neutron and gamma-ray spectra of 239 PuBe and 241 AmBe were measured and their dosimetric features were calculated. Neutron spectra were measured using a multisphere neutron spectrometer with a 6 LiI(Eu) scintillator. The 239 PuBe neutron spectrum was measured in an open environment, while the 241 AmBe neutron spectrum was measured in a closed environment. Gamma-ray spectra were measured using a NaI(Tl) scintillator using the same experimental conditions for both sources. The effect of measuring conditions for the 241 AmBe neutron spectrum indicates the presence of epithermal and thermal neutrons. The low-resolution neutron spectra obtained with the multisphere spectrometer allows one to calculate the dosimetric features of neutron sources. At 100 cm both sources produce approximately the same count rate as that of the 4.4 MeV gamma-ray per unit of alpha emitter activity

  19. Measurement of spectrum at the experimental 6.5 MW reactor in Vinca; Merenje spektra na eksperimentalnom 6,5 MW reaktoru u Vinci

    Energy Technology Data Exchange (ETDEWEB)

    Berovic, N; Boreli, F; Aleksic, N; Dragin, R [Boris Kidric Institute of nuclear sciences, Vinca, Belgrade (Yugoslavia)

    1961-07-01

    Since RA reactor is supplied with horizontal experimental channels which lead directly to the core fast neutron spectrum in the channel does not differ much from the neutron spectrum in the core. Spectrum was measured by 'telescope' for detecting scattered protons. Measuring procedure together with the measured spectrum are presented in this paper.

  20. Conceptual design, neutronic and radioprotection study of a fast neutron irradiation station at SINQ

    International Nuclear Information System (INIS)

    Zanini, L.; Baluc, N.; Simone, A. De; Eichler, R.; Joray, S.; Manfrin, E.; Pouchon, M.; Rabaioli, S.; Schumann, D.; Welte, J.; Zhernosekov, K.

    2011-12-01

    This comprehensive, illustrated report by the Paul Scherrer Institute PSI in Switzerland documents the proposals concerning the conceptual design, neutronic and radioprotection study of a fast neutron irradiation station at the PSI's Swiss Spallation Neutron Source SINQ facility. The need for fast neutron irradiation is discussed and the possibility of using SINQ as a fast neutron irradiation facility is considered. The production of isotopes, tracers and medical isotopes is discussed, as are fission and fusion reactor technologies. The characteristics of the neutron spectrum in SINQ are discussed. The neutronic and radioprotection calculations for an irradiation station at SINQ are looked at in detail and extensive examples of work done and results obtained are presented and discussed. Radioprotection issues are also looked at. Further contributions in the report cover the hot/cold irradiation station in the SINQ target. An appendix provides detailed drawings of the facility's pneumatic delivery system

  1. Neutron activation system for spectral measurements of pulsed ion diode neutron production

    International Nuclear Information System (INIS)

    Hanson, D.L.; Kruse, L.W.

    1980-02-01

    A neutron energy spectrometer has been developed to study intense ion beam-target interactions in the harsh radiation environment of a relativistic electron beam source. The main component is a neutron threshold activation system employing two multiplexed high efficiency Ge(Li) detectors, an annihilation gamma coincidence system, and a pneumatic sample transport. Additional constraints on the neutron spectrum are provided by total neutron yield and time-of-flight measurements. A practical lower limit on the total neutron yield into 4π required for a spectral measurement with this system is approx. 10 10 n where the neutron yield is predominantly below 4 MeV and approx. 10 8 n when a significant fraction of the yield is above 4 MeV. Applications of this system to pulsed ion diode neutron production experiments on Hermes II are described

  2. Energy response study of modified CR-39 neutron personnel dosimeter

    International Nuclear Information System (INIS)

    Sathian, Deepa; Bakshi, A.K.; Datta, D.; Nair, Sreejith S.; Sathian, V.; Mishra, Jitendra; Sen, Meghnath

    2018-01-01

    Personnel neutron dosimetry is an integral part of radiation protection. No single dosimeter provides the satisfactory energy response, sensitivity, angular dependence characteristics and accuracy necessary to meet the requirement of an ideal personnel neutron dosimeter. The response of a personnel neutron dosimeter is critically dependent upon the energy distribution of the neutron field. CR-39 personnel neutron dosimeters were typically calibrated in the standard neutron field of 252 Cf and 241 Am-Be in our laboratory, although actual neutron fields may vary from the calibration neutron spectrum. Recently the badge cassette of the personnel neutron dosimeter was changed due to frequent damage of the PVC badge used earlier. This paper discusses energy response of CR-39 solid state nuclear track detector loaded in this modified badge cassette as per latest ISO recommendation

  3. Physics of neutron emission in fission

    International Nuclear Information System (INIS)

    Lemmel, H.D.

    1989-06-01

    The document contains the proceedings of the IAEA Consultants' Meeting on the Physics of Neutron Emission in Fission, Mito City (Japan), 24-27 May 1988. Included are the conclusions and recommendations reached at the meeting and the papers presented by the meeting participants. These papers cover the following topics: Energy dependence of the number of fission neutrons ν-bar (3 papers), multiplicity distribution of fission neutrons (3 papers), competition between neutron and γ-ray emission (4 papers), the fission neutron yield in resonances (2 papers) and the energy spectrum of fission neutrons in experiment (9 papers), theory (4 papers) and evaluation (1 paper). A separate abstract was prepared for each of these papers. Refs, figs and tabs

  4. 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.

  5. Fast neutron sensitivity of polymer dosemeters

    International Nuclear Information System (INIS)

    Harper, M.W.; Pearson, D.W.; Moran, P.R.

    1975-01-01

    The responses of polymer thermocurrent dosemeters to fission spectrum and 14 MeV neutrons were measured. The dosemeters are in the form of disks 1 cm diam by 0.5 mm thick. Relative to Cobalt 60 gamma responses, teflon PTFE dosemeters show a 6 percent response to 14 MeV neutrons and a 5 percent response to fission neutrons on a tissue rad basis. Polymethylpentene dosemeters show a 49 percent response to 14 MeV neutrons and a 40 percent response to fission neutrons on a tissue rad basis when provided with adequate recoil proton buildup. The sensitivity of these dosemeters is limited to neutron doses greater than 10 rads by spurious background currents

  6. Neutron spectra of /sup 239/Pu-Be neutron sources

    Energy Technology Data Exchange (ETDEWEB)

    Kumar, A; Nagarajan, P S [Bhabha Atomic Research Centre, Bombay (India). Div. of Radiation Protection

    1977-01-01

    Neutron spectra of /sup 239/Pu-Be(..cap alpha..,n) sources have been calculated by using the most recent data on the differential cross sections and angular distributions. The contribution from the multibody break-up reaction /sup 9/Be(..cap alpha..,..cap alpha..n)/sup 8/Be has also been incorporated. Modifications to the primary spectrum due to the secondary interactions in the source such as elastic scattering with beryllium, oxygen and plutonium and the /sup 9/Be(n,2n) and /sup 239/Pu(n,f) reaction have been calculated for different strengths and geometries. The present calculation has shown that the spectrum changes considerably because of these events within the source by way of smearing of peaks and filling up of valleys and raising the low energy part of the spectrum. Increase in H/D value leads to channeling of extra neutrons into the equatorial plane at the cost of the neutrons along the axial direction. The present calculations show that inclusion of secondary interactions to the extent considered in this work does not account completely for the increased intensity in the lower energy end of the measured spectrum.

  7. Neutron therapy

    International Nuclear Information System (INIS)

    Riesler, Rudi

    1995-01-01

    Standard radiotherapy uses Xrays or electrons which have low LET (linear energy transfer); in contrast, particles such as neutrons with high LET have different radiobiological responses. In the late 1960s, clinical trials by Mary Catterall at the Hammersmith Hospital in London indicated that fast neutron radiation had clinical advantages for certain malignant tumours. Following these early clinical trials, several cyclotron facilities were built in the 1980s for fast neutron therapy, for example at the University of Washington, Seattle, and at UCLA. Most of these newer machines use extracted cyclotron proton beams in the range 42 to 66 MeV with beam intensities of 15 to 60 microamps. The proton beams are transported to dedicated therapy rooms, where neutrons are produced from beryllium targets. Second-generation clinical trials showed that accurate neutron beam delivery to the tumour site is more critical than for photon therapy. In order to achieve precise beam geometries, the extracted proton beams have to be transported through a gantry which can rotate around the patient and deliver beams from any angle; also the neutron beam outline (''field shape'') must be adjusted to extremely irregular shapes using a flexible collimation system. A therapy procedure has to be appropriately organized, with physicians, radiotherapists, nurses, medical physicists and other staff in attendance; other specialized equipment, such as CT or MRI scanners and radiation simulators must be made available. Neutron therapy is usually performed only in radiation oncology departments of major medical centres

  8. Neutron radiography

    International Nuclear Information System (INIS)

    Alaa eldin, M.T.

    2011-01-01

    The digital processing of the neutron radiography images gives the possibility for data quantification. In this case an exact relation between the measured neutron attenuation and the real macroscopic attenuation coefficient for every point of the sample is required. The assumption that the attenuation of the neutron beam through the sample is exponential is valid only in an ideal case where a monochromatic beam, non scattering sample and non background contribution are assumed. In the real case these conditions are not fulfilled and in dependence on the sample material we have more or less deviation from the exponential attenuation law. Because of the high scattering cross-sections of hydrogen (σs=80.26 barn) for thermal neutrons, the problem with the scattered neutrons at quantitative radiography investigations of hydrogenous materials (as PE, Oil, H 2 O, etc) is not trivial. For these strong scattering materials the neutron beam attenuation is no longer exponential and a dependence of the macroscopic attenuation coefficient on the material thickness and on the distance between the sample and the detector appears. When quantitative radiography (2 D) or tomography investigations (3 D) are performed, some image correction procedures for a description of the scattering effect are required. This thesis presents a method that can be used to enhance the neutron radiography image for objects with high scattering materials like hydrogen, carbon and other light materials. This method uses the Monte Carlo code, MCNP5, to simulate the neutron radiography process and get the flux distribution for each pixel of the image and determine the scattered neutrons distribution that causes the image blur and then subtract it from the initial image to improve its quality.

  9. Specification of fast neutron radiation quality from cell transformation data

    International Nuclear Information System (INIS)

    Coppola, M.

    1992-01-01

    Experimental data on the neoplastic transformation of C3H 10T1/2 cells measured at Casaccia after neutron and X-ray irradiation were used to determine neutron RBE values for the RSV-Tapiro fast reactor energy spectrum and for monoenergetic neutrons of 0.5, 1, and 6 MeV. In parallel, micro-dosimetric measurements provided the actual lineal energy distributions and related mean parameters for the reactor radiation. From these experiments, values of the neutron quality factor were derived for the reactor neutron energy spectrum and, in turn, for the other neutron energies tested. A mathematical expression giving a smooth dependence on neutron energy was also determined for the effective quality factor in the entire energy range examined. The results were compared with other proposals

  10. Neutronics codes

    International Nuclear Information System (INIS)

    Buckel, G.

    1983-01-01

    The objectives are the development, testing and cultivation of reliable, efficient and user-optimized neutron-physical calculation methods and conformity with users' requirements concerning design of power reactors, planning and analysis of experiments necessary for their protection as well as research on physical key problems. A short outline of available computing programmes for the following objectives is given: - Provision of macroscopic group constants, - Calculation of neutron flux distribution in transport theory and diffusion approximation, - Evaluation of neutron flux-distribution, - Execution of disturbance calculations for the determination reactivity coefficients, and - graphical representation of results. (orig./RW) [de

  11. Neutron microdosimetry at RARAF

    International Nuclear Information System (INIS)

    Kliauga, P.

    1986-01-01

    A comprehensive series of measurements of neutron microdosimetry spectra is underway at the RARAF facility. The neutrons generated at RARAF are semi-monoenergetic to monoenergetic, depending on energy. Thus far, measurements have concentrated on 15 MeV, with a few measurements done at 6 MeV. One of the main reasons for undertaking this project is dissatisfaction with the state of accuracy of microdosimetric measurements of neutrons, not only previous measurements done at RARAF, but reports in the literature from all over the world. Only a relatively modest amount of data has been taken for neutrons, as compared to photons, and the survey of dose mean lineal energy values done for the recent ICRU Report No. 36 (December 1983) reveals a spread of values far in excess of accepted estimates of statistical uncertainty (5-10%). One of the major motivations in undertaking this project, therefore, was to elucidate some of the factors, including experimental artifacts, which are important in contributing to systematic errors in measurements. Among the methods being employed are determination of the effect of various counter parameters on neutron spectra, and electronic parameters, also. Another important method of obtaining information is a comparison between different counters. This laboratory has access to perhaps a greater variety of microdosimetric proportional counters than any in the world, from the standard Rossi counter, to various wall-less types of differing geometries. Controlled comparisons of spectra from such differing counters using the same analysis technique can yield much information on the effect of counter geometry on the microdosimetric spectrum

  12. Use of the associated particle technique in the fast neutron spectroscopy

    International Nuclear Information System (INIS)

    Aquirre O, G.A.

    1978-01-01

    Selecting a neutrons monoenergetic source it was found that the nuclear reaction D(d,n) 3 He can be used to measure nuclear sections and differentials in elastic nuclear reactions through the associated particle technique; the neutron beam energy is directly determined in time of flight spectrum of the neutron. The flux is determined by the number of 3 He ions observed in the charged particle spectrum. The neutron flux can be increased increasing the solid angle of the neutrons beam in two magnitude orders according to the results of neutrons beam profile measures. (author)

  13. Quality factors for monoenergetic neutrons

    International Nuclear Information System (INIS)

    Cross, W.G.; Ing, H.

    1984-01-01

    Mean quality factors anti Q(E/sub n/), for the dose resulting from first interactons of monoenergetic neutrons in tissue, have been calculated at energies from thermal to 14.7 MeV. Knowledge of these values, along with kerma factors, allows the calculation of the heavy-particle dose equivalent for any known neutron spectrum in tissue. The partial quality factors for the dose delivered by C, N, and O elastic and inelastic recoils are found to be virtually independent of the scattering angular distributions and are given by simple expressions

  14. The GKSS cold neutron source

    International Nuclear Information System (INIS)

    Knop, W.; Wedderien, T.; Krull, W.

    1995-01-01

    The FRG-1 research reactor, in operation since 1958 at 5 MW power, is upgraded and refurbished many times to follow the changing demands on safe operation and the today needs for scientific research. This requires during the lifetime of the reactor many measures to follow these demands. Within the last years many additional activities have been made to overcome the ageing of the experiments, to change the experimental facilities and to increase the neutron flux and adapt the neutron spectrum to ensure good scientific utilization of the research reactor for the next 15 to 20 years. (orig./HP)

  15. Neutron reflectometry

    DEFF Research Database (Denmark)

    Klösgen-Buchkremer, Beate Maria

    2014-01-01

    of desired information. In the course, an introduction into the method and an overview on selected instruments at large scale facilities will be presented. Examples will be given that illustrate the potential of the method, mostly based on organic films. Results from the investigation of layered films......Neutron (and X-ray) reflectometry constitute complementary interfacially sensitive techniques that open access to studying the structure within thin films of both soft and hard condensed matter. Film thickness starts oxide surfaces on bulk substrates, proceeding to (pauci-)molecular layers and up...... films or films with magnetic properties. The reason is the peculiar property of neutron light since the mass of a neutron is close to the one of a proton, and since it bears a magnetic moment. The optical properties of matter, when interacting with neutrons, are described by a refractive index...

  16. Neutron scattering

    International Nuclear Information System (INIS)

    Furrer, A.

    1993-01-01

    This report contains the text of 16 lectures given at the Summer School and the report on a panel discussion entitled ''the relative merits and complementarities of x-rays, synchrotron radiation, steady- and pulsed neutron sources''. figs., tabs., refs

  17. Neutron storage

    International Nuclear Information System (INIS)

    Strelkov, A.V.

    2004-01-01

    The report is devoted to neutron storage (NS) and describes the history of experiments on the NS development. Great attention is paid to ultracold neutron (UCN) storage. The experiments on the UCN generation, transport, spectroscopy, storage and detection are described. Experiments on searching the UCN electric-dipole moment and electric charge are continued. Possible using of UCN for studying the nanoparticles is discussed [ru

  18. Neutron radiography

    International Nuclear Information System (INIS)

    Bayon, G.

    1989-01-01

    Neutronography or neutron radiography, a non-destructive test method which is similar in its principle to conventional X-ray photography, presently occupies a marginal position among non-destructive test methods (NDT) (no source of suitable performance or cost). Neutron radiography associated with the ORPHEE reactor permits industrial testing; it can very quickly meet a cost requirement comparable to that of conventional test methods. In 1988, 2500 parts were tested on this unit [fr

  19. Neutron detector

    International Nuclear Information System (INIS)

    Endo, Hiroshi.

    1993-01-01

    The device of the present invention detects neutrons in a reactor container under a high temperature and reduces the noise level in an FBR type reactor. That is, the detection section comprises a high heat resistant vessel containing a scintillator therein for detecting neutrons. Neutron signals sent from the detection section are inputted to a neutron measuring section by way of a signal transmission section. The detection section is disposed at the inside of the reactor container. Further, the signal transmission section is connected optically to the detection section. With such a constitution, since the detection section comprising the high temperature resistant vessel is disposed at the inside of the reactor container, neutron fluxes can be detected and measured at high sensitivity even under a high temperature circumstance. Since the signal transmission section is optically connected to the detection section, influence of radiation rays upon transmission of the neutron detection signals can be reduced. Accordingly, the noise level can be kept low. (I.S.)

  20. Neutron Scattering

    International Nuclear Information System (INIS)

    Fayer, Michael J.; Gee, Glendon W.

    2005-01-01

    The neutron probe is a standard tool for measuring soil water content. This article provides an overview of the underlying theory, describes the methodology for its calibration and use, discusses example applications, and identifies the safety issues. Soil water makes land-based life possible by satisfying plant water requirements, serving as a medium for nutrient movement to plant roots and nutrient cycling, and controlling the fate and transport of contaminants in the soil environment. Therefore, a successful understanding of the dynamics of plant growth, nutrient cycling, and contaminant behavior in the soil requires knowledge of the soil water content as well as its spatial and temporal variability. After more than 50 years, neutron probes remain the most reliable tool available for field monitoring of soil water content. Neutron probes provide integrated measurements over relatively large volumes of soil and, with proper access, allow for repeated sampling of the subsurface at the same locations. The limitations of neutron probes include costly and time-consuming manual operation, lack of data automation, and costly regulatory requirements. As more non-radioactive systems for soil water monitoring are developed to provide automated profiling capabilities, neutron-probe usage will likely decrease. Until then, neutron probes will continue to be a standard for reliable measurements of field water contents in soils around the globe