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Sample records for thermal neutron self-shielding

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

    International Nuclear Information System (INIS)

    Tzika, F.; Stamatelatos, I.E.

    2004-01-01

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

  2. Calculation of thermal neutron self-shielding correction factors for aqueous bulk sample prompt gamma neutron activation analysis using the MCNP code

    International Nuclear Information System (INIS)

    Nasrabadi, M.N.; Jalali, M.; Mohammadi, A.

    2007-01-01

    In this work thermal neutron self-shielding in aqueous bulk samples containing neutron absorbing materials is studied using bulk sample prompt gamma neutron activation analysis (BSPGNAA) with the MCNP code. The code was used to perform three dimensional simulations of a neutron source, neutron detector and sample of various material compositions. The MCNP model was validated against experimental measurements of the neutron flux performed using a BF 3 detector. Simulations were performed to predict thermal neutron self-shielding in aqueous bulk samples containing neutron absorbing solutes. In practice, the MCNP calculations are combined with experimental measurements of the relative thermal neutron flux over the sample's surface, with respect to a reference water sample, to derive the thermal neutron self-shielding within the sample. The proposed methodology can be used for the determination of the elemental concentration of unknown aqueous samples by BSPGNAA where knowledge of the average thermal neutron flux within the sample volume is required

  3. Self-shielding coefficient and thermal flux depression factor of voluminous sample in neutron activation analysis

    International Nuclear Information System (INIS)

    Noorddin Ibrahim; Rosnie Akang

    2009-01-01

    Full text: One of the major problems encountered during the irradiation of large inhomogeneous samples in performing activation analysis using neutron is the perturbation of the neutron field due to absorption and scattering of neutron within the sample as well as along the neutron guide in the case of prompt gamma activation analysis. The magnitude of this perturbation shown by self-shielding coefficient and flux depression depend on several factors including the average neutron energy, the size and shape of the sample, as well as the macroscopic absorption cross section of the sample. In this study, we use Monte Carlo N-Particle codes to simulate the variation of neutron self-shielding coefficient and thermal flux depression factor as a function of the macroscopic thermal absorption cross section. The simulation works was carried out using the high performance computing facility available at UTM while the experimental work was performed at the tangential beam port of Reactor TRIGA PUSPATI, Malaysia Nuclear Agency. The neutron flux measured along the beam port is found to be in good agreement with the simulated data. Our simulation results also reveal that total flux perturbation factor decreases as the value of absorption increases. This factor is close to unity for low absorbing sample and tends towards zero for strong absorber. In addition, sample with long mean chord length produces smaller flux perturbation than the shorter mean chord length. When comparing both the graphs of self-shielding factor and total disturbance, we can conclude that the total disturbance of the thermal neutron flux on the large samples is dominated by the self-shielding effect. (Author)

  4. Enhancement of thermal neutron self-shielding in materials surrounded by reflectors

    International Nuclear Information System (INIS)

    Cornelia Chilian; Gregory Kennedy

    2012-01-01

    Materials containing from 41 to 1124 mg chlorine and surrounded by polyethylene containers of various thicknesses, from 0.01 to 5.6 mm, were irradiated in a research reactor neutron spectrum and the 38 Cl activity produced was measured as a function of polyethylene reflector thickness. For the material containing the higher amount of chlorine, the 38 Cl specific activity decreased with increasing reflector thickness, indicating increased neutron self-shielding. It was found that the amount of neutron self-shielding increased by as much as 52% with increasing reflector thickness. This is explained by neutrons which have exited the material subsequently reflecting back into it and thus increasing the total mean path length in the material. All physical and empirical models currently used to predict neutron self-shielding have ignored this effect and need to be modified. A method is given for measuring the adjustable parameter of a self-shielding model for a particular sample size and combination of neutron reflectors. (author)

  5. Measurement of the thermal neutron self shielding coefficient in the Syrian miniature neutron source reactor inner irradiation site using the dy soils

    International Nuclear Information System (INIS)

    Khattab, K.; Khamis, I.

    2007-01-01

    Measurement of the thermal self shielding coefficient ( Gth ) in the Syrian Miniature Neutron Source Reactor (MNSR) inner irradiation site using Dy foils is presented in this paper. The thermal self shielding coefficient is measured as a function of the foil thickness or numbers. The mathematical equation which calculates the average relative radioactivity (Bq/g) versus the foil number is found as well.

  6. Thermal testing of solid neutron shielding materials

    International Nuclear Information System (INIS)

    Boonstra, R.H.

    1992-09-01

    Two legal-weight truck casks the GA-4 and GA-9, will carry four PWR and nine BWR spent fuel assemblies, respectively. Each cask has a solid neutron shielding material separating the steel body and the outer steel skin. In the thermal accident specified by NRC regulations in 10CFR Part 71, the cask is subjected to an 800 degree C environment for 30 minutes. The neutron shield need not perform any shielding function during or after the thermal accident, but its behavior must not compromise the ability of the cask to contain the radioactive contents. In May-June 1989 the first series of full-scale thermal tests was performed on three shielding materials: Bisco Products NS-4-FR, and Reactor Experiments RX-201 and RX-207. The tests are described in Thermal Testing of Solid Neutron Shielding Materials, GA-AL 9897, R. H. Boonstra, General Atomics (1990), and demonstrated the acceptability of these materials in a thermal accident. Subsequent design changes to the cask rendered these materials unattractive in terms of weight or adequate service temperature margin. For the second test series, a material specification was developed for a polypropylene based neutron shield with a softening point of at least 280 degree F. The neutron shield materials tested were boronated (0.8--4.5%) polymers (polypropylene, HDPE, NS-4). The Envirotech and Bisco materials are not polypropylene, but were tested as potential backup materials in the event that a satisfactory polypropylene could not be found

  7. Neutron self-shielding with k0-NAA irradiations

    International Nuclear Information System (INIS)

    Chilian, C.; Chambon, R.; Kennedy, G.

    2010-01-01

    A sample of SMELS Type II reference material was mixed with powdered Cd-nitrate neutron absorber and analysed by k 0 NAA for 10 elements. The thermal neutron self-shielding effect was found to be 34.8%. When flux monitors were irradiated sufficiently far from the absorbing sample, it was found that the self-shielding could be corrected accurately using an analytical formula and an iterative calculation. When the flux monitors were irradiated 2 mm from the absorbing sample, the calculations over-corrected the concentrations by as much as 30%. It is recommended to irradiate flux monitors at least 14 mm from a 10 mm diameter absorbing sample.

  8. Thermal neutron shield and method of manufacture

    Science.gov (United States)

    Brindza, Paul Daniel; Metzger, Bert Clayton

    2013-05-28

    A thermal neutron shield comprising concrete with a high percentage of the element Boron. The concrete is least 54% Boron by weight which maximizes the effectiveness of the shielding against thermal neutrons. The accompanying method discloses the manufacture of Boron loaded concrete which includes enriching the concrete mixture with varying grit sizes of Boron Carbide.

  9. Thermal testing of solid neutron shielding materials

    International Nuclear Information System (INIS)

    Boonstra, R.H.

    1990-03-01

    The GA-4 and GA-9 spent fuel shipping casks employ a solid neutron shielding material. During a hypothetical thermal accident, any combustion of the neutron shield must not compromise the ability of the cask to contain the radioactive contents. A two-phase thermal testing program was carried out to assist in selecting satisfactory shielding materials. In the first phase, small-scale screening tests were performed on nine candidate materials using ASTM procedures. From these initial results, three of the nine candidates were chosen for inclusion in the second phase of testing, These materials were Bisco Products NS-4-FR, Reactor Experiments 201-1, and Reactor Experiments 207. In the second phase, each selected material was fabricated into a test article which simulated a full-scale of neutron shield from the cask. The test article was heated in an environmental prescribed by NRC regulations. Results of this second testing phase showed that all three materials are thermally acceptable

  10. Thermal testing of solid neutron shielding materials

    International Nuclear Information System (INIS)

    Boonstra, R.N.

    1990-01-01

    The GA-4 and GA-9 spent fuel shipping casks employ a solid neutron shielding material. During a hypothetical thermal accident, any combustion of the neutron shield must not compromise the ability of the cask to contain the radioactive contents. A two-phase thermal testing program was carried out to assist in selecting satisfactory shielding materials. In the first phase, small-scale screening tests were performed on nine candidate materials using ASTM procedures. From these initial results, three of the nine candidates were chosen for inclusion in the second phase of testing. These materials were Bisco Products NS-4-FR, Reactor Experiments 201-1, and Reactor Experiments 207. In the second phase, each selected material was fabricated into a test article which simulated a full-scale section of neutron shield from the cask. The test article was heated in an environment prescribed by NRC regulations. Results of this second testing phase show that all three materials are thermally acceptable

  11. Thermal testing of solid neutron shielding materials

    International Nuclear Information System (INIS)

    Boonstra, R.H.

    1993-01-01

    In May-June 1989 the first series of full-scale thermal tests was performed on three shielding materials: Bisco Products NS-4-FR, and Reactor Experiments RX-201 and RX-207. The tests are described in Thermal Testing of Solid Neutron Shielding Materials, GA-A19897, R.H. Boonstra, General Atomics (1990), and demonstrated the acceptability of these materials in a thermal accident. Subsequent design changes to the cask rendered these materials unattractive in terms of weight or adequate service temperature margin. For the second test series a material specification was developed for a polypropylene based neutron shield with a softening point of at least 280degF. Table 1 lists the neutron shield materials tested. The Envirotech and Bisco materials are not polypropylene, but were tested as potential backup materials in the event that a satisfactory polypropylene could not be found. The Bisco modified NS-4 and Reactor Experiments HMPP are both acceptable materials from a thermal accident standpoint for use in the shipping cask. Tests of the Kobe PP-R01 and Envirotech HDPE were stopped for safety reasons, due to inability to deal with the heavy smoke, before completion of the 30-minute heating phase. However these materials may prove satisfactory if they could undergo the complete heating. (J.P.N.)

  12. Thermal, epithermal and thermalized neutron attenuation properties of ilmenite-serpentine heat resistant concrete shield

    International Nuclear Information System (INIS)

    Kany, A.M.I.; El-Gohary, M.I.; Kamal, S.M.

    1994-01-01

    Experimental measurements were carried out to study the attenuation properties of low-energy neutrons transmitted through unheated and preheated barriers of heavy-weight, highly hydrated and heat-resistant concrete shields. The concrete shields under investigation have been prepared from naturally occurring ilmenite and serpentine Egyptian ores. A collimated beam obtained from an Am-Be source was used as a source of neutrons, while the measurements of total thermal, epithermal, and thermalized neutron fluxes were performed using a BF-3 detector, multichannel analyzer and Cd filter. Results show that the ilmenite-serpentine concrete proved to be a better thermal, epithermal and thermalized neutron attenuator than the ordinary concrete especially at a high temperature of concrete exposure. (Author)

  13. Self-shielding for thick slabs in a converging neutron beam

    CERN Document Server

    Mildner, D F R

    1999-01-01

    We have previously given a correction to the neutron self-shielding for a thin slab to account for the increased average path length through the slab when irradiated in a converging neutron beam. This expression overstates the case for the self-shielding for a thick (or highly absorbing) slab. We give a better approximation to the increase in effective shielding correction for a slab placed in a converging neutron beam. It is negligible at large absorption mean free paths. (author)

  14. Calculation of self-shielding coefficients, flux depression and cadmium factor for thermal neutron flux measurement of the IPEN/MB-01 reactor

    International Nuclear Information System (INIS)

    Marques, Andre Luis Ferreira; Ting, Daniel Kao Sun; Mendonca, Arlindo Gilson

    1996-01-01

    A calculation methodology of Flux Depression, Self-Shielding and Cadmium Factors is presented, using the ANISN code, for experiments conducted at the IPEN/MB-01 Research Reactor. The correction factors were determined considering thermal neutron flux and 0.125 e 0.250 mm diameter of 197 Au wires. (author)

  15. Optimization of thermal neutron shield concrete mixture using artificial neural network

    Energy Technology Data Exchange (ETDEWEB)

    Yadollahi, A. [Engineering Department, Shahid Beheshti University, G.C., P.O. Box: 1983963113, Tehran (Iran, Islamic Republic of); Nazemi, E., E-mail: nazemi.ehsan@yahoo.com [Young Researchers and Elite Club, Kermanshah Branch, Islamic Azad University, Kermanshah (Iran, Islamic Republic of); Zolfaghari, A. [Engineering Department, Shahid Beheshti University, G.C., P.O. Box: 1983963113, Tehran (Iran, Islamic Republic of); Ajorloo, A.M. [Water and Environmental Engineering Department, Shahid Beheshti University, P.O. Box: 167651719, Tehran (Iran, Islamic Republic of)

    2016-08-15

    Highlights: • Colemanite was used in fabricating of thermal neutron shield concrete. • The Taguchi method was implemented to obtain the data set required for training the ANN. • Trained ANN predicted quality characteristics of thermal neutron shield. - Abstract: Colemanite is the most convenient boron mineral which has been widely used in construction of radiation shielding concrete in order to improve the capture of thermal neutrons. But utilization of Colemanite in radiation shielding concrete has a deleterious effect on both physical and mechanical properties. In the present work, Taguchi method and artificial neural network (ANN) were employed to find an optimal mixture of Colemanite based concrete in order to improve the boron content of concrete and increase thermal neutron absorption without violating the standards for physical and mechanical properties. Using Taguchi method for experimental design, 27 concrete samples with different mixtures were fabricated and tested. Water/cement ratio, cement quantity, volume fraction of Colemanite aggregate and silica fume quantity were selected as control factors, and compressive strength, ultrasonic pulse velocity and thermal neutron transmission ratio were considered as the quality responses. Obtained data from 27 experiments were used to train 3 ANNs. Four control factors were utilized as the inputs of 3 ANNs and 3 quality responses were used as the outputs, separately (each ANN for one quality response). After training the ANNs, 1024 different mixtures with different quality responses were predicted. At the final, optimum mixture was obtained among the predicted different mixtures. Results demonstrated that the optimal mixture of thermal neutron shielding concrete has a water–cement ratio of 0.38, cement content of 400 kg/m{sup 3}, a volume fraction Colemanite aggregate of 50% and silica fume–cement ratio of 0.15.

  16. Optimization of thermal neutron shield concrete mixture using artificial neural network

    International Nuclear Information System (INIS)

    Yadollahi, A.; Nazemi, E.; Zolfaghari, A.; Ajorloo, A.M.

    2016-01-01

    Highlights: • Colemanite was used in fabricating of thermal neutron shield concrete. • The Taguchi method was implemented to obtain the data set required for training the ANN. • Trained ANN predicted quality characteristics of thermal neutron shield. - Abstract: Colemanite is the most convenient boron mineral which has been widely used in construction of radiation shielding concrete in order to improve the capture of thermal neutrons. But utilization of Colemanite in radiation shielding concrete has a deleterious effect on both physical and mechanical properties. In the present work, Taguchi method and artificial neural network (ANN) were employed to find an optimal mixture of Colemanite based concrete in order to improve the boron content of concrete and increase thermal neutron absorption without violating the standards for physical and mechanical properties. Using Taguchi method for experimental design, 27 concrete samples with different mixtures were fabricated and tested. Water/cement ratio, cement quantity, volume fraction of Colemanite aggregate and silica fume quantity were selected as control factors, and compressive strength, ultrasonic pulse velocity and thermal neutron transmission ratio were considered as the quality responses. Obtained data from 27 experiments were used to train 3 ANNs. Four control factors were utilized as the inputs of 3 ANNs and 3 quality responses were used as the outputs, separately (each ANN for one quality response). After training the ANNs, 1024 different mixtures with different quality responses were predicted. At the final, optimum mixture was obtained among the predicted different mixtures. Results demonstrated that the optimal mixture of thermal neutron shielding concrete has a water–cement ratio of 0.38, cement content of 400 kg/m 3 , a volume fraction Colemanite aggregate of 50% and silica fume–cement ratio of 0.15.

  17. Self-shielding factors for TLD-600 and TLD-100 in an isotropic flux of thermal neutrons

    International Nuclear Information System (INIS)

    Horowitz, Y.S.; Dubi, A.; Ben Shahar, B.

    1976-01-01

    The applications of lithium fluoride thermoluminescent dosemeters in mixed n-γ environments, and the dependence of LiF-TL on linear energy transfer are both topics of current interest. Monte Carlo calculations have therefore been carried out to determine the thermal neutron absorption probability (and consequently the self-shielding factor) for an isotropic flux of neutrons impinging on different sized cylindrical samples of LiF TLD-100 and TLD-600. The calculations were performed for cylinders of radius up to 10 cm and heights of 0.1 to 1.5 cm. The Monte Carlo results were found to be significantly different from the analytic calculations for infinitely long cylinders, but, as expected, converged to the same value for (r/h) << 1. (U.K.)

  18. The problem of resonance self-shielding effect in neutron multigroup calculations

    International Nuclear Information System (INIS)

    Wang Qingming; Huang Jinghua

    1991-01-01

    It is not allowed to neglect the resonance self-shielding effect in hybrid blanket and fast reactor neutron designs. The authors discussed the importance as well as the method of considering the resonance self-shielding effect in hybrid blanket and fast reactor neutron multigroup calculations

  19. Removal, transportation and disposal of the Millstone 2 neutron thermal shield

    International Nuclear Information System (INIS)

    Snedeker, D.F.; Thomas, L.S.; Schmoker, D.S.; Cade, M.S.

    1985-01-01

    Some PWR reactors equipped with neutron thermal shields (NTS) have experienced severe neutron shield degradation to the extent that removal and disposal of these shields has become necessary. Due to the relative size and activation levels of the thermal shield, disposal techniques, remote material handling and transportation equipment must be carefully evaluated to minimize plant down time and maintain disposal costs at a minimum. This paper describes the techniques, equipment and methodology employed in the removal, transportation and disposal of the NTS at the Millstone 2 Nuclear Generating Station, a PWR facility owned and operated by Northeast Utilities of Hartford, CT. Specific areas addressed include: (1) remote underwater equipment and tooling for use in segmenting and loading the thermal shield in a disposal liner; (2) adaptation of the General Electric IF-300 Irradiated Fuel Cask for transportation of the NTS for disposal; (3) equipment and techniques used for cask handling and liner burial at the Low Level Radioactive Waste (LLRW) disposal facility

  20. Measurements and Monte-Carlo simulations of the particle self-shielding effect of B4C grains in neutron shielding concrete

    Science.gov (United States)

    DiJulio, D. D.; Cooper-Jensen, C. P.; Llamas-Jansa, I.; Kazi, S.; Bentley, P. M.

    2018-06-01

    A combined measurement and Monte-Carlo simulation study was carried out in order to characterize the particle self-shielding effect of B4C grains in neutron shielding concrete. Several batches of a specialized neutron shielding concrete, with varying B4C grain sizes, were exposed to a 2 Å neutron beam at the R2D2 test beamline at the Institute for Energy Technology located in Kjeller, Norway. The direct and scattered neutrons were detected with a neutron detector placed behind the concrete blocks and the results were compared to Geant4 simulations. The particle self-shielding effect was included in the Geant4 simulations by calculating effective neutron cross-sections during the Monte-Carlo simulation process. It is shown that this method well reproduces the measured results. Our results show that shielding calculations for low-energy neutrons using such materials would lead to an underestimate of the shielding required for a certain design scenario if the particle self-shielding effect is not included in the calculations.

  1. Evaluation of Neutron shielding efficiency of Metal hydrides

    Energy Technology Data Exchange (ETDEWEB)

    Jeon, Sang Hwan; Chae, San; Kim, Yong Soo [Hanyang University, Seoul (Korea, Republic of)

    2012-05-15

    Neutron shielding is achieved of interaction with material by moderation and absorption. Material that contains large amounts hydrogen atoms which are almost same neutron atomic weight is suited for fast neutron shielding material. Therefore, polymers containing high density hydrogen atom are being used for fast neutron shielding. On the other hand, composite materials containing high thermal neutron absorption cross section atom (Li, B, etc) are being used for thermal neutron shielding. However, these materials have low fast neutron absorption cross section. Therefore, these materials are not suited for fast neutron shielding. Hydrogen which has outstanding neutron energy reduction ability has very low thermal neutron absorption cross section, almost cannot be used for thermal neutron shielding. In this case, a large atomic number material (Pb, U, etc.) has been used. Thus, metal hydrides are considered as complement to concrete shielding material. Because metal hydrides contain high hydrogen density and elements with high atomic number. In this research neutron shielding performance and characteristic of nuclear about metal hydrides ((TiH{sub 2}, ZrH{sub 2}, HfH{sub 2}) is evaluated by experiment and MCNPX using {sup 252}Cf neutron source as purpose development shielding material to developed shielding material

  2. Enhancement of thermal neutron shielding of cement mortar by using borosilicate glass powder.

    Science.gov (United States)

    Jang, Bo-Kil; Lee, Jun-Cheol; Kim, Ji-Hyun; Chung, Chul-Woo

    2017-05-01

    Concrete has been used as a traditional biological shielding material. High hydrogen content in concrete also effectively attenuates high-energy fast neutrons. However, concrete does not have strong protection against thermal neutrons because of the lack of boron compound. In this research, boron was added in the form of borosilicate glass powder to increase the neutron shielding property of cement mortar. Borosilicate glass powder was chosen in order to have beneficial pozzolanic activity and to avoid deleterious expansion caused by an alkali-silica reaction. According to the experimental results, borosilicate glass powder with an average particle size of 13µm showed pozzolanic activity. The replacement of borosilicate glass powder with cement caused a slight increase in the 28-day compressive strength. However, the incorporation of borosilicate glass powder resulted in higher thermal neutron shielding capability. Thus, borosilicate glass powder can be used as a good mineral additive for various radiation shielding purposes. Copyright © 2017 Elsevier Ltd. All rights reserved.

  3. Neutron shielding material

    International Nuclear Information System (INIS)

    Nodaka, M.; Iida, T.; Taniuchi, H.; Yosimura, K.; Nagahama, H.

    1993-01-01

    From among the neutron shielding materials of the 'kobesh' series developed by Kobe Steel, Ltd. for transport and storage packagings, silicon rubber base type material has been tested for several items with a view to practical application and official authorization, and in order to determine its adaptability to actual vessels. Silicon rubber base type 'kobesh SR-T01' is a material in which, from among the silicone rubber based neutron shielding materials, the hydrogen content is highest and the boron content is most optimized. Its neutron shielding capability has been already described in the previous report (Taniuchi, 1986). The following tests were carried out to determine suitability for practical application; 1) Long-term thermal stability test 2) Pouring test on an actual-scale model 3) Fire test The experimental results showed that the silicone rubber based neutron shielding material has good neutron shielding capability and high long-term fire resistance, and that it can be applied to the advanced transport packaging. (author)

  4. Development of neutron shielding material for cask

    International Nuclear Information System (INIS)

    Najima, K.; Ohta, H.; Ishihara, N.; Matsuoka, T.; Kuri, S.; Ohsono, K.; Hode, S.

    2001-01-01

    Since 1980's Mitsubishi Heavy Industries, Ltd (MHI) has established transport and storage cask design 'MSF series' which makes higher payload and reliability for long term storage. MSF series transport and storage cask uses new-developed neutron shielding material. This neutron shielding material has been developed for improving durability under high condition for long term. Since epoxy resin contains a lot of hydrogen and is comparatively resistant to heat, many casks employ epoxy base neutron shielding material. However, if the epoxy base neutron shielding material is used under high temperature condition for a long time, the material deteriorates and the moisture contained in it is released. The loss of moisture is in the range of several percents under more than 150 C. For this reason, our purpose was to develop a high durability epoxy base neutron shielding material which has the same self-fire-extinction property, high hydrogen content and so on as conventional. According to the long-time heating test, the weight loss of this new neutron shielding material after 5000 hours heating has been lower than 0.04% at 150 C and 0.35% at 170 C. A thermal test was also performed: a specimen of neutron shielding material covered with stainless steel was inserted in a furnace under condition of 800 C temperature for 30 minutes then was left to cool down in ambient conditions. The external view of the test piece shows that only a thin layer was carbonized

  5. The determination of self-powered neutron detector sensitivity on thermal and epithermal neutron flux densities

    International Nuclear Information System (INIS)

    Erben, O.

    1980-01-01

    The coefficients of thermal and epithermal neutron flux density depression and self-shielding for the SPN detectors with vanadium, rhodium, silver and cobalt emitters are presented, (for cobalt SPN detectors the functions describing the absorbtion of neutrons along the emitter cross-section are also shown). Using these coefficients and previously published beta particle escape efficiencies, sensitivities are determined for the principal types of detectors produced by Les Cables de Lyon and SODERN companies. The experiments and their results verifying the validity of the theoretical work are described. (author)

  6. Self-shielding and burn-out effects in the irradiation of strongly-neutron-absorbing material

    International Nuclear Information System (INIS)

    Sekine, T.; Baba, H.

    1978-01-01

    Self-shielding and burn-out effects are discussed in the evaluation of radioisotopes formed by neutron irradiation of a strongly-neutron-absorbing material. A method of the evaluation of such effects is developed both for thermal and epithermal neutrons. Gadolinium oxide uniformly mixed with graphite powder was irradiated by reactor-neutrons together with pieces of a Co-Al alloy wire (the content of Co being 0.475%) as the neutron flux monitor. The configuration of the samples and flux monitors in each of two irradiations is illustrated. The yields of activities produced in the irradiated samples were determined by the γ-spectrometry with a Ge(Li) detector of a relative detection efficiency of 8%. Activities at the end of irradiation were estimated by corrections due to pile-up, self-absorption, detection efficiency, branching ratio, and decay of the activity. Results of the calculation are discussed in comparison with the observed yields of 153 Gd, 160 Tb, and 161 Tb for the case of neutron irradiation of disc-shaped targets of gadolinium oxide. (T.G.)

  7. New applications and developments in the neutron shielding

    Directory of Open Access Journals (Sweden)

    Uğur Fatma Aysun

    2017-01-01

    Full Text Available Shielding neutrons involve three steps that are slowing neutrons, absorption of neutrons, and impregnation of gamma rays. Neutrons slow down with thermal energy by hydrogen, water, paraffin, plastic. Hydrogenated materials are also very effective for the absorption of neutrons. Gamma rays are produced by neutron (radiation retention on the neutron shield, inelastic scattering, and degradation of activation products. If a source emits gamma rays at various energies, high-energy gamma rays sometimes specify shielding requirements. Multipurpose Materials for Neutron Shields; Concrete, especially with barium mixed in, can slow and absorb the neutrons, and shield the gamma rays. Plastic with boron is also a good multipurpose shielding material. In this study; new applications and developments in the area of neutron shielding will be discussed in terms of different materials.

  8. New applications and developments in the neutron shielding

    Science.gov (United States)

    Uğur, Fatma Aysun

    2017-09-01

    Shielding neutrons involve three steps that are slowing neutrons, absorption of neutrons, and impregnation of gamma rays. Neutrons slow down with thermal energy by hydrogen, water, paraffin, plastic. Hydrogenated materials are also very effective for the absorption of neutrons. Gamma rays are produced by neutron (radiation) retention on the neutron shield, inelastic scattering, and degradation of activation products. If a source emits gamma rays at various energies, high-energy gamma rays sometimes specify shielding requirements. Multipurpose Materials for Neutron Shields; Concrete, especially with barium mixed in, can slow and absorb the neutrons, and shield the gamma rays. Plastic with boron is also a good multipurpose shielding material. In this study; new applications and developments in the area of neutron shielding will be discussed in terms of different materials.

  9. Radiation shielding for neutron guides

    International Nuclear Information System (INIS)

    Ersez, T.; Braoudakis, G.; Osborn, J.C.

    2005-01-01

    Full text: Models of the neutron guide shielding for the out of bunker guides on the thermal and cold neutron beam lines of the OPAL Reactor (ANSTO) were constructed using the Monte Carlo code MCNP 4B. The neutrons that were not reflected inside the guides but were absorbed by the supermirror (SM) layers were noted to be a significant source of gammas. Gammas also arise from neutrons absorbed by the B, Si, Na and K contained in the glass. The proposed shielding design has produced compact shielding assemblies. These arrangements are consistent with safety requirements, floor load limits, and cost constraints. To verify the design a prototype was assembled consisting of 120mm thick Pb(96%)Sb(4%) walls resting on a concrete block. There was good agreement between experimental measurements and calculated dose rates for bulk shield regions. (authors)

  10. Radiation shielding for neutron guides

    International Nuclear Information System (INIS)

    Ersez, T.; Braoudakis, G.; Osborn, J.C.

    2006-01-01

    Models of the neutron guide shielding for the out of bunker guides on the thermal and cold neutron beam lines of the OPAL Reactor (ANSTO) were constructed using the Monte Carlo code MCNP 4B. The neutrons that were not reflected inside the guides but were absorbed by the supermirror (SM) layers were noted to be a significant source of gammas. Gammas also arise from neutrons absorbed by the B, Si, Na and K contained in the glass. The proposed shielding design has produced compact shielding assemblies. These arrangements are consistent with safety requirements, floor load limits, and cost constraints. To verify the design a prototype was assembled consisting of 120 mm thick Pb(96%)Sb(4%) walls resting on a concrete block. There was good agreement between experimental measurements and calculated dose rates for bulk shield regions

  11. Neutron radiation shielding properties of polymer incorporated self compacting concrete mixes.

    Science.gov (United States)

    Malkapur, Santhosh M; Divakar, L; Narasimhan, Mattur C; Karkera, Narayana B; Goverdhan, P; Sathian, V; Prasad, N K

    2017-07-01

    In this work, the neutron radiation shielding characteristics of a class of novel polymer-incorporated self-compacting concrete (PISCC) mixes are evaluated. Pulverized high density polyethylene (HDPE) material was used, at three different reference volumes, as a partial replacement to river sand in conventional concrete mixes. By such partial replacement of sand with polymer, additional hydrogen contents are incorporated in these concrete mixes and their effect on the neutron radiation shielding properties are studied. It has been observed from the initial set of experiments that there is a definite trend of reductions in the neutron flux and dose transmission factor values in these PISCC mixes vis-à-vis ordinary concrete mix. Also, the fact that quite similar enhanced shielding results are recorded even when reprocessed HDPE material is used in lieu of the virgin HDPE attracts further attention. Copyright © 2017 Elsevier Ltd. All rights reserved.

  12. Double-layer neutron shield design as neutron shielding application

    Science.gov (United States)

    Sariyer, Demet; Küçer, Rahmi

    2018-02-01

    The shield design in particle accelerators and other high energy facilities are mainly connected to the high-energy neutrons. The deep penetration of neutrons through massive shield has become a very serious problem. For shielding to be efficient, most of these neutrons should be confined to the shielding volume. If the interior space will become limited, the sufficient thickness of multilayer shield must be used. Concrete and iron are widely used as a multilayer shield material. Two layers shield material was selected to guarantee radiation safety outside of the shield against neutrons generated in the interaction of the different proton energies. One of them was one meter of concrete, the other was iron-contained material (FeB, Fe2B and stainless-steel) to be determined shield thicknesses. FLUKA Monte Carlo code was used for shield design geometry and required neutron dose distributions. The resulting two layered shields are shown better performance than single used concrete, thus the shield design could leave more space in the interior shielded areas.

  13. Neutron shielding properties of boron-containing ore and epoxy composites

    International Nuclear Information System (INIS)

    Li Zhifu; Xue Xiangxin

    2011-01-01

    Using the boron-containing iron ore concentrate and boron-rich slag as studying object, the starting materials were got after the specific green ore containing boron dressing in China and blast furnace separation respectively. Monte-Carlo method was used to study the effect of the boron-containing iron ore concentrate and boron-rich slag and their composites with epoxy on the neutron shielding abilities. The reasons that affecting the shielding materials properties was discussed and the suitable proportioning of boron-containing ore to epoxy composites was confirmed; the 14.1 MeV fast neutron removal cross section and the total thermal neutron attenuation coefficient were obtained and compared with that of the common used concrete. The results show that the shielding property of 14.1 MeV fast neutron is mainly concerned with the low-Z elements in the shielding materials, the thermal neutron shielding ability is mainly concerned with boron concentrate in the composite, the attenuation of the accompany γ-ray photon is mainly concerned with the high atom number elements content in the ore and the density of the shielding material. The optimum Janume fractions of composites are in the range of 0.4-0.6 and the fast neutron shielding properties are similar to concrete while the thermal neutron shielding properties are higher than concrete. The composites are expected to be used as biological concrete shields crack injection and filling of the anomalous holes through the concrete shields around the radiation fields or directly to be prepared as shielding materials.(authors)

  14. Self shielding in cylindrical fissile sources in the APNea system

    International Nuclear Information System (INIS)

    Hensley, D.

    1997-01-01

    In order for a source of fissile material to be useful as a calibration instrument, it is necessary to know not only how much fissile material is in the source but also what the effective fissile content is. Because uranium and plutonium absorb thermal neutrons so Efficiently, material in the center of a sample is shielded from the external thermal flux by the surface layers of the material. Differential dieaway measurements in the APNea System of five different sets of cylindrical fissile sources show the various self shielding effects that are routinely encountered. A method for calculating the self shielding effect is presented and its predictions are compared with the experimental results

  15. Development of paraffin and paraffin/bitumen composites with additions of B2O3 for thermal neutron shielding applications

    International Nuclear Information System (INIS)

    Toyen, Donruedee; Saenboonruang, Kiadtisak

    2017-01-01

    In this work, paraffin and paraffin/bitumen composites with additions of boron oxide (B 2 O 3 ) were prepared to evaluate the viscosity, flexural, and thermal neutron shielding properties for uses as thermal neutron shielding materials. The results showed that the addition of 3 wt% or 9 wt% bitumen to paraffin increased the overall flexural properties with the content of 9 wt% bitumen having the highest values. The improvement in flexural properties made the composites less brittle, stiffer, and longer-lasting. Furthermore, different contents of B 2 O 3 (0, 7, 14, 21, 28, and 35 wt%) were added to paraffin and paraffin/bitumen composites to investigate the effects of the B 2 O 3 contents. The results indicated that an increase in B 2 O 3 contents improved the shielding properties but slightly reduced the flexural properties. Specifically for 5-mm paraffin and 5-mm paraffin/bitumen samples with 35 wt% of B 2 O 3 , both samples could reduce neutron flux by more than 70%. The overall results suggested that the paraffin and paraffin/bitumen composites with additions of B 2 O 3 showed improved properties for utilization as effective thermal neutron shielding materials. (author)

  16. Research on shielding neutron efficiency of some boron-bearing fabric and transparent resin materials

    International Nuclear Information System (INIS)

    Chen Changmao; Liu Jinhua; Su Jingling; Wang Zheng

    1995-01-01

    The shielding neutron efficiency of boron-bearing materials developed recently is introduced. The thermal neutron shield ratios for two kinds of non-woven cloth with thickness of 58 mg/cm 2 and 153 mg/cm 2 are 51% and 79% respectively. Their mass attenuation coefficient for 0.186, 24.4 and 144 keV neutron are 1.56, 1.29 and 0.9 cm 2 /g respectively. The thermal neutron shield ratio is 85% for the natural boron-bearing transparent resin plate with the thickness of 0.59 g/cm 2 , and 97% for enriched boron or gadolinium bearing resin plate. The shield ratios of all three materials for 24.4 keV neutrons are 38%. The transparence of natural light for enriched boron-bearing resin plates shows no considerable change after they were exposed to thermal neutrons up to 6 Sv. After they were exposed up to 20 Sv, the transparence decreases to 50% but thermal neutron shield ratio does not change. The gadolinium-bearing plate has a very strong thermal neutron-capture gamma radiation and its dose-equivalent is greater than that of incident thermal neutrons

  17. Self-shielding models of MICROX-2 code: Review and updates

    International Nuclear Information System (INIS)

    Hou, J.; Choi, H.; Ivanov, K.N.

    2014-01-01

    Highlights: • The MICROX-2 code has been improved to expand its application to advanced reactors. • New fine-group cross section libraries based on ENDF/B-VII have been generated. • Resonance self-shielding and spatial self-shielding models have been improved. • The improvements were assessed by a series of benchmark calculations against MCNPX. - Abstract: The MICROX-2 is a transport theory code that solves for the neutron slowing-down and thermalization equations of a two-region lattice cell. The MICROX-2 code has been updated to expand its application to advanced reactor concepts and fuel cycle simulations, including generation of new fine-group cross section libraries based on ENDF/B-VII. In continuation of previous work, the MICROX-2 methods are reviewed and updated in this study, focusing on its resonance self-shielding and spatial self-shielding models for neutron spectrum calculations. The improvement of self-shielding method was assessed by a series of benchmark calculations against the Monte Carlo code, using homogeneous and heterogeneous pin cell models. The results have shown that the implementation of the updated self-shielding models is correct and the accuracy of physics calculation is improved. Compared to the existing models, the updates reduced the prediction error of the infinite multiplication factor by ∼0.1% and ∼0.2% for the homogeneous and heterogeneous pin cell models, respectively, considered in this study

  18. Neutron shieldings

    International Nuclear Information System (INIS)

    Tarutani, Kohei

    1979-01-01

    Purpose: To decrease the stresses resulted by the core bendings to the base of an entrance nozzle. Constitution: Three types of round shielding rods of different diameter are arranged in a hexagonal tube. The hexagonal tube is provided with several spacer pads receiving the loads from the core constrain mechanism at its outer circumference, a handling head for a fuel exchanger at its top and an entrance nozzle for self-holding the neutron shieldings and flowing heat-removing coolants at its bottom. The diameters for R 1 , R 2 and R 3 for the round shielding rods are designed as: 0.1 R 1 2 1 and 0.2 R 1 2 1 . Since a plurality of shielding rods of small diameter are provided, soft structure are obtained and a plurality of coolant paths are formed. (Furukawa, Y.)

  19. Development of silicone rubber-type neutron shielding material

    International Nuclear Information System (INIS)

    Do, Jae Bum; Cho, Soo Hang; Kim, Ik Soo; Oh, Seung Chul; Hong, Soon Seok; Noh, Sung Ki; Jeong, Duk Yeon.

    1997-06-01

    Because the exposure to radiation in the nuclear facilities can be fatal to human, it is important to reduce the radiation dose level to a tolerable level. The purpose of this study is to develop highly effective neutron shielding materials for the shipping and storage cask of radioactive materials or in the nuclear/radiation facilities. On this study, we developed silicone rubber based neutron shielding materials and their various material properties, including neutron shielding ability, fire resistance, combustion characteristics, radiation resistance, thermal and mechanical properties were evaluated experimentally. (author). 16 tabs., 17 figs., 25 refs

  20. 6Li-doped silicate glass for thermal neutron shielding

    International Nuclear Information System (INIS)

    Stone, C.A.; Blackburn, D.H.; Kauffman, D.A.; Cranmer, D.C.; Olmez, I.

    1994-01-01

    Glass formulations are described that contain high concentrations of 6 Li and are suitable for use as thermal neutron shielding. One formulation contained 31 mol% of 6 Li 2 O and 69 mol% of SiO 2 . Studies were performed on a second formulation that contained as much as 37 mol% of 6 Li 2 O and 59 mol% of SiO 2 , with 4 mol% Al 2 O 3 added to prevent crystallization at such high 6 Li 2 O concentrations. These lithium silicate glasses can be formed into a variety of shapes using conventional glass fabrication techniques. Examples include flat plates, disks, hollow cylinders, and other more complex geometries. Both in-beam and in-core experiments have been performed to study the use and durability of Li silicate glasses. In-core experiments show the glass can withstand the intense radiation fields near the core of a reactor. The neutron attenuation of the glasses used in these studies was 90%/mm. In-beam studies show that the glass is effective for reducing the gamma-ray and neutron fields near experiments. ((orig.))

  1. Experimental assessment on the thermal effects of the neutron shielding and heat-transfer fin of dual purpose casks on open pool fire

    International Nuclear Information System (INIS)

    Bang, Kyoung-Sik; Yu, Seung-Hwan; Lee, Ju-Chan; Seo, Ki-Seog; Choi, Woo-Seok

    2016-01-01

    Highlights: • An open pool fire test was performed to estimate not only the combustion effect of the neutron shielding but also the effect of the heat transfer fin of the dual purpose cask. • The heat transfer to the inside of the dual purpose cask was reduced, when the neutron shielding burns. • The surface temperatures are lower in the present of the heat transfer fins. • If inflammable material is used as the components of the cask, evaluating thermal integrity using the thermal test would be desirable. - Abstract: Dual purpose casks are used for storage and transport of spent nuclear fuel assemblies. They must therefore satisfy the requirements prescribed in the Korea Nuclear Safety Security Commission Act 2014-50, the IAEA Safety Standard Series No. SSR-6, and US 10 CFR Part 71. These regulatory guidelines classify the dual purpose cask as a Type B package and state that a Type B package must be able to withstand a temperature of 800 °C for a period of 30 min. NS-4-FR is used as neutron shielding of the dual purpose cask. Heat transfer fins are embedded to enhance heat transfer from the cask body to the outer-shell because the thermal conductivity of NS-4-FR is not good. However, accurately simulating not only the combustion effect of the neutron shielding but also the effect of the heat transfer fin in the thermal analysis is not easy. Therefore, an open pool fire test was conducted using a one-sixth slice of a real cask to estimate these effects at a temperature of 800 °C for a period of 30 min. The temperature at the central portion of the neutron shielding was lower when the neutron shielding in contact with the outer cask burned because the neutron shielding absorbed the surrounding latent heat as the neutron shielding burned. Therefore, the heat transfer to the inside of the dual purpose cask was reduced. The surface temperature was lower when a heat transfer fin was installed because the high heat generated by the flame was transferred to the

  2. Experimental assessment on the thermal effects of the neutron shielding and heat-transfer fin of dual purpose casks on open pool fire

    Energy Technology Data Exchange (ETDEWEB)

    Bang, Kyoung-Sik, E-mail: nksbang@kaeri.re.kr; Yu, Seung-Hwan; Lee, Ju-Chan; Seo, Ki-Seog; Choi, Woo-Seok

    2016-08-01

    Highlights: • An open pool fire test was performed to estimate not only the combustion effect of the neutron shielding but also the effect of the heat transfer fin of the dual purpose cask. • The heat transfer to the inside of the dual purpose cask was reduced, when the neutron shielding burns. • The surface temperatures are lower in the present of the heat transfer fins. • If inflammable material is used as the components of the cask, evaluating thermal integrity using the thermal test would be desirable. - Abstract: Dual purpose casks are used for storage and transport of spent nuclear fuel assemblies. They must therefore satisfy the requirements prescribed in the Korea Nuclear Safety Security Commission Act 2014-50, the IAEA Safety Standard Series No. SSR-6, and US 10 CFR Part 71. These regulatory guidelines classify the dual purpose cask as a Type B package and state that a Type B package must be able to withstand a temperature of 800 °C for a period of 30 min. NS-4-FR is used as neutron shielding of the dual purpose cask. Heat transfer fins are embedded to enhance heat transfer from the cask body to the outer-shell because the thermal conductivity of NS-4-FR is not good. However, accurately simulating not only the combustion effect of the neutron shielding but also the effect of the heat transfer fin in the thermal analysis is not easy. Therefore, an open pool fire test was conducted using a one-sixth slice of a real cask to estimate these effects at a temperature of 800 °C for a period of 30 min. The temperature at the central portion of the neutron shielding was lower when the neutron shielding in contact with the outer cask burned because the neutron shielding absorbed the surrounding latent heat as the neutron shielding burned. Therefore, the heat transfer to the inside of the dual purpose cask was reduced. The surface temperature was lower when a heat transfer fin was installed because the high heat generated by the flame was transferred to the

  3. Neutron guide shielding for the BIFROST spectrometer at ESS

    OpenAIRE

    Mantulnikovs, K.; Bertelsen, M.; Cooper-Jensen, Carsten P.; Lefmann, K.; Klinkby, E. B.

    2016-01-01

    We report on the study of fast-neutron background for the BIFROST spectrometerat ESS. We investigate the effect of background radiation induced by the interaction of fast neutrons from the source with the material of the neutron guide and devise a reasonable fast, thermal/cold neutron shielding solution for the current guide geometry using McStas and MCNPX. We investigate the effectiveness of the steel shielding around the guide by running simulations with three different steel thicknesses. T...

  4. Studsvik thermal neutron facility

    International Nuclear Information System (INIS)

    Pettersson, O.A.; Larsson, B.; Grusell, E.; Svensson, P.

    1992-01-01

    The Studsvik thermal neutron facility at the R2-0 reactor originally designed for neutron capture radiography has been modified to permit irradiation of living cells and animals. A hole was drilled in the concrete shielding to provide a cylindrical channel with diameter of 25.3 cm. A shielding water tank serves as an entry holder for cells and animals. The advantage of this modification is that cells and animals can be irradiated at a constant thermal neutron fluence rate of approximately 10 9 n cm -2 s -1 (at 100 kW) without stopping and restarting the reactor. Topographic analysis of boron done by neutron capture autoradiography (NCR) can be irradiated under the same conditions as previously

  5. Development of epoxy resin-type neutron shielding materials (I)

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Soo Haeng; Kim, Ik Soo; Shin, Young Joon; Do, Jae Bum; Ro, Seung Gy

    1997-12-01

    Because the exposure to radiation in the nuclear facilities can be fatal to human, it is important to reduce the radiation dose level to a tolerable level. The purpose of this study is to develop highly effective neutron shielding materials for the shipping and storage cask of radioactive materials or in the nuclear /radiation facilities. On this study, we developed epoxy resin based neutron shielding materials and their various materials properties, including neutron shielding ability, fire resistance, combustion characteristics, radiation resistance, thermal and mechanical properties were evaluated experimentally. (author). 31 refs., 22 tabs., 17 figs.

  6. Thermal shield support degradation in pressurized water reactors

    International Nuclear Information System (INIS)

    Sweeney, F.J.; Fry, D.N.

    1986-01-01

    Damage to the thermal shield support structures of three pressurized water reactors (PWRs) due to flow-induced vibrations was recently discovered during refueling. In two of the reactors, severe damage occurred to the thermal shield, and in one reactor the core support barrel (CSB) was damaged, necessitating extended outages for repairs. In all three reactors, several of the thermal shield supports were either loose, damaged, or missing. The three plants had been in operation for approximately 10 years before the damage was apparent by visual inspection. Because each of the three US PWR manufacturers have experienced thermal shield support degradation, the Nuclear Regulatory Commission requested that Oak Ridge National Laboratory analyze ex-core neutron detector noise data to determine the feasibility of detecting incipient thermal shield support degradation. Results of the noise data analysis indicate that thermal shield support degradation probably began early in the life of both severely damaged plants. The degradation was characterized by shifts in the resonant frequencies of core internal structures and the appearance of new resonances in the ex-core neutron detector noise. Both the data analyses and the finite element calculations indicate that these changes in resonant frequencies are less than 3 Hz. 11 refs., 16 figs

  7. LOFT shield tank steady state temperatures with addition of gamma and neutron shielding

    International Nuclear Information System (INIS)

    Kyllingstad, G.

    1977-01-01

    The effect of introducing a neutron and gamma shield into the annulus between the reactor vessel and the shield tank is analyzed. This addition has been proposed in order to intercept neutron streaming up the annulus during nuclear operations. Its installation will require removal of approximately 20- 1 / 2 inches of stainless steel foil insulation at the top of the annulus. The resulting conduction path is believed to result in increased water temperatures within the shield tank, possibly beyond the 150 0 F limit, and/or cooling of the reactor vessel nozzles such that adverse thermal stresses would be generated. A two dimensional thermal analysis using the finite element code COUPLE/MOD2 was done for the shield tank system illustrated in the figure (1). The reactor was assumed to be at full power, 55 MW (th), with a loop flow rate of 2.15 x 10 6 lbm/hr (268.4 kg/s) at 2250 psi (15.51 MPa). Calculations indicate a steady state shield tank water temperature of 140 0 F (60 0 C). This is below the 150 0 F (65.56 0 C) limit. Also, no significant changes in thermal gradients within the nozzle or reactor vessel wall are generated. A spacer between the gamma shield and the shield tank is recommended, however, in order to ensure free air circulation through the annulus

  8. Slow neutrons and secondary gamma ray distributions in concrete shields followed by reflecting layers

    International Nuclear Information System (INIS)

    Makarious, A.S.; Swilem, Y.I.; Awwad, Z.; Bayomy, T.

    1993-01-01

    Slow neutrons and secondary gamma ray distributions in concrete shields with and without a reflecting layer behind layer behind the concrete shield have been investigated first in case of using a bare reactor beam and then on using a B-4 C filtered beam. The total and capture secondary gamma ray coefficient (B gamma and B gamma C ), the ratio of the reflected thermal neutron (gamma) the ratio of the secondary gamma rays caused by reflected neutrons to those caused transmitted neutrons (Th I gamma/F I gamma) and the effect of inserting a blocking layer (a B-4 C layer) between the concrete shield and the reflector on the suppression of the produced secondary gamma rays have been investigated. It was found that the presence of the reflector layer behind the concrete shield reflects some thermal neutrons back to the concrete shields and so it increases the number of thermal neutrons at the interface between the concrete shield and the reflector. Also the capture secondary gamma rays was increased at the interface between the two medii due to the capture of the reflected thermal neutrons in the concrete shields. It was shown that B-gamma is higher than and that B g amma B gamma C and I gamma T h/ I gamma i f for the different concrete types is higher in case of using the graphite reflector than that in using either water or paraffin reflectors. Putting a blocking layer (B 4 C layer) between the concrete shield and the reflector decreases the produced secondary gamma rays due to the absorption of the reflected thermal neutrons. 17 figs

  9. Application of a calculational model for thermal neutrons through biological shields

    Energy Technology Data Exchange (ETDEWEB)

    Hathout, A M [Nuclear engineering safety department, national center for nuclear safety and radiation, Nasr City Cairo, (Egypt)

    1995-10-01

    In this work a computational program, based on the Boltzmann transport integrodifferential equation, is applied. The scattering kernel is represented by the synthetic scattering model. The behaviour of thermal neutron in hydrogenous materials, which can be used as biological shields, are studied. These materials are water, polyethylene, Oak-Ridge concrete, ordinary concrete and manganese concrete. The data obtained are presented in tables. The results are analysed and compared with similar experimental values. Safety evaluation and environmental impact are discussed. 2 tabs.

  10. Resonance self-shielding effect in uncertainty quantification of fission reactor neutronics parameters

    International Nuclear Information System (INIS)

    Chiba, Go; Tsuji, Masashi; Narabayashi, Tadashi

    2014-01-01

    In order to properly quantify fission reactor neutronics parameter uncertainties, we have to use covariance data and sensitivity profiles consistently. In the present paper, we establish two consistent methodologies for uncertainty quantification: a self-shielded cross section-based consistent methodology and an infinitely-diluted cross section-based consistent methodology. With these methodologies and the covariance data of uranium-238 nuclear data given in JENDL-3.3, we quantify uncertainties of infinite neutron multiplication factors of light water reactor and fast reactor fuel cells. While an inconsistent methodology gives results which depend on the energy group structure of neutron flux and neutron-nuclide reaction cross section representation, both the consistent methodologies give fair results with no such dependences.

  11. Characteristic Determination Of Self Shielding Factor And Cadmium Ratio Of Cylindrical Probe

    International Nuclear Information System (INIS)

    Hamzah, Amir; Budi R, Ita; Pinem, Suriam

    1996-01-01

    Determination of thermal, epithermal and total self shielding factor and cadmium ratio of cylindrical probe has been done by measurement and calculation. Self shielding factor can be determined by dividing probe activity to Al-alloy probe activity. Due to the lack of cylindrical probe made of Al-alloy, self shielding factor can be determined by parabolic extrapolation of measured activities to 0 cm radius to divide those activities. Theoretically, self shielding factor can be determined by making numerical solution of two dimensional integral equations using Romberg method. To simplify, the calculation is based on single collision theory with the assumption of monoenergetic neutron and isotropic distribution. For gold cylindrical probe, the calculation results are quite close to the measurement one with the relative discrepancy for activities, cadmium ratio and self shielding factor of bare probe are less then 11.5%, 3,5% and 1.5% respectively. The program can be used for the calculation of other kinds of cylindrical probes. Due to dependency to radius, cylindrical probe made of copper has the best characteristic of self shielding factor and cadmium ratio

  12. Neutron shielding material based on colemanite and epoxy resin

    International Nuclear Information System (INIS)

    Okuno, K.

    2005-01-01

    In recent years, there has been a need for compact shielding design such as self-shielding of a PET cyclotron or up-gradation of radiation machinery in existing facilities. In these cases, high performance shielding materials are needed. Concrete or polyethylene have been used for a neutron shield. However, for compact shielding, they fall short in terms of performance or durability. Therefore, a new type of neutron shielding material based on epoxy resin and colemanite has been developed. Slab attenuation experiments up to 40 cm for the new shielding material were carried out using a 252 Cf neutron source. Measurement was carried out using a REM-counter, and compared with calculation. The results show that the shielding performance is better than concrete and polyethylene mixed with 10 wt% boron oxide. From the result, we confirmed that the performance of the new material is suitable for practical use. (authors)

  13. RESONANCE SELF-SHIELDING EFFECT IN UNCERTAINTY QUANTIFICATION OF FISSION REACTOR NEUTRONICS PARAMETERS

    Directory of Open Access Journals (Sweden)

    GO CHIBA

    2014-06-01

    Full Text Available In order to properly quantify fission reactor neutronics parameter uncertainties, we have to use covariance data and sensitivity profiles consistently. In the present paper, we establish two consistent methodologies for uncertainty quantification: a self-shielded cross section-based consistent methodology and an infinitely-diluted cross section-based consistent methodology. With these methodologies and the covariance data of uranium-238 nuclear data given in JENDL-3.3, we quantify uncertainties of infinite neutron multiplication factors of light water reactor and fast reactor fuel cells. While an inconsistent methodology gives results which depend on the energy group structure of neutron flux and neutron-nuclide reaction cross section representation, both the consistent methodologies give fair results with no such dependences.

  14. Neutron sensitivity of prompt-response self-powered neutron detectors and the interval rule

    International Nuclear Information System (INIS)

    Molina Avila, J.; Carmolopes, M.

    1989-01-01

    This paper is devoted to the calculation of thermal s th and epithermal s epi sensitivities of cobalt prompt-response Self-Powered Neutron Detectors (SPNDs). The thermal sensitivity was obtained for a Maxwellian neutron field, and the effect of scattering on the self-shielding correction was taken into consideration in the second-collision approximation. The dependence of s th on the emitter radius R was studied in a wide region of R (0.025 to 0.2 cm). The differential and global epithermal sensitivities were calculated using a simple expression for the first-collision neutron absorption probability. Finally, a criterion to evaluate the accuracy of the parameters of the model was established in the form of some Interval Rule which is very sensitive to the radial dependence of the flux perturbation correction and other parameters of the model in both the thermal and epithermal regions

  15. Neutron shielding characteristics of nano-B2O3 dispersed Poly Vinyl Alcohol

    International Nuclear Information System (INIS)

    Kim, Jae Woo; Uhm, Young Rang; Lee, Min Ku; Lee, Hee Min; Rhee, Chang Kyu

    2008-01-01

    Neutron is sometimes beneficiary to human beings while they are unwanted for most cases same as the other radiations such as gamma, beta, and alpha, etc. do. Shielding for neutrons therefore is extremely important to keep the radiation environment safe. Especially, it is critical to absorb (or shield) neutrons generated from the spent fuel in a container/storage, nuclear reactor, and cyclotron, etc. In this regard, light materials containing neutron absorbers such as borated-polymers are very useful to shield neutrons in those radiation environments. This investigation is focused on the development of borated polymer-based materials whose neutron shielding efficiency is greatly enhanced by using nano sized boron compounds. Boron is well known as a thermal neutron absorber due to its large thermal neutron absorption cross-section (σ th = 760 b, b = 10 -2 - 4 cm 2 ). Although absorption of neutrons in the medium is mainly dependent on the boron atomic weight concentration, we firstly observed the size of boron particles also has an important role in neutron shielding. Mean free path of neutrons colliding with the smaller particles dispersed in the medium might be decreased when it is compared to the larger particles at the same atomic weight concentration. This means that the neutron shielding efficiency of a polymer mixed with the smaller boron compounds is higher than that of a polymer mixed with the larger boron compounds at the same atomic weight boron concentration

  16. Thermal neutron albedo measurements for multilithic reflectors

    International Nuclear Information System (INIS)

    Mehboob, Khurram; Ahmed, Raheel; Ali, Majid; Tabassam, Uzma

    2013-01-01

    Highlights: • Measurement of thermal neuron albedo for multilithic reflectors. • Modeling of experiments in MATLAB. • Comparison of numerical calculated and experimental values. • Study of thermal neutron albedo in different multilayered shielding. - Abstract: An experimental measurement of the thermal neutron (0.025 eV) albedo (αth) has been carried out for multilithic shielding by using Am–Be neutron source and BF 3 detector. The measured saturation value for the thermal albedo of paraffin wax has been found to be 0.734 ± 0.020, which is in close agreement to the corresponding value 0.83 quoted in the literature. The thermal neutron albedo has been measured for the multilayered shielding in copper–wood, copper–aluminum, wood–paraffin and paraffin–iron combinations in horizontal geometric configurations. Modeling and numerical simulation have been carried out by developing a MATLAB code which solves the diffusion equation in order to calculate the experimental results. Good agreement has been found between the numerical calculated and experimental results. The uncertainties in the measurements have also been calculated based on error propagation of the underlying Poisson distribution

  17. Measured neutron beam line shielding effectiveness of several iron/polyethylene configurations

    International Nuclear Information System (INIS)

    Legate, G.L.; Howe, M.L.; Mundis, R.L.

    1988-01-01

    Neutron and gamma-ray leakage measurements were taken at various stages of shield construction of neutron flight path 5 (the Lash-up flight path) at LANSCE, to compare the relative effectiveness of several configurations. Dose equivalent rates were determined for three categories: ''low-energy neutrons'', below 20 MeV; ''high- energy neutrons'', above 20 MeV; and gamma rays, as measured by hand-held survey instruments. The low energy neutrons were measured by activation of an indium foil in a paraffin-filled cadmium canister, sized to be generally insensitive above 20 MeV. High-energy neutrons were measured by (n,2n) production of Carbon 11 in a plastic scintillator with a 20-MeV threshold. Thermal neutrons were not measured at the shield-leakage test points. Room-scattered neutrons were observed by Albatross IV detector readings, which were taken beside the shield as a measure of variation of room background as the shield configuration changed. 1 fig., 1 tab

  18. Neutron guide shielding for the BIFROST spectrometer at ESS

    DEFF Research Database (Denmark)

    Mantulnikovs, K.; Bertelsen, M.; Cooper-Jensen, C.P.

    2016-01-01

    We report on the study of fast-neutron background for the BIFROST spectrometer at ESS. We investigate the effect of background radiation induced by the interaction of fast neutrons from the source with the material of the neutron guide and devise a reasonable fast, thermal/cold neutron shielding...... solution for the current guide geometry using McStas and MCNPX. We investigate the effectiveness of the steel shielding around the guide by running simulations with three different steel thicknesses. The same approach is used to study the efficiencies of the steel wall a flat cylinder pierced by the guide...... in the middle and the polyethylene layer. The final model presented here has a 3 cm thick steel shielding around the guide, 30 cm of polyethylene around the shielding, two 5 mm thick B4C layers and a steel wall at position Z = 38 m, being 1 m thick and 10 m in radius. The final model finally proves...

  19. Resonance Self-Shielding Methodologies in SCALE 6

    International Nuclear Information System (INIS)

    Williams, Mark L.

    2011-01-01

    SCALE 6 includes several problem-independent multigroup (MG) libraries that were processed from the evaluated nuclear data file ENDF/B using a generic flux spectrum. The library data must be self-shielded and corrected for problem-specific spectral effects for use in MG neutron transport calculations. SCALE 6 computes problem-dependent MG cross sections through a combination of the conventional Bondarenko shielding-factor method and a deterministic continuous-energy (CE) calculation of the fine-structure spectra in the resolved resonance and thermal energy ranges. The CE calculation can be performed using an infinite medium approximation, a simplified two-region method for lattices, or a one-dimensional discrete ordinates transport calculation with pointwise (PW) cross-section data. This paper describes the SCALE-resonance self-shielding methodologies, including the deterministic calculation of the CE flux spectra using PW nuclear data and the method for using CE spectra to produce problem-specific MG cross sections for various configurations (including doubly heterogeneous lattices). It also presents results of verification and validation studies.

  20. Neutron shielding material

    International Nuclear Information System (INIS)

    Suzuki, Shigenori; Iimori, Hiroshi; Kobori, Junzo.

    1980-01-01

    Purpose: To provide a neutron shielding material which incorporates preferable shielding capacity, heat resistance, fire resistance and workability by employing a mixture of thermosetting resin, polyethylene and aluminium hydroxide in special range ratio and curing it. Constitution: A mixture containing 20 to 60% by weight of thermosetting resin having preferable heat resistance, 10 to 40% by weight of polyethylene powder having high hydrogen atom density and 1000 to 60000 of molecular weight, and 15 to 55% by weight of Al(OH) 3 for imparting fire resistance and self-fire extinguishing property thereto is cured. At this time approx. 0.5 to 5% of curing catalyst of the thermosetting resin is contained in 100 parts by weight of the mixture. (Sekiya, K.)

  1. Measurement of thermal neutron cross-section and resonance integral for the 165Ho(n,γ) 166gHo reaction using electron linac-based neutron source

    Science.gov (United States)

    Nguyen, Van Do; Pham, Duc Khue; Kim, Tien Thanh; Kim, Guinyun; Lee, Manwoo; Kim, Kyung Sook; Kang, Heung-Sik; Cho, Moo-Hyun; Ko, In Soo; Namkung, Won

    2011-01-01

    The thermal neutron cross-section and the resonance integral of the 165Ho(n,γ) 166gHo reaction have been measured by the activation method using a 197Au(n,γ) 198Au monitor reaction as a single comparator. The high-purity natural Ho and Au foils with and without a cadmium shield case of 0.5 mm thickness were irradiated in a neutron field of the Pohang neutron facility. The induced activities in the activated foils were measured with a calibrated p-type high-purity Ge detector. The correction factors for the γ-ray attenuation ( Fg), the thermal neutron self-shielding ( Gth), the resonance neutron self-shielding ( Gepi) effects, and the epithermal neutron spectrum shape factor ( α) were taken into account. The thermal neutron cross-section for the 165Ho(n,γ) 166gHo reaction has been determined to be 59.7 ± 2.5 barn, relative to the reference value of 98.65 ± 0.09 barn for the 197Au(n,γ) 198Au reaction. By assuming the cadmium cut-off energy of 0.55 eV, the resonance integral for the 165Ho(n,γ) 166gHo reaction is 671 ± 47 barn, which is determined relative to the reference value of 1550 ± 28 barn for the 197Au(n,γ) 198Au reaction. The present results are, in general, good agreement with most of the previously reported data within uncertainty limits.

  2. Shielded regenerative neutron detector

    International Nuclear Information System (INIS)

    Terhune, J.H.; Neissel, J.P.

    1978-01-01

    An ion chamber type neutron detector is disclosed which has a greatly extended lifespan. The detector includes a fission chamber containing a mixture of active and breeding material and a neutron shielding material. The breeding and shielding materials are selected to have similar or substantially matching neutron capture cross-sections so that their individual effects on increased detector life are mutually enhanced

  3. Neutron shielding for a 252 Cf source

    International Nuclear Information System (INIS)

    Vega C, H.R.; Manzanares A, E.; Hernandez D, V.M.; Eduardo Gallego, Alfredo Lorente

    2006-01-01

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

  4. Neutron shielding performance of water-extended polyester

    International Nuclear Information System (INIS)

    Vega Carrillo, H.R.; Manzanares-Acuna, E.; Hernandez-Davila, V.M.; Vega Carrillo, H.R.; Hernandez-Davila, V.M.; Gallego, E.; Lorente, A.

    2006-01-01

    A Monte Carlo study to determine the shielding features to neutrons of water-extended polyester (WEP) was carried out. Materials with low atomic number are predominantly used for neutron shielding because these materials effectively attenuate neutrons, mainly through elastic and inelastic collisions. In addition to neutron attenuation properties, other desirable properties for neutron shielding materials include mechanical strength, stability, low cost, and ease of handling. During the selection of materials to design a neutron shield, prompt gamma production as well as radionuclide induced by neutron activation must be considered. In this investigation the Monte Carlo method (MCNP code) was used to evaluate the performance of a water-extended polyester shield designed for the transportation, storage, and use of a 252 Cf isotopic neutron source, for comparison the calculations were extended to water shielding, the bare source in vacuum and in air. (authors)

  5. Neutron shielding performance of water-extended polyester

    Energy Technology Data Exchange (ETDEWEB)

    Vega Carrillo, H.R.; Manzanares-Acuna, E.; Hernandez-Davila, V.M. [Zacatecas Univ. Autonoma, Nuclear Studies (Mexico); Vega Carrillo, H.R.; Hernandez-Davila, V.M. [Zacatecas Univ. Autonoma, Electric Engineering Academic Units (Mexico); Gallego, E.; Lorente, A. [Madrid Univ. Politecnica, cNuclear Engineering Department (Mexico)

    2006-07-01

    A Monte Carlo study to determine the shielding features to neutrons of water-extended polyester (WEP) was carried out. Materials with low atomic number are predominantly used for neutron shielding because these materials effectively attenuate neutrons, mainly through elastic and inelastic collisions. In addition to neutron attenuation properties, other desirable properties for neutron shielding materials include mechanical strength, stability, low cost, and ease of handling. During the selection of materials to design a neutron shield, prompt gamma production as well as radionuclide induced by neutron activation must be considered. In this investigation the Monte Carlo method (MCNP code) was used to evaluate the performance of a water-extended polyester shield designed for the transportation, storage, and use of a {sup 252}Cf isotopic neutron source, for comparison the calculations were extended to water shielding, the bare source in vacuum and in air. (authors)

  6. A simple method for correcting the neutron self-shielding effect of matrix and improving the analytical response in prompt gamma-ray neutron activation analysis

    International Nuclear Information System (INIS)

    Sudarshan, K.; Tripathi, R.; Nair, A.G.C.; Acharya, R.; Reddy, A.V.R.; Goswami, A.

    2005-01-01

    A simple method using an internal standard is proposed to correct for the self-shielding effect of B, Cd and Gd in a matrix. This would increase the linear dynamic range of PGNAA in analyzing samples containing these elements. The method is validated by analyzing synthetic samples containing large amounts of B, Cd, Hg and Gd, the elements having high neutron absorption cross-section, in aqueous solutions and solid forms. A simple Monte-Carlo simulation to find the extent of self-shielding in the matrix is presented. The method is applied to the analysis of titanium boride alloy containing large amount of boron. The satisfactory results obtained showed the efficacy of the method of correcting for the self-shielding effects in the sample

  7. The influence of rhodium burn-up on the sensitivity of rhodium self-powered neutron detectors

    International Nuclear Information System (INIS)

    Erben, O.

    1980-01-01

    Depression and self-shielding coefficients are presented for thermal and epithermal neutron flux densities. Functions are shown describing the distribution of beta particle sources on the emitter cross section for 0 to 50% rhodium burnup. The values are calculated of detector sensitivity to thermal and epithermal neutron flux densities for the said burnup for main types of rhodium SPN detectors made by SODERN. (J.B.)

  8. Neutron shielding performance of water-extended polyester

    International Nuclear Information System (INIS)

    Vega Carrillo, H.R.; Manzanares-Acuna, E.; Hernandez-Davila, V.M.; Vega Carrillo, H.R.; Gallegoc, E.; Lorentec, A.; Hernandez-Davila, V.M.

    2006-01-01

    A Monte Carlo study to determine the shielding features to neutrons of water-extended polyester was carried out. Materials with low atomic number are predominantly used for neutron shielding because these materials effectively attenuate neutrons, mainly through elastic and inelastic collisions. In addition to neutron attenuation properties, other desirable properties for neutron shielding materials include mechanical strength, stability, low cost, and ease of handling. During the selection of materials to design a neutron shield, prompt gamma production as well as radionuclide induced by neutron activation must be considered. In this investigation the Monte Carlo method (M.C.N.P. code) was used to evaluate the performance of a water-extended polyester shield designed for the transportation, storage, and use of a 252 Cf isotopic neutron source, for comparison the calculations were extended to water shielding, the bare source in vacuum and in air. (authors)

  9. Development of highly effective neutron shields and neutron absorbing materials

    International Nuclear Information System (INIS)

    Tsuda, K.; Matsuda, F.; Taniuchi, H.; Yuhara, T.; Iida, T.

    1993-01-01

    A wide range of materials, including polymers and hydrogen-occluded alloys that might be usable as the neutron shielding material were examined. And a wide range of materials, including aluminum alloys that might be usable as the neutron-absorbing material were examined. After screening, the candidate material was determined on the basis of evaluation regarding its adaptabilities as a high-performance neutron-shielding and neutron-absorbing material. This candidate material was manufactured for trial, after which material properties tests, neutron-shielding tests and neutron-absorbing tests were carried out on it. The specifications of this material were thus determined. This research has resulted in materials of good performance; a neutron-shielding material based on ethylene propylene rubber and titanium hydride, and a neutron-absorbing material based on aluminum and titanium hydride. (author)

  10. Development of highly effective neutron shielding material made of phenol-novolac type epoxy resin

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Soo Haeng; Jeong, Myeong Soo; Hong, Sun Seok; Lee, Won Kyoung; Kim, Ik Soo; Shin, Young Joon; Do, Jae Bum; Ro, Seung Gy; Oh, Seok Jin

    1998-06-01

    Because the exposure to radiation in the nuclear facilities can be fatal to human, it is important to reduce the radiation dose level to a tolerable level. The purpose of this study is to develop highly effective neutron shielding materials for the shipping and storage cask of radioactive materials or in the nuclear/radiation facilities. On this study, we developed epoxy resin based neutron shielding materials and their various material properties, including neutron shielding ability, fire resistance, combustion characteristics, radiation resistance, thermal and mechanical properties were evaluated experimentally. Especially we developed phenol-novolac type epoxy resin based neutron shielding materials and their characteristics were also evaluated. (author). 22 refs., 11 tabs., 21 figs

  11. Design of a permanent Cd-shielded epithermal neutron irradiation site in the Syrian Miniature Neutron Source Reactor

    International Nuclear Information System (INIS)

    Khattab, K.; Haddad, Kh.; Haj-Hassan, H.

    2008-01-01

    A Cd-shield (cylindrical shell 1 mm in thickness, 34 mm in diameter and 180 mm in length) was used to design a permanent epithermal neutron irradiation site for epithermal neutron activation analysis (ENAA) in the Syrian Miniature Neutron Source Reactor (MNSR). This site was achieved by shielding the surface of the aluminum tube of one of the outer irradiation sites. The calculated depression ratio of thermal neutron flux was 1/10. Homogeneity of the neutron flux in the first outer irradiation site has been found numerically using the WIMSD4 and CITATION codes and experimentally by irradiating five short copper wires using the outer irradiation capsule. Good agreement was obtained between the calculated and the measured results of the neutron flux distributions. (author)

  12. Design of a permanent Cd-shielded epithermal neutron irradiation site in the Syrian Miniature Neutron Source Reactor

    International Nuclear Information System (INIS)

    Khattab, K.; Haddad, Kh.; Haj-Hassan, H.

    2009-01-01

    A Cd-shield (cylindrical shell 1 mm in thickness, 34 mm in diameter and 180 mm in length) was used to design a permanent epithermal neutron irradiation site for epithermal neutron activation analysis (ENAA) in the Syrian Miniature Neutron Source Reactor (MNSR). This site was achieved by shielding the surface of the aluminum tube of one of the outer irradiation sites. The calculated depression ratio of thermal neutron flux was 1/10. Homogeneity of the neutron flux in the first outer irradiation site has been found numerically using the WIMSD4 and CITATION codes and experimentally by irradiating five short copper wires using the outer irradiation capsule. Good agreement was obtained between the calculated and the measured results of the neutron flux distributions. (author)

  13. An ''exact'' treatment of self-shielding and covers in neutron spectra determinations

    International Nuclear Information System (INIS)

    Griffin, P.J.; Kelly, J.G.

    1995-01-01

    Most neutron spectrum determination methodologies ignore self-shielding effects in dosimetry foils and treat covers with an exponential attenuation model. This work provides a quantitative analysis of the approximations in this approach. It also provides a methodology for improving the fidelity of the treatment of the dosimetry sensor response to a level consistent with the user's spectrum characterization approach. A library of correction functions for the energy-dependent sensor response has been compiled that addresses dosimetry foils/configurations in use at the Sandia National Laboratories Radiation Metrology Laboratory

  14. Transparent fast neutron shielding material and shielding method

    International Nuclear Information System (INIS)

    Nashimoto, Tetsuji; Katase, Haruhisa.

    1993-01-01

    Polyisobutylene having a viscosity average molecular weight of 20,000 to 80,000 and a hydrogen atom density of greater than 7.0 x 10 22 /cm 3 is used as a fast neutron shielding material. The shielding material is excellent in the shielding performance against fast neutrons, and there is no worry of leakage even when holes should be formed to a vessel. Further, it is excellent in fabricability, relatively safe even upon occurrence of fire and, in addition, it is transparent to enable to observe contents easily. (T.M.)

  15. Neutron shielding for a {sup 252} Cf source

    Energy Technology Data Exchange (ETDEWEB)

    Vega C, H.R.; Manzanares A, E.; Hernandez D, V.M. [Unidades Academicas de Estudios Nucleares e Ingenieria Electrica, Universidad Autonoma de Zacatecas, C. Cipres 10, Fracc. La Penuela, 98068 Zacatecas (Mexico); Eduardo Gallego, Alfredo Lorente [Depto. de Ingenieria Nuclear, ETS Ingenieros Industriales, Universidad Politecnica de Madrid, C. Jose Gutierrez Abascal 2, 28006 Madrid (Spain)]. e-mail: fermineutron@yahoo.com

    2006-07-01

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

  16. Study of filtration of reactor beam of neutrons with cadmium in a multilayer shielding containing boron carbide

    International Nuclear Information System (INIS)

    Megahid, R.M.; El-Kall, E.H.

    1986-01-01

    Experimental measurements were carried out to study the effect of cadmium on the distribution and attenuation of reactor thermal neutrons emitted from a reactor core and the new thermal neutrons produced in a heterogeneous shield of water, iron, iron + B 4 C and ordinary concrete. The measurements were made using a reactor beam of neutrons filtered with cadmium emitted from one of the horizontal channels of ET-RR-1. It is found that the presence of cadmium sheet at channel exit causes a marked decrease in the thickness of the shield required to attenuate the thermal neutron flux by a certain factor. 12 refs., 5 figures. (author)

  17. Theoretical evaluation of self-shielding factors due to scattering resonances in foils

    International Nuclear Information System (INIS)

    Selander, W.N.

    1960-06-01

    A semi-analytical method is given for evaluating self-shielding factors for activation measurements which use thin foils having neutron scattering resonances. The energy loss by scattering in the foil is taken into account. The energy-dependent neutron angular distribution is expanded as a double series, the coefficients of which are (energy dependent) solutions of an infinite set of coupled integral equations. These are truncated in some suitable manner and solved numerically. The leading term of the series is proportional to the average, or effective flux in the activation sample. The product of this terra and the neutron capture cross-section is integrated numerically over the resonance to give the resonance self-shielding correction. Figure 4 shows resonance self-shielding factors derived in this mariner for the 132ev resonance in Co-59 and figure 5 shows similar results for the two Mn-55 resonances at 337ev and 1080ev. Self-shielding factors for 1/v capture are not significantly different from unity. (author)

  18. The effect of cadmium shielding on the spatial neutron flux distribution inside one of the outer irradiation sites

    International Nuclear Information System (INIS)

    Shaaban, I.

    2009-06-01

    A permanent epithermal neutron irradiation facility was designed in the Syrian Miniature Neutron Source Reactor (MNSR) by using the cadmium (cylindrical vial 1.0 mm in thickness, 38.50 mm in diameter and 180 mm in length) as thermal neutron shielding material, for a permanent epithermal neutron activation analysis (ENAA). This site was designed by shielding the internal surface of the aluminum tube of the first outer irradiation site in the MNSR reactor. I was used the activation detectors 0.1143% Au-Al alloy foils with 0.1 mm thickness and 2.0 mm diameter for measurement the thermal neutron flux, epithermal and R c d=A b are/A c over ratio in the outer irradiation site. Distribution of the thermal neutron flux in the outer irradiation capsule has been found numerically using MCNP-4C code with and without cadmium shield, and experimentally by irradiating five copper wires using the outer irradiation capsule. Good agreements were obtained between the calculated and the measured results. (author)

  19. Neutron and gamma-ray spectra measurement on the model of the KS-150 reactor radial shielding

    International Nuclear Information System (INIS)

    Holman, M.; Hogel, J.; Marik, J.; Kovarik, K.; Franc, L.; Vespalec, R.

    1977-01-01

    A shortened model of the peripheral region of the KS-150 reactor core consisting of two rows of fuel elements and a reflector was constructed from the peripheral fuel elements of the KS-150 reactor core in an experiment on the TR-0 reactor. The mockup of the thermal shield (10 cm of steel), the pressure vessel (15 cm of steel) and the inner wall of the water biological shielding (2 cm of steel) of the KS-150 reactor were erected outside the TR-0 vessel. Fast neutron and gamma spectra were measured with a stilbene crystal scintillation spectrometer. The resonance neutron spectra were measured with 197 Au, 63 Cu and 23 Na resonance activation detectors. Fast neutron spectra inside the reactor were measured with a 10 mm diameter by 10 mm thick stilbene crystal spectrometer, outside the reactor with a 10 mm diameter by 10 mm thick and a 20 mm diameter by 20 mm thick stilbene crystal spectrometer. Neutron spectra in the energy regions of 1 eV to 3 keV and 0.6 MeV to 0.8 MeV were obtained on the core periphery, on the reflector half-thickness and in front of and behind the reactor thermal shield. Gamma spectra were obtained in front of and behind the thermal shield. It was found that the attenuation of neutron fluxes by the reflector and the thermal shield increased with increasing energy while gamma radiation attenuation decreased with increasing energy. It was not possible to obtain the neutron spectrum in the 10 to 600 keV energy range because suitable detection instrumentation was not available. (J.P.)

  20. Monte-Carlo simulations of neutron shielding for the ATLAS forward region

    CERN Document Server

    Stekl, I; Kovalenko, V E; Vorobel, V; Leroy, C; Piquemal, F; Eschbach, R; Marquet, C

    2000-01-01

    The effectiveness of different types of neutron shielding for the ATLAS forward region has been studied by means of Monte-Carlo simulations and compared with the results of an experiment performed at the CERN PS. The simulation code is based on GEANT, FLUKA, MICAP and GAMLIB. GAMLIB is a new library including processes with gamma-rays produced in (n, gamma), (n, n'gamma) neutron reactions and is interfaced to the MICAP code. The effectiveness of different types of shielding against neutrons and gamma-rays, composed from different types of material, such as pure polyethylene, borated polyethylene, lithium-filled polyethylene, lead and iron, were compared. The results from Monte-Carlo simulations were compared to the results obtained from the experiment. The simulation results reproduce the experimental data well. This agreement supports the correctness of the simulation code used to describe the generation, spreading and absorption of neutrons (up to thermal energies) and gamma-rays in the shielding materials....

  1. Resonance self-shielding effect analysis of neutron data libraries applied for the dual-cooled waste transmutation blanket of the fusion-driven subcritical system

    International Nuclear Information System (INIS)

    Liu Haibo; Wu Yican; Zheng Shanliang; Zhang Chunzao

    2004-01-01

    Based on the Fusion-Driven Subcritical System (FDS-I), the 25 groups, 175 groups and 620 groups neutron nuclear data libraries with/without resonance self-shielding correction are made with the Njoy and Transx codes, and the K eff and reaction rates are calculated with the Anisn code. The conclusion indicates that the resonance self-shielding effect affects the reaction rates strongly. (authors)

  2. Studying the shielding properties of lead glass composites using neutrons and gamma rays

    International Nuclear Information System (INIS)

    Osman, A.M.; El-Sarraf, M.A.; Abdel-Monem, A.M.; El-Sayed Abdo, A.

    2015-01-01

    Highlights: • Samples of sodalime silica glass loaded with different ratios of PbO were prepared. • Leaded glass composites were investigated for radiation shielding. • Experimental and theoretical attenuation parameters were studied. • Experimental and theoretical (MCNP5) results were in good agreement. - Abstract: The present work deals with the shielding properties of lead glass composites to find out its integrity for practical shielding applications and radiological safety. Composites of different lead oxide ratios (x = 0, 5, 10, 15 and 25 wt.%) have been prepared by the Nasser Glass and Crystal Company (Egypt). Attenuation measurements have been carried out using a collimated emitted beam from a fission 252 Cf (100 μg) neutron source, and the neutron–gamma spectrometer with stilbene scintillator. The pulse shape discriminating (P.S.D.) technique based on the zero cross-over method was used to discriminate between neutron and gamma-ray pulses. Thermal neutron fluxes were measured using the BF3 detector and thermal neutron detection system. The attenuation relations were used to evaluate fast neutron macroscopic effective removal cross-section Σ R-Meas (cm −1 ), gamma rays total attenuation coefficient μ (cm −1 ) and thermal neutron macroscopic cross-section Σ Meas (cm −1 ). Theoretical calculations have been achieved using MCNP5 code to calculate the same two parameters. Also, MERCSF-N program was used to calculate fast neutron macroscopic removal cross-section Σ R-MER (cm −1 ). Measured and MCNP5 calculated results have been compared and were found to be in reasonable agreement

  3. Effect of neutrons scattered from boundary of neutron field on shielding experiment

    International Nuclear Information System (INIS)

    Ogawa, Tatsuhiko; Abe, Takuya; Kosako, Toshiso; Iimoto, Takeshi

    2009-01-01

    Neutron shielding experiment with 49 cm-thick ordinary concrete was carried out at the reactor 'Yayoi' The University of Tokyo. System of this experiment is enclosed by heavy concrete where neutrons backscattered from heavy concrete likely affected neutron flux on the back surface of shielding concrete. Reaction rate of 197 Au(n, γ), cadmium covered 197 Au(n, γ) and 115 In(n, n') in the shielding concrete was measured using foil activation method. Neutron transport calculation was carried out in order to simulate reaction rate by calculating neutron spectra and convoluting with neutron capture cross-section in neutron shielding concrete. Comparison was made between calculated reaction rate and experimental one, and almost satisfactory agreement was found except for the back surface of shielding. To compose adequate simulation model, description of heavy concrete behind the shielding was thought to be of importance. For example, disregarding neutrons backscattered from heavy concrete, calculation underestimated reaction rate by the factor of 10. In another example, assuming that chemical composition of heavy concrete is equal to the composition adopted from a literature, the reaction rate was overestimated by factor of 5. By making the composition of heavy concrete equal to that based on facility design, overestimation was found to be the factor of 2. Therefore, adequate description of chemical composition of heavy concrete is found to be of importance in order to simulate neutron induced reaction rate on the back surface of neutron shielding concrete in shielding experiment performed in a system enclosed by heavy concrete. (author)

  4. Heat generation and temperature-rise in ordinary concrete due to capture of thermal neutrons

    International Nuclear Information System (INIS)

    Abdo, E.A.; Amin, E.

    1997-01-01

    The aim of this work is the evaluation of the heat generation and temperature-rise in local ordinary concrete as a biological shield due to capture of total thermal and reactor thermal neutrons. The total thermal neutron fluxes were measured and calculated. The channel number 2 of the ETRR-1 reactor was used in the measurements as a neutron source. Computer code ANISN (VAX version) and neutron multigroup cross-section library EURLiB-4 was used in the calculations. The heat generation and temperature-rise in local ordinary concrete were evaluated and calculated. The results were displayed in curves to show the distribution of thermal neutron fluxes and heat generation as well as temperature-rise with the shield thickness. The results showed that, the heat generation as well as the temperature-rise have their maximum values in the first layers of the shield thickness. 4 figs., 12 refs

  5. Physical, mechanical and neutron shielding properties of h-BN/Gd2O3/HDPE ternary nanocomposites

    Science.gov (United States)

    İrim, Ş. Gözde; Wis, Abdulmounem Alchekh; Keskin, M. Aker; Baykara, Oktay; Ozkoc, Guralp; Avcı, Ahmet; Doğru, Mahmut; Karakoç, Mesut

    2018-03-01

    In order to prepare an effective neutron shielding material, not only neutron but also gamma absorption must be taken into account. In this research, a polymer nanocomposite based novel type of multifunctional neutron shielding material is designed and fabricated. For this purpose, high density polyethylene (HDPE) was compounded with different amounts of hexagonal boron nitride (h-BN) and Gd2O3 nanoparticles having average particle size of 100 nm using melt-compounding technique. The mechanical, thermal and morphological properties of nanocomposites were investigated. As filler content increased, the absorption of both neutron and gamma fluxes increased despite fluctuating neutron absorption curves. Adding h-BN and Gd2O3 nano particles had a significant influence on both neutron and gamma attenuation properties (Σ, cm-1 and μ/ρ, cm-2/g) of ternary shields and they show an enhancement of 200-280%, 14-52% for neutron and gamma radiations, respectively, in shielding performance.

  6. Graphs of neutron cross sections in JSD1000 for radiation shielding safety analysis

    International Nuclear Information System (INIS)

    Yamano, Naoki

    1984-03-01

    Graphs of neutron cross sections and self-shielding factors in the JSD1000 library are presented for radiation shielding safety analysis. The compilation contains various reaction cross sections for 42 nuclides from 1 H to 241 Am in the energy range from 3.51 x 10 -4 eV to 16.5 MeV. The Bondarenko-type self-shielding factors of each reaction are given by the background cross sections from σ 0 = 0 to σ 0 = 10000. (author)

  7. A new formulation for resonance self-shielding factors

    International Nuclear Information System (INIS)

    Palma, Daniel A.P.; Martinez, Aquilino S.; Silva, Fernando C. da

    2007-01-01

    The activation technique allows either absolute or relative very precise neutron intensity measurements. This technique requires the knowledge of the Doppler broadening function to determine resonance self-shielding factors. In the present work a new formulation is proposed for the self-shielding factors where the Doppler broadening function is calculated using the Frobenius's method and compared to the values obtained from the four-pole Pade method. This calculation method is shown to be effective from the point of view of accuracy. (author)

  8. A study on the characteristics of modified and novolac type epoxy resin based neutron shielding material

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Soo Haeng; Hong, Sun Seok; Oh, Seung Chul; Do, Jae Bum [Korea Atomic Energy Research Institute, Taejon (Korea)

    1998-10-01

    Because the exposure to radiation in the nuclear facilities can be fatal to human, it is important to reduce the radiation dose level to a tolerable level. The purpose of this study is to develop highly effective neutron shielding materials for the shipping and storage cask of radioactive materials or in the nuclear/radiation facilities. In this study, we developed modified and novolac type epoxy resin based neutron shielding materials and their various material properties, including neutron shielding ability, prolonged time heat resistance, thermal and mechanical properties were evaluated experimently. (author). 31 refs., 27 figs., 16 tabs.

  9. Development of neutron shielding concrete containing iron content materials

    Science.gov (United States)

    Sariyer, Demet; Küçer, Rahmi

    2018-02-01

    Concrete is one of the most important construction materials which widely used as a neutron shielding. Neutron shield is obtained of interaction with matter depends on neutron energy and the density of the shielding material. Shielding properties of concrete could be improved by changing its composition and density. High density materials such as iron or high atomic number elements are added to concrete to increase the radiation resistance property. In this study, shielding properties of concrete were investigated by adding iron, FeB, Fe2B, stainless - steel at different ratios into concrete. Neutron dose distributions and shield design was obtained by using FLUKA Monte Carlo code. The determined shield thicknesses vary depending on the densities of the mixture formed by the additional material and ratio. It is seen that a combination of iron rich materials is enhanced the neutron shielding of capabilities of concrete. Also, the thicknesses of shield are reduced.

  10. Directional epithermal neutron detector

    International Nuclear Information System (INIS)

    Givens, W.W.; Mills, W.R. Jr.

    1986-01-01

    A borehole tool for epithermal neutron die-away logging of subterranean formations surrounding a borehole is described which consists of: (a) a pulsed source of fast neutrons for irradiating the formations surrounding a borehole, (b) at least one neutron counter for counting epithermal neutrons returning to the borehole from the irradiated formations, (c) a neutron moderating material, (d) an outer thermal neutron shield providing a housing for the counter and the moderating material, (e) an inner thermal neutron shield dividing the housing so as to provide a first compartment bounded by the inner thermal neutron shield and a first portion of the outer thermal neutron shield and a second compartment bounded by the inner thermal neutron shield and a second portion of the outer thermal neutron shield, the counter being positioned within the first compartment and the moderating material being positioned within the second compartment, and (f) means for positioning the borehole tool against one side of the borehole wall and azimuthally orienting the borehole tool such that the first chamber is in juxtaposition with the borehole wall, the formation epithermal neutrons penetrating into the first chamber through the first portion of the outer thermal neutron shield are detected by the neutron counter for die-away measurement, thereby maximizing the directional sensitivty of the neutron counter to formation epithermal neutrons, the borehole fluid epithermal neutrons penetrating into the second chamber through the second chamber through the second portion of the outer thermal neutron shield are largely slowed down and lowered in energy by the moderating material and absorbed by the inner thermal neutron shield before penetrating into the first chamber, thereby minimizing the directional sensitivity of the neutron counter to borehole fluid epithermal neutrons

  11. Thermal neutron radiative capture cross-section of 186W(n, γ)187W reaction

    International Nuclear Information System (INIS)

    Tan, V H; Son, P N

    2016-01-01

    The thermal neutron radiative capture cross section for 186 W(n, γ) 187 W reaction was measured by the activation method using the filtered neutron beam at the Dalat research reactor. An optimal composition of Si and Bi, in single crystal form, has been used as neutron filters to create the high-purity filtered neutron beam with Cadmium ratio of R cd = 420 and peak energy E n = 0.025 eV. The induced activities in the irradiated samples were measured by a high resolution HPGe digital gamma-ray spectrometer. The present result of cross section has been determined relatively to the reference value of the standard reaction 197 Au(n, γ) 198 Au. The necessary correction factors for gamma-ray true coincidence summing, and thermal neutron self-shielding effects were taken into account in this experiment by Monte Carlo simulations. (paper)

  12. Determination of self shielding factors and gamma attenuation effects for tree ring samples

    International Nuclear Information System (INIS)

    Dagistan Sahin; Kenan Uenlue

    2012-01-01

    Determination of tree ring chemistry using Neutron Activation Analysis (NAA) is part of an ongoing research between Penn State University (PSU) and Cornell University, The Malcolm and Carolyn Wiener Laboratory for Aegean and Near Eastern Dendrochronology. Tree-ring chemistry yields valuable data for environmental event signatures. These signatures are a complex function of elemental concentration. To be certain about concentration of signature elements, it is necessary to perform the measurements and corrections with the lowest error and maximum accuracy possible. Accurate and precise values of energy dependent neutron flux at dry irradiation tubes and detector efficiency for tree ring sample are calculated for Penn State Breazeale Reactor (PSBR). For the calculation of energy dependent and self shielding corrected neutron flux, detailed model of the TRIGA Mark III reactor at PSU with updated fuel compositions was prepared using the MCNP utility for reactor evolution (MURE) libraries. Dry irradiation tube, sample holder and sample were also included in the model. The thermal flux self-shielding correction factors due to the sample holder and sample for were calculated and verified with previously published values. The Geant-4 model of the gamma spectroscopy system, developed at Radiation Science and Engineering Center (RSEC), was improved and absolute detector efficiency for tree-ring samples was calculated. (author)

  13. A new formulation for resonance self-shielding factors

    Energy Technology Data Exchange (ETDEWEB)

    Palma, Daniel A.P.; Martinez, Aquilino S.; Silva, Fernando C. da [Universidade Federal, Rio de Janeiro, RJ (Brazil). Coordenacao dos Programas de Pos-graduacao de Engenharia. Programa de Engenharia Nuclear]. E-mail: aquilino@lmp.ufrj.br

    2007-07-01

    The activation technique allows either absolute or relative very precise neutron intensity measurements. This technique requires the knowledge of the Doppler broadening function to determine resonance self-shielding factors. In the present work a new formulation is proposed for the self-shielding factors where the Doppler broadening function is calculated using the Frobenius's method and compared to the values obtained from the four-pole Pade method. This calculation method is shown to be effective from the point of view of accuracy. (author)

  14. Design of incoming neutron-beam for detecting oil dirt

    International Nuclear Information System (INIS)

    Zhao Jingwu; Chen Xiaocheng; Alimujiang Naimaiti; Aierken Abuliemu

    2012-01-01

    For the technique of neutron back-scattering, the neutron counts are non-linear and have a tendency toward saturation because of the neutron self-shielding. As a result, the measurement accuracy is reduced and the measurement range is limited. Using a simply model and comparing with experimental data, it is shown that, in the measurement of the thickness of oil dirt, by adjusting the ratio of thermal to epithermal neutrons, the neutron self: shielding is weakened. As a result, the non-linearity can be reduced and the measurement accuracy and range can be improved. (authors)

  15. Development of Neutron Shielding Material for Cask and Accelerator

    International Nuclear Information System (INIS)

    Kang, Hee Young; Seo, Ki Seog; Lee, Byung Chul; Park, Chang Jae; Kim, Ho Dong

    2008-01-01

    The neutron shielding materials are used as a neutron shield for spent fuel shipping cask, beam accelerators and neutron generators. At early stage, the neutron attenuations of materials were evaluated with the cross sections. After that, benchmark or mock-up experiments on the multi-layer problem to confirm the shielding characteristics or to evaluate analysis accuracy were reported. Recently, the need to transport spent nuclear fuels is increasing due to the current limited storage capacity. The on-site storage capacity at some of nuclear power plants is expected to be full in near future. With a growing inventory of spent fuels at power plants, these spent fuels need to be transported to other storage facilities. Shipping casks have been developed to safely transport spent fuels that emit high neutrons and gamma-ray radiation. The external radiation level of the shipping cask from the spent fuel must be limited to meet the standards specified by the IAEA radioactive material package regulation, so it is important to develop a proper neutron shielding material for a shipping cask. Neutron shielding experiments and analyses on the shielding effects of materials have been conducted, and some experiments have been performed to examine the shielding effects of selected materials. The shielding experiments consist of evaluating not only the shielding effects of a material alone but also the effects of the material thickness. The experimental results were compared with those obtained by using the MCNP-5c code

  16. A practical neutron shielding design based on data-base interpolation

    International Nuclear Information System (INIS)

    Jiang, S.H.; Sheu, R.J.

    1993-01-01

    Neutron shielding design is an important part of the construction of nuclear reactors and high-energy accelerators. Neutron shielding design is also indispensable in the packaging and storage of isotopic neutron sources. Most efforts in the development of neutron shielding design have been concentrated on nuclear reactor shielding because of its huge mass and strict requirement of accuracy. Sophisticated computational tools, such as transport and Monte Carlo codes and detailed data libraries have been developed. In principle, now, neutron shielding, in spite of its complexity, can be designed in any detail and with fine accuracy. However, in most practical cases, neutron shielding design is accomplished with simplified methods. Unlike practical gamma-ray shielding design, where exponential attenuation coupled with buildup factors has been applied effectively and accurately, simplified neutron shielding design, either by using removal cross sections or by applying charts or tables of transmission factors such as the National Council on Radiation Protection and Measurements (NCRP) 38 (Ref. 1) for general neutron protection or to NCRP 51 (Ref. 2) for accelerator neutron shielding, is still very primitive and not well established. The available data are limited in energy range, materials, and thicknesses, and the estimated results are only roughly accurate. It is the purpose of this work to establish a simple, convenient, and user-friendly general-purpose computational tool for practical preliminary neutron shielding design that is reasonably accurate. A wide-range (energy, material, and thickness) data base of dose transmission factors has been generated by applying one-dimensional transport calculations in slab geometry

  17. Comprehensive analysis of shielding effectiveness for HDPE, BPE and concrete as candidate materials for neutron shielding

    International Nuclear Information System (INIS)

    Dhang, Prosenjit; Verma, Rishi; Shyam, Anurag

    2015-01-01

    In the compact accelerator based DD neutron generator, the deuterium ions generated by the ion source are accelerated after the extraction and bombarded to a deuterated titanium target. The emitted neutrons have typical energy of ∼2.45MeV. Utilization of these compact accelerator based neutron generators of yield up to 10 9 neutron/second (DD) is under active consideration in many research laboratories for conducting active neutron interrogation experiments. Requirement of an adequately shielded laboratory is mandatory for the effective and safe utilization of these generators for intended applications. In this reference, we report the comprehensive analysis of shielding effectiveness for High Density Polyethylene (HDPE), Borated Polyethylene (BPE) and Concrete as candidate materials for neutron shielding. In shielding calculations, neutron induced scattering and absorption gamma dose has also been considered along with neutron dose. Contemporarily any material with higher hydrogenous concentration is best suited for neutron shielding. Choice of shielding material is also dominated by practical issues like economic viability and availability of space. Our computational analysis results reveal that utilization of BPE sheets results in minimum wall thickness requirement for attaining similar range of attenuation in neutron and gamma dose. The added advantage of using borated polyethylene is that it reduces the effect of both neutron and gamma dose by absorbing neutron and producing lithium and alpha particle. It has also been realized that for deciding upon optimum thickness determination of any shielding material, three important factors to be necessarily considered are: use factor, occupancy factor and work load factor. (author)

  18. Rhodium self-powered neutron detector as a suitable on-line thermal neutron flux monitor in BNCT treatments.

    Science.gov (United States)

    Miller, Marcelo E; Sztejnberg, Manuel L; González, Sara J; Thorp, Silvia I; Longhino, Juan M; Estryk, Guillermo

    2011-12-01

    A rhodium self-powered neutron detector (Rh SPND) has been specifically developed by the Comisión Nacional de Energía Atómica (CNEA) of Argentina to measure locally and in real time thermal neutron fluxes in patients treated with boron neutron capture therapy (BNCT). In this work, the thermal and epithermal neutron response of the Rh SPND was evaluated by studying the detector response to two different reactor spectra. In addition, during clinical trials of the BNCT Project of the CNEA, on-line neutron flux measurements using the specially designed detector were assessed. The first calibration of the detector was done with the well-thermalized neutron spectrum of the CNEA RA-3 reactor thermal column. For this purpose, the reactor spectrum was approximated by a Maxwell-Boltzmann distribution in the thermal energy range. The second calibration was done at different positions along the central axis of a water-filled cylindrical phantom, placed in the mixed thermal-epithermal neutron beam of CNEA RA-6 reactor. In this latter case, the RA-6 neutron spectrum had been well characterized by both calculation and measurement, and it presented some marked differences with the ideal spectrum considered for SPND calibrations at RA-3. In addition, the RA-6 neutron spectrum varied with depth in the water phantom and thus the percentage of the epithermal contribution to the total neutron flux changed at each measurement location. Local (one point-position) and global (several points-positions) and thermal and mixed-field thermal neutron sensitivities were determined from these measurements. Thermal neutron flux was also measured during BNCT clinical trials within the irradiation fields incident on the patients. In order to achieve this, the detector was placed on patient's skin at dosimetric reference points for each one of the fields. System stability was adequate for this kind of measurement. Local mixed-field thermal neutron sensitivities and global thermal and mixed

  19. Rhodium self-powered neutron detector as a suitable on-line thermal neutron flux monitor in BNCT treatments

    Energy Technology Data Exchange (ETDEWEB)

    Miller, Marcelo E.; Sztejnberg, Manuel L.; Gonzalez, Sara J.; Thorp, Silvia I.; Longhino, Juan M.; Estryk, Guillermo [Comision Nacional de Energia Atomica, Av. del Libertador 8250, Ciudad de Buenos Aires 1429 (Argentina); Comision Nacional de Energia Atomica, Av. del Libertador 8250, Ciudad de Buenos Aires 1429, Argentina and CONICET, Av. Rivadavia 1917, Ciudad de Buenos Aires 1033 (Argentina); Comision Nacional de Energia Atomica, Av. del Libertador 8250, Ciudad de Buenos Aires 1429 (Argentina)

    2011-12-15

    Purpose: A rhodium self-powered neutron detector (Rh SPND) has been specifically developed by the Comision Nacional de Energia Atomica (CNEA) of Argentina to measure locally and in real time thermal neutron fluxes in patients treated with boron neutron capture therapy (BNCT). In this work, the thermal and epithermal neutron response of the Rh SPND was evaluated by studying the detector response to two different reactor spectra. In addition, during clinical trials of the BNCT Project of the CNEA, on-line neutron flux measurements using the specially designed detector were assessed. Methods: The first calibration of the detector was done with the well-thermalized neutron spectrum of the CNEA RA-3 reactor thermal column. For this purpose, the reactor spectrum was approximated by a Maxwell-Boltzmann distribution in the thermal energy range. The second calibration was done at different positions along the central axis of a water-filled cylindrical phantom, placed in the mixed thermal-epithermal neutron beam of CNEA RA-6 reactor. In this latter case, the RA-6 neutron spectrum had been well characterized by both calculation and measurement, and it presented some marked differences with the ideal spectrum considered for SPND calibrations at RA-3. In addition, the RA-6 neutron spectrum varied with depth in the water phantom and thus the percentage of the epithermal contribution to the total neutron flux changed at each measurement location. Local (one point-position) and global (several points-positions) and thermal and mixed-field thermal neutron sensitivities were determined from these measurements. Thermal neutron flux was also measured during BNCT clinical trials within the irradiation fields incident on the patients. In order to achieve this, the detector was placed on patient's skin at dosimetric reference points for each one of the fields. System stability was adequate for this kind of measurement. Results: Local mixed-field thermal neutron sensitivities and

  20. Rhodium self-powered neutron detector as a suitable on-line thermal neutron flux monitor in BNCT treatments

    International Nuclear Information System (INIS)

    Miller, Marcelo E.; Sztejnberg, Manuel L.; Gonzalez, Sara J.; Thorp, Silvia I.; Longhino, Juan M.; Estryk, Guillermo

    2011-01-01

    Purpose: A rhodium self-powered neutron detector (Rh SPND) has been specifically developed by the Comision Nacional de Energia Atomica (CNEA) of Argentina to measure locally and in real time thermal neutron fluxes in patients treated with boron neutron capture therapy (BNCT). In this work, the thermal and epithermal neutron response of the Rh SPND was evaluated by studying the detector response to two different reactor spectra. In addition, during clinical trials of the BNCT Project of the CNEA, on-line neutron flux measurements using the specially designed detector were assessed. Methods: The first calibration of the detector was done with the well-thermalized neutron spectrum of the CNEA RA-3 reactor thermal column. For this purpose, the reactor spectrum was approximated by a Maxwell-Boltzmann distribution in the thermal energy range. The second calibration was done at different positions along the central axis of a water-filled cylindrical phantom, placed in the mixed thermal-epithermal neutron beam of CNEA RA-6 reactor. In this latter case, the RA-6 neutron spectrum had been well characterized by both calculation and measurement, and it presented some marked differences with the ideal spectrum considered for SPND calibrations at RA-3. In addition, the RA-6 neutron spectrum varied with depth in the water phantom and thus the percentage of the epithermal contribution to the total neutron flux changed at each measurement location. Local (one point-position) and global (several points-positions) and thermal and mixed-field thermal neutron sensitivities were determined from these measurements. Thermal neutron flux was also measured during BNCT clinical trials within the irradiation fields incident on the patients. In order to achieve this, the detector was placed on patient's skin at dosimetric reference points for each one of the fields. System stability was adequate for this kind of measurement. Results: Local mixed-field thermal neutron sensitivities and global

  1. Connection factor calculation for isotopic neutron flux measurements with foil detectors

    International Nuclear Information System (INIS)

    Avila L, J.

    1987-01-01

    Thermal and resonance neutron self-shielding factors, neutron flux distortion and edge effects as well as a connection factor for neutron flux profile around a foil detector have been calculated. A general expression for resonance self shielding factor is presented in order to take into account the most important resonances for a given isotope. A computer program SPRESYTER.BAS was written and results for In-115 and Au-197 foils are given

  2. URR [Unresolved Resonance Region] computer code: A code to calculate resonance neutron cross-section probability tables, Bondarenko self-shielding factors, and self-indication ratios for fissile and fertile nuclides

    International Nuclear Information System (INIS)

    Leal, L.C.; de Saussure, G.; Perez, R.B.

    1990-01-01

    The URR computer code has been developed to calculate cross-section probability tables, Bondarenko self-shielding factors, and self-indication ratios for fertile and fissile isotopes in the unresolved resonance region. Monte Carlo methods are utilized to select appropriate resonance parameters and to compute the cross sections at the desired reference energy. The neutron cross sections are calculated by the single-level Breit-Wigner formalism with s-, p-, and d-wave contributions. The cross-section probability tables are constructed by sampling by Doppler broadened cross-sections. The various self-shielding factors are computer numerically as Lebesgue integrals over the cross-section probability tables

  3. Accelerator shield design of KIPT neutron source facility

    International Nuclear Information System (INIS)

    Zhong, Z.; Gohar, Y.

    2013-01-01

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

  4. Insufficient self-shielding correction in VITAMIN-B6

    International Nuclear Information System (INIS)

    Konno, Chikara; Ochiai, Kentaro; Ohnishi, Seiki

    2011-01-01

    We carried out a simple benchmark calculation test with a multigroup cross-section library VITAMIN-B6 generated from ENDF/B-VI. The model of this test consisted of an iron sphere of 1 m in radius with an isotropic 20 MeV neutron source in the center. Neutron spectra in the sphere were calculated with an Sn code ANISN and VITAMIN-B6 or FENDL/MG-1.1. A calculation with MCNP and ENDF/B-VI was carried out as a reference. The neutron spectra with ANISN and FENDL/MG-1.1 agreed with those with MCNP, while those with ANISN and VITAMIN-B6 were at most 50% different from those with MCNP. We uncovered that the discrepancy came from insufficient self-shielding correction due to the followings; 1) The smallest background cross section of 56 Fe in VITAMIN-B6 is 1. 2) The weighting flux used in generating VITAMIN-B6 is not adequate. VITAMIN-B6 should be revised for adequate self-shielding correction. (author)

  5. Shielding practice

    International Nuclear Information System (INIS)

    Sauermann, P.F.

    1985-08-01

    The basis of shielding practice against external irradiation is shown in a simple way. For most sources of radiation (point sources) occurring in shielding practice, the basic data are given, mainly in the form of tables, which are required to solve the shielding problems. The application of these data is explained and discussed using practical examples. Thickness of shielding panes of glove boxes for α and β radiation; shielding of sealed γ-radiography sources; shielding of a Co-60 radiation source, and of the manipulator panels for hot cells; damping factors for γ radiation and neutrons; shielding of fast and thermal neutrons, and of bremsstrahlung (X-ray tubes, Kr-85 pressure gas cylinders, 42 MeV betatrons, 20 MeV linacs); two-fold shielding (lead glass windows for hot cells, 14 MeV neutron generators); shielding against scattered radiation. (orig./HP) [de

  6. Thermal-hydraulic analysis and design improvement for coolant channel of ITER shield block

    International Nuclear Information System (INIS)

    Zhao Ling; Li Huaqi; Zheng Jiantao; Yi Jingwei; Kang Weishan; Chen Jiming

    2013-01-01

    As an important part for ITER, shield block is used to shield the neutron heat. The structure design of shield block, especially the inner coolant channel design will influence its cooling effect and safety significantly. In this study, the thermal-hydraulic analysis for shield block has been performed by the computational fluid dynamics software, some optimization suggestions have been proposed and thermal-hydraulic characteristics of the improved model has been analyzed again. The analysis results for improved model show that pressure drop through flow path near the inlet and outlet region of the shield block has been reduced, and the total pressure drop in cooling path has been reduced too; the uniformity of the mass flowrate distribution and the velocity distribution have been improved in main cooling branches; the local highest temperature of solid domain reduced considerably, which could avoid thermal stress becoming too large because of coolant effect unevenly. (authors)

  7. Development of neutron shielding material using metathesis-polymer matrix

    Energy Technology Data Exchange (ETDEWEB)

    Sakurai, Yoshinori E-mail: ysakurai@rri.kyoto-u.ac.jp; Sasaki, Akira; Kobayashi, Tooru

    2004-04-21

    A neutron shielding material using a metathesis-polymer matrix, which is a thermosetting resin, was developed. This shielding material has characteristics that can be controlled for different mixing ratios of neutron absorbers and for formation in the laboratory. Additionally, the elastic modulus can be changed at the hardening process, from a flexible elastoma to a mechanically tough solid. Experiments were performed at the Kyoto University Research Reactor in order to determine the important characteristics of this metathesis-polymer shielding material, such as neutron shielding performance, secondary gamma-ray generation and activation. The metathesis-polymer shielding material was shown to be practical and as effective as the other available shielding materials, which mainly consist of thermoplastic resin.

  8. DEMONR, Monte-Carlo Shielding Calculation for Neutron Flux and Neutron Spectra, Teaching Program

    International Nuclear Information System (INIS)

    Courtney, J. C.

    1987-01-01

    1 - Description of problem or function: DEMONR treats the behavior of neutrons in a slab shield. It is frequently used as a teaching tool. 2 - Method of solution: An unbiased Monte Carlo code calculates the number, energy, and direction of neutrons that penetrate or are reflected from a shield. 3 - Restrictions on the complexity of the problem: Only one shield may be used in each problem. The shield material may be a single element or a homogeneous mixture of elements with a single effective atomic weight. Only elastic scattering and neutron capture processes are allowed. The source is a point located on one face of the slab. It provides a cosine distribution of current. Monoenergetic or fission spectrum neutrons may be selected

  9. Thermal behavior of neutron shielding material, NS-4-FR, under long term storage conditions

    International Nuclear Information System (INIS)

    Yamada, N.; O-iwa, A.; Asano, R.; Horita, R.; Kusunoki, K.

    2004-01-01

    NS-4-FR, Epoxy-Resin, has been widely used as a neutron shielding material for casks. It is recognized that the resin will degrade during storage and loose weight under high temperature conditions. Most of the examinations for the resin degrading behavior were conducted with rather small bare resin specimens. However, the actual quantity of neutron shielding is quite large and is covered by the cask body. To confirm the degrading behavior of the resin under the long-term storage conditions, we performed the test on the specimen with the same cross-section as the actual design, Hitz B69. The resin test vessels were made out of stainless steel and equipped with flange

  10. Enhancement of thermal neutron attenuation of nano-B4C, -BN dispersed neutron shielding polymer nanocomposites

    International Nuclear Information System (INIS)

    Kim, Jaewoo; Lee, Byung-Chul; Uhm, Young Rang; Miller, William H.

    2014-01-01

    Highlights: • Preparation of B 4 C and BN nanopowders using a simple ball milling process. • Homogeneous dispersion and strong adhesion of nano-B 4 C and -BN with polymer matrix. • Enhancement of mechanical properties of the nanocomposites compared to their micro counterparts. • Enhancement of thermal neutron attenuation of the nanocomposites. - Abstract: Nano-sized boron carbide (B 4 C) and boron nitride (BN) powder were prepared using ball milling. Micro- and milled nano-powders were melt blended with high density polyethylene (HDPE) using a polymer mixer followed by hot pressing to fabricate sheet composites. The tensile and flexural strengths of HDPE nanocomposites were ∼20% higher than their micro counterparts, while those for latter decreased compared to neat HDPE. Thermal neutrons attenuation of the prepared HDPE nanocomposites was evaluated using a monochromatic ∼0.025 eV neutron beam. Thermal neutron attenuation of the HDPE nanocomposites was greatly enhanced compared to their micro counterparts at the same B-10 areal densities. Monte Carlo n-Particles (MCNP) simulations based on the lattice structure modeling also shows the similar filler size dependent thermal neutron absorption

  11. Enhancement in the microstructure and neutron shielding efficiency of sandwich type of 6061Al–B4C composite material via hot isostatic pressing

    International Nuclear Information System (INIS)

    Park, Jin-Ju; Hong, Sung-Mo; Lee, Min-Ku; Rhee, Chang-Kyu; Rhee, Won-Hyuk

    2015-01-01

    Highlights: • 6061Al–B 4 C neutron shielding composites are fabricated by sintering and HIP. • HIP process improves the wettability of B 4 C particles into 6061Al matrix. • Neutron attenuation performance can be enhanced by application of HIP process. - Abstract: Sandwich type of 6061Al–B 4 C composite plates, which are used as a thermal neutron absorber for spent nuclear fuel pool storage rack, were fabricated using two different consolidation ways as sintering and hot isostatic pressing (HIP) processes and their thermal neutron shielding efficiency was investigated as a function of B 4 C concentration ranging from 0 to 40 wt.%. For this purpose, two respective inner core compaction parts of sintered and HIPped neutron absorbing composite materials were first produced and then cladded them between two outer plates by HIP process. The application of HIP process provided not only a lead of excellent interfacial adhesion due to the improved wettability but also an enhancement of thermal neutron shielding efficiency owing to the more uniform dispersion of B 4 C particles

  12. Comparison of Thermal Neutron Flux Measured by Uranium 235 Fission Chamber and Rhodium Self-Powered Neutron Detector in MTR

    International Nuclear Information System (INIS)

    Fourmentel, D.; Filliatre, P.; Barbot, L.; Villard, J.-F.; Lyoussi, A.; Geslot, B.; Malo, J.-Y.; Carcreff, H.; Reynard-Carette, C.

    2013-06-01

    Thermal neutron flux is one of the most important nuclear parameter to be measured on-line in Material Testing Reactors (MTRs). In particular two types of sensors with different physical operating principles are commonly used: self-powered neutron detectors (SPND) and fission chambers with uranium 235 coating. This work aims to compare on one hand the thermal neutron flux evaluation given by these two types of sensors and on the other hand to compare these evaluations with activation dosimeter measurements, which are considered as the reference for absolute neutron flux assessment. This study was conducted in an irradiation experiment, called CARMEN-1, performed during 2012 in OSIRIS reactor (CEA Saclay - France). The CARMEN-1 experiment aims to improve the neutron and photon flux and nuclear heating measurements in MTRs. In this paper we focus on the thermal neutron flux measurements performed in CARMEN-1 experiment. The use of fission chambers to measure the absolute thermal neutron flux in MTRs is not very usual. An innovative calibration method for fission chambers operated in Campbell mode has been developed at the CEA Cadarache (France) and tested for the first time in the CARMEN-1 experiment. The results of these measurements are discussed, with the objective to measure with the best accuracy the thermal neutron flux in the future Jules Horowitz Reactor. (authors)

  13. Resonance self-shielding methodology of new neutron transport code STREAM

    International Nuclear Information System (INIS)

    Choi, Sooyoung; Lee, Hyunsuk; Lee, Deokjung; Hong, Ser Gi

    2015-01-01

    This paper reports on the development and verification of three new resonance self-shielding methods. The verifications were performed using the new neutron transport code, STREAM. The new methodologies encompass the extension of energy range for resonance treatment, the development of optimum rational approximation, and the application of resonance treatment to isotopes in the cladding region. (1) The extended resonance energy range treatment has been developed to treat the resonances below 4 eV of three resonance isotopes and shows significant improvements in the accuracy of effective cross sections (XSs) in that energy range. (2) The optimum rational approximation can eliminate the geometric limitations of the conventional approach of equivalence theory and can also improve the accuracy of fuel escape probability. (3) The cladding resonance treatment method makes it possible to treat resonances in cladding material which have not been treated explicitly in the conventional methods. These three new methods have been implemented in the new lattice physics code STREAM and the improvement in the accuracy of effective XSs is demonstrated through detailed verification calculations. (author)

  14. Method to produce a neutron shielding

    International Nuclear Information System (INIS)

    Merkle, H.J.

    1978-01-01

    The neutron shielding for armoured vehicles consists of preshaped plastic plates which are coated on the armoured vehicle walls by conversion of the thermoplast. Suitable plastics or thermoplasts are PVC, PVC acetate, or mixtures of these, into which more than 50% B, B 4 C, or BN is embedded. The colour of the shielding may be determined by the choice of the neutron absorber, e.g. a white colour for BN. The plates are produced using an extruder or calender. (DG) [de

  15. Preliminary neutron shielding calculations of the electronics in the EAST BES systems focusing on neutron induced displacement damage

    Energy Technology Data Exchange (ETDEWEB)

    Náfrádi, Gábor, E-mail: nafradi@reak.bme.hu [Institute of Nuclear Techniques (NTI), Budapest University of Technology and Economics (BME), H-1111 Budapest (Hungary); Kovácsik, Ákos, E-mail: kovacsik.akos@reak.bme.hu [Institute of Nuclear Techniques (NTI), Budapest University of Technology and Economics (BME), H-1111 Budapest (Hungary); Németh, József, E-mail: nemeth.jozsef@wigner.mta.hu [Institute for Particle and Nuclear Physics, Wigner Research Centre for Physics (Wigner RCP), Hungarian Academy of Sciences (HAS), POB 49, 1525 Budapest (Hungary); Pór, Gábor, E-mail: por@reak.bme.hu [Institute of Nuclear Techniques (NTI), Budapest University of Technology and Economics (BME), H-1111 Budapest (Hungary); Zoletnik, Sándor, E-mail: zoletnik.sandor@wigner.mta.hu [Institute for Particle and Nuclear Physics, Wigner Research Centre for Physics (Wigner RCP), Hungarian Academy of Sciences (HAS), POB 49, 1525 Budapest (Hungary)

    2016-11-15

    Monte Carlo N-Particle (MCNP) calculations were carried out to compare neutron shielding capabilities of three frequently used neutron shielding materials: polyethylene without neutron absorbers, polyethylene with boron absorbers and polyethylene with lithium absorbers, according to Non Ionizing Energy Loss (NIEL). The results of 1D shielding calculations showed that simple neutron moderating materials can provide sufficient and cheap shielding against 2.45 MeV and 14.1 MeV fusion neutrons, in terms of 1 MeV neutron equivalent flux, in silicon targets, which is the most commonly used material of electronic components. Based on these results a new shielding concept is proposed which can be taken into consideration where the reduction of displacement damage is the main goal and the free space available for shielding is limited. Based on this shielding concept detailed 3D calculations were carried out to describe the properties of the neutron shielding of the Beam Emission Spectroscopy (BES) system installed at the EAST tokamak.

  16. Design, fabrication, and properties of a continuous carbon-fiber reinforced Sm_2O_3/polyimide gamma ray/neutron shielding material

    International Nuclear Information System (INIS)

    Wang, Peng; Tang, Xiaobin; Chai, Hao; Chen, Da; Qiu, Yunlong

    2015-01-01

    Highlights: • Sm_2O_3 is used for neutron absorber instead of B_4C, and Sm_2O_3 has a good photon-shielding effect. • Carbon-fiber cloth and polyimide were used to enhance shielding materials’ mechanical behavior and thermal behavior. • Both Monte Carlo method and shielding test were used to evaluate shielding performance of the novel shielding material. - Abstract: The design and fabrication of shielding materials with good heat-resistance and mechanical properties is a major problem in the radiation shielding field. In this paper, based on gamma ray and neutron shielding theory, a continuous carbon-fiber reinforced Sm_2O_3/polyimide gamma ray/neutron shielding material was fabricated by hot-pressing method. The material's application behavior was subsequently evaluated using neutron shielding, photon shielding, mechanical tensile, and thermogravimetric analysis–differential scanning calorimetry tests. The results show that the tensile strength of the novel shielding material exceeds 200 MPa, which makes it of similar strength to aluminum alloy. The material does not undergo crosslinking and decomposition reactions at 300 °C and it can be used in such environments for long periods of time. The continuous carbon-fiber reinforced Sm_2O_3/polyimide material has a good shielding performance with respect to gamma rays and neutrons. The material thus has good prospects for use in fusion reactor system and nuclear waste disposal applications.

  17. Method for limiting movement of a thermal shield for a nuclear reactor, and thermal shield displacement limiter therefor

    International Nuclear Information System (INIS)

    Meuschke, R.E.; Boyd, C.H.

    1989-01-01

    This patent describes a method of limiting the movement of a thermal shield of a nuclear reactor. It comprises: machining at least four (4) pockets in upper portions of a thermal shield circumferentially about a core barrel of a nuclear reactor to receive key-wave inserts; tapping bolt holes in the pockets of the thermal shield to receive bolts; positioning key-wave inserts into the pockets of the thermal shield to be bolted in place with the bolt holes; machining dowel holes at least partially through the positioned key-way inserts and the thermal shield to receive dowel pins; positioning dowel pins in the dowel holes in the key-way insert and thermal shield to tangentially restrain movement of the thermal shield relative to the core barrel; sliding limiter keys into the key-way inserts and bolting the limiter keys to the core barrel to tangentially restrain movement of the thermal shield relative and the core barrel while allowing radial and axial movement of the thermal shield relative to the core barrel; machining dowel holes through the limiter key and at least partially through the core barrel to receive dowel pins; positioning dowel pins in the dowel holes in the limiter key and core barrel to restrain tangential movement of the thermal shield relative to the core barrel of the nuclear reactor

  18. Gamma self-shielding correction factors calculation for aqueous bulk sample analysis by PGNAA technique

    International Nuclear Information System (INIS)

    Nasrabadi, M.N.; Mohammadi, A.; Jalali, M.

    2009-01-01

    In this paper bulk sample prompt gamma neutron activation analysis (BSPGNAA) was applied to aqueous sample analysis using a relative method. For elemental analysis of an unknown bulk sample, gamma self-shielding coefficient was required. Gamma self-shielding coefficient of unknown samples was estimated by an experimental method and also by MCNP code calculation. The proposed methodology can be used for the determination of the elemental concentration of unknown aqueous samples by BSPGNAA where knowledge of the gamma self-shielding within the sample volume is required.

  19. An accuracy estimation on neutron penetration calculation through concrete shield with PALLAS codes using bunched component nuclides of concrete

    International Nuclear Information System (INIS)

    Sasamoto, Nobuo; Kotegawa, Hiroshi

    1984-11-01

    In order to improve computational efficiency of PALLAS code, an accuracy is estimated on the neutron penetration calculation through a concrete shield, using bunched component nuclides of concrete. The calculated fast neutron flux is observed to depend weakly on how the nuclides are bunched. Contrary to this, the calculated thermal neutron fluxes are strongly dependent on the manner of bunching, mainly due to the fact that iron cross section has exceptionally large negative sensitivity to thermal neutron flux. (author)

  20. SHREDI, Neutron Flux and Neutron Activation in 2-D Shields by Removal Diffusion

    International Nuclear Information System (INIS)

    Daneri, A.; Toselli, G.

    1976-01-01

    1 - Nature of physical problem solved: SHREDI is a removal - diffusion neutron shielding code. The program computes neutron fluxes and activations in bidimensional sections (x,y or r,z) of the shield. It is also possible to consider shielding points with the same y or z coordinate (mono-dimensional problems). 2 - Method of solution: The integrals which define the removal fluxes are computed in some shield points by means of a particular algorithm based on the Simpson's and trapezoidal rules. For the diffusion calculation the finite difference method is used. The removal sources are interpolated in all diffusion points by Chebyshev polynomials. 3 - Restrictions on the complexity of the problem: Maxima: number of removal energy groups NGR = 40; number of diffusion energy groups NGD = 40; number of the reactor core and shield materials NCMP = 50; number of core mesh points in r (or x) direction for integral calculation = 75; number of core mesh points in z (or y) direction for integral calculation = 75; number of core mesh points in theta (or z) direction for integral calculation = 75; number of shield mesh points for the neutron flux calculation in r (or x) direction NPX = 200; number of shield mesh points for the neutron flux calculation in z (or y) direction NPY = 200; n.b. (NPX * NPY) le 12000

  1. Development and testing of multigroup library with correction of self-shielding effects in fusion-fission hybrid reactor

    International Nuclear Information System (INIS)

    Zou Jun; He Zhaozhong; Zeng Qin; Qiu Yuefeng; Wang Minghuang

    2010-01-01

    A multigroup library HENDL2.1/SS (Hybrid Evaluated Nuclear Data Library/Self-Shielding) based on ENDF/B-VII.0 evaluate data has been generated using Bondarenko and flux calculator method for the correction of self-shielding effect of neutronics analyses. To validate the reliability of the multigroup library HENDL2.1/SS, transport calculations for fusion-fission hybrid system FDS-I were performed in this paper. It was verified that the calculations with the HENDL2.1/SS gave almost the same results with MCNP calculations and were better than calculations with the HENDL2.0/MG which is another multigroup library without self-shielding correction. The test results also showed that neglecting resonance self-shielding caused underestimation of the K eff , neutron fluxes and waste transmutation ratios in the multigroup calculations of FDS-I.

  2. Thermal neutron flux measurement using self-powered neutron detector (SPND) at out-core locations of TRIGA PUSPATI Reactor (RTP)

    Science.gov (United States)

    Ali, Nur Syazwani Mohd; Hamzah, Khaidzir; Mohamad Idris, Faridah; Hairie Rabir, Mohamad

    2018-01-01

    The thermal neutron flux measurement has been conducted at the out-core location using self-powered neutron detectors (SPNDs). This work represents the first attempt to study SPNDs as neutron flux sensor for developing the fault detection system (FDS) focusing on neutron flux parameters. The study was conducted to test the reliability of the SPND’s signal by measuring the neutron flux through the interaction between neutrons and emitter materials of the SPNDs. Three SPNDs were used to measure the flux at four different radial locations which located at the fission chamber cylinder, 10cm above graphite reflector, between graphite reflector and tank liner and fuel rack. The measurements were conducted at 750 kW reactor power. The outputs from SPNDs were collected through data acquisition system and were corrected to obtain the actual neutron flux due to delayed responses from SPNDs. The measurements showed that thermal neutron flux between fission chamber location near to the tank liner and fuel rack were between 5.18 × 1011 nv to 8.45 × 109 nv. The average thermal neutron flux showed a good agreement with those from previous studies that has been made using simulation at the same core configuration at the nearest irradiation facilities with detector locations.

  3. Electron accelerator shielding design of KIPT neutron source facility

    Energy Technology Data Exchange (ETDEWEB)

    Zhong, Zhao Peng; Gohar, Yousry [Argonne National Laboratory, Argonne (United States)

    2016-06-15

    The Argonne National Laboratory of the United States and the Kharkov Institute of Physics and Technology of the Ukraine have been collaborating on the design, development and construction of a neutron source facility at Kharkov Institute of Physics and Technology utilizing an electron-accelerator-driven subcritical assembly. The electron beam power is 100 kW using 100-MeV electrons. The facility was designed to perform basic and applied nuclear research, produce medical isotopes, and train nuclear specialists. The biological shield of the accelerator building was designed to reduce the biological dose to less than 5.0e-03 mSv/h during operation. The main source of the biological dose for the accelerator building is the photons and neutrons generated from different interactions of leaked electrons from the electron gun and the accelerator sections with the surrounding components and materials. The Monte Carlo N-particle extended code (MCNPX) was used for the shielding calculations because of its capability to perform electron-, photon-, and neutron-coupled transport simulations. The photon dose was tallied using the MCNPX calculation, starting with the leaked electrons. However, it is difficult to accurately tally the neutron dose directly from the leaked electrons. The neutron yield per electron from the interactions with the surrounding components is very small, ∼0.01 neutron for 100-MeV electron and even smaller for lower-energy electrons. This causes difficulties for the Monte Carlo analyses and consumes tremendous computation resources for tallying the neutron dose outside the shield boundary with an acceptable accuracy. To avoid these difficulties, the SOURCE and TALLYX user subroutines of MCNPX were utilized for this study. The generated neutrons were banked, together with all related parameters, for a subsequent MCNPX calculation to obtain the neutron dose. The weight windows variance reduction technique was also utilized for both neutron and photon dose

  4. URR [Unresolved Resonance Region] computer code: A code to calculate resonance neutron cross-section probability tables, Bondarenko self-shielding factors, and self-indication ratios for fissile and fertile nuclides

    International Nuclear Information System (INIS)

    Leal, L.C.; de Saussure, G.; Perez, R.B.

    1989-01-01

    The URR computer code has been developed to calculate cross-section probability tables, Bondarenko self-shielding factors, and self- indication ratios for fertile and fissile isotopes in the unresolved resonance region. Monte Carlo methods are utilized to select appropriate resonance parameters and to compute the cross sections at the desired reference energy. The neutron cross sections are calculated by the single-level Breit-Wigner formalism with s-, p-, and d-wave contributions. The cross-section probability tables are constructed by sampling the Doppler broadened cross-section. The various shelf-shielded factors are computed numerically as Lebesgue integrals over the cross-section probability tables. 6 refs

  5. Enhancement of thermal neutron attenuation of nano-B{sub 4}C, -BN dispersed neutron shielding polymer nanocomposites

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jaewoo, E-mail: kimj@kaeri.re.kr [Nuclear Materials Research Division, Korea Atomic Energy Research Institute, 111-989 Daeduck-daero, Yuseong-gu, Daejeon-si 305-353 (Korea, Republic of); WCI Quantum Beam based Radiation Research Center, Korea Atomic Energy Research Institute, 111-989 Daeduck-daero, Yuseong-gu, Daejeon-si 305-353 (Korea, Republic of); Missouri University Research Reactor, University of Missouri-Columbia, Columbia, MO 65211 (United States); Lee, Byung-Chul [Nuclear Reactor Core Design Division, Korea Atomic Energy Research Institute, 111-989 Daeduck-daero, Yuseong-gu, Daejeon-si 305-353 (Korea, Republic of); Uhm, Young Rang [Radioisotopes Research Division, Korea Atomic Energy Research Institute, 111-989 Daeduck-daero, Yuseong-gu, Daejeon-si 305-353 (Korea, Republic of); Miller, William H. [Missouri University Research Reactor, University of Missouri-Columbia, Columbia, MO 65211 (United States)

    2014-10-15

    Highlights: • Preparation of B{sub 4}C and BN nanopowders using a simple ball milling process. • Homogeneous dispersion and strong adhesion of nano-B{sub 4}C and -BN with polymer matrix. • Enhancement of mechanical properties of the nanocomposites compared to their micro counterparts. • Enhancement of thermal neutron attenuation of the nanocomposites. - Abstract: Nano-sized boron carbide (B{sub 4}C) and boron nitride (BN) powder were prepared using ball milling. Micro- and milled nano-powders were melt blended with high density polyethylene (HDPE) using a polymer mixer followed by hot pressing to fabricate sheet composites. The tensile and flexural strengths of HDPE nanocomposites were ∼20% higher than their micro counterparts, while those for latter decreased compared to neat HDPE. Thermal neutrons attenuation of the prepared HDPE nanocomposites was evaluated using a monochromatic ∼0.025 eV neutron beam. Thermal neutron attenuation of the HDPE nanocomposites was greatly enhanced compared to their micro counterparts at the same B-10 areal densities. Monte Carlo n-Particles (MCNP) simulations based on the lattice structure modeling also shows the similar filler size dependent thermal neutron absorption.

  6. Characterization of the Ljubljana TRIGA thermal column neutron radiographic facility

    International Nuclear Information System (INIS)

    Nemec, T.; Rant, J.; Kristof, E.; Glumac, B.

    1995-01-01

    An extensive characterization of the neutron beam of the existing neutron radiographic facility in the thermal column of the Ljubljana Triga Mark II research reactor is in progress. Neutron beam characteristics are needed to determine the effect of various neutron and gamma radiation on the neutron radiographic image. Commercially available medical scintillator converter screens based on Gd dioxy sulphite as well as Gd metal neutron converters are used to record neutron radiographic image. Thermal, epithermal and fast neutron fluxes were measured using Au and In activation detectors and cadmium ratio is determined. Neutron beam flux profiles are measured by film densitometry and by Au activation detector wires. By exposing films shielded by boral or lead plates individual contributions of thermal, epithermal neutrons and gamma radiation are estimated by densitometric measurements. By recording images of neutron image quality indicators BPI (Beam Purity Indicator) and SI (Sensitivity Indicator) produced by Riso, standard neutron radiography image characteristic are established. In gamma dosimetric measurements thermoluminescent detectors (CaF 2 Mn) are used. (author)

  7. Shielding experiments in different materials with 252Cf neutron spectra

    International Nuclear Information System (INIS)

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

    2008-01-01

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

  8. Neutron shielding properties of a new high-density concrete

    International Nuclear Information System (INIS)

    Lorente, A.; Gallego, E.; Vega Carrillo, H.R.; Mendez, R.

    2008-01-01

    The neutron shielding properties of a new high-density concrete (commercially available under the name Hormirad TM , developed in Spain by the company CT-RAD) have been characterized both experimentally and by Monte Carlo calculations. The shielding properties of this concrete against photons were previously studied and the material is being used to build bunkers, mazes and doors in medical accelerator facilities with good overall results. In this work, the objective was to characterize the material behaviour against neutrons, as well as to test alternative mixings including boron compounds in an effort to improve neutron shielding efficiency. With that purpose, Hormirad TM slabs of different thicknesses were exposed to an 241 Am-Be neutron source under controlled conditions in the neutron measurements laboratory of the Nuclear Engineering Department at UPM. The original mix, which includes a high fraction of magnetite, was then modified by adding different proportions of anhydrous borax (Na 2 B 4 O 7 ). In order to have a reference against common concrete used to shield medical accelerator facilities, the same experiment was repeated with ordinary (HA-25) concrete slabs. In parallel to the experiments, Monte Carlo calculations of the experiments were performed with MCNP5. The experimental results agree reasonably well with the Monte Carlo calculations. Therefore, the first and equilibrium tenth-value layers have been determined for the different types of concrete tested. The results show an advantageous behaviour of the Hormirad TM concrete, in terms of neutron attenuation against real thickness of the shielding. Borated concretes seem less practical since they did not show better neutron attenuation with respect to real thickness and their structural properties are worse. The neutron attenuation properties of Hormirad TM for typical neutron spectra in clinical LINAC accelerators rooms have been also characterized by Monte Carlo calculation. (author)

  9. Shielding evaluation of neutron generator hall by Monte Carlo simulations

    Energy Technology Data Exchange (ETDEWEB)

    Pujala, U.; Selvakumaran, T.S.; Baskaran, R.; Venkatraman, B. [Radiological Safety Division, Indira Gandhi Center for Atomic Research, Kalpakkam (India); Thilagam, L.; Mohapatra, D.K., E-mail: swathythila2@yahoo.com [Safety Research Institute, Atomic Energy Regulatory Board, Kalpakkam (India)

    2017-04-01

    A shielded hall was constructed for accommodating a D-D, D-T or D-Be based pulsed neutron generator (NG) with 4π yield of 10{sup 9} n/s. The neutron shield design of the facility was optimized using NCRP-51 methodology such that the total dose rates outside the hall areas are well below the regulatory limit for full occupancy criterion (1 μSv/h). However, the total dose rates at roof top, cooling room trench exit and labyrinth exit were found to be above this limit for the optimized design. Hence, additional neutron shielding arrangements were proposed for cooling room trench and labyrinth exits. The roof top was made inaccessible. The present study is an attempt to evaluate the neutron and associated capture gamma transport through the bulk shields for the complete geometry and materials of the NG-Hall using Monte Carlo (MC) codes MCNP and FLUKA. The neutron source terms of D-D, D-T and D-Be reactions are considered in the simulations. The effect of additional shielding proposed has been demonstrated through the simulations carried out with the consideration of the additional shielding for D-Be neutron source term. The results MC simulations using two different codes are found to be consistent with each other for neutron dose rate estimates. However, deviation up to 28% is noted between these two codes at few locations for capture gamma dose rate estimates. Overall, the dose rates estimated by MC simulations including additional shields shows that all the locations surrounding the hall satisfy the full occupancy criteria for all three types of sources. Additionally, the dose rates due to direct transmission of primary neutrons estimated by FLUKA are compared with the values calculated using the formula given in NCRP-51 which shows deviations up to 50% with each other. The details of MC simulations and NCRP-51 methodology for the estimation of primary neutron dose rate along with the results are presented in this paper. (author)

  10. Self-shielding factors

    International Nuclear Information System (INIS)

    Kaul, D.C.

    1982-01-01

    Throughout the last two decades many efforts have been made to estimate the effect of body self-shielding on organ doses from externally incident neutrons and gamma rays. These began with the use of simple geometry phantoms and have culminated in the use of detailed anthropomorphic phantoms. In a recent effort, adjoint Monte Carlo analysis techniques have been used to determine dose and dose equivalent to the active marrow as a function of energy and angle of neutron fluence externally incident on an anthropomorphic phantom. When combined with fluences from actual nuclear devices, these dose-to-fluence factors result in marrow dose values that demonstrate great sensitivity to variations in device type, range, and body orientation. Under a state-of-the-art radiation transport analysis demonstration program for the Japanese cities, sponsored by the Defense Nuclear Agency at the request of the National Council on Radiation Protection and Measurements, the marrow dose study referred to above is being repeated to obtain spectral distributions within the marrow for externally incident neutrons and gamma rays of arbitrary energy and angle. This is intended to allow radiobiologists and epidemiologists to select and to modify numbers of merit for correlation with health effects and to permit a greater understanding of the relationship between human and laboratory subject dosimetry

  11. Study of ceramic mixed boron element as a neutron shielding

    International Nuclear Information System (INIS)

    Ismail Mustapha; Mohd Reusmaazran Yusof; Md Fakarudin Ab Rahman; Nor Paiza Mohamad Hasan; Samihah Mustaffha; Yusof Abdullah; Mohamad Rabaie Shari; Airwan Affandi Mahmood; Nurliyana Abdullah; Hearie Hassan

    2012-01-01

    Shielding upon radiation should not be underestimated as it can causes hazard to health. Precautions on the released of radioactive materials should be well concerned and considered. Therefore, the combination of ceramic and boron make them very useful for shielding purpose in areas of low and intermediate neutron. A six grades of ceramic tile have been produced namely IMN05 - 5 % boron, IMN06 - 6 % boron, IMN07 - 7 % boron, IMN08 - 8 % boron, IMN09 - 9 % boron, IMN10 - 10 % boron from mixing, press and sintered process. Boron is a material that capable of absorbing and capturing neutron, so that neutron and gamma test were conducted to analyze the effectiveness of boron material in combination with ceramic as shielding. From the finding, percent reduction number of count per minute shows the ceramic tiles are capable to capture neutron. Apart from all the percentage of boron used, 10 % is the most effective shields since the percent reduction indicating greater neutron captured increased. (author)

  12. Shielding for neutrons produced by medical linear accelerators

    International Nuclear Information System (INIS)

    Rebello, Wilson F.; Silva, Ademir X.

    2007-01-01

    The shielding system called Multileaf Shielding (MLS) was designed in Brazil to be used for protection patients, who undergo radiotherapy treatment, against undesired neutrons produced in the medical linear accelerator heads. During the conceiving of the MLS it was necessary to evaluate its efficiency. For that purpose, several simulations using the Monte Carlo N-particle radiation transport code, MCNP5, were made, in order to evaluate the response of the new shielding system. The results showed a significant neutron dose reduction after the inclusion of the MLS. This work aims to presenting these simulation results. (author)

  13. Radiation shielding material characterization by non-destructive neutron radiography technique

    International Nuclear Information System (INIS)

    Hafizal Yazid; Azali Muhammad; Abdul Aziz Mohamed; Rafhayudi Jamro; Hishamuddin Husain

    2007-01-01

    Shielding property of boronated rubber was characterized easily by the use of neutron radiography technique. For 10 phr of boron carbide in the natural rubber composite, the ability to completely shield against neutron was found to have 8mm thickness and above for the neutron flux of 1.04 x 10 5 n/cm 2 s (author)

  14. Validation of a new 39 neutron group self-shielded library based on the nucleonics analysis of the Lotus fusion-fission hybrid test facility performed with the Monte Carlo code

    International Nuclear Information System (INIS)

    Pelloni, S.; Cheng, E.T.

    1985-02-01

    The Swiss LOTUS fusion-fission hybrid test facility was used to investigate the influence of the self-shielding of resonance cross sections on the tritium breeding and on the thorium ratios. Nucleonic analyses were performed using the discrete-ordinates transport codes ANISN and ONEDANT, the surface-flux code SURCU, and the version 3 of the MCNP code for the Li 2 CO 3 and the Li 2 O blanket designs with lead, thorium and beryllium multipliers. Except for the MCNP calculation which bases on the ENDF/B-V files, all nuclear data are generated from the ENDF/B-IV basic library. For the deterministic methods three NJOY group libraries were considered. The first, a 39 neutron group self-shielded library, was generated at EIR. The second bases on the same group structure as the first does and consists of infinitely diluted cross sections. Finally the third library was processed at LANL and consists of coupled 30+12 neutron and gamma groups; these cross sections are not self-shielded. The Monte Carlo analysis bases on a continuous and on a discrete 262 group library from the ENDF/B-V evaluation. It is shown that the results agree well within 3% between the unshielded libraries and between the different transport codes and theories. The self-shielding of resonance cross sections results in a decrease of the thorium capture rate and in an increase of the tritium breeding of about 6%. The remaining computed ratios are not affected by the self-shielding of cross sections. (Auth.)

  15. Estimation of dose distribution and neutron spectra in JCO critical accident by shielding calculations

    International Nuclear Information System (INIS)

    Sakamoto, Yukio

    2001-01-01

    The information about neutrons at the surrounding of JCO site in the critical accident is limited to survey results by neutron Rem counter in the period of accident and activation data very near the test facility measured after the shut down of accident. This caused the big uncertainty in the dose estimation by detailed shielding calculation codes. On the other hand, environmental activity data measured by radiochemical researchers included the information about fast neutrons inside of JCO site and thermal neutrons up to 1 km from test facility. It is important to grasp the actual circumstance and examine the executed evaluation of the critical accident as scientifically as possible. Therefore, it is meaningful for different field researchers to corporate and exchange the information. In the Technical Divisions of Radiation Science and Technology in Atomic Energy Society of Japan, the information about neutron spectra are released from their home page and three groups of JAERI/CRC, Sumitomo Atomic Energy Industry and Nuclear Power Engineering Corp. (NUPEC)/Mitsubishi Research Institute Inc. (MRI), tried the shielding calculation by Monte Carlo Code MCNP-4B. The procedures and main results of shielding calculations were reviewed in this report. The main difference of shielding calculation by three groups was density and water content of autoclaved light-weight concrete (ALC) as the wall and ceiling. From the result by NUPEC/MRI, it was estimated that the water content in ALC was from 0.05 g/cm 3 to 0.10 g/cm 3 . The behavior of dose equivalent attenuation obtained by shielding calculation was very similar with the measured data from 250 m to 1,700 m obtained by survey meter, TLD and monitoring post. For more exact dose estimation, more detail examination of density and water content of ALC will be needed. (author)

  16. A code for leakage neutron spectra through thick shields

    International Nuclear Information System (INIS)

    Nagarajan, P.S.; Sethulakshmi, P.; Raghavendran, C.P.

    1975-01-01

    An exponential transform Monte Carlo code has been developed for deep penetration of neutrons and the results of leakage neutron spectra of this code have been compared with those of a basic Monte Carlo code for small thickness. The development of the code and optimisation of certain transform parameters are discussed and results are presented for a few thick shields of concrete and water in the context of neutron monitoring in the environs of accelerator and reactor shields. (author)

  17. Neutronics investigation of advanced self-cooled liquid blanket systems in helical reactor

    International Nuclear Information System (INIS)

    Tanaka, T.; Sagara, A.; Muroga, T.; Youssef, M.Z.

    2006-10-01

    Neutronics performances of advanced self-cooled liquid blanket systems have been investigated in design activity of the helical-type reactor FFHR2. In the present study, a new three-dimensional (3-D) neutronics calculation system has been developed for the helical-type reactor to enhance quick feedback between neutronics evaluation and design modification. Using this new calculation system, advanced Flibe-cooled and Li-cooled liquid blanket systems proposed for FFHR2 have been evaluated to make clear design issues to enhance neutronics performance. Based on calculated results, modification of the blanket dimensions and configuration have been attempted to achieve the adequate tritium breeding ability and neutron shielding performance in the helical reactor. The total tritium breeding ratios (TBRs) obtained after modifying the blanket dimensions indicated that all the advanced blanket systems proposed for FFHR2 would achieve adequate tritium self-sufficiency by dimension adjustment and optimization of structures in the breeder layers. Issues in neutron shielding performance have been investigated quantitatively using 3-D geometry of the helical blanket system, support structures, poloidal coils etc. Shielding performance of the helical coils against direct neutrons from core plasma would achieve design target by further optimization of shielding materials. However, suppression of the neutron streaming and reflection through the divertor pumping areas in the original design is important issue to protect the poloidal coils and helical coils, respectively. Investigation of the neutron wall loading indicated that the peaking factor of the neutron wall load distribution would be moderated by the toroidal and helical effect of the plasma distribution in the helical reactor. (author)

  18. Study of neutron and gamma shielding by lead borate and bismuth lead borate glasses: transparent radiation shielding

    International Nuclear Information System (INIS)

    Singh, Vishwanath P.; Badiger, N.M.

    2013-01-01

    Radiation shielding for gamma and neutron is the prominent area in nuclear reactor technology, medical application, dosimetry and other industries. Shielding of these types of radiation requires an appropriate concrete with mixture of low-to-high Z elements which is an opaque medium. The transparent radiation shielding in visible light for gamma and neutron is also extremely essential in the nuclear facilities as lead window. Presently various types of lead equivalent glass oxides have been invented which are transparent as well as provide protection from radiation. In our study we have assessment of effectiveness of neutron and gamma radiation shielding of xPbO.(1-x) B 2 O 3 (x=0.15 to 0.60) and xBi 2 O 3 .(0.80-x) PbO.0.20 B 2 O 3 (x=0.10 to 0.70) transparent borate and bismuth glasses by NXCOM program. The neutron effective mass removal cross section, Σ R /ρ (cm 2 /g) of the lead, bismuth and boron oxides are given. We found invariable Σ R /ρ of various combinations of the lead borate glass for x=0.15 to 0.60 and bismuth lead borate glass for x=0.10 to 0.70. It is observed that the effective removal cross-section for fast neutron (cm -1 ) of lead borate reduces significantly whereas roughly constant for bismuth borate. The gamma mass attenuation coefficients (μ/ρ) of the glasses were also compared with possible experimental values and found comparable. High (μ/ρ) for gamma radiation of the bismuth glasses shows that it is better gamma shielding compared with lead containing glass. However lead borate glasses are better neutron shielding as the neutron removal coefficient are higher. Our investigation is very useful for nuclear reactor technology where prompt neutron of energy 17 MeV and gamma photon up to 10 MeV produced. (author)

  19. Assessment of fast and thermal neutron ambient dose equivalents around the KFUPM neutron source storage area using nuclear track detectors

    Energy Technology Data Exchange (ETDEWEB)

    Fazal-ur-Rehman [Physics Department, King Fahd University of Petroleum and Minerals, Dhahran 31261 (Saudi Arabia)]. E-mail: fazalr@kfupm.edu.sa; Al-Jarallah, M.I. [Physics Department, King Fahd University of Petroleum and Minerals, Dhahran 31261 (Saudi Arabia); Abu-Jarad, F. [Radiation Protection Unit, Environmental Protection Department, Saudi Aramco, P. O. Box 13027, Dhahran 31311 (Saudi Arabia); Qureshi, M.A. [Center for Applied Physical Sciences, King Fahd University of Petroleum and Minerals, Dhahran 31261 (Saudi Arabia)

    2005-11-15

    A set of five {sup 241}Am-Be neutron sources are utilized in research and teaching at King Fahd University of Petroleum and Minerals (KFUPM). Three of these sources have an activity of 16Ci each and the other two are of 5Ci each. A well-shielded storage area was designed for these sources. The aim of the study is to check the effectiveness of shielding of the KFUPM neutron source storage area. Poly allyl diglycol carbonate (PADC) Nuclear track detectors (NTDs) based fast and thermal neutron area passive dosimeters have been utilized side by side for 33 days to assess accumulated low ambient dose equivalents of fast and thermal neutrons at 30 different locations around the source storage area and adjacent rooms. Fast neutron measurements have been carried out using bare NTDs, which register fast neutrons through recoils of protons, in the detector material. NTDs were mounted with lithium tetra borate (Li{sub 2}B{sub 4}O{sub 7}) converters on their surfaces for thermal neutron detection via B10(n,{alpha})Li6 and Li6(n,{alpha})H3 nuclear reactions. The calibration factors of NTD both for fast and thermal neutron area passive dosimeters were determined using thermoluminescent dosimeters (TLD) with and without a polyethylene moderator. The calibration factors for fast and thermal neutron area passive dosimeters were found to be 1.33 proton tracks cm{sup -2}{mu}Sv{sup -1} and 31.5 alpha tracks cm{sup -2}{mu}Sv{sup -1}, respectively. The results show variations of accumulated dose with the locations around the storage area. The fast neutron dose equivalents rates varied from as low as 182nSvh{sup -1} up to 10.4{mu}Svh{sup -1} whereas those for thermal neutron ranged from as low as 7nSvh{sup -1} up to 9.3{mu}Svh{sup -1}. The study indicates that the area passive neutron dosimeter was able to detect dose rates as low as 7 and 182nSvh{sup -1} from accumulated dose for thermal and fast neutrons, respectively, which were not possible to detect with the available active neutron

  20. Thermal neutron equivalent doses assessment around KFUPM neutron source storage area using NTDs

    Energy Technology Data Exchange (ETDEWEB)

    Abu-Jarad, F.; Fazal-ur-Rehman; Al-Haddad, M.N.; Al-Jarrallah, M.I.; Nassar, R

    2002-07-01

    Area passive neutron dosemeters based on nuclear track detectors (NTDs) have been used for 13 days to assess accumulated low doses of thermal neutrons around neutron source storage area of the King Fahd University of Petroleum and Minerals (KFUPM). Moreover, the aim of this study is to check the effectiveness of shielding of the storage area. NTDs were mounted with the boron converter on their surface as one compressed unit. The converter is a lithium tetraborate (Li{sub 2}B{sub 4}O{sub 7}) layer for thermal neutron detection via {sup 10}B(N,{alpha}){sup 7}Li and {sup 6}Li(n,{alpha}){sup 3}H nuclear reactions. The area passive dosemeters were installed on 26 different locations around the source storage area and adjacent rooms. The calibration factor for NTD-based area passive neutron dosemeters was found to be 8.3 alpha tracks.cm{sup -2}.{mu}Sv{sup -1} using active snoopy neutron dosemeters in the KFUPM neutron irradiation facility. The results show the variation of accumulated dose with locations around the storage area. The range of dose rates varied from as low as 40 nSv.h{sup -1} up to 11 {mu}Sv.h{sup -1}. The study indicates that the area passive neutron dosemeter was able to detect accumulated doses as low as 40 nSv.h{sup -1}, which could not be detected with the available active neutron dosemeters. The results of the study also indicate that an additional shielding is required to bring the dose rates down to background level. The present investigation suggests extending this study to find the contribution of doses from fast neutrons around the neutron source storage area using NTDs through proton recoil. The significance of this passive technique is that it is highly sensitive and does not require any electronics or power supplies, as is the case in active systems. (author)

  1. Neutron dosimeter

    International Nuclear Information System (INIS)

    Bartko, J.; Schoch, K.F. Jr.; Congedo, T.V.; Anderson, S.L. Jr.

    1989-01-01

    This patent describes a nuclear reactor. It comprises a reactor core; a thermal shield surrounding the reactor core; a pressure vessel surrounding the thermal shield; a neutron dosimeter positioned outside of the thermal shield, the neutron dosimeter comprising a layer of fissile material and a second layer made of a material having an electrical conductivity which permanently varies as a function of its cumulative ion radiation dose; and means, outside the pressure vessel and electrically connected to the layer of second material, for measuring electrical conductivity of the layer of second material

  2. Thermal neutron capture and resonance integral cross sections of {sup 45}Sc

    Energy Technology Data Exchange (ETDEWEB)

    Van Do, Nguyen; Duc Khue, Pham; Tien Thanh, Kim [Institute of Physics, Vietnam Academy of Science and Technology, 10 Dao Tan, Hanoi (Viet Nam); Thi Hien, Nguyen [Institute of Physics, Vietnam Academy of Science and Technology, 10 Dao Tan, Hanoi (Viet Nam); Department of Physics and Center for High Energy Physics, Kyungpook National University, Daegu 702-701 (Korea, Republic of); Kim, Guinyun, E-mail: gnkim@knu.ac.kr [Department of Physics and Center for High Energy Physics, Kyungpook National University, Daegu 702-701 (Korea, Republic of); Kim, Kwangsoo [Department of Physics and Center for High Energy Physics, Kyungpook National University, Daegu 702-701 (Korea, Republic of); Shin, Sung-Gyun; Cho, Moo-Hyun [Department of Advanced Nuclear Engineering, Pohang University of Science and Technology, Pohang 790-784 (Korea, Republic of); Lee, Manwoo [Research Center, Dongnam Institute of Radiological and Medical Science, Busan 619-953 (Korea, Republic of)

    2015-11-01

    The thermal neutron cross section (σ{sub 0}) and resonance integral (I{sub 0}) of the {sup 45}Sc(n,γ){sup 46}Sc reaction have been measured relative to that of the {sup 197}Au(n,γ){sup 198}Au reaction by means of the activation method. High-purity natural scandium and gold foils without and with a cadmium cover of 0.5 mm thickness were irradiated with moderated pulsed neutrons produced from the Pohang Neutron Facility (PNF). The induced activities in the activated foils were measured with a high purity germanium (HPGe) detector. In order to improve the accuracy of the experimental results the counting losses caused by the thermal (G{sub th}) and resonance (G{sub epi}) neutron self-shielding, the γ-ray attenuation (F{sub g}) and the true γ-ray coincidence summing effects were made. In addition, the effect of non-ideal epithermal spectrum was also taken into account by determining the neutron spectrum shape factor (α). The thermal neutron cross-section and resonance integral of the {sup 45}Sc(n,γ){sup 46}Sc reaction have been determined relative to the reference values of the {sup 197}Au(n,γ){sup 198}Au reaction, with σ{sub o,Au} = 98.65 ± 0.09 barn and I{sub o,Au} = 1550 ± 28 barn. The present thermal neutron cross section has been determined to be σ{sub o,Sc} = 27.5 ± 0.8 barn. According to the definition of cadmium cut-off energy at 0.55 eV, the present resonance integral cross section has been determined to be I{sub o,Sc} = 12.4 ± 0.7 barn. The present results are compared with literature values and discussed.

  3. Measurement of thermal neutron cross-sections and resonance integrals for sup 7 sup 1 Ga(n,gamma) sup 7 sup 2 Ga and sup 7 sup 5 As(n,gamma) sup 7 sup 6 As by using sup 2 sup 4 sup 1 Am-Be isotopic neutron source

    CERN Document Server

    Karadag, M; Tan, M; Oezmen, A

    2003-01-01

    Thermal neutron cross-sections and resonance integrals for the sup 7 sup 1 Ga(n,gamma) sup 7 sup 2 Ga and sup 7 sup 5 As(n,gamma) sup 7 sup 6 As reactions were measured by the activation method. The experimental samples with and without a cylindrical Cd shield case in 1 mm wall thickness were irradiated in an isotropic neutron field of the sup 2 sup 4 sup 1 Am-Be neutron source. The induced activities in the samples were measured by high-resolution gamma-ray spectrometry with a calibrated reverse-electrode germanium detector. Thermal neutron cross-sections for 2200 m/s neutrons and resonance integrals for the sup 7 sup 1 Ga(n,gamma) sup 7 sup 2 Ga and sup 7 sup 5 As(n,gamma) sup 7 sup 6 As reactions have been obtained relative to the reference values, sigma sub 0 =13.3+-0.1 b and I sub 0 =14.0+-0.3 b for the sup 5 sup 5 Mn(n,gamma) sup 5 sup 6 Mn reaction as a single comparator. The necessary correction factors for gamma attenuation, thermal neutron and resonance neutron self-shielding effects were taken into...

  4. Shielding design study for the JAERI/KEK spallation neutron source

    International Nuclear Information System (INIS)

    Maekawa, Fujio; Teshigawara, Makoto; Konno, Chikara; Ikeda, Yujiro; Watanabe, Noboru

    2001-01-01

    Shielding design for the JAERI/KEK spallation neutron source was studied. Bulk shielding characteristics and optimization of a beam shutter were investigated by using Monte Carlo calculation code NMTC/JAM and MCNP with LA-150 neutron cross-section library. The following remarks were derived. (1) Neutron dose outside of the concrete shield at 6.6 m from the center is ∼10 μSv/hr regardless of angles with respect to the proton beam axis. The neutron dose can be reduced more than a factor of 30 by adding natural boron of 5 wt% in the concrete. (2) When a beam shutter position just outside the void vessel and the shutter length of 2 m are assumed, a shutter made of copper (1.7 m) with polyethylene (0.3 m) is the optimum in terms of shielding performance as well as cost merit. A shutter made of tungsten is not so effective. (3) Further studies are needed for optimization of beam shutter position. (author)

  5. 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

  6. Initial global 2-D shielding analysis for the Advanced Neutron Source core and reflector

    Energy Technology Data Exchange (ETDEWEB)

    Bucholz, J.A.

    1995-08-01

    This document describes the initial global 2-D shielding analyses for the Advanced Neutron Source (ANS) reactor, the D{sub 2}O reflector, the reflector vessel, and the first 200 mm of light water beyond the reflector vessel. Flux files generated here will later serve as source terms in subsequent shielding analyses. In addition to reporting fluxes and other data at key points of interest, a major objective of this report was to document how these analyses were performed, the phenomena that were included, and checks that were made to verify that these phenomena were properly modeled. In these shielding analyses, the fixed neutron source distribution in the core was based on the `lifetime-averaged` spatial power distribution. Secondary gamma production cross sections in the fuel were modified so as to account intrinsically for delayed fission gammas in the fuel as well as prompt fission gammas. In and near the fuel, this increased the low-energy gamma fluxes by 50 to 250%, but out near the reflector vessel, these same fluxes changed by only a few percent. Sensitivity studies with respect to mesh size were performed, and a new 2-D mesh distribution developed after some problems were discovered with respect to the use of numerous elongated mesh cells in the reflector. All of the shielding analyses were performed sing the ANSL-V 39n/44g coupled library with 25 thermal neutron groups in order to obtain a rigorous representation of the thermal neutron spectrum throughout the reflector. Because of upscatter in the heavy water, convergence was very slow. Ultimately, the fission cross section in the various materials had to be artificially modified in order to solve this fixed source problem as an eigenvalue problem and invoke the Vondy error-mode extrapolation technique which greatly accelerated convergence in the large 2-D RZ DORT analyses. While this was quite effective, 150 outer iterations (over energy) were still required.

  7. Initial global 2-D shielding analysis for the Advanced Neutron Source core and reflector

    International Nuclear Information System (INIS)

    Bucholz, J.A.

    1995-08-01

    This document describes the initial global 2-D shielding analyses for the Advanced Neutron Source (ANS) reactor, the D 2 O reflector, the reflector vessel, and the first 200 mm of light water beyond the reflector vessel. Flux files generated here will later serve as source terms in subsequent shielding analyses. In addition to reporting fluxes and other data at key points of interest, a major objective of this report was to document how these analyses were performed, the phenomena that were included, and checks that were made to verify that these phenomena were properly modeled. In these shielding analyses, the fixed neutron source distribution in the core was based on the 'lifetime-averaged' spatial power distribution. Secondary gamma production cross sections in the fuel were modified so as to account intrinsically for delayed fission gammas in the fuel as well as prompt fission gammas. In and near the fuel, this increased the low-energy gamma fluxes by 50 to 250%, but out near the reflector vessel, these same fluxes changed by only a few percent. Sensitivity studies with respect to mesh size were performed, and a new 2-D mesh distribution developed after some problems were discovered with respect to the use of numerous elongated mesh cells in the reflector. All of the shielding analyses were performed sing the ANSL-V 39n/44g coupled library with 25 thermal neutron groups in order to obtain a rigorous representation of the thermal neutron spectrum throughout the reflector. Because of upscatter in the heavy water, convergence was very slow. Ultimately, the fission cross section in the various materials had to be artificially modified in order to solve this fixed source problem as an eigenvalue problem and invoke the Vondy error-mode extrapolation technique which greatly accelerated convergence in the large 2-D RZ DORT analyses. While this was quite effective, 150 outer iterations (over energy) were still required

  8. Neutronic reactor thermal shield

    International Nuclear Information System (INIS)

    Lowe, P.E.

    1976-01-01

    A shield for a nuclear reactor includes at least two layers of alternating wide and narrow rectangular blocks so arranged that the spaces between blocks in adjacent layers are out of registry, each block having an opening therein equally spaced from the sides of the blocks and nearer the top of the block than the bottom, the distance from the top of the block to the opening in one layer being different from this distance in adjacent layers, openings in blocks in adjacent layers being in registry. 1 claim, 7 drawing figures

  9. Laboratory tests on neutron shields for gamma-ray detectors in space

    CERN Document Server

    Hong, J; Hailey, C J

    2000-01-01

    Shields capable of suppressing neutron-induced background in new classes of gamma-ray detectors such as CdZnTe are becoming important for a variety of reasons. These include a high cross section for neutron interactions in new classes of detector materials as well as the inefficient vetoing of neutron-induced background in conventional active shields. We have previously demonstrated through Monte-Carlo simulations how our new approach, supershields, is superior to the monolithic, bi-atomic neutron shields which have been developed in the past. We report here on the first prototype models for supershields based on boron and hydrogen. We verify the performance of these supershields through laboratory experiments. These experimental results, as well as measurements of conventional monolithic neutron shields, are shown to be consistent with Monte-Carlo simulations. We discuss the implications of this experiment for designs of supershields in general and their application to future hard X-ray/gamma-ray experiments...

  10. New shielding material development for compact accelerator-driven neutron source

    Directory of Open Access Journals (Sweden)

    Guang Hu

    2017-04-01

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

  11. Shielding of a neutron irradiator with {sup 241}Am-Be source

    Energy Technology Data Exchange (ETDEWEB)

    Oliveira, K.A.M. de; Crispim, V.R.; Silva, A.X., E-mail: koliveira@con.ufrj.b, E-mail: verginia@con.ufrj.b, E-mail: ademir@con.ufrj.b [Universidade Federal do Rio de Janeiro (PEN/COPPE/UFRJ), RJ (Brazil). Coordenacao dos Programas de Pos-Graduacao de Engenharia. Programa de Engenharia Nuclear; Fonseca, E.S., E-mail: evaldo@ird.gov.b [Instituto de Radioprotecao e Dosimetria (IRD/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

    2011-07-01

    The equivalent dose rates at 1.0 cm from the outer surface of the shielding of a neutron irradiation system that uses {sup 241}Am-Be source with activity of 185 GBq (5 Ci) were determined. A theoretical-experimental approach including case studies, through computer simulations with MCNP code was employed to calculate the best shielding thickness. Following the construction of the neutron irradiator, dose measurements were conducted in order to validate data obtained from simulation. The neutron irradiator shielding was designed in such a way to allow transport of the neutron radiography system for in loco inspections ensuring workers' radiologic safety. (author)

  12. Gamma ray and neutron shielding properties of some concrete materials

    International Nuclear Information System (INIS)

    Yilmaz, E.; Baltas, H.; Kiris, E.; Ustabas, I.; Cevik, U.; El-Khayatt, A.M.

    2011-01-01

    Highlights: → This study sheds light on the shielding properties of gamma-rays and neutrons for some concrete samples. → The experimental mass attenuation coefficients values were compared with theoretical values obtained using WinXCom. → Moreover, neutron shielding has been treated in terms of macroscopic removal cross-section (Σ R , cm -1 ) concept. → The NXcom program was employed to calculate the attenuation coefficients values of neutrons. → These values showed a change with energy and composition of the concrete samples. - Abstract: Shielding of gamma-rays and neutrons by 12 concrete samples with and without mineral additives has been studied. The total mass attenuation and linear attenuation coefficients, half-value thicknesses, effective atomic numbers, effective electron densities and atomic cross-sections at photons energies of 59.5 and 661 keV have been measured and calculated. The measured and calculated values were compared and a reasonable agreement has been observed. Also the recorded values showed a change with energy and composition of the concrete samples. In addition, neutron shielding has been treated in terms of macroscopic removal cross-section (Σ R , cm -1 ) concept. The WinXCom and NXcom programs were employed to calculate the attenuation coefficients of gamma-rays and neutrons, respectively.

  13. Evaluating secondary neutron doses of a refined shielded design for a medical cyclotron using the TLD approach

    International Nuclear Information System (INIS)

    Lin, Jye-Bin; Tseng, Hsien-Chun; Liu, Wen-Shan; Lin, Ding-Bang; Hsieh, Teng-San; Chen, Chien-Yi

    2013-01-01

    An increasing number of cyclotrons at medical centers in Taiwan have been installed to generate radiopharmaceutical products. An operating cyclotron generates immense amounts of secondary neutrons from reactions such the 18 O(p, n) 18 F, used in the production of FDG. This intense radiation can be hazardous to public health, particularly to medical personnel. To increase the yield of 18 F-FDG from 4200 GBq in 2005 to 48,600 GBq in 2011, Chung Shan Medical University Hospital (CSMUH) has prolonged irradiation time without changing the target or target current to meet requirements regarding the production 18 F. The CSMUH has redesigned the CTI Radioisotope Delivery System shield. The lack of data for a possible secondary neutron doses has increased due to newly designed cyclotron rooms. This work aims to evaluate secondary neutron doses at a CTI cyclotron center using a thermoluminescent dosimeter (TLD-600). Two-dimensional neutron doses were mapped and indicated that neutron doses were high as neutrons leaked through self-shielded blocks and through the L-shaped concrete shield in vault rooms. These neutron doses varied markedly among locations close to the H 2 18 O target. The Monte Carlo simulation and minimum detectable dose are also discussed and demonstrated the reliability of using the TLD-600 approach. Findings can be adopted by medical centers to identify radioactive hot spots and develop radiation protection. - Highlights: • Neutron doses were verified using TLD approach. • Neutron doses were increased at cyclotron centers. • Revised L-shaped shield suppresses effectively the neutrons. • Neutron dose can be attenuated to 1.13×10 6 %

  14. Inhomogeneity of neutron and gamma-ray attenuation in biological shields

    Energy Technology Data Exchange (ETDEWEB)

    El-bakkoush, F A; El-Ghobary, A M; Megahid, R M [Reactor and Neutron physics Department, Nuclear Research Center, A.E.A., Cairo (Egypt)

    1997-12-31

    Measurements have been carried-out to investigate the attenuation properties of some materials which are used as biological shields around nuclear radiation sources. Investigation was performed by measuring the transmitted fast neutron and gamma-spectra through cylindrical samples of magnetite- limonite, steel and cellulose shields. The neutron and gamma spectra were measured by a neutron-gamma spectrometer with stilbene scintillator. Discrimination between neutron and gamma pulses was achieved by a discrimination method. The obtained results are displayed in the form of neutron and gamma spectra and attenuation relations which are used to derive the total macroscopic cross-sections for neutrons and total linear attenuation coefficients for gamma-rays. The values of neutron and gamma relaxation lengths are also derived for the investigated materials. 10 figs., 1 tabs.

  15. Self Shielding in Nuclear Fissile Assay Using LSDS

    International Nuclear Information System (INIS)

    Lee, Yong Deok; Park, Chang Je; Park, Geun Il; Song, Kee Chan

    2012-01-01

    The new technology for isotopic fissile material contents assay is under development at KAERI using lead slowing down spectrometer(LSDS). LSDS is very sensitive to distinguish fission signals from each fissile isotope in spent and recycled fuel. The accumulation of spent fuel is current big issue. The amount of spent fuels will reach the maximum storage capacity of the pools soon. Therefore, an interim storage must be searched and it should be optimized in design by applying accurate fissile content. When the storage has taken effect, all the nuclear materials must be also specified and verified for safety, economics and management. Generally, the spent fuel from PWR has unburned ∼1 % U235, produced ∼0.5 % plutonium from decay chain, ∼3 % fission products, ∼ 0.1 % minor actinides (MA) and uranium remainder. About 1.5 % fissile materials still exist in the spent fuel. Therefore, for reutilization of fissile materials in spent fuel at SFR, resource material is produced through pyro process. Fissile material contents in resource material must be analyzed before fabricating SFR fuel for reactor safety and economics. In assay of fissile content of spent fuel and recycled fuel, intense radiation background gives limitation on the direct analysis of fissile materials. However, LSDS is not influenced by such a radiation background in fissile assay. Based on the decided geometry setup, self shielding parameter was calculated at the fuel assay zone by introducing spent fuel or pyro produced nuclear material. When nuclear material is inserted into the assay area, the spent fuel assembly or pyro recycled fuel material perturbs the spatial distribution of the slowing down neutrons in lead and the prompt fast fission neutrons produced by fissile materials are also perturbed. The self shielding factor is interpreted as that how much of absorption is created inside the fuel area when it is in the lead. Self shielding effect provides a non-linear property in the isotopic

  16. Radiological Shielding Design for the Neutron High-Resolution Backscattering Spectrometer EMU at the OPAL Reactor

    Directory of Open Access Journals (Sweden)

    Ersez Tunay

    2017-01-01

    Full Text Available The shielding for the neutron high-resolution backscattering spectrometer (EMU located at the OPAL reactor (ANSTO was designed using the Monte Carlo code MCNP 5-1.60. The proposed shielding design has produced compact shielding assemblies, such as the neutron pre-monochromator bunker with sliding cylindrical block shields to accommodate a range of neutron take-off angles, and in the experimental area - shielding of neutron focusing guides, choppers, flight tube, backscattering monochromator, and additional shielding elements inside the Scattering Tank. These shielding assemblies meet safety and engineering requirements and cost constraints. The neutron dose rates around the EMU instrument were reduced to < 0.5 µSv/h and the gamma dose rates to a safe working level of ≤ 3 µSv/h.

  17. Radiological Shielding Design for the Neutron High-Resolution Backscattering Spectrometer EMU at the OPAL Reactor

    Science.gov (United States)

    Ersez, Tunay; Esposto, Fernando; Souza, Nicolas R. de

    2017-09-01

    The shielding for the neutron high-resolution backscattering spectrometer (EMU) located at the OPAL reactor (ANSTO) was designed using the Monte Carlo code MCNP 5-1.60. The proposed shielding design has produced compact shielding assemblies, such as the neutron pre-monochromator bunker with sliding cylindrical block shields to accommodate a range of neutron take-off angles, and in the experimental area - shielding of neutron focusing guides, choppers, flight tube, backscattering monochromator, and additional shielding elements inside the Scattering Tank. These shielding assemblies meet safety and engineering requirements and cost constraints. The neutron dose rates around the EMU instrument were reduced to < 0.5 µSv/h and the gamma dose rates to a safe working level of ≤ 3 µSv/h.

  18. The shielding performance of multilayer composite shielding structures to 14.8 MeV fast neutrons

    International Nuclear Information System (INIS)

    Shen Zhiqiang; Kang Qing; Xu Jun; Wang Zhenggang; Lu Nan

    2014-01-01

    Cement-based round thin-layer samples mixed with 30% quality content of barite, and 20% quality content of carbide boron has Prepared, the same-diameter sliced samples of pure graphite and pure polyethylene has cut, then, samples combination and cross stack order has designed, formed four species Multilayer Composite shield structure, at last, neutron attenuation measurements has been done by experimental system of using 14.8 MeV neutrons from the 5SDH-2 accelerator and long counter composition, penetrating rate of samples and the shield structure to 14.8 MeV fast neutron has tested, and attenuation section has calculated. Results show that 14.8 MeV fast neutrons to higher penetration rates of thin layer samples, attenuation cross section of samples distinguish small between each other, must be increasing the thickness of the samples to reduce the experimental uncertainty; through composed of attenuation cross section and thickness parameters of composite structure, can more accurately predict the shielding ability of composite structures, error between calculation results and experimental results in 4%. (authors)

  19. Neutron shielding and activation of the MASTU device and surrounds

    Energy Technology Data Exchange (ETDEWEB)

    Taylor, David, E-mail: david.taylor@ccfe.ac.uk [EURATOM/CCFE Fusion Association, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Lilley, Steven; Turner, Andrew [EURATOM/CCFE Fusion Association, Culham Science Centre, Abingdon OX14 3DB (United Kingdom); Davis, Andrew [Now at College of Engineering, University of Wisconsin, Madison, WI 53706 (United States)

    2014-10-15

    Highlights: We model neutron shielding for the planned MASTU device; nadequacies in the existing shielding design are remedied; Levels of public exposure are considered; We model activated gamma emission for the device under a worst case scenario. Abstract: A significant functional upgrade is planned for the Mega Ampere Spherical Tokamak (MAST) device, located at Culham in the UK, including the implementation of a notably greater neutral beam injection power. This upgrade will cause the emission of a substantially increased intensity of neutron radiation for a substantially increased amount of time upon operation of the device. Existing shielding and activation precautions are shown to prove insufficient in some regards, and recommendations for improvements are made, including the following areas: shielding doors to MAST shielded facility enclosure (known as “the blockhouse”); north access tunnel; blockhouse roof; west cabling duct. In addition, some specific neutronic dose rate questions are addressed and answered; those discussed here relate to shielding penetrations and dose rate reflected from the air above the device (“skyshine”). It is shown that the alterations to shielding and area access reduce the dose rate in unrestricted areas from greater than 100 μSv/h to less than 2 μSv/h averaged over the working day. The tools used for this analysis are the MCNP (Monte Carlo N-particle) code, used to calculate the three-dimensional spatial distribution of neutron and photon dose rates in and around the device and its shields, and the nuclear inventory code FISPACT, run under the umbrella code MCR2S, used to calculate the time-dependent shutdown dose rate in the region of the device at several decay times.

  20. Attenuation of neutrons and gamma-rays in homogeneous and multilayered shields

    International Nuclear Information System (INIS)

    Abdo, A.E.; Megahid, R.M.

    1997-01-01

    Measurements were carried-out to compare the attenuation properties of homogeneous shields and shields of two layers and three layers for fast neutrons and total gamma-rays. These were performed by measuring the fast neutron and total gamma-ray spectra behind homogeneous shields of magnetite-limonite, ilmenite-ilmenite and magnetite-magnetite concretes. The two layers assembly consists of iron and one of the above mentioned concretes, while the three layers shield consists of water, iron and one of the previously mentioned concretes. All measurements were carried-out using a neutron-gamma spectrometer with stilbene scintillator coupled to a fast photo multi player tube. Separation between pulses of recoil protons and recoil electrons was achieved by a pulse shape discrimination technique. 3 tabs., 10 figs., 13 refs

  1. Neutron streaming studies along JET shielding penetrations

    Science.gov (United States)

    Stamatelatos, Ion E.; Vasilopoulou, Theodora; Batistoni, Paola; Obryk, Barbara; Popovichev, Sergey; Naish, Jonathan

    2017-09-01

    Neutronic benchmark experiments are carried out at JET aiming to assess the neutronic codes and data used in ITER analysis. Among other activities, experiments are performed in order to validate neutron streaming simulations along long penetrations in the JET shielding configuration. In this work, neutron streaming calculations along the JET personnel entrance maze are presented. Simulations were performed using the MCNP code for Deuterium-Deuterium and Deuterium- Tritium plasma sources. The results of the simulations were compared against experimental data obtained using thermoluminescence detectors and activation foils.

  2. The shielding of a 14 MeV neutron generator

    International Nuclear Information System (INIS)

    Brighton, D.R.

    1976-10-01

    The concrete masonry shield for a 14 MeV neutron generator was designed using data supplied by the manufacturer. Subsequent radiation surveys outside the shield showed doses higher than expected. Calculations indicated the sensitivity of dose transmission factors to concrete composition. The observed dose transmission factor agreed with that of Broerse but not with that of Hacke and Prudhomme. Measurements and calculations delineated the contribution that neutrons, scattered from the upper wall that supports the laboratory roof, made to the dose in adjoining areas. In redesigning the shield a compromise was made between additional cost and restrictions on the generator's duty cycle, which is automatically controlled to ensure personnel safety. (Author)

  3. Neutron shielding properties of a borated high-density glass

    Directory of Open Access Journals (Sweden)

    Saeed Aly Abdallah

    2017-01-01

    Full Text Available The neutron shielding properties of a borated high density glass system was characterized experimentally. The total removal macroscopic cross-section of fast neutrons, slow neutrons as well as the linear attenuation coefficient of total gamma rays, primary in addition to secondary, were measured experimentally under good geometric condition to characterize the attenuation properties of (75-x B2O3-1Li2O-5MgO-5ZnO-14Na2O-xBaO glassy system. Slabs of different thicknesses from the investigated glass system were exposed to a collimated beam of neutrons emitted from 252Cf and 241Am-Be neutron sources in order to measure the attenuation properties of fast and slow neutrons as well as total gamma rays. Results confirmed that barium borate glass was suitable for practical use in the field of radiation shielding.

  4. Evaluation of some resonance self-shielding procedures employed in high conversion light water reactor design

    International Nuclear Information System (INIS)

    Patino, N.E.; Abbate, M.J.; Sbaffoni, M.M.

    1990-01-01

    The procedures employed in the treatment of the resonance shielding effect have been identified as one of the causes of the large discrepancies found in the neutronic calculation of high conversion light water reactors (HCLWRs), indicating the need for a revision of the self-shielding procedures employed. In this work some well known techniques applied in HCLWR self-shielding calculations are evaluated; the study involves the comparison of methods for the generation of group constants, the analysis of the impact of considering some isotopes as infinitely diluted and the evaluation of the usual approximations utilized for the treatment of heterogeneities

  5. Neutron beam-line shield design for the protein crystallography instrument at the Lujan Center

    International Nuclear Information System (INIS)

    Russell, G.J.; Pitcher, E.J.; Muhrer, G.; Ferguson, P.D.

    2001-01-01

    We have developed a very useful methodology for calculating absolute total (neutron plus gamma-ray) dose equivalent rates for use in the design of neutron beam line shields at a spallation neutron source. We have applied this technique to the design of beam line shields for several new materials science instruments being built at the Manuel Lujan Jr. Neutron Scattering Center. These instruments have a variety of collimation systems and different beam line shielding issues. We show here some specific beam line shield designs for the Protein Crystallography Instrument. (author)

  6. Shielding calculations for the Intense Neutron Source Facility. Final report

    International Nuclear Information System (INIS)

    Battat, M.E.; Henninger, R.J.; Macdonald, J.L.; Dudziak, D.J.

    1978-06-01

    Results of shielding calculations for the Intnse Neutron Source (INS) facility are presented. The INS facility is designed to house two sources, each of which will produce D--T neutrons with intensities in the range from 1 to 3 x 10 15 n/s on a continuous basis. Topics covered include the design of the biological shield, use of two-dimensional discrete-ordinates results to specify the source terms for a Monte Carlo skyshine calculation, air activation, and dose rates in the source cell (after shutdown) due to activation of the biological shield

  7. Evaluation of neutron shielding made of cement type material

    International Nuclear Information System (INIS)

    Seshimo, Takuya; Nagai, Takayuki; Onose, Atsushi; Takuma, Yasuhisa; Tanuma, Hiroyuki; Otagawa, Masaaki

    1998-01-01

    We prepared boron-containing cement and evaluated the characteristics of this new cement. This is the material of neutron shielding which is lighter than existing one. The quality we aimed is: H ≥ 0.025 g/cm 3 , B ≥ 0.065 g/cm 3 , density ≤ 1.70 g/cm 3 . We made test pieces changing water powder ratio (W/P), adding amount of air entraining agent, adding amount of water reducing agent, and time of vibration, and then, evaluated the characteristics. The measured parameters are the air content, mortar flow and homogeneity for cement mortar, homogeneity and compressive strength for hardened one. From the results of these tests, we confirmed the possibility of making neutron shielding that can satisfy the aimed quality using this boron-containing cement. After all, we established the method of making the neutron shielding, and this method was used in the construction of RETF. (author)

  8. Neutron shielding behavior of thermoplastic natural rubber/boron carbide composites

    Science.gov (United States)

    Mat Zali, Nurazila; Yazid, Hafizal; Megat Ahmad, Megat Harun Al Rashid

    2018-01-01

    Many shielding materials have been designed against the harm of different types of radiation to the human body. Today, polymer-based lightweight composites have been chosen by the radiation protection industry. In the present study, thermoplastic natural rubber (TPNR) composites with different weight percent of boron carbide (B4C) fillers (0% to 30%) were fabricated as neutron shielding through melt blending method. Neutron attenuation properties of TPNR/B4C composites have been investigated. The macroscopic cross section (Σ), half value layer (HVL) and mean free path length (λ) of the composites have been calculated and the transmission curves have been plotted. The obtained results show that Σ, HVL and λ greatly depend on the B4C content. Addition of B4C fillers into TPNR matrix were found to enhance the macroscopic cross section values thus decrease the mean free path length (λ) and half value layer (HVL) of the composites. The transmission curves exhibited that the neutron transmission of the composites decreased with increasing shielding thickness. These results showed that TPNR/B4C composites have high potential for neutron shielding applications.

  9. Characterization of the internal background for thermal and fast neutron detection with CLLB

    Energy Technology Data Exchange (ETDEWEB)

    Woolf, Richard S., E-mail: richard.woolf@nrl.navy.mil; Phlips, Bernard F.; Wulf, Eric A.

    2016-12-01

    We report on a set of experiments conducted to determine what effects, if any, the internal background in the CLLB scintillation detector has on the thermal neutron detection performance. We conducted source measurements using an unmoderated and moderated {sup 252}Cf neutron/γ-ray source and long (48-h), unshielded and shielded, background measurements to characterize the internal background with and without a source present. These measurements allowed us to determine the 2-d event selections needed to isolate the thermal neutron peak observed in pulse shape vs. energy space and apply those selections to our background measurements. Our results indicate that the thermal neutron detection capabilities of the CLLB are marginally affected by the presence of internal background. An unmoderated 113-µCi {sup 252}Cf source at 15 cm from the detector yields a thermal neutron rate of 8×10{sup −2}/s cm{sup 3}, while moderating the source with 5 cm of polyethylene yields a thermal neutron rate of 5.5×10{sup −1}/s cm{sup 3}. The measured background rate for events that fall within the selected thermal neutron region is 1.2×10{sup −3}/s cm{sup 3}. Lastly, the potential for CLLB for detecting fast neutrons was investigated.

  10. A Physical Model of Pulsars as Gravitational Shielding and Oscillating Neutron Stars

    Directory of Open Access Journals (Sweden)

    Zhang T. X.

    2015-04-01

    Full Text Available Pulsars are thought to be fast rotating neutron stars, synchronously emitting periodic Dirac-delta-shape radio-frequency pulses and Lorentzian-shape oscillating X-rays. The acceleration of charged particles along the magnetic field lines of neutron stars above the magnetic poles that deviate from the rotating axis initiates coherent beams of ra- dio emissions, which are viewed as pulses of radiation whenever the magnetic poles sweep the viewers. However, the conventional lighthouse model of pulsars is only con- ceptual. The mechanism through which particles are accelerated to produce coherent beams is still not fully understood. The process for periodically oscillating X-rays to emit from hot spots at the inner edge of accretion disks remains a mystery. In addition, a lack of reflecting X-rays of the pulsar by the Crab Nebula in the OFF phase does not support the lighthouse model as expected. In this study, we develop a physical model of pulsars to quantitatively interpret the emission characteristics of pulsars, in accor- dance with the author’s well-developed five-dimensional fully covariant Kaluza-Klein gravitational shielding theory and the physics of thermal and accelerating charged par- ticle radiation. The results obtained from this study indicate that, with the significant gravitational shielding by scalar field, a neutron star nonlinearly oscillates and produces synchronous periodically Dirac-delta-shape radio-frequency pulses (emitted by the os- cillating or accelerating charged particles as well as periodically Lorentzian-shape os- cillating X-rays (as the thermal radiation of neutron stars whose temperature varies due to the oscillation. This physical model of pulsars broadens our understanding of neu- tron stars and develops an innovative mechanism to model the emissions of pulsars.

  11. Validation of SCALE code package on high performance neutron shields

    International Nuclear Information System (INIS)

    Bace, M.; Jecmenica, R.; Smuc, T.

    1999-01-01

    The shielding ability and other properties of new high performance neutron shielding materials from the KRAFTON series have been recently published. A comparison of the published experimental and MCNP results for the two materials of the KRAFTON series, with our own calculations has been done. Two control modules of the SCALE-4.4 code system have been used, one of them based on one dimensional radiation transport analysis (SAS1) and other based on the three dimensional Monte Carlo method (SAS3). The comparison of the calculated neutron dose equivalent rates shows a good agreement between experimental and calculated results for the KRAFTON-N2 material.. Our results indicate that the N2-M-N2 sandwich type is approximately 10% inferior as neutron shield to the KRAFTON-N2 material. All values of neutron dose equivalent obtained by SAS1 are approximately 25% lower in comparison with the SAS3 results, which indicates proportions of discrepancies introduced by one-dimensional geometry approximation.(author)

  12. AUTOSECOL: an automatic calculation of the self-shielding of heavy isotope resonances

    International Nuclear Information System (INIS)

    Grandotto-Biettoli, Marc.

    The formalism is based on separating both types of resonance effects: local energy effects creating a fine structure in the flux, and bulk effects resulting in a slow variation in the flux. Effective reaction rates are defined that, used as tables in a multigroup calculation of cells with a large pitch in regard to resonance widths, allow an exact account of the dependence of the effective integral upon fast variations in the flux. These tables are used to introduce this phenomenon of resonance self-shielding in the multigroup Apollo program for solving the neutron transport equation, they are derived from nuclear data with using some parameters relating to the physical state of the resonant isotope inside the fuel medium. The AUTOSECOL system provides a library of effective reaction rates for taking account of the resonance self-shielding effect on the neutron flux in nuclear reactor cells. Its versatility in regard to the methods previously used for solving the same problem allows a rapid testing of the consequences of considering the self-shielding effect of new isotope resonances, a following up of the evolution in nuclear data evaluation, and rapidly studying the interest lying in new data. Results obtained with AUTOSECOL are compared with those obtained when using the SECOL code for computing the effective reaction rates of 235 U, 239 Pu, 107 Ag, 109 Ag, and 241 Pu [fr

  13. Thermal neutron filter design for the neutron radiography facility at the LVR-15 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Soltes, Jaroslav [Research Centre Rez Ltd., Husinec - Rez 130, 250 68 Rez, (Czech Republic); Faculty of Nuclear Sciences and Physical Engineering, CTU in Prague, (Czech Republic); Viererbl, Ladislav; Lahodova, Zdena; Koleska, Michal; Vins, Miroslav [Research Centre Rez Ltd., Husinec - Rez 130, 250 68 Rez, (Czech Republic)

    2015-07-01

    In 2011 a decision was made to build a neutron radiography facility at one of the unused horizontal channels of the LVR-15 research reactor in Rez, Czech Republic. One of the key conditions for operating an effective radiography facility is the delivery of a high intensity, homogeneous and collimated thermal neutron beam at the sample location. Additionally the intensity of fast neutrons has to be kept as low as possible as the fast neutrons may damage the detectors used for neutron imaging. As the spectrum in the empty horizontal channel roughly copies the spectrum in the reactor core, which has a high ratio of fast neutrons, neutron filter components have to be installed inside the channel in order to achieve desired beam parameters. As the channel design does not allow the instalment of complex filters and collimators, an optimal solution represent neutron filters made of large single-crystal ingots of proper material composition. Single-crystal silicon was chosen as a favorable filter material for its wide availability in sufficient dimensions. Besides its ability to reasonably lower the ratio of fast neutrons while still keeping high intensities of thermal neutrons, due to its large dimensions, it suits as a shielding against gamma radiation from the reactor core. For designing the necessary filter dimensions the Monte-Carlo MCNP transport code was used. As the code does not provide neutron cross-section libraries for thermal neutron transport through single-crystalline silicon, these had to be created by approximating the theory of thermal neutron scattering and modifying the original cross-section data which are provided with the code. Carrying out a series of calculations the filter thickness of 1 m proved good for gaining a beam with desired parameters and a low gamma background. After mounting the filter inside the channel several measurements of the neutron field were realized at the beam exit. The results have justified the expected calculated values

  14. Nuclear characteristics of epoxy resin as a space environment neutron shielding

    Energy Technology Data Exchange (ETDEWEB)

    Adeli, Ruhollah [Nuclear Science and Technology Research Institute, Yazd (Iran, Islamic Republic of). Central Iran Research Complex; Shirmardi, Seyed Pezhman [Nuclear Science and Technology Research Institute, Tehran (Iran, Islamic Republic of). Radiation Application Research School; Mazinani, Saideh [Amirkabir Nanotechnology Research Institute, Tehran (Iran, Islamic Republic of); Ahmadi, Seyed Javad [Nuclear Science and Technology Research Institute, Tehran (Iran, Islamic Republic of). Nuclear Fuel Cycle Research School

    2017-03-15

    In recent years many investigations have been done for choosing applicable light neutron shielding in space environmental applications. In this study, we have considered the neutron radiation-protective characteristics of neat epoxy resin, a thermoplastic polymer material and have compared it with various candidate materials in neutron radiation protection such as Al 6061 alloy and Polyethylene. The aim of this investigation is the effect of type of moderator for fast neutron, notwithstanding neutron absorbers fillers. The nuclear interactions and the effective dose at shields have been studied with the Monte Carlo N-Particle transport code (MCNP), using variance reductions to reduce the relative error. Among the candidates, polymer matrix showed a better performance in attenuating fast neutrons and caused a lower neutron and secondary photon effective dose.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-11-01

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

  16. Benchmarking shielding simulations for an accelerator-driven spallation neutron source

    Directory of Open Access Journals (Sweden)

    Nataliia Cherkashyna

    2015-08-01

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

  17. Thermal neutron moderating device

    International Nuclear Information System (INIS)

    Takigami, Hiroyuki.

    1995-01-01

    In a thermal neutron moderating device, superconductive coils for generating magnetic fields capable of applying magnetic fields vertical to the longitudinal direction of a thermal neutron passing tube, and superconductive coils for magnetic field gradient for causing magnetic field gradient in the longitudinal direction of the thermal neutron passing tube are disposed being stacked at the outside of the thermal neutron passing tube. When magnetic field gradient is present vertically to the direction of a magnetic moment, thermal neutrons undergo forces in the direction of the magnetic field gradient in proportion to the magnetic moment. Then, the magnetic moment of the thermal neutrons is aligned with the direction vertical to the passing direction of the thermal neutrons, to cause the magnetic field gradient in the passing direction of the thermal neutrons. The speed of the thermal neutrons can be optionally selected and the wavelength can freely be changed by applying forces to the thermal neutrons and changing the extent and direction of the magnetic field gradient. Superconductive coils are used as the coils for generating magnetic fields and the magnetic field gradient in order to change extremely high energy of the thermal neutrons. (N.H.)

  18. A theoretical study of the fast-neutron attenuation in Ghanaian serpentine shields

    International Nuclear Information System (INIS)

    Akaho, E.H.K.; Anim-Sampong, S.

    1994-01-01

    Theoretical calculations were done to determine the suitability of local serpentine rocks for shielding fast neutrons. A coupled neutron-gamma library of 25 energy groups, IRAN3.LIB developed for ANISN/PC was used to generate nuclear data for the tested shields. Calculations were carried out assuming a P 3 scattering order for spherical geometry with S 6 angular quadrature. From the trends of attenuation and computer factors such as relaxation length and transmission there is the indication that the shielding properties of the local shields are better than the foreign serpentine shields used in this study. They are slightly inferior to ordinary concrete employed in shielding power reactors. (author). 9 refs.; 5 tabs.; 5 figs

  19. Neutronics and shielding issues of ADS

    International Nuclear Information System (INIS)

    Abderrahim, H. A.; Aoust, T.; Haeck, W.; Malambu, E.; Van den Eynde, G.; Gonzalez, E.; Vicente, C.; Martinez-Val, J. M.; Romanets, Y.; Vaz, P.

    2007-01-01

    implementation and deployment have in common the fact that they raise cutting edge scientific and technological problems, associated to the operation of the high-intensity proton accelerator, the high-power (in the multi-MegaWatt range) delivered to the target and the material damage in the target and surrounding structures. The thermal power in the core, the thermal-hydraulic aspects associated to the heat removal in steady state and also in transient mode, the subcriticality level of the system and the efficiency of the transmutation process, is particularly sensitive to the core design (geometry, number of subassemblies, fuel composition, among many other aspects). Neutronic and shielding issues and the computation and mapping of neutron fluxes and doses are important throughout all stages of design of these systems. In this paper, i) the main characteristics and parameters of the ADS systems previously alluded to will be reviewed ii) the neutronics and shielding calculations of relevance for the design of the ADS systems, for radiation damage and for radiation protection purposes will be extensively described

  20. Evaluation of Shielding Performance for Newly Developed Composite Materials

    Science.gov (United States)

    Evans, Beren Richard

    This work details an investigation into the contributing factors behind the success of newly developed composite neutron shield materials. Monte Carlo simulation methods were utilized to assess the neutron shielding capabilities and secondary radiation production characteristics of aluminum boron carbide, tungsten boron carbide, bismuth borosilicate glass, and Metathene within various neutron energy spectra. Shielding performance and secondary radiation data suggested that tungsten boron carbide was the most effective composite material. An analysis of the macroscopic cross-section contributions from constituent materials and interaction mechanisms was then performed in an attempt to determine the reasons for tungsten boron carbide's success over the other investigated materials. This analysis determined that there was a positive correlation between a non-elastic interaction contribution towards a material's total cross-section and shielding performance within the thermal and epi-thermal energy regimes. This finding was assumed to be a result of the boron-10 absorption reaction. The analysis also determined that within the faster energy regions, materials featuring higher non-elastic interaction contributions were comparable to those exhibiting primarily elastic scattering via low Z elements. This allowed for the conclusion that composite shield success within higher energy neutron spectra does not necessitate the use elastic scattering via low Z elements. These findings suggest that the inclusion of materials featuring high thermal absorption properties is more critical to composite neutron shield performance than the presence of constituent materials more inclined to maximize elastic scattering energy loss.

  1. A lumped parameter method of characteristics approach and multigroup kernels applied to the subgroup self-shielding calculation in MPACT

    International Nuclear Information System (INIS)

    Stimpson, Shane G.; Liu, Yuxuan; Collins, Benjamin S.; Clarno, Kevin T.

    2017-01-01

    An essential component of the neutron transport solver is the resonance self-shielding calculation used to determine equivalence cross sections. The neutron transport code, MPACT, is currently using the subgroup self-shielding method, in which the method of characteristics (MOC) is used to solve purely absorbing fixed-source problems. Recent efforts incorporating multigroup kernels to the MOC solvers in MPACT have reduced runtime by roughly 2×. Applying the same concepts for self-shielding and developing a novel lumped parameter approach to MOC, substantial improvements have also been made to the self-shielding computational efficiency without sacrificing any accuracy. These new multigroup and lumped parameter capabilities have been demonstrated on two test cases: (1) a single lattice with quarter symmetry known as VERA (Virtual Environment for Reactor Applications) Progression Problem 2a and (2) a two-dimensional quarter-core slice known as Problem 5a-2D. From these cases, self-shielding computational time was reduced by roughly 3–4×, with a corresponding 15–20% increase in overall memory burden. An azimuthal angle sensitivity study also shows that only half as many angles are needed, yielding an additional speedup of 2×. In total, the improvements yield roughly a 7–8× speedup. Furthermore given these performance benefits, these approaches have been adopted as the default in MPACT.

  2. A lumped parameter method of characteristics approach and multigroup kernels applied to the subgroup self-shielding calculation in MPACT

    Directory of Open Access Journals (Sweden)

    Shane Stimpson

    2017-09-01

    Full Text Available An essential component of the neutron transport solver is the resonance self-shielding calculation used to determine equivalence cross sections. The neutron transport code, MPACT, is currently using the subgroup self-shielding method, in which the method of characteristics (MOC is used to solve purely absorbing fixed-source problems. Recent efforts incorporating multigroup kernels to the MOC solvers in MPACT have reduced runtime by roughly 2×. Applying the same concepts for self-shielding and developing a novel lumped parameter approach to MOC, substantial improvements have also been made to the self-shielding computational efficiency without sacrificing any accuracy. These new multigroup and lumped parameter capabilities have been demonstrated on two test cases: (1 a single lattice with quarter symmetry known as VERA (Virtual Environment for Reactor Applications Progression Problem 2a and (2 a two-dimensional quarter-core slice known as Problem 5a-2D. From these cases, self-shielding computational time was reduced by roughly 3–4×, with a corresponding 15–20% increase in overall memory burden. An azimuthal angle sensitivity study also shows that only half as many angles are needed, yielding an additional speedup of 2×. In total, the improvements yield roughly a 7–8× speedup. Given these performance benefits, these approaches have been adopted as the default in MPACT.

  3. Elastomeric neutron shielding material and process of production

    International Nuclear Information System (INIS)

    John, G.; Knorr, W.

    1987-01-01

    Elastomeric neutron shielding material made of plastic with high hydrogen content, characterized in that the shielding material is a polymeric reaction product of a reaction between (a) polyol on the base of polybutadiene which compares with polyethylene with regard to hydrogen content, and (b) aliphatic diisocyanate, and in that the hydrogen content is higher than 8 weight per cent. (orig.) [de

  4. Dosimetry in Thermal Neutron Irradiation Facility at BMRR

    Directory of Open Access Journals (Sweden)

    Hu J.-P.

    2016-01-01

    Full Text Available Radiation dosimetry for Neutron Capture Therapy (NCT has been performed since 1959 at Thermal Neutron Irradiation Facility (TNIF of the three-megawatt light-water cooled Brookhaven Medical Research Reactor (BMRR. In the early 1990s when more effective drug carriers were developed for NCT, in which the eye melanoma and brain tumors in rats were irradiated in situ, extensive clinical trials of small animals began using a focused thermal neutron beam. To improve the dosimetry at irradiation facility, a series of innovative designs and major modifications made to enhance the beam intensity and to ease the experimental sampling at BMRR were performed; including (1 in-core fuel addition to increase source strength and balance flux of neutrons towards two ports, (2 out of core moderator remodeling, done by replacing thicker D2O tanks at graphite-shutter interfacial areas, to expedite neutron thermalization, (3 beam shutter upgrade to reduce strayed neutrons and gamma dose, (4 beam collimator redesign to optimize the beam flux versus dose for animal treatment, (5 beam port shielding installation around the shutter opening area (lithium-6 enriched polyester-resin in boxes, attached with polyethylene plates to reduce prompt gamma and fast neutron doses, (6 sample holder repositioning to optimize angle versus distance for a single organ or whole body irradiation, and (7 holder wall buildup with neutron reflector materials to increase dose and dose rate from scattered thermal neutrons. During the facility upgrade, reactor dosimetry was conducted using thermoluminescent dosimeters TLD for gamma dose estimate, using ion chambers to confirm fast neutron and gamma dose rate, and by the activation of gold-foils with and without cadmium-covers, for fast and thermal neutron flux determination. Based on the combined effect from the size and depth of tumor cells and the location and geometry of dosimeters, the measured flux from cadmium-difference method was 4–7

  5. Dosimetry in Thermal Neutron Irradiation Facility at BMRR

    Energy Technology Data Exchange (ETDEWEB)

    Hu, J. P. [Brookhaven National Lab. (BNL), Upton, NY (United States); Holden, N. E. [Brookhaven National Lab. (BNL), Upton, NY (United States); Reciniello, R. N.

    2014-05-23

    Radiation dosimetry for Neutron Capture Therapy (NCT) has been performed since 1959 at Thermal Neutron Irradiation Facility (TNIF) of the three-megawatt light-water cooled Brookhaven Medical Research Reactor (BMRR). In the early 1990s when more effective drug carriers were developed for NCT, in which the eye melanoma and brain tumors in rats were irradiated in situ, extensive clinical trials of small animals began using a focused thermal neutron beam. To improve the dosimetry at irradiation facility, a series of innovative designs and major modifications made to enhance the beam intensity and to ease the experimental sampling at BMRR were performed; including (1) in-core fuel addition to increase source strength and balance flux of neutrons towards two ports, (2) out of core moderator remodeling, done by replacing thicker D2O tanks at graphite-shutter interfacial areas, to expedite neutron thermalization, (3) beam shutter upgrade to reduce strayed neutrons and gamma dose, (4) beam collimator redesign to optimize the beam flux versus dose for animal treatment, (5) beam port shielding installation around the shutter opening area (lithium-6 enriched polyester-resin in boxes, attached with polyethylene plates) to reduce prompt gamma and fast neutron doses, (6) sample holder repositioning to optimize angle versus distance for a single organ or whole body irradiation, and (7) holder wall buildup with neutron reflector materials to increase dose and dose rate from scattered thermal neutrons. During the facility upgrade, reactor dosimetry was conducted using thermoluminescent dosimeters TLD for gamma dose estimate, using ion chambers to confirm fast neutron and gamma dose rate, and by the activation of gold-foils with and without cadmium-covers, for fast and thermal neutron flux determination. Based on the combined effect from the size and depth of tumor cells and the location and geometry of dosimeters, the measured flux from cadmium-difference method was 4 - 7

  6. Transmission test of the polyethylene shield against 40 and 65 MeV quasi monochrome neutron

    International Nuclear Information System (INIS)

    Nakao, Makoto; Nakamura, Takashi; Sakuya, Yoshimasa; Nauchi, Yasushi; Nakao, Noriaki; Tanaka, Susumu; Sakamoto, Yukio; Nakajima, Hiroshi; Nakane, Yoshihiro.

    1996-01-01

    Using 40 and 65 MeV quasi monochrome neutron of the AVF cyclotron installed at Takasaki Laboratory, Japan Atomic Energy Research Institute, the neutron energy spectra were measured after transmitting the polyethylene shield. Results of the shielding experiments using concrete and iron recognized as main shielding material were proposed previously. As data obtained in the experiments were useful for a bench-mark experiment to investigate for shielding calculation and sectional data set, a shielding calculation simulated with new experiment to compare with and investigate for the previous experimental data. As a result, it was found that calculation result of neutron flux transmitting through the polyethylene shield showed difference with increase of the shield thickness. And, reducing distance of the peak neutron was also found to be over-estimated in its calculation value, such as three and five times on 43 MeV at 120 and 180 cm thick, respectively. (G.K.)

  7. Unresolved resonance self shielding calculation: causes and importance of discrepancies

    International Nuclear Information System (INIS)

    Ribon, P.; Tellier, H.

    1986-01-01

    To compute the self shielding coefficient, it is necessary to know the point-wise cross-sections. In the unresolved resonance region, the parameters of each level are not known; only the average parameters. Therefore the authors simulate the point-wise cross-section by random sampling of the energy levels and resonance parameters with respect to the Wigner law and the x 2 distributions, and by computing the cross-section in the same way as in the resolved regions. The result of this statistical calculation obviously depends on the initial parameters but also on the method of sampling, on the formalism which is used to compute the cross-section or on the weighting neutron flux. In this paper, the authors survey the main phenomena which can induce discrepancies in self shielding computations. Results are given for typical dilutions which occur in nuclear reactors

  8. Determination of the neutron energy and spatial distributions of the neutron beam from the TSR-II in the large beam shield

    International Nuclear Information System (INIS)

    Clifford, C.E.; Muckenthaler, F.J.

    1976-01-01

    The TSR-II reactor of the ORNL Tower Shielding Facility has recently been relocated within a new, fixed shield. A principal feature of the new shield is a beam port of considerably larger area than that of its predecessor. The usable neutron flux has thereby been increased by a factor of approximately 200. The bare beam neutron spectrum behind the new shield has been experimentally determined over the energy range from 0.8 to 16 MeV. A high level of fission product gamma ray background prevented measurement of bare beam spectra below 0.8 MeV, however neutron spectra in the energy range from 8 keV to 1.4 MeV were obtained for two simple, calculable shielding configurations. Also measured in the present work were weighted integral flux distributions and fast neutron dose rates

  9. Optimization of Shielding- Collimator Parameters for ING-27 Neutron Generator Using MCNP5

    Directory of Open Access Journals (Sweden)

    Hegazy Aya Hamdy

    2018-01-01

    Full Text Available Neutron generators are now used in various fields. They produce only fast neutrons; D-D neutron generator produces 2.45 MeV neutrons and D-T produces 14.1 MeV neutrons. In order to optimize shielding-collimator parameters to achieve higher neutron flux at the investigated sample (The signal with lower neutron and gamma rays flux at the area of the detectors, design iterations are widely used. This work was applied to ROMASHA setup, TANGRA project, FLNP, Joint Institute for Nuclear Research. The studied parameters were; (1 shielding-collimator material, (2 Distance between the shielding-collimator assembly first plate and center of the neutron beam, and (3 thickness of collimator sheets. MCNP5 was used to simulate ROMASHA setup after it was validated on the experimental results of irradiation of Carbon-12 sample for one hour to detect its 4.44 MeV characteristic gamma line. The ratio between the signal and total neutron flux that enters each detector was calculated and plotted, concluding that the optimum shielding-collimator assembly is Tungsten of 5 cm thickness for each plate, and a distance of 2.3 cm. Also, the ratio between the signal and total gamma rays flux was calculated and plotted for each detector, leading to the previous conclusion but the distance was 1 cm.

  10. Optimization of Shielding- Collimator Parameters for ING-27 Neutron Generator Using MCNP5

    Science.gov (United States)

    Hegazy, Aya Hamdy; Skoy, V. R.; Hossny, K.

    2018-04-01

    Neutron generators are now used in various fields. They produce only fast neutrons; D-D neutron generator produces 2.45 MeV neutrons and D-T produces 14.1 MeV neutrons. In order to optimize shielding-collimator parameters to achieve higher neutron flux at the investigated sample (The signal) with lower neutron and gamma rays flux at the area of the detectors, design iterations are widely used. This work was applied to ROMASHA setup, TANGRA project, FLNP, Joint Institute for Nuclear Research. The studied parameters were; (1) shielding-collimator material, (2) Distance between the shielding-collimator assembly first plate and center of the neutron beam, and (3) thickness of collimator sheets. MCNP5 was used to simulate ROMASHA setup after it was validated on the experimental results of irradiation of Carbon-12 sample for one hour to detect its 4.44 MeV characteristic gamma line. The ratio between the signal and total neutron flux that enters each detector was calculated and plotted, concluding that the optimum shielding-collimator assembly is Tungsten of 5 cm thickness for each plate, and a distance of 2.3 cm. Also, the ratio between the signal and total gamma rays flux was calculated and plotted for each detector, leading to the previous conclusion but the distance was 1 cm.

  11. FENDL neutronics benchmark: Specifications for the calculational neutronics and shielding benchmark

    International Nuclear Information System (INIS)

    Sawan, M.E.

    1994-12-01

    During the IAEA Advisory Group Meeting on ''Improved Evaluations and Integral Data Testing for FENDL'' held in Garching near Munich, Germany in the period 12-16 September 1994, the Working Group II on ''Experimental and Calculational Benchmarks on Fusion Neutronics for ITER'' recommended that a calculational benchmark representative of the ITER design should be developed. This report describes the neutronics and shielding calculational benchmark available for scientists interested in performing analysis for this benchmark. (author)

  12. Effect of different lay-ups on the microstructure, mechanical properties and neutron transmission of neutron shielding fibre metal laminates

    International Nuclear Information System (INIS)

    Fu, Xuelong; Tang, Xiaobin; Hu, Yubing; Li, Huaguan; Tao, Jie

    2016-01-01

    A novel neutron shielding fibre metal laminates (NSFMLs) with different lay-ups, composed of stacking layers of AA6061 plates, neutron shielding composite and carbon fibre reinforced polyimide (CFRP), were fabricated using hot molding process in atmospheric environments. The microstructure, mechanical properties and neutron transmission of the NSFMLs were evaluated, respectively. The results indicated that the NSFMLs possessed good mechanical properties owing to the good interfacial adhesion of the components. Tensile strength and elastic modulus of the NSFMLs increased with the numbers of lay-ups, while the elongation to fracture exhibited obvious declining tendency. Flexural strength and modulus of the NSFMLs were improved obviously with the increasing of stacking layers. Neutron transmission of the NSFMLs decreased obviously with increasing the number of lay-ups, owing to the increase of "1"0B areal density. Besides, the effect of carbon fibres on the neutron shielding performance of the NSFMLs was also taken into consideration. - Highlights: • A novel neutron shielding fibre metal laminates (NSFMLs) with different lay-ups was successfully fabricated using hot molding process. • Mechanical properties of the NSFMLs were performed in accordance with relative standards. • Neutron transmission of the NSFMLs was conducted according to the testing results. • The effect of carbon fibres on the neutron transmission of the NSFMLs was also investigated.

  13. Effect of different lay-ups on the microstructure, mechanical properties and neutron transmission of neutron shielding fibre metal laminates

    Energy Technology Data Exchange (ETDEWEB)

    Fu, Xuelong [College of Material Science & Technology, Nanjing University of Aeronautics & Astronautics, Nanjing, 211100 (China); Department of Mechanical and Electronic Engineering, Jiangsu Polytechnic of Finance & Economics, Huai' an, 223003 (China); Tang, Xiaobin; Hu, Yubing; Li, Huaguan [College of Material Science & Technology, Nanjing University of Aeronautics & Astronautics, Nanjing, 211100 (China); Tao, Jie, E-mail: taojie@nuaa.edu.cn [College of Material Science & Technology, Nanjing University of Aeronautics & Astronautics, Nanjing, 211100 (China)

    2016-07-15

    A novel neutron shielding fibre metal laminates (NSFMLs) with different lay-ups, composed of stacking layers of AA6061 plates, neutron shielding composite and carbon fibre reinforced polyimide (CFRP), were fabricated using hot molding process in atmospheric environments. The microstructure, mechanical properties and neutron transmission of the NSFMLs were evaluated, respectively. The results indicated that the NSFMLs possessed good mechanical properties owing to the good interfacial adhesion of the components. Tensile strength and elastic modulus of the NSFMLs increased with the numbers of lay-ups, while the elongation to fracture exhibited obvious declining tendency. Flexural strength and modulus of the NSFMLs were improved obviously with the increasing of stacking layers. Neutron transmission of the NSFMLs decreased obviously with increasing the number of lay-ups, owing to the increase of {sup 10}B areal density. Besides, the effect of carbon fibres on the neutron shielding performance of the NSFMLs was also taken into consideration. - Highlights: • A novel neutron shielding fibre metal laminates (NSFMLs) with different lay-ups was successfully fabricated using hot molding process. • Mechanical properties of the NSFMLs were performed in accordance with relative standards. • Neutron transmission of the NSFMLs was conducted according to the testing results. • The effect of carbon fibres on the neutron transmission of the NSFMLs was also investigated.

  14. Radiation transport and shielding information, computer codes, and nuclear data for use in CTR neutronics research and development

    International Nuclear Information System (INIS)

    Santoro, R.T.; Maskewitz, B.F.; Roussin, R.W.; Trubey, D.K.

    1976-01-01

    The activities of the Radiation Shielding Information Center (RSIC) of the Oak Ridge National Laboratory are being utilized in support of fusion reactor technology. The major activities of RSIC include the operation of a computer-based information storage and retrieval system, the collection, packaging, and distribution of large computer codes, and the compilation and dissemination of processed and evaluated data libraries, with particular emphasis on neutron and gamma-ray cross-section data. The Center has acquired thirteen years of experience in serving fission reactor, weapons, and accelerator shielding research communities, and the extension of its technical base to fusion reactor research represents a logical progression. RSIC is currently working with fusion reactor researchers and contractors in computer code development to provide tested radiation transport and shielding codes and data library packages. Of significant interest to the CTR community are the 100 energy group neutron and 21 energy group gamma-ray coupled cross-section data package (DLC-37) for neutronics studies, a comprehensive 171 energy group neutron and 36 energy group gamma-ray coupled cross-section data base with retrieval programs, including resonance self-shielding, that are tailored to CTR application, and a data base for the generation of energy-dependent atomic displacement and gas production cross sections and heavy-particle-recoil spectra for estimating radiation damage to CTR structural components

  15. Analysis of shield for the nuclear ship MUTSU

    International Nuclear Information System (INIS)

    Fuse, Takayoshi; Takeuchi, Kiyoshi; Yamaji, Akio

    1975-01-01

    On the nuclear ship MUTSU, a higher-than-expected level of radiation was found, with output raised to 1.4 per cent. To investigate the radiation leakage, the analysis of the shielding problem utilized a four-step sequence of PALLAS-2DCY cylindrical r-z calculations with fixed sources distributions in the core. The neutron dose contours show the importance of streaming in the gap between the reactor vessel and the primary shield. Dominant consideration of thermal insulation exclude shielding from this area resulting in an imbalance in the shielding effectiveness. The neutron dose rate at the upper part of the reactor vessel is increased by neutrons incident on the head from cavity scattering. The calculation indicates that the neutron dose rate at the top of the primary shield is 5 rem/hr at 100 per cent output. (auth.)

  16. Application of the decoupling scheme on complex neutron-gamma shielding problems

    Energy Technology Data Exchange (ETDEWEB)

    Feher, S. [Institute of Nuclear Technology, Technical University of Budapest, Budapest (Hungary); Leege, P.F.A. de; Hoogenboom, J.E.; Kloosterman, J.L. [Interfaculty Reactor Institute, Delft University of Technology, Delft (Netherlands)

    2000-03-01

    Coupled neutron-gamma shielding calculations using S{sub n} transport theory can be time consuming, especially for two- and three-dimensional geometries. In general, the CPU time of these calculations increases stronger than linear with increasing number of neutron and gamma energy groups, and depends on the order of Legendre expansion and number of S{sub n} directions used. This fact induced the idea of the decoupling method, which seems applicable to accelerate coupled neutron-gamma shielding calculations. The data included in a combined neutron-gamma library can be readily separated into a library containing neutron data only and another library containing gamma data only. Separate calculations for neutrons and gammas are performed on complex geometries using a different Legendre order expansion for neutrons and gammas. CPU savings of 60 to 85% can be achieved for the two-dimensional DORT and three-dimensional TORT calculations respectively. (author)

  17. Under Water Thermal Cutting of the Moderator Vessel and Thermal Shield

    International Nuclear Information System (INIS)

    Loeb, A.; Sokcic-Kostic, M.; Eisenmann, B.; Prechtl, E.

    2007-01-01

    This paper presents the segmentation of the in 8 meter depth of water and for cutting through super alloyed moderator vessel and of the thermal shield of the MZFR stainless steel up to 130 mm wall thickness. Depending on the research reactor by means of under water plasma and contact arc metal cutting. The moderator vessel and the thermal shield are the most essential parts of the MZFR reactor vessel internals. These components have been segmented in 2005 by means of remotely controlled under water cutting utilizing a special manipulator system, a plasma torch and CAMC (Contact Arc Metal Cutting) as cutting tools. The engineered equipment used is a highly advanced design developed in a two years R and D program. It was qualified to cut through steel walls of more than 100 mm thickness in 8 meters water depth. Both the moderator vessel and the thermal shield had to be cut into such size that the segments could afterwards be packed into shielded waste containers each with a volume of roughly 1 m 3 . Segmentation of the moderator vessel and of the thermal shield was performed within 15 months. (author)

  18. Upgrade of the LHC magnet interconnections thermal shielding

    Energy Technology Data Exchange (ETDEWEB)

    Musso, Andrea; Barlow, Graeme; Bastard, Alain; Charrondiere, Maryline; Deferne, Guy; Dib, Gaëlle; Duret, Max; Guinchard, Michael; Prin, Hervé; Craen, Arnaud Vande; Villiger, Gilles [CERN European Organization for Nuclear Research, Meyrin 1211, Geneva 23, CH (Switzerland); Chrul, Anna [The Henryk Niewodniczanski Institute of Nuclear Physics, Polish Academy of Sciences, ul.Radzikowskiego 152, 31-324 Krakow (Poland); Damianoglou, Dimitrios [NTUA National Technical University of Athens, Heeron Polytechniou 9, 15780 Zografou (Greece); Strychalski, Michał [Wroclaw University of Technology, Faculty of Mechanical and Power Engineering, Wyb. Wyspianskiego 27, Wroclaw, 50-370 (Poland); Wright, Loren [Lancaster University, Bailrigg, Lancaster, LA1 4YW (United Kingdom)

    2014-01-29

    The about 1700 interconnections (ICs) between the Large Hadron Collider (LHC) superconducting magnets include thermal shielding at 50-75 K, providing continuity to the thermal shielding of the magnet cryostats to reduce the overall radiation heat loads to the 1.9 K helium bath of the magnets. The IC shield, made of aluminum, is conduction-cooled via a welded bridge to the thermal shield of the adjacent magnets which is actively cooled. TIG welding of these bridges made in the LHC tunnel at installation of the magnets induced a considerable risk of fire hazard due to the proximity of the multi-layer insulation of the magnet shields. A fire incident occurred in one of the machine sectors during machine installation, but fortunately with limited consequences thanks to prompt intervention of the operators. LHC is now undergoing a 2 years technical stop during which all magnet's ICs will have to be opened to consolidate the magnet electrical connections. The IC thermal shields will therefore have to be removed and re-installed after the work is completed. In order to eliminate the risk of fire hazard when re-welding, it has been decided to review the design of the IC shields, by replacing the welded bridges with a mechanical clamping which also preserves its thermal function. An additional advantage of this new solution is the ease in dismantling for maintenance, and eliminating weld-grinding operations at removal needing radioprotection measures because of material activation after long-term operation of the LHC. This paper describes the new design of the IC shields and in particular the theoretical and experimental validation of its thermal performance. Furthermore a status report of the on-going upgrade work in the LHC is given.

  19. Upgrade of the LHC magnet interconnections thermal shielding

    Science.gov (United States)

    Musso, Andrea; Barlow, Graeme; Bastard, Alain; Charrondiere, Maryline; Chrul, Anna; Damianoglou, Dimitrios; Deferne, Guy; Dib, Gaëlle; Duret, Max; Guinchard, Michael; Prin, Hervé; Strychalski, Michał; Craen, Arnaud Vande; Villiger, Gilles; Wright, Loren

    2014-01-01

    The about 1700 interconnections (ICs) between the Large Hadron Collider (LHC) superconducting magnets include thermal shielding at 50-75 K, providing continuity to the thermal shielding of the magnet cryostats to reduce the overall radiation heat loads to the 1.9 K helium bath of the magnets. The IC shield, made of aluminum, is conduction-cooled via a welded bridge to the thermal shield of the adjacent magnets which is actively cooled. TIG welding of these bridges made in the LHC tunnel at installation of the magnets induced a considerable risk of fire hazard due to the proximity of the multi-layer insulation of the magnet shields. A fire incident occurred in one of the machine sectors during machine installation, but fortunately with limited consequences thanks to prompt intervention of the operators. LHC is now undergoing a 2 years technical stop during which all magnet's ICs will have to be opened to consolidate the magnet electrical connections. The IC thermal shields will therefore have to be removed and re-installed after the work is completed. In order to eliminate the risk of fire hazard when re-welding, it has been decided to review the design of the IC shields, by replacing the welded bridges with a mechanical clamping which also preserves its thermal function. An additional advantage of this new solution is the ease in dismantling for maintenance, and eliminating weld-grinding operations at removal needing radioprotection measures because of material activation after long-term operation of the LHC. This paper describes the new design of the IC shields and in particular the theoretical and experimental validation of its thermal performance. Furthermore a status report of the on-going upgrade work in the LHC is given.

  20. Unresolved resonance self shielding calculation: causes and importance of discrepancies

    International Nuclear Information System (INIS)

    Ribon, P.; Tellier, H.

    1986-09-01

    To compute the self shielding coefficient, it is necessary to know the point-wise cross-sections. In the unresolved resonance region, we do not know the parameters of each level but only the average parameters. Therefore we simulate the point-wise cross-section by random sampling of the energy levels and resonance parameters with respect to the Wigner law and the X 2 distributions, and by computing the cross-section in the same way as in the resolved regions. The result of this statistical calculation obviously depends on the initial parameters but also on the method of sampling, on the formalism which is used to compute the cross-section or on the weighting neutron flux. In this paper, we will survey the main phenomena which can induce discrepancies in self shielding computations. Results are given for typical dilutions which occur in nuclear reactors. 8 refs

  1. SUBGR: A Program to Generate Subgroup Data for the Subgroup Resonance Self-Shielding Calculation

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Kang Seog [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2016-06-06

    The Subgroup Data Generation (SUBGR) program generates subgroup data, including levels and weights from the resonance self-shielded cross section table as a function of background cross section. Depending on the nuclide and the energy range, these subgroup data can be generated by (a) narrow resonance approximation, (b) pointwise flux calculations for homogeneous media; and (c) pointwise flux calculations for heterogeneous lattice cells. The latter two options are performed by the AMPX module IRFFACTOR. These subgroup data are to be used in the Consortium for Advanced Simulation of Light Water Reactors (CASL) neutronic simulator MPACT, for which the primary resonance self-shielding method is the subgroup method.

  2. Thermal Analysis of a SHIELD Electromigration Test Structure

    Energy Technology Data Exchange (ETDEWEB)

    Benson, David A.; Bowman, Duane J.; Mitchell, Robert T.

    1999-05-01

    The steady state and transient thermal behavior of an electromigration test structure was analyzed. The test structure was a Sandia SHIELD (Self-stressing HIgh fregquency rELiability Device) electromigration test device manufactured by an outside vendor. This device has a high frequency oscillator circuit, a buffer circuit to isolate and drive the metal line to the tested (DUT), the DUT to be electromigrated itself, a metal resistance thermometry monitor, and a heater elment to temperature accelerate the electromigration effect.

  3. Soil biological shield exposed to high energy neutrons; Zemlja kao bioloski stit od neutrona visokih energija

    Energy Technology Data Exchange (ETDEWEB)

    Simovic, R; Marinkovic, N [Institute of nuclear sciences Vinca, Belgrade (Yugoslavia)

    1993-04-15

    Shielding efficiency of soil biological shield exposed to high energy neutrons was investigated. Dose rate equivalents for neutrons, secondary gamma and gamma radiation were computed on the surface of soil slabs having different thicknesses. Yields of primary and secondary nuclear radiation in the total dose were evaluated. Influence of the incident neutron spectrum, water content and chemical composition of the material on its shielding efficiency was examined. It was found that the soil density and the water content determine the quality of biological shield, the influence of other factors being less important. Comparison of shielding efficiencies for soil with sand, brick and ordinary concrete shields was done.

  4. Evaluation of three partially volatile neutron shields for high-performance shipping casks

    International Nuclear Information System (INIS)

    Rack, H.J.; Pearson, H.S.

    1981-02-01

    The thermal stability and mechanical behavior of three partially volatile candidate neutron shield materials have been evaluated. The results indicate that silicone based rubbers, impregnated with elemental boron or boron carbide, Boro-silicone 236 and Bisco NS-I respectively are more thermally stable than are borated beechwoods, e.g., Permali JN. Mechanical property measurements indicated however that the compressive strength of the borated beechwood is 10 to 48 times higher than that of the silicone-based rubbers. The compressive strengths of the borated beechwood and boron carbide impregnated silicone rubber were substantially more sensitive to test temperature than was the compressive strength of the boron impregnated silicone rubber. Finally the compressive strengths and energy absorbing capability of the boron impregnated silicone rubber is not affected by prior thermal exposure at 425 0 K for 1000h

  5. Cooling Performance of TBM-shield Designed for Manufacturability

    International Nuclear Information System (INIS)

    Park, Seong Dae; Lee, Dong Won; Kim, Dong Jun; Yoon, Jae Sung; Ahn, Mu Young

    2016-01-01

    Helium cooled ceramic reflector (HCCR) test blanket module (TBM) is composed of four sub-modules and a common back manifold (BM). The associated shield is a water-cooled 316L(N)-IG block with internal cooling channels. The purpose of the TBM-shield is to make the condition with the allowable neutron flux and dose rate level. The radially continuous layers of water and structure were configured. The main purpose of the shield is to reduce the neutron flux by absorbing the neutron in the structure. The water could act as the moderator and cool down the structure which is heated due to the reaction with the neutrons. The moderated neutrons are easily absorbed by the structure. It could meet the criteria for the minimum neutron flux by increasing the thickness of structure. The formation of inside cooling channel in the TBM-shield should be considered while maintaining the allowable temperature range. In this work, a manufacturing process including the formation of inside cooling channel was presented. Current design and thermal analysis results for the TBM-shield were presented. The geometry of the shield blocks was considerably changed. The coolant channel was exposed to the outer surface of the TBM-shield. The overall manufacturing process is simplified compared with the previous process of CD model

  6. Cooling Performance of TBM-shield Designed for Manufacturability

    Energy Technology Data Exchange (ETDEWEB)

    Park, Seong Dae; Lee, Dong Won; Kim, Dong Jun; Yoon, Jae Sung [KAERI, Daejeon (Korea, Republic of); Ahn, Mu Young [NFRI, Daejeon (Korea, Republic of)

    2016-05-15

    Helium cooled ceramic reflector (HCCR) test blanket module (TBM) is composed of four sub-modules and a common back manifold (BM). The associated shield is a water-cooled 316L(N)-IG block with internal cooling channels. The purpose of the TBM-shield is to make the condition with the allowable neutron flux and dose rate level. The radially continuous layers of water and structure were configured. The main purpose of the shield is to reduce the neutron flux by absorbing the neutron in the structure. The water could act as the moderator and cool down the structure which is heated due to the reaction with the neutrons. The moderated neutrons are easily absorbed by the structure. It could meet the criteria for the minimum neutron flux by increasing the thickness of structure. The formation of inside cooling channel in the TBM-shield should be considered while maintaining the allowable temperature range. In this work, a manufacturing process including the formation of inside cooling channel was presented. Current design and thermal analysis results for the TBM-shield were presented. The geometry of the shield blocks was considerably changed. The coolant channel was exposed to the outer surface of the TBM-shield. The overall manufacturing process is simplified compared with the previous process of CD model.

  7. Neutron Buildup Factors Calculation for Support Vector Regression Application in Shielding Analysis

    International Nuclear Information System (INIS)

    Duckic, P.; Matijevic, M.; Grgic, D.

    2016-01-01

    In this paper initial set of data for neutron buildup factors determination using Support Vector Regression (SVR) method is prepared. The performance of SVR technique strongly depends on the quality of information used for model training. Thus it is very important to provide representable data to the SVR. SVR is a supervised type of learning so it demands data in the input/output form. In the case of neutron buildup factors estimation, the input parameters are the incident neutron energy, shielding thickness and shielding material and the output parameter is the neutron buildup factor value. So far the initial sets of data for different shielding configurations have been obtained using SCALE4.4 sequence SAS3. However, this results were obtained using group constants, thus the incident neutron energy was determined as the average value for each energy group. Obtained this way, the data provided to the SVR are fewer and therefore insufficient. More valuable information is obtained using SCALE6.2beta5 sequence MAVRIC which can perform calculations for the explicit incident neutron energy, which leads to greater maneuvering possibilities when active learning measures are employed, and consequently improves the quality of the developed SVR model.(author).

  8. Shielding calculations for the design of neutron radiography facility around PARR

    International Nuclear Information System (INIS)

    Ashraf, M.M.; Khan, A.R.

    1989-06-01

    Shielding calculations for neutron radiography facility, proposed to be established around PARR have been carried out using two group diffusion theory and shielding formulae. Gamma radiation penetration calculations have been carried out using simple attenuation methods. The fabrication and installation of the neutron radiography facility would provide the basis for designing a better collimating system and would help establish under water radiography facility for the inspection of highly radioactive materials and components etc. (orig./A.B.)

  9. Radiation shielding properties of high performance concrete reinforced with basalt fibers infused with natural and enriched boron

    Energy Technology Data Exchange (ETDEWEB)

    Zorla, Eyüp; Ipbüker, Cagatay [University of Tartu, Institute of Physics (Estonia); Biland, Alex [US Basalt Corp., Houston (United States); Kiisk, Madis [University of Tartu, Institute of Physics (Estonia); Kovaljov, Sergei [OÜ Basaltest, Tartu (Estonia); Tkaczyk, Alan H. [University of Tartu, Institute of Physics (Estonia); Gulik, Volodymyr, E-mail: volodymyr.gulik@gmail.com [Institute for Safety Problems of Nuclear Power Plants, Lysogirska 12, of. 201, 03028 Kyiv (Ukraine)

    2017-03-15

    Highlights: • Basalt fiber infused with natural and enriched boron in varying proportions. • Gamma-ray attenuation remains stable with addition of basalt-boron fiber. • Improvement in neutron shielding for nuclear facilities producing fast fission spectrum. • Basalt-boron fiber could decrease the shielding thickness in thermal spectrum reactors. - Abstract: The importance of radiation shielding is increasing in parallel with the expansion of the application areas of nuclear technologies. This study investigates the radiation shielding properties of two types of high strength concrete reinforced with basalt fibers infused with 12–20% boron oxide, containing varying fractions of natural and enriched boron. The gamma-ray shielding characteristics are analyzed with the help of the WinXCom, whereas the neutron shielding characteristics are modeled and computed by Monte Carlo Serpent code. For gamma-ray shielding, the attenuation coefficients of the studied samples do not display any significant variation due to the addition of basalt-boron fibers at any mixing proportion. For neutron shielding, the addition of basalt-boron fiber has negligible effects in the case of very fast neutrons (14 MeV), but it could considerably improve the neutron shielding of concrete for nuclear facilities producing a fast fission spectrum (e.g. with reactors as BN-800, FBTR) and thermal neutron spectrum (Light Water Reactors (LWR)). It was also found that basalt-boron fiber could decrease the thickness of radiation shielding material in thermal spectrum reactors.

  10. Radiation shielding properties of high performance concrete reinforced with basalt fibers infused with natural and enriched boron

    International Nuclear Information System (INIS)

    Zorla, Eyüp; Ipbüker, Cagatay; Biland, Alex; Kiisk, Madis; Kovaljov, Sergei; Tkaczyk, Alan H.; Gulik, Volodymyr

    2017-01-01

    Highlights: • Basalt fiber infused with natural and enriched boron in varying proportions. • Gamma-ray attenuation remains stable with addition of basalt-boron fiber. • Improvement in neutron shielding for nuclear facilities producing fast fission spectrum. • Basalt-boron fiber could decrease the shielding thickness in thermal spectrum reactors. - Abstract: The importance of radiation shielding is increasing in parallel with the expansion of the application areas of nuclear technologies. This study investigates the radiation shielding properties of two types of high strength concrete reinforced with basalt fibers infused with 12–20% boron oxide, containing varying fractions of natural and enriched boron. The gamma-ray shielding characteristics are analyzed with the help of the WinXCom, whereas the neutron shielding characteristics are modeled and computed by Monte Carlo Serpent code. For gamma-ray shielding, the attenuation coefficients of the studied samples do not display any significant variation due to the addition of basalt-boron fibers at any mixing proportion. For neutron shielding, the addition of basalt-boron fiber has negligible effects in the case of very fast neutrons (14 MeV), but it could considerably improve the neutron shielding of concrete for nuclear facilities producing a fast fission spectrum (e.g. with reactors as BN-800, FBTR) and thermal neutron spectrum (Light Water Reactors (LWR)). It was also found that basalt-boron fiber could decrease the thickness of radiation shielding material in thermal spectrum reactors.

  11. Collimator and shielding design for boron neutron capture therapy (BNCT) facility at TRIGA MARK II reactor

    International Nuclear Information System (INIS)

    Mohd Rafi Mohd Solleh; Abdul Aziz Tajuddin; Abdul Aziz Mohamed; Eid Mahmoud Eid Abdel Munem; Mohamad Hairie Rabir; Julia Abdul Karim; Yoshiaki, Kiyanagi

    2011-01-01

    The geometry of reactor core, thermal column, collimator and shielding system for BNCT application of TRIGA MARK II Reactor were simulated with MCNP5 code. Neutron particle lethargy and dose were calculated with MCNPX code. Neutron flux in a sample located at the end of collimator after normalized to measured value (Eid Mahmoud Eid Abdel Munem, 2007) at 1 MW power was 1.06 x 10 8 n/ cm 2 / s. According to IAEA (2001) flux of 1.00 x 10 9 n/ cm 2 / s requires three hours of treatment. Few modifications were needed to get higher flux. (Author)

  12. Characterization of the Shielded Neutron Source at Triangle Universities Nuclear Laboratory

    Science.gov (United States)

    Hobson, Chad; Finch, Sean; Howell, Calvin; Malone, Ron; Tornow, Wernew

    2016-09-01

    In 2015, Triangle Universities Nuclear Laboratory rebuilt its shielded neutron source (SNS) with the goal of improving neutron beam collimation and reducing neutron and gamma-ray backgrounds. Neutrons are produced via the 2H(d,n)3He reaction and then collimated by heavy shielding to form a beam. The SNS has the ability to produce both a rectangular and circular neutron beam through use of two collimators with different beam apertures. Our work characterized both the neutron beam profiles as well as the neutron and gamma-ray backgrounds at various locations around the SNS. This characterization was performed to provide researchers who use the SNS with beam parameters necessary to plan and conduct an experiment. Vertical and horizontal beam profiles were measured at two different distances from the neutron production cell by scanning a small plastic scintillator across the face of the beam at various energies for each collimator. Background neutron and gamma-ray intensities were measured using time-of-flight techniques at 10 MeV and 16 MeV with the rectangular collimator. We present results on the position and size of neutron beam as well as on the structure and magnitude of the backgrounds.

  13. Experimental investigation of thermal neutron analysis based landmine detection technology

    International Nuclear Information System (INIS)

    Zeng Jun; Chu Chengsheng; Ding Ge; Xiang Qingpei; Hao Fanhua; Luo Xiaobing

    2013-01-01

    Background: Recently, the prompt gamma-rays neutron activation analysis method is wildly used in coal analysis and explosive detection, however there were less application about landmine detection using neutron method especially in the domestic research. Purpose: In order to verify the feasibility of Thermal Neutron Analysis (TNA) method used in landmine detection, and explore the characteristic of this technology. Methods: An experimental system of TNA landmine detection was built based on LaBr 3 (Ce) fast scintillator detector and 252 Cf isotope neutron source. The system is comprised of the thermal neutron transition system, the shield system, and the detector system. Results: On the basis of the TNA, the wide energy area calibration method especially to the high energy area was investigated, and the least detection time for a typical mine was defined. In this study, the 72-type anti-tank mine, the 500 g TNT sample and several interferential objects are tested in loess, red soil, magnetic soil and sand respectively. Conclusions: The experimental results indicate that TNA is a reliable demining method, and it can be used to confirm the existence of Anti-Tank Mines (ATM) and large Anti-Personnel Mines (APM) in complicated condition. (authors)

  14. Neutron shielding and its impact on the ITER machine design

    International Nuclear Information System (INIS)

    Daenner, W.; El Guebaly, L.; Sawan, M.; Gohar, Y.; Maki, K.; Rado, V.; Schchipakin, O.; Zimin, S.

    1991-01-01

    This paper describes the efforts made in the frame of the ITER project to analyze the shielding of the superconducting magnets. First, the radiation limits to be achieved are specified as well as the neutron source in terms of wall loading on the first wall of the machine. Then the general shield concept is explained, including the most essential details of the various shield components. A brief section is devoted to the calculational tools, the data base, and the safety factors to be applied to the results obtained. The neutronics models of four different configurations are summarized as they were used to study the most critical parts of the machine. This section is followed by a presentation of the most important results from one-, two- and three-dimensional calculations. They are given for both the reference design and an improved one in which the critical regions are reinforced with respect to their shielding capability. It is concluded that the ITER shield layout just marginally meets the stated limits provided that some tungsten is included in the critical regions. A slight revision of the overall machine dimensions with the aim to achieve a less complex shield and a higher margin with respect to the limits is, however, seen the better solution. (orig.)

  15. Impact of thermal and intermediate energy neutrons on the semiconductor memories for the CERN accelerators

    CERN Document Server

    Cecchetto, Matteo; Gerardin, Simone

    A wide quantity of SRAM memories are employed along the Large Hadron Collider (LHC), the main CERN accelerator, and they are subjected to high levels of ionizing radiations which compromise the reliability of these devices. The Single Event Effect (SEE) qualification for components to be used in the complex high-energy accelerator at CERN relies on the characterization of two cross sections: 200-MeV protons and thermal neutrons. However, due to cost and time constraints, it is not always possible to characterize the SEE response of components to thermal neutrons, which is often regarded as negligible for components without borophosphosilicate glass (BPSG). Nevertheless, as recent studies show, the sensitivity of deep sub-micron technologies to thermal neutrons has increased owing to the presence of Boron 10 as a dopant and contact contaminant. The very large thermal neutron fluxes relative to high-energy hadron fluxes in some of the heavily shielded accelerator areas imply that even comparatively small therm...

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-08-11

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

  17. Experimental study on fast neutron streaming through grid-plate shield of a LMFBR

    International Nuclear Information System (INIS)

    Oka, Yoshiaki; Wakabayashi, Hiroaki; An, Shigehiro; Suzuki, Ikunori.

    1976-01-01

    Neutron streaming through the holes penetrating the grid plate shield of a prototype LMFBR was experimentally examined. The mockups of the grid plate shield were made of iron and aluminum. Experiments were conducted at the vertical column of ''YAYOI'', the fast neutron source reactor of University of Tokyo. A He-3 spectrometer was employed in order to measure the transmitted neutron spectrum, while rhodium and indium threshold foils were for the integral flux above specific energies and their spatial distributions in the form of reaction rates. The streaming factor for usual small bended holes is 1.28+-0.04 as to the integral neutron flux above 0.1 MeV and 1.30+-0.12 as to the reaction rate of indium foil. Use were made of the one and two dimensional neutron transport code ANISN and TWOTRAN for evaluation by computation. The reaction rates calculated by infinite slab model with ANISN code agree well with the experiments when normalized at the source point where neutrons are incident on the grid plate shield. (auth.)

  18. Validation of calculated self-shielding factors for Rh foils

    Science.gov (United States)

    Jaćimović, R.; Trkov, A.; Žerovnik, G.; Snoj, L.; Schillebeeckx, P.

    2010-10-01

    Rhodium foils of about 5 mm diameter were obtained from IRMM. One foil had thickness of 0.006 mm and three were 0.112 mm thick. They were irradiated in the pneumatic transfer system and in the carousel facility of the TRIGA reactor at the Jožef Stefan Institute. The foils were irradiated bare and enclosed in small cadmium boxes (about 2 g weight) of 1 mm thickness to minimise the perturbation of the local neutron flux. They were co-irradiated with 5 mm diameter and 0.2 mm thick Al-Au (0.1%) alloy monitor foils. The resonance self-shielding corrections for the 0.006 and 0.112 mm thick samples were calculated by the Monte Carlo simulation and amount to about 10% and 60%, respectively. The consistency of measurements confirmed the validity of self-shielding factors. Trial estimates of Q0 and k0 factors for the 555.8 keV gamma line of 104Rh were made and amount to 6.65±0.18 and (6.61±0.12)×10 -2, respectively.

  19. Alternative methodology for irradiation reactor experimental shielding calculation

    International Nuclear Information System (INIS)

    Vellozo, Sergio de Oliveira; Vital, Helio de Carvalho

    1996-01-01

    Due to a change in the project of the Experimental Irradiation Reactor, its shielding design had to be recalculated according to an alternative simplified analytical approach, since the standard transport calculations were temporarily unavailable. In the calculation of the new width for the shielding made up of steel and high-density concrete layers, the following radiation components were considered: fast neutrons and primary gammas (produced by fission and beta decay), from the core; and secondary gammas, produced by thermal neutron capture in the shielding. (author)

  20. Design of 6 Mev linear accelerator based pulsed thermal neutron source: FLUKA simulation and experiment

    Energy Technology Data Exchange (ETDEWEB)

    Patil, B.J., E-mail: bjp@physics.unipune.ac.in [Department of Physics, University of Pune, Pune 411 007 (India); Chavan, S.T.; Pethe, S.N.; Krishnan, R. [SAMEER, IIT Powai Campus, Mumbai 400 076 (India); Bhoraskar, V.N. [Department of Physics, University of Pune, Pune 411 007 (India); Dhole, S.D., E-mail: sanjay@physics.unipune.ac.in [Department of Physics, University of Pune, Pune 411 007 (India)

    2012-01-15

    The 6 MeV LINAC based pulsed thermal neutron source has been designed for bulk materials analysis. The design was optimized by varying different parameters of the target and materials for each region using FLUKA code. The optimized design of thermal neutron source gives flux of 3 Multiplication-Sign 10{sup 6}ncm{sup -2}s{sup -1} with more than 80% of thermal neutrons and neutron to gamma ratio was 1 Multiplication-Sign 10{sup 4}ncm{sup -2}mR{sup -1}. The results of prototype experiment and simulation are found to be in good agreement with each other. - Highlights: Black-Right-Pointing-Pointer The optimized 6 eV linear accelerator based thermal neutron source using FLUKA simulation. Black-Right-Pointing-Pointer Beryllium as a photonuclear target and reflector, polyethylene as a filter and shield, graphite as a moderator. Black-Right-Pointing-Pointer Optimized pulsed thermal neutron source gives neutron flux of 3 Multiplication-Sign 10{sup 6} n cm{sup -2} s{sup -1}. Black-Right-Pointing-Pointer Results of the prototype experiment were compared with simulations and are found to be in good agreement. Black-Right-Pointing-Pointer This source can effectively be used for the study of bulk material analysis and activation products.

  1. Using the shield for thermal energy storage in pulsar

    International Nuclear Information System (INIS)

    Sager, G.T.; Sze, D.K.; Wong, C.P.C.; Bathke, C.G.; Blanchard, J.P.; Brimer, C.; Cheng, E.T.; El-Guebaly, L.A.; Hasan, M.Z.; Najmabadi, F.; Sharafat, S.; Sviatoslavski, I.N.; Waganer, L.

    1995-01-01

    The PULSAR pulsed tokamak power plant design utilizes the outboard shield for thermal energy storage to maintain full 1000MW(e) output during the dwell period of 200s. Thermal energy resulting from direct nuclear heating is accumulated in the shield during the 7200s fusion power production phase. The maximum shield temperature may be much higher than that for the blanket because radiation damage is significantly reduced. During the dwell period, thermal power discharged from the shield and coolant temperature are simultaneously regulated by controlling the coolant mass flow rate at the shield inlet. This is facilitated by throttled coolant bypass. Design concepts using helium and lithium coolant have been developed. Two-dimensional time-dependent thermal hydraulic calculations were performed to confirm performance capabilities required of the design concepts. The results indicate that the system design and performance can accommodate uncertainties in material limits or the length of the dwell period. (orig.)

  2. Response of combined albedo-track neutron personnel dosimeters behind IHEP proton synchrotron shielding

    International Nuclear Information System (INIS)

    Sannikov, A.V.; Korshunova, E.P.

    1989-01-01

    The method of readings interpretation of combined albedo-track neutron personnel dosemeters based on calculationsl analysis of the detector responses in various neutron spectra is described. The measurements of dose equivalent responses have been performed in various points behind IHEP proton synchrotron shielding. It is shown that CDs with fission track detectors have a small dose equivalent response dispersion behind IHEP proton synchrotron shielding, that shows the promise of their using for neutron personnel monitoring, that shows the promise of their using for neutron personnel monitoring at high energy accelerators. 16 refs.; 7 figs.; 3 tabs

  3. Thermal design of top shield for PFBR

    International Nuclear Information System (INIS)

    Gajapathy, R.; Jalaludeen, S.; Selvaraj, A.; Bhoje, S.B.

    1988-01-01

    India's Liquid Metal Cooled Fast Breeder Reactor programme started with the construction of loop type 13MW(e) Fast Breeder Test Reactor (FBTR) which attained criticality in October 1985. With the experience of FBTR, the design work on pool type 500 MW(e) Prototype Fast Breeder Reactor (PFBR) which will be a forerunner for future commercial fast breeder reactors, has been started. The Top Shield forms the cover for the main vessel which contains the primary circuit. Argon cover gas separates the Top Shield from the free level of hot sodium pool (803K). The Top Shield which is of box type construction consists of control plug, two rotatable plugs and roof slab, assembled together, which provide biological shielding, thermal shielding and leak tight containment at the top of the main vessel. Heat is transferred from the sodium pool to the Top Shield through argon cover gas and through components supported by it and dipped in the sodium pool. The Top Shield should be maintained at the desired operating temperature by incorporating a cooling system inside it. Insulation may be provided below the bottom plate to reduce the heat load to the cooling system, if required. The thermal design of Top Shield consists of estimation of heat transfer to the Top Shield, selection of operating temperature, assessment of insulation requirement, design of cooling system and evaluation of transient temperature changes

  4. Resonance shielding in thermal reactor lattices

    International Nuclear Information System (INIS)

    Rothenstein, W.; Taviv, E.; Aminpour, M.

    1982-01-01

    The theoretical foundations of a new methodology for the accurate treatment of resonance absorption in thermal reactor lattice analysis are presented. This methodology is based on the solution of the point-energy transport equation in its integral or integro-differential form for a heterogeneous lattice using detailed resonance cross-section profiles. The methodology is applied to LWR benchmark analysis, with emphasis on temperature dependence of resonance absorption during fuel depletion, spatial and mutual self-shielding, integral parameter analysis and treatment of cluster geometry. The capabilities of the OZMA code, which implements the new methodology are discussed. These capabilities provide a means against which simpler and more rapid resonance absorption algorithms can be checked. (author)

  5. A comparison of neutron resonance absorption in thermal reactor lattices in the AUS neutronics code system with Monte Carlo calculations

    International Nuclear Information System (INIS)

    Robinson, G.S.

    1985-08-01

    The calculation of resonance shielding by the subgroup method, as incorporated in the MIRANDA module of the AUS neutronics code system, is compared with Monte Carlo calculatons for a number of thermal reactor lattices. For the large range of single rod and rod cluster lattices considered, AUS results for resonance absorption were high by up to two per cent

  6. CREST : a computer program for the calculation of composition dependent self-shielded cross-sections

    International Nuclear Information System (INIS)

    Kapil, S.K.

    1977-01-01

    A computer program CREST for the calculation of the composition and temperature dependent self-shielded cross-sections using the shielding factor approach has been described. The code includes the editing and formation of the data library, calculation of the effective shielding factors and cross-sections, a fundamental mode calculation to generate the neutron spectrum for the system which is further used to calculate the effective elastic removal cross-sections. Studies to explore the sensitivity of reactor parameters to changes in group cross-sections can also be carried out by using the facility available in the code to temporarily change the desired constants. The final self-shielded and transport corrected group cross-sections can be dumped on cards or magnetic tape in a suitable form for their direct use in a transport or diffusion theory code for detailed reactor calculations. The program is written in FORTRAN and can be accommodated in a computer with 32 K work memory. The input preparation details, sample problem and the listing of the program are given. (author)

  7. Cold neutron source with self-regulation

    International Nuclear Information System (INIS)

    Kawai, T.

    2003-01-01

    A way to increase the cold neutron flux is to cool moderator from where cold neutrons are extracted. Although various kinds of cooling system are considered, the closed thermo-siphon cooling system is adopted in many institutes. The notable feature of this system is to be able to keep the liquid level stable in the moderator cell against thermal disturbances, by using self-regulation, which allows a stable supply of cold neutrons. The main part of the closed thermo-siphon consists of a condenser, a moderator transfer tube and moderator cell, which is called the hydrogen cold system. When an extra heat load is applied to the hydrogen cold system having no flow resistance in a moderator transfer tube, the system pressure rises by evaporation of liquid hydrogen. Then the boiling point of hydrogen rises. The liquefaction capacity of the condenser is increasing with a rise of temperature, because a refrigerating power of the helium refrigerator increases linearly with temperature rise of the system. Therefore, the effect of thermal heat load increase is compensated and cancelled out. The closed thermo-siphon has this feature generally, when the moderator transfer tube is designed to be no flow resistance. The report reviews the concept of self-regulation, and how to design and construct the cold neutron source with self-regulation. (author)

  8. Development of self-powered neutron detectors for neutron flux monitoring in HCLL and HCPB ITER-TBM

    International Nuclear Information System (INIS)

    Angelone, M.; Klix, A.; Pillon, M.; Batistoni, P.; Fischer, U.; Santagata, A.

    2014-01-01

    Highlights: •Self powered neutron detector (SPND) is attractive neutron monitor for TBM in ITER. •In hard neutron spectra (e.g. TBM) there is the need to optimize their response. •Three state-of-the-art SPNDs were tested using fast and 14 MeV neutrons. •The response of SPNDs is much lower than in thermal neutron flux. •FISPACT calculations performed to find out candidate materials in hard spectra. -- Abstract: Self powered neutron detectors (SPND) have a number of interesting properties (e.g. small dimensions, capability to operate in harsh environments, absence of external bias), so they are attractive neutron monitors for TBM in ITER. However, commercially available SPNDs are optimized for operation in a thermal nuclear reactor where the neutron spectrum is much softer than that expected in a TBM. This fact can limit the use of SPND in a TBM since the effective cross sections for the production of beta emitters are much lower in a fast neutron spectrum. This work represents the first attempt to study SPNDs as neutron flux monitors for TBM. Three state-of-the-art SPND available on the market were bought and tested using fast neutrons at TAPIRO fast neutron source of ENEA Casaccia and with 14 MeV neutrons at the Frascati neutron generator (FNG). The results clearly indicate that in fast neutron spectra, the response of SPNDs is much lower than in thermal neutron flux. Activation calculations were performed using the FISPACT code to find out possible material candidates for SPND suitable for operation in TBM neutron spectra

  9. Development of self-powered neutron detectors for neutron flux monitoring in HCLL and HCPB ITER-TBM

    Energy Technology Data Exchange (ETDEWEB)

    Angelone, M., E-mail: maurizio.angelone@enea.it [Associazione ENEA-EURATOM sulla FusioneENEA C.R. Frascati, Via E. Fermi 45, 00044 Frascati (Italy); Klix, A. [Association KIT-EURATOM, Karlsruhe Institute of Technology (KIT), Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Pillon, M.; Batistoni, P. [Associazione ENEA-EURATOM sulla FusioneENEA C.R. Frascati, Via E. Fermi 45, 00044 Frascati (Italy); Fischer, U. [Association KIT-EURATOM, Karlsruhe Institute of Technology (KIT), Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Santagata, A. [ENEA C.R. Casaccia, via Anguillarese Km. 1,300, 00100 Roma (Italy)

    2014-10-15

    Highlights: •Self powered neutron detector (SPND) is attractive neutron monitor for TBM in ITER. •In hard neutron spectra (e.g. TBM) there is the need to optimize their response. •Three state-of-the-art SPNDs were tested using fast and 14 MeV neutrons. •The response of SPNDs is much lower than in thermal neutron flux. •FISPACT calculations performed to find out candidate materials in hard spectra. -- Abstract: Self powered neutron detectors (SPND) have a number of interesting properties (e.g. small dimensions, capability to operate in harsh environments, absence of external bias), so they are attractive neutron monitors for TBM in ITER. However, commercially available SPNDs are optimized for operation in a thermal nuclear reactor where the neutron spectrum is much softer than that expected in a TBM. This fact can limit the use of SPND in a TBM since the effective cross sections for the production of beta emitters are much lower in a fast neutron spectrum. This work represents the first attempt to study SPNDs as neutron flux monitors for TBM. Three state-of-the-art SPND available on the market were bought and tested using fast neutrons at TAPIRO fast neutron source of ENEA Casaccia and with 14 MeV neutrons at the Frascati neutron generator (FNG). The results clearly indicate that in fast neutron spectra, the response of SPNDs is much lower than in thermal neutron flux. Activation calculations were performed using the FISPACT code to find out possible material candidates for SPND suitable for operation in TBM neutron spectra.

  10. Mechanical performance optimization of neutron shielding material based on short carbon fiber reinforced B4C/epoxy resin

    International Nuclear Information System (INIS)

    Wang Peng; Tang Xiaobin; Chen Feida; Chen Da

    2013-01-01

    To satisfy engineering requirements for mechanics performance of neutron shielding material, short carbon fiber was used to reinforce the traditional containing B 4 C neutron shielding material and effects of fiber content, length and surface treatment to mechanics performance of material was discussed. Based on Americium-Beryllium neutron source, material's neutron shielding performance was tested. The result of experiment prove that tensile strength of material which the quality ratio of resin and fiber is 5:1 is comparatively excellent for 10wt% B 4 C of carbon fiber reinforced epoxy resin. The tensile properties of material change little with the fiber length ranged from 3-10 mm The treatment of fiber surface with silane coupling agent KH-550 can increase the tensile properties of materials by 20% compared with the untreated of that. A result of shielding experiment that the novel neutron shielding material can satisfy the neutron shielding requirements can be obtained by comparing with B 4 C/polypropylene materials. The material has good mechanical properties and wide application prospect. (authors)

  11. Thermophysical Properties of Heat Resistant Shielding Material

    International Nuclear Information System (INIS)

    Porter, W.D.

    2004-01-01

    This project was aimed at determining thermal conductivity, specific heat and thermal expansion of a heat resistant shielding material for neutron absorption applications. These data are critical in predicting the structural integrity of the shielding under thermal cycling and mechanical load. The measurements of thermal conductivity and specific heat were conducted in air at five different temperatures (-31 F, 73.4 F, 140 F, 212 F and 302 F). The transient plane source (TPS) method was used in the tests. Thermal expansion tests were conducted using push rod dilatometry over the continuous range from -40 F (-40 C) to 302 F (150 C)

  12. The Spallation Neutron Source (SNS) conceptual design shielding analysis

    International Nuclear Information System (INIS)

    Johnson, J.O.; Odano, N.; Lillie, R.A.

    1998-03-01

    The shielding design is important for the construction of an intense high-energy accelerator facility like the proposed Spallation Neutron Source (SNS) due to its impact on conventional facility design, maintenance operations, and since the cost for the radiation shielding shares a considerable part of the total facility costs. A calculational strategy utilizing coupled high energy Monte Carlo calculations and multi-dimensional discrete ordinates calculations, along with semi-empirical calculations, was implemented to perform the conceptual design shielding assessment of the proposed SNS. Biological shields have been designed and assessed for the proton beam transport system and associated beam dumps, the target station, and the target service cell and general remote maintenance cell. Shielding requirements have been assessed with respect to weight, space, and dose-rate constraints for operating, shutdown, and accident conditions. A discussion of the proposed facility design, conceptual design shielding requirements calculational strategy, source terms, preliminary results and conclusions, and recommendations for additional analyses are presented

  13. A New Method for Predicting the Penetration and Slowing-Down of Neutrons in Reactor Shields

    Energy Technology Data Exchange (ETDEWEB)

    Hjaerne, L; Leimdoerfer, M

    1965-05-15

    A new approach is presented in the formulation of removal-diffusion theory. The 'removal cross-section' is redefined and the slowing-down between the multigroup diffusion equations is treated with a complete energy transfer matrix rather than in an age theory approximation. The method, based on the new approach contains an adjustable parameter. Examples of neutron spectra and thermal flux penetrations are given in a number of differing shield configurations and the results compare favorably with experiments and Moments Method calculations.

  14. A New Method for Predicting the Penetration and Slowing-Down of Neutrons in Reactor Shields

    International Nuclear Information System (INIS)

    Hjaerne, L.; Leimdoerfer, M.

    1965-05-01

    A new approach is presented in the formulation of removal-diffusion theory. The 'removal cross-section' is redefined and the slowing-down between the multigroup diffusion equations is treated with a complete energy transfer matrix rather than in an age theory approximation. The method, based on the new approach contains an adjustable parameter. Examples of neutron spectra and thermal flux penetrations are given in a number of differing shield configurations and the results compare favorably with experiments and Moments Method calculations

  15. Analysis of coupled neutron-gamma radiations, applied to shieldings in multigroup albedo method

    International Nuclear Information System (INIS)

    Dunley, Leonardo Souza

    2002-01-01

    The principal mathematical tools frequently available for calculations in Nuclear Engineering, including coupled neutron-gamma radiations shielding problems, involve the full Transport Theory or the Monte Carlo techniques. The Multigroup Albedo Method applied to shieldings is characterized by following the radiations through distinct layers of materials, allowing the determination of the neutron and gamma fractions reflected from, transmitted through and absorbed in the irradiated media when a neutronic stream hits the first layer of material, independently of flux calculations. Then, the method is a complementary tool of great didactic value due to its clarity and simplicity in solving neutron and/or gamma shielding problems. The outstanding results achieved in previous works motivated the elaboration and the development of this study that is presented in this dissertation. The radiation balance resulting from the incidence of a neutronic stream into a shielding composed by 'm' non-multiplying slab layers for neutrons was determined by the Albedo method, considering 'n' energy groups for neutrons and 'g' energy groups for gammas. It was taken into account there is no upscattering of neutrons and gammas. However, it was considered that neutrons from any energy groups are able to produce gammas of all energy groups. The ANISN code, for an angular quadrature order S 2 , was used as a standard for comparison of the results obtained by the Albedo method. So, it was necessary to choose an identical system configuration, both for ANISN and Albedo methods. This configuration was six neutron energy groups and eight gamma energy groups, using three slab layers (iron aluminum - manganese). The excellent results expressed in comparative tables show great agreement between the values determined by the deterministic code adopted as standard and, the values determined by the computational program created using the Albedo method and the algorithm developed for coupled neutron

  16. Design of hyper-thermal neutron irradiation fields for neutron capture therapy in KUR-heavy water neutron irradiation facility. Mounting of hyper-thermal neutron converter in therapeutic collimator

    International Nuclear Information System (INIS)

    Sakurai, Y.; Kobayashi, T.

    2001-01-01

    Neutron capture therapy (NCP) using thermal neutron needs to improve of depth dose distribution in a living body. Epi-thermal neutron following moderation of fast neutron is usually used for improving of the depth dose distribution. The moderation method of fast neutron, however, gets mixed some of high energy neutron which give some of serious effects to a living body, and involves the difficulty for collimation of thermal neutron to the diseased part. Hyper-thermal neutrons, which are in an energy range of 0.1-3 eV at high temperature side of thermal neutron, are under consideration for application to the NCP. The hyper-thermal neutrons can be produced by up-scattering of thermal neutron in a high temperature material. Fast neutron components in collimator for the NCP reduce on application of the up-scattering method. Graphite at high temperature (>1000k) is used as a hyper-thermal neutron converter. The hyper-thermal neutron converter is planted to mount on therapeutic collimator which is located at the nearest side of patient for the NCP. Total neutron flux, ratio of hyper-thermal neutron to total neutron, and ratio of gamma-ray dose to neutron flux are calculated as a function of thickness of the graphite converter using monte carlo code MCNP-V4B. (M. Suetake)

  17. Shield design for next-generation, low-neutron-fluence, superconducting tokamaks

    International Nuclear Information System (INIS)

    Lee, V.D.; Gohar, Y.

    1985-01-01

    A shield design using stainless steel (SST), water, boron carbide, lead, and concrete materials was developed for the next-generation tokamak device with superconducting toroidal field (TF) coils and low neutron fluence. A device such as the Tokamak Fusion Core Experiment (TFCX) is representative of the tokamak design which could use this shield design. The unique feature of this reference design is that a majority of the bulk steel in the shield is in the form of spherical balls with two small, flat spots. The balls are purchased from ball-bearing manufacturers and are added as bulk shielding to the void areas of builtup, structural steel shells which form the torus cavity of the plasma chamber. This paper describes the design configuration of the shielding components

  18. Shield design for next-generation, low-neutron-fluence, superconducting tokamaks

    International Nuclear Information System (INIS)

    Lee, V.D.; Gohar, Y.

    1985-01-01

    A shield design using stainless steel (SST), water, boron carbide, lead, and concrete materials was developed for the next-generation tokamak device with superconducting toroidal field (TF) coils and low neutron fluence. A device such as the Tokamak Fusion Core Experiment (TFCX) is representative of the tokamak design which could use this shield design. The unique feature of this reference design is that a majority of the bulk steel in the shield is in the form of spherical balls with two small, flat spots. The balls are purchased from ball-bearing manufacturers and are added as bulk shielding to the void areas of built-up, structural steel shells which form the torus cavity of the plasma chamber. This paper describes the design configuration of the shielding components

  19. 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.))

  20. Radiation transport and shielding information, computer codes, and nuclear data for use in CTR neutronics research and development

    International Nuclear Information System (INIS)

    Santoro, R.T.; Maskewitz, B.F.; Roussin, R.W.; Trubey, D.K.

    1976-01-01

    The activities of the Radiation Shielding Information Center (RSIC) of the Oak Ridge National Laboratory are being utilized in support of fusion reactor technology. The major activities of RSIC include the operation of a computer-based information storage and retrieval system, the collection, packaging, and distribution of large computer codes, and the compilation and dissemination of processed and evaluated data libraries, with particular emphasis on neutron and gamma-ray cross-section data. The Center has acquired thirteen years of experience in serving fission reactor, weapons, and accelerator shielding research communities, and the extension of its technical base to fusion reactor research represents a logical progression. RSIC is currently working with fusion reactor researchers and contractors in computer code development to provide tested radiation transport and shielding codes and data library packages. Of significant interest to the CTR community are the 100 energy group neutron and 21 energy group gamma-ray coupled cross-section data package (DLC-37) for neutronics studies, a comprehensive 171 energy group neutron and 36 energy group gamma-ray coupled cross-section data base with retrieval programs, including resonance self-shielding, that are tailored to CTR application, and a data base for the generation of energy-dependent atomic displacement and gas production cross sections and heavy-particle-recoil spectra for estimating radiation damage to CTR structural components. Since 1964, the Center has been involved in the international exchange of information, encouraged and supported by both government and interagency agreements; and to achieve an equally viable and successful program in fusion research, the reciprocal exchange of CTR data and computing technology is encouraged and welcomed

  1. A Wavelet-Based Finite Element Method for the Self-Shielding Issue in Neutron Transport

    International Nuclear Information System (INIS)

    Le Tellier, R.; Fournier, D.; Ruggieri, J. M.

    2009-01-01

    This paper describes a new approach for treating the energy variable of the neutron transport equation in the resolved resonance energy range. The aim is to avoid recourse to a case-specific spatially dependent self-shielding calculation when considering a broad group structure. This method consists of a discontinuous Galerkin discretization of the energy using wavelet-based elements. A Σ t -orthogonalization of the element basis is presented in order to make the approach tractable for spatially dependent problems. First numerical tests of this method are carried out in a limited framework under the Livolant-Jeanpierre hypotheses in an infinite homogeneous medium. They are mainly focused on the way to construct the wavelet-based element basis. Indeed, the prior selection of these wavelet functions by a thresholding strategy applied to the discrete wavelet transform of a given quantity is a key issue for the convergence rate of the method. The Canuto thresholding approach applied to an approximate flux is found to yield a nearly optimal convergence in many cases. In these tests, the capability of such a finite element discretization to represent the flux depression in a resonant region is demonstrated; a relative accuracy of 10 -3 on the flux (in L 2 -norm) is reached with less than 100 wavelet coefficients per group. (authors)

  2. Application of a simple analytical model to estimate effectiveness of radiation shielding for neutrons

    International Nuclear Information System (INIS)

    Frankle, S.C.; Fitzgerald, D.H.; Hutson, R.L.; Macek, R.J.; Wilkinson, C.A.

    1993-01-01

    Neutron dose equivalent rates have been measured for 800-MeV proton beam spills at the Los Alamos Meson Physics Facility. Neutron detectors were used to measure the neutron dose levels at a number of locations for each beam-spill test, and neutron energy spectra were measured for several beam-spill tests. Estimates of expected levels for various detector locations were made using a simple analytical model developed for 800-MeV proton beam spills. A comparison of measurements and model estimates indicates that the model is reasonably accurate in estimating the neutron dose equivalent rate for simple shielding geometries. The model fails for more complicated shielding geometries, where indirect contributions to the dose equivalent rate can dominate

  3. Utilizing the slowing-down-time technique for benchmarking neutron thermalization in graphite

    International Nuclear Information System (INIS)

    Zhou, T.; Hawari, A. I.; Wehring, B. W.

    2007-01-01

    Graphite is the moderator/reflector in the Very High Temperature Reactor (VHTR) concept of Generation IV reactors. As a thermal reactor, the prediction of the thermal neutron spectrum in the VHTR is directly dependent on the accuracy of the thermal neutron scattering libraries of graphite. In recent years, work has been on-going to benchmark and validate neutron thermalization in 'reactor grade' graphite. Monte Carlo simulations using the MCNP5 code were used to design a pulsed neutron slowing-down-time experiment and to investigate neutron slowing down and thermalization in graphite at temperatures relevant to VHTR operation. The unique aspect of this experiment is its ability to observe the behavior of neutrons throughout an energy range extending from the source energy to energies below 0.1 eV. In its current form, the experiment is designed and implemented at the Oak Ridge Electron Linear Accelerator (ORELA). Consequently, ORELA neutron pulses are injected into a 70 cm x 70 cm x 70 cm graphite pile. A furnace system that surrounds the pile and is capable of heating the graphite to a centerline temperature of 1200 K has been designed and built. A system based on U-235 fission chambers and Li-6 scintillation detectors surrounds the pile. This system is coupled to multichannel scaling instrumentation and is designed for the detection of leakage neutrons as a function of the slowing-down-time (i.e., time after the pulse). To ensure the accuracy of the experiment, careful assessment was performed of the impact of background noise (due to room return neutrons) and pulse-to-pulse overlap on the measurement. Therefore, the entire setup is surrounded by borated polyethylene shields and the experiment is performed using a source pulse frequency of nearly 130 Hz. As the basis for the benchmark, the calculated time dependent reaction rates in the detectors (using the MCNP code and its associated ENDF-B/VI thermal neutron scattering libraries) are compared to measured

  4. Calculation And Design Of A New Configuration For Radiation Shielding At Neutron Beam No.3 For Fundamental And Applied Researches

    International Nuclear Information System (INIS)

    Vuong Huu Tan; Tran Tuan Anh; Nguyen Kien Cuong; Nguyen Canh Hai; Nguyen Xuan Hai; Pham Ngoc Son; Ho Huu Thang

    2011-01-01

    The tangential horizontal channel of No. 3 of the Dalat Research Reactor has been opened and used during the 1990s. The utilizations of the thermal neutron beam at this channel were the Neutron Radiography and the Prompt Gamma Neutron Activation Analysis method (PGNAA). At present, the neutron beam used for nuclear structure data researches based on the Summing of Amplitude Coincident Pulses system (SACP). Beside, several related research equipments have been set up and operated for the research purposes. A renovation of the neutron channel, therefore, will play an important role in safe and effective utilizations of the neutron beam in fields of nuclear physic training and researches. A new configuration for radiation shielding has been simulated by MCNP code. The calculated results of dose rates for neutron and gamma at working positions are in range of dose rate limit. (author)

  5. Attenuation of fast neutron in concretes for biological shielding

    International Nuclear Information System (INIS)

    Labrada, A.; Chavez, A.; Gonzalez Mateu, D.; Desdin, F.; Tenjeiro, J.I.; Tellez, E.

    1993-01-01

    The attenuation of neutrons emitted by an 10 6 n/s. Am-Be source, in concretes elaborated with different aggregates is discussed in this paper. Two measurement methods were used an dosimetric system with Bonner spheres and 6 LiI(Eu) detector, and LAVSAN dielectric nuclear track detectors - with 238 U converts. The concretes elaborated with magnetite is reported as the best for neutron shielding while the Bauxite is not advisable for this purpose

  6. Up-dating of the RA-0 reactor shielding. Gamma and neutron isodoses

    International Nuclear Information System (INIS)

    Murua, Carlos A.; Chautemps, Norma A.; Ackerley, Alejandro F.; Alexeiew, Vladimiro

    1999-01-01

    A comparative analysis of the historical shielding configurations of the RA-0 reactor is performed and the comparison methodology is described. The gamma and neutron dose mapping of the last two stages of the reactor shielding has been carried out and the results are analysed

  7. Monte Carlo simulations of a D-T neutron generator shielding for landmine detection

    International Nuclear Information System (INIS)

    Reda, A.M.

    2011-01-01

    Shielding for a D-T sealed neutron generator has been designed using the MCNP5 Monte Carlo radiation transport code. The neutron generator will be used in field for the detection of explosives, landmines, drugs and other 'threat' materials. The optimization of the detection of buried objects was started by studying the signal-to-noise ratio for different geometric conditions. - Highlights: → A landmine detection system based on neutron fast/slow analysis has been designed. → Shielding for a D-T sealed neutron generator tube has been designed using Monte Carlo radiation transport code. → Detection of buried objects was started by studying the signal-to-noise ratio for different geometric conditions. → The signal-to-background ratio optimized at one position for all depths.

  8. Dismantling system of concrete thermal shielding walls

    International Nuclear Information System (INIS)

    Machida, Nobuhiro; Saiki, Yoshikuni; Ono, Yorimasa; Tokioka, Masatake; Ogino, Nobuyuki.

    1985-01-01

    Purpose: To enable safety and efficient dismantling of concrete thermal shielding walls in nuclear reactors. Method: Concrete thermal shielding walls are cut and dismantled into dismantled blocks by a plasma cutting tool while sealing the top opening of bioshielding structures. The dismantled blocks are gripped and conveyed. The cutting tool is remote-handled while monitoring on a television receiver. Slugs and dusts produced by cutting are removed to recover. Since the dismantling work is carried out while sealing the working circumstance and by the remote control of the cutting tool, the operators' safety can be secured. Further, since the thermal sealing walls are cut and dismantled into blocks, dismantling work can be done efficiently. (Moriyama, K.)

  9. Neutron dosimetry program at Mound - problems and solutions

    International Nuclear Information System (INIS)

    Winegardner, M.K.

    1991-01-01

    The Mound personnel neutron dosimetry program utilizes TLD albedo technology. The neutron dosimeter design incorporates a two-element spectrometer for site-specific neutron quality determination and empirical application of field neutron calibration factors. Design elements feature two Li(6)F (TLD- 600) chips for neutron detection and one Li(7)F (TLD-700) chip for gamma compensation of the TLD- 600 chips. One TLD-600 chip is Cadmium shielded on the front side of the dosimeter, the other is Cadmium shielded from the back side. Tin filters are placed opposite of the Cadmium shield on each of the TLD-600 chips and on both sides of the TLD-700 chip for symmetrically equivalent gamma absorption characteristics. Neutron quality determination is accomplished by the albedo neutron-to- incident thermal neutron response ratio above the Cadmium cutoff. This front Cadmium shielded-to-back Cadmium shielded response ratio, compensated for the presence of gamma radiation, provides the basis for neutron energy calibration via the albedo response curve

  10. Equivalent-spherical-shield neutron dose calculations

    International Nuclear Information System (INIS)

    Russell, G.J.; Robinson, H.

    1988-01-01

    Neutron doses through 162-cm-thick spherical shields were calculated to be 1090 and 448 mrem/h for regular and magnetite concrete, respectively. These results bracket the measured data, for reinforced regular concrete, of /approximately/600 mrem/h. The calculated fraction of the high-energy (>20 MeV) dose component also bracketed the experimental data. The measured and calculated doses were for a graphite beam stop bombarded with 100 nA of 800-MeV protons. 6 refs., 2 figs., 1 tab

  11. Design and characterization of a novel neutron shield for BNCT in an experimental model of oral cancer in the hamster cheek pouch at RA-3

    International Nuclear Information System (INIS)

    Pozzi, E.C.C.; Curotto, P.; Monti Hughes, A.; Nigg, D.W.; Schwint, A.E.; Trivillin, V.A.; Thorp, S.I.

    2013-01-01

    Our research group at the Radiation Pathology Division of the Department of Radiobiology (National Atomic Energy Commission) has previously demonstrated the therapeutic efficacy of different BNCT protocols to treat oral cancer in an experimental hamster cheek pouch model. In particular, to perform studies in this experimental model at the thermal facility constructed at RA-3, we designed and constructed a shielding device for thermal neutrons, to be able to expose the cheek pouch while minimizing the dose to the rest of the body. This device allowed for the irradiation of one animal at a time. Given the usage rate of the device, the aim of the present study was to design and construct an optimized version of the existing shielding device that would allow for the simultaneous irradiation of 2 animals at the thermal facility of RA-3. Taking into account the characteristics of the neutron source and preliminary biological assays, we designed the shielding device for the body of the animal, i.e. a rectangular shaped box with double acrylic walls. The space between the walls contains a continuous filling of 6Li 2 CO 3 (95% enriched in 6Li), approximately 6 mm thick. Two small windows interrupt the shield at one end of the box through which the right pouch of each hamster is everted out onto an external acrylic shelf for exposure to the neutron flux. The characterization of the shielding device showed that the neutron flux was equivalent at both irradiation positions confirming that we were able to design and construct a new shielding device that allows for the irradiation of 2 animals at the same time at the thermal facility of RA-3. This new version of the shielding device will reduce the number of interventions of the reactor operators, reducing occupational exposure to radiation and will make the procedure more efficient for researchers. In addition, we addressed the generation of tritium as a product of the capture reaction in lithium. It was considered as a

  12. Comparison of MCNP4C and experimental results on neutron and gamma ray shielding effects for materials

    Energy Technology Data Exchange (ETDEWEB)

    Cha, Kyoon Ho; Lee, Eun Ki [KEPRI, Taejon (Korea, Republic of)

    2004-07-01

    MCNP code is a general-purpose Monte Carlo radiation transport code that can numerically simulate neutron, photon, and electron transport. Increasing the speed of computing machine is making numerical transport simulation more attractive and has led to the widespread use of such code. This code can be used for general radiation shielding and criticality accident alarm system related dose calculations, so that the version 4C2 of this code was used to evaluate the shielding effect against neutron and gamma ray experiments. The Ueki experiments were used for neutron shielding effects for materials, and the Kansas State University (KSU) photon skyshine experiments of 1977 were tested for gamma ray shielding effects.

  13. Neutron shielding studies on an advanced molten salt fast reactor design

    International Nuclear Information System (INIS)

    Merk, Bruno; Konheiser, Jörg

    2014-01-01

    Highlights: • Material damage due to irradiation has already been discovered at the MSRE. • Neutronic analysis of MSFR with curved blanket wall geometry. • Neutron fluence limit at the wall of the outer vessel can be kept for 80 years. • Shielded MSFR core will be of same dimension than a SFR core. - Abstract: The molten salt reactor technology has gained some new interest. In contrast to the historic molten salt reactors, the current projects are based on designing a molten salt fast reactor. Thus the shielding becomes significantly more challenging than in historic concepts. One very interesting and innovative result of the most recent EURATOM project on molten salt reactors – EVOL – is the fluid flow optimized design of the inner reactor vessel using curved blanket walls. The developed structure leads to a very uniform flow distribution. The design avoids all internal structures. Based on this new geometry a model for neutron physics calculation is presented. The major steps are: the modeling of the curved geometry in the unstructured mesh neutron transport code HELIOS and the determination of the real neutron flux and power distribution for this new geometry. The developed model is then used for the determination of the neutron fluence distribution in the inner and outer wall of the system. Based on these results an optimized shielding strategy is developed for the molten salt fast reactor to keep the fluence in the safety related outer vessel below expected limit values. A lifetime of 80 years can be assured, but the size of the core/blanket system will be comparable to a sodium cooled fast reactor. The HELIOS results are verified against Monte-Carlo calculations with very satisfactory agreement for a deep penetration problem

  14. Bench-mark experiments to study the neutron distribution in a heterogeneous reactor shielding

    International Nuclear Information System (INIS)

    Bolyatko, V.V.; Vyrskij, M.Yu.; Mashkovich, V.P.; Nagaev, R.Kh.; Prit'mov, A.P.; Sakharov, V.K.; Troshin, V.S.; Tikhonov, E.G.

    1981-01-01

    The bench-mark experiments performed at the B-2 facility of the BR-10 reactor to investigate the spatial and energy neutron distributions are described. The experimental facility includes the neutron beam channel with a slide, a mo shielding composition investigated consisted of sequential layers of steel (1KH18N9T) and graphite slabs. The neutron spectra were measured by activation method, a set of treshold and resonance detectors having been used. The detectors made it possible to obtain the absolute neutron spectra in the 1.4 eV-10 MeV range. The comparison of calculations with the results of the bench-mark experiments made it possible to prove the neutron transport calculational model realized in the ROZ-9 and ARAMAKO-2F computer codes and evaluate the validity of the ARAMAKO constants for the class of shielding compositions in question [ru

  15. The TRIUMF thermal neutron facility as planned for operation by 1978

    International Nuclear Information System (INIS)

    Arrott, A.S.; Templeton, T.L.; Thorson, I.M.; Blaby, R.E.; Burgerjon, J.J.

    1977-08-01

    The concepts of the thermal neutron facility have been considerably modified since they were first put forth in 1971. The move has been toward simplification. This report describes the basic vacuum tank structure, its surrounding steel shielding and the concrete structure. The vacuum tank contains a target, moderator and reflector and has ports for the extraction of thermal neutron beams. It also has capabilities for producing mesons and for irradiation of targets in the primary proton beam. The system has been designed with flexibility for modification to meet possible future demands for irradiation facilities, radiography, or pulsed operation. The targets can be easily changed, and it is planned to do this to meet the heat transfer problems as they arise on going to higher beam currents. Feasibility studies for Pb-Bi and Pb targets have been carried out. The Pb target was chosen because of safety considerations and simpler design. (author)

  16. Pulsed thermal neutron source at the fast neutron generator.

    Science.gov (United States)

    Tracz, Grzegorz; Drozdowicz, Krzysztof; Gabańska, Barbara; Krynicka, Ewa

    2009-06-01

    A small pulsed thermal neutron source has been designed based on results of the MCNP simulations of the thermalization of 14 MeV neutrons in a cluster-moderator which consists of small moderating cells decoupled by an absorber. Optimum dimensions of the single cell and of the whole cluster have been selected, considering the thermal neutron intensity and the short decay time of the thermal neutron flux. The source has been built and the test experiments have been performed. To ensure the response is not due to the choice of target for the experiments, calculations have been done to demonstrate the response is valid regardless of the thermalization properties of the target.

  17. Fail-safe neutron shutter used for thermal neutron radiography

    International Nuclear Information System (INIS)

    Sachs, R.D.; Morris, R.A.

    1976-11-01

    A fail-safe, reliable, easy-to-use neutron shutter was designed, built, and put into operation at the Omega West Reactor, Los Alamos Scientific Laboratory. The neutron shutter will be used primarily to perform thermal neutron radiography, but is also available for a highly collimated source of thermal neutrons [neutron flux = 3.876 x 10 6 (neutrons)/(cm 2 .s)]. Neutron collimator sizes of either 10.16 by 10.16 cm or 10.16 by 30.48 cm are available

  18. Neutron shielding verification measurements and simulations for a 235-MeV proton therapy center

    International Nuclear Information System (INIS)

    Newhauser, W.D.; Titt, U.; Dexheimer, D.; Yan, X.; Nill, S.

    2002-01-01

    The neutron shielding at the Massachusetts General Hospital's 235-MeV proton therapy facility was investigated with measurements, analytical calculations, and realistic three-dimensional Monte Carlo simulations. In 37 of 40 cases studied, the analytical calculations predicted higher neutron dose equivalent rates outside the shielding than the measured, typically by more than a factor of 10, and in some cases more than 100. Monte Carlo predictions of dose equivalent at three locations are, on average, 1.1 times the measured values. Except at one location, all of the analytical model predictions and Monte Carlo simulations overestimate neutron dose equivalent

  19. Measurement of neutron sensitivity of self powered neutron detectors

    International Nuclear Information System (INIS)

    Mahant, A.K.; Yeshuraja, V.; Ghodke, Shobha

    2005-01-01

    Self powered neutron detectors (SPNDs ) will form the part of Reactor Instrumentation in the upcoming 500 MWe power reactors. ECIL has developed Vanadium and Cobalt SPNDs for NPCIL to be used in regulation and protection channels. Experimental determination of neutron sensitivity of the vanadium and cobalt Self Powered Neutron Detectors (SPNDs) was carried out in A-l location of Apsara reactor at BARC. The measurements involved determination of total detector signal, its various components and the thermal neutron flux at the detector location. The paper describes the experimental techniques used to measure various parameters required to evaluate the neutron sensitivity of the SPNDs and also the parameters required to ascertain the integrity of SPNDs. Neutron flux measurement was done by gold foil irradiation technique. The predominant signal component from the vanadium SPND is Ib the current due to activation of the vanadium emitter, it forms about 85% of the total signal. The other components I n,γ due to the capture gamma rays of 52 V and I externalγ produced by the external reactor gamma rays contribute about 10% and 5% respectively to the total signal. Whereas in the cobalt SPND the main signal component is due to the capture gamma rays of 60 Co and accounts for about the 95% of the total signal. Remaining 5% signal is due to external reactor gamma rays. (author)

  20. Development and application of high performance liquid shielding materials

    International Nuclear Information System (INIS)

    Miura, Toshimasa; Omata, Sadao; Otano, Naoteru; Hirao, Yoshihiro; Kanai, Yasuji

    1998-01-01

    Development of liquid shielding material with good performance for neutron and γ-ray was investigated. Lead, hydrogen and boron were selected as the elements of shielding materials which were made by the ultraviolet curing method. Good performance shielding materials with about 1 mm width to neutron and gamma ray were produced by mixing lead, boron compound and ultraviolet curing monomer with many hydrogens. The shielding performance was the same as a concrete with two times width. The activation was very small such as 1/10 6 -1/10 8 of the standard concrete. The weight and the external appearance did not charged from room temperature to 100degC. Polyfunctional monomer had good thermal resistance. This shielding material was applied to double bending cylindrical duct and annulus ring duct. The results proved the shielding materials developed had good performance. (S.Y.)

  1. Evaluation of neutron shielding properties of lead glass using bubble detector

    International Nuclear Information System (INIS)

    Viswanathan, S.; Vishwa Prasad, K.; Srinivasan, T.K.; Ponraju, D.

    1999-01-01

    Neutron shielding properties of lead glass had been studied using a 241 Am-Be neutron source. Indigenously developed bubble detector was used as neutron detector. Attenuation curves were determined experimentally for the lead glass under the conditions of broad beam geometry. Theoretical calculations were made using Monte Carlo code MCNP3. Measurements were made for polyethylene and concrete to serve as reference. The measured and calculated neutron removal cross sections of lead glass, polyethylene and concrete are reported in this paper. Good agreement is observed between the experimental results and theoretical calculations. (author)

  2. Using FLUKA to Study Concrete Square Shield Performance in Attenuation of Neutron Radiation Produced by APF Plasma Focus Neutron Source

    Science.gov (United States)

    Nemati, M. J.; Habibi, M.; Amrollahi, R.

    2013-04-01

    In 2010, representatives from the Nuclear Engineering and physics Department of Amirkabir University of Technology (AUT) requested development of a project with the objective of determining the performance of a concrete shield for their Plasma Focus as neutron source. The project team in Laboratory of Nuclear Engineering and physics department of Amirkabir University of Technology choose some shape of shield to study on their performance with Monte Carlo code. In the present work, the capability of Monte Carlo code FLUKA will be explored to model the APF Plasma Focus, and investigating the neutron fluence on the square concrete shield in each region of problem. The physical models embedded in FLUKA are mentioned, as well as examples of benchmarking against future experimental data. As a result of this study suitable thickness of concrete for shielding APF will be considered.

  3. A shielding design for an accelerator-based neutron source for boron neutron capture therapy

    Energy Technology Data Exchange (ETDEWEB)

    Hawk, A.E.; Blue, T.E. E-mail: blue.1@osu.edu; Woollard, J.E

    2004-11-01

    Research in boron neutron capture therapy (BNCT) at The Ohio State University Nuclear Engineering Department has been primarily focused on delivering a high quality neutron field for use in BNCT using an accelerator-based neutron source (ABNS). An ABNS for BNCT is composed of a proton accelerator, a high-energy beam transport system, a {sup 7}Li target, a target heat removal system (HRS), a moderator assembly, and a treatment room. The intent of this paper is to demonstrate the advantages of a shielded moderator assembly design, in terms of material requirements necessary to adequately protect radiation personnel located outside a treatment room for BNCT, over an unshielded moderator assembly design.

  4. Development of Neutron and Photon Shielding Calculation System for Workstation (NPSS-W)

    International Nuclear Information System (INIS)

    Shimizu, Yoshio; Nojiri, Ichiro; Odajima, Akira; Sasaki, Toshihisa; Kurosawa, Naohiro

    1998-01-01

    In plant designs and safety evaluations of nuclear fuel cycle facilities, it is important to evaluate the direct radiation and the skyshine (air-scattered photon radiation) from facilities reasonably. The Neutron and Photon Shielding Calculation System for Workstation (NPSS-W) was developed. The NPSS-W can carry out the shielding calculations of the photon and the neutron easily and rapidly. The NPSS-W can easily calculate the radiation source intensity by ORIGEN-S and the dose equivalent rate by SN transport calculational codes, which are ANISN and DOT3.5. The NPSS-W consists of five modules, which named CAL1, CAL2, CAL3, CAL4, CAL5). Some kinds of shielding calculational systems are calculated. The user's manual of NPSS-W, the examples of calculations for each module and the output data are appended. (author)

  5. Neutron transmission benchmark problems for iron and concrete shields in low, intermediate and high energy proton accelerator facilities

    Energy Technology Data Exchange (ETDEWEB)

    Nakane, Yoshihiro; Sakamoto, Yukio [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Hayashi, Katsumi [and others

    1996-09-01

    Benchmark problems were prepared for evaluating the calculation codes and the nuclear data for accelerator shielding design by the Accelerator Shielding Working Group of the Research Committee on Reactor Physics in JAERI. Four benchmark problems: transmission of quasi-monoenergetic neutrons generated by 43 MeV and 68 MeV protons through iron and concrete shields at TIARA of JAERI, neutron fluxes in and around an iron beam stop irradiated by 500 MeV protons at KEK, reaction rate distributions inside a thick concrete shield irradiated by 6.2 GeV protons at LBL, and neutron and hadron fluxes inside an iron beam stop irradiated by 24 GeV protons at CERN are compiled in this document. Calculational configurations and neutron reaction cross section data up to 500 MeV are provided. (author)

  6. Thermal neutron flux distribution in ET-RR-2 reactor thermal column

    Directory of Open Access Journals (Sweden)

    Imam Mahmoud M.

    2002-01-01

    Full Text Available The thermal column in the ET-RR-2 reactor is intended to promote a thermal neutron field of high intensity and purity to be used for following tasks: (a to provide a thermal neutron flux in the neutron transmutation silicon doping, (b to provide a thermal flux in the neutron activation analysis position, and (c to provide a thermal neutron flux of high intensity to the head of one of the beam tubes leading to the room specified for boron thermal neutron capture therapy. It was, therefore, necessary to determine the thermal neutron flux at above mentioned positions. In the present work, the neutron flux in the ET-RR-2 reactor system was calculated by applying the three dimensional diffusion depletion code TRITON. According to these calculations, the reactor system is composed of the core, surrounding external irradiation grid, beryllium block, thermal column and the water reflector in the reactor tank next to the tank wall. As a result of these calculations, the thermal neutron fluxes within the thermal column and at irradiation positions within the thermal column were obtained. Apart from this, the burn up results for the start up core calculated according to the TRITION code were compared with those given by the reactor designer.

  7. Sensitivity Calculation of Vanadium Self-Powered Neutron Detector

    International Nuclear Information System (INIS)

    Cha, Kyoon Ho

    2011-01-01

    Self-powered neutron detector (SPND) is being widely used to monitor the reactor core of the nuclear power plants. The SPND contains a neutron-sensitive metallic emitter surrounded by a ceramic insulator. Currently, the rhodium SPND has been used in many nuclear power plants. The lifetime of rhodium is too short (about 3∼5 years) to operate the nuclear power plant economically. The vanadium (V) SPND is also primarily sensitive to neutrons like rhodium, but is a somewhat slower reaction time as that of a rhodium SPND. The benefit of vanadium over rhodium is its low depletion rate, which is a factor of 7 times less than that of rhodium. For this reason, a vanadium SPND has been being developed to replace the rhodium SPND which is used in OPR1000. Some Monte Carlo simulations were accomplished to calculate the initial sensitivity of vanadium emitter material and alumina (Al 2 O 3 ) insulator with a cylindrical geometry. An MCNP-X code was used to simulate some factors (neutron self shielding factor and electron escape probability from the emitter) necessary to calculate the sensitivity of vanadium detector. The simulation results were compared with some theoretical and experimental values. The method presented here can be used to analyze the optimum design of the vanadium SPND

  8. Combination thermal and radiation shield for well logging apparatus

    International Nuclear Information System (INIS)

    Wilson, B.F.

    1984-01-01

    A device for providing both thermal protection and radiation shielding for components such as radiation detectors within a well logging instrument comprises a thermally insulative flask containing a weldment filled with a mass of eutectic material which undergoes a change of state e.g. melting at a temperature which will provide an acceptable thermal environment for such components for extended time periods. The eutectic material which is preferably a bismuth (58%)/tin (42%) alloy has a specific gravity (> 8.5) facilitating its use as a radiation shield and is distributed around the radiation detectors so as to selectively impede the impinging of the detectors by radiation. The device is incorporated in a skid of a well logging instrument for measuring γ backscatter. A γ source is located either above or within the protective shielding. (author)

  9. Program GROUPIE (version 79-1): calculation of Bondarenko self-shielded neutron cross sections and multiband parameters from data in the ENDF/B format

    International Nuclear Information System (INIS)

    Cullen, D.E.

    1980-01-01

    Program GROUPIE reads evaluated data in the ENDF/B format and uses these data to calculate Bondarenko self-shielded cross sections and multiband parameters. To give as much generality as possible, the program allows the user to specify arbitrary energy groups and an arbitrary energy groups and an arbitrary energy-dependent neutron spectrum (weighing function). To guarantee the accuracy of the results, all integrals are performed analytically; in no case is iteration or any approximate form of integration used. The output from this program includes both listings and multiband parameters suitable for use either in a normal multigroup transport calculation or in a multiband transport calculation. A listing of the source deck is available on request

  10. 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)

  11. A Sensitivity Study on the Radiation Shield of KSPR Space Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Cerba, S.; Lee, Hyun Chul; Lim, Hong Sik; Noh, Jae Man [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    The idea of a space reactor was realised some decades ago and since that time several research activities have been performed into this field. The US National Aeronautics and Space Administration (NASA) has been developing a small fast reactor called as fission power system (FPS) for deep space mission, where highly enriched uranium (HEU) is used as fuel. On the other hand, other researchers have also surveyed a thermal reactor concept with low enriched uranium (LEU) for space applications. One of the main concerns in terms of a space reactor is the total size and the mass of the system including the reactor itself as well as the radiation shield. Since the reactor core is a source of neutrons and gamma photons of various energies, which may cause severe damage on the electronics of the space stations, the questions related to the development of a radiation shield should be address appropriately. The proposal of a radiation shield for a small space reactor is discussed in this paper. The requirements for the radiation shield have been addressed in terms of maximal absorbed doses and neutron flounces during 10 years of operation. In this study a radiation shield design for a small space reactor was investigated. All the presented calculations were performed using the multi-purpose stochastic MCNP code with temperature dependent continuous energy ENDF/B VII.0 neutron and photon cross section libraries. The aim of this study was to design a neutron and gamma shield that can meet the requirements of 250 Gy absorbed during 10 years of reactor operation. The comparison with a fast reactor design showed that high content of {sup 238}U strongly influences the shielding mass. This phenomenon is due to the higher photon production in case of the KSPR design and therefore the use of high {sup 235}U enrichments and the operation in fast neutron spectrum may be more desirable. In case if the KSPR space reactor the best shielding performance was achieved while utilizing a multi

  12. Neutron activation analysis of archaeological artifacts using the conventional relative method: a realistic approach for analysis of large samples

    International Nuclear Information System (INIS)

    Bedregal, P.S.; Mendoza, A.; Montoya, E.H.; Cohen, I.M.; Universidad Tecnologica Nacional, Buenos Aires; Oscar Baltuano

    2012-01-01

    A new approach for analysis of entire potsherds of archaeological interest by INAA, using the conventional relative method, is described. The analytical method proposed involves, primarily, the preparation of replicates of the original archaeological pottery, with well known chemical composition (standard), destined to be irradiated simultaneously, in a well thermalized external neutron beam of the RP-10 reactor, with the original object (sample). The basic advantage of this proposal is to avoid the need of performing complicated effect corrections when dealing with large samples, due to neutron self shielding, neutron self-thermalization and gamma ray attenuation. In addition, and in contrast with the other methods, the main advantages are the possibility of evaluating the uncertainty of the results and, fundamentally, validating the overall methodology. (author)

  13. Self-shielding characteristics of aqueous self-cooled blankets for next generation fusion devices

    International Nuclear Information System (INIS)

    Pelloni, S.; Cheng, E.T.; Embrechts, M.J.

    1987-11-01

    The present study examines self-shielding characteristics for two aqueous self-cooled tritium producing driver blankets for next generation fusion devices. The aqueous Self-Cooled Blanket concept (ASCB) is a very simple blanket concept that relies on just structural material and coolant. Lithium compounds are dissolved in water to provide for tritium production. An ASCB driver blanket would provide a low technology and low temperature environment for blanket test modules in a next generation fusion reactor. The primary functions of such a blanket would be shielding, energy removal and tritium production. One driver blanket considered in this study concept relates to the one proposed for the Next European Torus (NET), while the second concept is indicative for the inboard shield design for the Engineering Test Reactor proposed by the USA (TIBER II/ETR). The driver blanket for NET is based on stainless steel for the structural material and aqueous solution, while the inboard shielding blanket for TIBER II/ETR is based on a tungsten/aqueous solution combination. The purpose of this study is to investigate self-shielding and heterogeneity effects in aqueous self-cooled blankets. It is found that no significant gains in tritium breeding can be achieved in the stainless steel blanket if spatial and energy self-shielding effects are considered, and the heterogeneity effects are also insignificant. The tungsten blanket shows a 5 percent increase in tritium production in the shielding blanket when energy and spatial self-shielding effects are accounted for. However, the tungsten blanket shows a drastic increase in the tritium breeding ratio due to heterogeneity effects. (author) 17 refs., 9 figs., 9 tabs

  14. Measurement and simulation of thermal neutron flux distribution in the RTP core

    Science.gov (United States)

    Rabir, Mohamad Hairie B.; Jalal Bayar, Abi Muttaqin B.; Hamzah, Na'im Syauqi B.; Mustafa, Muhammad Khairul Ariff B.; Karim, Julia Bt. Abdul; Zin, Muhammad Rawi B. Mohamed; Ismail, Yahya B.; Hussain, Mohd Huzair B.; Mat Husin, Mat Zin B.; Dan, Roslan B. Md; Ismail, Ahmad Razali B.; Husain, Nurfazila Bt.; Jalil Khan, Zareen Khan B. Abdul; Yakin, Shaiful Rizaide B. Mohd; Saad, Mohamad Fauzi B.; Masood, Zarina Bt.

    2018-01-01

    The in-core thermal neutron flux distribution was determined using measurement and simulation methods for the Malaysian’s PUSPATI TRIGA Reactor (RTP). In this work, online thermal neutron flux measurement using Self Powered Neutron Detector (SPND) has been performed to verify and validate the computational methods for neutron flux calculation in RTP calculations. The experimental results were used as a validation to the calculations performed with Monte Carlo code MCNP. The detail in-core neutron flux distributions were estimated using MCNP mesh tally method. The neutron flux mapping obtained revealed the heterogeneous configuration of the core. Based on the measurement and simulation, the thermal flux profile peaked at the centre of the core and gradually decreased towards the outer side of the core. The results show a good agreement (relatively) between calculation and measurement where both show the same radial thermal flux profile inside the core: MCNP model over estimation with maximum discrepancy around 20% higher compared to SPND measurement. As our model also predicts well the neutron flux distribution in the core it can be used for the characterization of the full core, that is neutron flux and spectra calculation, dose rate calculations, reaction rate calculations, etc.

  15. Neutron shielding material and a process for producing the same

    International Nuclear Information System (INIS)

    Tadokoro, S.; Segawa, H.

    1982-01-01

    A neutron shielding material comprises a polymerization product of a monomer mixture of an alkyl methacrylate or styrene and a boric acid containing a polyol constituent. Such a material may be formed into transparent sheets with high mechanical strength

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-05-01

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

  17. Investigating spatial self-shielding and temperature effects for homogeneous and double heterogeneous pebble models with MCNP

    International Nuclear Information System (INIS)

    Li, J.; Nuenighoff; Pohl, C.; Allelein, H.J.

    2010-01-01

    The gas-cooled, high temperature reactor (HTR) represents a valuable option for the future development of nuclear technology, because of its excellent safety features. One main safety feature is the negative temperature coefficient which is due to the Doppler broadening of the (n,y) resonance absorption cross section. A second important effect is the spatial self-shielding due to the double heterogeneous geometry of a pebble bed reactor. At FZ-Juelich two reactor analysis codes have been developed: VSOP for core design and MGT for transient analysis. Currently an update of the nuclear cross section libraries to ENDF/B-VII.0 of both codes takes place. In order to take the temperature dependency as well as the spatial self-shielding into account the absorption cross sections σ (n,y) for the resonance absorbers like 232 Th and 238 U have to be provided as function of incident neutron energy, temperature and nuclide concentration. There are two reasons for choosing the Monte-Carlo approach to calculate group wise cross sections. First, the former applied ZUT-DGL code to generate the resonance cross section tables for MGT is so far not able to handle the new resonance description based on Reich-Moore instead of Single-level Breit-Wigner. Second, the rising interest in PuO 2 fuel motivated an investigation on the generation of group wise cross sections describing thermal resonances of 240 Pu and 242 Pu. (orig.)

  18. 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.

  19. Characterization of neutron flux spectra in the irradiation sites of a 37 GBq {sup 241}Am-Be isotopic source

    Energy Technology Data Exchange (ETDEWEB)

    Yücel, Haluk [Ankara University, Institute of Nuclear Sciences, 06100 Tandogan, Ankara (Turkey); Budak, Mustafa Guray, E-mail: mbudak@gazi.edu.tr [Gazi University, Gazi Education Faculty, 06500 Teknikokullar, Ankara (Turkey); Karadag, Mustafa [Gazi University, Gazi Education Faculty, 06500 Teknikokullar, Ankara (Turkey); Yüksel, Alptuğ Özer [Ankara University, Institute of Nuclear Sciences, 06100 Tandogan, Ankara (Turkey)

    2014-11-01

    Highlights: • An irradiation unit was installed using a 37 GBq {sup 241}Am-Be neutron source. • The source neutrons moderated by using both water and paraffin. • Irradiation unit was shielded by boron oxide and lead against neutrons and gammas. • There are two sites for irradiations, one of them has a pneumatic transfer system. • Cadmium ratio method was used for irradiation site characterization. - Abstract: For the applicability of instrumental neutron activation analysis (NAA) technique, an irradiation unit with a 37 GBq {sup 241}Am-Be neutron source was installed at Institute of Nuclear Sciences of Ankara University. Design and configuration properties of the irradiation unit are described. It has two different sample irradiation positions, one is called site #1 having a pneumatic sample transfer system and the other is site #2 having a location for manual use. In order to characterize neutron flux spectra in the irradiation sites, the measurement results were obtained for thermal (φ{sub th}) and epithermal neutron fluxes (φ{sub epi}), thermal to epithermal flux ratio (f) and epithermal spectrum shaping factors (α) by employing cadmium ratios of gold (Au) and molybdenum (Mo) monitors. The activities produced in these foils were measured by using a p-type, 44.8% relative efficiency HPGe well detector. For the measured γ-rays, self-absorption and true coincidence summing effects were taken into account. Additionally, thermal neutron self-shielding and resonance neutron self-shielding effects were taken into account in the measured results. For characterization of site #1, the required parameters were found to be φ{sub th} = (2.11 ± 0.05) × 10{sup 3} n cm{sup −2} s{sup −1}, φ{sub epi} = (3.32 ± 0.17) × 10{sup 1} n cm{sup −2} s{sup −1}, f = 63.6 ± 1.5, α = 0.045 ± 0.009, respectively. Similarly, those parameters were measured in site #2 as φ{sub th} = (1.49 ± 0.04) × 10{sup 3} n cm{sup −2} s{sup −1}, φ{sub epi} = (2.93 ± 0

  20. Elastic removal self-shielding factors for light and medium nuclides with strong-resonance scattering

    International Nuclear Information System (INIS)

    Nakagawa, Masayuki; Ishiguro, Yukio; Tokuno, Yukio.

    1978-01-01

    The self-shielding factors for elastic removal cross sections of light and medium weight nuclides were calculated for the parameter, σ 0 within the conventional concept of the group constant sets. The numerical study were performed for obtaining a simple and accurate method. The present results were compared with the exact values and the conventional ones, and shown to be remarkably improved. It became apparent that the anisotropy of the elastic scattering did not affect to the self-shielding factors though it did to the infinite dilution cross sections. With use of the present revised set, the neutron flux were calculated in an iron medium and in a prototype FBR and compared with those by the fine spectrum calculations and the conventional set. The present set showed the considerable improvement in the vicinity of the large resonance regions of sodium, iron and oxygen. (auth.)

  1. Radiation protection commissioning of neutron beam instruments at the OPAL research reactor

    International Nuclear Information System (INIS)

    Parkes, Alison; Saratsopoulos, John; Deura, Michael; Kenny, Pat

    2008-01-01

    The neutron beam facilities at the 20 MW OPAL Research Reactor were commissioned in 2007 and 2008. The initial suite of eight neutron beam instruments on two thermal neutron guides, two cold neutron guides and one thermal beam port located at the reactor face, together with their associated shielding were progressively installed and commissioned according to their individual project plans. Radiation surveys were systematically conducted as reactor power was raised in a step-wise manner to 20 MW in order to validate instrument shielding design and performance. The performance of each neutron guide was assessed by neutron energy spectrum and flux measurements. The activation of beam line components, decay times assessments and access procedures for Bragg Institute beam instrument scientists were established. The multiple configurations for each instrument and the influence of operating more than one instrument or beamline simultaneously were also tested. Areas of interest were the shielding around the secondary shutters, guide shield and bunker shield interfaces and monochromator doors. The shielding performance, safety interlock checks, improvements, radiation exposures and related radiation protection challenges are discussed. This paper discusses the health physics experience of commissioning the OPAL Research Reactor neutron beam facilities and describes health physics results, actions taken and lessons learned during commissioning. (author)

  2. Calculation of self–shielding factor for neutron activation experiments using GEANT4 and MCNP

    Energy Technology Data Exchange (ETDEWEB)

    Romero–Barrientos, Jaime, E-mail: jaromero@ing.uchile.cl [Comisión Chilena de Energía Nuclear, Nueva Bilbao 12501, Las Condes, Santiago (Chile); Universidad de Chile, DFI, Facultad de Ciencias Físicas Y Matemáticas, Avenida Blanco Encalada 2008, Santiago (Chile); Molina, F. [Comisión Chilena de Energía Nuclear, Nueva Bilbao 12501, Las Condes, Santiago (Chile); Aguilera, Pablo [Comisión Chilena de Energía Nuclear, Nueva Bilbao 12501, Las Condes, Santiago (Chile); Universidad de Chile, Depto. de Física, Facultad de Ciencias, Las Palmeras 3425, Ñuñoa, Santiago (Chile); Arellano, H. F. [Universidad de Chile, DFI, Facultad de Ciencias Físicas Y Matemáticas, Avenida Blanco Encalada 2008, Santiago (Chile)

    2016-07-07

    The neutron self–shielding factor G as a function of the neutron energy was obtained for 14 pure metallic samples in 1000 isolethargic energy bins from 1·10{sup −5}eV to 2·10{sup 7}eV using Monte Carlo simulations in GEANT4 and MCNP6. The comparison of these two Monte Carlo codes shows small differences in the final self–shielding factor mostly due to the different cross section databases that each program uses.

  3. Piping structural design for the ITER thermal shield manifold

    Energy Technology Data Exchange (ETDEWEB)

    Noh, Chang Hyun, E-mail: chnoh@nfri.re.kr [ITER Korea, National Fusion Research Institute, Daejeon 305-333 (Korea, Republic of); Chung, Wooho, E-mail: whchung@nfri.re.kr [ITER Korea, National Fusion Research Institute, Daejeon 305-333 (Korea, Republic of); Nam, Kwanwoo; Kang, Kyoung-O. [ITER Korea, National Fusion Research Institute, Daejeon 305-333 (Korea, Republic of); Bae, Jing Do; Cha, Jong Kook [Korea Marine Equipment Research Institute, Busan 606-806 (Korea, Republic of); Kim, Kyoung-Kyu [Mecha T& S, Jinju-si 660-843 (Korea, Republic of); Hamlyn-Harris, Craig; Hicks, Robby; Her, Namil; Jun, Chang-Hoon [ITER Organization, Route de Vinon-sur-Verdon, CS 90 046, 13067 St. Paul Lez Durance Cedex (France)

    2015-10-15

    Highlights: • We finalized piping design of ITER thermal shield manifold for procurement. • Support span is determined by stress and deflection limitation. • SQP, which is design optimization method, is used for the pipe design. • Benchmark analysis is performed to verify the analysis software. • Pipe design is verified by structural analyses. - Abstract: The thermal shield (TS) provides the thermal barrier in the ITER tokamak to minimize heat load transferred by thermal radiation from the hot components to the superconducting magnets operating at 4.2 K. The TS is actively cooled by 80 K pressurized helium gas which flows from the cold valve box to the cooling tubes on the TS panels via manifold piping. This paper describes the manifold piping design and analysis for the ITER thermal shield. First, maximum allowable span for the manifold support is calculated based on the simple beam theory. In order to accommodate the thermal contraction in the manifold feeder, a contraction loop is designed and applied. Sequential Quadratic Programming (SQP) method is used to determine the optimized dimensions of the contraction loop to ensure adequate flexibility of manifold pipe. Global structural behavior of the manifold is investigated when the thermal movement of the redundant (un-cooled) pipe is large.

  4. In-core neutron flux measurements at PARR using self powered neutron detector

    International Nuclear Information System (INIS)

    Hussain, A.; Ansari, S.A.

    1989-10-01

    This report describes experimental reactor physics measure ments at PARR using the in-core neutron detectors. Rhodium self powered neutron detectors (SPND) were used in the PARR core and several measurements were made aimed at detector calibration, response time determination and neutron flux measurements. The detectors were calibrated at low power using gold foils and full power by the thermal channel. Based on this calibration it was observed that the detector response remains almost linear throughout the power range. The self powered detectors were used for on-line determination of absolute neutron flux in the core as well as the spatial distribution of neutron flux or reactor power. The experimental, axial and horizontal flux mapping results at certain locations in the core are presented. The total response time of rhodium detector was experimentally determined to be about 5 minutes, which agree well with the theoretical results. Because of longer response time of SPND of the detectors it is not possible to use them in the reactor protection system. (author). 10 figs

  5. Self-Shielding Treatment to Perform Cell Calculation for Seed Furl In Th/U Pwr Using Dragon Code

    Directory of Open Access Journals (Sweden)

    Ahmed Amin El Said Abd El Hameed

    2015-08-01

    Full Text Available Time and precision of the results are the most important factors in any code used for nuclear calculations. Despite of the high accuracy of Monte Carlo codes, MCNP and Serpent, in many cases their relatively long computational time leads to difficulties in using any of them as the main calculation code. Usually, Monte Carlo codes are used only to benchmark the results. The deterministic codes, which are usually used in nuclear reactor’s calculations, have limited precision, due to the approximations in the methods used to solve the multi-group transport equation. Self- Shielding treatment, an algorithm that produces an average cross-section defined over the complete energy domain of the neutrons in a nuclear reactor, is responsible for the biggest error in any deterministic codes. There are mainly two resonance self-shielding models commonly applied: models based on equivalence and dilution and models based on subgroup approach. The fundamental problem with any self-shielding method is that it treats any isotope as there are no other isotopes with resonance present in the reactor. The most practical way to solve this problem is to use multi-energy groups (50-200 that are chosen in a way that allows us to use all major resonances without self-shielding. In this paper, we perform cell calculations, for a fresh seed fuel pin which is used in thorium/uranium reactors, by solving 172 energy group transport equation using the deterministic DRAGON code, for the two types of self-shielding models (equivalence and dilution models and subgroup models Using WIMS-D5 and DRAGON data libraries. The results are then tested by comparing it with the stochastic MCNP5 code.  We also tested the sensitivity of the results to a specific change in self-shielding method implemented, for example the effect of applying Livolant-Jeanpierre Normalization scheme and Rimman Integration improvement on the equivalence and dilution method, and the effect of using Ribbon

  6. Developmental testing of partially volatile neutron shields for high-performance shipping casks

    International Nuclear Information System (INIS)

    Pope, R.B.; Allen, G.C.; Rack, H.J.; Joseph, B.J.; Dupree, S.A.

    1980-01-01

    Results of the phase one tests have demonstrated that the neutron-shielding concept described in this paper is a viable design option for spent fuel shipping casks. The tests have shown that the Boro-silicone 236 shield is superior to the other shield materials considered. Repeated TGA, aging and fire tests demonstrated the reliability of the data. A second phase of the test program is now being pursued where the Boro-silicone 236 is injected into all-steel slab sections, and cured in place. 5 tables

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

    International Nuclear Information System (INIS)

    Clayton, C.G.; Spackman, R.

    1988-01-01

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

  8. Neutron streaming analysis for shield design of FMIT Facility

    International Nuclear Information System (INIS)

    Carter, L.L.

    1980-12-01

    Applications of the Monte Carlo method have been summarized relevant to neutron streaming problems of interest in the shield design for the FMIT Facility. An improved angular biasing method has been implemented to further optimize the calculation of streaming and this method has been applied to calculate streaming within a double bend pipe

  9. Adaptive algorithms for a self-shielding wavelet-based Galerkin method

    International Nuclear Information System (INIS)

    Fournier, D.; Le Tellier, R.

    2009-01-01

    The treatment of the energy variable in deterministic neutron transport methods is based on a multigroup discretization, considering the flux and cross-sections to be constant within a group. In this case, a self-shielding calculation is mandatory to correct sections of resonant isotopes. In this paper, a different approach based on a finite element discretization on a wavelet basis is used. We propose adaptive algorithms constructed from error estimates. Such an approach is applied to within-group scattering source iterations. A first implementation is presented in the special case of the fine structure equation for an infinite homogeneous medium. Extension to spatially-dependent cases is discussed. (authors)

  10. Neutron Radiation Shielding For The NIF Streaked X-Ray Detector (SXD) Diagnostic

    Energy Technology Data Exchange (ETDEWEB)

    Song, P; Holder, J; Young, B; Kalantar, D; Eder, D; Kimbrough, J

    2006-11-02

    The National Ignition Facility (NIF) at Lawrence Livermore National Laboratory (LLNL) is preparing for the National Ignition Campaign (NIC) scheduled in 2010. The NIC is comprised of several ''tuning'' physics subcampaigns leading up to a demonstration of Inertial Confinement Fusion (ICF) ignition. In some of these experiments, time-resolved x-ray imaging of the imploding capsule may be required to measure capsule trajectory (shock timing) or x-ray ''bang-time''. A capsule fueled with pure tritium (T) instead of a deutriun-tritium (DT) mixture is thought to offer useful physics surrogacy, with reduced yields of up to 5e14 neutrons. These measurements will require the use of the NIF streak x-ray detector (SXD). The resulting prompt neutron fluence at the planned SXD location ({approx}1.7 m from the target) would be {approx}1.4e9/cm{sup 2}. Previous measurements suggest the onset of significant background at a neutron fluence of {approx} 1e8/cm{sup 2}. The radiation damage and operational upsets which starts at {approx}1e8 rad-Si/sec must be factored into an integrated experimental campaign plan. Monte Carlo analyses were performed to predict the neutron and gamma/x-ray fluences and radiation doses for the proposed diagnostic configuration. A possible shielding configuration is proposed to mitigate radiation effects. The primary component of this shielding is an 80 cm thickness of Polyethylene (PE) between target chamber center (TCC) and the SXD diagnostic. Additionally, 6-8 cm of PE around the detector provide from the large number of neutrons that scatter off the inside of the target chamber. This proposed shielding configuration reduces the high-energy neutron fluence at the SXD by approximately a factor {approx}50.

  11. Neutron Radiation Shielding For The NIF Streaked X-Ray Detector (SXD) Diagnostic

    International Nuclear Information System (INIS)

    Song, P; Holder, J; Young, B; Kalantar, D; Eder, D; Kimbrough, J

    2006-01-01

    The National Ignition Facility (NIF) at Lawrence Livermore National Laboratory (LLNL) is preparing for the National Ignition Campaign (NIC) scheduled in 2010. The NIC is comprised of several ''tuning'' physics subcampaigns leading up to a demonstration of Inertial Confinement Fusion (ICF) ignition. In some of these experiments, time-resolved x-ray imaging of the imploding capsule may be required to measure capsule trajectory (shock timing) or x-ray ''bang-time''. A capsule fueled with pure tritium (T) instead of a deutriun-tritium (DT) mixture is thought to offer useful physics surrogacy, with reduced yields of up to 5e14 neutrons. These measurements will require the use of the NIF streak x-ray detector (SXD). The resulting prompt neutron fluence at the planned SXD location (∼1.7 m from the target) would be ∼1.4e9/cm 2 . Previous measurements suggest the onset of significant background at a neutron fluence of ∼ 1e8/cm 2 . The radiation damage and operational upsets which starts at ∼1e8 rad-Si/sec must be factored into an integrated experimental campaign plan. Monte Carlo analyses were performed to predict the neutron and gamma/x-ray fluences and radiation doses for the proposed diagnostic configuration. A possible shielding configuration is proposed to mitigate radiation effects. The primary component of this shielding is an 80 cm thickness of Polyethylene (PE) between target chamber center (TCC) and the SXD diagnostic. Additionally, 6-8 cm of PE around the detector provide from the large number of neutrons that scatter off the inside of the target chamber. This proposed shielding configuration reduces the high-energy neutron fluence at the SXD by approximately a factor ∼50

  12. Neutron-photon multigroup cross sections for neutron energies up to 400 MeV: HILO86R

    International Nuclear Information System (INIS)

    Kotegawa, Hiroshi; Nakane, Yoshihiro; Hasegawa, Akira; Tanaka, Shun-ichi

    1993-02-01

    A macroscopic multigroup cross section library of 66 neutron and 22 photon groups for neutron energies up to 400 MeV: HILO86R is prepared for 10 typical shielding materials; water, concrete, iron, air, graphite, polyethylene, heavy concrete, lead, aluminum and soil. The library is a revision of the DLC-119/HILO86, in which only the cross sections below 19.6 MeV have been exchanged with a group cross section processed from the JENDL-3 microscopic cross section library. In the HILO86R library, self shielding factors are used to produce effective cross sections for neutrons less than 19.6 MeV considering rather coarse energy meshes. Energy spectra and dose attenuation in water, concrete and iron have been compared among the HILO, HILO86 and HILO86R libraries for different energy neutron sources. Significant discrepancy has been observed in the energy spectra less than a couple of MeV energy in iron among the libraries, resulting large difference in the dose attenuation. The difference was attributed to the effect of self-shielding factor, namely to the difference between infinite dilution and effective cross sections. Even for 400 MeV neutron source the influence of the self-shielding factor is significant, nevertheless only the cross sections below 19.6 MeV are exchanged. (author)

  13. Thermal neutron inelastic scattering and it's application to the material science

    International Nuclear Information System (INIS)

    Li Zhuqi

    1986-01-01

    A brief description of the elementary scattering theory of the interaction between the thermal neutrons and the condensed matter is given and the characteristics related to the experimental method of the thermal neutrons inelastic scattering is described. Expressions of the phonons dispersion, density of the phonon state and the self-diffusion coefficient at the some conditions are also introduced. Some examples of describing diagram of the phonon dispersion, density of the phonons state and selfdiffusion coefficient measured by different authors are given

  14. The properties of neutron shielding and flame retardant of EVA polymer after modified by EB accelerator

    Science.gov (United States)

    Wang, Guo-hui; He, Man-li; Jiang, Dan-feng; He, Fan; Chang, Shu-quan; Dai, Yao-dong

    2017-11-01

    According to the requirements for neutron shielding and flame retardant properties of some nuclear devices, a new kind of polymer composite materials based on ethylene and vinyl acetate (EVA) polymer have been studied. EVA is the copolymer of ethylene and vinyl acetate, It can be used as materials for applications due to its flexibility, good processability, and low cost. Insulating EVA can be used for cable sheath, automotive sound damping and many other appication. Boron nitride (BN), zinc borate (ZB), magnesium hydroxide (MH) and EVA consisted the compounds with the properties of neutron shielding and flame retardant. With increasing of the contents of BN and ZB, the neutron shielding performance of materials increased up to 33.08%. With the increasing contents of MH and ZB as flame retardant, oxygen index of material have been improved. The elongation at break and tensile strength of material decreased with the increasing of filler powders. Sheet E was chosen and modified by electron beam accelerator in different doses. After modification by electron beam irradiation the sheets showed varying degrees of transformation in the OI, neutron shielding rate and mechanical properties.

  15. Calculation of neutron fluxes in biological shield of the TRIGA Mark II reactor

    International Nuclear Information System (INIS)

    Bozic, M.; Zagar, T.; Ravnik, M.

    2001-01-01

    The complete calculation of neutron fluxes in biological shield and verification with experimental results is presented. Calculated results are obtained with TORT code (TORT-Three Dimensional Oak Ridge Discrete Ordinates Neutron/Photon Transport Code). Experimental results used for comparison are available from irradiation experiment with selected type of concrete and other materials in irradiation channel 4 in TRIGA Mark II reactor. These experimental results were used as a benchmark. Homogeneous type of problem (without inserted irradiation channel) and problem with asymmetry (inserted beam port 4, filled with different materials) were of interest for neutron flux calculation. Deviation from material data set up as original parameters is also considered (first of all presence of water in concrete and density of concrete) for type of concrete in biological shield and for selected type of concrete in irradiation channel. BUGLE-96 (47 neutron energy groups) library is used. Excellent agreement between calculated and experimental results for reaction rate is received.(author)

  16. Self powered neutron detectors

    International Nuclear Information System (INIS)

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

    1975-01-01

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

  17. Study on bulk shielding for a spallation neutron source facility in the high-intensity proton accelerator project

    CERN Document Server

    Maekawa, F; Takada, H; Teshigawara, M; Watanabe, N

    2002-01-01

    Under the JAERI-KEK High-Intensity Proton Accelerator Project, a spallation neutron source driven by a 3 GeV-1 MW proton beam is planed to be constructed in a main part of the Materials and Life Science Facility. This report describes results of a study on bulk shielding performance of a biological shield for the spallation neutron source by means of a Monte Carlo calculation method, that is important in terms of radiation safety and cost reduction. A shielding configuration was determined as a reference case by considering preliminary studies and interaction with other components, then shielding thickness that was required to achieve a target dose rate of 1 mu Sv/h was derived. Effects of calculation conditions such as shielding materials and dimensions on the shielding performance was investigated by changing those parameters. By taking all the results and design margins into account, a shielding configuration that was identified as the most appropriate was finally determined as follows. An iron shield regi...

  18. 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)

  19. Online In-Core Thermal Neutron Flux Measurement for the Validation of Computational Methods

    International Nuclear Information System (INIS)

    Mohamad Hairie Rabir; Muhammad Rawi Mohamed Zin; Yahya Ismail

    2016-01-01

    In order to verify and validate the computational methods for neutron flux calculation in RTP calculations, a series of thermal neutron flux measurement has been performed. The Self Powered Neutron Detector (SPND) was used to measure thermal neutron flux to verify the calculated neutron flux distribution in the TRIGA reactor. Measurements results obtained online for different power level of the reactor. The experimental results were compared to the calculations performed with Monte Carlo code MCNP using detailed geometrical model of the reactor. The calculated and measured thermal neutron flux in the core are in very good agreement indicating that the material and geometrical properties of the reactor core are modelled well. In conclusion one can state that our computational model describes very well the neutron flux distribution in the reactor core. Since the computational model properly describes the reactor core it can be used for calculations of reactor core parameters and for optimization of RTP utilization. (author)

  20. 3-dimensional shielding design for a spallation neutron source facility in the high-intensity proton accelerator project

    Energy Technology Data Exchange (ETDEWEB)

    Tamura, Masaya; Maekawa, Fujio [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2003-03-01

    Evaluation of shielding performance for a 1 MW spallation neutron source facility in the Materials and Life Science Facility being constructed in the High-Intensity Proton Accelerator Project (J-PARC) is important from a viewpoint of radiation safety and optimization of arrangement of components. This report describes evaluated results for the shielding performance with modeling three-dimensionally whole structural components including gaps between them in detail. A Monte Carlo calculation method with MCNPX2.2.6 code and LA-150 library was adopted. Streaming and void effects, optimization of shield for cost reduction and optimization of arrangement of structures such as shutters were investigated. The streaming effects were investigated quantitatively by changing the detailed structure of components and gap widths built into the calculation model. Horizontal required shield thicknesses were ranged from about 6.5 m to 7.5 m as a function of neutron beam line angles. A shutter mechanism for a horizontal neutron reflectometer that was directed downward was devised, and it was shown that the shielding performance of the shutter was acceptable. An optimal biological shield configuration was finally determined according to the calculated results. (author)

  1. GROUPIE2007, Bondarenko Self-Shielded Cross sections from ENDF/B

    International Nuclear Information System (INIS)

    2007-01-01

    1 - Description of problem or function - GROUPIE reads evaluated data in ENDF/B Format and uses these to calculate unshielded group averaged Cross sections, Bondarenko self-shielded Cross sections, and multiband parameters. The program allows the user to specify arbitrary energy groups and an arbitrary energy-dependent neutron spectrum (weighting function). IAEA0849/15: This version include the updates up to January 30, 2007. Changes in ENDF/B-VII Format and procedures, as well as the evaluations themselves, make it impossible for versions of the ENDF/B pre-processing codes earlier than PREPRO 2007 (2007 Version) to accurately process current ENDF/B-VII evaluations. The present code can handle all existing ENDF/B-VI evaluations through release 8, which will be the last release of ENDF/B-VI. 2 - Modifications from previous versions: Groupie VERS. 2007-1 (Jan. 2007): checked against all ENDF/B-VII; increased page size from 120,000 to 600,000 points. 3 - Method of solution: All integrals are performed analytically; in no case is iteration or any approximate form of integration used. GROUPIE reads either the 0 deg. Kelvin Cross sections or the Doppler broadened Cross sections to calculate the self-shielded Cross sections and multiband parameters for 25 values of the 'background' Cross sections (representing the combined effects of all other isotopes and of leakage). 4 - Restrictions on the complexity of the problem: GROUPIE requires that the energy-dependent neutron spectrum and all Cross sections be given in tabular form, with linear interpolation between tabulated values. There is no limit to the size of the table used to describe the spectrum, so the spectrum may be described in as much detail as required. - If only unshielded averages are calculated, the program can handle up to 3000 groups. If self-shielded averages and/or multiband parameters are calculated, the program can handle up to 175 groups. These limits can easily be extended. - The program only uses the

  2. RASH D - A mercury programme for neutron shielding calculations

    International Nuclear Information System (INIS)

    Bendall, D.E.

    1962-08-01

    An improved version of an earlier neutron shielding programme (RASH B) is described. The new programme is also written in Mercury Autocode and solves a set of multigroup diffusion equations in one dimension. It differs from RASH B in that distributed source terms may be introduced into all the groups if required. Some other improvements are also included. (author)

  3. Shielding modification design of the N.S. Mutsu

    International Nuclear Information System (INIS)

    Yamaji, A.; Miyakoshi, J.; Kageyama, T.; Futamura, Y.

    1983-01-01

    Shielding modification design of the N.S. Mutsu was performed for reducing the radiation doses outside the primary and the secondary shields by providing shields for neutrons streaming through the air gap between the pressure vessel and the primary shield. This was accomplished by replacing parts of the shields and adding new shields in the upper and lower sections of both primary and secondary shields, and also replacing the thermal insulator in the gap. The shielding design calculations were made using one- and two-dimensional discrete ordinates codes and also a point kernel code. Special attention was paid to the calculations of, (1) the neutrons streaming through the gap between the pressure vessel and the primary shield, (2) the radiations transmitted through the radial shield of the core in the primary shield, (3) the radiations transmitted through the upper and lower sections of the secondary shield, and (4) the dose rate equivalent in the accommodation area. Their calculational accuracies were estimated by analyzing various experiments. To support the modification, a variety of experiments and tests were carried out, which were material tests, cooling test of the primary shield, mechanical strength test of the double bottom, trial fabrication tests of new shields, performance degradation test of heavy concrete and duct streaming experiment in the secondary shield. (author)

  4. Uncertainty Analysis with Considering Resonance Self-shielding Effect

    Energy Technology Data Exchange (ETDEWEB)

    Han, Tae Young [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    If infinitely diluted multi-group cross sections were used for the sensitivity, the covariance data from the evaluated nuclear data library (ENDL) was directly applied. However, in case of using a self-shielded multi-group cross section, the covariance data should be corrected considering self-shielding effect. Usually, implicit uncertainty can be defined as the uncertainty change by the resonance self-shielding effect as described above. MUSAD ( Modules of Uncertainty and Sensitivity Analysis for DeCART ) has been developed for a multiplication factor and cross section uncertainty based on the generalized perturbation theory and it, however, can only quantify the explicit uncertainty by the self-shielded multi-group cross sections without considering the implicit effect. Thus, this paper addresses the implementation of the implicit uncertainty analysis module into the code and the numerical results for the verification are provided. The implicit uncertainty analysis module has been implemented into MUSAD based on infinitely-diluted cross section-based consistent method. The verification calculation was performed on MHTGR 350 Ex.I-1a and the differences with McCARD result decrease from 40% to 1% in CZP case and 3% in HFP case. From this study, it is expected that MUSAD code can reasonably produce the complete uncertainty on VHTR or LWR where the resonance self-shielding effect should be significantly considered.

  5. Uncertainty Analysis with Considering Resonance Self-shielding Effect

    International Nuclear Information System (INIS)

    Han, Tae Young

    2016-01-01

    If infinitely diluted multi-group cross sections were used for the sensitivity, the covariance data from the evaluated nuclear data library (ENDL) was directly applied. However, in case of using a self-shielded multi-group cross section, the covariance data should be corrected considering self-shielding effect. Usually, implicit uncertainty can be defined as the uncertainty change by the resonance self-shielding effect as described above. MUSAD ( Modules of Uncertainty and Sensitivity Analysis for DeCART ) has been developed for a multiplication factor and cross section uncertainty based on the generalized perturbation theory and it, however, can only quantify the explicit uncertainty by the self-shielded multi-group cross sections without considering the implicit effect. Thus, this paper addresses the implementation of the implicit uncertainty analysis module into the code and the numerical results for the verification are provided. The implicit uncertainty analysis module has been implemented into MUSAD based on infinitely-diluted cross section-based consistent method. The verification calculation was performed on MHTGR 350 Ex.I-1a and the differences with McCARD result decrease from 40% to 1% in CZP case and 3% in HFP case. From this study, it is expected that MUSAD code can reasonably produce the complete uncertainty on VHTR or LWR where the resonance self-shielding effect should be significantly considered

  6. Thermal shielding device in LMFBR type reactors

    International Nuclear Information System (INIS)

    Nakamura, Hiroshi.

    1985-01-01

    Purpose: To improve the soundness and earthquake proofness of mounting structures to a reactor vessel in a thermal shielding device comprising a plurality of tightly closed casings evacuated or shield with heat insulation gases, by reducing the wall thickness and weight of the casing. Constitution: the thermal shielding body comprises tightly closed casings and compressing core materials for preventing the deformation of the casings. The tightly closed casing is in the shape of a hollow vessel, completely sealed in gastight manner, and evacuated or sealed with heat insulation gases at a low pressure of about less than 0.5 kg/cm 2 G, such that the inner pressure is lower than the outer pressure. Compressing core materials made of porous metals or porous ceramics are contained to the inside of the casing. In this way, the wall thickness of the tightly closed casing can be reduced significantly as compared with the conventional case, whereby the mounting work on the site to the reactor container on the field can remarkably be improved and high reliability can be maintained at the mounting portion. (Kamimura, M.)

  7. 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)

  8. Neutron dose equivalent next to the target shield of a neutron therapy facility using an LET counter

    International Nuclear Information System (INIS)

    Stinchcomb, T.G.; Kuchnir, F.T.

    1981-01-01

    The use of a spherical tissue-equivalent proportional counter for measurements of the lineal energy (y) and derivations of the linear energy transfer (LET) for fast neutrons has the advantage of giving distributions of dose and dose equivalent as functions of either LET or y. A measurement next to the target shielding of the neutron therapy facility at the University of Chicago Hospitals and Clinics (UCHC) is described, and the data processing is outlined. The distributions are presented and compared to those from measurements in the neutron beam. The average quality factors are presented

  9. Manufacture and testing of the CTB&SBB thermal shield for the ITER magnet feeder system

    Energy Technology Data Exchange (ETDEWEB)

    Lu, Kun; Liu, Chen; Song, Yuntao; Feng, Hansheng; Ding, Kaizhong, E-mail: kzding@ipp.ac.cn; Wang, Tanbin; Ji, Hui

    2015-10-15

    The system of International Thermonuclear Experimental Reactor (ITER) feeders is responsible for the power, helium cooling, and instrumentation of the magnets of the coil terminal box and S-bend box (CTB&SBB) thermal shield outside the cryostat. An 80-K rectangular Al thermal shield is hung inside the CTB&SBB to reduce the thermal radiation heat loads of 4.5-K helium. The American Society of Interventional Pain Physicians (ASIPP) will supply all the 31 sets of feeders for ITER. A manufactured prototype of CTB&SBB thermal shield is first quality-tested before the commencement of the series production. First, a detailed configuration of the rectangular Al thermal shield is presented in this article. The paper also presents more information on the manufacturing process of the thermal shield, especially the welding process, the procedure for ensuring good weld quality, and the use of a specially designed tool to ensure <5-mm deformation on such a 7.3-m-long thermal shield during welding. In addition, the cold test and results, including the cooling process with 13-bar and 17.5-g/s 80-K He gas, and the temperature distribution on different panels of the thermal shield are presented. The whole process of manufacture and testing lays a good foundation for the series production of the thermal shield.

  10. Neutron flux measurements in C-9 capsule pressure tube

    International Nuclear Information System (INIS)

    Barbos, D.; Roth, C. S.; Gugiu, D.; Preda, M.

    2001-01-01

    C-9 capsule is a fuel testing facility in which the testing consists of a daily cycle ranging between the limits 100% power to 50% power. C-9 in-pile section with sample holder an instrumentation are introduced in G-9 and G-10 experimental channels. The experimental fuel channel has a maximum value when the in-pile section (pressure tube) is in G-9 channel and minimum value in G-10 channel. In this paper the main goals are determination or measurements of: - axial thermal neutron flux distribution in C-9 pressure tube both in G-9 and G-10 channel; - ratio of maximum neutron flux value in G-9 and the same value in G-9 channel and the same value in G-10 channel; - neutron flux-spectrum. On the basis of axial neutron flux distribution measurements, the experimental fuel element in sample holder position in set. Both axial neutron flux distribution of thermal neutrons and neutron flux-spectrum were performed using multi- foil activation technique. Activation rates were obtained by absolute measurements of the induced activity using gamma spectroscopy methods. To determine the axial thermal neutron flux distribution in G-9 and G-10, Cu 100% wire was irradiated at the reactor power of 2 MW. Ratio between the two maximum values, in G-9 and G-10 channels, is 2.55. Multi-foil activation method was used for neutron flux spectrum measurements. The neutron spectra and flux were obtained from reaction rate measurements by means of SAND 2 code. To obtain gamma-ray spectra, a HPGe detector connected to a multichannel analyzer was used. The spectrometer is absolute efficiency calibrated. The foils were irradiated at 2 MW reactor power in previously determined maximum flux position resulted from wire measurements. This reaction rates were normalized for 10 MW reactor power. Neutron self shielding corrections for the activation foils were applied. The self-shielding corrections are computed using Monte Carlo simulation methods. The measured integral flux is 1.1·10 14 n/cm 2 s

  11. Systems for neutronic, thermohydraulic and shielding calculation in personal computers

    International Nuclear Information System (INIS)

    Villarino, E.A.; Abbate, P.; Lovotti, O.; Santini, M.

    1990-01-01

    The MTR-PC (Materials Testing Reactors-Personal Computers) system has been developed by the Nuclear Engineering Division of INVAP S.E. with the aim of providing working conditions integrated with personal computers for design and neutronic, thermohydraulic and shielding analysis for reactors employing plate type fuel. (Author) [es

  12. Development of a transportable neutron radiography system for non-destructive tests application

    International Nuclear Information System (INIS)

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

    1999-01-01

    This paper presents a study of a transportable neutron radiography system utilizing californium-252. Studies about moderation, collimation and shielding are showed. A Monte Carlo Code, MCNP3b, has been used to obtain a maximum and more homogeneous thermal neutron flux in the collimator outlet next to the image plain, and an adequate radiation shielding to attend radiological protection rules. With the presented collimator, it was possible to obtain for the thermal neutron flux, at the collimator outlet and next to the image plain, a L/D ratio 7,5, for neutron flux up to 6 X 10 -6 cm -2 .s -1 per neutron source. (author)

  13. Shielding performances analysis for the IFMIF test facility based on high-fidelity Monte Carlo neutronic calculations

    Energy Technology Data Exchange (ETDEWEB)

    Kondo, Keitaro, E-mail: kondo.keitaro@jaea.go.jp; Arbeiter, Frederik; Fischer, Ulrich; Lu, Lei; Qiu, Yuefeng; Tian, Kuo

    2015-10-15

    Highlights: • A detailed geometry model with pipe penetrations and gaps was prepared for the IFMIF test cell. • The neutron streaming effect due to gaps and pipes with shielding plugs was investigated. • The present analysis revealed that the streaming effect can be mitigated by some counter measures. • Occupational workers can access to the room above the test cell during operation. - Abstract: The IFMIF Test Cell (TC) design was developed and optimized in the EVEDA phase, and finally the reference TC design was proposed. The present study is devoted to further investigations of open issues on the reference TC design. In order to examine the neutron streaming effect caused by pipe penetrations and gaps around removable shielding plugs, a new geometry model for neutronic analyses has been prepared directly from engineering CAD data by utilizing the McCad conversion software. All removable shielding plugs are separately described in the model and a detailed description of pipes was incorporated into the model. The calculation result suggests that the streaming effect is mitigated if the pipe penetration is designed appropriately, while the gaps around the shielding plugs above the TC have large impact on the radiation dose in the access cell. The concept of the reference TC design has been basically validated from the neutronics point of view, although the streaming effect should be compensated by the shielding capability of the test cell cover plate so that occupational workers can access to the access cell during operation.

  14. Shielding performances analysis for the IFMIF test facility based on high-fidelity Monte Carlo neutronic calculations

    International Nuclear Information System (INIS)

    Kondo, Keitaro; Arbeiter, Frederik; Fischer, Ulrich; Lu, Lei; Qiu, Yuefeng; Tian, Kuo

    2015-01-01

    Highlights: • A detailed geometry model with pipe penetrations and gaps was prepared for the IFMIF test cell. • The neutron streaming effect due to gaps and pipes with shielding plugs was investigated. • The present analysis revealed that the streaming effect can be mitigated by some counter measures. • Occupational workers can access to the room above the test cell during operation. - Abstract: The IFMIF Test Cell (TC) design was developed and optimized in the EVEDA phase, and finally the reference TC design was proposed. The present study is devoted to further investigations of open issues on the reference TC design. In order to examine the neutron streaming effect caused by pipe penetrations and gaps around removable shielding plugs, a new geometry model for neutronic analyses has been prepared directly from engineering CAD data by utilizing the McCad conversion software. All removable shielding plugs are separately described in the model and a detailed description of pipes was incorporated into the model. The calculation result suggests that the streaming effect is mitigated if the pipe penetration is designed appropriately, while the gaps around the shielding plugs above the TC have large impact on the radiation dose in the access cell. The concept of the reference TC design has been basically validated from the neutronics point of view, although the streaming effect should be compensated by the shielding capability of the test cell cover plate so that occupational workers can access to the access cell during operation.

  15. Epithermal interrogation of fissile waste

    International Nuclear Information System (INIS)

    Coop, K.L.; Hollas, C.L.

    1996-01-01

    Self-shielding of interrogating thermal neutrons in lumps of fissile material can be a major source of error in transuranic waste assay using the widely employed differential dieaway technique. We are developing a new instrument, the combined thermal/epithermal neutron (CTEN) interrogation instrument to detect the occurrence of self- shielding and mitigate its effects. Neutrons are moderated in the graphite walls of the CTEN instrument to provide an interrogating flux of epithermal and thermal neutrons. The induced prompt fission neutrons are detected in proportional counters. We report the results of measurements made with the CTEN instrument, using minimal and highly self-shielding plutonium and uranium sources in 55 gallon drums containing a variety of mock waste matrices. Fissile isotopes and waste forms for which the method is most applicable, and limitations associated with the hydrogen content of the waste package/matrix are described

  16. Neutron/photon/electron shielding study for a laser-fusion facility

    International Nuclear Information System (INIS)

    Thompson, W.L.

    1977-01-01

    A Monte Carlo shielding study encompassing neutron, photon, and electron transport has been conducted for the High Energy Gas Laser Facility at the Los Alamos Scientific Laboratory. This paper describes the application of the Monte Carlo technique and several variance reduction schemes to the study. The calculations involve a geometry which is complicated in all three dimensions, a very intense 14 MeV neutron source, skyshine and deep penetrations. The facility design with 1.83 m concrete walls and a 1.52 m concrete roof is based on these calculations

  17. The Benchmark experiment on stainless steel bulk shielding at the Frascati neutron generator

    International Nuclear Information System (INIS)

    Batistoni, P.; Angelone, M.; Martone, M.; Pillon, M.; Rado, V.

    1994-11-01

    In the framework of the European Technology Program for NET/ITER, ENEA (Italian Agency for New Technologies, Energy and Environment) - Frascati and CEA (Commissariat a L'Energie Atomique) - Cadarache collaborated on a Bulk Shield Benchmark Experiment using the 14-MeV Frascati Neutron Generator (FNG). The aim of the experiment was to obtain accurate experimental data for improving the nuclear database and methods used in shielding designs, through a rigorous analysis of the results. The experiment consisted of the irradiation of a stainless steel block by 14-MeV neutrons. The neutron reaction rates at different depths inside the block were measured by fission chambers and activation foils characterized by different energy response ranges. The experimental results have been compared with numerical results calculated using both S N and Monte Carlo transport codes and as transport cross section library the European Fusion File (EFF). In particular, the present report describes the experimental and numerical activity, including neutron measurements and Monte Carlo calculations, carried out by the ENEA Italian Agency for New Technologies, Energy and Environment) team

  18. Measurement of neutron diffusion length in heavy concrete

    International Nuclear Information System (INIS)

    Krejci, D.

    2007-04-01

    Using an aluminium sampler filled with heavy concrete the neutron diffusion length was determined, measuring thermal and fast neutrons over the whole beam hole with various threshold detectors using gold samples. These calculations should describe the neutron distribution in the whole concrete shield of the reactor and contribute to the investigation of the activation of the concrete shield using reactor parameters like operating time, power and neutron flux. Instrumentation, activation and positioning of the samples in the beam hole of the TRIGA Mark II reactor are described. (nevyjel)

  19. The shielding calculation for the CN guide shielding assembly in HANARO

    International Nuclear Information System (INIS)

    Kim, H. S.; Lee, B. C.; Lee, K. H.; Kim, H.

    2006-01-01

    The cold neutron research facility in HANARO is under construction. The area including neutron guides and rotary shutter in the reactor hall should be shielded by the guide shielding assembly which is constructed of heavy concrete blocks and structure. The guide shielding assembly is divided into 2 parts, A and B. Part A is about 6.4 meters apart from the reactor biological shield and it is constructed of heavy concrete blocks whose density is above 4.0g/cm 3 . And part B is a fixed heavy concrete structure whose density is above 3.5g/cm 3 . The rotary shutter is also made with heavy concrete whose density is above 4.0g/cm 3 and includes 5 neutron guides inside. It can block the neutron beam by rotating when CNS is not operating. The dose criterion outside the guide shielding assembly is established as 12.5 μSv/hr which is also applied to reactor shielding in HANARO

  20. Shielding concerns at a spallation source

    International Nuclear Information System (INIS)

    Russell, G.J.; Robinson, H.; Legate, G.L.; Woods, R.

    1989-01-01

    Neutrons produced by 800-MeV proton reactions at the Los Alamos Neutron Scattering Center spallation neutron source cause a variety of challenging shielding problems. We identify several characteristics distinctly different from reactor shielding and compute the dose attenuation through an infinite slab/shield composed of iron (100 cm) and borated polyethylene (15 cm). Our calculations show that (for an incident spallation spectrum characteristic of neutrons leaking from a tungsten target at 90/degree/) the dose through the shield is a complex mixture of neutrons and gamma rays. High-energy (> 20 MeV) neutron production from the target is ≅5% of the total, yet causes ≅68% of the dose at the shield surface. Primary low-energy (< 20 MeV) neutrons from the target contribute negligibly (≅0.5%) to the dose at the shield surface yet cause gamma rays, which contribute ≅31% to the total dose at the shield surface. Low-energy neutrons from spallation reactions behave similarly to neutrons with a fission spectrum distribution. 6 refs., 8 figs., 1 tab

  1. MARS14 deep-penetration calculation for the ISIS target station shielding

    International Nuclear Information System (INIS)

    Nakao, Noriaki; Nunomiya, Tomoya; Iwase, Hiroshi; Nakamura, Takashi

    2004-01-01

    The calculation of neutron penetration through a thick shield was performed with a three-dimensional multi-layer technique using the MARS14(02) Monte Carlo code to compare with the experimental shielding data in 1998 at the ISIS spallation neutron source facility of Rutherford Appleton Laboratory. In this calculation, secondary particles from a tantalum target bombarded by 800-MeV protons were transmitted through a bulk shield of approximately 3-m-thick iron and 1-m-thick concrete. To accomplish this deep-penetration calculation, a three-dimensional multi-layer technique and energy cut-off method were used considering a spatial statistical balance. Finally, the energy spectra of neutrons behind the very thick shield could be calculated down to the thermal energy with good statistics, and the calculated results typically agree well within a factor of two with the experimental data over a broad energy range. The 12 C(n,2n) 11 C reaction rates behind the bulk shield were also calculated, which agree with the experimental data typically within 60%. These results are quite impressive in calculation accuracy for deep-penetration problem

  2. Basic design of shield blocks for a spallation neutron source under the high-intensity proton accelerator project

    Energy Technology Data Exchange (ETDEWEB)

    Yoshida, Katsuhiko; Maekawa, Fujio; Takada, Hiroshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    2003-03-01

    Under the JAERI-KEK High-Intensity Proton Accelerator Project (J-PARC), a spallation neutron source driven by a 3 GeV-1 MW proton beam is planed to be constructed as a main part of the Materials and Life Science Facility. Overall dimensions of a biological shield of the neutron source had been determined by evaluation of shielding performance by Monte Carlo calculations. This report describes results of design studies on an optimum dividing scheme in terms of cost and treatment and mechanical strength of shield blocks for the biological shield. As for mechanical strength, it was studied whether the shield blocks would be stable, fall down or move to a horizontal direction in case of an earthquake of seismic intensity of 5.5 (250 Gal) as an abnormal load. For ceiling shielding blocks being supported by both ends of the long blocks, maximum bending moment and an amount of maximum deflection of their center were evaluated. (author)

  3. Basic design of shield blocks for a spallation neutron source under the high-intensity proton accelerator project

    CERN Document Server

    Yoshida, K; Takada, H

    2003-01-01

    Under the JAERI-KEK High-Intensity Proton Accelerator Project (J-PARC), a spallation neutron source driven by a 3 GeV-1 MW proton beam is planed to be constructed as a main part of the Materials and Life Science Facility. Overall dimensions of a biological shield of the neutron source had been determined by evaluation of shielding performance by Monte Carlo calculations. This report describes results of design studies on an optimum dividing scheme in terms of cost and treatment and mechanical strength of shield blocks for the biological shield. As for mechanical strength, it was studied whether the shield blocks would be stable, fall down or move to a horizontal direction in case of an earthquake of seismic intensity of 5.5 (250 Gal) as an abnormal load. For ceiling shielding blocks being supported by both ends of the long blocks, maximum bending moment and an amount of maximum deflection of their center were evaluated.

  4. Self-powered neutron detector of high sensitivity

    International Nuclear Information System (INIS)

    Brixy, H.; Spillekothen, H.G.; Benninghofen, G.; Serafin, N.

    1983-01-01

    A self-powered neutron detector is proposed, consisting of three concentrically arranged electrically conducting tubes; where the central one forms the emitter and the inner and outer ones form the collector and where the tubes are electrically insulated from each other by insulating material. The emitter consists of a material with a high absorption cross-section for thermal neutrons, particularly of gadolinium, and is provided with an auxiliary emitter layer on the inside or the outside. With suitable dimensions and material, the auxiliary emitter layer increases the yield of electrons. (orig./HP) [de

  5. Shielding Design and Radiation Shielding Evaluation for LSDS System Facility

    International Nuclear Information System (INIS)

    Kim, Younggook; Kim, Jeongdong; Lee, Yongdeok

    2015-01-01

    As the system characteristics, the target in the spectrometer emits approximately 1012 neutrons/s. To efficiently shield the neutron, the shielding door designs are proposed for the LSDS system through a comparison of the direct shield and maze designs. Hence, to guarantee the radiation safety for the facility, the door design is a compulsory course of the development of the LSDS system. To improve the shielding rates, 250x250 covering structure was added as a subsidiary around the spectrometer. In this study, the evaluations of the suggested shielding designs were conducted using MCNP code. The suggested door design and covering structures can shield the neutron efficiently, thus all evaluations of all conditions are satisfied within the public dose limits. From the Monte Carlo code simulation, Resin(Indoor type) and Tungsten(Outdoor type) were selected as the shielding door materials. From a comparative evaluation of the door thickness, In and Out door thickness was selected 50 cm

  6. Thermal neutron detection by activation of CaSO4:Dy + KBr thermoluminescent phosphors

    International Nuclear Information System (INIS)

    Gordon, A.M.P.L.; Muccillo, R.

    1979-01-01

    Thermoluminescence (TL) studies to detect thermal neutrons were performed in cold-pressed CaSO 4 :0,1%Dy + KBr samples. The detection is based on the self-irradiation of the CaSO 4 :Dy TL phosphor by the Br isotopes activated by exposure to a mixed neutron-gamma field. (Author) [pt

  7. A Monte Carlo Library Least Square approach in the Neutron Inelastic-scattering and Thermal-capture Analysis (NISTA) process in bulk coal samples

    Science.gov (United States)

    Reyhancan, Iskender Atilla; Ebrahimi, Alborz; Çolak, Üner; Erduran, M. Nizamettin; Angin, Nergis

    2017-01-01

    A new Monte-Carlo Library Least Square (MCLLS) approach for treating non-linear radiation analysis problem in Neutron Inelastic-scattering and Thermal-capture Analysis (NISTA) was developed. 14 MeV neutrons were produced by a neutron generator via the 3H (2H , n) 4He reaction. The prompt gamma ray spectra from bulk samples of seven different materials were measured by a Bismuth Germanate (BGO) gamma detection system. Polyethylene was used as neutron moderator along with iron and lead as neutron and gamma ray shielding, respectively. The gamma detection system was equipped with a list mode data acquisition system which streams spectroscopy data directly to the computer, event-by-event. A GEANT4 simulation toolkit was used for generating the single-element libraries of all the elements of interest. These libraries were then used in a Linear Library Least Square (LLLS) approach with an unknown experimental sample spectrum to fit it with the calculated elemental libraries. GEANT4 simulation results were also used for the selection of the neutron shielding material.

  8. The stainless steel bulk shielding benchmark experiment at the Frascati Neutron Generator (FNG)

    International Nuclear Information System (INIS)

    Batistoni, P.; Angelone, M.; Martone, M.; Petrizzi, L.; Pillon, M.; Rado, V.; Santamarina, A.; Abidi, I.; Gastaldi, G.; Joyer, P.; Marquette, J.P.; Martini, M.

    1994-01-01

    In the framework of the European Technology Program for NET/ITER, ENEA (Ente Nazionale per le Nuove Tecnologie, l'Energia e l'Ambiente), Frascati and CEA (Commissariat a l'Energie Atomique), Cadarache, are collaborating on a bulk shielding benchmark experiment using the 14 MeV Frascati Neutron Generator (FNG). The aim of the experiment is to obtain accurate experimental data for improving the nuclear database and methods used in the shielding designs, through a rigorous analysis of the results. The experiment consists of the irradiation of a stainless steel block by 14 MeV neutrons. The neutron flux and spectra at different depths, up to 65 cm inside the block, are measured by fission chambers and activation foils characterized by different energy response ranges. The γ-ray dose measurements are performed with ionization chambers and thermo-luminescent dosimeters (TLD). The first results are presented, as well as the comparison with calculations using the cross section library EFF (European Fusion File). ((orig.))

  9. Neutron and Gamma Shielding Evaluation for KN-12 Spent Nuclear Fuel Transport Cask

    Energy Technology Data Exchange (ETDEWEB)

    Cho, I. J.; Min, D. K.; Lee, J. C.; You, G. S.; Yoon, J. S. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Chang, G. H.; Jeong, Y. C.; Ko, Y. W. [Korea Hydro and Nuclear Power Co., LTD., Kori (Korea, Republic of)

    2007-07-01

    The CASTOR KN-12 is designed to transport 12 intact PWR spent fuel assemblies for dry and wet transportation conditions. The overall cask length is 480.1 cm with a wall thickness 37.5 cm. Shield for the KN-12 is maintained by the thick walled cask body and the lid. For neutron shielding, polyethylene rods (PE) are arranged in longitudinal boreholes in the vessel wall and PE-plates are inserted between the cask lid and lid side shock absorber and between the cask bottom and bottom steel plate. The shielding evaluation of the cask has been performed with MCNP to confirm the shielding integrity of cask for pre-service inspection of transport cask.

  10. ParShield: A computer program for calculating attenuation parameters of the gamma rays and the fast neutrons

    International Nuclear Information System (INIS)

    Elmahroug, Y.; Tellili, B.; Souga, C.; Manai, K.

    2015-01-01

    Highlights: • Description of the theoretical method used by the ParShield program. • Description of the ParShield program. • Test and validation the ParShield program. - Abstract: This study aims to present a new computer program called ParShield which determines the neutron and gamma-ray shielding parameters. This program can calculate the total mass attenuation coefficients (μ t ), the effective atomic numbers (Z eff ) and the effective electron densities (N eff ) for gamma rays and it can also calculate the effective removal cross-sections (Σ R ) for fast neutrons for mixtures and compounds. The results obtained for the gamma rays by using ParShield were compared with the results calculated by the WinXcom program and the measured results. The obtained values of (Σ R ) were tested by comparing them with the measured results,the manually calculated results and with the results obtained by using MERCSFN program and an excellent agreement was found between them. The ParShield program can be used as a fast and effective tool to choose and compare the shielding materials, especially for the determination of (Z eff ) and (N eff ), there is no other programs in the literature which can calculate

  11. Thermal neutron capture cross-section and resonance integral measurements of {sup 139}La(n, γ){sup 140}La and {sup 140}Ce(n, γ){sup 141}Ce using a Am-Be neutron source

    Energy Technology Data Exchange (ETDEWEB)

    Panikkath, Priyada; Mohanakrishnan, P. [Manipal University, Manipal Centre for Natural Sciences, Karnataka (India)

    2017-03-15

    Thermal neutron capture cross-sections and resonance integrals of {sup 139}La(n, γ){sup 140}La and {sup 140}Ce (n, γ){sup 141}Ce are measured with respect to reference reactions {sup 197}Au(n, γ){sup 198}Au and {sup 55}Mn(n, γ){sup 56}Mn using the neutron activation technique. Measurements are carried out using neutrons from an Am-Be source located inside a concrete bunker. Two different methods are used for determining self-shielding factors of activation foils as well as for finding the epithermal neutron spectrum shape factor. For {sup 139}La with reference to {sup 197}Au and {sup 55}Mn the measured thermal cross sections are 9.24 ± 0.25 b and 9.28 ± 0.37 b, respectively, while the measured resonance integrals are 12.18 ± 0.67 b and 11.81 ± 0.94 b, respectively. For {sup 140}Ce with reference to {sup 197}Au and {sup 55}Mn the measured thermal cross sections are 0.44 ± 0.01 b and 0.44 ± 0.02 b, respectively, while the measured resonance integrals are 0.55 ± 0.03 b and 0.54 ± 0.04 b, respectively. The present measurements are compared with earlier measurements and evaluations. Presently estimated values confirm the established {sup 139}La(n, γ){sup 140}La cross-sections. The presently measured thermal capture cross-section {sup 140}Ce(n, γ){sup 141}Ce, though lower than the evaluated data, is having higher accuracy compared to previous measurements with large uncertainties. The resonance integral measured is higher (like most previous measurements) than most evaluations requiring a revision of the evaluated data. (orig.)

  12. Thermal neutrons thermoluminescence dosimetry using CaF2 + KBr e CaSO4: Dy + Br

    International Nuclear Information System (INIS)

    Leite, A.M.P.

    1979-01-01

    Cold-pressed samples of CaF 2 + KBr and CaSO 4 :Dy + KBr have been used in the thermal neutron detection by the thermoluminescence technique. The amount of 100 mg of the TL phosphor added to 80 mg of KBr showed to be the optimum mixture regarding sensitivity as well as the handling of the dosimeters. The detection is based on the self-irradiation of the phosphor by the Br isotopes activated by exposure to a neutron-gama field. The prompt dose and consequentely the gama contribution are erased by post-irradiation thermal annealing. A linear dependence has been found between the TL self-induced signal and the thermal neutron flux in the range 10 6 n.cm -2 .seg -1 -10 -12 n.cm -2 .seg -1 . The minimum detectable fluence has benn determined as 10 9 n.cm -2 and 10 6 n.cm -2 using pellets of CaF 2 + KBr and CaSO 4 :Dy + KBr, respectively. The main results suggest the use of CaSO 4 :Dy + KBr pellets and TL as a complementary technique for thermal neutron detection. (author) [pt

  13. Self-Shielding Of Transmission Lines

    Energy Technology Data Exchange (ETDEWEB)

    Christodoulou, Christos [Univ. of New Mexico, Albuquerque, NM (United States)

    2017-03-01

    The use of shielding to contend with noise or harmful EMI/EMR energy is not a new concept. An inevitable trade that must be made for shielding is physical space and weight. Space was often not as much of a painful design trade in older larger systems as they are in today’s smaller systems. Today we are packing in an exponentially growing number of functionality within the same or smaller volumes. As systems become smaller and space within systems become more restricted, the implementation of shielding becomes more problematic. Often, space that was used to design a more mechanically robust component must be used for shielding. As the system gets smaller and space is at more of a premium, the trades starts to result in defects, designs with inadequate margin in other performance areas, and designs that are sensitive to manufacturing variability. With these challenges in mind, it would be ideal to maximize attenuation of harmful fields as they inevitably couple onto transmission lines without the use of traditional shielding. Dr. Tom Van Doren proposed a design concept for transmission lines to a class of engineers while visiting New Mexico. This design concept works by maximizing Electric field (E) and Magnetic Field (H) field containment between operating transmission lines to achieve what he called “Self-Shielding”. By making the geometric centroid of the outgoing current coincident with the return current, maximum field containment is achieved. The reciprocal should be true as well, resulting in greater attenuation of incident fields. Figure’s 1(a)-1(b) are examples of designs where the current centroids are coincident. Coax cables are good examples of transmission lines with co-located centroids but they demonstrate excellent field attenuation for other reasons and can’t be used to test this design concept. Figure 1(b) is a flex circuit design that demonstrate the implementation of self-shielding vs a standard conductor layout.

  14. Developing light nano-composites with improved mechanical properties for neutron shielding

    Energy Technology Data Exchange (ETDEWEB)

    Jamali, F. [Shiraz Univ. of Medical Sciences (Iran, Islamic Republic of). School of Medicine; Mortazavi, S.M.J. [Shiraz Univ. of Medical Sciences (Iran, Islamic Republic of). Dept. of Medical Physics and Medical Engineering; Shiraz Univ. of Medical Sciences (Iran, Islamic Republic of). The Center for Radiological Research; Kardan, M. [Nuclear Science and Technology Institute, Tehran (Iran, Islamic Republic of). Radiation Application School; Mosleh-Shirazi, M.A. [Shiraz Univ. of Medical Sciences (Iran, Islamic Republic of). Radiotherapy Dept.; Sina, S. [Shiraz Univ. of Medical Sciences (Iran, Islamic Republic of). Radiation Research Center; Rahpeyma, J.

    2017-12-15

    Although radiation exposures in manned space missions are normally below the limits recommended to NASA by NCRP, in long-duration deep space exploratory missions astronauts may receive relatively high doses of ionizing radiation. Novel light polyethylene-based composites can be considered as effective radiation shields in space explorations. However, normally these composites cannot provide desired mechanical properties. Over the past several years our laboratories have focused on developing efficient methods for both physical and biological protection of the crew in long term space missions. In this study carbon nanotubes and either nano-sized or micro-sized boron carbide (B{sub 4}C) fillers were incorporated into the continuous phase of low density polyethylene (LDPE). In the next phase, the mechanical characteristics of the composites as well as their neutron attenuation properties were studied. Findings of this study indicated enhanced mechanical properties accompanied by an enhanced shielding efficiency for neutrons at some specific weight fraction of the fillers.

  15. Measurement of thermal neutron fluence with CaSO4 thermoluminescent phosphors

    International Nuclear Information System (INIS)

    Liu Jinhua; Su Jingling; Wei Zemin

    1984-01-01

    During neutron irradiation, some TL phosphors were activated. After leaving the irradiation field the TL phosphor produced self-irradiation. The TL output of self-dose was only related to the original neutron fluence and independent of the γ-radiation. Several CaSO 4 TL phosphors were made. They were CaSO 4 :Dy, CaSO 4 :Dy-Teflon, CaSO 4 :Dy mixed with Dy 2 O 3 , CaSO 4 :Mn mixed with Dy 2 O 3 . The linearity, and lower detection limits of these TL phosphors were measured. The thermal neutron response of CaSO 4 :Mn mixed with Dy 2 O 3 was 64 R/(10 10 cm -2 ) and the lower detection limit was 1.3x10 5 cm -2

  16. MPACT Subgroup Self-Shielding Efficiency Improvements

    International Nuclear Information System (INIS)

    Stimpson, Shane; Liu, Yuxuan; Collins, Benjamin S.; Clarno, Kevin T.

    2016-01-01

    Recent developments to improve the efficiency of the MOC solvers in MPACT have yielded effective kernels that loop over several energy groups at once, rather that looping over one group at a time. These kernels have produced roughly a 2x speedup on the MOC sweeping time during eigenvalue calculation. However, the self-shielding subgroup calculation had not been reevaluated to take advantage of these new kernels, which typically requires substantial solve time. The improvements covered in this report start by integrating the multigroup kernel concepts into the subgroup calculation, which are then used as the basis for further extensions. The next improvement that is covered is what is currently being termed as ''Lumped Parameter MOC''. Because the subgroup calculation is a purely fixed source problem and multiple sweeps are performed only to update the boundary angular fluxes, the sweep procedure can be condensed to allow for the instantaneous propagation of the flux across a spatial domain, without the need to sweep along all segments in a ray. Once the boundary angular fluxes are considered to be converged, an additional sweep that will tally the scalar flux is completed. The last improvement that is investigated is the possible reduction of the number of azimuthal angles per octant in the shielding sweep. Typically 16 azimuthal angles per octant are used for self-shielding and eigenvalue calculations, but it is possible that the self-shielding sweeps are less sensitive to the number of angles than the full eigenvalue calculation.

  17. Nuclear reactions and self-shielding effects of gamma-ray database for nuclear materials

    Energy Technology Data Exchange (ETDEWEB)

    Fujita, Mitsutane; Noda, Tetsuji [National Research Institute for Metals, Tsukuba, Ibaraki (Japan)

    2001-03-01

    A database for transmutation and radioactivity of nuclear materials is required for selection and design of materials used in various nuclear reactors. The database based on the FENDL/A-2.0 on the Internet and the additional data collected from several references has been developed in NRIM site of 'Data-Free-Way' on the Internet. Recently, the function predicted self-shielding effect of materials for {gamma}-ray was added to this database. The user interface for this database has been constructed for retrieval of necessary data and for graphical presentation of the relation between the energy spectrum of neutron and neutron capture cross section. It is demonstrated that the possibility of chemical compositional change and radioactivity in a material caused by nuclear reactions can be easily retrieved using a browser such as Netscape or Explorer. (author)

  18. Nuclear reactions and self-shielding effects of gamma-ray database for nuclear materials

    International Nuclear Information System (INIS)

    Fujita, Mitsutane; Noda, Tetsuji

    2001-01-01

    A database for transmutation and radioactivity of nuclear materials is required for selection and design of materials used in various nuclear reactors. The database based on the FENDL/A-2.0 on the Internet and the additional data collected from several references has been developed in NRIM site of 'Data-Free-Way' on the Internet. Recently, the function predicted self-shielding effect of materials for γ-ray was added to this database. The user interface for this database has been constructed for retrieval of necessary data and for graphical presentation of the relation between the energy spectrum of neutron and neutron capture cross section. It is demonstrated that the possibility of chemical compositional change and radioactivity in a material caused by nuclear reactions can be easily retrieved using a browser such as Netscape or Explorer. (author)

  19. Health physics aspects in disposal of self powered neutron detectors

    International Nuclear Information System (INIS)

    Deokar, D.V.; Tibrewala, S.K.; Singh, K.K.; Purohit, R.G.; Tripathi, R.M.

    2014-01-01

    Self Powered Neutron Detectors (SPNDs) are being used in reactor core for neutron flux measurement at Nuclear Power Plants. After their useful life, SPNDs are replaced and are disposed off in Tile holes. The Cobalt SPNDs having activity in the range of 35 to 160 TBq were encompassed in carbon steel canister. The canister having dose 25 to 50 Sv/h at 1 meter were transported in shielded flask for disposal in specially designed Tile hole at Solid Waste Management Facility (SWMF) at Tarapur. To keep personal exposures As Low As Reasonably Achievable (ALARA) the disposal operation was carried out remotely from a shielded cabin placed at a distance of 50 meter from the disposal site. During the disposal radiation measurements were carried out remotely by installing radiations monitors at a distance of 10 m, 25 m, and 50 m from the Tile hole. Estimations of radiation levels were carried out before jobs were taken up. Disposal of 70 numbers of Cobalt SPNDs was carried out by implementing ALARA. The decrease in collective dose is achieved due to improved operational practices, mock-up trials, effective monitoring program and safety compliance at various stages of operation

  20. EL-2 reactor: Thermal neutron flux distribution

    International Nuclear Information System (INIS)

    Rousseau, A.; Genthon, J.P.

    1958-01-01

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

  1. Two-dimensional shielding benchmarks for iron at YAYOI, (1)

    International Nuclear Information System (INIS)

    Oka, Yoshiaki; An, Shigehiro; Kasai, Shigeru; Miyasaka, Shun-ichi; Koyama, Kinji.

    The aim of this work is to assess the collapsed neutron and gamma multigroup cross sections for two dimensional discrete ordinate transport code. Two dimensional distributions of neutron flux and gamma ray dose through a 70cm thick and 94cm square iron shield were measured at the fast neutron source reactor ''YAYOI''. The iron shield was placed over the lead reflector in the vertical experimental column surrounded by heavy concrete wall. The detectors used in this experiment were threshold detectors In, Ni, Al, Mg, Fe and Zn, sandwitch resonance detectors Au, W and Co, activation foils Au for neutrons and thermoluminescence detectors for gamma ray dose. The experimental results were compared with the calculated ones by the discrete ordinate transport code ANISN and TWOTRAN. The region-wise, coupled neutron-gamma multigroup cross-sections (100n+20gamma, EURLIB structure) were generated from ENDF/B-IV library for neutrons and POPOP4 library for gamma-ray production cross-sections by using the code system RADHEAT. The effective microscopic neutron cross sections were obtained from the infinite dilution values applying ABBN type self-shielding factors. The gamma ray production multigroup cross-sections were calculated from these effective microscopic neutron cross-sections. For two-dimensional calculations the group constants were collapsed into 10 neutron groups and 3 gamma groups by using ANISN. (auth.)

  2. Semiconductor Thermal Neutron Detector

    Directory of Open Access Journals (Sweden)

    Toru Aoki

    2014-02-01

    Full Text Available The  CdTe  and  GaN  detector  with  a  Gd  converter  have  been developed  and  investigated  as  a  neutron  detector  for neutron  imaging.  The  fabricated  Gd/CdTe  detector  with  the  25  mm  thick  Gd  was  designed  on  the  basis  of  simulation results  of  thermal  neutron  detection  efficiency  and  spatial  resolution.  The  Gd/CdTe  detector  shows  the  detection  of neutron  capture  gamma  ray  emission  in  the  155Gd(n,  g156Gd,  157Gd(n,  g158Gd  and  113Cd(n,  g114Cd  reactions  and characteristic X-ray emissions due to conversion-electrons generated inside the Gd film. The observed efficient thermal neutron detection with the Gd/CdTe detector shows its promise in neutron radiography application. Moreover, a BGaN detector has also investigated to separate neutron signal from gamma-ray clearly. 

  3. 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.

  4. Propagation of thermal neutrons in mock-up screw-shaped steel elements with water protection; Propagation des neutrons thermiques dans des fausses cartouches d'acier en helice dans une protection d'eau. Programme tournesol 3

    Energy Technology Data Exchange (ETDEWEB)

    Devillers, C L; Lanore, J M [Commissariat a l' Energie Atomique, Fontenay-aux-Roses (France). Centre d' Etudes Nucleaires

    1967-07-01

    This report treats the streaming of thermal neutrons in a cylindrical duct in light water. The duct contains a spiral iron shield. Transmission and reflection matrices are used to describe the probabilities for the thermal neutrons to be absorbed or to be scattered on the surfaces. The neutron paths across the void are represented by geometrical matrices. The numerical resolution is performed by the Monte-Carlo method. (authors) [French] Dans ce rapport on traite un probleme de fuites de neutrons thermiques dans un canal cylindrique plonge dans l'eau et obture par un ecran helicoidal en acier. On utilise des matrices de transmission-reflexion pour decrire les probabilites d'absorption et de diffusion des neutrons sur les parois et l'helicoide et des matrices de correspondance geometrique pour representer la propagation dans le vide. La resolution numerique se fait par une methode de Monte-Carlo. (auteur)

  5. Neutron Thermalization and Reactor Spectra. Vol. II. Proceedings of the Symposium on Neutron Thermalization and Reactor Spectra

    International Nuclear Information System (INIS)

    1968-01-01

    Proceedings of a Symposium organized by the IAEA and held at Ann Arbor, Michigan, USA, 17 - 21 July 1967. The meeting was attended by 143 participants from 24 Member States and one international organization. Contents: (Vol.I) Theory of neutron thermalization (15 papers); Scattering law (20 papers); Angular, space, temperature and time dependence of neutron spectra (9 papers). (Vol.II) Measurement of thermal neutron spectra and spectral indices, and comparison with theory (17 papers); Time-dependent problems in neutron thermalization (12 papers). Each paper is in its original language (61 English, 1 French and 11 Russian) and is preceded by an abstract in English with one in the original language if this is not English. Discussions are in English.

  6. Comparative measurements of independent yields of 239Pu fission fragments induced by thermal and resonance neutrons

    International Nuclear Information System (INIS)

    Gundorin, N.A.; Kopach, Y.N.; Telezhnikov, S.A.

    1994-01-01

    The independent yields of 239 Pu fission fragments by means of gamma-spectroscopy method were measured for light and heavy groups on the IBR-30 reactor in Dubna. Comparative analysis of experimental data for fission induced by thermal and resonance neutrons was performed. The possibilities to increase the measurement's precision consist of the employment of a HPGe detector with high efficiency and its open-quotes activeclose quotes shielding in the gamma spectrometer, as well as a high speed electronics system. In this way the number of identified fragments will be increased and independent yields will be measured to a precision of 1-3%. Measurements at the source with shorter neutron pulse duration to increase neutron energy resolution will be possible after the reconstruction of a modern neutron source in Dubna in accordance with the IREN project

  7. Boron filled siloxane polymers for radiation shielding

    Science.gov (United States)

    Labouriau, Andrea; Robison, Tom; Shonrock, Clinton; Simmonds, Steve; Cox, Brad; Pacheco, Adam; Cady, Carl

    2018-03-01

    The purpose of the present work was to evaluate changes to structure-property relationships of 10B filled siloxane-based polymers when exposed to nuclear reactor radiation. Highly filled polysiloxanes were synthesized with the intent of fabricating materials that could shield high neutron fluences. The newly formulated materials consisted of cross-linked poly-diphenyl-methylsiloxane filled with natural boron and carbon nanofibers. This polymer was chosen because of its good thermal and chemical stabilities, as well as resistance to ionizing radiation thanks to the presence of aromatic groups in the siloxane backbone. Highly isotopically enriched 10B filler was used to provide an efficient neutron radiation shield, and carbon nanofibers were added to improve mechanical strength. This novel polymeric material was exposed in the Annular Core Research Reactor (ACRR) at Sandia National Labs to five different neutron/gamma fluxes consisting of very high neutron fluences within very short time periods. Thermocouples placed on the specimens recorded in-situ temperature changes during radiation exposure, which agreed well with those obtained from our MCNP simulations. Changes in the microstructural, thermal, chemical, and mechanical properties were evaluated by SEM, DSC, TGA, FT-IR NMR, solvent swelling, and uniaxial compressive load measurements. Our results demonstrate that these newly formulated materials are well-suitable to be used in applications that require exposure to different types of ionizing conditions that take place simultaneously.

  8. Proposal for a radiation shielding study aiming the implantation of neutrons beam shutter in the J-9 radiation channel of the Argonauta reactor of the Nuclear Engineering Institute

    Energy Technology Data Exchange (ETDEWEB)

    Xavier, Larissa R.P.; Cardoso, Domingos D’Oliveira, E-mail: larissa.xavier@cnen.gov.br, E-mail: domingosoliveiralvr71@gmail.com [Instituto Militar de Engenharia (IME), Rio de Janeiro, RJ (Brazil); Ferreira, Francisco José de Oliveira; Voi, Dante Luiz, E-mail: fferreira@ien.gov.br, E-mail: dante@ien.gov.br [Instituto de Engenharia Nuclear (IEN/CNEN-RJ), Rio de Janeiro, RJ (Brazil)

    2017-07-01

    Argonauta, the only nuclear research reactor situated in Rio de Janeiro, located at the Institute of Nuclear Engineering (IEN), regularly serves a network of users focused on research and development, and also provides its infrastructure for experimental classes and completion work course. Due to increasing demand for non-destructive thermal neutron assays and production of radioisotopes, there is a search for new procedures and/or devices that optimize users' exposure to neutrons. The implementation of mechanisms that allow access to the irradiation channels without the reactor being turned off and with a shielding configuration that limits the occupational doses at this location is very useful for the operation of the reactor. In order to achieve this, the present work proposes the establishment of a neutron beam shutter of the J-9 irradiation channel of the IEN's Argonauta reactor. In a first step, experimental measurements were made in the irradiation channel of the reactor using a BF3 detector, which is coupled to a spectrometer. In this phase, the neutron beam was aligned to the spectrometer, and different materials were used as shields, aiming the attenuation of the beam. To validate and/or change the configuration of the barrier that best meets the material irradiation needs, a second planned phase is involving the neutron flux simulation of the reactor and the various shields with different boundary conditions using the particle transport code, Monte Carlo N-Particle Extended (MCNP- X). (author)

  9. Proposal for a radiation shielding study aiming the implantation of neutrons beam shutter in the J-9 radiation channel of the Argonauta reactor of the Nuclear Engineering Institute

    International Nuclear Information System (INIS)

    Xavier, Larissa R.P.; Cardoso, Domingos D’Oliveira; Ferreira, Francisco José de Oliveira; Voi, Dante Luiz

    2017-01-01

    Argonauta, the only nuclear research reactor situated in Rio de Janeiro, located at the Institute of Nuclear Engineering (IEN), regularly serves a network of users focused on research and development, and also provides its infrastructure for experimental classes and completion work course. Due to increasing demand for non-destructive thermal neutron assays and production of radioisotopes, there is a search for new procedures and/or devices that optimize users' exposure to neutrons. The implementation of mechanisms that allow access to the irradiation channels without the reactor being turned off and with a shielding configuration that limits the occupational doses at this location is very useful for the operation of the reactor. In order to achieve this, the present work proposes the establishment of a neutron beam shutter of the J-9 irradiation channel of the IEN's Argonauta reactor. In a first step, experimental measurements were made in the irradiation channel of the reactor using a BF3 detector, which is coupled to a spectrometer. In this phase, the neutron beam was aligned to the spectrometer, and different materials were used as shields, aiming the attenuation of the beam. To validate and/or change the configuration of the barrier that best meets the material irradiation needs, a second planned phase is involving the neutron flux simulation of the reactor and the various shields with different boundary conditions using the particle transport code, Monte Carlo N-Particle Extended (MCNP- X). (author)

  10. Induced radioactivity in Bevatron concrete radiation shielding blocks

    International Nuclear Information System (INIS)

    Moeller, G.C.; Donahue, R.J.

    1994-07-01

    The Bevatron accelerated protons up to 6.2 GeV and heavy ions up to 2.1 GeV/amu. It operated from 1954 to 1993. Radioactivity was induced in some concrete radiation shielding blocks by prompt radiation. Prompt radiation is primarily neutrons and protons that were generated by the Bevatron's primary beam interactions with targets and other materials. The goal was to identify the gamma-ray emitting nuclides (t 1/2 > 0.5 yr) that could be present in the concrete blocks and estimate the depth at which the maximum radioactivity presently occurs. It is shown that the majority of radioactivity was produced via thermal neutron capture by trace elements present in concrete. The depth of maximum thermal neutron flux, in theory, corresponds with the depth of maximum induced activity. To estimate the depth at which maximum activity occurs in the concrete blocks, the LAHET Code System was used to calculate the depth of maximum thermal neutron flux. The primary beam interactions that generate the neutrons are also modeled by the LAHET Code System

  11. Adapting an x-ray/debris shield to the cascade ICF power plant: Neutronics issues

    International Nuclear Information System (INIS)

    Tobin, M.T.

    1990-01-01

    A neutronics analysis has been carried out to determine the effects on the Cascade ICF reactor concept of adding a solid-lithium x-ray and debris shield to each ICF capsule. Results indicate that tritium breeding in LiAlO 2 is possible with a modest isotopic enhancement in 6 Li (to 15%). The shallow-burial index is greater than 1 (indicating that deep burial may be required) if the blanket is kept in the reactor for more than 2.5 yr. Nine percent of the total thermal power is unrecoverable. Parts of the chamber wall may require replacement once during the reactor life due to radiation damage. Part of the SiC chamber end cap must be replaced annually. The reactor may not require any nuclear-grade construction. 20 refs., 4 figs., 3 tabs

  12. Thermal Neutron Capture and Thermal Neutron Burn-up of K isomeric state of 177mLu: a way to the Neutron Super-Elastic Scattering cross section

    International Nuclear Information System (INIS)

    Roig, O.; Belier, G.; Meot, V.; Daugas, J.-M.; Romain, P.; Aupiais, J.; Jutier, Ch.; Le Petit, G.; Letourneau, A.; Marie, F.; Veyssiere, Ch.

    2006-01-01

    Thermal neutron radiative capture and burn-up measurements of the K isomeric state in 177Lu form part of an original method to indirectly obtain the neutron super-elastic scattering cross section at thermal energy. Neutron super-elastic scattering, also called neutron inelastic acceleration, occurs during the neutron collisions with an excited nuclear level. In this reaction, the nucleus could partly transfer its excitation energy to the scattered neutron

  13. Study of a transportable neutron radiography system

    International Nuclear Information System (INIS)

    Souza, S.N.A. de.

    1991-05-01

    This work presents a study a transportable neutron radiography system for a 185 GBq 241 Am-Be (α, η) source with a neutron yield roughly 1,25 x 10 7 n/s. Studies about moderation, collimation and shielding are showed. In these studies, a calculation using Transport Theory was carried out by means of transport codes ANISN and DOT (3.5). Objectives were: to obtain a maximum and more homogeneous thermal neutron flux in the collimator outlet to the image plain, and an adequate radiation shielding to attend radiological protection rules. With the presented collimator, it was possible to obtain for the thermal neutron flux, at the collimator outlet and next to the image plain, a L/D ratio of 14, for neutron fluxes up to 4,09 x 10 2 n.cm -2 .s -1 . Considering the low intensity of the source, it is a good value. Studies have also been carried out for L/D ratios of 22 and 30, giving thermal neutron fluxes at the image plain of 1,27 x 10 2 n.cm -2 .s -1 and 2,65 x 10 2 n.cm -2 .s -1 , respectively. (author). 30 refs, 39 figs, 9 tabs

  14. Resonance self-shielding calculation with regularized random ladders

    Energy Technology Data Exchange (ETDEWEB)

    Ribon, P.

    1986-01-01

    The straightforward method for calculation of resonance self-shielding is to generate one or several resonance ladders, and to process them as resolved resonances. The main drawback of Monte Carlo methods used to generate the ladders, is the difficulty of reducing the dispersion of data and results. Several methods are examined, and it is shown how one (a regularized sampling method) improves the accuracy. Analytical methods to compute the effective cross-section have recently appeared: they are basically exempt from dispersion, but are inevitably approximate. The accuracy of the most sophisticated one is checked. There is a neutron energy range which is improperly considered as statistical. An examination is presented of what happens when it is treated as statistical, and how it is possible to improve the accuracy of calculations in this range. To illustrate the results calculations have been performed in a simple case: nucleus /sup 238/U, at 300 K, between 4250 and 4750 eV.

  15. Neutron flux measurements at the TRIGA reactor in Vienna for the prediction of the activation of the biological shield

    International Nuclear Information System (INIS)

    Merz, Stefan; Djuricic, Mile; Villa, Mario; Boeck, Helmuth; Steinhauser, Georg

    2011-01-01

    The activation of the biological shield is an important process for waste management considerations of nuclear facilities. The final activity can be estimated by modeling using the neutron flux density rather than the radiometric approach of activity measurements. Measurement series at the TRIGA reactor Vienna reveal that the flux density next to the biological shield is in the order of 10 9 cm -2 s -1 at maximum power; but it is strongly influenced by reactor installations. The data allow the estimation of the final waste categorization of the concrete according to the Austrian legislation. - Highlights: → Neutron activation is an important process for the waste management of nuclear facilities. → Biological shield of the TRIGA reactor Vienna has been topic of investigation. → Flux values allow a categorization of the concrete concerning radiation protection legislation. → Reactor installations are of great importance as neutron sources into the biological shield. → Every installation shows distinguishable flux profiles.

  16. New Improvements in Mixture Self-Shielding Treatment with APOLLO2 Code

    International Nuclear Information System (INIS)

    Coste-Delclaux, M.

    2006-01-01

    Full text of the presentation follows: APOLLO2 is a modular multigroup transport code developed at the CEA in Saclay (France). Previously, the self-shielding module could only treat one resonant isotope mixed with moderator isotopes. Consequently, the resonant mixture self-shielding treatment was an iterative one. Each resonant isotope of the mixture was treated separately, the other resonant isotopes of the mixture being then considered as moderator isotopes, that is to say non-resonant isotopes. This treatment could be iterated. Recently, we have developed a new method that consists in treating the resonant mixture as a unique entity. A main feature of APOLLO2 self-shielding module is that some implemented models are very general and therefore very powerful and versatile. We can give, as examples, the use of probability tables in order to describe the microscopic cross-section fluctuations or the TR slowing-down model that can deal with any resonance shape. The self-shielding treatment of a resonant mixture was developed essentially thanks to these two models. The goal of this paper is to describe the improvements on the self-shielding treatment of a resonant mixture and to present, as an application, the calculation of the ATRIUM-10 BWR benchmark. We will conclude by some prospects on remaining work in the self-shielding domain. (author)

  17. Monte Carlo validation of self shielding and void effect calculations

    International Nuclear Information System (INIS)

    Tellier, H.; Coste, M.; Raepsaet, C.; Soldevila, M.; Van der Gucht, C.

    1995-01-01

    The self shielding validation and the void effect are studied with Monte Carlo method. The satisfactory comparison obtained between the APOLLO 2 results of the self shielding effect and the TRIPOLI and MCNP results allows us to be confident in the multigroup transport code. (K.A.)

  18. Concrete shielding of neutron radiations of plasma focus and dose examination by FLUKA

    Science.gov (United States)

    Nemati, M. J.; Amrollahi, R.; Habibi, M.

    2013-07-01

    Plasma Focus (PF) is among those devices which are used in plasma investigations, but this device produces some dangerous radiations after each shot, which generate a hazardous area for the operators of this device; therefore, it is better for the operators to stay away as much as possible from the area, where plasma focus has been placed. In this paper FLUKA Monte Carlo simulation has been used to calculate radiations produced by a 4 kJ Amirkabir plasma focus device through different concrete shielding concepts with various thicknesses (square, labyrinth and cave concepts). The neutron yield of Amirkabir plasma focus at varying deuterium pressure (3-9 torr) and two charging voltages (11.5 and 13.5 kV) is (2.25 ± 0.2) × 108 neutrons/shot and (2.88 ± 0.29) × 108 neutrons/shot of 2.45 MeV, respectively. The most influential shield for the plasma focus device among these geometries is the labyrinth concept on four sides and the top with 20 cm concrete.

  19. 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

  20. 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

  1. Simbol-X Mirror Module Thermal Shields: II-Small Angle X-Ray Scattering Measurements

    Science.gov (United States)

    Barbera, M.; Ayers, T.; Collura, A.; Nasillo, G.; Pareschi, G.; Tagliaferri, G.

    2009-05-01

    The formation flight configuration of the Simbol-X mission implies that the X-ray mirror module will be open to Space on both ends. In order to reduce the power required to maintain the thermal stability and, therefore, the high angular resolution of the shell optics, a thin foil thermal shield will cover the mirror module. Different options are presently being studied for the foil material of these shields. We report results of an experimental investigation conducted to verify that the scattering of X-rays, by interaction with the thin foil material of the thermal shield, will not significantly affect the performances of the telescope.

  2. Simbol-X Mirror Module Thermal Shields: II-Small Angle X-Ray Scattering Measurements

    International Nuclear Information System (INIS)

    Barbera, M.; Ayers, T.; Collura, A.; Nasillo, G.; Pareschi, G.; Tagliaferri, G.

    2009-01-01

    The formation flight configuration of the Simbol-X mission implies that the X-ray mirror module will be open to Space on both ends. In order to reduce the power required to maintain the thermal stability and, therefore, the high angular resolution of the shell optics, a thin foil thermal shield will cover the mirror module. Different options are presently being studied for the foil material of these shields. We report results of an experimental investigation conducted to verify that the scattering of X-rays, by interaction with the thin foil material of the thermal shield, will not significantly affect the performances of the telescope.

  3. RADSHI: shielding calculation program for different geometries sources

    International Nuclear Information System (INIS)

    Gelen, A.; Alvarez, I.; Lopez, H.; Manso, M.

    1996-01-01

    A computer code written in pascal language for IBM/Pc is described. The program calculates the optimum thickness of slab shield for different geometries sources. The Point Kernel Method is employed, which enables the obtention of the ionizing radiation flux density. The calculation takes into account the possibility of self-absorption in the source. The air kerma rate for gamma radiation is determined, and with the concept of attenuation length through the equivalent attenuation length the shield is obtained. The scattering and the exponential attenuation inside the shield material is considered in the program. The shield materials can be: concrete, water, iron or lead. It also calculates the shield for point isotropic neutron source, using as shield materials paraffin, concrete or water. (authors). 13 refs

  4. Structural Design and Thermal Analysis for Thermal Shields of the MICE Coupling Magnets

    International Nuclear Information System (INIS)

    Green, Michael A.; Pan, Heng; Liu, X.K.; Wang, Li; Wu, Hong; Chen, A.B.; Guo, X.L.

    2009-01-01

    A superconducting coupling magnet made from copper matrix NbTi conductors operating at 4 K will be used in the Muon Ionization Cooling Experiment (MICE) to produce up to 2.6 T on the magnet centerline to keep the muon beam within the thin RF cavity indows. The coupling magnet is to be cooled by two cryocoolers with a total cooling capacity of 3 W at 4.2 K. In order to keep a certain operating temperature margin, the most important is to reduce the heat leakage imposed on cold surfaces of coil cold mass assembly. An ntermediate temperature shield system placed between the coupling coil and warm vacuum chamber is adopted. The shield system consists of upper neck shield, main shields, flexible connections and eight supports, which is to be cooled by the first stage cold heads of two ryocoolers with cooling capacity of 55 W at 60 K each. The maximum temperature difference on the shields should be less than 20 K, so the thermal analyses for the shields with different thicknesses, materials, flexible connections for shields' cooling and structure design for heir supports were carried out. 1100 Al is finally adopted and the maximum temperature difference is around 15 K with 4 mm shield thickness. The paper is to present detailed analyses on the shield system design.

  5. Analysis of a shield design for a DT neutron generator test facility.

    Science.gov (United States)

    Chichester, D L; Pierce, G D

    2007-10-01

    Independent numerical simulations have been performed using the MCNP5 and SCALE5 radiation transport codes to evaluate the effectiveness of a concrete facility designed to shield personnel from neutron radiation emitted from DT neutron generators. The analysis considered radiation source terms of 14.1 MeV monoenergetic neutrons located at three discrete locations within the two test vaults in the facility, calculating neutron and photon dose rates at 44 locations around the facility using both codes. In addition, dose rate contours were established throughout the facility using the MCNP5 mesh tally feature. Neutron dose rates calculated outside of the facility are predicted to be below 0.01 mrem/h at all locations when all neutron generator source terms are operating within the facility. Similarly, the neutron dose rate in one empty test vault when the adjacent test vault is being utilized is also less then 0.01 mrem/h. For most calculation locations outside the facility the photon dose rates were less then the neutron dose rates by a factor of 10 or more.

  6. Attenuation of thermal neutron through graphite

    International Nuclear Information System (INIS)

    Adib, M.; Ismaail, H.; Fathaallah, M.; Abbas, Y.; Habib, N.; Wahba, M.

    2004-01-01

    Calculation of the nuclear capture, thermal diffuse and Bragg scattering cross-sections as a function of graphite temperature and crystalline from for neutron energies from 1 me V< E<10 eV were carried out. Computer programs have been developed which allow calculation for the graphite hexagonal closed-pack structure in its polycrystalline form and pyrolytic one. I The calculated total cross-section for polycrystalline graphite were compared with the experimental values. An overall agreement is indicated between the calculated values and experimental ones. Agreement was also obtained for neutron cross-section measured for oriented pyrolytic graphite at room and liquid nitrogen temperatures. A feasibility study for use of graphite in powdered form as a cold neutron filter is details. The calculated attenuation of thermal neutrons through large mosaic pyrolytic graphite show that such crystals can be used effectively as second order filter of thermal neutron beams and that cooling improve their effectiveness

  7. Characterization and Neutron Shielding Behavior of Dehydrated Magnesium Borate Minerals Synthesized via Solid-State Method

    Directory of Open Access Journals (Sweden)

    Azmi Seyhun Kipcak

    2013-01-01

    Full Text Available Magnesium borates are one of the major groups of boron minerals that have good neutron shielding performance. In this study, dehydrated magnesium borates were synthesized by solid-state method using magnesium oxide (MgO and boron oxide (B2O3, in order to test their ability of neutron shielding. After synthesizing the dehydrated magnesium borates, characterizations were done by X-ray Diffraction (XRD, fourier transform infrared (FT-IR, Raman spectroscopy, and scanning electron microscopy (SEM. Also boron oxide (B2O3 contents and reaction yields (% were calculated. XRD results showed that seven different types of dehydrated magnesium borates were synthesized. 1000°C reaction temperature, 240 minutes of reaction time, and 3 : 2, 1 : 1 mole ratios of products were selected and tested for neutron transmission. Also reaction yields were calculated between 84 and 88% for the 3 : 2 mole ratio products. The neutron transmission experiments revealed that the 3 : 2 mole ratio of MgO to B2O3 neutron transmission results (0.618–0.655 was better than the ratio of 1 : 1 (0.772–0.843.

  8. Neutron multiplication and shielding problems in PWR spent-fuel shipping casks

    International Nuclear Information System (INIS)

    Devillers, C.

    1976-01-01

    In order to evaluate the degree of accuracy of computational methods used for the shield design of spent-fuel shipping casks, comparisons were made between biological dose rate calculations and measurements at the surface of a cask carrying three PWR fuel assemblies (the fuel being successively wet and dry). The experimental methods used provide ksub(eff) with an accuracy of 0.024. Neutron multiplication coefficients provided by the APOLLO and DOT-3 codes are located within the uncertainty range of the experimentally derived values. The APOLLO plus DOT codes for neutron source calculations and ANISN plus DOT codes for neutron transmission calculations provide neutron dose rate predictions in agreement with measurements to within 10%. The PEPIN 76 code used for deriving fission product γ-rays and the point kernel code MERCURE 4 treating the γ-ray transmission give γ dose rate predictions that generally differ from measurements by less than 25%

  9. Thermal neutron diffusion parameters in homogeneous mixtures

    Energy Technology Data Exchange (ETDEWEB)

    Drozdowicz, K.; Krynicka, E. [Institute of Nuclear Physics, Cracow (Poland)

    1995-12-31

    A physical background is presented for a computer program which calculates the thermal neutron diffusion parameters for homogeneous mixtures of any compounds. The macroscopic absorption, scattering and transport cross section of the mixture are defined which are generally function of the incident neutron energy. The energy-averaged neutron parameters are available when these energy dependences and the thermal neutron energy distribution are assumed. Then the averaged diffusion coefficient and the pulsed thermal neutron parameters (the absorption rare and the diffusion constant) are also defined. The absorption cross section is described by the 1/v law and deviations from this behaviour are considered. The scattering cross section can be assumed as being almost constant in the thermal neutron region (which results from the free gas model). Serious deviations are observed for hydrogen atoms bound in molecules and a special study in the paper is devoted to this problem. A certain effective scattering cross section is found in this case on a base of individual exact data for a few hydrogenous media. Approximations assumed for the average cosine of the scattering angle are also discussed. The macroscopic parameters calculated are averaged over the Maxwellian energy distribution for the thermal neutron flux. An information on the input data for the computer program is included. (author). 10 refs, 4 figs, 5 tabs.

  10. Thermal neutron diffusion parameters in homogeneous mixtures

    International Nuclear Information System (INIS)

    Drozdowicz, K.; Krynicka, E.

    1995-01-01

    A physical background is presented for a computer program which calculates the thermal neutron diffusion parameters for homogeneous mixtures of any compounds. The macroscopic absorption, scattering and transport cross section of the mixture are defined which are generally function of the incident neutron energy. The energy-averaged neutron parameters are available when these energy dependences and the thermal neutron energy distribution are assumed. Then the averaged diffusion coefficient and the pulsed thermal neutron parameters (the absorption rare and the diffusion constant) are also defined. The absorption cross section is described by the 1/v law and deviations from this behaviour are considered. The scattering cross section can be assumed as being almost constant in the thermal neutron region (which results from the free gas model). Serious deviations are observed for hydrogen atoms bound in molecules and a special study in the paper is devoted to this problem. A certain effective scattering cross section is found in this case on a base of individual exact data for a few hydrogenous media. Approximations assumed for the average cosine of the scattering angle are also discussed. The macroscopic parameters calculated are averaged over the Maxwellian energy distribution for the thermal neutron flux. An information on the input data for the computer program is included. (author). 10 refs, 4 figs, 5 tabs

  11. Deep-penetration calculations in concrete and iron for shielding of proton therapy accelerators

    International Nuclear Information System (INIS)

    Sheu, Rong-Jiun; Chen, Yen-Fu; Lin, Uei-Tyng; Jiang, Shiang-Huei

    2012-01-01

    Proton accelerators in the energy range of approximately 200 MeV have become increasingly popular for cancer treatment in recent years. These proton therapy facilities usually involve bulky concrete or iron in their shielding design or accelerator structure. Simple shielding data, such as source terms or attenuation lengths for various proton energies and materials are useful in designing accelerator shielding. Understanding the appropriateness or uncertainties associated with these data, which are largely generated from Monte Carlo simulations, is critical to the quality of a shielding design. This study demonstrated and investigated the problems of deep-penetration calculations on the estimation of shielding parameters through an extensive comparison between the FLUKA and MCNPX calculations for shielding against a 200-MeV proton beam hitting an iron target. Simulations of double-differential neutron production from proton bombardment were validated by comparison with experimental data. For the concrete shielding, the FLUKA calculated depth–dose distributions were consistent with the MCNPX results, except for some discrepancies in backward directions. However, for the iron shielding, if FLUKA is used inappropriately then overestimation of neutron attenuation can be expected as shown by this work because of the multigroup treatment for low-energy neutrons in FLUKA. Two neutron energy group structures, three degrees of self-shielding correction, and two iron compositions were considered in this study. Significant variation of the resulting attenuation lengths indicated the importance of problem-dependent multigroup cross sections and proper modeling of iron composition in deep-penetration calculations.

  12. Cryogenic magnetic coil and superconducting magnetic shield for neutron electric dipole moment searches

    Science.gov (United States)

    Slutsky, S.; Swank, C. M.; Biswas, A.; Carr, R.; Escribano, J.; Filippone, B. W.; Griffith, W. C.; Mendenhall, M.; Nouri, N.; Osthelder, C.; Pérez Galván, A.; Picker, R.; Plaster, B.

    2017-08-01

    A magnetic coil operated at cryogenic temperatures is used to produce spatial, relative field gradients below 6 ppm/cm, stable for several hours. The apparatus is a prototype of the magnetic components for a neutron electric dipole moment (nEDM) search, which will take place at the Spallation Neutron Source (SNS) at Oak Ridge National Laboratory using ultra-cold neutrons (UCN). That search requires a uniform magnetic field to mitigate systematic effects and obtain long polarization lifetimes for neutron spin precession measurements. This paper details upgrades to a previously described apparatus [1], particularly the introduction of super-conducting magnetic shielding and the associated cryogenic apparatus. The magnetic gradients observed are sufficiently low for the nEDM search at SNS.

  13. Development of heat resistant concrete and its application to concrete casks. Improvement of neutron shielding performance of concrete in high temperature environment

    International Nuclear Information System (INIS)

    Owaki, Eiji; Hata, Akihito; Sugihara, Yutaka; Shimojo, Jun; Taniuchi, Hiroaki; Mantani, Kenichi

    2003-01-01

    Heat resistant concrete with hydrogen, which is able to shield neutron at more than 100degC, was developed. Using this new type concrete, a safety concrete cask having the same concept of metal casks was designed and produced. The new type cask omitted the inhalation and exhaust vent of the conventional type concrete casks. The new concrete consists of Portland cement added calcium hydroxide, iron powder and iron fiber. It showed 2.17 g/cm 3 density, 10.8 mass% water content, 1.4 W/(m·K) thermal conductivity at 150degC. Increasing of heat resistance made possible to produce the perfect sealing type structure, which had high shielding performance of radiation no consideration for streaming of radiation. Moreover, a monitor of sealing can be set. General view of concrete casks, outer view of 1/3 scaled model, cask storage system in the world, properties of new developed heat resistant concrete, results of shielding calculation are contained. (S.Y.)

  14. Method and apparatus for determining the dose value of neutrons

    International Nuclear Information System (INIS)

    Burgkhardt, B.; Piesch, E.

    1976-01-01

    A method is provided for determining the dose value of neutrons leaving a body as thermal and intermediate neutrons after having been scattered in the body. A first dose value of thermal and intermediate neutrons is detected on the surface of the body by means of a first detector for neutrons which is shielded against thermal and intermediate neutrons not emerging from the body. A second detector is used to measure a second dose value of the thermal and intermediate neutrons not emerging from the body. A first correction factor based on the first and second values is obtained from a calibration diagram and is applied to the first dose value to determine a first corrected first dose value. 21 Claims, 6 Drawing Figures

  15. Calculation of neutron shielding for a real loaded C-30 cask by code DORT

    International Nuclear Information System (INIS)

    Lacina, J.

    1999-01-01

    Measured neutron dose rates of real loaded C-30 casks for WWER spent fuel assemblies are compared with calculated values in the frame of benchmark calculation task. The part of this benchmark task concerning neutron shielding was calculated. Neutron sources values were taken from data presented by V. Chrapciak during the eighth symposium Atomic Energy Research, Bystrice pod Perstejnem in 1998 and the data about cask from the article of the same author from the Atomic Energy Research working group E meeting at Stolpen in 1998. (Author)

  16. Preparation of the in-house neutron detectors and the software needed to process experimental data

    International Nuclear Information System (INIS)

    Haddad, Kh.; Haj-Hassan, H.; Helal, W.

    2007-04-01

    In - house neutron activation detectors were prepared in this work using pure commercial gold. The neutron self-shielding factors in the foils for both thermal and epithermal neutrons have been determined experimentally. The work shows good results repeatability and good agreement with certified activation monitors. the software KHW for neutron flux measurements using local and standards gold foils was designed and performed locally. it deals as well with irradiated uranium spectrums to calculate some important fission product ratios for neutron flux measurement. Some experiments were performed to investigate the possibility of using uranium, produced in the pilot plant, as fission neutron detector. The results shows the possibility of using fission product ratios to determine the cooling time of the samples. It shows also the possibility of using fission and activation product ratios as an indicators of neutron fluences ratios.(author)

  17. 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.

  18. Radiation shielding design of BNCT treatment room for D-T neutron source.

    Science.gov (United States)

    Pouryavi, Mehdi; Farhad Masoudi, S; Rahmani, Faezeh

    2015-05-01

    Recent studies have shown that D-T neutron generator can be used as a proper neutron source for Boron Neutron Capture Therapy (BNCT) of deep-seated brain tumors. In this paper, radiation shielding calculations have been conducted based on the computational method for designing a BNCT treatment room for a recent proposed D-T neutron source. By using the MCNP-4C code, the geometry of the treatment room has been designed and optimized in such a way that the equivalent dose rate out of the treatment room to be less than 0.5μSv/h for uncontrolled areas. The treatment room contains walls, monitoring window, maze and entrance door. According to the radiation protection viewpoint, dose rate results of out of the proposed room showed that using D-T neutron source for BNCT is safe. Copyright © 2015 Elsevier Ltd. All rights reserved.

  19. Calculation of the neutrons shielding in cyclotron accelerator

    International Nuclear Information System (INIS)

    Ribeiro, Martha S.; Sanches, Matias P.; Rodrigues, Demerval L.

    2000-01-01

    The objective of radioprotection in cyclotron facilities is to reduce the dose levels in the workplaces to classify them like supervised areas. In this way, the radiation dose rates in areas occupied by workers during cyclotron operations should not exceed 7,5 μSv/h. In controlled areas these levels are not observed and some rigorous controls must be exerted by administrative procedures or protection mechanisms. The Cyclotron Laboratory at IPEN-CNEN/SP has a cyclotron model Cyclone 30, 30 MeV, used for research and it is also used for radioisotopes production for medical diagnosis and therapeutical applications. Among them, 123 I, 67 Ga and 18 F can be pointed. When accelerator is operating, failures in perforations and paths that conduce to room accelerator can be occur and thus, the dose levels are higher than that established by law. For this reason, a review for shielding structure was necessary in order to optimize radiation dose. The purpose of this work was to determine the shielding thickness and adequate material to diminish the dose rates in workplaces to a value below 7,5 μSv/h. It was used a method to employ the equivalent dose value in the facility areas for neutrons fluency rate for the principal reactions in target irradiation processes. The purposed shielding for the vault doors ensures dose levels lower than established limits to supervised areas. (author)

  20. Measurements of neutron flux in the RA reactor; Merenje karakteristika neutronskog fluksa u reaktoru RA

    Energy Technology Data Exchange (ETDEWEB)

    Raisic, N [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

    1961-12-15

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

  1. Research on preparation and performance of graphite cement-based materials used for fast neutron shielding

    International Nuclear Information System (INIS)

    Xu Jun; Kang Qing; Shen Zhiqiang; Wang Zhenggang; Wang Zhiqiang

    2014-01-01

    Measurements have been carried out to investigate the 14.8 MeV neutron attenuation properties for 3 kinds of cement-graphite composites. In comparison with the void group, the 14.8 MeV neutron attenuation properties of cement-graphite composites raised not clearly in 8 mm thickness, and drop not remarkably in 40 mm thickness; with the increase of graphite content and the thickness, the 14.8 MeV neutron attenuation properties were enhanced clearly. The data may be useful to the radiation shielding design of neutron. (authors)

  2. Radioprotection shielding for neutrons induced by the reaction (2H (40 MeV, 12C

    Directory of Open Access Journals (Sweden)

    Fadil M.

    2017-01-01

    Full Text Available In the framework of design studies for SPIRAL2, the simulation of the neutron flux generated by 40 MeV deuterons on a thick 12C target was performed and compared to experimental data. The calculation of the dose rate of these neutrons allowed to compare four materials being considered for radioprotection shielding: barites, gypsum, ordinary concrete and heavy concrete. The simulated map of the neutron dose rate in the production building shows a very high dose rate around the neutron source and in the environment of some of the accelerator equipment.

  3. Neutron activation measurements in research reactor concrete shield

    International Nuclear Information System (INIS)

    Zagar, T.; Ravnik, M.; Bozic, M.

    2001-01-01

    The results of activation measurement inside TRIGA research reactor concrete shielding are given. Samples made of ordinary and barytes concrete together with gold and nickel foils were irradiated in the reactor body. Long-lived neutron-induced gamma-ray-emitting radioactive nuclides in the samples were measured with HPGe detector. The most active longlived radioactive nuclides in ordinary concrete samples were found to be 60 Co and 152 Eu and in barytes concrete samples 60 Co, 152 Eu and 133 Ba. Measured activity density of all nuclides was found to decrease almost linearly with depth in logarithmic scale.(author)

  4. Resonance region neutronics of unit cells in fast and thermal reactors

    International Nuclear Information System (INIS)

    Salehi, A.A.; Driscoll, M.J.; Deutsch, O.L.

    1977-05-01

    A method has been developed for generating resonance-self-shielded cross sections based upon an improved equivalence theorem, which appears to allow extension of the self-shielding-factor (Bondarenko f-factor) method, now mainly applied to fast reactors, to thermal reactors as well. The method is based on the use of simple prescriptions for the ratio of coolant-to-fuel region-averaged fluxes, in the equations defining cell averaged cross sections. Linearization of the dependence of these functions on absorber optical thickness is found to be a necessary and sufficient condition for the existence of an equivalence theorem. Results are given for cylindrical, spherical and slab geometries. The functional form of the flux ratio relations is developed from theoretical considerations, but some of the parameters are adjusted to force-fit numerical results. Good agreement over the entire range of fuel and coolant optical thicknesses is demonstrated with numerical results calculated using the ANISN program in the S 8 P 1 option

  5. Theory of neutron resonance cross sections for safety applications

    International Nuclear Information System (INIS)

    Froehner, F.H.

    1992-09-01

    Neutron resonances exert a strong influence on the behaviour of nuclear reactors, especially on their response to the temperature changes accompanying power excursions, and also on the efficiency of shielding materials. The relevant theory of neutron resonance cross sections including the practically important approximations is reviewed, both for the resolved and the unresolved resonance region. Numerical techniques for Doppler broadening of resonances are presented, and the construction of group constants and especially of self-shielding factors for neutronics calculations is outlined. (orig.) [de

  6. Trial production of hyper-thermal neutron generator for Neutron Capture Therapy (NCT) and its radiation properties

    International Nuclear Information System (INIS)

    Sakurai, Yoshinori; Kobayashi, Toru

    1999-01-01

    In NCT, it was at first important to give a cancer portion to radiation dose required for its recovery. By finding out that whole cross-section of water comprising of a living body decreased monotonously with increase of neutron energy from about 100 barn against thermal neutron, became about 40 barn at about 0.5 eV and kept constant to 40 barn till at about 100 eV, application of thermal neutron shifted to higher temperature side, called Hyper thermal neutron, to NCT is proposed. The Hyper thermal neutron radiation can be expected to have similar controllability to that of the thermal neutron radiation. In 1977 fiscal year, a trial Hyper thermal neutron generator was produced on a base of up-to-date investigation results. As a part of property evaluation of the generator, evaluation of energy spectra in the Hyper thermal neutron generated at LINAC by TOF was conducted to confirm shift of the spectra to high temperature side. And, a Fantom experiment at KUR heavy water neutron radiation facility was also conducted to confirm effect of improvement in deep portion dose distribution. (G.K.)

  7. Test of thermal shields for early warning station detectors

    DEFF Research Database (Denmark)

    Petersen, Jesper

    1997-01-01

    The properties of thermal shields around NaI crystal scintillators for early warning stations have been checked in order to assure that external temperature variations cannot influence the stability of the measurements....

  8. Simultaneous thermal neutron decay time and porosity logging system

    International Nuclear Information System (INIS)

    Smith, H.D. Jr.; Smith, M.P.; Schultz, W.E.

    1979-01-01

    A simultaneous pulsed neutron porosity and thermal neutron capture cross section logging system is provided for radiological well logging of subsurface earth formations. A logging tool provided with a 14 MeV pulsed neutron source, an epithermal neutron detector, and a combination gamma ray and fast neutron detector is moved through a borehole. Repetitive bursts of neutrons irradiate the earth formations; and, during the bursts, the fast neutron and epithermal neutron populations are sampled. During the interval between bursts the thermal neutron capture gamma ray population is sampled in two or more time intervals. The fast and epithermal neutron population measurements are combined to provide a measurement of formation porosity phi. The capture gamma ray measurements are combined to provide a simultaneous determination of the thermal neutron capture cross section Σ

  9. Modeling of neutron and photon transport in iron and concrete radiation shields by using Monte Carlo method

    CERN Document Server

    Žukauskaitėa, A; Plukienė, R; Ridikas, D

    2007-01-01

    Particle accelerators and other high energy facilities produce penetrating ionizing radiation (neutrons and γ-rays) that must be shielded. The objective of this work was to model photon and neutron transport in various materials, usually used as shielding, such as concrete, iron or graphite. Monte Carlo method allows obtaining answers by simulating individual particles and recording some aspects of their average behavior. In this work several nuclear experiments were modeled: AVF 65 (AVF cyclotron of Research Center of Nuclear Physics, Osaka University, Japan) – γ-ray beams (1-10 MeV), HIMAC (heavy-ion synchrotron of the National Institute of Radiological Sciences in Chiba, Japan) and ISIS-800 (ISIS intensive spallation neutron source facility of the Rutherford Appleton laboratory, UK) – high energy neutron (20-800 MeV) transport in iron and concrete. The calculation results were then compared with experimental data.compared with experimental data.

  10. Shielding plugs

    International Nuclear Information System (INIS)

    Makishima, Kenji.

    1986-01-01

    Purpose: In shielding plugs of an LMFBR type reactor, to restrain natural convection of heat in an annular space between a thermal shield layer and a shield shell, to prevent the lowering of heat-insulation performance, and to alleviate a thermal stress in a reactor container and the shield shell. Constitution: A ring-like leaf spring split in the direction of height is disposed in an annular space between a thermal shield layer and a shield shell. In consequence, the space is partitioned in the direction of height and, therefore, if axial temperature conditions and space width are the same and the space is low, the natural convection is hard to occur. Thus the rise of upper surface temperature of the shielding plugs can prevent the lowering of the heat insulation performance which will result in the increment of shielding plug cooling capacity, thereby improving reliability. In the meantime, since there is mounted an earthquake-resisting support, the thermal shield layer will move for a slight gap in case of an earthquake, being supported by the earthquake-resisting support, and the movement of the thermal shield layer is restricted, thereby maintaining integrity without increasing the stroke of the ring-like spring. (Kawakami, Y.)

  11. Shielding technology for high energy radiation production facility

    International Nuclear Information System (INIS)

    Lee, Byung Chul; Kim, Heon Il

    2004-06-01

    In order to develop shielding technology for high energy radiation production facility, references and data for high energy neutron shielding are searched and collected, and calculations to obtain the characteristics of neutron shield materials are performed. For the evaluation of characteristics of neutron shield material, it is chosen not only general shield materials such as concrete, polyethylene, etc., but also KAERI developed neutron shields of High Density PolyEthylene (HDPE) mixed with boron compound (B 2 O 3 , H 2 BO 3 , Borax). Neutron attenuation coefficients for these materials are obtained for later use in shielding design. The effect of source shape and source angular distribution on the shielding characteristics for several shield materials is examined. This effect can contribute to create shielding concept in case of no detail source information. It is also evaluated the effect of the arrangement of shield materials using current shield materials. With these results, conceptual shielding design for PET cyclotron is performed. The shielding composite using HDPE and concrete is selected to meet the target dose rate outside the composite, and the dose evaluation is performed by configuring the facility room conceptually. From the result, the proper shield configuration for this PET cyclotron is proposed

  12. Self-shielding of hydrogen in the IGM during the epoch of reionization

    Science.gov (United States)

    Chardin, Jonathan; Kulkarni, Girish; Haehnelt, Martin G.

    2018-04-01

    We investigate self-shielding of intergalactic hydrogen against ionizing radiation in radiative transfer simulations of cosmic reionization carefully calibrated with Lyα forest data. While self-shielded regions manifest as Lyman-limit systems in the post-reionization Universe, here we focus on their evolution during reionization (redshifts z = 6-10). At these redshifts, the spatial distribution of hydrogen-ionizing radiation is highly inhomogeneous, and some regions of the Universe are still neutral. After masking the neutral regions and ionizing sources in the simulation, we find that the hydrogen photoionization rate depends on the local hydrogen density in a manner very similar to that in the post-reionization Universe. The characteristic physical hydrogen density above which self-shielding becomes important at these redshifts is about nH ˜ 3 × 10-3 cm-3, or ˜20 times the mean hydrogen density, reflecting the fact that during reionization photoionization rates are typically low enough that the filaments in the cosmic web are often self-shielded. The value of the typical self-shielding density decreases by a factor of 3 between redshifts z = 3 and 10, and follows the evolution of the average photoionization rate in ionized regions in a simple fashion. We provide a simple parameterization of the photoionization rate as a function of density in self-shielded regions during the epoch of reionization.

  13. The Transmission of Thermal and Fast Neutrons in Air Filled Annular Ducts through Slabs of Iron and Heavy Water

    Energy Technology Data Exchange (ETDEWEB)

    Nilsson, J; Sandlin, R

    1964-12-15

    An investigation has been carried out concerning the transmission of thermal and fast neutrons in air filled annular ducts through laminated Fe-D{sub 2}O shields. Measurements have been made with annular air gaps of 0.5, 1.0, 1.5 and 2.0 cm, at a duct length of half a meter. The neutron fluxes were determined with a foil activation technique. The thermal flux was theoretically and experimentally divided into three components, a streaming, a leakage and an albedo component. The fast flux was similarly divided into a streaming component and a 'leakage' component. A calculational model to predict the components was then developed and fitted, to the data obtained by experiments. The model reported here for prediction of neutron attenuation in ducted configurations may be applied to straight annular ducts of arbitrary dimensions and material configurations but is especially designed for the problems met with in short ducts.

  14. Three frequency modulated combination thermal neutron lifetime log and porosity

    International Nuclear Information System (INIS)

    Paap, H.J.; Arnold, D.M.; Smith, M.P.

    1976-01-01

    Methods are disclosed for measuring simultaneously the thermal neutron lifetime of the borehole fluid and earth formations in the vicinity of a well borehole, together with the formation porosity. A harmonically intensity modulated source of fast neutrons is used to irradiate the earth formations with fast neutrons at three different modulation frequencies. Intensity modulated clouds of thermal neutrons at each of the three modulation frequencies are detected by dual spaced detectors and the relative phase shift of the thermal neutrons with respect to the fast neutrons is determined at each of the three modulation frequencies at each detector. These measurements are then combined to determine simultaneously the thermal neutron decay time of the borehole fluid, the thermal neutron decay time of surrounding earth formation media and the porosity of the formation media

  15. The resonance self-shielding calculation with regularized random ladders

    International Nuclear Information System (INIS)

    Ribon, P.

    1986-01-01

    The straightforward method for calculation of resonance self-shielding is to generate one or several resonance ladders, and to process them as resolved resonances. The main drawback of Monte Carlo methods used to generate the ladders, is the difficulty of reducing the dispersion of data and results. Several methods are examined, and it is shown how one (a regularized sampling method) improves the accuracy. Analytical methods to compute the effective cross-section have recently appeared: they are basically exempt from dispersion, but are inevitably approximate. The accuracy of the most sophisticated one is checked. There is a neutron energy range which is improperly considered as statistical. An examination is presented of what happens when it is treated as statistical, and how it is possible to improve the accuracy of calculations in this range. To illustrate the results calculations have been performed in a simple case: nucleus 238 U, at 300 K, between 4250 and 4750 eV. (author)

  16. Method for determining thermal neutron decay times of earth formations

    International Nuclear Information System (INIS)

    Arnold, D.M.

    1976-01-01

    A method is disclosed for measuring the thermal neutron decay time of earth formations in the vicinity of a well borehole. A harmonically intensity modulated source of fast neutrons is used to irradiate the earth formations with fast neutrons at three different intensity modulation frequencies. The tangents of the relative phase angles of the fast neutrons and the resulting thermal neutrons at each of the three frequencies of modulation are measured. First and second approximations to the earth formation thermal neutron decay time are derived from the three tangent measurements. These approximations are then combined to derive a value for the true earth formation thermal neutron decay time

  17. Activation measurements for thermal neutrons. Part J. Evaluation of thermal neutron transmission factors

    International Nuclear Information System (INIS)

    Egbert, Stephen D.

    2005-01-01

    In order to relate thermal neutron activation measurements in samples to the calculated free-in-air thermal neutron activation levels given in Chapter 3, use is made of sample transmission factors. Transmission factors account for the modification of the fluence and activation at each sample's in situ location. For the purposes of this discussion, the transmission factor (TF) is defined as the ratio of the in situ sample activation divided by the free-in-air (FIA) activation at a height of 1 m above ground at the same ground range. The procedures for calculation of TF's and example results are presented in this section. (author)

  18. Modelling of neutron and photon transport in iron and concrete radiation shieldings by the Monte Carlo method - Version 2

    CERN Document Server

    Žukauskaite, A; Plukiene, R; Plukis, A

    2007-01-01

    Particle accelerators and other high energy facilities produce penetrating ionizing radiation (neutrons and γ-rays) that must be shielded. The objective of this work was to model photon and neutron transport in various materials, usually used as shielding, such as concrete, iron or graphite. Monte Carlo method allows obtaining answers by simulating individual particles and recording some aspects of their average behavior. In this work several nuclear experiments were modeled: AVF 65 – γ-ray beams (1-10 MeV), HIMAC and ISIS-800 – high energy neutrons (20-800 MeV) transport in iron and concrete. The results were then compared with experimental data.

  19. Revisiting the stamm'ler self-shielding method

    International Nuclear Information System (INIS)

    Hebert, A.

    2004-01-01

    The generalized Stamm'ler method is been used in lattice codes such as PHOENIX, WIMS-AECL and DRAGON-IST for computing self-shielded cross sections, prior to the main flux calculation. This method is handicapped by deficiencies, such as its low accuracy and its inability to represent distributed self-shielding effects in a fuel rod or across a fuel bundle. The paper describes improvements that could be made to the generalized Stamm'ler method in order to mitigate these two defects. A validation is presented for the case of 238 U nuclides located in different geometries. The isotopic absorption rates obtained with the proposed numerical scheme are compared with exact values obtained with a fine-group elastic slowing-down calculation in the resolved energy domain. (author)

  20. GEM-based thermal neutron beam monitors for spallation sources

    International Nuclear Information System (INIS)

    Croci, G.; Claps, G.; Caniello, R.; Cazzaniga, C.; Grosso, G.; Murtas, F.; Tardocchi, M.; Vassallo, E.; Gorini, G.; Horstmann, C.; Kampmann, R.; Nowak, G.; Stoermer, M.

    2013-01-01

    The development of new large area and high flux thermal neutron detectors for future neutron spallation sources, like the European Spallation Source (ESS) is motivated by the problem of 3 He shortage. In the framework of the development of ESS, GEM (Gas Electron Multiplier) is one of the detector technologies that are being explored as thermal neutron sensors. A first prototype of GEM-based thermal neutron beam monitor (bGEM) has been built during 2012. The bGEM is a triple GEM gaseous detector equipped with an aluminum cathode coated by 1μm thick B 4 C layer used to convert thermal neutrons to charged particles through the 10 B(n, 7 Li)α nuclear reaction. This paper describes the results obtained by testing a bGEM detector at the ISIS spallation source on the VESUVIO beamline. Beam profiles (FWHM x =31 mm and FWHM y =36 mm), bGEM thermal neutron counting efficiency (≈1%), detector stability (3.45%) and the time-of-flight spectrum of the beam were successfully measured. This prototype represents the first step towards the development of thermal neutrons detectors with efficiency larger than 50% as alternatives to 3 He-based gaseous detectors

  1. Generation and testing of the shielding data library EURLIB for fission and fusion technology

    International Nuclear Information System (INIS)

    Caglioti, E.; Hehn, G.; Herrnberger, V.; Mattes, M.; Nicks, R.; Penkuhn, H.

    1977-01-01

    For the common field of core physics and shielding, the CSEWG group structure of 239 fast neutron groups had been proposed, of which the 100 neutron groups of the EURLIB Library is a sub-set for shielding. This standard group Library EURLIB had been initiated by the NEA-specialist group on shielding benchmarks in 1974. The wide acceptance of the Library for interpretation of benchmarks in the NEA program represents an important step forward in the standardization of group data which is the basic requirement for a useful collaboration. On the other side the interpretation of a series of different benchmark experiments with the EURLIB Library provides the best check of the cross section data for neutron and gamma-rays showing the needs for further improvements. The paper describes the joint work of IKE, Stuttgart and EURATOM, Ispra in generating multigroup libraries for neutron and gamma-rays. Special effort has been devoted to improve the flux weighting for both types of radiation and proper treatment of thermal neutrons. The coupled multigroup Library of 100 neutron and 20 gamma groups is collapsed into few group structures for typical designs of LWR, LMFBR, gas cooled and thermonuclear reactors. The work for optimal few group representation is done in cooperation with EIR, Wurenlingen. The testing of the EURLIB Library is a common effort of several institutions participating in the NEA shielding benchmark program

  2. Infinite slab-shield dose calculations

    International Nuclear Information System (INIS)

    Russell, G.J.

    1989-01-01

    I calculated neutron and gamma-ray equivalent doses leaking through a variety of infinite (laminate) slab-shields. In the shield computations, I used, as the incident neutron spectrum, the leakage spectrum (<20 MeV) calculated for the LANSCE tungsten production target at 90 degree to the target axis. The shield thickness was fixed at 60 cm. The results of the shield calculations show a minimum in the total leakage equivalent dose if the shield is 40-45 cm of iron followed by 20-15 cm of borated (5% B) polyethylene. High-performance shields can be attained by using multiple laminations. The calculated dose at the shield surface is very dependent on shield material. 4 refs., 4 figs., 1 tab

  3. Application of nuclear pumped laser to an optical self-powered neutron detector

    Science.gov (United States)

    Yamanaka, N.; Takahashi, H.; Iguchi, T.; Nakazawa, M.; Kakuta, T.; Yamagishi, H.; Katagiri, M.

    1996-05-01

    A Nuclear Pumped Laser (NPL) using 3He/Ne/Ar gas mixture is investigated for a purpose of applying to an optical self-powered neutron detector. Reactor experiments and simulations on lasing mechanism have been made to estimate the best gas pressure and mixture ratios on the threshold input power density (or thermal neutron flux) in 3He/Ne/Ar mixture. Calculational results show that the best mixture pressure is 3He/Ne/Ar=2280/60/100 Torr and thermal neutron flux threshold 5×1012 n/cm2 sec, while the reactor experiments made in the research reactor ``YAYOI'' of the University of Tokyo and ``JRR-4'' of JAERI also demonstrate that excitational efficiency is maximized in a similar gas mixture predicted by the calculation.

  4. Standard Test Method for Determining Thermal Neutron Reaction Rates and Thermal Neutron Fluence Rates by Radioactivation Techniques

    CERN Document Server

    American Society for Testing and Materials. Philadelphia

    2008-01-01

    1.1 The purpose of this test method is to define a general procedure for determining an unknown thermal-neutron fluence rate by neutron activation techniques. It is not practicable to describe completely a technique applicable to the large number of experimental situations that require the measurement of a thermal-neutron fluence rate. Therefore, this method is presented so that the user may adapt to his particular situation the fundamental procedures of the following techniques. 1.1.1 Radiometric counting technique using pure cobalt, pure gold, pure indium, cobalt-aluminum, alloy, gold-aluminum alloy, or indium-aluminum alloy. 1.1.2 Standard comparison technique using pure gold, or gold-aluminum alloy, and 1.1.3 Secondary standard comparison techniques using pure indium, indium-aluminum alloy, pure dysprosium, or dysprosium-aluminum alloy. 1.2 The techniques presented are limited to measurements at room temperatures. However, special problems when making thermal-neutron fluence rate measurements in high-...

  5. PMMA/MWCNT nanocomposite for proton radiation shielding applications

    Science.gov (United States)

    Li, Zhenhao; Chen, Siyuan; Nambiar, Shruti; Sun, Yonghai; Zhang, Mingyu; Zheng, Wanping; Yeow, John T. W.

    2016-06-01

    Radiation shielding in space missions is critical in order to protect astronauts, spacecraft and payloads from radiation damage. Low atomic-number materials are efficient in shielding particle-radiation, but they have relatively weak material properties compared to alloys that are widely used in space applications as structural materials. However, the issues related to weight and the secondary radiation generation make alloys not suitable for space radiation shielding. Polymers, on the other hand, can be filled with different filler materials for reinforcement of material properties, while at the same time provide sufficient radiation shielding function with lower weight and less secondary radiation generation. In this study, poly(methyl-methacrylate)/multi-walled carbon nanotube (PMMA/MWCNT) nanocomposite was fabricated. The role of MWCNTs embedded in PMMA matrix, in terms of radiation shielding effectiveness, was experimentally evaluated by comparing the proton transmission properties and secondary neutron generation of the PMMA/MWCNT nanocomposite with pure PMMA and aluminum. The results showed that the addition of MWCNTs in PMMA matrix can further reduce the secondary neutron generation of the pure polymer, while no obvious change was found in the proton transmission property. On the other hand, both the pure PMMA and the nanocomposite were 18%-19% lighter in weight than aluminum for stopping the protons with the same energy and generated up to 5% fewer secondary neutrons. Furthermore, the use of MWCNTs showed enhanced thermal stability over the pure polymer, and thus the overall reinforcement effects make MWCNT an effective filler material for applications in the space industry.

  6. Development of HANARO ST3 shield

    International Nuclear Information System (INIS)

    Park, K. N.; Lee, J. S.; Shim, H. S.

    2004-12-01

    This report contains the design, fabrication and accurate installation of ST3 shield, which would be installed at ST3 beam port of HANARO. At first, we designed and fabricated ST3 shield casemate composed of 14 blocks. We filled it with heavy concrete, lead ingot and polyethylene that mixed B 4 C powder and epoxy. The average filling density of total shield casemate was 4.7g/cm 3 . The developed ST3 shield was installed at the ST3 beam port and the accuracy of installation for each beam path and channel was evaluated. We found that the extraction of neutron beam to meet the requirement of neutron spectrometer is possible. Also, we developed ancillary equipment such as BGU, quick shutter and exterior shield door for the effective opening and closing of neutron beam. As a result of this study, it was found that neutron spectrometer such as neutron reflectometer and high intensity powder diffractomater can be installed at the ST3 beam port

  7. ANALISIS KESELAMATAN TERMOHIDROLIK BULK SHIELDING REAKTOR KARTINI

    Directory of Open Access Journals (Sweden)

    Azizul Khakim

    2015-10-01

    Full Text Available ABSTRAK ANALISIS KESELAMATAN TERMOHIDROLIK BULK SHIELDING REAKTOR KARTINI. Bulk shielding merupakan fasilitas yang terintegrasi dengan reaktor Kartini yang berfungsi sebagai penyimpanan sementara bahan bakar bekas. Fasilitas ini merupakan fasilitas yang termasuk dalam struktur, sistem dan komponen (SSK yang penting bagi keselamatan. Salah satu fungsi keselamatan dari sistem penanganan dan penyimpanan bahan bakar adalah mencegah kecelakaan kekritisan yang tak terkendali dan membatasi naiknya temperatur bahan bakar. Analisis keselamatan paling kurang harus mencakup analisis keselamatan dari sisi neutronik dan termo hidrolik Bulk shielding. Analisis termo hidrolik ditujukan untuk memastikan perpindahan panas dan proses pendinginan bahan bakar bekas berjalan baik dan tidak terjadi akumulasi panas yang mengancam integritas bahan bakar. Code tervalidasi PARET/ANL digunakan untuk analisis pendinginan dengan mode konveksi alam. Hasil perhitungan menunjukkan bahwa mode pendinginan konvekasi alam cukup memadai dalam mendinginkan panas sisa tanpa mengakibatkan kenaikan temperatur bahan bakar yang signifikan. Kata kunci: Bulk shielding, bahan bakar bekas, konveksi alam, PARET.   ABSTRACT THERMAL HYDRAULIC SAFETY ANALYSIS OF BULK SHIELDING KARTINI REACTOR. Bulk shielding is an integrated facility to Kartini reactor which is used for temporary spent fuels storage. The facility is one of the structures, systems and components (SSCs important to safety. Among the safety functions of fuel handling and storage are to prevent any uncontrolable criticality accidents and to limit the fuel temperature increase. Safety analyses should, at least, cover neutronic and thermal hydraulic calculations of the bulk shielding. Thermal hydraulic analyses were intended to ensure that heat removal and the process of the spent fuels cooling takes place adequately and no heat accumulation that challenges the fuel integrity. Validated code, PARET/ANL was used for analysing the

  8. Real-time thermal neutron radiographic detection systems

    International Nuclear Information System (INIS)

    Berger, H.; Bracher, D.A.

    1976-01-01

    Systems for real-time detection of thermal neutron images are reviewed. Characteristics of one system are presented; the data include contrast, resolution and speed of response over the thermal neutron intensity range 2.5 10 3 n/cm 2 -sec to 10 7 n/cm 2 -sec

  9. Attenuation of Thermal Neutrons by Crystalline Silicon

    International Nuclear Information System (INIS)

    Adib, M.; Habib, N.; Ashry, A.; Fathalla, M.

    2002-01-01

    A simple formula is given which allows to calculate the contribution of the total neutron cross - section including the Bragg scattering from different (hkt) planes to the neutron * transmission through a solid crystalline silicon. The formula takes into account the silicon form of poly or mono crystals and its parameters. A computer program DSIC was developed to provide the required calculations. The calculated values of the total neutron cross-section of perfect silicon crystal at room and liquid nitrogen temperatures were compared with the experimental ones. The obtained agreement shows that the simple formula fits the experimental data with sufficient accuracy .A good agreement was also obtained between the calculated and measured values of polycrystalline silicon in the energy range from 5 eV to 500μ eV. The feasibility study on using a poly-crystalline silicon as a cold neutron filter and mono-crystalline as a thermal neutron one is given. The optimum crystal thickness, mosaic spread, temperature and cutting plane for efficiently transmitting the thermal reactor neutrons, while rejecting both fast neutrons and gamma rays accompanying the thermal ones for the mono crystalline silicon are also given

  10. Parameters measurement for the thermal neutron beam in the thermal column hole of Xi’an pulse reactor

    Institute of Scientific and Technical Information of China (English)

    2010-01-01

    The distribution of the neutron spectra in the thermal column hole of Xi’an pulse reactor was measured with the time-of-flight method.Compared with the thermal Maxwellian theory neutron spectra,the thermal neutron spectra measured is a little softer,and the average neutron energy of the experimental spectra is about 0.042±0.01 eV.The thermal neutron fluence rate at the front end of thermal column hole,measured with gold foil activation techniques,is about 1.18×105 cm-2 s-1.The standard uncertainty of the measured thermal neutron fluence is about 3%.The spectra-averaged cross section of 197Au(n,γ) determined by the experimental thermal neutron spectra is(92.8±0.93) ×10-24 cm2.

  11. Feasibility study on using imaging plates to estimate thermal neutron fluence in neutron-gamma mixed fields

    International Nuclear Information System (INIS)

    Fujibuchi, T.; Tanabe, Y.; Sakae, T.; Terunuma, T.; Isobe, T.; Kawamura, H.; Yasuoka, K.; Matsumoto, T.; Harano, H.; Nishiyama, J.; Masuda, A.; Nohtomi, A.

    2011-01-01

    In current radiotherapy, neutrons are produced in a photonuclear reaction when incident photon energy is higher than the threshold. In the present study, a method of discriminating the neutron component was investigated using an imaging plate (IP) in the neutron-gamma-ray mixed field. Two types of IP were used: a conventional IP for beta- and gamma rays, and an IP doped with Gd for detecting neutrons. IPs were irradiated in the mixed field, and the photo-stimulated luminescence (PSL) intensity of the thermal neutron component was discriminated using an expression proposed herein. The PSL intensity of the thermal neutron component was proportional to thermal neutron fluence. When additional irradiation of photons was added to constant neutron irradiation, the PSL intensity of the thermal neutron component was not affected. The uncertainty of PSL intensities was approximately 11.4 %. This method provides a simple and effective means of discriminating the neutron component in a mixed field. (authors)

  12. Uranium self-shielding in fast reactor blankets

    Energy Technology Data Exchange (ETDEWEB)

    Kadiroglu, O.K.; Driscoll, M.J.

    1976-03-01

    The effects of heterogeneity on resonance self-shielding are examined with particular emphasis on the blanket region of the fast breeder reactor and on its dominant reaction--capture in /sup 238/U. The results, however, apply equally well to scattering resonances, to other isotopes (fertile, fissile and structural species) and to other environments, so long as the underlying assumptions of narrow resonance theory apply. The heterogeneous resonance integral is first cast into a modified homogeneous form involving the ratio of coolant-to-fuel fluxes. A generalized correlation (useful in its own right in many other applications) is developed for this ratio, using both integral transport and collision probability theory to infer the form of correlation, and then relying upon Monte Carlo calculations to establish absolute values of the correlation coefficients. It is shown that a simple linear prescription can be developed for the flux ratio as a function of only fuel optical thickness and the fraction of the slowing-down source generated by the coolant. This in turn permitted derivation of a new equivalence theorem relating the heterogeneous self-shielding factor to the homogeneous self-shielding factor at a modified value of the background scattering cross section per absorber nucleus. A simple version of this relation is developed and used to show that heterogeneity has a negligible effect on the calculated blanket breeding ratio in fast reactors.

  13. Chemical warfare agents identification by thermal neutron detection

    International Nuclear Information System (INIS)

    Liu Boxue; Ai Xianyun; Tan Daoyuan; Zhang Dianqin

    2000-01-01

    The hydrogen concentration determination by thermal neutron detection is a non-destructive, fast and effective method to identify chemical warfare agents and TNT that contain different hydrogen fraction. When an isotropic neutron source is used to irradiate chemical ammunition, hydrogen atoms of the agent inside shell act as a moderator and slow down neutrons. The number of induced thermal neutrons depends mainly upon hydrogen content of the agent. Therefore measurement of thermal neutron influence can be used to determine hydrogen atom concentration, thereby to determine the chemical warfare agents. Under a certain geometry three calibration curves of count rate against hydrogen concentration were measured. According to the calibration curves, response of a chemical agent or TNT could be calculated. Differences of count rate among chemical agents and TNT for each kind of shells is greater than five times of standard deviations of count rate for any agent, so chemical agents or TNT could be identified correctly. Meanwhile, blast tube or liquid level of chemical warfare agent could affect the response of thermal neutron count rate, and thereby the result of identification. (author)

  14. Neutron flux calculations for criticality safety analysis using the narrow resonance approximations. Vol. 2

    Energy Technology Data Exchange (ETDEWEB)

    Hathout, A M [National Center for Nuclear Safety and Radiation Control, NC-NSRC, Atomic Energy Authority, Cairo (Egypt)

    1996-03-01

    The narrow resonance approximation is applicable for all low-energy resonances and the heaviest nuclides. It is of great importance in neutron calculations, hence, fertile isotopes do not undergo fission at resonance energies. The effect of overestimating the self shielded group averaged cross-section data for a given resonance nuclide can be fairly serious. In the present work, a detailed study, and derivation of the problem of self-shielding are carried-out through the information of Hansen-roach library which is used for criticality safety analysis. The intermediate neutron flux spectrum is analyzed, using the narrow resonance approximation. The resonance self-shielded values of various cross-sections are determined. 4 figs., 3 tabs.

  15. Shielding implications for secondary neutrons and photons produced within the patient during IMPT

    International Nuclear Information System (INIS)

    DeMarco, J.; Kupelian, P.; Santhanam, A.; Low, D.

    2013-01-01

    Purpose: Intensity modulated proton therapy (IMPT) uses a combination of computer controlled spot scanning and spot-weight optimized planning to irradiate the tumor volume uniformly. In contrast to passive scattering systems, secondary neutrons and photons produced from inelastic proton interactions within the patient represent the major source of emitted radiation during IMPT delivery. Various published studies evaluated the shielding considerations for passive scattering systems but did not directly address secondary neutron production from IMPT and the ambient dose equivalent on surrounding occupational and nonoccupational work areas. Thus, the purpose of this study was to utilize Monte Carlo simulations to evaluate the energy and angular distributions of secondary neutrons and photons following inelastic proton interactions within a tissue-equivalent phantom for incident proton spot energies between 70 and 250 MeV.Methods: Monte Carlo simulation methods were used to calculate the ambient dose equivalent of secondary neutrons and photons produced from inelastic proton interactions in a tissue-equivalent phantom. The angular distribution of emitted neutrons and photons were scored as a function of incident proton energy throughout a spherical annulus at 1, 2, 3, 4, and 5 m from the phantom center. Appropriate dose equivalent conversion factors were applied to estimate the total ambient dose equivalent from secondary neutrons and photons.Results: A reference distance of 1 m from the center of the patient was used to evaluate the mean energy distribution of secondary neutrons and photons and the resulting ambient dose equivalent. For an incident proton spot energy of 250 MeV, the total ambient dose equivalent (3.6 × 10 −3 mSv per proton Gy) was greatest along the direction of the incident proton spot (0°–10°) with a mean secondary neutron energy of 71.3 MeV. The dose equivalent decreased by a factor of 5 in the backward direction (170°–180°) with a mean

  16. Shielding implications for secondary neutrons and photons produced within the patient during IMPT

    Energy Technology Data Exchange (ETDEWEB)

    DeMarco, J.; Kupelian, P.; Santhanam, A.; Low, D. [UCLA Department of Radiation Oncology, University of California Los Angeles, Los Angeles, California 90095 (United States)

    2013-07-15

    Purpose: Intensity modulated proton therapy (IMPT) uses a combination of computer controlled spot scanning and spot-weight optimized planning to irradiate the tumor volume uniformly. In contrast to passive scattering systems, secondary neutrons and photons produced from inelastic proton interactions within the patient represent the major source of emitted radiation during IMPT delivery. Various published studies evaluated the shielding considerations for passive scattering systems but did not directly address secondary neutron production from IMPT and the ambient dose equivalent on surrounding occupational and nonoccupational work areas. Thus, the purpose of this study was to utilize Monte Carlo simulations to evaluate the energy and angular distributions of secondary neutrons and photons following inelastic proton interactions within a tissue-equivalent phantom for incident proton spot energies between 70 and 250 MeV.Methods: Monte Carlo simulation methods were used to calculate the ambient dose equivalent of secondary neutrons and photons produced from inelastic proton interactions in a tissue-equivalent phantom. The angular distribution of emitted neutrons and photons were scored as a function of incident proton energy throughout a spherical annulus at 1, 2, 3, 4, and 5 m from the phantom center. Appropriate dose equivalent conversion factors were applied to estimate the total ambient dose equivalent from secondary neutrons and photons.Results: A reference distance of 1 m from the center of the patient was used to evaluate the mean energy distribution of secondary neutrons and photons and the resulting ambient dose equivalent. For an incident proton spot energy of 250 MeV, the total ambient dose equivalent (3.6 Multiplication-Sign 10{sup -3} mSv per proton Gy) was greatest along the direction of the incident proton spot (0 Degree-Sign -10 Degree-Sign ) with a mean secondary neutron energy of 71.3 MeV. The dose equivalent decreased by a factor of 5 in the

  17. URR-PACK: Calculating Self-Shielding in the Unresolved Resonance Energy Range

    International Nuclear Information System (INIS)

    Cullen, Dermott E.; Trkov, Andrej

    2016-07-01

    This report describes HOW to calculate self-shielding in the unresolved resonance region (URR), in terms of the computer codes we provide to allow a user to do these calculations himself. Here we only describe HOW to calculate; a longer companion report describes in detail WHY it is necessary to include URR self-shielding.

  18. Neutron-photon multigroup cross sections for neutron energies less than or equal to400 MeV. Revision 1

    International Nuclear Information System (INIS)

    Alsmiller, R.G. Jr.; Barnes, J.M.; Drischler, J.D.

    1986-01-01

    For a variety of applications, e.g., accelerator shielding design, neutrons in radiotherapy, radiation damage studies, etc., it is necessary to carry out transport calculations involving medium-energy (greater than or equal to20 MeV) neutrons. A previous paper described neutron-photon multigroup cross sections in the ANISN format for neutrons from thermal to 400 MeV. In the present paper the cross-section data presented previously have been revised to make them agree with available experimental data. 7 refs., 1 fig

  19. Thermal design of top shield

    International Nuclear Information System (INIS)

    Raghupathy, S.; Velusamy, K.; Parthasarathy, U.; Ghosh, D.; Selvaraj, P.; Chellapandi, P.; Chetal, S.C.

    2005-01-01

    Full text of publication follows: Prototype Fast Breeder Reactor (PFBR) is a 500 MWe, sodium cooled, pool type fast reactor. The top shield forms the top cover for the main vessel (MV) and includes roof slab (RS), large rotatable plug (LRP), small rotatable plug (SRP) and control Plug (CP). RS, LRP and SRP are box type structures consisting of top and bottom plates stiffened by radial stiffeners and vertical penetration shells. TS is exposed to argon cover gas provided above sodium pool on the bottom side and reactor containment building air at the top. Heat transfer takes place through the argon cover gas to the bottom plate of TS. Annular gaps are formed between the components supported on TS and the component penetrations through which cellular convection takes place. A single thermal shield provided below TS reduces the heat flux to the bottom plate to 1.15 kW/m 2 . The MV (SS 316 LN) is welded to RS (carbon steel A48 P2) through a dissimilar metal weld. A step in RS and an anti convection barrier (ACB) outside RS are provided to limit the temperature at the MV-RS junction. The MV is surrounded by safety vessel (SV) and reactor vault made of concrete. Thermal insulation is provided outside SV to limit the heat transfer to the reactor vault. The design requirements of TS are to maintain the operating temperature at 383-393 K, limit the temperature difference (ΔT) across the height of TS to 20 / 100 K under normal operation/loss of cooling, provide minimum annular gap size at the component penetrations, provide a nearly linear temperature gradient in the CP portion within the height of TS, maintain the temperature of top plate of CP > 383 K, limit the ΔT across the top plate of CP to 2 K, limit the temperature near the inflatable / backup seal to 393 K, limit the temperature at the MV-RS junction and the heat flux to the reactor vault. The total heat transferred to TS is estimated to be 210 kW. A dedicated closed loop cooling system with a total flow rate of 10

  20. Study and development of new dosemeters for thermal neutrons

    International Nuclear Information System (INIS)

    Urena N, F.

    1998-01-01

    An alanine-boron compound, alanine hydroborate, was synthesized and chemically characterized to be used for thermal neutrons fluence measurements. The synthesis of the compound was made by reacting the amino acid alanine with boric acid in three different media: acidic, neutral and alkaline. Physicochemical analysis showed that the alkaline medium is favorable for the synthesis of the alanine hydroborate. The compound was evaluated as a thermal neutron fluence detector by the detection of the free radical yield upon neutron thermal irradiation by Electron Paramagnetic Resonance (EPR). The present work also studies the EPR-signal response of the three preparations to thermal neutron irradiation (φ = 5 x 10 7 n/cm 2 -s). The following EPR signal parameters of the samples were investigated: peak-to-peak signal intensity vs. thermal neutron fluence Φ = φ Δt ; where Δt = 1, 5, 10, 20, 40, 60, 80, 90, 100, 110 and 120 h. , peak-to-peak signal intensity vs. microwave power, signal fading; repeatability, batch homogeneity, stability and zero dose response. It is concluded that these new products could be used in thermal neutron fluence estimations. (Author)

  1. EL-2 reactor: Thermal neutron flux distribution; EL-2: Repartition du flux de neutrons thermiques

    Energy Technology Data Exchange (ETDEWEB)

    Rousseau, A; Genthon, J P [Commissariat a l' Energie Atomique, Saclay (France). Centre d' Etudes Nucleaires

    1958-07-01

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

  2. Neutronics shielding analysis of the last mirror-beam duct system for a laser fusion power reactor

    International Nuclear Information System (INIS)

    Ragheb, M.M.H.; Klein, A.C.

    1981-01-01

    A Monte Carlo three-dimensional neutronics analysis for the last mirror-beam duct system for the SOLASE conceptual laser-driven fusion power reactor design is presented. Detailed geometric configurations including the reactor cavity, the two last mirrors, and the three-section two-right-angle bends duct are modeled. Measurements are given of the dimensions and compositions of the reactor components, and of neutron scalar fluxes, spatial dependencies and neutron volumetric heating rates for the cases of aluminum or Boral as laser beam duct liners, and ordinary concrete or lead mortar as shield material. A three-dimensional modeling of laser-driven reactor penetrations is employed. The particle leakage is found to be excessively high for the configuration of the conceptual design considered and the advantages and disadvantages of various solutions, such as the use of Boral as a duct liner and the use of lead mortar instead of ordinary concrete as a shield material, are considered

  3. A Combined Shielding Design for a Neutron Generator and a Linear Accelerator at Soreq NRC

    International Nuclear Information System (INIS)

    Epstein, L.

    2014-01-01

    A new radiography facility is designed at Soreq NRC. The facility will hold a neutron generator that produces 1.73·109 n/s with an energy of 14 MeV and a linear accelerator that accelerates electrons to an energy of 9 MeV. The two radiation sources will be installed in 2 separate laboratories that will be built in an existing building. Each laboratory will have its own machine and control room. The dose rates around the sources were calculated using the FLUKA Monte Carlo code(1,2). The annual doses were calculated in several regions around the generator and the accelerator laboratories in accordance with the occupancy in each area. The calculated annual doses were compared with the dose limits specified in the Safety at Work Regulations(3) and the IAEC Standard for Protection against Ionizing Radiation. The shielding was designed to comply with the following dose constraints: 0.3 mSv/y for members of the public and 2 mSv/y for radiation workers. Each radiation source is planned to produce radiation for a maximum of 500 hours per year. The dose rate in the direct beam of the accelerator is 30 Gy/min at 1 m from the source and it will be surrounded by a collimator with an opening of 30N-tilde horizontally and 2 mm vertically, 3 m from the radiation source. The leakage radiation dose will not be greater than 1.5 mGy/min (0.005% of the direct beam, according to the manufacturer). The leakage radiation will be produced isotropically. The neutron generator will be surrounded by a shielding made of a 10 cm iron cylinder (density 7.87 g/cm3), surrounded by 50 cm of borated polyethylene (atomic percent: H (13.8%), C (82.2%), B (4%), density: 0.92 g/cm3) and 5 cm of lead (density 11.35 g/cm3). The neutron generator shielding was not designed or required in the present shielding design but was considered in the shielding calculations

  4. Shielding analysis method applied to nuclear ship 'MUTSU' and its evaluation based on experimental analyses

    International Nuclear Information System (INIS)

    Yamaji, Akio; Miyakoshi, Jun-ichi; Iwao, Yoshiaki; Tsubosaka, Akira; Saito, Tetsuo; Fujii, Takayoshi; Okumura, Yoshihiro; Suzuoki, Zenro; Kawakita, Takashi.

    1984-01-01

    Procedures of shielding analysis are described which were used for the shielding modification design of the Nuclear Ship ''MUTSU''. The calculations of the radiation distribution on board were made using Sn codes ANISN and TWOTRAN, a point kernel code QAD and a Monte Carlo code MORSE. The accuracies of these calculations were investigated through the analysis of various shielding experiments: the shield tank experiment of the Nuclear Ship ''Otto Hahn'', the shielding mock-up experiment for ''MUTSU'' performed in JRR-4, the shielding benchmark experiment using the 16 N radiation facility of AERE Harwell and the shielding effect experiment of the ship structure performed in the training ship ''Shintoku-Maru''. The values calculated by the ANISN agree with the data measured at ''Otto Hahn'' within a factor of 2 for fast neutrons and within a factor of 3 for epithermal and thermal neutrons. The γ-ray dose rates calculated by the QAD agree with the measured values within 30% for the analysis of the experiment in JRR-4. The design values for ''MUTSU'' were determined in consequence of these experimental analyses. (author)

  5. Monte Carlo simulation of explosive detection system based on a Deuterium-Deuterium (D-D) neutron generator.

    Science.gov (United States)

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

    2014-12-01

    An explosive detection system based on a Deuterium-Deuterium (D-D) neutron generator has been simulated using the Monte Carlo N-Particle Transport Code (MCNP5). Nuclear-based explosive detection methods can detect explosives by identifying their elemental components, especially nitrogen. Thermal neutron capture reactions have been used for detecting prompt gamma emission (10.82MeV) following radiative neutron capture by (14)N nuclei. The explosive detection system was built based on a fully high-voltage-shielded, axial D-D neutron generator with a radio frequency (RF) driven ion source and nominal yield of about 10(10) fast neutrons per second (E=2.5MeV). Polyethylene and paraffin were used as moderators with borated polyethylene and lead as neutron and gamma ray shielding, respectively. The shape and the thickness of the moderators and shields are optimized to produce the highest thermal neutron flux at the position of the explosive and the minimum total dose at the outer surfaces of the explosive detection system walls. In addition, simulation of the response functions of NaI, BGO, and LaBr3-based γ-ray detectors to different explosives is described. Copyright © 2014 Elsevier Ltd. All rights reserved.

  6. Measurements of Neutron and Gamma Attenuation in Massive Laminated Shields of Concrete and a Study of the Accuracy of some Methods of Calculation

    Energy Technology Data Exchange (ETDEWEB)

    Aalto, E; Nilsson, R

    1964-09-15

    Extensive neutron and gamma attenuation measurements have been performed in magnetite and ordinary concrete up to a depth of 2 metres in order to study the accuracy attainable by some shield calculation methods. The effect of thin, heavy layers (Pb) has also been studied. Experimental facilities and instrumentation, especially the foil detection methods used for thermal and epithermal neutrons, are described in some detail. Great weight is laid upon a thorough error analysis. The fluxes measured are compared to those calculated by an earlier version of the British 18-group removal method (RASH B{sub 3}), by an improved removal method (NRN) developed at AB Atomenergi, and by numerical integration of the Boltzmann equation (NIOBE). The results show that shielding calculations with the newer methods give fluxes that are generally within a factor of 2-3 from the true values. A greater accuracy seems to be difficult to obtain in practice in spite of possible improvements in the mathematical solution of the transport problem. The greatest errors originate in the translation between the true and calculation geometries in the uncertainty of material properties in the case of concrete, and in approximations and inaccuracies of radiation sources.

  7. Measurements of Neutron and Gamma Attenuation in Massive Laminated Shields of Concrete and a Study of the Accuracy of some Methods of Calculation

    International Nuclear Information System (INIS)

    Aalto, E.; Nilsson, R.

    1964-09-01

    Extensive neutron and gamma attenuation measurements have been performed in magnetite and ordinary concrete up to a depth of 2 metres in order to study the accuracy attainable by some shield calculation methods. The effect of thin, heavy layers (Pb) has also been studied. Experimental facilities and instrumentation, especially the foil detection methods used for thermal and epithermal neutrons, are described in some detail. Great weight is laid upon a thorough error analysis. The fluxes measured are compared to those calculated by an earlier version of the British 18-group removal method (RASH B 3 ), by an improved removal method (NRN) developed at AB Atomenergi, and by numerical integration of the Boltzmann equation (NIOBE). The results show that shielding calculations with the newer methods give fluxes that are generally within a factor of 2-3 from the true values. A greater accuracy seems to be difficult to obtain in practice in spite of possible improvements in the mathematical solution of the transport problem. The greatest errors originate in the translation between the true and calculation geometries in the uncertainty of material properties in the case of concrete, and in approximations and inaccuracies of radiation sources

  8. Shielding benchmark test

    International Nuclear Information System (INIS)

    Kawai, Masayoshi

    1984-01-01

    Iron data in JENDL-2 have been tested by analyzing shielding benchmark experiments for neutron transmission through iron block performed at KFK using CF-252 neutron source and at ORNL using collimated neutron beam from reactor. The analyses are made by a shielding analysis code system RADHEAT-V4 developed at JAERI. The calculated results are compared with the measured data. As for the KFK experiments, the C/E values are about 1.1. For the ORNL experiments, the calculated values agree with the measured data within an accuracy of 33% for the off-center geometry. The d-t neutron transmission measurements through carbon sphere made at LLNL are also analyzed preliminarily by using the revised JENDL data for fusion neutronics calculation. (author)

  9. TORE-SUPRA: design of thermal radiation shield at 80 K

    International Nuclear Information System (INIS)

    Aymar, R.; Cordier, J.J.; Deschamps, P.; Gauthier, A.; Perin, J.P.

    1982-09-01

    The TORE-SUPRA superconducting toroidal magnet operating at liquid helium temperature, must be protected against thermal radiation from the vessels. For this purpose, stainless steel heat shields, cooled at 80 K, are positioned between coil casings at 4.5 K and the vessels, and constitute a double stiff toroid which completely surrounds the magnet. Mockups have been manufactured to study their design and operating problems. Calculations have also been made to analyse the mechanical behaviour of these shields

  10. Experimental characterization of semiconductor-based thermal neutron detectors

    Energy Technology Data Exchange (ETDEWEB)

    Bedogni, R., E-mail: roberto.bedogni@lnf.infn.it [IFNF—LNF, via E. Fermi n. 40, 00044 Frascati, Roma (Italy); Bortot, D.; Pola, A.; Introini, M.V.; Lorenzoli, M. [Politecnico di Milano, Dipartimento di Energia, via La Masa 34, 20156 Milano (Italy); INFN—Milano, Via Celoria 16, 20133 Milano (Italy); Gómez-Ros, J.M. [IFNF—LNF, via E. Fermi n. 40, 00044 Frascati, Roma (Italy); CIEMAT, Av. Complutense 40, 28040 Madrid (Spain); Sacco, D. [IFNF—LNF, via E. Fermi n. 40, 00044 Frascati, Roma (Italy); INAIL—DIT, Via di Fontana Candida 1, 00040 Monteporzio Catone (Italy); Esposito, A.; Gentile, A.; Buonomo, B. [IFNF—LNF, via E. Fermi n. 40, 00044 Frascati, Roma (Italy); Palomba, M.; Grossi, A. [ENEA Triga RC-1C.R. Casaccia, via Anguillarese 301, 00060 S. Maria di Galeria, Roma (Italy)

    2015-04-21

    In the framework of NESCOFI@BTF and NEURAPID projects, active thermal neutron detectors were manufactured by depositing appropriate thickness of {sup 6}LiF on commercially available windowless p–i–n diodes. Detectors with different radiator thickness, ranging from 5 to 62 μm, were manufactured by evaporation-based deposition technique and exposed to known values of thermal neutron fluence in two thermal neutron facilities exhibiting different irradiation geometries. The following properties of the detector response were investigated and presented in this work: thickness dependence, impact of parasitic effects (photons and epithermal neutrons), linearity, isotropy, and radiation damage following exposure to large fluence (in the order of 10{sup 12} cm{sup −2})

  11. RSMASS: A simple model for estimating reactor and shield masses

    International Nuclear Information System (INIS)

    Marshall, A.C.; Aragon, J.; Gallup, D.

    1987-01-01

    A simple mathematical model (RSMASS) has been developed to provide rapid estimates of reactor and shield masses for space-based reactor power systems. Approximations are used rather than correlations or detailed calculations to estimate the reactor fuel mass and the masses of the moderator, structure, reflector, pressure vessel, miscellaneous components, and the reactor shield. The fuel mass is determined either by neutronics limits, thermal/hydraulic limits, or fuel damage limits, whichever yields the largest mass. RSMASS requires the reactor power and energy, 24 reactor parameters, and 20 shield parameters to be specified. This parametric approach should be applicable to a very broad range of reactor types. Reactor and shield masses calculated by RSMASS were found to be in good agreement with the masses obtained from detailed calculations

  12. Thermal neutron imaging in an active interrogation environment

    International Nuclear Information System (INIS)

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

    2009-01-01

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

  13. Concrete shielding for nuclear ship 'Mutsu'

    International Nuclear Information System (INIS)

    Nagase, Tetsuo; Saito, Tetsuo

    1983-01-01

    The repair works of the shielding for the nuclear ship ''Mutsu'' were completed in August, 1982. For the primary shielding, serpentine concrete was adopted as it contains a large quantity of water required for neutron shielding, and in the secondary shielding at the upper part of the reactor containment vessel, the original shielding was abolished, and the heavy concrete (high water content, high density concrete) which is effective for neutron and gamma-ray shielding was newly adopted. In this report, the design and construction using these shielding concrete are outlined. In September, 1974, Mutsu caused radiation leak during the test, and the cause was found to be the fast neutrons streaming through a gap between the reactor pressure vessel and the primary shielding. The repair works were carried out in the Sasebo Shipyard. The outline of the repair works of the shielding is described. The design condition for the shielding, the design standard for the radiation dose outside and inside the ship, the method of shielding analysis and the performance required for shielding concrete are reported. The selection of materials, the method of construction and mixing ratio, the evaluation of the soundness and properties of concrete, and the works of placing the shielding concrete are outlined. (Kako, I.)

  14. Analysis of Neutron Flux Using Monte Carlo Methods

    International Nuclear Information System (INIS)

    Picha, Roppon

    2007-08-01

    Full text: The energy profile of neutrons from a fission reactor core and a neutron irradiation setup are simulated. The neutron doses deposited inside casings of aluminum, cadmium, and tantalum are studied via MCNP simulations to estimate the doses received by materials with different types of shielding. It is found that the difference in dose reduction between cadmium and tantalum is most pronounced at the thermal energy region

  15. Study of the behavior of thermal shield support system for the French CPO series plants

    International Nuclear Information System (INIS)

    Bellet, S.; Roux, P.; Bhandari, D.R.; Schwirian, R.E.; Yu, C.; Matarazzo, J.C.; Singleton, N.R.

    1996-01-01

    Degradation/failure of thermal shield support system in PWRs has been observed in the US as well as in foreign plants. In almost all the cases, remedial actions were put in place at very high economic costs to the utilities only after the failures had occurred. This paper presents the results of a comprehensive study to predict the long term behavior of a thermal shield support system due to flow-induced vibratory loads and thermal transients. Excellent agreement from the system finite model between the measured plant test data on the barrel/thermal shield beam and shell mode frequencies and the flexure strains confirms the basic structural behavior and physics of the flow induced vibrations. Loads and stresses on the support bolts and the flexures were determined to predict the fatigue life of the components

  16. Several problems in accelerator shielding study

    International Nuclear Information System (INIS)

    Nakamura, Takashi; Hirayama, Hideo; Ban, Shuichi.

    1980-01-01

    Recently, the utilization of accelerators has increased rapidly, and the increase of accelerating energy and beam intensity is also remarkable. The studies on accelerator shielding have become important, because the amount of radiation emitted from accelerators increased, the regulation of the dose of environmental radiation was tightened, and the cost of constructing shielding rose. As the plans of constructing large accelerators have been made successively, the survey on the present state and the problems of the studies on accelerator shielding was carried out. Accelerators are classified into electron accelerators and proton accelerators in view of the studies on shielding. In order to start the studies on accelerator shielding, first, the preparation of the cross section data is indispensable. The cross sections for generating Bremsstrahlung, photonuclear reactions generating neutrons, generation of neutrons by hadrons, nuclear reaction of neutrons and generation of gamma-ray by hadrons are described. The generation of neutrons and gamma-ray as the problems of thick targets is explained. The shielding problems are complex and diversified, but in this paper, the studies on the shielding, by which basic data are obtainable, are taken up, such as beam damping and side wall shielding. As for residual radioactivity, main nuclides and the difference of residual radioactivity according to substances have been studied. (J.P.N.)

  17. 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

  18. Development of temperature related thermal neutron scattering database for MCNP

    International Nuclear Information System (INIS)

    Mei Longwei; Cai Xiangzhou; Jiang Dazhen; Chen Jingen; Guo Wei

    2013-01-01

    Based on ENDF/B-Ⅶ neutron library, the thermal neutron scattering library S(α, β) for molten salt reactor moderators was developed. The temperatures of this library were chose as the characteristic temperature of the molten salt reactor. The cross section of the thermal neutron scattering of ACE format was investigated, and this library was also validated by the benchmarks of ICSBEP. The uncertainties shown in the validation were in reasonable range when compared with the thermal neutron scattering library tmccs which included in the MCNP data library. It was proved that the thermal neutron scattering library processed in this study could be used in the molten salt reactor design. (authors)

  19. Criticality analysis of thermal reactors for two energy groups applying Monte Carlo and neutron Albedo method

    International Nuclear Information System (INIS)

    Terra, Andre Miguel Barge Pontes Torres

    2005-01-01

    The Albedo method applied to criticality calculations to nuclear reactors is characterized by following the neutron currents, allowing to make detailed analyses of the physics phenomena about interactions of the neutrons with the core-reflector set, by the determination of the probabilities of reflection, absorption, and transmission. Then, allowing to make detailed appreciations of the variation of the effective neutron multiplication factor, keff. In the present work, motivated for excellent results presented in dissertations applied to thermal reactors and shieldings, was described the methodology to Albedo method for the analysis criticality of thermal reactors by using two energy groups admitting variable core coefficients to each re-entrant current. By using the Monte Carlo KENO IV code was analyzed relation between the total fraction of neutrons absorbed in the core reactor and the fraction of neutrons that never have stayed into the reflector but were absorbed into the core. As parameters of comparison and analysis of the results obtained by the Albedo method were used one dimensional deterministic code ANISN (ANIsotropic SN transport code) and Diffusion method. The keff results determined by the Albedo method, to the type of analyzed reactor, showed excellent agreement. Thus were obtained relative errors of keff values smaller than 0,78% between the Albedo method and code ANISN. In relation to the Diffusion method were obtained errors smaller than 0,35%, showing the effectiveness of the Albedo method applied to criticality analysis. The easiness of application, simplicity and clarity of the Albedo method constitute a valuable instrument to neutronic calculations applied to nonmultiplying and multiplying media. (author)

  20. 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.

  1. Design of analytical instrumentation with D-T sealed neutron generators

    International Nuclear Information System (INIS)

    Qiao Yahua; Wu Jizong; Zheng Weiming; Liu Quanwei; Zhang Min

    2008-01-01

    Analytical instrumentation with D-T sealed neutron generators source activation, The 14 MeV D-T sealed neutron tube with 10 9 n · s -1 neutron yield is used as generator source. The optimal structure of moderator and shield was achieved by MC computing.The instrumentation's configuration is showed. The instrumentation is made up of the SMY-DT50.8-2.1 sealed neutron tube and the high-voltage power supply system, which center is the sealed neutron generators. 6 cm Pb and 20 cm polythene is chosen as moderator, Pb, polythene and 10 cm boron-PE was chosen as shield .The sample box is far the source from 9 cm, the measurement system were made up of HPGe detector and the sample transforming system. After moderator and shield, the thermal neutron fluence rate at the point of sample is 0.93 × 10 6 n · s -1 cm -2 , which is accorded with design demand, and the laboratory and surroundings reaches the safety standard of the dose levels. (authors)

  2. Thermal neutron standard fields with the KUR heavy water facility

    International Nuclear Information System (INIS)

    Kanda, K.; Kobayashi, K.; Shibata, T.

    1978-01-01

    A heavy water facility attached to the KUR (Kyoto University Reactor, swimming pool type, 5 MW) yields pure thermal neutrons in the Maxwellian distribution. The facility is faced to the core of KUR and it contains about 2 tons of heavy water. The thickness of the layer is about 140 cm. The neutron spectrum was measured with the time of flight technique using a fast chopper. The measured spectrum was in good agreement with the Maxwellian distribution in all energy region for thermal neutrons. The neutron temperature was slightly higher than the heavy water temperature. The contamination of epithermal and fast neutrons caused by photo-neutrons of the γ-n reaction of heavy water was very small. The maximum intensity of thermal neutrons is 3x10 11 n/cm 2 sec. When the bismuth scatterer is attached, the gamma rays contamination is eliminated by the ratio of 0.05 of gamma rays to neutrons in rem. This standard neutron field has been used for such experiments as thermal neutron cross section measurement, detector calibration, activation analysis, biomedical purposes etc. (author)

  3. Studying the ability to use basalt in preparing radiation shielding concrete and the properties of the resulted concrete

    International Nuclear Information System (INIS)

    Alhajali, S.; Yousef, S.; Kanbour, M.; Naoum, B.

    2010-12-01

    Basalt is widespread rocks in the lands of Syria. This kind of rocks has high density relatively, high insulation properties and, mechanical and heat resistance. In this work several kinds of basalt rocks, which were collected from several sites, were studied. The analyses which were done, shows that the basalt rocks collected from Shahba, Nba'a Al-Sakhr and Almana'a mountain are suitable for high efficient gamma radiation shielding, but with low efficiency for neutron shielding, especially for thermal and epithermal neutrons. (author)

  4. Thermodynamic considerations on self-regulating characteristics of a cold neutron source with a closed thermosiphon

    International Nuclear Information System (INIS)

    Kawai, Takeshi; Utsuro, Masahiko; Ogino, Fumimaru.

    1991-01-01

    The present report describes that a cold neutron source (CNS) having a closed-thermosiphon cooling loop shows a self-regulating characteristic under thermal disturbances if the effect of the moderator transfer tube is negligible. Due to this property, the liquid level in the moderator cell is kept almost constant under thermal disturbances. The thermodynamic meaning of the self-regulating property in the idealized closed-thermosiphon and the effect of the moderator transfer tube to the self-regulation are described. (author)

  5. Effects of neutron source ratio on nuclear characteristics of D-D fusion reactor blankets and shields

    International Nuclear Information System (INIS)

    Nakashima, Hideki; Nakao, Yasuyuki; Ohta, Masao

    1978-01-01

    An examination is made of the dependence shown by the nuclear characteristics of the blanket and shield of D-D fusion reactors on S sub( d d)/S sub( d t), the ratio between the 2.45 MeV neutrons resulting from the D-D reaction and those of 14.06 MeV from the D-T reaction. Also, an estimate is presented of this neutron source ratio S sub( d d)/S sub( d t) for the case of D-D reactors, taken as an example. It is shown that an increase of S sub( d d)/S sub( d t) reduces the amount of nuclear heating per unit source neutron, while at the same time improving the shielding characteristics. This is accountable to lowering of the energy and penetrability of incident neutrons into the blanket brought about by the increase of S sub( d d)/S sub( d t). The value of S sub( d d)/S sub( d t) in a steady state D-D fusioning plasma core is estimated to be 1.46 -- 1.72 for an ion temperature ranging from 60 -- 180 keV. The reductions obtained on H sub( t)sup( b) (total heating in the blanket), H sub( t)sup( m g)/H sub( t)sup( b) (shielding indicator = ratio between total heating in superconducting magnet and that in the blanket) and phi sup( m g)/phi sup( w) (ratio of fast neutron fluxes between that at the magnet inner surface and that at the first wall inner surface) brought about by increasing S sub( d d)/S sub( d t) from unity to the value cited above do not differ to any appreciable extent, whichever is adopted among the design models considered here, the differences being at most about 10, 15 and 25%, respectively, for these three parameters. These results would broaden the validity of the conclusion derived in the previous paper for the case of S sub( d d)/S sub( d t) = 1.0, that the blanket-shield concept would appear to be the most suitable for D-D fusion reactors. (author)

  6. Nuclear data for radiation shielding

    International Nuclear Information System (INIS)

    Miyasaka, Shunichi; Takahashi, Hiroshi.

    1976-01-01

    The third shielding expert conference was convened in Paris in Oct. 1975 for exchanging informations about the sensitivity evaluation of nuclear data in shielding calculation and integral bench mark experiment. The requirements about nuclear data presented at present from the field of nuclear design do not reflect sufficiently the requirements of shielding design, therefore it was the object to gather the requirements about nuclear data from the field of shielding. The nuclides used for shielding are numerous, and the nuclear data on these isotopes are required. Some of them cannot be ignored as the source of secondary γ-ray or in view of the radioactivation of materials. The requirements for the nuclear data of neutrons in the field of shielding are those concerning the reaction cross sections producing secondary γ-ray, the reaction cross sections including the production of secondary neutrons, elastic scattering cross sections, and total cross sections. The topics in the Paris conference about neutron shielding data are described, such as the methodology of sensitivity evaluation, the standardization of group constant libraries, the bench mark experiment on iron and sodium, and the cross section of γ-ray production. In the shielding of nuclear fission reactors, the γ-ray production owing to nuclear fission reaction is also important. In (d, t) fusion reactors, high energy neutrons are generated, and high energy γ-ray is emitted through giant E1 resonance. (Kako, I.)

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

    Science.gov (United States)

    Vermeeren, L.; Wéber, M.

    2003-06-01

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

  8. Experimental study of neutron streaming through steel-walled annular ducts in reactor shields

    International Nuclear Information System (INIS)

    Toshimas, M.; Nobuo, S.

    1983-01-01

    For the purpose of providing experimental data to assess neutron streaming calculations, neutron flux measurements were performed along the axes of the steel-walled annular ducts set up in a water shield of the pool-type reactor JRR-4. An annular duct simulated the air gap around the main coolant pipe. Another duct simulated the streaming path around the primary circulating pump of the integrated-type marine reactor. A 90-deg bend annular duct was also studied. In a set of measurements, the distance Z between the core center and the duct axis and the annular gap width delta were taken as parameters, that is, Z = 0, 80, and 160 cm and delta = 2.2, 4.7, and 10.1 cm. The reaction rates and the fluxes measured by the activation method are given in terms of absolute magnitude within an accuracy of + or - 30%. An empirical formula is derived based on those measured data, which describes the axial distribution of the neutron flux in the steel-walled annular duct in reactor shields. It is expressed by a simple function of the axial distance in units of the square root of the line-of-sight area, S /SUB l/ . The accuracy of the formula is examined by taking into account the duct location with respect to the reactor core, the neutron energy, the steel wall thickness, and the media outside of the steel wall. The accuracy of the formula is, in general, <30% in the axial distance between 3√S /SUB l/ and 30√S /SUB l/

  9. Development of EASYQAD version β. A visualization code system for gamma and neutron shielding calculations

    International Nuclear Information System (INIS)

    Kim, Jae Cheon; Kim, Soon Young; Lee, Hwan Soo; Ha, Pham Nhu Viet; Kim, Jong Kyung

    2008-01-01

    EASYQAD version β was developed by MATLAB GUI (Graphical User Interface) as a visualization code system based on QAD-CGGP-A point-kernel code for convenient shielding calculations of gammas and neutrons. It consists of four graphic interface modules including GEOMETRY, INPUT, OUTPUT, and SHIELD. These modules were compiled in C++ programming language by using the MATLAB Compiler Toolbox to form a stand-along code system that can be run on the Windows XP operating system without MATLAB installation. In addition, EASYQAD version β has user-friendly graphical interfaces and, additionally, many useful functions in comparison with QAD- CGGP-A such as common material library, line and grid detectors, and multi-group energy calculations so as to increase its applicability in the field of radiation shielding analysis. It is a powerful tool for non-experts to analyze easily the shielding problems without special training. Therefore, EASYOAD version β is expected to contribute effectively to the development of radiation shielding analysis by providing users in medical and industrial fields with an efficient radiation shielding code. (author)

  10. Concrete shielding for nuclear ship 'Mutsu'

    International Nuclear Information System (INIS)

    Nagase, Tetsuo; Nakajima, Tadao; Okumura, Tadahiko; Saito, Tetsuo

    1983-01-01

    The nuclear ship ''Mutsu'' was constructed in 1970 as the fourth in the world. On September 1, 1974, during the power raising test in the Pacific Ocean, radiation leak was detected. As the result of investigation, it was found that the cause was the fast neutrons streaming through the gap between the reactor pressure vessel and the primary shield. In order to repair the shielding facility, the Japan Nuclear Ship Research Development Agency carried out research and development and shielding design. It was decided to adopt serpentine concrete for the primary shield, which is the excellent moderator of fast neutrons even at high temperature, and heavy concrete for the secondary shield, which is effective for shielding both gamma ray and neutron beam. The repair of shielding was carried out in the Sasebo Shipyard, and completed in August, 1982. The outline of the repair work is reported. The weight increase was about 300 t. The conditions of the shielding design, the method of shielding analysis, the performance required for the shielding concrete, the preliminary experiment on heavy concrete and the construction works of serpentine concrete and heavy concrete are described. (Kako, I.)

  11. Neutron flux measurement and thermal power calibration of the IAN-R1 TRIGA reactor

    Energy Technology Data Exchange (ETDEWEB)

    Sarta Fuentes, Jose A.; Castiblanco Bohorquez, Luis A

    2008-10-29

    The IAN-R1 TRIGA reactor in Colombia was initially fueled with MTR-HEU enriched to 93% U-235, operated since 1965 at 10 kW, and was upgraded to 30 kW in 1980. General Atomics achieved in 1997 the conversion of HEU fuel to LEU fuel TRIGA type, and upgraded the reactor power to 100 kW. Since the IAN-R1 TRIGA reactor was in an extended shutdown during seven years, it was necessary to repeat some results of the commissioning test conducted in 1997. The thermal power calibration was carried out using the calorimetric method. The reactor was operated approximately at 20 kW during 3.5 hours, with manual power corrections since the automatic control system failed and with the forced refrigeration off. During the calorimetric experiment, the pool temperature was measured with a RTD which is installed near to the core. The dates were collected in intervals of 30 minutes. For establishing thermal power reactor, the water temperature versus the running were registered. For a calculated tank volume of 16 m{sup 3}, the tank constant calculated for the IAN-R1 TRIGA reactor is 0.0539 C/kW-hr. The reactor power determined was 19 kW. The core configuration is a rectangular grid plate that holds a combination of 4-rod and 3-rod clusters. The core contains 50 fuel rods with LEU fuel TRIGA (UZr H1.6) type enriched to 19.7%. The radial reflector consists of twenty graphite elements six of which are used for isotope production. The top an bottom reflectors are the cylindrical graphite end reflectors which are installed above and below of the active fuel section in each fuel rod. The spatial dependence of thermal neutron flux was measured axially in the 3-rod clusters 4C, 3D, 5E and in the 4F graphite element. The spatial distribution of the thermal neutron was determined using a self-powered detector and the absolute value of thermal neutron flux was determined by a gold activation detector. The (n, b- ) reaction is applied to determine the relative spatial distribution of thermal

  12. Experimental validation of thermal design of top shield for a pool type SFR

    International Nuclear Information System (INIS)

    Aithal, Sriramachandra; Babu, V. Rajan; Balasubramaniyan, V.; Velusamy, K.; Chellapandi, P.

    2016-01-01

    Highlights: • Overall thermal design of top shield in a SFR is experimentally verified. • Air jet cooling is effective in ensuring the temperatures limits for top shield. • Convection patterns in narrow annulus are in line with published CFD results. • Wire mesh insulation ensures gradual thermal gradient at top portion of main vessel. • Under loss of cooling scenario, sufficient time is available for corrective action. - Abstract: An Integrated Top Shield Test Facility towards validation of thermal design of top shield for a pool type SFR has been conceived, constructed & commissioned. Detailed experiments were performed in this experimental facility having full-scale features. Steady state temperature distribution within the facility is measured for various heater plate temperatures in addition to simulating different operating states of the reactor. Following are the important observations (i) jet cooling system is effective in regulating the roof slab bottom plate temperature and thermal gradient across roof slab simulating normal operation of reactor, (ii) wire mesh insulation provided in roof slab-main vessel annulus is effective in obtaining gradual thermal gradient along main vessel top portion and inhibiting the setting up of cellular convection within annulus and (iii) cellular convection with four distinct convective cells sets in the annular gap between roof slab and small rotatable plug measuring ∼ϕ4 m in diameter & gap width varying from 16 mm to 30 mm. Repeatability of results is also ensured during all the above tests. The results presented in this paper is expected to provide reference data for validation of thermal hydraulic models in addition to serving as design validation of jet cooling system for pool type SFR.

  13. Thermalization of monoenergetic neutrons in a concrete room

    Energy Technology Data Exchange (ETDEWEB)

    Vega C, H.R.; Manzanares A, E.; Hernandez D, V.M.; Mercado, G.A. [UAZ, A.P. 336, 98000 Zacatecas (Mexico); Iniguez, M.P.; Martin M, A. [Universidad de Valladolid, (Spain)

    2006-07-01

    The thermalization of neutrons from monoenergetic neutron sources in a concrete room has been studied. During calibration of neutron detectors it is mandatory to make corrections due to neutron scattering produced by the room walls, therefore this factor must be known in advance. The scattered neutrons are thermalized and produce a neutron field that is directly proportional to source strength and inversely proportional to room total wall-surfaces, the proportional coefficient has been calculated for neutrons whose energy goes from 1 eV to 20 MeV. This coefficient was calculated using Monte Carlo methods for 150, 200 and 300 cm-radius spherical cavity, where monoenergetic neutrons were located at the center, along the spherical cavity radius neutron spectra were calculated at several source-to-detector distances inside the cavity. The obtained coefficient is almost three times larger than the factor normally utilized. (Author)

  14. An optimized ultra-fine energy group structure for neutron transport calculations

    International Nuclear Information System (INIS)

    Huria, Harish; Ouisloumen, Mohamed

    2008-01-01

    This paper describes an optimized energy group structure that was developed for neutron transport calculations in lattices using the Westinghouse lattice physics code PARAGON. The currently used 70-energy group structure results in significant discrepancies when the predictions are compared with those from the continuous energy Monte Carlo methods. The main source of the differences is the approximations employed in the resonance self-shielding methodology. This, in turn, leads to ambiguous adjustments in the resonance range cross-sections. The main goal of developing this group structure was to bypass the self-shielding methodology altogether thereby reducing the neutronic calculation errors. The proposed optimized energy mesh has 6064 points with 5877 points spanning the resonance range. The group boundaries in the resonance range were selected so that the micro group cross-sections matched reasonably well with those derived from reaction tallies of MCNP for a number of resonance absorbers of interest in reactor lattices. At the same time, however, the fast and thermal energy range boundaries were also adjusted to match the MCNP reaction rates in the relevant ranges. The resulting multi-group library was used to obtain eigenvalues for a wide variety of reactor lattice numerical benchmarks and also the Doppler reactivity defect benchmarks to establish its adequacy. (authors)

  15. New evaluation of thermal neutron scattering libraries for light and heavy water

    Directory of Open Access Journals (Sweden)

    Marquez Damian Jose Ignacio

    2017-01-01

    Full Text Available In order to improve the design and safety of thermal nuclear reactors and for verification of criticality safety conditions on systems with significant amount of fissile materials and water, it is necessary to perform high-precision neutron transport calculations and estimate uncertainties of the results. These calculations are based on neutron interaction data distributed in evaluated nuclear data libraries. To improve the evaluations of thermal scattering sub-libraries, we developed a set of thermal neutron scattering cross sections (scattering kernels for hydrogen bound in light water, and deuterium and oxygen bound in heavy water, in the ENDF-6 format from room temperature up to the critical temperatures of molecular liquids. The new evaluations were generated and processable with NJOY99 and also with NJOY-2012 with minor modifications (updates, and with the new version of NJOY-2016. The new TSL libraries are based on molecular dynamics simulations with GROMACS and recent experimental data, and result in an improvement of the calculation of single neutron scattering quantities. In this work, we discuss the importance of taking into account self-diffusion in liquids to accurately describe the neutron scattering at low neutron energies (quasi-elastic peak problem. To improve modeling of heavy water, it is important to take into account temperature-dependent static structure factors and apply Sköld approximation to the coherent inelastic components of the scattering matrix. The usage of the new set of scattering matrices and cross-sections improves the calculation of thermal critical systems moderated and/or reflected with light/heavy water obtained from the International Criticality Safety Benchmark Evaluation Project (ICSBEP handbook. For example, the use of the new thermal scattering library for heavy water, combined with the ROSFOND-2010 evaluation of the cross sections for deuterium, results in an improvement of the C/E ratio in 48 out of

  16. New evaluation of thermal neutron scattering libraries for light and heavy water

    Science.gov (United States)

    Marquez Damian, Jose Ignacio; Granada, Jose Rolando; Cantargi, Florencia; Roubtsov, Danila

    2017-09-01

    In order to improve the design and safety of thermal nuclear reactors and for verification of criticality safety conditions on systems with significant amount of fissile materials and water, it is necessary to perform high-precision neutron transport calculations and estimate uncertainties of the results. These calculations are based on neutron interaction data distributed in evaluated nuclear data libraries. To improve the evaluations of thermal scattering sub-libraries, we developed a set of thermal neutron scattering cross sections (scattering kernels) for hydrogen bound in light water, and deuterium and oxygen bound in heavy water, in the ENDF-6 format from room temperature up to the critical temperatures of molecular liquids. The new evaluations were generated and processable with NJOY99 and also with NJOY-2012 with minor modifications (updates), and with the new version of NJOY-2016. The new TSL libraries are based on molecular dynamics simulations with GROMACS and recent experimental data, and result in an improvement of the calculation of single neutron scattering quantities. In this work, we discuss the importance of taking into account self-diffusion in liquids to accurately describe the neutron scattering at low neutron energies (quasi-elastic peak problem). To improve modeling of heavy water, it is important to take into account temperature-dependent static structure factors and apply Sköld approximation to the coherent inelastic components of the scattering matrix. The usage of the new set of scattering matrices and cross-sections improves the calculation of thermal critical systems moderated and/or reflected with light/heavy water obtained from the International Criticality Safety Benchmark Evaluation Project (ICSBEP) handbook. For example, the use of the new thermal scattering library for heavy water, combined with the ROSFOND-2010 evaluation of the cross sections for deuterium, results in an improvement of the C/E ratio in 48 out of 65

  17. Simultaneous thermal neutron decay time and porosity logging system

    International Nuclear Information System (INIS)

    Shultz, W.E.

    1980-01-01

    A method for simultaneously determining the porosity and thermal neutron capture cross-section of earth formations in the vicinity of a well borehole is claimed. It comprises the following steps: passing a well tool into a cased well borehole. The tool has a pulsed source of fast neutrons, a combination fast neutron and gamma ray detector and an epithermal neutron detector; repetitively irradiating the earth formations in the vicinity of the borehole with bursts of fast neutrons; detecting the fast neutron and epithermal neutron populations in the borehole (during the neutron bursts) and generating first and second measurement signals; detecting for second and third time intervals during the time between the neutron bursts, the gamma radiation present in the borehole due to the capture of thermalized neutrons by the nuclei of elements comprising the earth formations and generating third and fourth measurement signals; and combining the first and second measurement signals according to a predetermined relationship to derive an indication of the porosity of the earth formations and combining the third and fourth measurement signals to derive an indication of the thermal neutron capture cross-section of the earth formations

  18. Characteristics of the quarry as shielding for "2"4"1AmBe neutrons and monoenergetic photons

    International Nuclear Information System (INIS)

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

    2016-09-01

    Shielding is an important element in radiation protection since allows the management of radiation sources. Currently there are different materials of natural or anthropogenic origin that are used as shielding for both photons and neutrons. The quarry is a material of natural origin and abundant in our country, which is used in construction or for the manufacture of sculptures, however its characteristics as shielding have not been reported. In this paper we report some of the properties of the quarry as shielding for monoenergetic photons and for neutrons produced by an isotopic neutron source of "2"4"1AmBe. A quarry piece was used to determine its density and its chemical composition, with the XCOM code the elemental composition was determined and the mass interaction and total attenuation coefficients of the quarry were determined with photons of 10"-"3 to 10"-"5 MeV; the interaction coefficients included coherent dispersion, photoelectric absorption, Compton dispersion and the production of pairs in the nuclear and electronic field. Using the MCNP5 code, a narrow geometry attenuation experiment was modeled and the photon fluence was estimated that reaches a point detector at a distance of 42 cm from a point source, isotropic and monoenergetic photon when the source and the point detector were added quarry pieces of different thicknesses. The reduction of the number of photons as a function of the thickness of the quarry was used to determine the coefficient of linear attenuation of the quarry before photons of 0.03, 0.07, 0.1, 0.3, 1, 2 and 3 MeV that were the same as those calculated with the XCOM code. With the MCNP, the K a and H(10) transmission curves were also calculated. This same model was used to determined the variation of the "2"4"1AmBe neutron spectrum as a function of quarry thickness, as well as the E_R_O_T and H(10) transmission curves. (Author)

  19. Three-Dimensional (X,Y,Z) Deterministic Analysis of the PCA-Replica Neutron Shielding Benchmark Experiment using the TORT-3.2 Code and Group Cross Section Libraries for LWR Shielding and Pressure Vessel Dosimetry

    OpenAIRE

    Pescarini Massimo; Orsi Roberto; Frisoni Manuela

    2016-01-01

    The PCA-Replica 12/13 (H2O/Fe) neutron shielding benchmark experiment was analysed using the ORNL TORT-3.2 3D SN code. PCA-Replica, specifically conceived to test the accuracy of nuclear data and transport codes employed in LWR shielding and radiation damage calculations, reproduces a PWR ex-core radial geometry with alternate layers of water and steel including a PWR pressure vessel simulator. Three broad-group coupled neutron/photon working cross section libraries in FIDO-ANISN format with ...

  20. Nuclear reactor shield including magnesium oxide

    International Nuclear Information System (INIS)

    Rouse, C.A.; Simnad, M.T.

    1981-01-01

    An improvement is described for nuclear reactor shielding of a type used in reactor applications involving significant amounts of fast neutron flux. The reactor shielding includes means providing structural support, neutron moderator material, neutron absorber material and other components, wherein at least a portion of the neutron moderator material is magnesium in the form of magnesium oxide either alone or in combination with other moderator materials such as graphite and iron

  1. Nuclear reactor neutron shielding

    Science.gov (United States)

    Speaker, Daniel P; Neeley, Gary W; Inman, James B

    2017-09-12

    A nuclear reactor includes a reactor pressure vessel and a nuclear reactor core comprising fissile material disposed in a lower portion of the reactor pressure vessel. The lower portion of the reactor pressure vessel is disposed in a reactor cavity. An annular neutron stop is located at an elevation above the uppermost elevation of the nuclear reactor core. The annular neutron stop comprises neutron absorbing material filling an annular gap between the reactor pressure vessel and the wall of the reactor cavity. The annular neutron stop may comprise an outer neutron stop ring attached to the wall of the reactor cavity, and an inner neutron stop ring attached to the reactor pressure vessel. An excore instrument guide tube penetrates through the annular neutron stop, and a neutron plug comprising neutron absorbing material is disposed in the tube at the penetration through the neutron stop.

  2. Thermal neutron detectors based on complex oxide crystals

    CERN Document Server

    Ryzhikov, V; Volkov, V; Chernikov, V; Zelenskaya, O

    2002-01-01

    The ways of improvement of spectrometric quality of CWO and GSO crystals have been investigated with the aim of their application in thermal neutron detectors based on radiation capture reactions. The efficiency of the neutron detection by these crystals was measured, and the obtained data were compared with the results for sup 6 LiI(Tl) crystals. It is shown that the use of complex oxide crystals and neutron-absorption filters for spectrometry of thermal and resonance neutrons could be a promising method in combination with computer data processing. Numerical calculations are reported for spectra of gamma-quanta due to radiation capture of the neutrons. To compensate for the gamma-background lines, we used a crystal pair of heavy complex oxides with different sensitivity to neutrons.

  3. Accuracy estimation for intermediate and low energy neutron transport calculation with Monte Carlo code MCNP

    International Nuclear Information System (INIS)

    Kotegawa, Hiroshi; Sasamoto, Nobuo; Tanaka, Shun-ichi

    1987-02-01

    Both ''measured radioactive inventory due to neutron activation in the shield concrete of JPDR'' and ''measured intermediate and low energy neutron spectra penetrating through a graphite sphere'' are analyzed using a continuous energy model Monte Carlo code MCNP so as to estimate calculational accuracy of the code for neutron transport in thermal and epithermal energy regions. Analyses reveal that MCNP calculates thermal neutron spectra fairly accurately, while it apparently over-estimates epithermal neutron spectra (of approximate 1/E distribution) as compared with the measurements. (author)

  4. 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

  5. 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

  6. Integrated neutron/gamma-ray portal monitors for nuclear safeguards

    International Nuclear Information System (INIS)

    Fehlau, P.E.

    1994-01-01

    Radiation monitoring is one nuclear-safeguards measure used to protect against the theft of special nuclear materials (SNM) by pedestrians departing from SNM access areas. The integrated neutron/gamma-ray portal monitor is an ideal radiation monitor for the task when the SNM is plutonium. It achieves high sensitivity for detecting both bare and shielded plutonium by combining two types of radiation detector. One type is a neutron-chamber detector, comprising a large, hollow, neutron moderator that contains a single thermal-neutron proportional counter. The entrance wall of each chamber is thin to admit slow neutrons from plutonium contained in a moderating shield, while the other walls are thick to moderate fast neutrons from bare or lead-shielded plutonium so that they can be detected. The other type of detector is a plastic scintillator that is primarily for detecting gamma rays from small amounts of unshielded plutonium. The two types of detector are easily integrated by making scintillators part of the thick back wall of each neutron chamber or by inserting them into each chamber void. The authors compared the influence of the two methods of integration on detecting neutrons and gamma rays, and they examined the effectiveness of other design factors and the methods for signal detection as well

  7. Using thermalizers in measuring 'Ukryttia' object's FCM neutron fluxes

    CERN Document Server

    Krasnyanskaya, O G; Odinokin, G I; Pavlovich, V N

    2003-01-01

    The results of research of a thermalizer (heater) width influence on neutron thermalization efficiency during FCM neutron flux measuring in the 'Ukryttia' are described. The calculations of neutron flux densities were performed by the Monte-Carlo method with the help of computer code MCNP-4C for FCM different models.Three possible installations of detectors were considered: on FCM surface,inside the FCM, and inside the concrete under the FCM layer. It was shown,that in order to increase the sensitivity of neutron detectors in intermediate and fast neutrons field,and consequently, to decrease the dependence of the readings of spectral distribution of neutron flux,it is necessary to position the detector inside the so-called thermalizer or heater. The most reasonable application of thick 'heaters' is the situation, when the detector is placed on FCM surface.

  8. SB2. Experiment on secondary gamma-ray production cross sections arising from thermal-neutron capture in each of 14 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 arising from thermal-neutron capture in iron, nitrogen, sodium, aluminum, copper, titanium, calcium, potassium, chlorine, silicon, ickel, zinc, barium, sulfur and a type 321 stainless steel. 1 figure, 30 tables

  9. Analysis of the sensitivity concept of self-powered neutron detector (SPND)

    International Nuclear Information System (INIS)

    Moreira, O.; Lescano, H.

    2012-01-01

    Self powered neutron detectors (SPND) are widely used to monitor the neutron flux, either in nuclear as in irradiation facilities and medical treatments. However, the physical meaning of the parameter that is used to relate the detector signal (an electrical current) with the neutron flux, i.e., the sensitivity of the detector, has not been sufficiently analyzed. Since the definition of sensitivity, ε=i/φ is calculated for particular reactor conditions, i.e., for thermal neutrons at room temperature, it does not take into account the deviation originated from other conditions of temperature (above ambient), as found for example in nuclear power plants. In this work we calculated the microscopic cross section weighted with the neutron flux, defined in the usual way. This weighted microscopic cross section reveals the no proportionality between the absorption rate and the neutron flux, exhibiting the problem that the SPND current signal has to properly represent the neutron flux (author)

  10. LANSCE steady state unperturbed thermal neutron fluxes at 100 μA

    International Nuclear Information System (INIS)

    Russell, G.J.

    1989-01-01

    The ''maximum'' unperturbed, steady state thermal neutron flux for LANSCE is calculated to be 2 /times/ 10 13 n/cm 2 -s for 100 μA of 800-MeV protons. This LANSCE neutron flux is a comparable entity to a steady state reactor thermal neutron flux. LANSCE perturbed steady state thermal neutron fluxes have also been calculated. Because LANSCE is a pulsed neutron source, much higher ''peak'' (in time) neutron fluxes can be generated than at a steady state reactor source. 5 refs., 5 figs

  11. In-beam background suppression shield

    DEFF Research Database (Denmark)

    Santoro, V.; Cai, Xiao Xiao; DiJulio, D. D.

    2015-01-01

    The long (3 ms) proton pulse of the European Spallation Source (ESS) gives rise to unique and potentially high backgrounds for the instrument suite. In such a source an instrument's capabilities will be limited by its Signal to Noise (S/N) ratio. The instruments with a direct view of the moderator......, which do not use a bender to help mitigate the fast neutron background, are the most challenging. For these beam lines we propose the innovative shielding of placing blocks of material directly into the guide system, which allow a minimum attenuation of the cold and thermal fluxes relative...... to the background suppression. This shielding configuration has been worked into a beam line model using Geant4. We study particularly the advantages of single crystal sapphire and silicon blocks....

  12. Response of six neutron survey meters in mixed fields of fast and thermal neutrons.

    Science.gov (United States)

    Kim, S I; Kim, B H; Chang, I; Lee, J I; Kim, J L; Pradhan, A S

    2013-10-01

    Calibration neutron fields have been developed at KAERI (Korea Atomic Energy Research Institute) to study the responses of commonly used neutron survey meters in the presence of fast neutrons of energy around 10 MeV. The neutron fields were produced by using neutrons from the (241)Am-Be sources held in a graphite pile and a DT neutron generator. The spectral details and the ambient dose equivalent rates of the calibration fields were established, and the responses of six neutron survey meters were evaluated. Four single-moderator-based survey meters exhibited an under-responses ranging from ∼9 to 55 %. DINEUTRUN, commonly used in fields around nuclear reactors, exhibited an over-response by a factor of three in the thermal neutron field and an under-response of ∼85 % in the mixed fields. REM-500 (tissue-equivalent proportional counter) exhibited a response close to 1.0 in the fast neutron fields and an under-response of ∼50 % in the thermal neutron field.

  13. Adaptation of the HCPB DEMO TBM as breeding blanket for ITER : Neutronic and thermal analyses

    International Nuclear Information System (INIS)

    Aquaro, D.; Morellini, D.; Cerullo, N.

    2006-01-01

    Two breeding blanket are presently developed in Europe for the DEMO reactor: the first one, the Helium Cooled Lithium Lead (HCLL) uses a liquid breeder while the other , the Helium Cooled Pebble Bed (HCPB), uses a solid breeder in form of pebble bed. The modules of these blankets, called Test Blanket Modules (TBM) will be located in correspondence of the equatorial ports of ITER in order to be tested. ITER FEAT was designed with shielding blankets, therefore in the final stage of the experiment, in the foreseen tritium -deuterium operation phase, the tritium will be supplied to the reactor and not produced inside it. Since the production of tritium is of main importance for the feasibility of a nuclear fusion reactor, perhaps in the ITER final stage, the shielding blanket could be substituted by means of a breeding blanket. The geometry and composition of this breeding blanket would be, of course, similar to that of TBM which demonstrated to have the best performances. This paper illustrates a neutronic and thermal analysis of an hypothetical triziogen blanket for ITER FEAT made similar to a HCPB test module. The main aims of the performed analyses are to determine the Tritium Breeding Ratio (TBR) considering different solid breeders (Li 4 SiO 4 and Li 2 TiO 3 ) with different enrichment in 6 Li and different structural materials (a 9%CRWVTa reduced activation ferritic martensitic steel (EUROFER) or ceramic matrix composites like SiCf/SiC). The breeding blanket design is compared considering the highest value of TBR and the verification of the temperature constraints ( 550 o C for the steel, 950 o C for the breeder and 650 o C for the Beryllium). The neutronic analyses have been performed by means of MCNP-4C code and the thermal analyses using the MSC-MARC code. A TBR about equal 1 was obtained with a SiCf/SiC structural material and a Li 4 SiO 4 breeder. The performed analyses have to be considered preliminary and an academic exercise, nevertheless they could give

  14. Simbol-X Mirror Module Thermal Shields: I-Design and X-Ray Transmission

    Science.gov (United States)

    Collura, A.; Barbera, M.; Varisco, S.; Basso, S.; Pareschi, G.; Tagliaferri, G.; Ayers, T.

    2009-05-01

    The Simbol-X mission is designed to fly in formation flight configuration. As a consequence, the telescope has both ends open to space, and thermal shielding at telescope entrance and exit is required to maintain temperature uniformity throughout the mirrors. Both mesh and meshless solutions are presently under study for the shields. We discuss the design and the X-ray transmission.

  15. Simbol-X Mirror Module Thermal Shields: I - Design and X-Ray Transmission

    International Nuclear Information System (INIS)

    Collura, A.; Varisco, S.; Barbera, M.; Basso, S.; Pareschi, G.; Tagliaferri, G.; Ayers, T.

    2009-01-01

    The Simbol-X mission is designed to fly in formation flight configuration. As a consequence, the telescope has both ends open to space, and thermal shielding at telescope entrance and exit is required to maintain temperature uniformity throughout the mirrors. Both mesh and meshless solutions are presently under study for the shields. We discuss the design and the X-ray transmission.

  16. A design study on hyper-thermal neutron irradiation field for neutron capture therapy at Kyoto University Reactor

    International Nuclear Information System (INIS)

    Sakurai, Y.; Kobayashi, T.

    2000-01-01

    A study about the installation of a hyper-thermal neutron converter to a clinical collimator was performed, as a series of the design study on a hyper-thermal neutron irradiation field at the Heavy Water Neutron Irradiation Facility of Kyoto University Reactor. From the parametric-surveys by Monte Carlo calculation, it was confirmed that the practical irradiation field of hyper-thermal neutrons would be feasible by the modifications of the clinical collimator and the bismuth-layer structure. (author)

  17. Eddy current and mechanical support of the wendelstein 7-X thermal shield

    International Nuclear Information System (INIS)

    Shim, S. Y.; Nagel, M.; Schauer, F.

    2004-01-01

    The machine which equipped huge magnet such as fusion reactor must be in need of special care on the emergency process. Rapid drop down magnetic field generate noticeable induced current, eddy current, and it causes strong electromagnetic forces on mechanical structure. The Wendelstein 7-X consists with 5 pentagonal shaped modules, plasma vessel, and each module can be divided into two symmetric half modules. Each half-module is going to be covered by 20 pieces of plasma vessel thermal shield (PVTS). The subject of this calculation is to find appropriate support positions for PVTS which can withstand self-weight of PVTS and electromagnetic force during the emergency case within our design criterion. We report the calculation procedure and results with half-module of PVTS

  18. Calculation of neutron shielding using an unidimensional model of transportation in formulation of discrete ordinates with scattering linearly anisotropic and a speed

    International Nuclear Information System (INIS)

    Libotte, Rafael Barbosa; Alves Filho, Hermes; Oliva, Amaury Muñoz

    2017-01-01

    The physical phenomenon of transport of neutral particles in a host environment is of interest in various scientific applications, e.g., nuclear reactors, shielding calculations, radiological protection, nuclear medicine, agronomy, materials science, oil prospecting, etc. In all these areas there is a need for an accurate description of the transport of the particles in the host medium. In this class of applications are the neutron shielding problems, also referred to as 'fixed-source' problems, where the interaction of the particles with the medium does not produce new neutrons, i.e., non-multiplicative medium. In this context, the development of tools that model these problems is relevant and of a beneficial return to society. In this work, we propose the development of deterministic mathematical and computational modeling of neutron transport using the linearized equation of Boltzmann applied to neutron shielding problems. Here we present also the development of a spectro-nodal method (coarse mesh) considering the scattering phenomenon as being linearly anisotropic. We show the results using a computational application, developed in Java language, version 1.8.0 9 1

  19. New approximations for the Doppler broadening function applied to the calculation of resonance self-shielding factors

    International Nuclear Information System (INIS)

    Palma, Daniel A.; Goncalves, Alessandro C.; Martinez, Aquilino S.; Silva, Fernando C.

    2008-01-01

    The activation technique allows much more precise measurements of neutron intensity, relative or absolute. The technique requires the knowledge of the Doppler broadening function ψ(x,ξ) to determine the resonance self-shielding factors in the epithermal range G epi (τ,ξ). Two new analytical approximations for the Doppler broadening function ψ(x,ξ) are proposed. The approximations proposed are compared with other methods found in literature for the calculation of the ψ(x,ξ) function, that is, the 4-pole Pade method and the Frobenius method, when applied to the calculation of G epi (τ,ξ). The results obtained provided satisfactory accuracy. (authors)

  20. New approximations for the Doppler broadening function applied to the calculation of resonance self-shielding factors

    Energy Technology Data Exchange (ETDEWEB)

    Palma, Daniel A. [CEFET QUIMICA de Nilopolis/RJ, Rio de Janeiro (Brazil); Goncalves, Alessandro C.; Martinez, Aquilino S.; Silva, Fernando C. [COPPE/UFRJ - Programa de Engenharia Nuclear, Rio de Janeiro (Brazil)

    2008-07-01

    The activation technique allows much more precise measurements of neutron intensity, relative or absolute. The technique requires the knowledge of the Doppler broadening function psi(x,xi) to determine the resonance self-shielding factors in the epithermal range G{sub epi} (tau,xi). Two new analytical approximations for the Doppler broadening function psi(x,xi) are proposed. The approximations proposed are compared with other methods found in literature for the calculation of the psi(x,xi) function, that is, the 4-pole Pade method and the Frobenius method, when applied to the calculation of G{sub epi} (tau,xi). The results obtained provided satisfactory accuracy. (authors)