WorldWideScience

Sample records for reactor shielding concretes

  1. Evaluation of ilmenite serpentine concrete and ordinary concrete as nuclear reactor shielding

    International Nuclear Information System (INIS)

    Abulfaraj, W.H.; Kamal, S.M.

    1994-01-01

    The present study involves adapting a formal decision methodology to the selection of alternative nuclear reactor concrete shielding. Multiattribute utility theory is selected to accommodate decision maker's preferences. Multiattribute utility theory (MAU) is here employed to evaluate two appropriate nuclear reactor shielding concretes in terms of effectiveness to determine the optimal choice in order to meet the radiation protection regulations. These concretes are Ordinary concrete (O.C.) and Illmenite Serpentile concrete (I.S.C.). These are normal weight concrete and heavy weight heat resistive concrete, respectively. The effectiveness objective of the nuclear reactor shielding is defined and structured into definite attributes and subattributes to evaluate the best alternative. Factors affecting the decision are dose received by reactor's workers, the material properties as well as cost of concrete shield. A computer program is employed to assist in performing utility analysis. Based upon data, the result shows the superiority of Ordinary concrete over Illmenite Serpentine concrete. (Author)

  2. 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)

  3. Activation of TRIGA Mark II research reactor concrete shield

    International Nuclear Information System (INIS)

    Zagar, Tomaz; Ravnik, Matjaz; Bozic, Matjaz

    2002-01-01

    To determine neutron activation inside the TRIGA research reactor concrete body a special sample-holder for irradiation inside horizontal channel was developed and tested. In the sample-holder various samples can be irradiated at different concrete shielding depths. In this paper the description of the sample-holder, experiment conditions and results of long-lived activation measurements are given. Long-lived neutron-induced gamma-ray-emitting radioactive nuclides in the samples were measured with HPGe detector. The most active long-lived 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. Radiation distribution through serpentine concrete using local materials and its application as a reactor biological shield

    International Nuclear Information System (INIS)

    Kansouh, W.A.

    2012-01-01

    Highlights: ► New serpentine concrete was made and examined as a reactor biological shield. ► Ilmenite–limonite concrete is a better reactor biological shield. ► New serpentine concrete is a better reactor fast neutrons shield than ordinary and hematite–serpentine concretes. ► Serpentine concrete has lower properties as a reactor total gamma rays shields. - Abstract: In the present work attempt has been made to estimate the shielding parameters of the new serpentine concrete (density = 2.4 g/cm 3 ) using local materials on the shielding parameters for two types of heat resistant concretes, namely hematite–serpentine (density = 2.5 g/cm 3 ) and ilmenite–limonite (density = 2.9 g/cm 3 ). Shielding parameters for ordinary concrete (density = 2.3 g/cm 3 ) were also discussed. These parameters were determined experimentally for serpentine concrete and compared with previously published values for other concretes, which had also been obtained using local materials. The leakage spectra of reactor fast neutrons and total gamma photon beams from cylindrical samples of these concrete shields were also investigated using a collimated beam from ET-RR-1 reactor. A neutron–gamma spectrometer was used in order to obtain pulse height spectra of reactor fast neutrons and the total gamma rays leakage through the investigated concrete samples. These spectra were utilized to obtain the energy spectra required in these investigations. Removal cross section Σ R (E n ) and linear attenuation coefficient μ(E g ) for reactor fast neutrons and total gamma rays and their relative coefficients were evaluated and presented. Measured results were compared with those previously measured for other concretes. The results show that ilmenite–limonite concrete is a better reactor biological shield than the other three concretes. Serpentine concrete under investigation is a better reactor fast neutrons shield than ordinary and hematite–serpentine concretes. Serpentine concrete

  5. Numerical simulation of a reinforced concrete shield around a nuclear reactor

    International Nuclear Information System (INIS)

    Mahama, Mumuni Salifu

    1996-02-01

    Ghana currently operates a Research Reactor and other nuclear facilities including a Gamma Irradiation Facility, a Radiographic Non-Destructive Testing laboratory and would be operating in the nearest future a Radiotherapy Centre. Each of these has a concrete radiation shield as a major safety device. In carrying out its functions, a concrete radiation shield may be subjected to thermal and mechanical stresses. A facility for analysing these stresses is desirable. Two computer codes have been developed under this programme for radiation shielding computation and stress analysis of cylindrical reactor shields. (au)

  6. Radiological characterization of the concrete biological shield of the APSARA reactor

    OpenAIRE

    Srinivasan Priya; Srinivasan Panchapakesan; Thomas Shibu; Gopalakrishnan R.K.; Goswami A.

    2013-01-01

    The first Indian research reactor, APSARA, was utilized for various R&D programmes from 1956 until its shutdown in 2009. The biological shield of the reactor developed residual activity due to neutron irradiation during the operation of the reactor. Dose rate mapping and in-situ gamma spectrometry of the concrete structures of the reactor pool were carried out. Representative concrete samples collected from various locations were subjected to high-resolution gamma spectrometry analysis....

  7. Production of a datolite-based heavy concrete for shielding nuclear reactors and megavoltage radiotherapy rooms

    International Nuclear Information System (INIS)

    Mortazavi, S. M. J.; Mosleh-Shirazi, M.A.; Baradaran-Ghahfarokhi, M.; Siavashpour, Z.; Farshadi, A.; Ghafoori, M.; Shahvar, A.

    2010-01-01

    Biological shielding of nuclear reactors has always been a great concern and decreasing the complexity and expense of these installations is of great interest. In this study, we used datolite and galena minerals for production of a high performance heavy concrete. Materials and Methods: Datolite and galena minerals which can be found in many parts of Iran were used in the concrete mix design. To measure the gamma radiation attenuation of the Datolite and galena concrete samples, they were exposed to both narrow and wide beams of gamma rays emitted from a cobalt-60 radiotherapy unit. An Am-Be neutron source was used for assessing the shielding properties of the samples against neutrons. To test the compression strengths, both types of concrete mixes (Datolite and galena and ordinary concrete) were investigated. Results: The concrete samples had a density of 4420-4650 kg/m 3 compared to that of ordinary concrete (2300-2500 kg/m 3 ) or barite high density concrete (up to 3500 kg/m 3 ). The measured half value layer thickness of the Datolite and galena concrete samples for cobalt-60 gamma rays was much less than that of ordinary concrete (2.56 cm compared to 6.0 cm). Furthermore, the galena concrete samples had a significantly higher compressive strength as well as 20% more neutron absorption. Conclusion: The Datolite and galena concrete samples showed good shielding/engineering properties in comparison with other reported samples made, using high-density materials other than depleted uranium. It is also more economic than the high-density concretes. Datolite and galena concrete may be a suitable option for shielding nuclear reactors and megavoltage radiotherapy rooms.

  8. Production of an economic high-density concrete for shielding megavoltage radiotherapy rooms and nuclear reactors

    International Nuclear Information System (INIS)

    Mortazavi, S. M. J.; Mosleh-Shirazi, M. A.; Maheri, M. R.; Haji-pour, A.; Yousefnia, H.; Zolghadri, S.

    2007-01-01

    In megavoltage radiotherapy rooms, ordinary concrete is usually used due to its low construction costs, although higher density concrete are sometimes used, as well. The use of high-density concrete decreases the required thickness of the concrete barrier; hence, its disadvantage is its high cost. In a nuclear reactor, neutron radiation is the most difficult to shield. A method for production of economic high-density concrete witt, appropriate engineering properties would be very useful. Materials and Methods: Galena (Pb S) mineral was used to produce of a high-density concrete. Galena can be found in many parts of Iran. Two types of concrete mixes were produced. The water-to-concrete (w/c) ratios of the reference and galena concrete mixes were 0.53 and 0.25, respectively. To measure the gamma radiation attenuation of Galena concrete samples, they were exposed to a narrow beam of gamma rays emitted from a cobalt-60 therapy unit. Results: The Galena mineral used in this study had a density of 7400 kg/m 3 . The concrete samples had a density of 4800 kg/m 3 . The measured half value layer thickness of the Galena concrete samples for cobalt 60 gamma rays was much less than that of ordinary concrete (2.6 cm compared to 6.0 cm). Furthermore, the galena concrete samples had significantly higher compressive strength (500 kg/cm 2 compared to 300 kg/cm 2 ). Conclusion: The Galena concrete samples made in our laboratories had showed good shielding/engineering properties in comparison with all samples made by using high-density materials other than depleted uranium. Based on the preliminary results, Galena concrete is maybe a suitable option where high-density concrete is required in megavoltage radiotherapy rooms as well as nuclear reactors

  9. Analysis of crack-formation in the shielding concrete of a TRIGA Mark II reactor

    International Nuclear Information System (INIS)

    Linsbauer, H.; Maydl, P.

    1978-01-01

    Within a short time after the start-up of the reactor several cracks appeared at the concrete surface and the number and width of the cracks had grown till now. Experimental and theoretical analysis were made in order to investigate the origin of the cracks and to prevent further crack increase. Crack movement was measured by inductive gages and simultaneously the temperature of the cooling water in the reactor tank at the top and at the bottom as well as the air and the concrete temperature were recorded. The calculations of the thermal stresses were made in two independent ways: 1. Analytically, simulating the shielding concrete as an infinite hollow cylinder of constant thickness and 2. Using the Finite Element method, for a better description of the geometry. It was concluded that the cracks of the shielding concrete are exclusively caused by the thermal stresses. The thermal insulation at the lower part of the shielding is not effective. The structural system of the shielding concrete as a monolithic block without joints produces automatically tensile stresses

  10. Activation of concrete samples from the biological shield of the ASTRA reactor

    International Nuclear Information System (INIS)

    Smecka, F.

    2006-09-01

    Drill cores from the biological shield of the ASTRA reactor in Seibersdorf were taken and milled because of the different size of the Baryt crystals in the concrete in order to get homogenous samples. The powder samples were put into bore holes of a graphite block which was placed into the thermal column of the TRIGA Mark II reactor. The block was irradiated for 10 minutes at a reactor power of 25 kW. After one hour the dose rate was examined and the samples were ready for further save handling. The gamma spectrum was measured with a Ge detector and the results were compared with simulation data. (nevyjel)

  11. Development of a low activation concrete shielding wall by multi-layered structure for a fusion reactor

    International Nuclear Information System (INIS)

    Sato, Satoshi; Maegawa, Toshio; Yoshimatsu, Kenji; Sato, Koichi; Nonaka, Akira; Takakura, Kosuke; Ochiai, Kentaro; Konno, Chikara

    2011-01-01

    A multi-layered concrete structure has been developed to reduce induced activity in the shielding for neutron generating facilities such as a fusion reactor. The multi-layered concrete structure is composed of: (1) an inner low activation concrete, (2) a boron-doped low activation concrete as the second layer, and (3) ordinary concrete as the outer layer of the neutron shield. With the multi-layered concrete structure the volume of boron is drastically decreased compared to a monolithic boron-doped concrete. A 14 MeV neutron shielding experiment with multi-layered concrete structure mockups was performed at FNS and several reaction rates and induced activity in the mockups were measured. This demonstrated that the multi-layered concrete effectively reduced low energy neutrons and induced activity.

  12. Shielding concretes for liquid sodium cooled nuclear reactors

    International Nuclear Information System (INIS)

    Massa, G.; De Stefano, R.; Chatterji, S.; Maniscalco, V.

    1983-01-01

    The scope of the research was to develop concretes with low water loss and high strength with time during continuing operation in relatively high temperature environment. The required physical properties of the concrete are: - density greater than or equal to 3.8 t/m 3 - 28 days compressive strength greater than 500 kg/cm 2 - retained water at 200 0 C higher than 10 litres per cubic meter. The optimum mixture is determined by the following processes: - selection of materials capable of yielding concretes of the required density, - chemical/physical analysis of the mix components, - optimization/determination of required sieve analyses with quantitative analysis of each sieve group, - determination of the fineness modulus drawn from the selected sieve analyses, - preparation of various mixtures with the criteria of minimizing water content, obtaining high workability (to meet the needs of the various casting operations) and 28 days compressive strength of at least 500 kilograms per square centimeter, and achieving retention of the required physical properties at 200 degrees centigrades, - study of mixture response to variations in temperature with the scope of determining the modulus of elasticity. (orig./HP)

  13. REACTOR SHIELD

    Science.gov (United States)

    Wigner, E.P.; Ohlinger, L.E.; Young, G.J.; Weinberg, A.M.

    1959-02-17

    Radiation shield construction is described for a nuclear reactor. The shield is comprised of a plurality of steel plates arranged in parallel spaced relationship within a peripheral shell. Reactor coolant inlet tubes extend at right angles through the plates and baffles are arranged between the plates at right angles thereto and extend between the tubes to create a series of zigzag channels between the plates for the circulation of coolant fluid through the shield. The shield may be divided into two main sections; an inner section adjacent the reactor container and an outer section spaced therefrom. Coolant through the first section may be circulated at a faster rate than coolant circulated through the outer section since the area closest to the reactor container is at a higher temperature and is more radioactive. The two sections may have separate cooling systems to prevent the coolant in the outer section from mixing with the more contaminated coolant in the inner section.

  14. Application of SCALE 6.1 MAVRIC Sequence for Activation Calculation in Reactor Primary Shield Concrete

    International Nuclear Information System (INIS)

    Kim, Yong IL

    2014-01-01

    Activation calculation requires flux information at desired location and reaction cross sections for the constituent elements to obtain production rate of activation products. Generally it is not an easy task to obtain fluxes or reaction rates with low uncertainties in a reasonable time for deep penetration problems by using standard Monte Carlo methods. The MAVRIC (Monaco with Automated Variance Reduction using Importance Calculations) sequence in SCALE 6.1 code package is intended to perform radiation transport on problems that are too challenging for standard, unbiased Monte Carlo methods. And the SCALE code system provides plenty of ENDF reaction types enough to consider almost all activation reactions in the nuclear reactor materials. To evaluate the activation of the important isotopes in primary shield, SCALE 6.1 MAVRIC sequence has been utilized for the KSNP reactor model and the calculated results are compared to the isotopic activity concentration of related standard. Related to the planning for decommission, the activation products in concrete primary shield such as Fe-55, Co-60, Ba-133, Eu-152, and Eu-154 are identified as important elements according to the comparisons with related standard for exemption. In this study, reference data are used for the concrete compositions in the activation calculation to see the applicability of MAVRIC code to the evaluation of activation inventory in the concrete primary shield. The composition data of trace elements as shown in Table 1 are obtained from various US power plant sites and accordingly they have large variations in quantity due to the characteristics of concrete composition. In practical estimation of activation radioactivity for a specific plant related to decommissioning, rigorous chemical analysis of concrete samples of the plant would first have to be performed to get exact information for compositions of concrete. Considering the capability of solving deep penetration transport problems and richness

  15. Radiation shielding concrete

    International Nuclear Information System (INIS)

    Kunishima, Shigeru.

    1990-01-01

    The radiation shielding concretes comprise water, cement, fine aggregates consisting of serpentines and blown mist slags, coarse aggregates consisting of serpentines and kneading materials. Since serpentines containing a relatively great amount of water of crystallization in rocks as coarse aggregates and fine aggregates, the hydrogen content in the radiation shielding concretes is increased and the neutron shielding effect is improved. In addition, since serpentines are added as the fine aggregates and blown mists slags of a great specific gravity are used, the specific gravity of the shielding concretes is increased to improve the γ-ray shielding effect. Further, by the use of the kneading material having a water reducing effect and fluidizing effect, and by the bearing effect of the spherical blown mist slags used as the fine aggregates, concrete fluidity can be increased. Accordingly, workability of the radiation shielding concretes can be improved. (T.M.)

  16. 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.)

  17. Concrete radiation shielding

    International Nuclear Information System (INIS)

    Kaplan, M.F.

    1989-01-01

    The increased use of nuclear energy has given rise to a growth in the amount of artificially produced radiation and radioactive materials. The design and construction of shielding to protect people, equipment and structures from the effects of radiation has never been more important. Experience has shown that concrete is an effective, versatile and economical material for the construction of radiation shielding. This book provides information on the principles governing the interaction of radiation with matter and on relevant nuclear physics to give the engineer an understanding of the design and construction of concrete shielding. It covers the physical, mechanical and nuclear properties of concrete; the effects of elevated temperatures and possible damage to concrete due to radiation; basic procedures for the design of concrete radiation shields and finally the special problems associated with their construction and cost. Although written primarily for engineers concerned with the design and construction of concrete shielding, the book also reviews the widely scattered data and information available on this subject and should therefore be of interest to students and those wishing to research further in this field. (author)

  18. Radiation shield for PWR reactors

    International Nuclear Information System (INIS)

    Esenov, Amra; Pustovgar, Andrey

    2013-01-01

    One of the chief structures of a reactor pit is a 'dry' shield. Setting up a 'dry' shield includes the technologically complex process of thermal processing of serpentinite concrete. Modern advances in the area of materials technology permit avoiding this complex and demanding procedure, and this significantly decreases the duration, labor intensity, and cost of setting it up. (orig.)

  19. 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.)

  20. Concrete shielding exterior to iron

    International Nuclear Information System (INIS)

    Yurista, P.; Cossairt, D.

    1983-08-01

    A rule of thumb at Fermilab has been to use 3 feet of concrete exterior to iron shielding. A recent design of a shield with a severe dimensional constraint has prompted a re-evaluation of this rule of thumb and has led to the following calculations of the concrete thickness required to nullify this problem. 4 references, 4 figures

  1. 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.

  2. Beta Bremsstrahlung dose in concrete shielding

    Energy Technology Data Exchange (ETDEWEB)

    Manjunatha, H.C., E-mail: manjunatha@rediffmail.com [Department of Physics, Government college for women, Kolar 563101, Karnataka (India); Chandrika, B.M. [Shravana, 592, Ist Cross, Behind St.Anne s School, PC Extension, Kolar 563101, Karnataka (India); Rudraswamy, B. [Department of Physics, Bangalore University, Bangalore 560056, Karnataka (India); Sankarshan, B.M. [Shravana, 592, Ist Cross, Behind St.Anne s School, PC Extension, Kolar 563101, Karnataka (India)

    2012-05-11

    In a nuclear reactor, beta nuclides are released during nuclear reactions. These betas interact with shielding concrete and produces external Bremsstrahlung (EB) radiation. To estimate Bremsstrahlung dose and shield efficiency in concrete, it is essential to know Bremsstrahlung distribution or spectra. The present work formulated a new method to evaluate the EB spectrum and hence Bremsstrahlung dose of beta nuclides ({sup 32}P, {sup 89}Sr, {sup 90}Sr-{sup 90}Y, {sup 90}Y, {sup 91}Y, {sup 208}Tl, {sup 210}Bi, {sup 234}Pa and {sup 40}K) in concrete. The Bremsstrahlung yield of these beta nuclides in concrete is also estimated. The Bremsstrahlung yield in concrete due to {sup 90}Sr-{sup 90}Y is higher than those of other given nuclides. This estimated spectrum is accurate because it is based on more accurate modified atomic number (Z{sub mod}) and Seltzer's data, where an electron-electron interaction is also included. Presented data in concrete provide a quick and convenient reference for radiation protection. The present methodology can be used to calculate the Bremsstrahlung dose in nuclear shielding materials. It can be quickly employed to give a first pass dose estimate prior to a more detailed experimental study. - Highlights: Black-Right-Pointing-Pointer Betas released in a nuclear reactor interact with shielding concrete and produces Bremsstrahlung. Black-Right-Pointing-Pointer The present work formulated a new method to evaluate the Bremsstrahlung spectrum and dose in concrete. Black-Right-Pointing-Pointer Presented data in concrete provide a quick and convenient reference for radiation protection.

  3. Studies on the use of the hematitic heavy concrete in the shielding of nuclear reactors

    International Nuclear Information System (INIS)

    Mello, M.J. de.

    1980-12-01

    The main difficulties and and the necessary adaptations in the usual techniques for the regular concrete, are intended to show. In particular the problems of dosing and the methods of vibration an homogenization in the setting of heavy concrete will need more detailed research. It was showed possible to obtain densities of 4 g/cm 3 and more. The atenuation coeficients obtained for the range of energy of gamma rays studied are much greater than the ones obtained with most concretes. The interpretation of measurements of these coeficients showed agreement with the characteristics of the materials and measuring methods used. The matching of the measured atenuation coeficients with the values calculated from the analysis of the mean composition of the heavy concrete also showed coherence with a deviation of 4% in most cases. (Author) [pt

  4. Special concrete shield selection using the analytic hierarchy process

    International Nuclear Information System (INIS)

    Abulfaraj, W.H.

    1994-01-01

    Special types of concrete radiation shields that depend on locally available materials and have improved properties for both neutron and gamma-ray attenuation were developed by using plastic materials and heavy ores. The analytic hierarchy process (AHP) is implemented to evaluate these types for selecting the best biological radiation shield for nuclear reactors. Factors affecting the selection decision are degree of protection against neutrons, degree of protection against gamma rays, suitability of the concrete as building material, and economic considerations. The seven concrete alternatives are barite-polyethylene concrete, barite-polyvinyl chloride (PVC) concrete, barite-portland cement concrete, pyrite-polyethylene concrete, pyrite-PVC concrete, pyrite-portland cement concrete, and ordinary concrete. The AHP analysis shows the superiority of pyrite-polyethylene concrete over the others

  5. Shields for nuclear reactors

    International Nuclear Information System (INIS)

    Aspden, G.J.

    1984-01-01

    The patent concerns shields for nuclear reactors. The roof shield comprises a normally fixed radial outer portion, a radial inner portion rotatable about a vertical axis, and a connection between the inner and outer portions. In the event of hypothecal core disruption conditions, a cantilever system on the inner wall allows the upward movement of the inner wall, in order to prevent loss of containment. (UK)

  6. 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.)

  7. Reactor head shielding apparatus

    International Nuclear Information System (INIS)

    Schukei, G.E.; Roebelen, G.J.

    1992-01-01

    This patent describes a nuclear reactor head shielding apparatus for mounting on spaced reactor head lifting members radially inwardly of the head bolts. It comprises a frame of sections for mounting on the lifting members and extending around the top central area of the head, mounting means for so mounting the frame sections, including downwardly projecting members on the frame sections and complementary upwardly open recessed members for fastening to the lifting members for receiving the downwardly projecting members when the frame sections are lowered thereto with lead shielding supported thereby on means for hanging lead shielding on the frame to minimize radiation exposure or personnel working with the head bolts or in the vicinity thereof

  8. Mechanical properties of JPDR biological shield concrete

    International Nuclear Information System (INIS)

    Idei, Yoshio; Kamata, Hiroshi; Akutsu, Youichi; Onizawa, Kunio; Nakajima, Nobuya; Sukegawa, Takenori; Kakizaki, Masayoshi.

    1990-11-01

    Plant life of nuclear power plant will be determined by the aging degradation of main components and structures because of the difficulty and the cost of the replacement. These components are the reactor pressure vessel, concrete structures and cables. Authors have performed the investigation of JPDR biological shield which was the succeeded in first generating electricity in Japan and is now being decommissioned in JAERI. The test core samples were bored from the shield concrete and tested to obtain the mechanical properties. Test results are summarized as below, (1) Peak value of fast neutron dose was estimated as 1 x 10 18 n/cm 2 which is equivalent to the dose at the end of life for commercial power reactor. (2) Averaged compressive strength of all specimens had been increased about 20 % compared with initial design strength. (3) It was identified that the compressive strength had a little trend to increase with the increase of neutron dose within the dose range obtained in this study. (4) Tensile strength, Elastic modulus and Poisson's ratio showed little effect of neutron dose. (5) It was suggested that the inside and the mid-section liners were effective to keep the water in concrete and to avoid the reduction in strength. (author)

  9. Core test reactor shield cooling system analysis

    International Nuclear Information System (INIS)

    Larson, E.M.; Elliott, R.D.

    1971-01-01

    System requirements for cooling the shield within the vacuum vessel for the core test reactor are analyzed. The total heat to be removed by the coolant system is less than 22,700 Btu/hr, with an additional 4600 Btu/hr to be removed by the 2-inch thick steel plate below the shield. The maximum temperature of the concrete in the shield can be kept below 200 0 F if the shield plug walls are kept below 160 0 F. The walls of the two ''donut'' shaped shield segments, which are cooled by the water from the shield and vessel cooling system, should operate below 95 0 F. The walls of the center plug, which are cooled with nitrogen, should operate below 100 0 F. (U.S.)

  10. 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

  11. Activation of the concrete in the bio shield of ITER

    International Nuclear Information System (INIS)

    Kalcheva, S.

    2005-02-01

    Calculations of neutron spectra in different parts of the tokamak building of ITER are performed. A computational geometry model of the tokamak building is prepared using MCNP-4C. The model includes adequate material composition and geometry description of the main parts of the tokamak for PPCS plant model A: toroidal field coils, vacuum vessel, shield, blanket structure, first wall, divertor, 14.1 MeV neutron source. The design and the dimensions of the bio shield are taken from the current ITER design. MCNP calculations of the neutron spectra in the bio shield (concrete) of ITER are performed, using the neutron spectra in TF coils calculated at UKAEA as external neutron source. The neutron spectra in the concrete calculated by MCNP are used as input data in the code EASY99 for estimations of the activation of the concrete in the bio shield around the tokamak. The time evolutions of the maximum (in the bio shield floor) and minimum (in the bio shield side walls) specific activity (Bq/kg) and dose rate (Sv/h.) of the main dominant nuclides in the concrete are evaluated and compared for 3 different concrete types, used as biological shield in the PWR and BR3 reactors. (author)

  12. The optimum shielding for a power reactor using local components

    International Nuclear Information System (INIS)

    AlHajali, S.; Kharita, M. H.; Yousef, S.; Naoom, B.; Al-Nassar, M.

    2009-07-01

    Some local concrete mixtures have been picked out (selected) to be studied as shielding concrete for prospective nuclear power reactor in Syria. This research has interested in the attenuation of gamma radiation and neutron fluxes by these local concretes in the ordinary conditions. In addition to the heat effect on the shielding and physical properties of local concrete. Furthermore the neutron activation of the elements of the local concrete mixtures have been studied that for selection the low-activation materials (low dose rate and short half life radioisotopes). In this way biological shielding for nuclear reactor can be safe during operation of nuclear power reactor, in addition to be low radioactive waste after decommissioning the reactor. (author)

  13. Radiation shielding for fission reactors

    Energy Technology Data Exchange (ETDEWEB)

    Oka, Yoshiaki [Tokyo Univ., Nuclear Engineering Research Laboratory, Tokyo (Japan)

    2000-03-01

    Radiation shielding aspects relating fission reactors have been reviewed. Domestic activities in the past five years have been mainly described concerning nuclear data, calculation methods, shielding and skyshine experiments, Advanced Boiling Water Reactor (ABWR), Advanced Pressurized Water Reactor (APWR), High Temperature Engineering Test Reactor (HTTR), Experimental and Prototype Fast Reactors (JOYO, MONJU), Demonstration FBR, core shroud replacement of BWR, and spent fuel transportation cask and vessel. These studies have valuable information in safety and cost reduction issues of fission reactor design for not only existing reactors but also new reactor concepts in the next century. It has been concluded that we should maintain existing shielding technologies and improve these data and methods for coming generations in the next millennium. (author)

  14. Nuclear reactor cavity streaming shield

    International Nuclear Information System (INIS)

    Klotz, R.J.; Stephen, D.W.

    1978-01-01

    The upper portion of a nuclear reactor vessel supported in a concrete reactor cavity has a structure mounted below the top of the vessel between the outer vessel wall and the reactor cavity wall which contains hydrogenous material which will attenuate radiation streaming upward between vessel and the reactor cavity wall while preventing pressure buildup during a loss of coolant accident

  15. Modular reactor head shielding system

    International Nuclear Information System (INIS)

    Jacobson, E. B.

    1985-01-01

    An improved modular reactor head shielding system is provided that includes a frame which is removably assembled on a reactor head such that no structural or mechanical alteration of the head is required. The shielding system also includes hanging assemblies to mount flexible shielding pads on trolleys which can be moved along the frame. The assemblies allow individual pivoting movement of the pads. The pivoting movement along with the movement allowed by the trolleys provides ease of access to any point on the reactor head. The assemblies also facilitate safe and efficient mounting of the pads directly to and from storage containers such that workers have additional shielding throughout virtually the entire installation and removal process. The flexible shielding pads are designed to interleave with one another when assembled around the reactor head for substantially improved containment of radiation leakage

  16. Radiation shield for nuclear reactors

    International Nuclear Information System (INIS)

    Weissenfluh, J.A.

    1978-01-01

    A shield for use with nuclear reactor systems to attenuate radiation resulting from reactor operation is described. The shield comprises a container preferably of a thin, flexible or elastic material, which may be in the form of a bag, a mattress, a toroidal segment or toroid or the like filled with radiation attenuating liuid. Means are provided in the container for filling and draining the container in place. Due to its flexibility, the shield readily conforms to irregularities in surfaces with which it may be in contact in a shielding position

  17. Radiation shielding for fusion reactors

    International Nuclear Information System (INIS)

    Santoro, R.T.

    2000-01-01

    Radiation shielding requirements for fusion reactors present different problems than those for fission reactors and accelerators. Fusion devices, particularly tokamak reactors, are complicated by geometry constraints that complicate disposition of fully effective shielding. This paper reviews some of these shielding issues and suggested solutions for optimizing the machine and biological shielding. Radiation transport calculations are essential for predicting and confirming the nuclear performance of the reactor and, as such, must be an essential part of the reactor design process. Development and optimization of reactor components from the first wall and primary shielding to the penetrations and containment shielding must be carried out in a sensible progression. Initial results from one-dimensional transport calculations are used for scoping studies and are followed by detailed two- and three-dimensional analyses to effectively characterize the overall radiation environment. These detail model calculations are essential for accounting for the radiation leakage through ports and other penetrations in the bulk shield. Careful analysis of component activation and radiation damage is cardinal for defining remote handling requirements, in-situ replacement of components, and personnel access at specific locations inside the reactor containment vessel. (author)

  18. Abrasive water jet cutting technique for biological shield concrete dismantlement

    International Nuclear Information System (INIS)

    Konno, T.; Narazaki, T.; Yokota, M.; Yoshida, H.; Miura, M.; Miyazaki, Y.

    1987-01-01

    The Japan Atomic Energy Research Institute (JAERI) is developing the abrasive-water jet cutting system to be applied to dismantling the biological shield walls of the JPDR as a part of the reactor dismantling technology development project. This is a total system for dismantling highly activated concrete. The concrete biological shield wall is cut into blocks by driving the abrasive-water jet nozzle, which is operated with a remote, automated control system. In this system, the concrete blocks are removed to a container, while the slurry and dust/mist which are generated during cutting are collected and treated, both automatically. It is a very practical method and will quite probably by used for actual dismantling of commercial power reactors in the future because it can minimize workers' exposure to radioactivity during dismantling, contributes to preventing diffusion of radiation, and reduces the volume of contaminated secondary waste

  19. Concrete for γ radiation shielding

    International Nuclear Information System (INIS)

    Azevedo e Souza, A.C. de; Rogers, John Douglas

    1980-01-01

    The attenuation characteristics of γ radiation in concrete slabs, considering their mechanical resistence and densities were determined. One heavy concrete which was used, was prepared using as additives iron ore and Fe 2 O 3 pellets in various grain sizes. Fortran programs were used for analysing data and determining the absorption coefficients and attenuation factors. (Author) [pt

  20. Concrete for. gamma. radiation shielding

    Energy Technology Data Exchange (ETDEWEB)

    de Azevedo e Souza, A.C. (Rio de Janeiro Univ. (Brazil). Inst. de Quimica); Rogers, J D [Rio de Janeiro Univ. (Brazil). Coordenacao dos Programas de Pos-graduacao de Engenharia

    1980-06-01

    The attenuation characteristics of ..gamma.. radiation in concrete slabs, considering their mechanical resistence and densities were determined. One heavy concrete which was used, was prepared using as additives iron ore and Fe/sub 2/ O/sub 3/ pellets in various grain sizes. Fortran programs were used for analysing data and determining the absorption coefficients and attenuation factors.

  1. Radiation shielding structure for concrete structure

    International Nuclear Information System (INIS)

    Oya, Hiroshi

    1998-01-01

    Crack inducing members for inducing cracks in a predetermined manner are buried in a concrete structure. Namely, a crack-inducing member comprises integrally a shielding plate and extended plates situated at the center of a wall and inducing plates vertically disposed to the boundary portion between them with the inducing plates being disposed each in a direction perforating the wall. There are disposed integrally a pair of the inducing plate spaced at a predetermined horizontal distance on both sides of the shielding plate so as to form a substantially crank-shaped cross section and extended plates formed in the extending direction of the shielding plate, and the inducing plates are disposed each in a direction perforating the wall. Then, cracks generated when stresses are exerted can be controlled, and generation of cracks passing through the concrete structure can be prevented reliably. The reliability of a radiation shielding effect can be enhanced remarkably. (N.H.)

  2. 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.

  3. Removal of concrete layers from biological shields by microwaves

    International Nuclear Information System (INIS)

    Wace, P.F.; Harker, A.H.; Hills, D.L.

    1990-01-01

    A comprehensive literature review has been carried out, to provide information for an experimental programme and equipment design. Mathematical modelling of the microwave and power fields in a concrete block, both steel reinforced and unreinforced, subjected to a microwave attack at two frequencies, has been carried out and estimates of the likely temperature rise with time obtained. A method of launching microwaves into concrete has been established from theoretical considerations and from the findings of the literature review. Equipment for laboratory trials has been designed and assembled using an 896 MHz, 25 kW microwave generator. Reinforced concrete blocks, 0.6 m in dimension and representing the concrete in a Magnox reactor biological shield, have been attacked at different power levels and the surface removed to the depth of the reinforcing steel (100 mm). Outline proposals for the design of a remotely operated prototype microwave machine for stripping the surface of large concrete test panels have been prepared. (author)

  4. CONCRETE REACTOR CONTAINMENT

    Energy Technology Data Exchange (ETDEWEB)

    Lumb, Ralph F.; Hall, William F.; Fruchtbaum, Jacob

    1963-06-15

    The results of various leak-rate tests demonstrate the practicality of concrete as primary containment for the maximum credible accident for a research reactor employing plate-type fuel and having a power in excess of one megawatt. Leak-test time was shortened substantially by measuring the relaxation time for overpressure decay, which is a function of leak rate. (auth)

  5. Estimation of temperature distribution in a reactor shield

    International Nuclear Information System (INIS)

    Agarwal, R.A.; Goverdhan, P.; Gupta, S.K.

    1989-01-01

    Shielding is provided in a nuclear reactor to absorb the radiations emanating from the core. The energy of these radiations appear in the form of heat. Concrete which is commonly used as a shielding material in nuclear power plants must be able to withstand the temperatures and temperature gradients appearing in the shield due to this heat. High temperatures lead to dehydration of the concrete and in turn reduce the shielding effectiveness of the material. Adequate cooling needs to be provided in these shields in order to limit the maximum temperature. This paper describes a method to estimate steady state and transient temperature distribution in reactor shields. The results due to loss of coolant in the coolant tubes have been studied and presented in the paper. (author). 5 figs

  6. Slipforming of reinforced concrete shield building

    International Nuclear Information System (INIS)

    Hsieh, M.C.; King, J.R.

    1982-01-01

    The unique design and construction features of slipforming the heavily reinforced concrete cylindrical shield walls at the Satsop nuclear plant in Washington, D.C. site are presented. The shield walls were designed in compliance with seismic requirements which resulted in the need for reinforcing steel averaging 326 kg/m/sup 3/. A 7.6 m high, three-deck moving platform was designed to permit easy installation of the reinforcing steel, embedments, and blockouts, and to facilitate concrete placement and finishing. Two circular box trusses, one on each side of the shield wall, were used in combination with a spider truss to meet both the tolerance and strength requirements for the slipform assembly

  7. Radiation shielding performance of some concrete

    International Nuclear Information System (INIS)

    Akkurt, I.; Akyildirim, H.; Mavi, B.; Kilincarslan, S.; Basyigit, C.

    2007-01-01

    The energy consumption is increasing with the increased population of the world and thus new energy sources were discovered such as nuclear energy. Besides using nuclear energy, nuclear techniques are being used in a variety of fields such as medical hospital, industry, agriculture or military issue, the radiation protection becomes one of the important research fields. In radiation protection, the main rules are time, distance and shielding. The most effective radiation shields are materials which have a high density and high atomic number such as lead, tungsten which are expensive. Alternatively the concrete which produced using different aggregate can be used. The effectiveness of radiation shielding is frequently described in terms of the half value layer (HVL) or the tenth value layer (TVL). These are the thicknesses of an absorber that will reduce the radiation to half, and one tenth of its intensity respectively. In this study the radiation protection properties of different types of concrete will be discussed

  8. 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.

  9. Radiation shield for nuclear reactors

    International Nuclear Information System (INIS)

    Weissenfluh, J.A.

    1980-01-01

    A reusable radiation shield for use in a reactor installation comprises a thin-walled, flexible and resilient container, made of plastic or elastomeric material, containing a hydrogenous fluid with boron compounds in solution. The container can be filled and drained in position and the fluid can be recirculated if required. When not in use the container can be folded and stored in a small space. The invention relates to a shield to span the top of the annular space between a reactor vessel and the primary shield. For this purpose a continuous toroidal container or a series of discrete segments is used. Other forms can be employed for different purposes, e.g. mattress- or blanket-like forms can be draped over potential sources of radiation or suspended from a mobile carrier and placed between a worker and a radiation source. (author)

  10. Shield design of concrete wall between decay tank room and primary pump room in TRIGA facility

    International Nuclear Information System (INIS)

    Khan, M. J. H.; Rahman, M.; Haque, A.; Zulquarnain, A.; Ahmed, F. U.; Bhuiyan, S. I.

    2007-01-01

    The objective of this study is to recommend the radiation protection design parameters from the shielding point of view for concrete wall between the decay tank room and the primary pump room in TRIGA Mark-II research reactor facility. The shield design for this concrete wall has been performed with the help of Point-kernel Shielding Code Micro-Shield 5.05 and this design was also validated based on the measured dose rate values with Radiation Survey Meter (G-M Counter) considering the ICRP-60 (1990) recommendations for occupational dose rate limit (10 μSv/hr). The recommended shield design parameters are: (i) thickness of 114.3 cm Ilmenite-Magnetite Concrete (IMC) or 129.54 cm Ordinary Reinforced Concrete (ORC) for concrete wall A (ii) thickness of 66.04 cm Ilmenite-Magnetite Concrete (IMC) or 78.74 cm Ordinary Reinforced Concrete (ORC) for concrete wall B and (iii) door thickness of 3.175 cm Mild Steel (MS) on the entrance of decay tank room. In shielding efficiency analysis, the use of I-M concrete in the design of this concrete wall shows that it reduced the dose rate by a factor of at least 3.52 times approximately compared to ordinary reinforced concrete

  11. Study and application of construction technology of shielding concrete

    International Nuclear Information System (INIS)

    Wu Chongming; Ding Dexin; Chen Liangzhu; Zhao Jingfa; Li Shilong

    2008-01-01

    Process and techniques such as mixing,transportation and pouring have been studied. The construction technology for the shielding concrete with different densities has been summarized. The technology for the common concrete is quite different from that of shielding concrete, especially when its density is more than 4000 kg/m3. Application and practices have shown that different construction technologies shall be used for shielding concretes with different densities, and thus to ensure its uniformity and construction quality. (authors)

  12. Reactor vessel head permanent shield

    International Nuclear Information System (INIS)

    Hankinson, M.F.; Leduc, R.J.; Richard, J.W.; Malandra, L.J.

    1989-01-01

    A nuclear reactor is described comprising: a nuclear reactor pressure vessel closure head; control rod drive mechanisms (CRDMs) disposed within the closure head so as to project vertically above the closure head; cooling air baffle means surrounding the control rod drive mechanisms for defining cooling air paths relative to the control rod drive mechanisms; means defined within the periphery of the closure head for accommodating fastening means for securing the closure head to its associated pressure vessel; lifting lugs fixedly secured to the closure head for facilitating lifting and lowering movements of the closure head relative to the pressure vessel; lift rods respectively operatively associated with the plurality of lifting lugs for transmitting load forces, developed during the lifting and lowering movements of the closure head, to the lifting lugs; upstanding radiation shield means interposed between the cooling air baffle means and the periphery of the enclosure head of shielding maintenance personnel operatively working upon the closure head fastening means from the effects of radiation which may emanate from the control rod drive mechanisms and the cooling air baffle means; and connecting systems respectively associated with each one of the lifting lugs and each one of the lifting rods for connecting each one of the lifting rods to a respective one of each one of the lifting lugs, and for simultaneously connecting a lower end portion of the upstanding radiation shield means to each one of the respective lifting lugs

  13. 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)

  14. Research reactor decommissioning experience - concrete removal and disposal -

    International Nuclear Information System (INIS)

    Manning, Mark R.; Gardner, Frederick W.

    1990-01-01

    Removal and disposal of neutron activated concrete from biological shields is the most significant operational task associated with research reactor decommissioning. During the period of 1985 thru 1989 Chem-Nuclear Systems, Inc. was the prime contractor for complete dismantlement and decommissioning of the Northrop TRIGA Mark F, the Virginia Tech Argonaut, and the Michigan State University TRIGA Mark I Reactor Facilities. This paper discusses operational requirements, methods employed, and results of the concrete removal, packaging, transport and disposal operations for these (3) research reactor decommissioning projects. Methods employed for each are compared. Disposal of concrete above and below regulatory release limits for unrestricted use are discussed. This study concludes that activated reactor biological shield concrete can be safely removed and buried under current regulations

  15. Prestressed concrete pressure vessels for nuclear reactors - 1973

    Energy Technology Data Exchange (ETDEWEB)

    1977-01-01

    This standard deals with the design, construction, inspection and testing of prestressed concrete pressure vessels for nuclear reactors. Such pressure vessels serve the dual purpose of shielding and containing gas cooled nuclear reactors and are a form of civil engineering structure requiring particularly high integrity, and ensured leak tightness. (Metric)

  16. 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

  17. Radiation Damage In Reactor Cavity Concrete

    Energy Technology Data Exchange (ETDEWEB)

    Field, Kevin G [ORNL; Le Pape, Yann [ORNL; Naus, Dan J [ORNL; Remec, Igor [ORNL; Busby, Jeremy T [ORNL; Rosseel, Thomas M [ORNL; Wall, Dr. James Joseph [Electric Power Research Institute (EPRI)

    2015-01-01

    License renewal up to 60 years and the possibility of subsequent license renewal to 80 years has established a renewed focus on long-term aging of nuclear generating stations materials, and recently, on concrete. Large irreplaceable sections of most nuclear generating stations include concrete. The Expanded Materials Degradation Analysis (EMDA), jointly performed by the Department of Energy, the Nuclear Regulatory Commission and Industry, identified the urgent need to develop a consistent knowledge base on irradiation effects in concrete. Much of the historical mechanical performance data of irradiated concrete does not accurately reflect typical radiation conditions in NPPs or conditions out to 60 or 80 years of radiation exposure. To address these potential gaps in the knowledge base, The Electric Power Research Institute and Oak Ridge National Laboratory are working to disposition radiation damage as a degradation mechanism. This paper outlines the research program within this pathway including: (i) defining the upper bound of the neutron and gamma dose levels expected in the biological shield concrete for extended operation (80 years of operation and beyond), (ii) determining the effects of neutron and gamma irradiation as well as extended time at temperature on concrete, (iii) evaluating opportunities to irradiate prototypical concrete under accelerated neutron and gamma dose levels to establish a conservative bound and share data obtained from different flux, temperature, and fluence levels, (iv) evaluating opportunities to harvest and test irradiated concrete from international NPPs, (v) developing cooperative test programs to improve confidence in the results from the various concretes and research reactors, (vi) furthering the understanding of the effects of radiation on concrete (see companion paper) and (vii) establishing an international collaborative research and information exchange effort to leverage capabilities and knowledge.

  18. Characterization of Radiation Fields for Assessing Concrete Degradation in Biological Shields of NPPs

    Science.gov (United States)

    Remec, Igor; Rosseel, Thomas M.; Field, Kevin G.; Pape, Yann Le

    2017-09-01

    Life extensions of nuclear power plants (NPPs) to 60 years of operation and the possibility of subsequent license renewal to 80 years have renewed interest in long-term material degradation in NPPs. Large irreplaceable sections of most nuclear generating stations are constructed from concrete, including safety-related structures such as biological shields and containment buildings; therefore, concrete degradation is being considered with particular focus on radiation-induced effects. Based on the projected neutron fluence values (E > 0.1 MeV) in the concrete biological shields of the US pressurized water reactor fleet and the currently available data on radiation effects on concrete, some decrease in mechanical properties of concrete cannot be ruled out during extended operation beyond 60 years. An expansion of the irradiated concrete database is desirable to ensure reliable risk assessment for extended operation of nuclear power plants.

  19. Investigation and assessment of lead slag concrete as nuclear shields

    International Nuclear Information System (INIS)

    Zaghloul, Y.R.

    2009-01-01

    The present work is concerned with the efficiency of heavy weight concrete as a shielding material in constructing nuclear installations as well as for radioactive wastes disposal facilities.In this context, lead slag was used as a replacement for fine aggregates in heavy concrete shields that include local heavy weight aggregates (namely; barite and ilmenite) as well as normal concrete includes dolomite and sand as coarse and fine aggregates, as a reference. The effect of different percentages of lead slag was investigated to assess the produced lead slag concrete as a nuclear shielding material. The different properties (physical, mechanical and nuclear) of the produced lead slag concrete were investigated. The results obtained showed that increasing the lead slag percentage improving the investigated properties of the different concrete mixes. In addition, ilmenite concrete with 20% lead slag showed the best results for all the investigated properties.

  20. The AA disappearing under concrete shielding

    CERN Multimedia

    CERN PhotoLab

    1982-01-01

    When the AA started up in July 1980, the machine stood freely in its hall, providing visitors with a view through the large window in the AA Control Room. The target area, in which the high-intensity 26 GeV/c proton beam from the PS hit the production target, was heavily shielded, not only towards the outside but also towards the AA-Hall. However, electrons and pions emanating from the target with the same momentum as the antiprotons, but much more numerous, accompanied these through the injection line into the AA ring. The pions decayed with a half-time corresponding to approximately a revolution period (540 ns), whereas the electrons lost energy through synchrotron radiation and ended up on the vacuum chamber wall. Electrons and pions produced the dominant component of the radiation level in the hall and the control room. With operation times far exceeding original expectations, the AA had to be buried under concrete shielding in order to reduce the radiation level by an order of magnitude.

  1. Determination of shielding parameters for different types of concretes by Monte Carlo methods

    International Nuclear Information System (INIS)

    Aminian, A.; Nematollahi, M. R.

    2007-01-01

    The chose of a suitable concrete composition for a biological reactor shield remain as a research target up to now. In the present study the attempts has been made to estimate the influence of the concrete aggregates on the shielding parameters for three type of ordinary, serpentine and steel magnetite concrete by Monte Carlo N-Particle (MCNP ) transport code. MCNP calculations have been performed in order to obtain the leakage of neutrons, photons and electrons from dry shield. Also the mass attenuation coefficients and the liner attenuation coefficient are estimated for neutron and photon in those energies in range of actual energy which exist out of pressure vessel of power reactor in the cavity for the investigated concretes. The concrete densities ranged from 2.3 to 5.11 g/cm 3 . These calculations were done in the condition of a typical commercial Pressurized Water Reactor (PWR). The results show that Steel-magnetite concrete, with high density (5.11 g/cm 3 ) and constituents of relatively high atomic number, is an effective shield for both photons and neutrons

  2. 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

  3. Heavy density concrete for nuclear radiation shielding and power stations: [Part]2

    International Nuclear Information System (INIS)

    Singha Roy, P.K.

    1987-01-01

    This article is the second part of the paper entitled 'Heavy density concrete for nuclear radiation shielding and power stations'. In this part, some of the important properties of heavy density concrete are discussed. They include density, water retentivity, air content, permeability with special reference to concrete mixes used in India's nuclear power reactors. All these properties are affected to various extents by heating. Indian shield concrete is rarely subjected to temperatures above 60degC during its life, because of thermal shield protection. During placement, the maximum anticipated rise in temperature due to heat of hydration is restricted to around 45degC by chilling, if necessary to reduce shrinkage stresses and cracks. (M.G.B.)

  4. Modelling the electrical properties of concrete for shielding effectiveness prediction

    International Nuclear Information System (INIS)

    Sandrolini, L; Reggiani, U; Ogunsola, A

    2007-01-01

    Concrete is a porous, heterogeneous material whose abundant use in numerous applications demands a detailed understanding of its electrical properties. Besides experimental measurements, material theoretical models can be useful to investigate its behaviour with respect to frequency, moisture content or other factors. These models can be used in electromagnetic compatibility (EMC) to predict the shielding effectiveness of a concrete structure against external electromagnetic waves. This paper presents the development of a dispersive material model for concrete out of experimental measurement data to take account of the frequency dependence of concrete's electrical properties. The model is implemented into a numerical simulator and compared with the classical transmission-line approach in shielding effectiveness calculations of simple concrete walls of different moisture content. The comparative results show good agreement in all cases; a possible relation between shielding effectiveness and the electrical properties of concrete and the limits of the proposed model are discussed

  5. Fractionation of plutonium in environmental and bio-shielding concrete samples using dynamic sequential extraction

    DEFF Research Database (Denmark)

    Qiao, Jixin; Hou, Xiaolin

    2010-01-01

    Fractionation of plutonium isotopes (238Pu, 239,240Pu) in environmental samples (i.e. soil and sediment) and bio-shielding concrete from decommissioning of nuclear reactor were carried out by dynamic sequential extraction using an on-line sequential injection (SI) system combined with a specially...

  6. Gamma radiation shielding analysis of lead-flyash concretes

    International Nuclear Information System (INIS)

    Singh, Kanwaldeep; Singh, Sukhpal; Dhaliwal, A.S.; Singh, Gurmel

    2015-01-01

    Six samples of lead-flyash concrete were prepared with lead as an admixture and by varying flyash content – 0%, 20%, 30%, 40%, 50% and 60% (by weight) by replacing cement and keeping constant w/c ratio. Different gamma radiation interaction parameters used for radiation shielding design were computed theoretically and measured experimentally at 662 keV, 1173 keV and 1332 keV gamma radiation energy using narrow transmission geometry. The obtained results were compared with ordinary-flyash concretes. The radiation exposure rate of gamma radiation sources used was determined with and without lead-flyash concretes. - Highlights: • Concrete samples with lead as admixture were casted with flyash replacing 0%, 20%, 30%, 40%, 50% and 60% of cement content (by weight). • Gamma radiation shielding parameters of concretes for different gamma ray sources were measured. • The attenuation results of lead-flyash concretes were compared with the results of ordinary flyash concretes

  7. Technical specifications for the bulk shielding reactor

    International Nuclear Information System (INIS)

    1986-05-01

    This report provides information concerning the technical specifications for the Bulk Shielding Reactor. Areas covered include: safety limits and limiting safety settings; limiting conditions for operation; surveillance requirements; design features; administrative controls; and monitoring of airborne effluents. 10 refs

  8. Shielded transport containers for reactor waste

    International Nuclear Information System (INIS)

    Grundfelt, B.; Eriksson, E.

    The report presents that part of risk analysis which deals with the frequency of breakdowns and the damage on containers. The report focusses on shielded containers made of reinforced concrete. Also a container made of steel is referred to the cases of breakdown are closely allied to collisions with ships. The frequency of breakdowns which might damage the containers is low in all respects, namely 1.10 -5 per year or lower for the shielded container. (G.B.)

  9. Shielding design to obtain compact marine reactor

    International Nuclear Information System (INIS)

    Yamaji, Akio; Sako, Kiyoshi

    1994-01-01

    The marine reactors equipped in previously constructed nuclear ships are in need of the secondary shield which is installed outside the containment vessel. Most of the weight and volume of the reactor plants are occupied by this secondary shield. An advanced marine reactor called MRX (Marine Reactor X) has been designed to obtain a more compact and lightweight marine reactor with enhanced safety. The MRX is a new type of marine reactor which is an integral PWR (The steam generator is installed in the pressure vessel.) with adopting a water-filled containment vessel and a new shielding design method of no installation of the secondary shield. As a result, MRX is considerably lighter in weight and more compact in size as compared with the reactors equipped in previously constructed nuclear ships. For instance, the plant weight and volume of the containment vessel of MRX are about 50% and 70% of those of the Nuclear Ship MUTSU, in spite of the power of MRX is 2.8 times as large as the MUTSU's reactor. The shielding design calculation was made using the ANISN, DOT3.5, QAD-CGGP2 and ORIGEN codes. The computational accuracy was confirmed by experimental analyses. (author)

  10. Using natural local materials for developing special radiation shielding concretes, and deduction of its shielding characteristics

    International Nuclear Information System (INIS)

    Kharita, M. H.; Takeyeddin, M.; Al-Nassar, M.; Yousef, S.

    2006-06-01

    Concrete is considered as the most important material to be used for radiation shielding in facilities contain radioactive sources and radiation generating machines. The concrete shielding properties may vary depending on the construction of the concrete, which is highly relative to the composing aggregates i.e. aggregates consist about 70 - 80% of the total weight of normal concrete. In this project tow types of concrete used in Syria (in Damascus and Aleppo) had been studied and their shielding properties were defined for gamma ray from Cs-137 and Co-60 sources, and for neutrons from Am-Be source. About 10% reduction in HVL was found in the comparison between the tow concrete types for both neutrons and gammas. Some other types of concrete were studied using aggregates from different regions in Syria, to improve the shielding properties of concrete, and another 10% of reduction was achieved in comparison with Damascene concrete (20% in comparison with the concrete from Aleppo) for both neutrons and gamma rays. (author)

  11. Study of local Agregate for Gamma radiation concrete shield

    International Nuclear Information System (INIS)

    Tochrul-Binowo; Endro-Kismolo; Darsono

    1996-01-01

    Investigation on the composition of gamma radiation concrete shield made of local barite, manganese fine and coarse aggregates from Kulon Progo, Yogyakarta has been done. The purpose of the research was to find out the quality of these local material for an aggregate of gamma radiation concrete shield. The research was done where each mineral was used as coarse aggregate and the fine aggregate from Kulon Progo was used as fine basic aggregate. Firstly a normal concrete was made by mixing cement, fine aggregate, coarse aggregate and water at a weight ratio of cement: fine aggregate: coarse: water 1: 2.304: 3.456: 0.58. The gamma radiation absorption capacity of the concrete tested by using Cs-137 as source standard. The same method was done on barite concrete at the weight ratio of cement: fine aggregate: barite aggregate: water 1: 2.303: 3.456: 0.58 and manganese concrete at the weight ratio of cement: fine aggregate: manganese aggregate: and water 1: 1.896: 2.844: 0.58. The result of the study showed that the gamma radiation absorption capacity of barite aggregate was greater than that of normal concrete and manganese concrete. The coefficient linear attenuation (for 6.0 cm thickness) of each concrete were μ barite concrete = 0.23071 cm -1 , μ manganese concrete = 0.08401 cm -1 and μ normal concrete = 0.1669 cm -1

  12. Investigation of water content in primary upper shield of high temperature engineering test reactor (HTTR)

    International Nuclear Information System (INIS)

    Sumita, Junya; Sawa, Kazuhiro; Mogi, Haruyoshi; Itahashi, Shuuji; Kitami, Toshiyuki; Akutu, Youichi; Fuchita, Yasuhiro; Kawaguchi, Toru; Moriya, Masahiro

    1999-09-01

    A primary upper shield of the High Temperature Engineering Test Reactor (HTTR) is composed of concrete (grout) which is packed into iron frames. The main function of the primary upper shield is to attenuate neutron and gamma ray from the core, that leads to satisfy dose equivalent rate limit of operating floor and stand-pipe room. Water content in the concrete is one of the most important things because it strongly affects neutron-shielding ability. Then, we carried out out-of-pile experiments to investigate relationship between temperature and water content in the concrete. Based on the experimental results, a hydrolysis-diffusion model was developed to investigate water release behavior from the concrete. The model showed that water content used for shielding design in the primary upper shield of the HTTR will be maintained if temperature during operating life is under 110degC. (author)

  13. Investigating Radiation Shielding Properties of Different Mineral Origin Heavyweight Concretes

    Science.gov (United States)

    Basyigit, Celalettin; Uysal, Volkan; Kilinçarslan, Şemsettin; Mavi, Betül; Günoǧlu, Kadir; Akkurt, Iskender; Akkaş, Ayşe

    2011-12-01

    The radiation although has hazardous effects for human health, developing technologies bring lots of usage fields to radiation like in medicine and nuclear power station buildings. In this case protecting from undesirable radiation is a necessity for human health. Heavyweight concrete is one of the most important materials used in where radiation should be shielded, like those areas. In this study, used heavyweight aggregates of different mineral origin (Limonite, Siderite), in order to prepare different series in concrete mixtures and investigated radiation shielding properties. The experimental results on measuring the radiation shielding, the heavyweight concrete prepared with heavyweight aggregates of different mineral origin show that, are useful radiation absorbents when they used in concrete mixtures.

  14. Investigating Radiation Shielding Properties of Different Mineral Origin Heavyweight Concretes

    International Nuclear Information System (INIS)

    Basyigit, Celalettin; Uysal, Volkan; Kilincarslan, Semsettin; Akkas, Ayse; Mavi, Betuel; Guenoglu, Kadir; Akkurt, Iskender

    2011-01-01

    The radiation although has hazardous effects for human health, developing technologies bring lots of usage fields to radiation like in medicine and nuclear power station buildings. In this case protecting from undesirable radiation is a necessity for human health. Heavyweight concrete is one of the most important materials used in where radiation should be shielded, like those areas. In this study, used heavyweight aggregates of different mineral origin (Limonite, Siderite), in order to prepare different series in concrete mixtures and investigated radiation shielding properties. The experimental results on measuring the radiation shielding, the heavyweight concrete prepared with heavyweight aggregates of different mineral origin show that, are useful radiation absorbents when they used in concrete mixtures.

  15. Development of special radiation shielding concretes using natural local materials and evaluation of their shielding characteristics

    International Nuclear Information System (INIS)

    Kharita, M. H.; Takeyeddin, M.; Al-Nassri, M.; Yousef, S.

    2008-01-01

    Concrete is one of the most important materials used for radiation shielding in facilities containing radioactive sources and radiation generating machines. The concrete shielding properties may vary depending on the composite of the concrete. Aggregates is the largest constituent (about 70-80% of the total weight of normal concrete). The aim of this work is to develop special concrete with good shielding properties for gamma and neutrons, using natural local materials. For this reason two types of typical concrete widely used in Syria (in Damascus and Aleppo) and four other types of concrete, using aggregates from different regions, have been prepared. The shielding properties of these six types were studied for gamma ray (from Cs-137 and Co-60 sources)and for neutrons (from am-Be source). A reduction of about 10% in the HVL was obtained for the concrete from Damascus in comparison with that from Aleppo, for both neutrons and gammas. One of the other four types of concrete (from Rajo site, mostly Hematite), was found to further reduce the HVL by about 10% for both neutrons and gamma rays.(author)

  16. 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.

  17. 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.

  18. MEANS FOR SHIELDING AND COOLING REACTORS

    Science.gov (United States)

    Wigner, E.P.; Ohlinger, L.A.; Young, G.J.; Weinberg, A.M.

    1959-02-10

    Reactors of the water-cooled type and a means for shielding such a rcactor to protect operating personnel from harmful radiation are discussed. In this reactor coolant tubes which contain the fissionable material extend vertically through a mass of moderator. Liquid coolant enters through the bottom of the coolant tubes and passes upwardly over the fissionable material. A shield tank is disposed over the top of the reactor and communicates through its bottom with the upper end of the coolant tubes. A hydrocarbon shielding fluid floats on the coolant within the shield tank. With this arrangements the upper face of the reactor can be opened to the atmosphere through the two superimposed liquid layers. A principal feature of the invention is that in the event radioactive fission products enter thc coolant stream. imposed layer of hydrocarbon reduces the intense radioactivity introduced into the layer over the reactors and permits removal of the offending fuel material by personnel shielded by the uncontaminated hydrocarbon layer.

  19. Build-up Factor Calculation for Ordinary Concrete, Baryte Concrete and Blast-furnace Slugges Concrete as γ Radiation Shielding

    International Nuclear Information System (INIS)

    Isman MT; Elisabeth Supriatni; Tochrul Binowo

    2002-01-01

    Calculation of build up factor ordinary concrete, baryte concrete and blast-furnace sludge concrete have been carried out. The calculations have been carried out by dose rate measurement of Cs 137 source before and after passing through shielding. The investigated variables were concrete type, thickness of concrete and relative possession of concrete. Concrete type variables are ordinary concrete, baryte concrete and blast sludge furnace concrete. The thickness variables were 6, 12, 18, 24, 30 and 36 cm. The relative position variables were dose to the source and close to detector. The result showed that concrete type and position did not have significant effect to build-up factor value, while the concrete thickness (r) and the attenuation coefficient (μ) were influenced to the build-up factor. The higher μr value the higher build-up factor value. (author)

  20. Shield structure for a nuclear reactor

    International Nuclear Information System (INIS)

    Rouse, C.A.; Simnad, M.T.

    1979-01-01

    An improved nuclear reactor shield structure is described for use where there are significant amounts of fast neutron flux above an energy level of approximately 70 keV. The shield includes structural supports and neutron moderator and absorber systems. A portion at least of the neutron moderator material is magnesium oxide either alone or in combination with other moderator materials such as graphite and iron. (U.K.)

  1. Investigation of the use of Galena concrete in electromagnetic radiation shielding

    International Nuclear Information System (INIS)

    Egwuonwu, G. N.; Bukar, P. H.; Avaa, A.

    2011-01-01

    Galena samples, collected from Ishiagu, south-eastern Nigeria, were used to make high density concretes for experimental radiation shielding. The concretes were molded into cylindrical tablets of various densities and volumes in order to ascertain their attenuation capability to some electromagnetic radiations. Blue visible light and gamma-ray sourced from cobalt-60, were transmitted through the concretes and detected with the aid of Op-Amp and digital Geiger-Muller Counter respectively. The absorption coefficients of the samples of thicknesses in the range of 1.00 - 5.00 cm were determined. Results show that for a typical galena concrete of average density 2.33gcm -3 , the absorption coefficient is about 1.186 cm -1 for the blue light and 0.495cm -1 for gamma-ray. For this density, 4.45cm of the galena concrete reduces the gamma-ray intensity by 90% and its half value layer thickness is 1.40cm. The investigation however, suggests the shielding properties of the galena sourced from Ishiagu. A database of shielding strength for the in situ galena was established hence, can serve as suitable platform for quality and quantity control in radiation shielding technology in radiotherapy treatment rooms and nuclear reactors.

  2. Concrete mix design for X-and gamma shielding

    International Nuclear Information System (INIS)

    Mohamad Pauzi Ismail; Noor Azreen Masenwat; Suhairy Sani; Abdul Bakhri Muhammad; Mohd Kamal Shah Shamsuddin; Rahmad Abd Rashid

    2012-01-01

    The design of X-ray or gamma ray radiographic exposure room requires some calculations on shielding to provide safe operation of the facility and minimum exposure to radiation workers. Careful design can lead to economical installations with minimal barriers. The design depends on such factors as: maximum energy, maximum intensity, permitted full-body dosage, workload, use factor, occupancy factor, maximum dose output and shielding materials. Choice of material for a barrier depends on convenience and cost. The radiographic exposure room is usually made of normal concrete with density of about 2.3 - 2.4 g/ cc. Normal concrete is often used for construction of exposure room because of cheap and ease of construction. This paper explained and discussed the optimum mix design for normal concrete used for X-and gamma shielding. (author)

  3. Heavy concrete shieldings made of recycled radio-active steel

    International Nuclear Information System (INIS)

    Holland, D.; Quade, U.; Sappok, M.; Heim, H.

    1998-01-01

    Maintenance and decommissioning of nuclear installations will generate increasing quantities of radioactively contaminated metallic residues. For many years, Siempelkamp has been melting low-level radioactive scrap in order to re-use it for containers of nuclear industry. Another new recycling path has recently been developed by producing steel granules from the melt. These granules are used as replacement for hematite (iron ore) in the production of heavy concrete shieldings. In the CARLA plant (central plant for the recycling of low-level radioactive waste) of Siempelkamp Nuklear- und Umwelttechnik GmbH and Co., the scrap is melted in a medium frequency induction furnace. The liquid iron is poured into a cooling basin through a water jet, which splits the iron into granules. The shape of these granules is determined by various factors, such as water jet speed, pouring rate of the liquid iron and different additives to the melt. In this process, massive spheres with diameters ranging from 1 to 8 mm can be produced which add to the density of heavy concrete elements for optimum shielding. In close cooperation with Boschert, which indeed is an expert for the production of concrete shieldings, a new technology for manufacturing heavy concrete shieldings, containing low-level radioactive steel granules, has been developed. The portion of steel granules in the concrete is approx. 50 weight-%. A concrete density between 2.4 kg/dm 3 and 4.0 kg/dm 3 is available. The compressive strength for the concrete reaches values up to 65 MPa. Different types of Granulate Shielding Casks (GSC) are offered by Siempelkamp. The most famous one is the GSC 200 for 200 1 drums, which has already been qualified for final storage of radioactive wastes at the German Morsleben final repository (ERAM). This newly developed recycling process further increases the quantities of low-level radioactive metallic wastes available for recycling. Expensive storage area can thus be saved respectively

  4. 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)

  5. Review of ORNL-TSF shielding experiments for the gas-cooled Fast Breeder Reactor Program

    International Nuclear Information System (INIS)

    Abbott, L.S.; Ingersoll, D.T.; Muckenthaler, F.J.; Slater, C.O.

    1982-01-01

    During the period between 1975 and 1980 a series of experiments was performed at the ORNL Tower Shielding Facility in support of the shield design for a 300-MW(e) Gas Cooled Fast Breeder Demonstration Plant. This report reviews the experiments and calculations, which included studies of: (1) neutron streaming in the helium coolant passageways in the GCFR core; (2) the effectiveness of the shield designed to protect the reactor grid plate from radiation damage; (3) the adequacy of the radial shield in protecting the PCRV (prestressed concrete reactor vessel) from radiation damage; (4) neutron streaming between abutting sections of the radial shield; and (5) the effectiveness of the exit shield in reducing the neutron fluxes in the upper plenum region of the reactor

  6. The effect of some organic and non-organic additions on the shielding and mechanical properties of radiation shielding concrete

    International Nuclear Information System (INIS)

    Kharita, M. H.; Yousef, S.; Al-Nassar, M.

    2011-04-01

    Few studies on the effect of some additives on the shielding properties of concrete have been carried out in this research. These studies included the effect of carbon powder, boron compounds, and waste polyethylene. The effect of water to cement ratio has been studied too. The research results showed that carbon powder and some boron compounds could be used to improve shielding concrete properties, and the possibility to add waste polyethylene in shielding concrete without effects on shielding properties. No significant effect for water to cement ratio on shielding properties of concrete. (author)

  7. 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

  8. Criticality study of PWR fuel elements separated by concrete shields

    International Nuclear Information System (INIS)

    Guillemot, M.; Colomb, G.; Maubert, L.

    1984-01-01

    The development of nuclear energy involved a need of neutronic shield studies to reduce the size and the cost of nuclear liquid storage units, of storage pools, and of transport casks. A concrete has been developed of the laboratory of concretes and coatings of Saclay; the validation of calculation codes including this material, as also the presence of free water and linked water necessitated an experimental confirmation of which program is presented here [fr

  9. A study on the effect of crack in concrete structure in the point of radiation shielding

    International Nuclear Information System (INIS)

    Lee, Chang-Min; Lee, Yoon-Hee; Lee, Kun-Jai; Cho, Cheon-Hyung; Choi, Byung-Il; Lee, Heung-Young

    2005-01-01

    The saturation of South Korea's at-reactor (AR) spent fuel storage pools has created a necessity for additional spent fuel storage capacity. Because the South Korean government has a plan to increase the number of nuclear power plants to 27 units by 2016, the increase of spent nuclear fuel generation will be accelerated. Because there is no concrete plan for spent unclear fuel permanent disposal, the Korea hydraulic nuclear power company is planning to construct dry storage facility. Spent nuclear fuel from CANDU type nuclear power plant will be stored in MACSTOR-400 composed by reinforced concrete. Because it is new model, it has to be licensed. Life time estimation is needed for licensing. Deterioration of reinforced concrete structure is currently of great concern for life time estimation. The most significant form of deterioration is reinforcement corrosion that gives rise to crack the concrete structure. In this study, in order to estimate the life time of MACSTOR, the tendency of crack creation, propagation and the effect of crack in concrete structure against radiation shielding are investigated. Crack creation and propagation depends on concrete cover thickness and c/d ratio. The surface dose rate at the concrete shield in MACSTOR is simulated by MCNP code about several cases. Generally in the case of point source, surface dose rate depends on shape, width and length of crack. In the case of MACSTOR-400, It is estimated that crack is not dominant factor in the point of radiation shielding in less than 0.4mm of crack width. Above results will be helpful to estimate the life time of concrete structure as radiation shield

  10. Analyses and testing of model prestressed concrete reactor vessels with built-in planes of weakness

    International Nuclear Information System (INIS)

    Dawson, P.; Paton, A.A.; Fleischer, C.C.

    1990-01-01

    This paper describes the design, construction, analyses and testing of two small scale, single cavity prestressed concrete reactor vessel models, one without planes of weakness and one with planes of weakness immediately behind the cavity liner. This work was carried out to extend a previous study which had suggested the likely feasibility of constructing regions of prestressed concrete reactor vessels and biological shields, which become activated, using easily removable blocks, separated by a suitable membrane. The paper describes the results obtained and concludes that the planes of weakness concept could offer a means of facilitating the dismantling of activated regions of prestressed concrete reactor vessels, biological shields and similar types of structure. (author)

  11. Technical specifications: Tower Shielding Reactor II

    International Nuclear Information System (INIS)

    1979-02-01

    The technical specifications define the key limitations that must be observed for safe operation of the Tower Shielding Reactor II (TSR-II) and an envelope of operation within which there is reasonable assurance that these limits cannot be exceeded. The specifications were written to satisfy the requirements of the Department of Energy (DOE) Manual Chapter 0540, September 1, 1972

  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. Construction of concrete hot cells; requirements for shielding windows for concrete walls with different densities

    International Nuclear Information System (INIS)

    1987-10-01

    The shielding windows form part of the basic equipment of hot cells for remote handling, as defined in standard DIN 25 420 part 1. The draft standard in hand is intended to specify the design and manufacture requirements, especially with regard to main dimensions, sight quality, shielding effects, and radiation resistance. The standard refers to three types of shielding window with surface area design (product of density and wall thickness) corresponding to concrete walls of the densities 2.4, 3.4, and 4.0 g/cm 3 . The windows fit to three types of concrete of common usage, and the design is made for Co-60 radiation, with attenuation factors of about 10 4 , 10 6 , or 10 7 . For concrete walls with densities between these data, a shielding window suitable to the next higher density data is to be chosen. (orig./HP) [de

  14. Preliminary shielding analysis of VHTR reactors

    International Nuclear Information System (INIS)

    Flaspoehler, Timothy M.; Petrovic, Bojan

    2011-01-01

    Over the last 20 years a number of methods have been established for automated variance reduction in Monte Carlo shielding simulations. Hybrid methods rely on deterministic adjoint and/or forward calculations to generate these parameters. In the present study, we use the FWCADIS method implemented in MAVRIC sequence of the SCALE6 package to perform preliminary shielding analyses of a VHTR reactor. MAVRIC has been successfully used by a number of researchers for a range of shielding applications, including modeling of LWRs, spent fuel storage, radiation field throughout a nuclear power plant, study of irradiation facilities, and others. However, experience in using MAVRIC for shielding studies of VHTRs is more limited. Thus, the objective of this work is to contribute toward validating MAVRIC for such applications, and identify areas for potential improvement. A simplified model of a prismatic VHTR has been devised, based on general features of the 600 MWt reactor considered as one of the NGNP options. Fuel elements have been homogenized, and the core region is represented as an annulus. However, the overall mix of materials and the relatively large dimensions of the spatial domain challenging the shielding simulations have been preserved. Simulations are performed to evaluate fast neutron fluence, dpa, and other parameters of interest at relevant positions. The paper will investigate and discuss both the effectiveness of the automated variance reduction, as well as applicability of physics model from the standpoint of specific VHTR features. (author)

  15. Beta induced Bremsstrahlung dose rate in concrete shielding

    International Nuclear Information System (INIS)

    Manjunatha, H.C.

    2013-01-01

    Dosimetric study of beta-induced Bremsstrahlung in concrete is importance in the field of radiation protection. The efficiency, intensity and dose rate of beta induced Bremsstrahlung by 113 pure beta nuclides in concrete shielding is computed. The Bremsstrahlung dosimetric parameters such as the efficiency (yield), Intensity and dose rate of Bremsstrahlung are low for 199 Au and high for 104 Tc in concrete. The efficiency, Intensity and dose rate of Bremsstrahlung increases with maximum energy of beta nuclide (Emax) and modified atomic number (Zmod) of the target. The estimated Bremsstrahlung efficiency, Intensity and dose rate are useful in the calculations photon track-length distributions. These parameters are useful to determine the quality and quantity of the radiation (known as the source term). Precise estimation of this source term is very important in planning of radiation shielding. (author)

  16. Electromagnetic Shielding Characteristics of Eco-Friendly Foamed Concrete Wall

    Directory of Open Access Journals (Sweden)

    Sung-Sil Cho

    2017-01-01

    Full Text Available The electromagnetic shielding characteristics according to the material composition of foamed concrete, which was manufactured to reduce environmental pollution and to economically apply it in actual building walls, were researched herein. Industrial by-products such as ladle furnace slag (LFS, gypsum, and blast furnace slag (BFS were added to manufacture foamed concrete with enhanced functionalities such as lightweight, heat insulation, and sound insulation. The electrical characteristics such as permittivity and loss tangent according to the foam and BFS content were calculated and measured. Free space measurement was used to measure the electromagnetic shielding characteristics of the actually manufactured foamed concrete. It was confirmed that electromagnetic signals were better blocked when the foam content was low and the BFS content was high in the measured frequency bands (1–8 GHz and that approximately 90% of the electromagnetic signals were blocked over 4 GHz.

  17. Attenuation of gamma radiation in concrete shields

    International Nuclear Information System (INIS)

    Azevedo e Souza, A.C. de.

    1978-12-01

    The attenuation characteristics of γ radiation in concrete layers considering their mechanical resistence and densities were determined. A 137 Cs source was used in a 'good geometry' arrangement to eliminate the effects of the buildup factor. The ordinary and the heavy concrete were irradiated and for the latter it was used as additives iron ore and Fe 2 O 3 pellets in various grain sizes. The detection system consisted of a 2' x 2' NaI (Tl) crystal coupled to a photomultiplier tube and the associated electronic equipment. FORTRAN programs were used for determining the absorption coefficients and the attenuation factors. These programs calculate photopeak areas eliminating all contributions due to Compton effect and background. (Author) [pt

  18. Important aspects of radiation shielding for fusion reactor tokamaks

    International Nuclear Information System (INIS)

    Abdou, M.A.

    1977-01-01

    Radiation shielding is a key subsystem in tokamak reactors. Design of this shield must evolve from economic and technological trade-off studies that account for the strong interrelations among the various components of the reactor system. These trade-offs are examined for the bulk shield on the inner side of the torus and for the special shields of major penetrations. Results derived are applicable for a large class of tokamak-type reactors

  19. Three-dimensional fabric reinforced concrete finds first use in reactor building

    International Nuclear Information System (INIS)

    Akihama, S.; Nakagava, H.

    1989-01-01

    It is reported about creation of concrete reinforced with synthetic fibers by Japanese firm Kadzima. Synthetic material with three-dimensional orientation of fibers is produced of roving impreganted with synthetic resin. The reinforcement produced is submerged into the concrete matrix. The compression strength of such a material makes up 58 MPa. The new material is used for constructing the nuclear reactor shielding containers

  20. ORNL fusion reactor shielding integral experiments

    International Nuclear Information System (INIS)

    Santoro, R.T.; Alsmiller, R.G. Jr.; Barnes, J.M.; Chapman, G.T.

    1980-01-01

    Integral experiments that measure the neutron and gamma-ray energy spectra resulting from the attenuation of approx. 14 MeV T(D,n) 4 He reaction neutrons in laminated slabs of stainless steel type 304, borated polyethylene, and a tungsten alloy (Hevimet) and from neutrons streaming through a 30-cm-diameter iron duct (L/D = 3) imbedded in a concrete shield have been performed. The facility, the NE-213 liquid scintillator detector system, and the experimental techniques used to obtain the measured data are described. The two-dimensional discrete ordinates radiation transport codes, calculational models, and nuclear data used in the analysis of the experiments are reviewed

  1. Nuclear data requirements for fusion reactor shielding

    International Nuclear Information System (INIS)

    Abdou, M.A.

    1979-01-01

    The nuclear data requirements for experimental, demonstration and commercial fusion reactors are reviewed. Particular emphasis is given to the shield as well as major reactor components of concern to the nuclear performance. The nuclear data requirements are defined as a result of analyzing four key areas. These are the most likely candidate materials, energy range, types of needed nuclear data, and the required accuracy in the data. Deducing the latter from the target goals for the accuracy in prediction is also discussed. A specific proposal of measurements is recommended. Priorities for acquisition of data are also assigned. (author)

  2. Determination of gamma ray shielding parameters of rocks and concrete

    Science.gov (United States)

    Obaid, Shamsan S.; Gaikwad, Dhammajyot K.; Pawar, Pravina P.

    2018-03-01

    Gamma shielding parameters such as mass attenuation coefficient (μ/ρ), effective atomic number (Zeff) and electron density (Neff) have been measured and calculated for rocks and concrete in the energy range 122-1330 keV. The measurements have been carried out at 122, 356, 511, 662, 1170, 1275, 1330 keV gamma ray energies using a gamma spectrometer includes a NaI(Tl) scintillation detector and MCA card. The atomic and electronic cross sections have also been investigated. Experimental and calculated (WinXCom) values were compared, and good agreement has been observed within the experimental error. The obtained results showed that feldspathic basalt, compact basalt, volcanic rock, dolerite and pink granite are more efficient than the sandstone and concrete for gamma ray shielding applications.

  3. 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)

  4. Shielding effects of concrete and foam external pipeline coatings

    International Nuclear Information System (INIS)

    Barlo, T.J.; Werner, D.P.

    1992-01-01

    The research project began in July, 1986 and was completed in December, 1990. The objectives of the research were: To determine whether concrete and urethane foam-barrier coatings shield the pipe from cathodic-protection current, To determine whether the barrier coatings also effectively shield the pipe from the environment, thus reducing the need for cathodic protection, To determine what levels of cathodic protection will be required to overcome shielding, and To establish what types of barrier coatings are most compatible with obtaining adequate levels of cathodic protection. To achieve these objectives, laboratory experiments were conducted with five barrier coating materials. These materials were (1) 2-lb/ft 3 , closed-cell urethane foam, (2) 3-lb/ft 3 , closed-cell urethane foam, (3) concrete barrier material, (4) glass fiber-reinforced concrete barrier material, and (5) sand. The barrier materials, whole and intentionally cracked, were applied to the bare, FBE-coated, and tape-coated steel specimens. The specimens were tested in aqueous electrolytes at room temperature and 140 degree F with no protection, protection to -0.95 V, and overprotection to -1.2 V (Cu/CuSO 4 )

  5. 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.)

  6. A Shielding Analysis of Hot Cell for a 10 MW Research Reactor

    International Nuclear Information System (INIS)

    Alnajjar, Alaaddin; Park, Chang Je; Roh, Gyuhong; Lee, Byunchul

    2013-01-01

    In this paper, a shielding analysis has been performed for the hot cell in a 10 MW research reactor. Two kinds of shielding analysis code systems are used such as MCNPX2.7 and M-Shield7. The first one is Monte Carlo stochastic code and the second one is a deterministic point kernel code. The results are compared in this study. In order to obtain source term, the ORIGEN-S code is used for different kinds of source. Four kinds of sources are taken into consideration. From the simulation, it is also proposed that the proper thickness of shielding material and the maximum source capacity in the hot cell. This study shows preliminary analysis results of hot cell shielding for 10MW research reactor. Total four different source terms are considered such as spent fuel assembly, Ir-192, Mo-99, and I-131. For shielding material, general concrete, heavy concrete, and lead are used. MCNPX code is mainly used for a simplified hot cell model and the result are nearly consistent when compared with M-Shield code. Required shielding thickness and the hot cell capacity are also obtained for various criterion of surface dose rates

  7. Methodology of shielding calculation for nuclear reactors

    International Nuclear Information System (INIS)

    Maiorino, J.R.; Mendonca, A.G.; Otto, A.C.; Yamaguchi, Mitsuo

    1982-01-01

    A methodology of calculation that coupling a serie of computer codes in a net that make the possibility to calculate the radiation, neutron and gamma transport, is described, for deep penetration problems, typical of nuclear reactor shielding. This net of calculation begining with the generation of constant multigroups, for neutrons and gamma, by the AMPX system, coupled to ENDF/B-IV data library, the transport calculation of these radiations by ANISN, DOT 3.5 and Morse computer codes, up to the calculation of absorbed doses and/or equivalents buy SPACETRAN code. As examples of the calculation method, results from benchmark n 0 6 of Shielding Benchmark Problems - ORNL - RSIC - 25, namely Neutron and Secondary Gamma Ray fluence transmitted through a Slab of Borated Polyethylene, are presented. (Author) [pt

  8. Photon spectrum behind biological shielding of the LVR-15 research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Klupak, V.; Viererbl, L.; Lahodova, Z.; Marek, M.; Vins, M. [Research Centre Rez Ltd., Husinec-Rez 130 (Czech Republic)

    2011-07-01

    The LVR-15 reactor is a light water research reactor situated at the Research Centre Rez, near Prague. It operates as a multipurpose facility with a maximum thermal power of 10 MW. The reactor core usually contains from 28 to 32 fuel assemblies with a total mass of {sup 235}U of about 5 kg. Emitted radiation from the fuel caused by fission is shielded by moderating water, a steel reactor vessel, and heavy concrete. This paper deals with measurement and analysis of the gamma spectrum near the outer surface of the concrete wall, behind biological shielding, mainly in the 3- to 10-MeV energy range. A portable HPGe detector with a portable multichannel analyzer was used to measure gamma spectra. The origin of energy lines in gamma detector spectra was identified. (authors)

  9. 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)

  10. Work for radiation shielding concrete in large-scaled radiation facilities

    International Nuclear Information System (INIS)

    Konomi, Shinzo; Sato, Shoni; Otake, Takao.

    1980-01-01

    This paper reports the radiation shielding concrete work in the construction of radiation laboratory facilities of Electrotechnical Laboratory, a Japanese Government agency for the research and development of electronic technology. The radiation shielding walls of the facilities are made of ordinary concrete, heavy weight concrete and raw iron ore. This paper particularly relates the use of ordinary concrete which constitutes the majority of such concretes. The concrete mix was determined so as to increase its specific gravity for better shielding effect, to improve mass concrete effect and to advance good workability. The tendency of the concrete to decrease its specific gravity and the temperature variations were also made on how to place concrete to secure good shielding effect and uniform quality. (author)

  11. Design of radiation shields in nuclear reactor core

    International Nuclear Information System (INIS)

    Mousavi Shirazi, A.; Daneshvar, Sh.; Aghanajafi, C.; Jahanfarnia, Gh.; Rahgoshay, M.

    2008-01-01

    This article consists of designing radiation shields in the core of nuclear reactors to control and restrain the harmful nuclear radiations in the nuclear reactor cores. The radiation shields protect the loss of energy. caused by nuclear radiation in a nuclear reactor core and consequently, they cause to increase the efficiency of the reactor and decrease the risk of being under harmful radiations for the staff. In order to design these shields, by making advantages of the O ppenheim Electrical Network m ethod, the structure of the shields are physically simulated and by obtaining a special algorithm, the amount of optimized energy caused by nuclear radiations, is calculated

  12. Analytical determination of traced elements in concrete samples used in nuclear reactors of the European Community

    International Nuclear Information System (INIS)

    May, S.; Piccot, D.

    1984-01-01

    In reactor dismantling residual radioacting of concrete used, especially in biological shield can brought problems for treatment and disposal. Radioactivity of concrete from reactors can be forecasted if element content is known. Elements producing long life radionuclides are: chlorine, calcium nickel, cobalt, niobium, europium and samarium. Neutron activation analysis is used for determination of these elements whithout chemical separation for Ca, Co, Eu and Sm and with radiochemical separation for Cl, Ni and Nb. A lot of elements, less interesting are also determined by gamma spectrometry after irradiation. It was possible to determine 29 elements in 21 concrete samples from different European Community reactors

  13. Study on Basic Characteristics for the Development of Radiation Shielding High-Weight Concrete

    Energy Technology Data Exchange (ETDEWEB)

    Mun, Young Bum; Lee, Jea Hyung; Choi, Hyun Kook [Sungshin Cement CO., Sejong (Korea, Republic of); Oh, Jeong Hwan; Choi, Soo Seok [Jeju National University, Jeju (Korea, Republic of)

    2016-05-15

    It is planned to build a power plant more than 6 units. Although the demand of a nuclear power plant is going to increase, the attention for radiation shielding is relatively in a low level. Concrete is one of the excellent and widely used shielding materials. Since the radiation shielding of a given material is proportional to density and thickness, a high-weight concrete with high-weight aggregate which is higher than normal concrete is used for radiation shielding. However, there are a few studies and references about radiation shielding concrete. Therefore, it is required to find a high-weight aggregate. The purpose of this paper is the development of a highweight concrete to improve radiation shielding capability. The radiation shielding rate of high-weight concrete is higher than that of reference concrete. It is confirmed that the density of aggregate and the unit weight of concreate is proportional to the radiation shielding rate. In addition, the chemical composition of aggregate has also has an important effect on γ-ray shielding. Therefore, high weight aggregates of higher density are essentially required to improve radiation shielding capability. The compressive strength of a high weight concrete is better than that of reference concrete. Slump and air contents, however, are slightly increased with by-product aggregates.

  14. 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.

  15. Evaluation of the performance of peridotite aggregates for radiation shielding concrete

    International Nuclear Information System (INIS)

    Wang, Jinjun; Li, Guofeng; Meng, Dechuan

    2014-01-01

    Highlights: • Using peridotite rich in crystal water as aggregates of radiation-shielding concrete. • Performance of peridotite concrete is simulated and compared with ordinary concrete. • Performance of concrete samples is tested. • Neutron shielding performance can be significantly enhanced by peridotite aggregates. - Abstract: Peridotite is a kind of material that is rich in crystal water. In this paper, peridotite is used as fine and coarse aggregates for radiation shielding concrete. The transmission data of different concrete thickness and different energy neutron are calculated using Monte-Carlo method. The neutron shielding performance of the peridotite concrete samples are tested using 241 Am-Be neutron source. The results show that the peridotite is an excellent neutron shielding material

  16. Upper shielding body in LMFBR type reactors

    International Nuclear Information System (INIS)

    Shoji, Koichi.

    1986-01-01

    Purpose: Preference is given to the strength and thermal insulation of a roof slab thereby ensuring axial size and improving the operationability upon inserting the control rod in the upper shielding body of LMFBR type reactors. Constitution: In an upper shielding body in which a large rotational plug is rotatably mounted to a circular hole formed at an eccentric position of a roof slab, while a small rotational plug is rotatably mounted to a circular hole disposed at an eccentric position of the large rotational plug and the reactor core upper mechanisms are supported on the small rotational plug, heat insulation layers are attached to the inside of the inner circumferential wall of the roof slab and the outer circumferential wall of the large rotational plug. By attaching the heat insulation layers, the heat conduction between the roof slab and the large rotational plug can be suppressed remarkably, by which occurrence of specific heat pass or local generation of large thermal stresses can be avoided even if difference is resulted to the temperature distribution between them. In this way, functions taking advantage of respective features of the roof slab and the small rotational plug can be obtained to achieve the purpose. (Kamimura, M.)

  17. Study and installation of concrete shielding in the civil engineering of nuclear construction (1960)

    International Nuclear Information System (INIS)

    Dubois, F.

    1960-01-01

    The object of this report is to give technical information about high density concretes which have become very important for radiation biological shielding. The most generally used heavy aggregates (barytes, ilmenite, ferrophosphorus, limonite, magnetite and iron punching) to make these concretes are investigated from the point of view prospecting and physical and chemical characteristics. At first, a general survey of shielding concretes is made involving the study of components, mixing and placing methods, then, a detailed investigation of some high density concretes: barytes concrete, with incorporation of iron punching or iron shot, ferrophosphorus concrete, ilmenite concrete and magnetite concrete, more particularly with regard to grading and mix proportions and testing process. To put this survey in concrete form, two practical designs are described such as they have been carried out at the Saclay Nuclear Station. Specifications are given for diverse concretes and for making the proton-synchrotron 'Saturne' shielding blocks. (author) [fr

  18. 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

  19. Economic analysis of sectional concrete blocks uses in biological shieldings

    International Nuclear Information System (INIS)

    Ivanov, V.N.

    1977-01-01

    The relative economy of different structural embodiments of the biological protection of a research reactor has been evaluated. The alternatives include cast in-situ concrete and prefabricated blocks with different linear dimension tolerances (+-2, +-5 and +-7 mm). The cost-benefit estimates have been done according to the reduced cost calculated for the final products - the erected structures. It has been found that the optimum tolerances for 6 meter-long blocks are not less than +-5 mm for the other linear dimensions. The optimum concrete block volume for dismountable structures is 1 to 1.5 m 3 and for prefabricated protection structures -more than 4 m 3

  20. Calculation of a concrete shielding for an ILU-8 D electron accelerator

    International Nuclear Information System (INIS)

    Helal, A.; Imam, A.

    1996-01-01

    A concrete shielding for an electron accelerator of 1 MeV is suggested to replace its structural steel shielding. The thickness of such a shield is calculated. The calculational model used is based on standard and transmission curves given in the literature. The calculated concrete shielding is generally adequate to attenuate the accelerator produced radiation to a level 1 μ Gy/h or less at any point outside of the vault enclosure. 5 figs

  1. Calculation of a concrete shielding for an ILU-8 D electron accelerator

    Energy Technology Data Exchange (ETDEWEB)

    Helal, A [Nuclear Research Center, AEA, Cairo (Egypt); Imam, A [National Center for Nuclear Safety and Radiation Control, AEA, Cairo (Egypt)

    1997-12-31

    A concrete shielding for an electron accelerator of 1 MeV is suggested to replace its structural steel shielding. The thickness of such a shield is calculated. The calculational model used is based on standard and transmission curves given in the literature. The calculated concrete shielding is generally adequate to attenuate the accelerator produced radiation to a level 1 {mu} Gy/h or less at any point outside of the vault enclosure. 5 figs.

  2. 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

  3. 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)

  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. Embrittlement of the Shippingport reactor shield tank

    International Nuclear Information System (INIS)

    Chopra, O.K.; Shack, W.J.

    1989-01-01

    Surveillance specimens from the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory showed an unexpectedly high degree of embrittlement relative to the data obtained on similar materials in Materials Testing Reactors (MTRs). The results suggest a possible negative flux effect and raise the issue of embrittlement of the pressure vessel support structures of commercial light water reactors. To help resolve this issues, a program was initiated to characterize the irradiation embrittlement of the neutron shield tank (NST) from the decommissioned Shippingport reactor. The Shippingport NST operated at 55 degree C (130 degree F) and was fabricated from A212 Grade B steel, similar to the vessel material in HFIR. The inner wall of the NST was exposed to a total maximum fluence of ∼ 6 x 10 17 n/cm 2 (E > 1 MeV) over a life of 9.25 effective full power years. This corresponds to a fast flux of 2.1 x 10 9 n/cm 2 x s and is comparable to the conditions for the HFIR surveillance specimens. The results indicate that irradiation increases the 15 ft x lb Charpy transition temperature (CTT) by ∼25 degree C (45 degree F) and decreases the upper shelf energy. The shift in CTT is not as severe as that observed in the HFIR surveillance specimens and is consistent with that expected from the MTR data base. However, the actual value of CTT is high, and the toughness at service temperature is low, even when compared with the HFIR data. The increase in yield stress is ∼50 MPa, which is comparable to the HFIR data. The results also indicate a lower impact strength and higher transition temperature for the TL orientation than that for the LT orientation. Some effects of the location across the thickness of the wall are also observed for the LT specimens; CTT is slightly greater for the specimens from the inner region of the wall

  6. Method for temporary shielding of reactor vessel internals

    International Nuclear Information System (INIS)

    Grimm, N.P.; Sejvar, J.

    1991-01-01

    This patent describes a method for shielding stored internals for reactor vessel annealing. It comprises removing nuclear fuel from the reactor vessel containment building; removing and storing upper and lower core internals under water in a refueling canal storage area; assembling a support structure in the refueling canal between the reactor vessel and the stored internals; introducing vertical shielding tanks individually through a hatch in the containment building and positioning each into the support structure; introducing horizontal shielding tanks individually through a hatch in the containment building and positioning each above the stored internals and vertical tanks; draining water from the refueling canal to the level of a flange of the reactor vessel; placing an annealing apparatus in the reactor vessel; pumping the remaining water from the reactor vessel; and annealing the reactor vessel

  7. Gravity Scaling of a Power Reactor Water Shield

    International Nuclear Information System (INIS)

    Reid, Robert S.; Pearson, J. Boise

    2008-01-01

    Water based reactor shielding is being considered as an affordable option for potential use on initial lunar surface reactor power systems. Heat dissipation in the shield from nuclear sources must be rejected by an auxillary thermal hydraulic cooling system. The mechanism for transferring heat through the shield is natural convection between the core surface and an array of thermosyphon radiator elements. Natural convection in a 100 kWt lunar surface reactor shield design has been previously evaluated at lower power levels (Pearson, 2006). The current baseline assumes that 5.5 kW are dissipated in the water shield, the preponderance on the core surface, but with some volumetric heating in the naturally circulating water as well. This power is rejected by a radiator located above the shield with a surface temperature of 370 K. A similarity analysis on a water-based reactor shield is presented examining the effect of gravity on free convection between a radiation shield inner vessel and a radiation shield outer vessel boundaries. Two approaches established similarity: 1) direct scaling of Rayleigh number equates gravity-surface heat flux products, 2) temperature difference between the wall and thermal boundary layer held constant on Earth and the Moon. Nussult number for natural convection (laminar and turbulent) is assumed of form Nu = CRa n . These combined results estimate similarity conditions under Earth and Lunar gravities. The influence of reduced gravity on the performance of thermosyphon heat pipes is also examined

  8. Discussion on the standardization of concrete composition for radiation shielding design 2. Evaluation of the effect of the composition variance on the shielding property

    International Nuclear Information System (INIS)

    Ogata, Tomohiro; Kimura, Ken-ichi; Nakata, Mikihiro; Okuno, Koichi; Ishikawa, Tomoyuki

    2017-01-01

    Radiation Shielding Material Standardization Working Group of AESJ has been organized to establish Japanese standard concrete composition for radiation shielding design. We have collected concrete composition data to organize a representative concrete composition data. Neutron and Gamma dose rates penetrated through several concrete compositions are calculated by one dimensional discrete ordinate code ANISN. Effects of the variation of concrete composition on the neutron and gamma dose are evaluated. In this paper, recent standardization activity is summarized. (author)

  9. Basic nuclear data and reactor shielding design formulaire PROPANE Do

    International Nuclear Information System (INIS)

    Estiot, J.C.; Salvatores, M.; Trapp, J.P.

    1979-01-01

    This paper presents a calculational scheme - formulaire PROPANE - to calculate the deep neutron penetration in the fast reactor shield. The emphasis is put on the multigroup data and method approximations. The performances of this formulaire are presented

  10. Water and Regolith Shielding for Surface Reactor Missions

    Science.gov (United States)

    Poston, David I.; Ade, Brian J.; Sadasivan, Pratap; Leichliter, Katrina J.; Dixon, David D.

    2006-01-01

    This paper investigates potential shielding options for surface power fission reactors. The majority of work is focused on a lunar shield that uses a combination of water in stainless-steel cans and lunar regolith. The major advantage of a water-based shield is that development, testing, and deployment should be relatively inexpensive. This shielding approach is used for three surface reactor concepts: (1) a moderated spectrum, NaK cooled, Hastalloy/UZrH reactor, (2) a fast-spectrum, NaK-cooled, SS/UO2 reactor, and (3) a fast-spectrum, K-heat-pipe-cooled, SS/UO2 reactor. For this study, each of these reactors is coupled to a 25-kWt Stirling power system, designed for 5 year life. The shields are designed to limit the dose both to the Stirling alternators and potential astronauts on the surface. The general configuration used is to bury the reactor, but several other options exist as well. Dose calculations are presented as a function of distance from reactor, depth of buried hole, water boron concentration (if any), and regolith repacked density.

  11. Water and Regolith Shielding for Surface Reactor Missions

    International Nuclear Information System (INIS)

    Poston, David I.; Sadasivan, Pratap; Dixon, David D.; Ade, Brian J.; Leichliter, Katrina J.

    2006-01-01

    This paper investigates potential shielding options for surface power fission reactors. The majority of work is focused on a lunar shield that uses a combination of water in stainless-steel cans and lunar regolith. The major advantage of a water-based shield is that development, testing, and deployment should be relatively inexpensive. This shielding approach is used for three surface reactor concepts: (1) a moderated spectrum, NaK cooled, Hastalloy/UZrH reactor, (2) a fast-spectrum, NaK-cooled, SS/UO2 reactor, and (3) a fast-spectrum, K-heat-pipe-cooled, SS/UO2 reactor. For this study, each of these reactors is coupled to a 25-kWt Stirling power system, designed for 5 year life. The shields are designed to limit the dose both to the Stirling alternators and potential astronauts on the surface. The general configuration used is to bury the reactor, but several other options exist as well. Dose calculations are presented as a function of distance from reactor, depth of buried hole, water boron concentration (if any), and regolith repacked density

  12. 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)

  13. Study and application of high-density concrete in radiation-shielding experiment

    International Nuclear Information System (INIS)

    Wu Chongming; Ding Dexin; Xiao Xuefu; Wang Shaolin; Lin Xingjun; Shen Yuanyuan

    2008-01-01

    According to the demand for research and construction project, a series of systematic experiments and studies on shielding γ-ray radiation concrete with the density of 4.60 t/m 3 were made in such aspects as mix ratio design, construction technology, uniformly shielding etc. Such issues as uniformity in the construction and compactness were solved. The ray test method for uniformly shielding concrete was presented and some technical steps for this high-density concrete used in the process of test design or construction were summed up. A series of tests and practical applications show that this technology of mix ratio design and construction is feasible. (authors)

  14. Monte Carlo based radial shield design of typical PWR reactor

    Energy Technology Data Exchange (ETDEWEB)

    Gul, Anas; Khan, Rustam; Qureshi, M. Ayub; Azeem, Muhammad Waqar; Raza, S.A. [Pakistan Institute of Engineering and Applied Sciences, Islamabad (Pakistan). Dept. of Nuclear Engineering; Stummer, Thomas [Technische Univ. Wien (Austria). Atominst.

    2016-11-15

    Neutron and gamma flux and dose equivalent rate distribution are analysed in radial and shields of a typical PWR type reactor based on the Monte Carlo radiation transport computer code MCNP5. The ENDF/B-VI continuous energy cross-section library has been employed for the criticality and shielding analysis. The computed results are in good agreement with the reference results (maximum difference is less than 56 %). It implies that MCNP5 a good tool for accurate prediction of neutron and gamma flux and dose rates in radial shield around the core of PWR type reactors.

  15. Space nuclear reactor shields for manned and unmanned applications

    International Nuclear Information System (INIS)

    McKissock, B.I.; Bloomfield, H.S.

    1990-01-01

    Missions which use nuclear reactor power systems require radiation shielding of payload and/or crew areas to predetermined dose rates. Since shielding can become a significant fraction of the total mass of the system, it is of interest to show the effect of various parameters on shield thickness and mass for manned and unmanned applications. Algorithms were developed to give the thicknesses needed if reactor thermal power, separation distances and dose rates are given as input. The thickness algorithms were combined with models for four different shield geometries to allow tradeoff studies of shield volume and mass for a variety of manned and unmanned missions. The shield design tradeoffs presented in this study include the effects of: higher allowable dose rates; radiation hardened electronics; shorter crew exposure times; shield geometry; distance of the payload and/or crew from the reactor; and changes in the size of the shielded area. Specific NASA missions that were considered in this study include unmanned outer planetary exploration, manned advanced/evolutionary space station and advanced manned lunar base. (author)

  16. Space nuclear reactor shields for manned and unmanned applications

    International Nuclear Information System (INIS)

    Mckissock, B.I.; Bloomfield, H.S.

    1989-01-01

    Missions which use nuclear reactor power systems require radiation shielding of payload and/or crew areas to predetermined dose rates. Since shielding can become a significant fraction of the total mass of the system, it is of interest to show the effect of various parameters on shield thickness and mass for manned and unmanned applications. Algorithms were developed to give the thicknesses needed if reactor thermal power, separation distances, and dose rates are given as input. The thickness algorithms were combined with models for four different shield geometries to allow tradeoff studies of shield volume and mass for a variety of manned and unmanned missions. Shield design tradeoffs presented in this study include the effects of: higher allowable dose rates; radiation hardened electronics; shorter crew exposure times; shield geometry; distance of the payload and/or crew from the reactor; and changes in the size of the shielded area. Specific NASA missions that were considered in this study include unmanned outer planetary exploration, manned advanced/evolutionary space station, and advanced manned lunar base

  17. Characterisation of the inventory of radioisotopes induced in the biological shield a WWER-440 reactor

    International Nuclear Information System (INIS)

    Feher, S.; Czifrus, Sz.; Zsolnay, E.M.; Szondi, E.

    2001-01-01

    A significant part of the radwaste originating from the decommissioning of NPPs is made up of the activated concrete and steel components of the biological shield. The paper presents the results of studies aimed at the determination of the amount of radionuclides accumulating in the serpentinous and ordinary concrete shield around the WWER-440 reactors of the Paks NPP. For the calculations, the reactor, vessel and shield were modelled in detail both in terms of geometry and material composition. The spatial and energy distribution of the activating neutron spectrum was determined by certain modules of SCALE 4.3 and the code TORT in two and three dimensions, while the activation was calculated using ORIGEN-S for 22 geometrical regions. The results showed that the activity of the concrete structures at final shutdown after 30 years of operation is approximately 50 TBq, which decreases to 20, 12, 1.1 TBq and 27 GBq after 1 month, 1 year, 10 and 100 years, respectively (Authors)

  18. Dismantling method for reactor shielding wall and device therefor

    International Nuclear Information System (INIS)

    Akagawa, Katsuhiko.

    1995-01-01

    A ring member having an outer diameter slightly smaller than an inner diameter of a reactor shielding wall to be dismantled is lowered in the inside of the reactor shielding wall while keeping a horizontal posture. A cutting device is disposed at the lower peripheral edge of the ring member. The cutting device can move along the peripheral edge of the circular shape of the ring member. The ring member is urged against the inner surface of the reactor shielding wall by using an urging member to immobilize the ring member. Then, the cutting device is operated to cut the reactor shielding wall into a plurality of ring-like blocks at a plurality of inner horizontal ribs or block connection ribs. Then, the blocks of the cut reactor shielding wall are supported by the ring member, and transported out of the reactor container by a lift. The cut blocks transported to the outside are finely dismantled for every block in a closed chamber. (I.N.)

  19. Status of reactor-shielding research in the US

    International Nuclear Information System (INIS)

    Maienshein, F.C.

    1980-01-01

    While reactor programs change, shielding analysis methods are improved slowly. Version-V of ENDF/B provides improved data and Version-VI will be cost effective in advanced fission reactors are to be developed in the US. Benchmarks for data and methods validation are collected and distributed in the US in two series, one primarily for FBR-related experiments and one for LWR calculational methods. For LWR design, cavity streaming is now handled adequately, if with varying degrees of elegance. Investigations of improved detector response for LWRs rely upon transport methods. The great potential importance of pressure-vessel damage is dreflected in widespread studies to aid in the prediction of neutron fluences in vessels. For LMFBRS, the FFTF should give attenuation results on an operating reactor. For larger power reactors, the advantages of alternate shield materials appear compelling. A few other shielding studies appear to require experimental confirmation if LMFBRs are to be economically competitive. A coherent shielding program for the GCFR is nearing completion. For the fusion-reactor program, methods verification is under way, practical calculations are well advanced for test devices such as the TFTR and FMIT, and consideration is now given to shielding problems of large reactors, as in the ETF study

  20. High Density Radiation Shielding Concretes for Hot Cells of 99mTc Project

    International Nuclear Information System (INIS)

    Sakr, K.

    2006-01-01

    High density concrete [more than 3.6 ton/m 3 (3.6x10 3 kg/m 3 )] was prepared to be used as a radiation shielding concrete (RSC) for hot-cells in gel technetium project at inshas to attenuate gamma radiation emitted from radioactive sources. different types of concrete were prepared by mixing local mineral aggregates mainly gravel and ilmenite . iron shots were added to the concrete mixture proportion as partial replacement of heavy aggregates to increase its density. the physical properties of prepared concrete in both plastic and hardened phases were investigated. compressive strength and radiation attenuation of gamma rays were determined. Results showed that ilmenite concrete mixed with iron shots had the highest density suitable to be use as RSC according to the chinese hot cell design requirements. Recommendations to avoid some technical problems of manufacturing radiation shielding concrete were maintained

  1. 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

  2. Shielding device for control rod in nuclear reactor

    International Nuclear Information System (INIS)

    Sakamaki, Kazuo; Tomatsu, Tsutomu.

    1995-01-01

    The device of the present invention shields radiation emitted from control rods to greatly reduce an operator's radiation exposure even if reactor water level is lowered and the upper portion of the control rod is exposed upon inspection of a BWR type reactor. Namely, a shield assembly has a structure comprising a set of four columnar shields in a two-row and two-column arrangement, which can be inserted into a control rod guide tube. Upon conducting inspection, the control rod is lowered into the control rod guide tube, and in this state, the columnar shields of the shield assembly are inserted to the control rod in the control rod guide tube. With such procedures, the upper portion of the control rod protruded from the control rod guide tube is covered with the shield assembly. As a result, radiation leaked from the control rod is shielded. Accordingly, irradiation in the reactor due to leaked radiation can be prevented thereby enabling to reduce an operator's radiation exposure. (I.S.)

  3. Study of some health physics parameters of bismuth-ground granulated blast furnace slag shielding concretes

    Energy Technology Data Exchange (ETDEWEB)

    Kumar, Sandeep, E-mail: sandeep0078monu@gmail.com [Department of Physics, Punjabi University, Patiala (India); Singh, Sukhpal, E-mail: sukhpal-78@rediffmail.com [Department of Basic and Applied Sciences, Punjabi University, Patiala (India)

    2016-05-06

    The Bismuth-ground granulated blastfurnace slang (Bi-GGBFS) concrete samples were prepared. The weight percentage of different elements present in Bi-GGBFS Shielding concretewas evaluated by Energy Dispersive X-ray Microanalysis (EDX). The exposure rate and absorbed dose rate characteristics were calculated theoretically for radioactive sources namely {sup 241}Am and {sup 137}Cs. Our calculations reveal that the Bi-GGBFS concretes are effective in shielding material for gamma radiations.

  4. Study of some health physics parameters of bismuth-ground granulated blast furnace slag shielding concretes

    Science.gov (United States)

    Kumar, Sandeep; Singh, Sukhpal

    2016-05-01

    The Bismuth-ground granulated blastfurnace slang (Bi-GGBFS) concrete samples were prepared. The weight percentage of different elements present inBi-GGBFS Shielding concretewas evaluated by Energy Dispersive X-ray Microanalysis (EDX). The exposure rate and absorbed dose rate characteristics were calculated theoretically for radioactive sources namely 241Am and 137Cs. Our calculations reveal that the Bi-GGBFS concretes are effective in shielding material for gamma radiations.

  5. Study of some health physics parameters of bismuth-ground granulated blast furnace slag shielding concretes

    International Nuclear Information System (INIS)

    Kumar, Sandeep; Singh, Sukhpal

    2016-01-01

    The Bismuth-ground granulated blastfurnace slang (Bi-GGBFS) concrete samples were prepared. The weight percentage of different elements present in Bi-GGBFS Shielding concretewas evaluated by Energy Dispersive X-ray Microanalysis (EDX). The exposure rate and absorbed dose rate characteristics were calculated theoretically for radioactive sources namely "2"4"1Am and "1"3"7Cs. Our calculations reveal that the Bi-GGBFS concretes are effective in shielding material for gamma radiations.

  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. Radiological shielding of low power compact reactor: calculation and design

    International Nuclear Information System (INIS)

    Marino, Raul

    2004-01-01

    The development of compact reactors becoming a technology that offers great projection and innumerable use possibilities, both in electricity generation and in propulsion.One of the requirements for the operation of this type of reactor is that it must include a radiological shield that will allow for different types of configurations and that, may be moved with the reactor if it needs to be transported.The nucleus of a reactor emits radiation, mainly neutrons and gamma rays in the heat of power, and gamma radiation during the radioactive decay of fission products.This radiation must be restrained in both conditions of operation to avoid it affecting workers or the public.The combination of different materials and properties in layers results in better performance in the form of a decrease in radiation, hence causing the dosage outside the reactor, whether in operation or shut down, to fall within the allowed limits.The calculations and design of radiological shields is therefore of paramount importance in reactor design.The choice of material and the design of the shield have a strong impact on the cost and the load capacity, the latter being one of the characteristics to optimize.The imposed condition of design is that the reactor can be transported together with the decay shield in a standard container of 40 foot [es

  8. Preliminary shielding design evaluation for reactor assembly of SMART

    International Nuclear Information System (INIS)

    Kim, Kyo Youn; Kang, Chang M.; Kim, Ha Yong; Zee, Sung Quun; Chang, Moon Hee

    1999-03-01

    This report describes a preliminary evaluations of SMART shielding design near the reactor core by using the DORT two-dimensional discrete ordinates transport code. The results indicate that maximum neutron fluence at the bottom of reactor vessel is 1.64x10 17 n/cm 2 and that on the radial surface of reactor vessel is 6.71x10 16 n/cm 2 . These results meet the requirement, 1.0x10 20 n/cm 2 , in 10 CFR 50.61 and the integrity of SMART reactor vessel is confirmed during the lifetime of reactor. (Author). 20 refs., 11 tabs., 8 figs

  9. Construction of reactor vessel bottom of prestressed reinforced concrete

    International Nuclear Information System (INIS)

    Sitnikov, M.I.; Metel'skij, V.P.

    1980-01-01

    Methods are described for building reactor vessel bottoms of prestressed reinforced concrete during NPPs construction in Great Britain, France, Germany (F.R.) and the USA. Schematic of operations performed in succession is presented. Considered are different versions of one of the methods for concreting a space under a facing by forcing concrete through a hole in the facing. The method provides tight sticking of the facing to the reactor vessel bottom concrete

  10. Shielding assessment of the ETRR-1 Reactor Under power upgrading

    Energy Technology Data Exchange (ETDEWEB)

    Ahmad, E E [Reactor Department, Nuclear Research Center, Atomic Energy Authority, Cairo (Egypt)

    1997-12-31

    The assessment of existing shielding of the ETRR-1 reactor in case of power upgrading is presented and discussed. It was carried out using both the present EK-10 type fuel elements and some other types of fuel elements with different enrichments. The shielding requirements for the ETRR-1 when power is upgraded are also discussed. The optimization curves between the upgraded reactor power and the shield thickness are presented. The calculation have been made using the ANISN code with the DLC-75 data library. The results showed that the present shield necessitates an additional layer of steel with thickness of 10.20 and 25 cm. When its power is upgraded to 3, 6 and 10 MWt in order to cutoff all neutron energy groups to be adequately safe under normal operating conditions. 4 figs.

  11. Gamma dose from activation of internal shields in IRIS reactor.

    Science.gov (United States)

    Agosteo, Stefano; Cammi, Antonio; Garlati, Luisella; Lombardi, Carlo; Padovani, Enrico

    2005-01-01

    The International Reactor Innovative and Secure is a modular pressurised water reactor with an integral design. This means that all the primary system components, such as the steam generators, pumps, pressuriser and control rod drive mechanisms, are located inside the reactor vessel, which requires a large diameter. For the sake of better reliability and safety, it is desirable to achieve the reduction of vessel embrittlement as well as the lowering of the dose beyond the vessel. The former can be easily accomplished by the presence of a wide downcomer, filled with water, which surrounds the core region, while the latter needs the presence of additional internal shields. An optimal shielding configuration is under investigation, for reducing the ex-vessel dose due to activated internals and for limiting the amount of the biological shielding. MCNP 4C calculations were performed to evaluate the neutron and the gamma dose during operation and the 60Co activation of various shields configurations. The gamma dose beyond the vessel from activation of its structural components was estimated in a shutdown condition, with the Monte Carlo code FLUKA 2002 and the MicroShield software. The results of the two codes are in agreement and show that the dose is sufficiently low, even without an additional shield.

  12. Gamma dose from activation of internal shields in IRIS reactor

    International Nuclear Information System (INIS)

    Agosteo, S.; Cammi, A.; Garlati, L.; Lombardi, C.; Padovani, E.

    2005-01-01

    The International Reactor Innovative and Secure is a modular pressurised water reactor with an integral design. This means that all the primary system components, such as the steam generators, pumps, pressurizer and control rod drive mechanisms, are located inside the reactor vessel, which requires a large diameter. For the sake of better reliability and safety, it is desirable to achieve the reduction of vessel embrittlement as well as the lowering of the dose beyond the vessel. The former can be easily accomplished by the presence of a wide downcomer, filled with water, which surrounds the core region, while the latter needs the presence of additional internal shields. An optimal shielding configuration is under investigation, for reducing the ex-vessel dose due to activated internals and for limiting the amount of the biological shielding. MCNP 4C calculations were performed to evaluate the neutron and the gamma dose during operation and the 60 Co activation of various shields configurations. The gamma dose beyond the vessel from activation of its structural components was estimated in a shutdown condition, with the Monte Carlo code FLUKA 2002 and the MicroShield software. The results of the two codes are in agreement and show that the dose is sufficiently low, even without an additional shield. (authors)

  13. Concept of spatial channel theory applied to reactor shielding analysis

    International Nuclear Information System (INIS)

    Williams, M.L.; Engle, W.W. Jr.

    1977-01-01

    The concept of channel theory is used to locate spatial regions that are important in contributing to a shielding response. The method is analogous to the channel-theory method developed for ascertaining important energy channels in cross-section analysis. The mathematical basis for the theory is shown to be the generalized reciprocity relation, and sample problems are given to exhibit and verify properties predicted by the mathematical equations. A practical example is cited from the shielding analysis of the Fast Flux Test Facility performed at Oak Ridge National Laboratory, in which a perspective plot of channel-theory results was found useful in locating streaming paths around the reactor cavity shield

  14. Shielding efficiency of metal hydrides and borohydrides in fusion reactors

    DEFF Research Database (Denmark)

    Singh, Vishvanath P.; Badiger, Nagappa M.; Gerward, Leif

    2016-01-01

    at energies 0.015 MeV to15 MeV, and for penetration depths up to 40 mean free paths. Fast-neutron shielding efficiency has been characterized by the effective neutron removal cross-section. It is shown that ZrH2 and VH2 are very good shielding materials for gamma rays and fast neutrons due to their suitable...... combination of low-and high-Z elements. The present work should be useful for the selection and design of blankets and shielding, and for dose evaluation for components in fusion reactors....

  15. 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)

  16. Summary of the fifth international conference on reactor shielding

    International Nuclear Information System (INIS)

    Roussin, R.W.; Abbott, L.S.; Bartine, D.E.

    1977-01-01

    The Fifth International Conference on Reactor Shielding was held April 18-23, 1977 in Knoxville, Tennessee. The meeting was the largest in the series and attracted participants from 34 countries. The 10 invited papers and 10 of the contributed papers, selected as being representative of the Conference by the Technical Program Committee, are published in this issue of ATOMKERNENERGIE. This collection of papers demonstrates that the field of nuclear reactor shielding has developed into a mature discipline while retaining a definite vitality. (orig.) [de

  17. Evaluation of the concrete shield compositions from the 2010 criticality accident alarm system benchmark experiments at the CEA Valduc SILENE facility

    International Nuclear Information System (INIS)

    Miller, Thomas Martin; Celik, Cihangir; Dunn, Michael E; Wagner, John C; McMahan, Kimberly L; Authier, Nicolas; Jacquet, Xavier; Rousseau, Guillaume; Wolff, Herve; Savanier, Laurence; Baclet, Nathalie; Lee, Yi-kang; Trama, Jean-Christophe; Masse, Veronique; Gagnier, Emmanuel; Naury, Sylvie; Blanc-Tranchant, Patrick; Hunter, Richard; Kim, Soon; Dulik, George Michael; Reynolds, Kevin H.

    2015-01-01

    In October 2010, a series of benchmark experiments were conducted at the French Commissariat a l'Energie Atomique et aux Energies Alternatives (CEA) Valduc SILENE facility. These experiments were a joint effort between the United States Department of Energy Nuclear Criticality Safety Program and the CEA. The purpose of these experiments was to create three benchmarks for the verification and validation of radiation transport codes and evaluated nuclear data used in the analysis of criticality accident alarm systems. This series of experiments consisted of three single-pulsed experiments with the SILENE reactor. For the first experiment, the reactor was bare (unshielded), whereas in the second and third experiments, it was shielded by lead and polyethylene, respectively. The polyethylene shield of the third experiment had a cadmium liner on its internal and external surfaces, which vertically was located near the fuel region of SILENE. During each experiment, several neutron activation foils and thermoluminescent dosimeters (TLDs) were placed around the reactor. Nearly half of the foils and TLDs had additional high-density magnetite concrete, high-density barite concrete, standard concrete, and/or BoroBond shields. CEA Saclay provided all the concrete, and the US Y-12 National Security Complex provided the BoroBond. Measurement data from the experiments were published at the 2011 International Conference on Nuclear Criticality (ICNC 2011) and the 2013 Nuclear Criticality Safety Division (NCSD 2013) topical meeting. Preliminary computational results for the first experiment were presented in the ICNC 2011 paper, which showed poor agreement between the computational results and the measured values of the foils shielded by concrete. Recently the hydrogen content, boron content, and density of these concrete shields were further investigated within the constraints of the previously available data. New computational results for the first experiment are now available

  18. Recycling radioactive scrap metal by producing concrete shielding with steel granules

    International Nuclear Information System (INIS)

    Sappok, M.

    1996-01-01

    Siempelkamp foundry at Krefeld, Germany, developed a method for recycling radioactively contaminated steel from nuclear installations. The material is melted and used for producing shielding plates, containers, etc., on a cast-iron basis. Because the percentage of stainless steel has recently increased significantly, problems in the production of high-quality cast iron components have also grown. The metallurgy, the contents of nickel and chromium especially, does not allow for the recycling of stainless steel in a percentage to make this process economical. In Germany, the state of the art is to use shielded concrete containers for the transport of low active waste; this concrete is produced by using hematite as an additive for increasing shielding efficiency. The plan was to produce steel granules from radioactive scrap metal as a substitute for hematite in shielding concrete

  19. High-performance heavy concrete as a multi-purpose shield

    International Nuclear Information System (INIS)

    Mortazavi, S. M. J.; Mosleh-Shirazi, M. A.; Roshan-Shomal, P.; Raadpey, N.; Baradaran-Ghahfarokhi, M.

    2010-01-01

    Concrete has long been used as a shield against high-energy photons and neutrons. In this study, colemanite and galena minerals (CoGa) were used for the production of an economical high-performance heavy concrete. To measure the gamma radiation attenuation of the CoGa concrete samples, they were exposed to a narrow beam of gamma rays emitted from a 60 Co radiotherapy unit. An Am-Be neutron source was used for assessing the shielding properties of the samples against neutrons. The compression strengths of both types of concrete mixes (CoGa and reference concrete) were investigated. The range of the densities of the heavy concrete samples was 4100-4650 kg m -3 , whereas it was 2300-2600 kg m -3 in the ordinary concrete reference samples. The half-value layer of the CoGa concrete samples for 60 Co gamma rays was 2.49 cm; much less than that of ordinary concrete (6.0 cm). Moreover, CoGa concrete samples had a 10% greater neutron absorption compared with reference concrete. (authors)

  20. 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.

  1. 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.

  2. Tungsten-based composite materials for fusion reactor shields

    International Nuclear Information System (INIS)

    Greenspan, E.; Karni, Y.

    1985-01-01

    Composite tungsten-based materials were recently proposed for the heavy constituent of compact fusion reactor shields. These composite materials will enable the incorporation of tungsten - the most efficient nonfissionable inelastic scattering (as well as good neutron absorbing and very good photon attenuating) material - in the shield in a relatively cheap way and without introducing voids (so as to enable minimizing the shield thickness). It is proposed that these goals be achieved by bonding tungsten powder, which is significantly cheaper than high-density tungsten, with a material having the following properties: good shielding ability and relatively low cost and ease of fabrication. The purpose of this work is to study the effectiveness of the composite materials as a function of their composition, and to estimate the economic benefit that might be gained by the use of these materials. Two materials are being considered for the binder: copper, second to tungsten in its shielding ability, and iron (or stainless steel), the common fusion reactor shield heavy constituent

  3. 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)

  4. Neutron activation of building materials used in the reactor shield

    International Nuclear Information System (INIS)

    Hernandez, A.T.; Perez, G.; D'Alessandro, K.

    1993-01-01

    Cuban concretes and their main components (mineral aggregates and cement) were investigated through long-lived activation products induced by neutrons from a reactor. The multielemental content in the materials studied was obtained by neutron activation analysis in an IBR-2 reactor and gamma activation analysis in an MT-25 microtron from Join Institute of Nuclear Research of Dubna. After irradiation of building materials for 30 years by a neutron flow of unitary density, induced radioactivity was calculated according to experimental data. The comparative evaluation of different concretes aggregates and two types of cement related to the activation properties is discussed

  5. Status of reactor shielding research in the United States

    International Nuclear Information System (INIS)

    Bartine, D.E.

    1983-01-01

    Shielding research in the United States continues to place emphasis on: (1) the development and refinement of shielding design calculational methods and nuclear data; and (2) the performance of confirmation experiments, both to evaluate specific design concepts and to verify specific calculational techniques and input data. The successful prediction of the radiation levels observed within the now-operating Fast Flux Test Facility (FFTF) has demonstrated the validity of this two-pronged approach, which has since been applied to US fast breeder reactor programs and is now being used to determine radiation levels and possible further shielding needs at operating light water reactors, especially under accident conditions. A similar approach is being applied to the back end of the fission fuel cycle to verify that radiation doses at fuel element storage and transportation facilities and within fuel reprocessing plants are kept at acceptable levels without undue economic penalties

  6. 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.

  7. Radiation resistance of concrete of nuclear reactor vessel

    International Nuclear Information System (INIS)

    Belyakov, V.V.; Denisov, A.V.; Korenevskij, V.V.; Muzalevskij, L.P.; Dubrovskij, V.B.; Ivanov, D.A.; Nazarov, I.L.; Sashin, N.L.

    1992-01-01

    Results of calculational-experimental determination of radiation resistance for concrete bases on limestone gravel and quartz sand, which are the most perspective materials for manufacturing prestressed concrete of the VG-400 reactor vessel are considered. Material samples under investigation were irradiated in the channels of the IBR-2 research reactor for the purpose of the calcultional result verification

  8. Heavy concrete exerting shielding effects particularly against gamma radiation

    International Nuclear Information System (INIS)

    Valenta, D.; Oravec, J.; Racek, M.

    1990-01-01

    The heavy concrete contains synthetic iron(III) oxide in amounts of 5 to 100% with respect to the aggregate content. The oxide has smooth grains, no more than 4 mm in size. The remaining aggregate has grains up to 32 mm in size and a specific weight of 3500 to 5200 kg.m -3 . The remaining concrete components are cement, water and plasticizer. The mixture is homogeneous and is well suited to feeding by means of concrete pumps. (M.D.)

  9. 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.)

  10. Operating manual for the Bulk Shielding Reactor

    International Nuclear Information System (INIS)

    1983-04-01

    The BSR is a pool-type reactor. It has the capabilities of continuous operation at a power level of 2 MW or at any desired lower power level. This manual presents descriptive and operational information. The reactor and its auxillary facilities are described from physical and operational viewpoints. Detailed operating procedures are included which are applicable from source-level startup to full-power operation. Also included are procedures relative to the safety of personnel and equipment in the areas of experiments, radiation and contamination control, emergency actions, and general safety. This manual supercedes all previous operating manuals for the BSR

  11. Operating manual for the Bulk Shielding Reactor

    International Nuclear Information System (INIS)

    1987-03-01

    The BSR is a pool-type reactor. It has the capabilities of continuous operation at a power level of 2 MW or at any desired lower power level. This manual presents descriptive and operational information. The reactor and its auxiliary facilities are described from physical and operational viewpoints. Detailed operating procedures are included which are applicable from source-level startup to full-power operation. Also included are procedures relative to the safety of personnel and equipment in the areas of experiments, radiation and contamination control, emergency actions, and general safety. This manual supersedes all previous operating manuals for the BSR

  12. Operating manual for the Bulk Shielding Reactor

    Energy Technology Data Exchange (ETDEWEB)

    1987-03-01

    The BSR is a pool-type reactor. It has the capabilities of continuous operation at a power level of 2 MW or at any desired lower power level. This manual presents descriptive and operational information. The reactor and its auxiliary facilities are described from physical and operational viewpoints. Detailed operating procedures are included which are applicable from source-level startup to full-power operation. Also included are procedures relative to the safety of personnel and equipment in the areas of experiments, radiation and contamination control, emergency actions, and general safety. This manual supersedes all previous operating manuals for the BSR.

  13. Operating manual for the Bulk Shielding Reactor

    Energy Technology Data Exchange (ETDEWEB)

    1983-04-01

    The BSR is a pool-type reactor. It has the capabilities of continuous operation at a power level of 2 MW or at any desired lower power level. This manual presents descriptive and operational information. The reactor and its auxillary facilities are described from physical and operational viewpoints. Detailed operating procedures are included which are applicable from source-level startup to full-power operation. Also included are procedures relative to the safety of personnel and equipment in the areas of experiments, radiation and contamination control, emergency actions, and general safety. This manual supercedes all previous operating manuals for the BSR.

  14. Loss of coolant analysis for the tower shielding reactor 2

    International Nuclear Information System (INIS)

    Radcliff, T.D.; Williams, P.T.

    1990-06-01

    The operational limits of the Tower Shielding Reactor-2 (TSR-2) have been revised to account for placing the reactor in a beam shield, which reduces convection cooling during a loss-of-coolant accident (LOCA). A detailed heat transfer analysis was performed to set operating time limits which preclude fuel damage during a LOCA. Since a LOCA is survivable, the pressure boundary need not be safety related, minimizing seismic and inspection requirements. Measurements of reactor component emittance for this analysis revealed that aluminum oxidized in water may have emittance much higher than accepted values, allowing higher operating limits than were originally expected. These limits could be increased further with analytical or hardware improvements. 5 refs., 7 figs

  15. Shield materials recommended for space power nuclear reactors

    Science.gov (United States)

    Kaszubinski, L. J.

    1973-01-01

    Lithium hydride is recommended for neutron attenuation and depleted uranium is recommended for gamma ray attenuation. For minimum shield weights these materials must be arranged in alternate layers to attenuate the secondary gamma rays efficiently. In the regions of the shield near the reactor, where excessive fissioning occurs in the uranium, a tungsten alloy is used instead. Alloys of uranium such as either the U-0.5Ti or U-8Mo are available to accommodate structural requirements. The zone-cooled casting process is recommended for lithium hydride fabrication. Internal honeycomb reinforcement to control cracks in the lithium hydride is recommended.

  16. 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.

  17. 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.

  18. Shielding considerations for advanced space nuclear reactor systems

    International Nuclear Information System (INIS)

    Angelo, J.P. Jr.; Buden, D.

    1982-01-01

    To meet the anticipated future space power needs, the Los Alamos National Laboratory is developing components for a compact, 100 kW/sub e/-class heat pipe nuclear reactor. The reactor uses uranium dioxide (UO 2 ) as its fuel, and is designed to operate around 1500 k. Heat pipes are used to remove thermal energy from the core without the use of pumps or compressors. The reactor heat pipes transfer mal energy to thermoelectric conversion elements that are advanced versions of the converters used on the enormously successful Voyager missions to the outer planets. Advanced versions of this heat pipe reactor could also be used to provide megawatt-level power plants. The paper reviews the status of this advanced heat pipe reactor and explores the radiation environments and shielding requirements for representative manned and unmanned applications

  19. PS buildings : reinforced concrete structure for shielding "bridge" pillar

    CERN Multimedia

    CERN PhotoLab

    1956-01-01

    The PS ring traverses the region between the experimental halls South and North (buildings Nos 150 and 151) under massive bridge-shaped concrete beams. This pillar stands at the S-W end of the structure.

  20. Shielding plug for LMFBR type reactors

    International Nuclear Information System (INIS)

    Hashiguchi, Ko.

    1979-01-01

    Purpose: To enable effective removal of liquid metals deposited, if any, in the gaps between a rotary plug and a fixed plug in LMFBR type reactors. Constitution: A plate incorporated with a heater and capable of projecting in a gap between a rotary plug and a fixed plug, and a scraper connected in perpendicular to it are provided to the rotary plug. Solidified liquid metals such as sodium deposited in the gap are effectively removed by the heating with the heater and the scraping action due to the rotation. (Horiuchi, T.)

  1. Improvement of top shield analysis technology for CANDU 6 reactor

    International Nuclear Information System (INIS)

    Kim, Kyo Yoon; Jin, Young Kwon; Lee, Sung Hee; Moon, Bok Ja; Kim, Yong Il

    1996-07-01

    As for Wolsung NPP unit 1, radiation shielding analysis was performed by using neutron diffusion codes, one-dimensional discrete ordinates code ANISN, and analytical methods. But for Wolsung NPP unit 2, 3, and 4, two-dimensional discrete ordinates code DOT substituted for neutron diffusion codes. In other words, the method of analysis and computer codes used for radiation shielding of CANDU 6 type reactor have been improved. Recently Monte Carlo MCNP code has been widely utilized in the field of radiation physics and other radiation related areas because it can describe an object sophisticately by use of three-dimensional modelling and can adopt continuous energy cross-section library. Nowadays Monte Carlo method has been reported to be competitive to discrete ordinate method in the field of radiation shielding and the former has been known to be superior to the latter for complex geometry problem. However, Monte Carlo method had not been used for radiation streaming calculation in the shielding design of CANDU type reactor. Neutron and gamma radiations are expected to be streamed from calandria through the penetrations to reactivity mechanism deck (R/M deck) because many reactivity control units which are established on R/M deck extend from R/M deck to calandria within penetrations, which are provided by guide tube extensions. More precise estimation of radiation streaming is required because R/M deck is classified as an accessible area where atomic worker can access when necessary. Therefore neutron and gamma dose rates were estimated using MCNP code on the R/M deck in the top shield system of CANDU 6 reactor. 9 tabs., 17 figs., 21 refs. (Author)

  2. The removal of concrete layers from biological shields by microwaves

    International Nuclear Information System (INIS)

    Hills, D.L.

    1989-01-01

    Concrete blocks reinforced with steel bars have been subjected to microwave attack at a frequency of 896 MHz at power levels up to 25 kW. The surface concrete has been explosively removed to the depth of the reinforcement, 10 cm, at a rate of about 2 litres per kWh. Heating was localized around the point of attack, with temperatures up to 300 0 C at the fractured face being attained. A simple mathematical model of the propagation and absorption of micro-waves was used to estimate the temperature rise of concrete at microwave frequencies of 896 wand 2450 MHz, at different power levels with and without the presence of reinforcing bars. This demonstrated that reinforcement is expected to significantly increase the temperature rise in the concrete between the irradiated surface and the reinforcement, and that near-surface heating should be more rapid at the higher frequency. There was reasonable agreement between predicted and observed temperature at the higher power levels. Further desk and laboratory studies are proposed before proceeding to a fullscale practical demolition machine and the requirements for a prototype remotely-operated demonstration system have been identified. This consists of a static generator of high power (at least 50 kW) transmitting microwaves via a steerable waveguide to a remote applicator mounted on a simple three-axis manipulator capable of traversing realistically large concrete test panels

  3. SU-E-T-90: Concrete Forward-Scatter Fractions for Radiotherapy Shielding Applications

    International Nuclear Information System (INIS)

    Tanny, S; Parsai, E

    2014-01-01

    Purpose: There is little instruction within the primary shielding guidance document NCRP 151 for vault designs where the primary beam intercepts the maze. We have conducted a Monte-Carlo study to characterize forward-scattered radiation from concrete barriers with the intent of quantifying what amount of additional shielding outside the primary beam is needed in this situation. Methods: We reproduced our vault in MCNP 5 and simulated spectra obtained from the literature and from our treatment planning system for 10 and 18 MV beams. Neutron and gamma-capture contributions were not simulated. Energy deposited was scored at isocenter in a water phantom, within various cells that comprised the maze, and within cells that comprised the vault door. Tracks were flagged that scattered from within the maze to the door and their contributions were tallied separately. Three different concrete mixtures found in the literature were simulated. An empirically derived analytic equation was used for comparison, utilizing patient scatter fractions to approximate the scatter from concrete. Results: Our simulated data confirms that maze-scattered radiation is a significant contribution to total photon dose at the door. It contributes between 20-35% of the photon shielding workload. Forward-scatter fractions for concrete were somewhat dependent on concrete composition and the relative abundance of higher-Z elements. Scatter fractions were relatively insensitive to changes in the primary photon spectrum. Analytic results were of the same magnitude as simulated results. Conclusions: Forward-scattered radiation from the maze barrier needs to be included in the photon workload for shielding calculations in non-standard vault designs. Scatter fractions will vary with concrete composition, but should be insensitive to spectral changes between machine manufacturers. Further plans for investigation include refined scatter fractions for various concrete compositions, scatter fraction

  4. SU-E-T-90: Concrete Forward-Scatter Fractions for Radiotherapy Shielding Applications

    Energy Technology Data Exchange (ETDEWEB)

    Tanny, S; Parsai, E [University of Toledo Medical Center, Toledo, OH (United States)

    2014-06-01

    Purpose: There is little instruction within the primary shielding guidance document NCRP 151 for vault designs where the primary beam intercepts the maze. We have conducted a Monte-Carlo study to characterize forward-scattered radiation from concrete barriers with the intent of quantifying what amount of additional shielding outside the primary beam is needed in this situation. Methods: We reproduced our vault in MCNP 5 and simulated spectra obtained from the literature and from our treatment planning system for 10 and 18 MV beams. Neutron and gamma-capture contributions were not simulated. Energy deposited was scored at isocenter in a water phantom, within various cells that comprised the maze, and within cells that comprised the vault door. Tracks were flagged that scattered from within the maze to the door and their contributions were tallied separately. Three different concrete mixtures found in the literature were simulated. An empirically derived analytic equation was used for comparison, utilizing patient scatter fractions to approximate the scatter from concrete. Results: Our simulated data confirms that maze-scattered radiation is a significant contribution to total photon dose at the door. It contributes between 20-35% of the photon shielding workload. Forward-scatter fractions for concrete were somewhat dependent on concrete composition and the relative abundance of higher-Z elements. Scatter fractions were relatively insensitive to changes in the primary photon spectrum. Analytic results were of the same magnitude as simulated results. Conclusions: Forward-scattered radiation from the maze barrier needs to be included in the photon workload for shielding calculations in non-standard vault designs. Scatter fractions will vary with concrete composition, but should be insensitive to spectral changes between machine manufacturers. Further plans for investigation include refined scatter fractions for various concrete compositions, scatter fraction

  5. RADIO SHIELDING PROPERTIES OF CONCRETE BASED ON SHUNGITE NANOMATERIALS

    Directory of Open Access Journals (Sweden)

    BELOUSOVA Elena Sergeevna

    2013-04-01

    Full Text Available Modifications of shielding construction materials based on Portland cement with the addition of powder nanomaterial shungite were developed. Attenuation and re­flection of electromagnetic radiation for obtained materials were studied. Recommen­dations for using are given.

  6. An Analysis of Radiation Penetration through the U-Shaped Cast Concrete Joints of Concrete Shielding in the Multipurpose Gamma Irradiator of BATAN

    Science.gov (United States)

    Ardiyati, Tanti; Rozali, Bang; Kasmudin

    2018-02-01

    An analysis of radiation penetration through the U-shaped joints of cast concrete shielding in BATAN’s multipurpose gamma irradiator has been carried out. The analysis has been performed by calculating the radiation penetration through the U-shaped joints of the concrete shielding using MCNP computer code. The U-shaped joints were a new design in massive concrete construction in Indonesia and, in its actual application, it is joined by a bonding agent. In the MCNP simulation model, eight detectors were located close to the observed irradiation room walls of the concrete shielding. The simulation results indicated that the radiation levels outside the concrete shielding was less than the permissible limit of 2.5 μSv/h so that the workers could safely access electrical room, control room, water treatment facility and outside irradiation room. The radiation penetration decreased as the density of material increased.

  7. Testing of plain and fibrous concrete single cavity prestressed concrete reactor vessel models

    International Nuclear Information System (INIS)

    Oland, C.B.

    1985-01-01

    Two single-cavity prestressed concrete reactor vessel (PCRV) models were fabricated and tested to failure to demonstrate the structural response and ultimate pressure capacity of models cast from high-strength concretes. Concretes with design compressive strengths in excess of 70 MPa (10,000 psi) were developed for this investigation. One model was cast from plain concrete and failed in shear at the head region. The second model was cast from fiber reinforced concrete and failed by rupturing the circumferential prestressing at the sidewall of the structure. The tests also demonstrated the capabilities of the liner system to maintain a leak-tight pressure boundary. 3 refs., 4 figs

  8. 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

  9. Efficiency of steel-concrete compositions in a side shielding of high-energy proton accelerators

    International Nuclear Information System (INIS)

    Getmanov, V.B.; Kryuchkov, V.P.; Lebedev, V.N.

    1983-01-01

    Aiming at the study of efficiency of application of heavy concretes with the density up to 6.3 g/cm -3 with iron-ore aggregate and steel scrap with shot the calculational study on high-energy radiation attenuation in the accelerator side shield has been carried out. The calculation is made for five concretes with the density 2.38; 3.66; 4.68; 5.34; 6.30 g x cm -3 and for pure iron. The real chemical composition of each concrete, including hydrogen, is taken into account. The real spectrum of hadron generated in the materiai of evacuated ionguide wall under the effect of the 70 GeV proton beam incident on the wall at a narrow angle THETA -3 ensuring the same ratio of the dose or hadron fluence with the energy > 20 MeV attenuation is accepted as a relative shield efficiency of the material. It is shown, that for steel-concrete compositions with the density > 5.6 gxcm -3 the relative shield efficiency decreases sharply. It is also shown, that aplication of concretes with the density 3.6-3.7 gxcm -3 is expedient and economically profitable

  10. Aggregate effects on γ-ray shielding characteristics and compressive strength on concrete

    International Nuclear Information System (INIS)

    Oh, Jeong Hwan; Choi, Soo Seok; Mun, Young Bun; Lee, Jae Hyung; Choi, Hyun Kook

    2016-01-01

    We observed the γ-ray shielding characteristics and compressive strength of five types of concrete using general aggregates and high-weight aggregates. The aggregates were classified into fine aggregate and coarse aggregate according to the average size. The experimental results obtained an attenuation coefficient of 0.371 cm-1 from a concrete with the oxidizing slag sand (OSS) and oxidizing slag gravel (OSG) for a γ-ray of "1"3"7Cs, which is improved by 2% compared with a concrete with typical aggregates of sand and gravel. In the unit weight measurement, a concrete prepared by iron ore sand (IOS) and OSG had the highest value of 3,175 kg·m"-"3. Although the unit weight of the concrete with OSS and OSG was 3,052 kg·m"-"3, which was lower than the maximum unit weight condition by 123 kg·m"-"3, its attenuation coefficient was improved by 0.012 cm-1. The results of chemical analysis of aggregates revealed that the magnesium content in oxidizing slag was lower than that in iron ore, while the calcium content was higher. The concrete with oxidizing slag aggregates demonstrated enhanced γ-ray shielding performance due to a relatively high calcium content compared with the concrete with OSS and OSG in spite of a low unit weight. All sample concretes mixed with high-weight aggregates had higher compressive strength than the concrete with typical sand and gravel. When OSS and IOS were used, the highest compressive strength was 50.2 MPa, which was an improvement by 45% over general concrete, which was achieved after four weeks of curing

  11. Aggregate effects on γ-ray shielding characteristics and compressive strength on concrete

    Energy Technology Data Exchange (ETDEWEB)

    Oh, Jeong Hwan; Choi, Soo Seok [Jeju National University, Jeju (Korea, Republic of); Mun, Young Bun; Lee, Jae Hyung; Choi, Hyun Kook [Sungshin Cement Co., Ltd, Sejong (Korea, Republic of)

    2016-12-15

    We observed the γ-ray shielding characteristics and compressive strength of five types of concrete using general aggregates and high-weight aggregates. The aggregates were classified into fine aggregate and coarse aggregate according to the average size. The experimental results obtained an attenuation coefficient of 0.371 cm-1 from a concrete with the oxidizing slag sand (OSS) and oxidizing slag gravel (OSG) for a γ-ray of {sup 137}Cs, which is improved by 2% compared with a concrete with typical aggregates of sand and gravel. In the unit weight measurement, a concrete prepared by iron ore sand (IOS) and OSG had the highest value of 3,175 kg·m{sup -3}. Although the unit weight of the concrete with OSS and OSG was 3,052 kg·m{sup -3}, which was lower than the maximum unit weight condition by 123 kg·m{sup -3}, its attenuation coefficient was improved by 0.012 cm-1. The results of chemical analysis of aggregates revealed that the magnesium content in oxidizing slag was lower than that in iron ore, while the calcium content was higher. The concrete with oxidizing slag aggregates demonstrated enhanced γ-ray shielding performance due to a relatively high calcium content compared with the concrete with OSS and OSG in spite of a low unit weight. All sample concretes mixed with high-weight aggregates had higher compressive strength than the concrete with typical sand and gravel. When OSS and IOS were used, the highest compressive strength was 50.2 MPa, which was an improvement by 45% over general concrete, which was achieved after four weeks of curing.

  12. Waterproofing shielding for concrete in wet and dry storage

    International Nuclear Information System (INIS)

    Gorin, N.; Scherbina, A.; Urusov, S.

    2007-01-01

    One of main reliability and safety criteria for constructions, designed for wet and dry storage of radioactive materials and waste, is the long-term ability to maintain the waterproofing properties in the conditions of high radiation load. The base structural material of these constructions is concrete (cooling ponds, different storage for spent nuclear fuel and waste, etc.). The provision of reliable concrete waterproofing is very important for decreasing risks of radioactive substances ingress to environment and moisture penetration to objects from outside, and also for construction life extension. In the process of long-term operation, some concrete constructions, erected already few decades ago, are gradually losing their waterproofing and this circumstance involves severe operational and ecological threats. Therefore advanced effective concrete waterproofing technologies both for erection of new objects and for repairing of operating constructions are in extreme demand. The paper is devoted to the solution of this problem proposed by Russian Federal Nuclear Centre (RFNC-VNIITF, Snezhinsk). The paper contains the developed criteria established for the search for optimal materials, the 'integral capillary systems' (ICS) principal of operation, methods and results of the tests, and also the experience of ICS application on real objects. (author)

  13. Integrated thermal analysis of top-shield and reactor vault of Indian FBR-600

    International Nuclear Information System (INIS)

    Rajendrakumar, M.; Velusamy, K.; Selvaraj, P.

    2015-01-01

    The design for next generation fast breeder reactors (FBR-600) has been commenced with enhanced safety and improved economy as the main targets. The Top Shield (TS) of Prototype Fast Breeder Reactor (PFBR) is a box type structure consisting of Roof Slab (RS), Small Rotatable Plug (SRP), and Large Rotatable Plug (LRP). The large box type structure with many penetrations posed difficulties during manufacturing. Because of the required high load carrying capabilities, a dome shaped thick plate roof slab is conceived for FBR-600. Main Vessel (MV) which holds the primary sodium and associated components is welded to the RS through a triple joint. Reactor vault (RV) is a thick concrete structure which supports MV and Safety Vessel (SV). The temperature of RV concrete has to be less than 338 K (65°C) under normal operating heat loads (full and part load conditions) and less than 363 K (90°C) under Safety Grade Decay Heat Removal (SGDHR) conditions with one cooling loop in service. The temperature in the component penetrations of the RS should be greater than 120°C to avoid sodium aerosol deposition. Similarly, the temperature of the LRP and SRP has to be ∼120°C to protect the elastomeric seals provided to these structures. Further, the heat load to RV transferred by direct conduction by roof slab support has to be minimum. To meet these conflicting thermal requirements, detailed multi-physics CFD calculations have been performed to finalize, (i) the insulation requirements on the top of roof slab, (ii) number and position of reflective insulation plates below the bottom plate of roof slab/rotating plugs, (iii) air flow rate for various zones of the top shield and (iv) water flow rate and pitch of water cooling pipes for the reactor vault. (author)

  14. Large-scale experiment with laying shielding concrete at Dukovany nuclear power plant

    International Nuclear Information System (INIS)

    Hoenig, A.; Svoboda, R.; Rosa, J.

    1984-01-01

    In some places the concrete walls of the biological shielding are so thin that it is not possible to control the density of the deposited concrete. An experiment was therefore carried out which was to demonstrate that concrete may be deposited by sinking through concrete tubes or by a concrete pump from a height of 8 metres. Two test walls A and B were concreted using the non-standardized method and the third wall was used as the standard. The following tests were conducted on the two non-standardized walls: test of miscibility of extra-heavy concrete, ultrasonic test of homogeneity, and samples were taken for tests of density. Density was determined radiometrically using a narrow gamma beam. Statistical evaluation of the results showed that the homogeneity of density of the concrete was the best in the standard wall, in walls A and B the variation coefficient did not exceed 8 per mille. An exception was made to the rule and concrete with a max. 16 mm grain size was deposited from the height of 8 m on condition of strict observance of production technology. (J.P.)

  15. Review of analysis methods for prestressed concrete reactor vessels

    International Nuclear Information System (INIS)

    Dodge, W.G.; Bazant, Z.P.; Gallagher, R.H.

    1977-02-01

    Theoretical and practical aspects of analytical models and numerical procedures for detailed analysis of prestressed concrete reactor vessels are reviewed. Constitutive models and numerical algorithms for time-dependent and nonlinear response of concrete and various methods for modeling crack propagation are discussed. Published comparisons between experimental and theoretical results are used to assess the accuracy of these analytical methods

  16. Mechanical properties of concrete for power reactor at high temperatures

    International Nuclear Information System (INIS)

    Kawase, Kiyotaka; Tanaka, Hitoshi; Nakano, Masayuki

    1985-01-01

    The purpose of this study is to investigate the mechanical properties of concrete for power reactor at high temperature. This paper presents the creep behavior of concrete at high temperature and the cause by which a specified aggregate is broken at a specified high temperature. The creep coefficient at high temperature is smaller than that at ordinary temperature. (author)

  17. Radiometric determination of density of fresh shielding concrete (in situ) in the nuclear industry

    International Nuclear Information System (INIS)

    Honig, A.

    1985-01-01

    Methods of radiometric determination of density have been in recent years elaborated in detail and successfully. But on the market no instruments are available for measuring fresh concrete when it is possible to repair inhomogeneities, if any, even before hardening, and thus to guarantee safety of biological protection of nuclear reactors. The paper describes an analog and digital radiation density meter and their application in the inspection of radiation protection concrete walls. By repairing defective, insufficiently dense locations still in the course of concrete placement it is possible to attain a laboratory quality of the concrete even under on-site conditions

  18. Tools and applications for core design and shielding in fast reactors

    International Nuclear Information System (INIS)

    Rachamin, Reuven

    2013-01-01

    Outline: • Modeling of SFR cores using the Serpent-DYN3D code sequence; • Core shielding assessment for the design of FASTEF-MYRRHA; • Neutron shielding studies on an advanced Molten Salt Fast Reactor (MSFR) design

  19. 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.

  20. Quality control in high thickness concrete walls for shielding

    International Nuclear Information System (INIS)

    Arcama, J.A.; San Pedro, Marcelo; Cannistracci, C.A.

    1983-01-01

    After evaluating different methods of non-destructive testing, of fast execution and quick results, with low operative cost, and suitable to verify the homogeneity and the shielding power of the walls of process cells for radiochemical use, under construction in the Centro Atomico Ezeiza, it was decided to employ the ultrasound method over the whole surface to be examined, with subsequent verification of the results on isolated zones by means of radiometry and gammagraphy. This procedure proved to be satisfactory. The cell's characteristics, the tests performed and their results, which were statistically evaluated by means of a computer program, implemented to his effect, are described. (C.A.K.) [es

  1. Comparison of calculational methods for liquid metal reactor shields

    International Nuclear Information System (INIS)

    Carter, L.L.; Moore, F.S.; Morford, R.J.; Mann, F.M.

    1985-09-01

    A one-dimensional comparison is made between Monte Carlo (MCNP), discrete ordinances (ANISN), and diffusion theory (MlDX) calculations of neutron flux and radiation damage from the core of the Fast Flux Test Facility (FFTF) out to the reactor vessel. Diffusion theory was found to be reasonably accurate for the calculation of both total flux and radiation damage. However, for large distances from the core, the calculated flux at very high energies is low by an order of magnitude or more when the diffusion theory is used. Particular emphasis was placed in this study on the generation of multitable cross sections for use in discrete ordinates codes that are self-shielded, consistent with the self-shielding employed in the generation of cross sections for use with diffusion theory. The Monte Carlo calculation, with a pointwise representation of the cross sections, was used as the benchmark for determining the limitations of the other two calculational methods. 12 refs., 33 figs

  2. Comparative study of the shield of concrete blocks with hematite in relation to common concrete blocks

    International Nuclear Information System (INIS)

    Costa, Paulo R.; Buerger, Andre A.; Naccache, Veronica K.; Priszkulnik, Simao

    2012-01-01

    The present work shows results of an empirical evaluation of the transmission properties of two radioprotection materials: an ordinary concrete and an ordinary concrete mixed with hematite. It was used techniques of x-ray spectroscopy and measurements of the air-kerma transmitted through these two materials in order to compare the transmission properties for each one. (author)

  3. Heavy density concrete for nuclear radiation shielding and power stations: [Part]3

    International Nuclear Information System (INIS)

    Singha Roy, P.K.

    1987-01-01

    This article is the third part of the paper entitled 'Heavy density concrete for nuclear radiation shielding and power stations'. Specific considerations relevant to natural but manufactured heavy aggregates like haematite used in India are briefly discussed. They include water-cement ratio, strength versus water-cement ratio, mix design strength and aggregate grading. Some typical mix proportions in haematite concretes used in India are given. Equipment for heavy density concrete is mentioned. Quality control methods and tests for heavy density concrete are described under the heading: type and chemical composition of the rock, specific gravity and surface absorption of the aggregates, grading of aggregates, cement, batching, mixing, compressive strength, and density. Construction aspects such as form work, placement, vibration, finishing, and temperature control are discussed. Finally it is pointed out that for optimising the design and economy of heavy density concrete, it is necessary to carry out country-wide survey of suitable materials, to study their properties, suitability and effectiveness in shielding radiation. (M.G.B.)

  4. Dry storage technologies: keys to choosing among metal casks, concrete shielded steel canister modules and vaults

    International Nuclear Information System (INIS)

    Roland, V.; Solignac, Y.; Chiguer, M.; Guenon, Y.

    2003-01-01

    time. Then the key criterion is maximum modularity. Furthermore, the up front capital costs requirement for this type of solution is minimal, so depending on the chosen discount rate of the investor, they have an additional attraction. Those smaller modules allow to change course in back end policy more easily. Priority of modularity yields two other solutions, dual-purpose metal casks of the TN24TM family or dual purpose or single purpose concrete shielded welded canisters such as NUHOMS. These solutions, implemented by COGEMA LOGISTICS, TRANSNUCLEAR Inc. and FRAMATOME-ANP, are very flexible and have been adapted also to quite different fuels. Among what influences the choice, we can consider: in favor of metal casks (minimal ancillary equipment, ready to move to final or centralized repository or reprocessing or other ISFSI, compact systems, easy rearrangement, easy handling), in favor of concrete shielded canisters based systems (economics when initial quantity is sufficient to spread out up front equipment, significant cost-shielding advantage, easy local production of the relatively light canisters). Both approaches, when transportable, are also a factor for public acceptance because of the non-permanent characteristics and because transport licensing refers to internationally recognized rules, standards and methods. (authors)

  5. Method of shielding a liquid-metal-cooled reactor

    International Nuclear Information System (INIS)

    Sayre, R.K.

    1978-01-01

    The primary heat transport system of a nuclear reactor - particularly for a liquid-metal-cooled fast-breeder reactor - is shielded and protected from leakage by establishing and maintaining a bed of a powdered oxide closely and completely surrounding all components thereof by passing a gas upwardly therethrough at such a rate as to slightly expand the bed to the extent that the components of the system are able to expand without damage and yet the particles of a the bed remain close enough so that the bed acts as a guard vessel for the system. Preferably the gas contains 1 to 10% oxygen and the gas is passed upwardly through the bed at such a rate that the lower portion of the bed is a fixed bed while the upper portion is a fluidized bed, the line of demarcation therebetween being high enough that the fixed bed portion of the bed serves as guard vessel for the system

  6. Improved Nuclear Reactor and Shield Mass Model for Space Applications

    Science.gov (United States)

    Robb, Kevin

    2004-01-01

    New technologies are being developed to explore the distant reaches of the solar system. Beyond Mars, solar energy is inadequate to power advanced scientific instruments. One technology that can meet the energy requirements is the space nuclear reactor. The nuclear reactor is used as a heat source for which a heat-to-electricity conversion system is needed. Examples of such conversion systems are the Brayton, Rankine, and Stirling cycles. Since launch cost is proportional to the amount of mass to lift, mass is always a concern in designing spacecraft. Estimations of system masses are an important part in determining the feasibility of a design. I worked under Michael Barrett in the Thermal Energy Conversion Branch of the Power & Electric Propulsion Division. An in-house Closed Cycle Engine Program (CCEP) is used for the design and performance analysis of closed-Brayton-cycle energy conversion systems for space applications. This program also calculates the system mass including the heat source. CCEP uses the subroutine RSMASS, which has been updated to RSMASS-D, to estimate the mass of the reactor. RSMASS was developed in 1986 at Sandia National Laboratories to quickly estimate the mass of multi-megawatt nuclear reactors for space applications. In response to an emphasis for lower power reactors, RSMASS-D was developed in 1997 and is based off of the SP-100 liquid metal cooled reactor. The subroutine calculates the mass of reactor components such as the safety systems, instrumentation and control, radiation shield, structure, reflector, and core. The major improvements in RSMASS-D are that it uses higher fidelity calculations, is easier to use, and automatically optimizes the systems mass. RSMASS-D is accurate within 15% of actual data while RSMASS is only accurate within 50%. My goal this summer was to learn FORTRAN 77 programming language and update the CCEP program with the RSMASS-D model.

  7. Rapid detailed characterization of concrete shielding blocks utilizing internal natural radionuclides for calibration

    International Nuclear Information System (INIS)

    McDonald, R.J.; Smith, A.R.; Norman, E.B.; Cowles, D.

    1995-10-01

    Following many years of productive work, the SuperHILAC and Bevalac accelerators at Lawrence Berkeley National Laboratory were closed, leaving thousands of concrete shielding blocks available for reuse or disposal. The process history of these blocks as shielding precludes free release pending radiological characterization. This paper presents a method for the rapid characterization of gamma-ray-emitting radioisotopes in large samples of earth-like materials: concrete shielding blocks in this case. Active regions are identified with a sensitive radiation-survey instrument and then examined in detail with a high-efficiency lead-shielded Ge spectrometer. Naturally-occurring gamma-ray emissions from the decays of uranium, thorium, and potassium are used to calibrate the spectrometer. A simple relationship exists between the observed counting rate in a characteristic gamma ray and the activity in the block. This method, taking only tens of minutes per sample at the nano-Curie/gram sensitivity level, replaces much of the expensive coring and laboratory analysis methods needed otherwise

  8. 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.

  9. PKI, Gamma Radiation Reactor Shielding Calculation by Point-Kernel Method

    International Nuclear Information System (INIS)

    Li Chunhuai; Zhang Liwu; Zhang Yuqin; Zhang Chuanxu; Niu Xihua

    1990-01-01

    1 - Description of program or function: This code calculates radiation shielding problem of gamma-ray in geometric space. 2 - Method of solution: PKI uses a point kernel integration technique, describes radiation shielding geometric space by using geometric space configuration method and coordinate conversion, and makes use of calculation result of reactor primary shielding and flow regularity in loop system for coolant

  10. Prestressed concrete reactor vessels: review of design and failure criteria

    International Nuclear Information System (INIS)

    Endebrock, E.G.

    1975-03-01

    The design and failure criteria of prestressed concrete reactor vessels (PCRVs) are reviewed along with the analysis methods. The mechanical properties of concrete under multiaxial stresses are not adequately quantified or described to permit an accurate analysis of a PCRV. Structural analysis of PCRVs almost universally utilizes a finite element which encounters difficulties in numerical solution of the governing equations and in treatment of fractured elements. (U.S.)

  11. Biological shielding design and qualification of concreting process for construction of electron beam irradiation facility

    International Nuclear Information System (INIS)

    Petwal, V.C.; Kumar, P.; Suresh, N.; Parchani, G.; Dwivedi, J.; Thakurta, A.C.

    2011-01-01

    A technology demonstration facility for irradiation of food and agricultural products is being set-up by RRCAT at Indore. The facility design is based on linear electron accelerator with maximum beam power of 10 kW and can be operated either in electron mode at 10 MeV or photon modes at 5/7.5 MeV. Biological shielding has been designed in accordance with NCRP 51 to achieve dose rate at all accessible points outside the irradiation vault less than the permissible limit of 0.1 mR/hr. In addition to radiation attenuation property, concrete must have satisfactory mechanical properties to meet the structural requirements. There are number of site specific variables which affect the structural, thermal and radiological properties of concrete, leading to considerable difference in actual values and design values. Hence it is essential to establish a suitable site and environmental specific process to cast the concrete and qualify the process by experimental measurement. For process qualification we have cast concrete test blocks of different thicknesses up to 3.25 m and evaluated the radiological and mechanical properties by radiometry, ultrasonic and mechanical tests. In this paper we describe the biological shielding design of the facility and analyse the results of tests carried out for qualification of the process. (author)

  12. Requirements for thermal insulation on prestressed concrete reactor vessels

    International Nuclear Information System (INIS)

    Neylan, A.J.; Wistrom, J.D.

    1979-01-01

    During the past decade, extensive design, construction, and operating experience on concrete pressure vessels for gas-cooled reactor applications has accumulated. Excellent experience has been obtained to date on the structural components (concrete, prestressing systems, liners, penetrations, and closures) and the thermal insulation. Three fundamentally different types of insulation systems have been employed to ensure the satisfactory performance of this component, which is critical to the overall success of the prestressed concrete reactor vessel (PCRV). Although general design criteria have been published, the requirements for design, materials, and construction are not rigorously addressed in any national or international code. With the more onerous design conditions being imposed by advanced reactor systems, much greater attention has been directed to advance the state of the art of insulation systems for PCRVs. This paper addresses some of the more recent developments in this field being performed by General Atomic Company and others. (author)

  13. Effect of elevated temperatures on heavy concrete structural strength in Qinshan phase 3 CANDU 6 reactor buildings

    International Nuclear Information System (INIS)

    Alikhan, S.; Khan, A.F.; Chen, S.

    2005-01-01

    Heavy concrete is commonly used inside the Qinshan Phase 3 CANDU 6 reactor buildings for radiation shielding functions in order to provide access to key areas during reactor operation. In some cases, the heavy concrete elements are also structural elements. Concerns have been raised about the functional performance of the heavy concrete structural elements, specifically the primary heat transport pump (PHTS) supporting slabs, surrounding the feeder cabinets when subjected to elevated temperatures between 42 degree C and 121 degree C and their corresponding temperature gradients on a long-term basis during the normal operation of the plant. This paper presents the results of a test investigation on the strength of heavy concrete under elevated temperature conditions being experienced by the heavy concrete structural elements around the feeder cabinet to confirm that these structural elements meet their functional requirements. The loading conditions consist subjecting the specimens to the elevated temperatures and temperature gradient noted during commissioning, including the effect of epoxy coating. The heavy concrete mix proportion and materials of the test samples (ilmenite aggregate and Portland cement) are identical to those used for heavy concrete structural elements surrounding the feeder cabinet. Subsequent to the confirmation of the functional requirements of the heavy concrete structural elements, alarm limits are recommended for these structural elements. (authors)

  14. Study of local Agregate for Gamma radiation concrete shield; Studi pemakaian Agregat lokal pada pembuatan beton perisai radiasi Gamma

    Energy Technology Data Exchange (ETDEWEB)

    Tochrul-Binowo,; Endro-Kismolo,; Darsono, [Yogyakarta Nuclear Research Centre, National Atomic Energy Agency, Yogyakarta (Indonesia)

    1996-04-15

    Investigation on the composition of gamma radiation concrete shield made of local barite, manganese fine and coarse aggregates from Kulon Progo, Yogyakarta has been done. The purpose of the research was to find out the quality of these local material for an aggregate of gamma radiation concrete shield. The research was done where each mineral was used as coarse aggregate and the fine aggregate from Kulon Progo was used as fine basic aggregate. Firstly a normal concrete was made by mixing cement, fine aggregate, coarse aggregate and water at a weight ratio of cement: fine aggregate: coarse: water 1: 2.304: 3.456: 0.58. The gamma radiation absorption capacity of the concrete tested by using Cs-137 as source standard. The same method was done on barite concrete at the weight ratio of cement: fine aggregate: barite aggregate: water 1: 2.303: 3.456: 0.58 and manganese concrete at the weight ratio of cement: fine aggregate: manganese aggregate: and water 1: 1.896: 2.844: 0.58. The result of the study showed that the gamma radiation absorption capacity of barite aggregate was greater than that of normal concrete and manganese concrete. The coefficient linear attenuation (for 6.0 cm thickness) of each concrete were {mu} barite concrete = 0.23071 cm{sup -1}, {mu} manganese concrete = 0.08401 cm{sup -1} and {mu} normal concrete = 0.1669 cm{sup -1}.

  15. 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

  16. Shield design and streaming calculations for the sodium cooled PEC reactor

    International Nuclear Information System (INIS)

    Prosperi, M.; Tavoni, R.; Travaglini, N.

    1977-01-01

    This paper summarises the shielding calculations carried out for the PEC reactor. A brief description of calculation methods and of the work carried out to set them up is given; the most representative calculations with the relative isoflux curves are also referred. A general outline is then given for the main shielding problems of the PEC reactor

  17. Concrete Shielding For Radiation Safety And Unexpected Dangerous Inside Cobalt-60 Industrial Irradiator

    International Nuclear Information System (INIS)

    Keshk, A.B.; Aly, R.A.

    2011-01-01

    The study shows a proposed destruction inside one of three cobalt-60 industrial irradiators to determine and reduce the negative results, to improve and modify emergency plan to face terrorism works. The results show the performance of concrete shielding (walls and ceiling) contains the bad effect of dynamic pressures. The explosion forces are prevented to destructive by performance of their concrete shielding, which will contain the most components of devastated systems inside each irradiator after explosion. Shield penetration like electrical cable tunnels, pushers holes, hole with removable plug, product boxes openings, lens opening and ozone duct are affected badly by destruction. Through probability of transporting, some of devastated parts of broken radioactive cobalt- 60 pencils from inside radiation concreter room to outside (surrounded environment) are maintained and causing very danger radiation exposure by gamma rays outside irradiator. A necessity needs to modify emergency plan to prevent any explosive materials to enter inside the main building (irradiation sale) and also discovering any explosive materials which are placed inside the product boxes before passing to inside irradiator. The minimizing radiation exposure (2 mrem/h) inside underground radiation shelters are maintained by reducing radiation dose exerted from a nuclear explosion of 20 kT about 1 km away to a safe value, and calculating the protective factors of radiation main building basements are more than 40 (safety factor) as they are located under ground level, are surrounded by sandy soil and are constructed by concrete. The study shows the proposed basements of the main building maintain success to use as under ground safe radiation shelter (during emergency) with separate safe radiation trace. It begins from the main opening of irradiation sale and leads to underground proposed shelter through modified main stair

  18. Space reactor system and subsystem investigations: assessment of technology issues for the reactor and shield subsystem. SP-100 Program

    International Nuclear Information System (INIS)

    Atkins, D.F.; Lillie, A.F.

    1983-01-01

    As part of Rockwell's effort on the SP-100 Program, preliminary assessment has been completed of current nuclear technology as it relates to candidate reactor/shield subsystems for the SP-100 Program. The scope of the assessment was confined to the nuclear package (to the reactor and shield subsystems). The nine generic reactor subsystems presented in Rockwell's Subsystem Technology Assessment Report, ESG-DOE-13398, were addressed for the assessment

  19. Nuclear data needs for fast breeder reactor shielding

    International Nuclear Information System (INIS)

    Oblow, E.M.; Perey, F.G.

    1978-11-01

    A review of neutron and gamma-ray cross section data needs for fast reactor shielding is presented in light of the recent advances made in assessing these needs through sensitivity studies. Total and partial cross sections and energy and angular distribution data for neutrons are surveyed as well as gamma-ray production cross sections. The strengths and deficiencies of currently available benchmark-quality integral experiments are also discussed with respect to their use in creating adjusted cross section libraries for design work. The availability of first round covariance data in ENDF/B-IV and plans for ENDF/B-V are also reviewed. This latter information makes it possible to quantitatively assess the quality of current cross section data libraries and also puts adjustment and data assessment procedures on a firmer basis

  20. The investigation of gamma and neutron shielding properties of concrete including basalt fibre for nuclear energy applications

    International Nuclear Information System (INIS)

    Nulk, H.; Ipbuker, C.; Gulik, V.; Tkaczyk, A.; Biland, A.

    2015-01-01

    In this study, we would like to draw attention to the prospect of basalt fibre as the main component for concrete reinforcement of NPP. This work describes the computational study of gamma attenuation parameters, the effective atomic number Z(eff) and the effective electron density N e (eff), of relatively light-weight concrete with chopped basalt fibre used as reinforcement in different mixture rates. We can draw the following conclusions. Basalt fibre is a relatively cheap material that can be used as reinforcement instead of metallic fibers. Basalt fibre has a similar specific gravity to that of concrete elements. Basalt fibre has high chemical and abrasion resistance. Basalt fibre has almost 10 times the tensile strength of steel re-bars. Gamma-ray attenuation coefficients increase with addition of basalt fibre into concrete in every case. The effective atomic number of the concrete increases with the addition of basalt fibre. The results show that basalt fibre reinforced concrete have improved shielding properties against gamma rays in comparison with regular concrete. This result is based on a regular concrete with only basalt fiber reinforcement. We estimate that with addition of standard aggregates for radiation shielding concrete, such as barite, magnetite or hematite, the shielding properties will increase exponentially

  1. 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.

  2. Heat transfer in reactor cavity during core-concrete interaction

    International Nuclear Information System (INIS)

    Adroguer, B.; Cenerino, G.

    1989-08-01

    In the unlikely event of a severe accident in a nuclear power plant, the core may melt through the vessel and slump into the concrete reactor cavity. The hot mixture of the core material called corium interacts thermally with the concrete basemat. The WECHSL code, developed at K.f.K. Karlsruhe in Germany is used at the Protection and Nuclear Safety Institute (I.P.S.N.) of CEA to compute this molten corium concrete interaction (MCCI). Some uncertainties remain in the partition of heat from the corium between the basemat and the upper surrounding structures in the cavity where the thermal conditions are not computer. The CALTHER code, under development to perform a more mechanistic evaluation of the upward heat flux has been linked to WECHSL-MOD2 code. This new version enables the modelling of the feedback effects from the conditions in the cavity to the MCCI and the computation of the fraction of upward flux directly added to the cavity atmosphere. The present status is given in the paper. Preliminary calculations of the reactor case for silicate and limestone common sand (L.C.S.) concretes are presented. Significant effects are found on concrete erosion, gases release and temperature of the upper part of corium, particularly for L.C.S. concrete

  3. Prediction of the strength of concrete radiation shielding based on LS-SVM

    International Nuclear Information System (INIS)

    Juncai, Xu; Qingwen, Ren; Zhenzhong, Shen

    2015-01-01

    Highlights: • LS-SVM was introduced for prediction of the strength of RSC. • A model for prediction of the strength of RSC was implemented. • The grid search algorithm was used to optimize the parameters of the LS-SVM. • The performance of LS-SVM in predicting the strength of RSC was evaluated. - Abstract: Radiation-shielding concrete (RSC) and conventional concrete differ in strength because of their distinct constituents. Predicting the strength of RSC with different constituents plays a vital role in radiation shielding (RS) engineering design. In this study, a model to predict the strength of RSC is established using a least squares-support vector machine (LS-SVM) through grid search algorithm. The algorithm is used to optimize the parameters of the LS-SVM on the basis of traditional prediction methods for conventional concrete. The predicted results of the LS-SVM model are compared with the experimental data. The results of the prediction are stable and consistent with the experimental results. In addition, the studied parameters exhibit significant effects on the simulation results. Therefore, the proposed method can be applied in predicting the strength of RSC, and the predicted results can be adopted as an important reference for RS engineering design

  4. Measurement of 36Cl induced in shielding concrete of various accelerator facilities

    International Nuclear Information System (INIS)

    Bessho, K.; Matsumura, H.; Matsuhiro, T.

    2003-01-01

    The concentrations of 36 Cl induced in shielding concrete of the various accelerators has been measured by accelerator mass spectrometry. For three kinds of accelerator facilities, SF cyclotron (Center for Nuclear Study, the University of Tokyo), 300 MeV electron LINAC (Laboratory of Nuclear Science, Tohoku University), and 12 GeV proton synchrotron (High Energy Accelerator Research Organization), the depth profiles of 36 Cl/ 35 Cl ratios in concrete samples near the beam lines were analyzed. The depth profiles of 36 Cl/ 35 Cl are consistent with those of the radioactive concentrations of 152 Eu and 60 Co, which are formed by thermal neutron capture reactions. These results imply that 36 Cl formed in shielding concrete of these accelerators is mainly produced by thermal neutron capture of 35 Cl. The maximum 36 Cl/ 35 Cl ratio of 3x10 -8 (300 MeV electron LINAC, depth of 8 cm) corresponds to the specific radioactivity of 2x10 -3 Bq/g, which is not serious for radioactive waste management in reconstruction or decommissioning of accelerator facilities, compared with specific radioactivity of 3 H, 152 Eu and 60 Co. (author)

  5. Reactors with pressure vessel in pre-stressed concrete

    International Nuclear Information System (INIS)

    Devillers, Christian; Lafore, Pierre

    1964-12-01

    After having proposed a general description of the evolution of the general design of reactors with a vessel in pre-stressed concrete, this report outlines the interest of this technical solution of a vessel in pre-stressed concrete with integrated exchangers, which is to replace steel vessel. This solution is presented as much safer. The authors discuss the various issues related to protection: inner and outer biological protection of the vessel, material protection (against heating, steel irradiation, Wigner effect, and moderator radiolytic corrosion). They report the application of calculation methods: calculation of vessel concrete heating, study of the intermediate zone in integrated reactors, neutron spectrum and flows in the core of a graphite pile

  6. Nonlinear analysis of prestressed concrete reactor pressure vessels

    International Nuclear Information System (INIS)

    Berg, S.; Loeseth, S.; Holand, I.

    1977-01-01

    A computational model for circular symmetric reinforced concrete shell problems is described. The model is based on the Finite Element Method. Non-linear stress-strain constitutive relations are used for the concrete, the reinforcement and for the liner. The reinforcement layers may be of different steel qualities. Each layer may be given a specified prestressing. This can be done at the beginning of the computations or the specific reinforcement layer can be considered inactive until a specified level of loading is reached. Thus, the prestressing procedure may also be analyzed in detail. Bond-slip effects are not accounted for. However, no bond may be assumed for prestressing cables by inserting special reinforcement elements. Several models of prestressed concrete reactor pressure vessels which have been tested up to rupture have been analysed. Analytical (numerical) models for reinforced concrete are also discussed on a more general basis. (Auth.)

  7. Shielding of radiation fields generated by 252Cf in a concrete maze. Part 2 -- Simulation

    International Nuclear Information System (INIS)

    Fasso, A.; Ipe, N.E.; Reyna, A.

    1998-03-01

    A streaming experiment performed in a concrete maze of shape and size typical of a radiotherapy room was simulated with the Monte Carlo program FLUKA. The purpose of the calculation was to test the performance of the code in the low energy neutron range, and at the same time to provide additional information which could help in optimizing shielding of medical facilities. Instrument responses were calculated at different maze locations for several experimental configurations and were compared with measurements. In addition, neutron and gamma fluence, ambient dose equivalent and effective dose were calculated at the same positions. Both sources used in the experiment, namely a bare 252 Cf source and one shielded by a tungsten shell 5 cm thick, were considered in the simulation

  8. Shielding of radiation fields generated by {sup 252}Cf in a concrete maze. Part 2 -- Simulation

    Energy Technology Data Exchange (ETDEWEB)

    Fasso, A.; Ipe, N.E.; Reyna, A. [Stanford Univ., CA (US). Stanford Linear Accelerator Center; McCall, R.C. [McCall Associates, Woodside, CA (US)

    1998-03-01

    A streaming experiment performed in a concrete maze of shape and size typical of a radiotherapy room was simulated with the Monte Carlo program FLUKA. The purpose of the calculation was to test the performance of the code in the low energy neutron range, and at the same time to provide additional information which could help in optimizing shielding of medical facilities. Instrument responses were calculated at different maze locations for several experimental configurations and were compared with measurements. In addition, neutron and gamma fluence, ambient dose equivalent and effective dose were calculated at the same positions. Both sources used in the experiment, namely a bare {sup 252}Cf source and one shielded by a tungsten shell 5 cm thick, were considered in the simulation.

  9. Calculations of the photon dose behind concrete shielding of high energy proton accelerators

    International Nuclear Information System (INIS)

    Dworak, D.; Tesch, K.; Zazula, J.M.

    1992-02-01

    The photon dose per primary beam proton behind lateral concrete shieldings was calculated by using an extension of the Monte Carlo particle shower code FLUKA. The following photon-producing processes were taken into account: capture of thermal neutrons, deexcitation of nuclei after nuclear evaporation, inelastic neutron scattering and nuclear reactions below 140 MeV, as well as photons from electromagnetic cascades. The obtained ratio of the photon dose to the neutron dose equivalent varies from 8% to 20% and it well compares with measurements performed recently at DESY giving a mean ratio of 14%. (orig.)

  10. Availability of special local rock materials for using in radiation shielding concrete

    International Nuclear Information System (INIS)

    Rammah, S.; Al-Hent, R.; Aissa, M.; Yousef, S.

    2003-11-01

    Concrete is an excellent and versatile material for using in radiation shielding of nuclear power plants, hot cells and medical facilities that deal with ionizing radiations, Because it is easy controlled with composition and density by using aggregates with high specific gravity such as Barite, Hematite, Magnetite, or minerals with high hydrogen content such as Serpentine. This research offered the essential information about local resources rocks and minerals can be used in this inclination, as aggregates for heavy/high hydrion concrete. The present work indicates that iron ores, which located in RAJO-EFREEN is better than other locations like ANTI-LEBANON or AL-KADMOUS. While the heavy beach sands in AL-BASSIT are the best compared with other locations on the Syrian seaside, because it has acceptable percentage of heavy mineral. Barite concretions were found in KALAMON, HOMS and other sites, which its percentages approach 50%, but however in small quantities. Finally, high hydrion concrete can be used by Serpentinite were found with high Serpentine percentage in BAYER and BASSIT blocks. (author)

  11. Finite element analysis of prestressed concrete reactor vessels

    International Nuclear Information System (INIS)

    Smith, P.D.; Cook, W.A.; Anderson, C.A.

    1977-01-01

    Several present and proposed gas-cooled reactors use concrete pressure vessels. In addition, concrete is almost universally used for the secondary containment structures of water-cooled reactors. Regulatory agencies must have means of assuring that these concrete structures perform their containment functions during normal operation and after extreme conditions of transient overpressure and high temperature. The NONSAP nonlinear structural analysis program has been extensively modified to provide one analytical means of assessing the safety of reinforced concrete pressure vessels and containments. Several structural analysis codes were studied to evaluate their ability to model the nonlinear static and dynamic behavior of three-dimensional structures. The NONSAP code was selected because of its availability and because of the ease with which it can be modified. In particular, the modular structure of this code allows ready addition of specialized material models. Major modifications have been the development of pre- and post-processors for mesh generation and graphics, the addition of an out-of-core solver, and the addition of constitutive models for reinforced concrete subject to either long-term or short-term loads. Emphasis was placed on development of a three-dimensional analysis capability

  12. Material problems in accident analysis of prestressed concrete reactor vessels

    International Nuclear Information System (INIS)

    Bazant, Z.P.

    1977-01-01

    Due to their very high energy absorption capability, as well as their inherent safety advantages, prestressed concrete reactor vessels are presently being keenly studied as the basic barrier to contain hypothetical core disruptive accidents in a fast breeder reactor. One problem investigated is the nonlinear constitutive behavior and failure criteria for concrete. Previously, a comprehensive theory, called endochronic theory, has been shown to satisfy all basic currently known features of test data. Nevertheless uncertainty still exists with regard to non-proportional loading paths, for which good test data are lacking at present. An extension of the endochronic theory which correlates best with general experimental evidence and includes fracturing terms is given, and a comparison with vertex-type hardening in plasticity is made. A second problem which must be analysed in accident situations is the high temperature shock on the concrete walls (due to liquid sodium, up to 850 0 C). Refining a previous crude formulation, a rational model for calculating moisture and heat transfer and pore pressures in concrete subjected to thermal shock is presented. In conclusion, a new design concept, in which the concrete vessel is completely dehydrated and kept hot throughout its service life in order to substantially improve its response to thermal shock as well as liquid sodium contact, is described. (Auth.)

  13. A model to predict moisture conditions in concrete reactor containments

    International Nuclear Information System (INIS)

    Ahs, M.; Nilsson, L.O.; Poyet, S.; L'Hostis, V.

    2015-01-01

    Moisture has an impact in many of the degradation mechanisms that appear in the structures of a nuclear power plant. Moisture conditions in a reactor containment wall have been simulated by using a hygro-thermal model of drying concrete. Methods to estimate the temperature dependency of the sorption isotherms and moisture transport properties is suggested and applied in the model. This temperature dependency is included as there is a temperature gradient present through the containment wall. The hygro-thermal model was applied on a full scale 3D model of a real reactor containment building and the concrete relative humidity has been computed at 4 different moments: 1, 10, 20 and 30 years. The results show that the major part of the concrete is not dried at all even after 30 years of operation. It is also clear that the temperature distribution inside the whole concrete volume is affected by the variable boundary conditions. It was concluded that the suggested hygro-thermal model was appropriate to use as a method to estimate the existing conditions in a PWR reactor containment wall

  14. Engineering and Fabrication Considerations for Cost-Effective Space Reactor Shield Development

    International Nuclear Information System (INIS)

    Berg, Thomas A.; Disney, Richard K.

    2004-01-01

    Investment in developing nuclear power for space missions cannot be made on the basis of a single mission. Current efforts in the design and fabrication of the reactor module, including the reactor shield, must be cost-effective and take into account scalability and fabricability for planned and future missions. Engineering considerations for the shield need to accommodate passive thermal management, varying radiation levels and effects, and structural/mechanical issues. Considering these challenges, design principles and cost drivers specific to the engineering and fabrication of the reactor shield are presented that contribute to lower recurring mission costs

  15. 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.

  16. Rapid detailed characterization of concrete shielding blocks utilizing internal natural radionuclides for calibration

    International Nuclear Information System (INIS)

    McDonald, R.J.; Smith, A.R.; Hurley, D.L.; Norman, E.B.; Schoonover, M.R.

    1998-01-01

    Following many years of productive research, the 184-inch Cyclotron, the SuperHILAC, and the BEVALAC accelerators at the Berkeley Laboratory were closed, leaving thousands of concrete shielding blocks available for reuse, recycling, or disposal. The process history of these blocks precludes free release pending radiological characterization. This paper describes a procedure whereby a high efficiency shielded germanium spectrometer is used to rapidly characterize natural and man-made activity within the blocks. The spectrometer is moved up to the block and 5 minutes of data are collected at the point on the block that registers highest on a micro-R meter. Sensitivity is better than 1 pCi/g (0.037 Bq/g) for Co-60 and Eu-152, the prominent man-made activities observed. One-time calibration of the detector system is obtained from a sample of concrete, drilled with a hammer drill, counted in our low-background facility, and compared to crushed rock with known U, Th, and K activity. A simple relationship exists between the counts/minute observed in a characteristic gamma-ray peak and the activity in the block. (author)

  17. Radiation shielding provided by residential houses in Japan in reactor accidents accompanied with atmospheric release

    International Nuclear Information System (INIS)

    Yamaguchi, Yasuhiro; Minami, Kentaro

    1991-01-01

    The present report describes the radiation shielding effect of houses in Japan against the radioactive cloud resulting from a major reactor accident accompanied with atmospheric release. The shielding factor of houses, the ratio of indoor exposure rate to outdoor one, has been studied for the semi-infinite and finite clouds which contain γ-emitting radionuclides released from a reactor facility. The shielding factor of houses against γ-rays from the radioactive cloud decreases gradually with release delay time and keeps a minimum during the period from 50 to 1000 hours after reactor shutdown while 133 Xe predominates in the cloud. Radioiodines mixed in the cloud raise slightly the shielding factor, and the factor depends little on the shape of the cloud. A set of shielding factors for the use of emergency planning was consequently proposed as 0.4 for simple ferroconcrete residential house and 0.9 for other ordinary ones. (author)

  18. ANS shielding standards for light-water reactors

    International Nuclear Information System (INIS)

    Trubey, D.K.

    1982-01-01

    The purpose of the American Nuclear Society Standards Subcommittee, ANS-6, Radiation Protection and Shielding, is to develop standards for radiation protection and shield design, to provide shielding information to other standards-writing groups, and to develop standard reference shielding data and test problems. A total of seven published ANS-6 standards are now current. Additional projects of the subcommittee, now composed of nine working groups, include: standard reference data for multigroup cross sections, gamma-ray absorption coefficients and buildup factors, additional benchwork problems for shielding problems and energy spectrum unfolding, power plant zoning design for normal and accident conditions, process radiation monitors, and design for postaccident radiological conditions

  19. Minimum thickness blanket-shield for fusion reactors

    International Nuclear Information System (INIS)

    Karni, Y.; Greenspan, E.

    1989-01-01

    A lower bound on the minimum thickness fusion reactor blankets can be designed to have, if they are to breed 1.267 tritons per fusion neutron, is identified by performing a systematic nucleonic optimization of over a dozen different blanket concepts which use either Be, Li 17 Pb 83 , W or Zr for neutron multiplication. It is found that Be offers minimum thickness blankets; that the blanket and shield (B/S) thickness of Li 17 Pb 83 based blankets which are supplemented by Li 2 O and/or TiH 2 are comparable to the thickness of Be based B/S; that of the Be based blankets, the aqueous self-cooled one offers one of the most compact B/S; and that a number of blanket concepts might enable the design of B/S which is approximately 12 cm and 39 cm thinner than the B/S thickness of, respectively, conventional self-cooled Li 17 Pb 83 and Li blankets. Aqueous self-cooled tungsten blankets could be useful for experimental fusion devices provided they are designed to be heterogeneous. (orig.)

  20. Shielding data for hadron-therapy ion accelerators: Attenuation of secondary radiation in concrete

    CERN Document Server

    Agosteo, S; Sagia, E; Silari, M

    2014-01-01

    The secondary radiation field produced by seven different ion species (from hydrogen to nitrogen), impinging onto thick targets made of either iron or ICRU tissue, was simulated with the FLUKA Monte Carlo code, and transported through thick concrete shields: the ambient dose equivalent was estimated and shielding parameters evaluated. The energy for each ion beam was set in order to reach a maximum penetration in ICRU tissue of 290 mm (equivalent to the therapeutic range of 430 MeV/amu carbon ions). Source terms and attenuation lengths are given as a function of emission angle and ion species, along with fits to the Monte Carlo data, for shallow depth and deep penetration in the shield. Trends of source terms and attenuation lengths as a function of neutron emission angle and ion species impinging on tar- get are discussed. A comparison of double differential distributions of neutrons with results from similar simulation works reported in the literature is also included. The aim of this work is to provide shi...

  1. 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.

  2. Measurement of concentrations of {gamma}-ray emitters induced in the concrete shield of the JAERI electron linac facility

    Energy Technology Data Exchange (ETDEWEB)

    Endo, Akira; Kawasaki, Katsuya; Kikuchi, Masamitsu [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Harada, Yasunori

    1997-07-01

    Measurement has been made to study distributions of {gamma}-ray emitters induced in the concrete shield of the JAERI electron linac facility. Core boring was carried out at seven positions to take samples from the concrete shield, and {gamma}-ray counting rates and {gamma}-ray spectra of these samples were measured with a NaI(Tl) detector and a Ge semiconductor detector, respectively. The following radionuclides were detected in the concrete samples: {sup 60}Co, {sup 134}Cs, {sup 152}Eu and {sup 154}Eu generated through thermal neutron capture reaction, and {sup 22}Na and {sup 54}Mn generated through nuclear reactions by bremsstrahlung and fast neutrons. The relation between the distributions of {gamma}-ray emitters, as a function of the depth of concrete, and the positions of core boring is discussed. (author)

  3. Effect of Gamma Ray Energies and Steel Fiber addition by Weight on some Shielding Properties of Limestone Concrete

    International Nuclear Information System (INIS)

    Abd El-Latifa, A.A.; Ikraiam, F.A.; Abd El-Latifa, A.A.; Abd Elazziz, A.; Abd Elazziz, A.

    2010-01-01

    The mass attenuation coefficient , the build up factor , the half value thickness X 1/2 , and tenth value thickness X 1/10 of fiber concrete , 0% , 1% , 2%, 3%, and 4% by weight fiber content were measured at different gamma ray energies in MeV, 0.511,1.274 from Na-22 ,1.17 ,1.33 from Co-60 and 0.662 from Cs-137 . Appreciable variations were noted in the former nuclear parameters, due to the changes in the fiber content and gamma ray energies .A comparison of shielding properties of concrete with fiber content and reference sample(concrete without fiber ) have proven that the addition of steel fibers by weight to concrete have a potential application as a radiation shielding

  4. EXPERIMENTAL EVALUATION OF THE THERMAL PERFORMANCE OF A WATER SHIELD FOR A SURFACE POWER REACTOR

    International Nuclear Information System (INIS)

    Reid, Robert S.; Pearson, J. Bosie; Stewart, Eric T.

    2007-01-01

    Water based reactor shielding is being investigated for use on initial lunar surface power systems. A water shield may lower overall cost (as compared to development cost for other materials) and simplify operations in the setup and handling. The thermal hydraulic performance of the shield is of significant interest. The mechanism for transferring heat through the shield is natural convection. Natural convection in a 100 kWt lunar surface reactor shield design is evaluated with 2 kW power input to the water in the Water Shield Testbed (WST) at the NASA Marshall Space Flight Center. The experimental data from the WST is used to validate a CFD model. Performance of the water shield on the lunar surface is then predicted with a CFD model anchored to test data. The experiment had a maximum water temperature of 75 C. The CFD model with 1/6-g predicts a maximum water temperature of 88 C with the same heat load and external boundary conditions. This difference in maximum temperature does not greatly affect the structural design of the shield, and demonstrates that it may be possible to use water for a lunar reactor shield.

  5. EXPERIMENTAL EVALUATION OF THE THERMAL PERFORMANCE OF A WATER SHIELD FOR A SURFACE POWER REACTOR

    Energy Technology Data Exchange (ETDEWEB)

    REID, ROBERT S. [Los Alamos National Laboratory; PEARSON, J. BOSIE [Los Alamos National Laboratory; STEWART, ERIC T. [Los Alamos National Laboratory

    2007-01-16

    Water based reactor shielding is being investigated for use on initial lunar surface power systems. A water shield may lower overall cost (as compared to development cost for other materials) and simplify operations in the setup and handling. The thermal hydraulic performance of the shield is of significant interest. The mechanism for transferring heat through the shield is natural convection. Natural convection in a 100 kWt lunar surface reactor shield design is evaluated with 2 kW power input to the water in the Water Shield Testbed (WST) at the NASA Marshall Space Flight Center. The experimental data from the WST is used to validate a CFD model. Performance of the water shield on the lunar surface is then predicted with a CFD model anchored to test data. The experiment had a maximum water temperature of 75 C. The CFD model with 1/6-g predicts a maximum water temperature of 88 C with the same heat load and external boundary conditions. This difference in maximum temperature does not greatly affect the structural design of the shield, and demonstrates that it may be possible to use water for a lunar reactor shield.

  6. Experimental Evaluation of the Thermal Performance of a Water Shield for a Surface Power Reactor

    International Nuclear Information System (INIS)

    Pearson, J. Boise; Stewart, Eric T.; Reid, Robert S.

    2007-01-01

    Water based reactor shielding is being investigated for use on initial lunar surface power systems. A water shield may lower overall cost (as compared to development cost for other materials) and simplify operations in the setup and handling. The thermal hydraulic performance of the shield is of significant interest. The mechanism for transferring heat through the shield is natural convection. Natural convection in a 100 kWt lunar surface reactor shield design is evaluated with 2 kW power input to the water in the Water Shield Testbed (WST) at the NASA Marshall Space Flight Center. The experimental data from the WST is used to validate a CFD model. Performance of the water shield on the lunar surface is then predicted with a CFD model anchored to test data. The experiment had a maximum water temperature of 75 deg. C. The CFD model with 1/6-g predicts a maximum water temperature of 88 deg. C with the same heat load and external boundary conditions. This difference in maximum temperature does not greatly affect the structural design of the shield, and demonstrates that it may be possible to use water for a lunar reactor shield

  7. 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

  8. 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.

  9. 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.

  10. Design and analysis of multicavity prestressed concrete reactor vessels

    International Nuclear Information System (INIS)

    Goodpasture, D.W.; Burdette, E.G.; Callahan, J.P.

    1977-01-01

    During the past 25 years, a rather rapid evolution has taken place in the design and use of prestressed concrete reactor vessels (PCRVs). Initially the concrete vessel served as a one-to-one replacement for its steel counterpart. This was followed by the development of the integral design which led eventually to the more recent multicavity vessel concept. Although this evolution has seen problems in construction and operation, a state-of-the-art review which was recently conducted by the Oak Ridge National Laboratory indicated that the PCRV has proven to be a satisfactory and inherently safe type of vessel for containment of gas-cooled reactors from a purely functional standpoint. However, functionalism is not the only consideration in a demanding and highly competitive industry. A summary is presented of the important considerations in the design and analysis of multicavity PCRVs together with overall conclusions concerning the state of the art of these vessels

  11. Shielding Calculations for PUSPATI TRIGA Reactor (RTP) Fuel Transfer Cask with Micro shield

    International Nuclear Information System (INIS)

    Nurhayati Ramli; Ahmad Nabil Abdul Rahim; Ariff Shah Ismail

    2011-01-01

    The shielding calculations for RTP fuel transfer cask was performed by using computer code Micro shield 7.02. Micro shield is a computer code designed to provide a model to be used for shielding calculations. The results of the calculations can be obtained fast but the code is not suitable for complex geometries with a shielding composed of more than one material. Nevertheless, the program is sufficient for As Low As Reasonable Achievable (ALARA) optimization calculations. In this calculation, a geometry based on the conceptual design of RTP fuel transfer cask was modeled. Shielding material used in the calculations were lead (Pb) and stainless steel 304 (SS304). The results obtained from these calculations are discussed in this paper. (author)

  12. Research and development of the prestressed concrete reactor vessel

    International Nuclear Information System (INIS)

    Shiozawa, Shoji; Omata, Ippei; Nakamura, Norio

    1975-01-01

    Compared with the steel reactor vessel, the prestressed concrete reactor vessel (PCRV) is said to be superior in safety and economy. One of the characteristics of the high temperature gas cooled reactor (HTGR) is the adoption of the PCRV instead of the steel reactor vessel to ensure safety. In order to improve safety characteristics, it is necessary for the PCRV to be provided with more reliable functions. When the multi-purpose HTGR or the gas cooled fast breeder reactor (GCFR) are realized in future, more severe conditions of technology will be imposed on the PCRV, and accordingly, technical developments are now increasingly required. IHI is now proceeding with the technical research and development on the PCRV, in which a basic study of its liner cooling system has already been completed. In this study applying a large cylindrical PCRV model, comparison was made between experimental data and analyses concerning the liner cooling system, and the results of analytical technique have been evaluated. The analytical technique established this time is applicable to the estimation of temperature distribution in the concrete of a large PCRV and also to the evaluation of the liner cooling system. (auth.)

  13. Design of the shield door and transporter for the Culham Conceptual Tokamak Reactor Mark II

    International Nuclear Information System (INIS)

    Guthrie, J.A.S.

    1980-04-01

    In the Culham Conceptual Tokamak Reactor MK II access to the interior for blanket maintenance is through large openings in the fixed shield structure closed by removable shield doors when the reactor is operational. This report describes the design of the 200 tonne doors and the associated special-purpose remote operating transporter manipulator. The design, which has not been optimised, generally uses available commercial equipment and state-of-the-art techniques. (U.K.)

  14. Transient thermal creep of nuclear reactor pressure vessel type concretes

    International Nuclear Information System (INIS)

    Khoury, G.A.

    1983-01-01

    The immediate aim of the research was to study the transient thermal strain behaviour of four AGR type nuclear reactor concretes during first time heating in an unsealed condition to 600 deg. C. The work being also relevant to applications of fire exposed concrete structures. The programme was, however, expanded to serve a second more theoretical purpose, namely the further investigation of the strain development of unsealed concrete under constant, transient and cyclic thermal states in particular and the effect of elevated temperatures on concrete in general. The range of materials investigated included seven different concretes and three types of cement paste. Limestone, basalt, gravel and lightweight aggregates were employed as well as OPC and SRC cements. Cement replacements included pfa and slag. Test variables comprised two rates of heating (0.2 and 1 deg. C/minute), three initial moisture contents (moist as cast, air-dry and oven dry at 105 deg. C), two curing regimes (bulk of tests represented mass cured concrete), five stress levels (0, 10, 20, 30 and a few tests at 60% of the cold strength), two thermal cycles and levels of test temperature up to 720 deg. C. Supplementary, dilatometry, TGA and DTA tests were performed at CERL on individual samples of aggregate and cement paste which helped towards explaining the observed trends in the concretes. A simple formula was developed which relates the elastic thermal stresses generated from radial temperature gradients to the solution obtained from the transient heat conduction equation. Thermal stresses can, therefore, be minimized by reductions in the radius of the specimen and the rate of heating The results were confirmed by finite element analysis which indicate( tensile stresses in the central region and compressive stresses near the surf ace during heating which are reversed during cooling. It is shown that the temperature gradients, pore pressures and tensile thermal stresses during both heating and

  15. Apparatus for sealing a rotatable shield plug in a liquid metal nuclear reactor

    International Nuclear Information System (INIS)

    Winkleblack, R.K.

    1980-01-01

    An apparatus for sealing a rotatable shield plug in a nuclear reactor having liquid metal coolant is described. The apparatus includes a dip -ring seal adapted to provide a fluid barrier between the liquid metal and the atmosphere and to permit rotation of the shield plug. The apparatus also includes a static seal for the rotatable shield plug located between the dip-ring seal and the liquid metal. The static seal isolates the dip-ring seal from the liquid metal vapor during operation at power and can be disengaged for rotation of the shield plug

  16. Development and investigation of the prestressed reinforced concrete vessels for the water cooled reactors in the FRG

    International Nuclear Information System (INIS)

    Medovikov, A.I.; Bogopol'skij, V.G.; Nikolaev, Yu.B.; Konevskij, V.N.

    1980-01-01

    An analysis of calculation results for characteristics of stress-strained state of reactor vessel made of prestressed reinforced concrete is presented. Experimental data obtained during the investigation into a model of reactor vessel top cover are given. Thermal shielding system both for boiling water and pressurized-water reactors has been considered and its working capacity has been evaluated. An analysis of experimental data show correctness of the method assumed for calculation of the reactor top cover which permits to exactly determine its stressed-strained state as well as the nature of crack propagation in the vessel and the structure supporting power. Ceramics is suggested to be used as a heat-insulating material

  17. Verification of using SABINE-3.1 code for calculations of radioactive inventory in reactor shield

    International Nuclear Information System (INIS)

    Moukhamadeev, R.; Suvorov, A.

    2000-01-01

    This report presents the results of calculations of radioactive inventory and doses of activation radiation for the International Benchmark Calculations of Radioactive Inventory for Fission Reactor Decommissioning, IAEA, and measurements of activation doses in shield of WWER-440 (Armenian NPP), using one-dimension modified code SABINE-3.1. For decommissioning of NPP it is very important to evaluate in correct manner radioactive inventory in reactor construction and shield materials. One-dimension code SABINE-3.1 (removing-diffusion method for neutron calculation) was modified to perform calculation of radioactive inventory in reactor shield materials and dose from activation photons behind them. These calculations are carried out on the base of nuclear constant system ABBN-78 and new library of activation data for a number of long-lived isotopes, prepared by authors on the base of [9], which present at shield materials as microimpurities and manage radiation situation under the decay more than 1 year. (Authors)

  18. Shielding design study of the demonstration fast breeder reactor. 2. Shielding design on the basis of the JASPER analysis

    International Nuclear Information System (INIS)

    Suzuoki, Zenro; Tabayashi, Masao; Handa, Hiroyuki; Iida, Masaaki; Takemura, Morio

    2000-01-01

    Conceptual shielding design has been performed for the Demonstration Fast Breeder Reactor (DFBR) to achieve further optimization and reduction of the plant construction cost. The design took into account its implications in overall plant configuration such as reduction of shields in the core, adoption of fission gas plenum in the lower portion of fuel assemblies, and adoption of gas expansion modules. Shielding criteria applied for the design are to secure fast neutron fluence on in-vessel structures as well as responses of the nuclear instrumentation system and to restrict secondary sodium activation. The design utilized the cross sections and the one- and two-dimensional discrete ordinates transport codes, whose verification had been performed by the JASPER experiment analysis. Correction factors yielded by the JASPER analysis were applied to the design calculations to obtain design values with improved accuracy. Design margins, which are defined by the ratios of the design criteria to the design values, were more than two for all shielding issues of interest, showing the adequacy of the shielding design of the DFBR. (author)

  19. Primary shield displacement and bowing

    International Nuclear Information System (INIS)

    Scott, K.V.

    1978-01-01

    The reactor primary shield is constructed of high density concrete and surrounds the reactor core. The inlet, outlet and side primary shields were constructed in-place using 2.54 cm (1 in) thick steel plates as the forms. The plates remained as an integral part of the shields. The elongation of the pressure tubes due to thermal expansion and pressurization is not moving through the inlet nozzle hardware as designed but is accommodated by outward displacement and bowing of the inlet and outlet shields. Excessive distortion of the shields may result in gas seal failures, intolerable helium gas leaks, increased argon-41 emissions, and shield cooling tube failures. The shield surveillance and testing results are presented

  20. Structure shielding from cloud and fallout gamma ray sources for assessing the consequences of reactor accidents

    International Nuclear Information System (INIS)

    Burson, Z.G.; Profio, A.E.

    1975-12-01

    Radiation shielding provided by transportation vehicles and structures typical of where people live and work were estimated for cloud and fallout gamma-ray sources resulting from a hypothetical reactor accident. Dose reduction factors are recommended for a variety of situations for realistically assessing the consequences of reactor accidents

  1. Nonlinear analysis of prestressed concrete reactor pressure vessels

    International Nuclear Information System (INIS)

    Connor, J.J.

    1975-01-01

    The numerical procedures for predicting the nonlinear behavior of a prestressed concrete reactor vessel over its design life are discussed. The numerical models are constructed by combining three-dimensional isoparametric finite elements which simulate the concrete, thin shell elements which simulate steel linear plates, and layers of reinforcement steel, and axial elements for discrete prestressing cables. Nonlinearity under compressive stress, multi-dimensional cracking, shrinkage and stress/temperature induced creep of concrete are considered in addition to the elasti-plastic behavior of the liner and reinforcing steel. Various failure theories for concrete have been proposed recently. Also, there are alternative strategies for solving the discrete system equations over the design life, accounting for test loads, pressure and temperature operational loads, creep unloading and abnormal loads. The proposed methods are reviewed, and a new formulation developed by the authors is described. A number of comparisons with experimental tests results and other numerical schemes are presented. These examples demonstrate the validity of the formulation and also provide valuable information concerning the cost and accuracy of the various solution strategies i.e., total vs. incremental loading and initial vs. tangent stiffness. Finally, the analysis of an actual PCRV is described. Stress contours and cracking patterns in the region of cutouts corresponding to operational pressure and temperature loads are illustrated. The effects of creep, unloading, and creep recovery are then shown. Lastly, a strategy for assessing the performance over its design life is discussed

  2. Aging management of light water reactor concrete containments

    International Nuclear Information System (INIS)

    Shah, V.N.; Hookhman, C.J.

    1994-01-01

    This paper evaluates aging of light water reactor concrete containments and identifies three degradation mechanisms that have potential to cause widespread aging damage after years of satisfactory experience: alkali-silica reaction, corrosion of reinforcing steel, and sulfate attack. The evaluation is based on a comprehensive review of the relevant technical literature. Low-alkali cement and slow-reacting aggregates selected according to ASTM requirements cause deleterious alkali-silica reactions. Low concentrations of chloride ions can initiate corrosion of the reinforcing steel if the hydroxyl ions are sufficiently reduced by carbonation, leaching, or magnesium sulfate attack. Magnesium sulfate attack on concrete can cause loss of strength and cementitious properties after long exposure. Techniques to detect and mitigate these long-term aging effects are discussed

  3. Finite element analysis of prestressed concrete reactor vessels

    International Nuclear Information System (INIS)

    Smith, P.D.; Cook, W.A.; Anderson, C.A.

    1977-01-01

    This paper discusses the development of a finite element code suitable for the safety analysis of prestressed concrete reactor vessels. The project has involved modification of a general purpose computer code to handle reinforced concrete structures as well as comparison of results obtained with the code against published experimental data. The NONSAP nonlinear structural analysis program was selected for the ease with which it can be modified to encompass problems peculiar to nuclear reactors. Pre- and post-processors have been developed for mesh generation and for graphical display of response variables. An out-of-core assembler and solver have been developed for the analysis of large three dimensional problems. The constitutive model for short term loads forms an orthotropic stress-strain relationship in which the concrete and the reinforcing steel are treated as a composite. The variation of stiffness and strength of concrete under multiaxial stress states is accounted for. Cracks are allowed to form at element integration points based on a three dimensional failure envelope in stress space. Composite tensile and shear properties across a crack are modified to account for bond degradation and for dowel action of the reinforcement. The constitutive law for creep is base on the expansion of the usual creep compliance function in the form of a Dirichlet exponential series. Empirical creep data are then fit to the Dirichlet series approximation by means of a least squares procedure. The incremental deformation process is subsequently reduced to a series of variable stiffness elasticity problems in which the past stress history is represented by a finite number of hidden material variables

  4. Meso-scale modeling of irradiated concrete in test reactor

    International Nuclear Information System (INIS)

    Giorla, A.; Vaitová, M.; Le Pape, Y.; Štemberk, P.

    2015-01-01

    Highlights: • A meso-scale finite element model for irradiated concrete is developed. • Neutron radiation-induced volumetric expansion is a predominant degradation mode. • Confrontation with expansion and damage obtained from experiments is successful. • Effects of paste shrinkage, creep and ductility are discussed. - Abstract: A numerical model accounting for the effects of neutron irradiation on concrete at the mesoscale is detailed in this paper. Irradiation experiments in test reactor (Elleuch et al., 1972), i.e., in accelerated conditions, are simulated. Concrete is considered as a two-phase material made of elastic inclusions (aggregate) subjected to thermal and irradiation-induced swelling and embedded in a cementitious matrix subjected to shrinkage and thermal expansion. The role of the hardened cement paste in the post-peak regime (brittle-ductile transition with decreasing loading rate), and creep effects are investigated. Radiation-induced volumetric expansion (RIVE) of the aggregate cause the development and propagation of damage around the aggregate which further develops in bridging cracks across the hardened cement paste between the individual aggregate particles. The development of damage is aggravated when shrinkage occurs simultaneously with RIVE during the irradiation experiment. The post-irradiation expansion derived from the simulation is well correlated with the experimental data and, the obtained damage levels are fully consistent with previous estimations based on a micromechanical interpretation of the experimental post-irradiation elastic properties (Le Pape et al., 2015). The proposed modeling opens new perspectives for the interpretation of test reactor experiments in regards to the actual operation of light water reactors.

  5. Meso-scale modeling of irradiated concrete in test reactor

    Energy Technology Data Exchange (ETDEWEB)

    Giorla, A. [Oak Ridge National Laboratory, One Bethel Valley Road, Oak Ridge, TN 37831 (United States); Vaitová, M. [Czech Technical University, Thakurova 7, 166 29 Praha 6 (Czech Republic); Le Pape, Y., E-mail: lepapeym@ornl.gov [Oak Ridge National Laboratory, One Bethel Valley Road, Oak Ridge, TN 37831 (United States); Štemberk, P. [Czech Technical University, Thakurova 7, 166 29 Praha 6 (Czech Republic)

    2015-12-15

    Highlights: • A meso-scale finite element model for irradiated concrete is developed. • Neutron radiation-induced volumetric expansion is a predominant degradation mode. • Confrontation with expansion and damage obtained from experiments is successful. • Effects of paste shrinkage, creep and ductility are discussed. - Abstract: A numerical model accounting for the effects of neutron irradiation on concrete at the mesoscale is detailed in this paper. Irradiation experiments in test reactor (Elleuch et al., 1972), i.e., in accelerated conditions, are simulated. Concrete is considered as a two-phase material made of elastic inclusions (aggregate) subjected to thermal and irradiation-induced swelling and embedded in a cementitious matrix subjected to shrinkage and thermal expansion. The role of the hardened cement paste in the post-peak regime (brittle-ductile transition with decreasing loading rate), and creep effects are investigated. Radiation-induced volumetric expansion (RIVE) of the aggregate cause the development and propagation of damage around the aggregate which further develops in bridging cracks across the hardened cement paste between the individual aggregate particles. The development of damage is aggravated when shrinkage occurs simultaneously with RIVE during the irradiation experiment. The post-irradiation expansion derived from the simulation is well correlated with the experimental data and, the obtained damage levels are fully consistent with previous estimations based on a micromechanical interpretation of the experimental post-irradiation elastic properties (Le Pape et al., 2015). The proposed modeling opens new perspectives for the interpretation of test reactor experiments in regards to the actual operation of light water reactors.

  6. Optimization of a partially non-magnetic primary radiation shielding for the triple-axis spectrometer PANDA at the Munich high-flux reactor FRM-II

    CERN Document Server

    Pyka, N M; Rogov, A

    2002-01-01

    Monte Carlo simulations have been used to optimize the monochromator shielding of the polarized cold-neutron triple-axis spectrometer PANDA at the Munich high-flux reactor FRM-II. By using the Monte Carlo program MCNP-4B, the density of the total spectrum of incoming neutrons and gamma radiation from the beam tube SR-2 has been determined during the three-dimensional diffusion process in different types of heavy concrete and other absorbing material. Special attention has been paid to build a compact and highly efficient shielding, partially non-magnetic, with a total biological radiation dose of less than 10 mu Sv/h at its outsides. Especially considered was the construction of an albedo reducer, which serves to reduce the background in the experiment outside the shielding. (orig.)

  7. Prestressed concrete pressure vessels for boiling water reactors

    International Nuclear Information System (INIS)

    Menon, S.

    1979-12-01

    Following a general description of the Scandinavian cooperative project on prestressed concrete pressure vessels for boiling water reactors, detailed discussion is given in four appendices of the following aspects: the verification programme of tests and studies, the development and testing of a liner venting system, a preliminary safety philosophy and comparative assessment of cold and hot liners. Vessel failure probability is briefly discussed and some figures presented. The pressure gradients in the vessel wall resulting from various stipulated linear cracks, with a liner venting system are presented graphically. (JIW)

  8. Completely integrated prestressed-concrete reactor pressure vessel, type 'Star'

    International Nuclear Information System (INIS)

    Neunert, B.; Jueptner, G.; Kumpf, H.

    1975-01-01

    The star support vessel is suitable for the connection to all primary circuit systems consisting of a main vessel and a number of satellite vessels around and connected to it, i.e. for LWR, HTR and process reactor. It must be made clear, however, that the PWR in particular with its components does not appear to be suited for the optimum incorporation in a prestressed-concrete pressure vessel system, no matter what kind. There are clear concepts about modifications which, however, require considerable development expenditure. (orig./LH) [de

  9. Shear strength of end slabs of prestressed concrete reactor vessels

    International Nuclear Information System (INIS)

    Cheung, K.C.; Gotschall, H.L.; Liu, T.C.

    1975-01-01

    Prestressed concrete reactor vessels (PCRV's) have been adopted for primary containments in most large high-temperature gas-cooled reactor installations. The most common configuration for PCRVs is a right-vertical cylinder with thick end slabs. In order to assess the integrity of a PCRV it is necessary to predict the ultimate strength of the end slabs. The complexity of the basic mechanism of shear failure in the PCRV end slabs has thus far prohibited the development of a completely analytical solution. However, many experimental investigations of PCRV end slabs have been conducted over the past decade. This information makes it possible to establish empirical formulae for the ultimate strength of PCRV end slabs. The basis and development of an empirical shear-flexure interaction expression is presented. (Auth.)

  10. Electromagnetic characterization and shielding effectiveness of concrete composite reinforced with carbon nanotubes in the mobile phones frequency band

    Energy Technology Data Exchange (ETDEWEB)

    Micheli, D., E-mail: davide.micheli@uniroma1.it [“Sapienza” University of Rome, Department of Astronautic, Electric and Energy Engineering (DIAEE), Via Salaria 851, 00184 Rome (Italy); Pastore, R.; Vricella, A.; Morles, R.B.; Marchetti, M.; Delfini, A. [“Sapienza” University of Rome, Department of Astronautic, Electric and Energy Engineering (DIAEE), Via Salaria 851, 00184 Rome (Italy); Moglie, F.; Primiani, V. Mariani [Università Politecnica delle Marche, Department of Information Engineering (DII), Via Brecce Bianche 12, Ancona (Italy)

    2014-10-15

    Highlights: • The frequency band 0.75–1.12 GHz is exploited in mobile phone radio access network. • A lot of nanomaterial is needed for the measurement and no literature is available. • The manufacturing procedure is usually used for preparation of concrete composite. • High EM absorbing walls could be used to mitigate the human exposure to EM fields. • A shielding effectiveness of 50 dB is obtained for a 15 cm thick wall–3 wt% of CNT. - Abstract: The electromagnetic properties of carbon nanotube powder reinforced concretes are numerically and experimentally characterized. This typology of composite material is built by following the simple procedure usually adopted for the on-site concrete production. The dielectric parameters are investigated by means of waveguide measurements in the frequency band 0.75–1.12 GHz that is currently exploited in mobile phone radio access networks. The obtained results are used to compute the electromagnetic shielding effectiveness of large wall-shaped concrete structures. A shielding effectiveness up to 50 dB is obtained for a 15 cm thick wall when the carbon nanotube inclusion is raised up to 3 wt%.

  11. Electromagnetic characterization and shielding effectiveness of concrete composite reinforced with carbon nanotubes in the mobile phones frequency band

    International Nuclear Information System (INIS)

    Micheli, D.; Pastore, R.; Vricella, A.; Morles, R.B.; Marchetti, M.; Delfini, A.; Moglie, F.; Primiani, V. Mariani

    2014-01-01

    Highlights: • The frequency band 0.75–1.12 GHz is exploited in mobile phone radio access network. • A lot of nanomaterial is needed for the measurement and no literature is available. • The manufacturing procedure is usually used for preparation of concrete composite. • High EM absorbing walls could be used to mitigate the human exposure to EM fields. • A shielding effectiveness of 50 dB is obtained for a 15 cm thick wall–3 wt% of CNT. - Abstract: The electromagnetic properties of carbon nanotube powder reinforced concretes are numerically and experimentally characterized. This typology of composite material is built by following the simple procedure usually adopted for the on-site concrete production. The dielectric parameters are investigated by means of waveguide measurements in the frequency band 0.75–1.12 GHz that is currently exploited in mobile phone radio access networks. The obtained results are used to compute the electromagnetic shielding effectiveness of large wall-shaped concrete structures. A shielding effectiveness up to 50 dB is obtained for a 15 cm thick wall when the carbon nanotube inclusion is raised up to 3 wt%

  12. Application of MCNP code in shielding calculation of minitype fast reactor

    International Nuclear Information System (INIS)

    He Keyu; Han Weishi

    2008-01-01

    An accurate shielding calculation model has been set up for the minitype sodium-cooled fast reactor (MFR) based on MCNP code and particular calculation of its primary shielding parameters has been carried out. The results indicate that the photon and neutron flux density of MFR has rapidly fallen to a low-level. The material for the shielding layer outside of main container is primarily of carbon steel, which can be design as a shielding structure satisfying the safety code. The sodium activation in primary circuit is extremely limited and it is simple to shield from. Both the output of helium in reflector and burn up of boron-10 in control rod are very small. These materials can be used for several cycle lives. (authors)

  13. Assessment of Mass Fraction and Melting Temperature for the Application of Limestone Concrete and Siliceous Concrete to Nuclear Reactor Basemat Considering Molten Core–Concrete Interaction

    Directory of Open Access Journals (Sweden)

    Hojae Lee

    2016-04-01

    Full Text Available Severe accident scenarios in nuclear reactors, such as nuclear meltdown, reveal that an extremely hot molten core may fall into the nuclear reactor cavity and seriously affect the safety of the nuclear containment vessel due to the chain reaction caused by the reaction between the molten core and concrete. This paper reports on research focused on the type and amount of vapor produced during the reaction between a high-temperature molten core and concrete, as well as on the erosion rate of concrete and the heat transfer characteristics at its vicinity. This study identifies the mass fraction and melting temperature as the most influential properties of concrete necessary for a safety analysis conducted in relation to the thermal interaction between the molten core and the basemat concrete. The types of concrete that are actually used in nuclear reactor cavities were investigated. The H2O content in concrete required for the computation of the relative amount of gases generated by the chemical reaction of the vapor, the quantity of CO2 necessary for computing the cooling speed of the molten core, and the melting temperature of concrete are evaluated experimentally for the molten core–concrete interaction analysis.

  14. Assessment of mass fraction and melting temperature for the application of limestone concrete and siliceous concrete to nuclear reactor basemat considering molten core-concrete interaction

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Ho Jae; Kim, Do Gyeum [Korea Institute of Civil Engineering and Building Technology, Goyang (Korea, Republic of); Cho, Jae Leon [Korea Hydro and Nuclear Power Co., Ulsan (Korea, Republic of); Yoon, Eui Sik [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of); Cho, Myung Suk [Korea Hydro and Nuclear Power Co., Central Research Institute, Daejeon (Korea, Republic of)

    2016-04-15

    Severe accident scenarios in nuclear reactors, such as nuclear meltdown, reveal that an extremely hot molten core may fall into the nuclear reactor cavity and seriously affect the safety of the nuclear containment vessel due to the chain reaction caused by the reaction between the molten core and concrete. This paper reports on research focused on the type and amount of vapor produced during the reaction between a high-temperature molten core and concrete, as well as on the erosion rate of concrete and the heat transfer characteristics at its vicinity. This study identifies the mass fraction and melting temperature as the most influential properties of concrete necessary for a safety analysis conducted in relation to the thermal interaction between the molten core and the basemat concrete. The types of concrete that are actually used in nuclear reactor cavities were investigated. The H2O content in concrete required for the computation of the relative amount of gases generated by the chemical reaction of the vapor, the quantity of CO2 necessary for computing the cooling speed of the molten core, and the melting temperature of concrete are evaluated experimentally for the molten core-concrete interaction analysis.

  15. 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

  16. Radiation shielding considerations for the repair and maintenance of a swimming pool-type tokamak reactor

    International Nuclear Information System (INIS)

    Seki, Y.; Mori, S.

    1984-01-01

    The radiation shielding relevant to the repair and maintenance of a swimming pool-type tokamak reactor is considered. The dose rate during the reactor operation can be made low enough for personnel access into the reactor room if a 2m thick water layer is installed above the magnet cryostat. The dose rate 24 h after shutdown is such that the human access is allowed above the magnet cryostat. Sufficient water layer thickness is provided in the inboard space for the operation of automatic welder/cutter while retaining the magnet shielding capability. Some forced cooling is required for the decay heat removal in the first wall. The penetration shield thickness around the neutral beam injector port is estimated to be barely sufficient in terms of the magnet radiation damage. (orig.)

  17. Nonlinear analysis of end slabs in prestressed concrete reactor vessels

    International Nuclear Information System (INIS)

    Abdulrahman, H.O.

    1978-01-01

    A procedure for the nonlinear analysis of end slabs is prestressed concrete reactor vessels (PCRVs), based on the finite element method, is presented. The applicability of the procedure to the ultimate load analysis of small-scale models of the primary containment of nuclear reactors is shown. Material nonlinearity only is considered. The procedure utilizes the four-node linear quadrilateral isoparametric element with the choice of incorporating the nonconforming modes. This element is used for modeling the vessel as an axisymmetric solid. Concrete is assumed to be an isotropic material in the elastic range. The compressive stresses are judged according to a special form of the Mohr-Coulomb criterion. The nonlinear problem was solved using a generalized Newton-Raphson procedure. A detailed example problem of a pressure vessel with penetrations is presented. This is followed by a summary of the other cases studied. The solutions obtained match very closely the measured response of the test vessels under increasing internal pressure up to failure. The procedure is thus adequate for the assessment of the ultimate load behavior and failure of actual pressure vessels with a moderate demand on human and computational resources

  18. Prestressed concrete vessels suitable for helium high temperature reactors

    International Nuclear Information System (INIS)

    Lockett, G.E.; Kinkead, A.N.

    1967-02-01

    In considering prestressed concrete vessels for use with helium cooled high temperature reactors, a number of new problems arise and projected designs involve new approaches and new solutions. These reactors, having high coolant outlet temperature from the core and relatively high power densities, can be built into compact designs which permit usefully high working pressures. Consequently, steam generators and circulating units tend to be small. Although circuit activity can be kept quite low with coated particle fuels, designs which involve entry for subsequent repair are not favoured, and coupled with the preferred aim of using fully shop fabricated units within the designs with removable steam generators which involve no tube welding inside the vessel. A particular solution uses a number of slim cylindrical assemblies housed in the wall of the pressure vessel and this vessel design concept is presented. The use of helium requires very high sealing standards and one of the important requirements is a vessel design which permits leak testing during construction, so that a repair seal can be made to any faulty part in a liner seam. Very good demountable joint seals can be made without particular difficulty and Dragon experience is used to provide solutions which are suitable for prestressed concrete vessel penetrations. The concept layout is given of a vessel meeting these requirements; the basis of design is outlined and special features of importance discussed. (author)

  19. Evaluation of tritiated water retention capacity of fusion reactor concrete building

    International Nuclear Information System (INIS)

    Numata, S.; Fujii, Y.; Okamoto, M.

    1992-01-01

    In this paper the diffusion of tritiated water vapor into concrete walls is studied to evaluate tritiated water retention capacity of a fusion reactor concrete building. Using a model of the tritiated water diffusion determined form experimental results, depth profiles of tritiated water in concrete are calculated in the case of being exposed to air containing tritiated water vapor during the normal operational condition of a fusion reactor. A 0.5-m-thick concrete is sufficient for reactor hall walls from a viewpoint of the tritium containment

  20. Crack analysis of multicavity prestressed concrete reactor vessels

    International Nuclear Information System (INIS)

    Gallix, R.; Liu, T.C.; Lu, S.C.H.

    1975-01-01

    A new method to perform the crack analysis of non-axisymmetric, multicavity prestressed concrete reactor vessels (PCRV's) subjected to hypothetical overpressure by using an axisymmetric two-dimensional finite element computer code is presented. Concrete, steel liner, bonded reinforcing steel and prestressing steel elements are modeled. The limiting tensile strain criterion is adopted for concrete cracking. The steel elements are assumed to be elastic/perfectly plastic. Von Mises yield criterion and Prandtl-Reuss flow equations define the behavior of the liner in the range of plastic deformations. An orthotropic stress-strain constitutive law is utilized for cracked concrete elements. To account for the presence of penetrations and secondary cavities in the PCRV, a modified finite element model based on the concept of effective moduli is adopted. The pressure in these cavities is simulated by equivalent axisymmetric pressure distributions. In the analysis, the pressure is applied incrementally. For a given pressure, the displacements, strains, and stresses are computed. The state of strains or stresses is then examined against the cracking or yield criteria. If cracking or yield is indicated, the stiffness and load matrices for the cracked and yielding elements are recomputed and a new equilibrium is sought. This procedure is repeated until the desired convergence of the solution is achieved. The validity of the adopted approach utilizing the two-dimensional finite element method for overpressure analyses of non-axisymmetric PCRV's is demonstrated through comparisons with two multicavity PCRV scale models. A reliable and conservative estimate of PCRV behavior under overpressure is obtained

  1. The computer code system for reactor radiation shielding in design of nuclear power plant

    International Nuclear Information System (INIS)

    Li Chunhuai; Fu Shouxin; Liu Guilian

    1995-01-01

    The computer code system used in reactor radiation shielding design of nuclear power plant includes the source term codes, discrete ordinate transport codes, Monte Carlo and Albedo Monte Carlo codes, kernel integration codes, optimization code, temperature field code, skyshine code, coupling calculation codes and some processing codes for data libraries. This computer code system has more satisfactory variety of codes and complete sets of data library. It is widely used in reactor radiation shielding design and safety analysis of nuclear power plant and other nuclear facilities

  2. 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.

  3. Method for the construction of a nuclear reactor with a prestressed concrete pressure vessel

    International Nuclear Information System (INIS)

    Schoening, J.; Schwiers, H.G.

    1981-01-01

    Method for the construction of nuclear reactors with prestressed concrete pressure vessel, providing during the initial stage of construction of the prestressed concrete pressure vessel a support structure around the liner. This enables an early mounting of core components in clean conditions as well as load reductions for final concreting in layers of the prestressed concrete pressure vessel. By applying the support structure, the overall assembly time of these nuclear power plant is considerably reduced without extra cost. (orig.) [de

  4. Safety evaluation for packaging (onsite) for the concrete-shielded RH TRU drum for the 327 Postirradiation Testing Laboratory

    International Nuclear Information System (INIS)

    Smith, R.J.

    1998-01-01

    This safety evaluation for packaging authorizes onsite transport of Type B quantities of radioactive material in the Concrete Shielded Remote-Handled Transuranic Waste (RH TRU) Drum per HNF-PRO-154, Responsibilities and Procedures for all Hazardous Material Shipments. The drum will be used for transport of 327 Building legacy waste from the 300 Area to a solid waste storage facility on the Hanford Site

  5. Safety evaluation for packaging (onsite) for concrete-shielded RHTRU waste drum for the 327 postirradiation testing laboratory

    Energy Technology Data Exchange (ETDEWEB)

    Adkins, H.E.

    1996-10-29

    This safety evaluation for packaging authorizes onsite transport of Type B quantities of radioactive material in the Concrete- Shielded Remote-Handled Transuranic Waste (RH TRU) Drum per WHC-CM-2-14, Hazardous Material Packaging and Shipping. The drum will be used for transport of 327 Building legacy waste from the 300 Area to the Transuranic Waste Storage and Assay Facility in the 200 West Area and on to a Solid Waste Storage Facility, also in the 200 Area.

  6. Monte Carlo based radial shield design of typical PWR reactor

    Energy Technology Data Exchange (ETDEWEB)

    Gul, Anas; Khan, Rustam; Qureshi, M. Ayub; Azeem, Muhammad Waqar; Raza, S.A. [Pakistan Institute of Engineering and Applied Sciences, Islamabad (Pakistan). Dept. of Nuclear Engineering; Stummer, Thomas [Technische Univ. Wien (Austria). Atominst.

    2017-04-15

    This paper presents the radiation shielding model of a typical PWR (CNPP-II) at Chashma, Pakistan. The model was developed using Monte Carlo N Particle code [2], equipped with ENDF/B-VI continuous energy cross section libraries. This model was applied to calculate the neutron and gamma flux and dose rates in the radial direction at core mid plane. The simulated results were compared with the reference results of Shanghai Nuclear Engineering Research and Design Institute (SNERDI).

  7. Benchmark experiments of dose distributions in phantom placed behind iron and concrete shields at the TIARA facility

    International Nuclear Information System (INIS)

    Nakane, Yoshihiro; Sakamoto, Yukio; Tsuda, Shuichi

    2004-01-01

    To verify the calculation methods used for the evaluations of neutron dose at the radiation shielding design of the high-intensity proton accelerator facility (J-PARC), dose distributions in a plastic phantom of 30x30x30 cm 3 slab placed behind iron and concrete test shields were measured by using a tissue equivalent proportional counter for 65-MeV quasi-monoenergetic neutrons generated from the 7 Li(p,n) reactions with 68-MeV protons at the TIARA facility. Dose distributions in the phantom were calculated by using the MCNPX and the NMTC/JAM-MCNP codes with the flux-to-dose conversion coefficients prepared for the shielding design of the facility. The comparison results show the calculated results were in good agreement with the measured ones within 20%. (author)

  8. Design of the prestressed concrete reactor vessel for gas-cooled heating reactors

    International Nuclear Information System (INIS)

    Becker, G.; Notheisen, C.; Steffen, G.

    1987-01-01

    The GHR pebble bed reactor offers a simple, safe and economic possibility of heat generation. An essential component of this concept is the prestressed concrete reactor vessel. A system of cooling pipes welded to the outer surface of the liner is used to transfer the heat from the reactor to the intermediate circuit. The high safety of this vessel concept results from the clear separation of the functions of the individual components and from the design principle of the prestressed conncrete. The prestressed concrete structure is so designed that failure can be reliably ruled out under all operating and accident conditions. Even in the extremely improbable event of failure of all decay heat removal systems when decay heat and accumulated heat are transferred passively by natural convection only, the integrity of the vessel remains intact. For reasons of plant availability the liner and the liner cooling system shall be designed so as to ensure safe elimination of failure over the total operating life. The calculations which were peformed partly on the basis of extremely adverse assumption, also resulted in very low loads. The prestressed concrete vessel is prefabricated to the greatest possible extent. Thus a high quality and optimized fabrication technology can be achieved especially for the liner and the liner cooling system. (orig./HP)

  9. Radiation safety aspects during nondestructive testing of reactor shielding components by gamma radiometry

    International Nuclear Information System (INIS)

    Viswanathan, S.; Jose, M.T.; Venkatraman, B.

    2016-01-01

    In nuclear facilities, effective shielding of radioactive components and structures are essential to ensure radiation protection to operating personnel. The shield structures are made of lead, steel and concrete with varying thickness of up to 1200 mm. It needs to be verified for shielding integrity, presence of voids, blowholes and defects to avoid exposure to workers and to public at large. Radiometry using gamma source serves as excellent tool for non-destructive examination of such structures and components. Gamma sources of high activity up to 50 Curies (gamma camera type) depending on the thickness of component have to be used. During the testing exposure to the operating personnel needs to be minimized, this requires certain safety procedures to be followed. This paper focuses the methodology to be adapted by means of selection of source, effective training of personnel, compliance with safety requirements and maintenance of source devices

  10. Shear strength of end slabs of prestressed concrete nuclear reactor vessels

    International Nuclear Information System (INIS)

    Reins, J.D.; Quiros, J.L. Jr.; Schnobrich, W.C.; Sozen, M.A.

    1976-07-01

    The report summarizes the experimental and part of the analytical work carried out in connection with an investigation of the structural strength of prestressed concrete reactor vessels. The project is part of the Prestressed Concrete Reactor Vessel Program of the Oak Ridge National Laboratory sponsored by ERDA. The objective of the current phase of the work is to develop procedures to determine the shear strength of flat end slabs of reactor vessels with penetrations

  11. Acoustic emission technique for leak detection in an end shield of a pressurized heavy water reactor

    International Nuclear Information System (INIS)

    Kalyanasundaram, P.; Jayakumar, T.; Raj, B.

    1989-01-01

    This paper discusses the successful application of the Acoustic Emission Technique (AET) for detection and location of leak paths present on the inaccessible side of an end shield of a Pressurized Heavy Water Reactor (PHWR). The methodology was based on the fact that air and water leak AE signals have different characteristic features. Baseline data was generated from a sound end-shield of a PHWR for characterizing the background noise. A mock up end-shield system with saw cut leak paths was used to verify the validity of the methodology. It was found that air leak signals under pressurisation (as low as 3 psi) could be detected by frequency domain analysis. Signals due to air leaks from various locations of a defective end-shield were acquired and analysed. It was possible to detect and locate leak paths. Presence of detected leak paths were further confirmed by alternate test. (orig.)

  12. Prestressed concrete nuclear reactor containment structures. Revision 3

    International Nuclear Information System (INIS)

    Reuter, H.R.; Chang-Lo, P.L.C.; Pfeifer, B.W.; Shah, G.H.; Whitcraft, J.S.

    1975-02-01

    A discussion of the techniques and procedures used for the design of prestressed concrete nuclear reactor containment structures is presented. A physical description of Bechtel designed containment structures is presented. The design bases and load combinations are given for anticipated conditions of service. Reference design documents which include industry codes, specifications, AEC Regulatory Guides, Bechtel Topical Reports and additional criteria as appropriate to containment design are listed. Stepwise procedures typically followed by Bechtel for design of containments is discussed and design examples are presented. A description of currently used analytical methods and the practical application of these methods for containment design is also presented. The principal containment construction materials are identified and codes of practice pertaining to construction procedures are listed. Preoperational structural testing procedures and post-operational surveillance programs are furnished along with results of tests on completed containment structures. (U.S.)

  13. High performance inboard shield design for the compact TIBER-II test reactor: Appendix A-2

    International Nuclear Information System (INIS)

    El-Guebaly, L.A.; Sviatoslavsky, I.N.

    1987-01-01

    The compactness of the TIBER-II reactor has placed a premium on the design of a high performance inboard shield to protect the inner legs of the toroidal field (TF) coils. The available space for shield is constrained to 48 cm and the use of tungsten is mandatory to protect the magnet against the 1.53 MW/m 2 neutron wall loading. The primary requirement for the shield is to limit the fast neutron fluence to 10 19 n/cm 2 . In an optimization study, the performance of various candidate materials for protecting the magnet was examined. The optimum shield consists of a 40 cm thick W layer, followed by an 8 cm thick H 2 O/LiNO 3 layer. The mechanical design of the shield calls for tungsten blocks within SS stiffened panels. All the coolant channels are vertical with more of them in the front where there is a high heat load. The coolant pressure is 0.2 MPa and the maximum structural surface temperature is 0 C. The effects of the detailed mechanical design of the shield and the assembly gaps between the shield sectors on the damage in the magnet were analyzed and peaking factors of ∼2 were found at the hot spots. 2 refs., 6 figs., 2 tabs

  14. Comparison of neutron fluxes obtained by 2-D and 3-D geometry with different shielding libraries in biological shield of the TRIGA MARK II reactor

    International Nuclear Information System (INIS)

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

    2003-01-01

    Neutron fluxes in different spatial locations in biological shield are obtained with TORT code (TORT-Three Dimensional Oak Ridge Discrete Ordinates Neutron/Photon Transport Code). Libraries used with TORT code were BUGLE-96 library (coupled library with 47 neutron groups and 20 gamma groups) and VITAMIN-B6 library (coupled library with 199 neutron groups and 42 gamma groups). BUGLE-96 library is derived from VITAMIN-B6 library. 2-D and 3-D models for homogeneous type of problem (without inserted beam port 4) and problem with asymmetry (non-homogeneous problem; inserted beam port 4, filled with different materials) were of interest for neutron flux calculation. The main purpose is to verify the possibility for using 2-D approximation model instead of large 3-D model in some calculations. Another purpose of this paper was to compare neutron spectral constants obtained from neutron fluxes (3-D model) determined with smaller BUGLE-96 library with new constants obtained from fluxes calculated with bigger VITAMIN-B6 library. These neutron spectral constants are used in isotopic calculation with SCALE code package (ORIGEN-S). In past only neutron spectral constants determined by neutron fluxes from BUGLE-96 library were used. Experimental results used for isotopic composition comparison are available from irradiation experiment with selected type of concrete and other materials in beam port 4 (irradiation channel 4) in TRIGA Mark II reactor. These experimental results were used as a benchmark in this paper. (author)

  15. Prestressed concrete reactor vessel thermal cylinder model study

    International Nuclear Information System (INIS)

    Callahan, J.P.; Canonico, D.A.; Richardson, M.; Corum, J.M.; Dodge, W.G.; Robinson, G.C.; Whitman, G.D.

    1977-01-01

    The thermal cylinder experiment was designed both to provide information for evaluating the capability of analytical methods to predict the time-dependent stress-strain behavior of a 1 / 6 -scale model of the barrel section of a single-cavity prestressed concrete reactor vessel and to demonstrate the structural behavior under design and off-design thermal conditions. The model was a thick-walled cylinder having a height of 1.22 m, a thickness of 0.46 m, and an outer diameter of 2.06 m. It was prestressed both axially and circumferentially and subjected to 4.83 MPa internal pressure together with a thermal crossfall imposed by heating the inner surface to 338.8 K and cooling the outer surface to 297.1 K. The initial 460 days of testing were divided into time periods that simulated prestressing, heatup, reactor operation, and shutdown. At the conclusion of the simulated operating period, the model was repressurized and subjected to localized heating at 505.4 K for 84 days to produce an off-design hot-spot condition. Comparisons of experimental data with calculated values obtained using the SAFE-CRACK finite-element computer program showed that the program was capable of predicting time-dependent behavior in a vessel subjected to normal operating conditions, but that it was unable to accurately predict the behavior during off-design hot-spot heating. Readings made using a neutron and gamma-ray backscattering moisture probe showed little, if any, migration of moisture in the concrete cross section. Destructive examination indicated that the model maintained its basic structural integrity during localized hot-spot heating

  16. Radiation shielding and effective atomic number studies in different types of shielding concretes, lead base and non-lead base glass systems for total electron interaction: A comparative study

    International Nuclear Information System (INIS)

    Kurudirek, Murat

    2014-01-01

    Highlights: • Radiation shielding calculations for concretes and glass systems. • Assigning effective atomic number for the given materials for total electron interaction. • Glass systems generally have better shielding ability than concretes. - Abstract: Concrete has been widely used as a radiation shielding material due to its extremely low cost. On the other hand, glass systems, which make everything inside visible to observers, are considered as promising shielding materials as well. In the present work, the effective atomic numbers, Z eff of some concretes and glass systems (industrial waste containing glass, Pb base glass and non-Pb base glass) have been calculated for total electron interaction in the energy region of 10 keV–1 GeV. Also, the continuous slowing down approximation (CSDA) ranges for the given materials have been calculated in the wide energy region to show the shielding effectiveness of the given materials. The glass systems are not only compared to different types of concretes but also compared to the lead base glass systems in terms of shielding. Moreover, the obtained results for total electron interaction have been compared to the results for total photon interaction wherever possible. In general, it has been observed that the glass systems have superior properties than most of the concretes over the high-energy region with respect to the electron interaction. Also, glass systems without lead show better electron stopping than lead base glasses at some energy regions as well. Along with the photon attenuation capability, it is seen that Fly Ash base glass systems have not only greater electron stopping capability but also have greater photon attenuation especially in high energy region when compared with standard shielding concretes

  17. Radiation shielding and effective atomic number studies in different types of shielding concretes, lead base and non-lead base glass systems for total electron interaction: A comparative study

    Energy Technology Data Exchange (ETDEWEB)

    Kurudirek, Murat, E-mail: mkurudirek@gmail.com

    2014-12-15

    Highlights: • Radiation shielding calculations for concretes and glass systems. • Assigning effective atomic number for the given materials for total electron interaction. • Glass systems generally have better shielding ability than concretes. - Abstract: Concrete has been widely used as a radiation shielding material due to its extremely low cost. On the other hand, glass systems, which make everything inside visible to observers, are considered as promising shielding materials as well. In the present work, the effective atomic numbers, Z{sub eff} of some concretes and glass systems (industrial waste containing glass, Pb base glass and non-Pb base glass) have been calculated for total electron interaction in the energy region of 10 keV–1 GeV. Also, the continuous slowing down approximation (CSDA) ranges for the given materials have been calculated in the wide energy region to show the shielding effectiveness of the given materials. The glass systems are not only compared to different types of concretes but also compared to the lead base glass systems in terms of shielding. Moreover, the obtained results for total electron interaction have been compared to the results for total photon interaction wherever possible. In general, it has been observed that the glass systems have superior properties than most of the concretes over the high-energy region with respect to the electron interaction. Also, glass systems without lead show better electron stopping than lead base glasses at some energy regions as well. Along with the photon attenuation capability, it is seen that Fly Ash base glass systems have not only greater electron stopping capability but also have greater photon attenuation especially in high energy region when compared with standard shielding concretes.

  18. Advanced Neutron Source Reactor zoning, shielding, and radiological optimization guide

    International Nuclear Information System (INIS)

    Westbrook, J.L.; DeVore, J.R.

    1995-08-01

    In the design of major nuclear facilities, it is important to protect both humans and equipment excessive radiation dose. Past experience has shown that it is very effective to apply dose reduction principles early in the design of a nuclear facility both to specific design features and to the manner of operation of the facility, where they can aid in making the facility more efficient and cost-effective. Since the appropriate choice of radiological controls and practices varies according to the case, each area of the facility must be analyzed for its radiological impact, both by itself and in interactions with other areas. For the Advanced Neutron Source (ANS) project, a large relational database will be used to collect facility information by system and relate it to areas. The database will also hold the facility dose and shielding information as it is produced during the design process. This report details how the ANS zoning scheme was established and how the calculation of doses and shielding are to be done

  19. Seismic and cask drop excitation evaluation of the tower shielding reactor

    International Nuclear Information System (INIS)

    Harris, S.P.; Stover, R.L.; Johnson, J.J.; Sumodobila, B.N.

    1989-01-01

    During the current shutdown of the Tower Shielding Reactor II (TSR-II), analyses were performed to determine the effect of nearby cask drops on the structural and mechanical integrity of the reactor. This evaluation was then extended to include the effects of earthquakes. Several analytic models were developed to simulate the effects of earthquake and cask drop excitation. A coupled soil-structure model was developed. As a result of the analyses, several hardware modifications and enhancements were implemented to ensure reactor integrity during future operations. 6 figs

  20. Seismic and cask drop excitation evaluation of the Tower Shielding Reactor

    International Nuclear Information System (INIS)

    Stover, R.L.; Harris, S.P.; Johnson, J.J.; Sumodobila, B.N.

    1989-01-01

    During the current shutdown of the Tower Shielding Reactor II (TSR-II), analyses were performed to determine the effect of nearby cask drops on the structural and mechanical integrity of the reactor. This evaluation was then extended to include the effects of earthquakes. Several analytic models were developed to simulate the effects of earthquake and cask drop excitation. A coupled soil-structure model was developed. As a result of the analyses, several hardware modifications and enhancements were implemented to ensure reactor integrity during future operations

  1. Boiling water reactor radiation shielded Control Rod Drive Housing Supports

    Energy Technology Data Exchange (ETDEWEB)

    Baversten, B.; Linden, M.J. [ABB Combustion Engineering Nuclear Operations, Windsor, CT (United States)

    1995-03-01

    The Control Rod Drive (CRD) mechanisms are located in the area below the reactor vessel in a Boiling Water Reactor (BWR). Specifically, these CRDs are located between the bottom of the reactor vessel and above an interlocking structure of steel bars and rods, herein identified as CRD Housing Supports. The CRD Housing Supports are designed to limit the travel of a Control Rod and Control Rod Drive in the event that the CRD vessel attachement went to fail, allowing the CRD to be ejected from the vessel. By limiting the travel of the ejected CRD, the supports prevent a nuclear overpower excursion that could occur as a result of the ejected CRD. The Housing Support structure must be disassembled in order to remove CRDs for replacement or maintenance. The disassembly task can require a significant amount of outage time and personnel radiation exposure dependent on the number and location of the CRDs to be changed out. This paper presents a way to minimize personal radiation exposure through the re-design of the Housing Support structure. The following paragraphs also delineate a method of avoiding the awkward, manual, handling of the structure under the reactor vessel during a CRD change out.

  2. Fiber reinforced concrete as a material for nuclear reactor containment buildings

    International Nuclear Information System (INIS)

    Mallikarjuna; Banthia, N.; Mindess, S.

    1991-01-01

    The fiber reinforced concrete as a constructional material for nuclear reactor containment buildings calls for an examination of its individual characteristics and potentialities due to its inherent superiority over normal plain and reinforced concrete. In the present investigation, first, to study the static behavior of straight, hooked-end and crimped fibers, recently developed nonlinear three-dimensional interface (contact) element has been used in conjunction with the eight nodded hexahedron and two nodded bar elements for concrete and steel fiber respectively. Then impact tests were carried out on fiber reinforced concrete beams with an instrumented drop weight impact machine. Two different concrete mixes were tested: normal strength and high strength concrete specimens. Fibers in the concrete mix found to significantly increase the ductility and the impact resistance of the composite. Deformed fibers increase peak pull-out load and pull-out distance, and perform better in the steel fiber reinforced concrete (SFRC) structures. (author)

  3. Study of temperature effect on the physical properties of ilmenite-serpentine heat resistant concrete radiation shields

    International Nuclear Information System (INIS)

    Kany, A.M.I.; EL-Fouly, M.M.; EL-Gohary, M.I.; Makatious, A.S.; Kamal, S.M.

    1990-01-01

    A series of experimental studies have been carried out to determine the change in unit weigh, compressive strength, water content and neutron macroscopic cross section of a new type of concrete shields made from egyptian ilmenite and serpentine ores when heated for long period at temperatures up to 600 degree C. Results show that the unit weight of the cure concrete has a value of 2.98 Ton/M 3 and decreases with increasing temperature, while the compressive strength reaches a maximum value of 19 Ton/M 2 at 100 degree C. The differential thermal analysis (D.T.A.) of this concrete shows three endothermic peaks at 100 degree C, 48 degree C and 740 degree C. Also, the thermogravimetry analysis (T.G.A.) shows that the cure concrete retains about 11% water content of the total sample weigh and still retains 4.5% of its initial value when heated for long period at 600 degree C. Results also show that the neutron macroscopic cross section (for neutrons of energies < 1 MeV) of the ilmenite-serpentine heat resistant concrete decreases to 18.6% of its initial value after heating to 600 degree C

  4. Shielding properties of the ordinary concrete loaded with micro- and nano-particles against neutron and gamma radiations.

    Science.gov (United States)

    Mesbahi, Asghar; Ghiasi, Hosein

    2018-06-01

    The shielding properties of ordinary concrete doped with some micro and nano scaled materials were studied in the current study. Narrow beam geometry was simulated using MCNPX Monte Carlo code and the mass attenuation coefficient of ordinary concrete doped with PbO 2 , Fe 2 O 3 , WO 3 and H 4 B (Boronium) in both nano and micro scales was calculated for photon and neutron beams. Mono-energetic beams of neutrons (100-3000 keV) and photons (142-1250 keV) were used for calculations. The concrete doped with nano-sized particles showed higher neutron removal cross section (7%) and photon attenuation coefficient (8%) relative to micro-particles. Application of nano-sized material in the composition of new concretes for dual protection against neutrons and photons are recommended. For further studies, the calculation of attenuation coefficients of these nano-concretes against higher energies of neutrons and photons and different particles are suggested. Copyright © 2018 Elsevier Ltd. All rights reserved.

  5. Light Water Reactor Sustainability Program: survey of models for concrete degradation

    International Nuclear Information System (INIS)

    2014-01-01

    Concrete has been used in the construction of nuclear facilities because of two primary properties: its structural strength and its ability to shield radiation. Concrete structures have been known to last for hundreds of years, but they are also known to deteriorate in very short periods of time under adverse conditions. The use of concrete in nuclear facilities for containment and shielding of radiation and radioactive materials has made its performance crucial for the safe operation of the facility. The goal of this report is to review and document the main aging mechanisms of concern for concrete structures in nuclear power plants (NPPs) and the models used in simulations of concrete aging and structural response of degraded concrete structures. This is in preparation for future work to develop and apply models for aging processes and response of aged NPP concrete structures in the Grizzly code. To that end, this report also provides recommendations for developing more robust predictive models for aging effects of performance of concrete.

  6. Light Water Reactor Sustainability Program: Survey of Models for Concrete Degradation

    Energy Technology Data Exchange (ETDEWEB)

    Spencer, Benjamin W. [Idaho National Lab. (INL), Idaho Falls, ID (United States). Fuel Modeling and Simulation; Huang, Hai [Idaho National Lab. (INL), Idaho Falls, ID (United States). Energy and Environment Science and Technology

    2014-08-01

    Concrete is widely used in the construction of nuclear facilities because of its structural strength and its ability to shield radiation. The use of concrete in nuclear facilities for containment and shielding of radiation and radioactive materials has made its performance crucial for the safe operation of the facility. As such, when life extension is considered for nuclear power plants, it is critical to have predictive tools to address concerns related to aging processes of concrete structures and the capacity of structures subjected to age-related degradation. The goal of this report is to review and document the main aging mechanisms of concern for concrete structures in nuclear power plants (NPPs) and the models used in simulations of concrete aging and structural response of degraded concrete structures. This is in preparation for future work to develop and apply models for aging processes and response of aged NPP concrete structures in the Grizzly code. To that end, this report also provides recommendations for developing more robust predictive models for aging effects of performance of concrete.

  7. Shielding properties of protective thin film coatings and blended concrete compositions for high level waste storage packages

    International Nuclear Information System (INIS)

    Fusco, Michael A.; Winfrey, Leigh; Bourham, Mohamed A.

    2016-01-01

    Highlights: • Measured linear attenuation coefficients are the same for bare and coated steels. • Gamma mean free path is much larger than coating thickness; buildup is negligible. • ‘Concrete-6’ reduces exposure rate outside spent fuel cask significantly over ordinary concrete. - Abstract: Various thin film coatings have been proposed to protect stainless steel high level waste (HLW) containers from premature failure due to localized corrosion, hydrogen embrittlement, and mechanical wear. These coatings include TiN, ZrO 2 , MoS 2 , TiO 2 , and Al 2 O 3 , to be deposited either in multiple layers or as a thicker, single-layer composite. Linear attenuation coefficients of these materials have been simulated using MicroShield and measured experimentally for various photon energies. Additionally, spent fuel casks with overpacks made of two different types of concrete were simulated to compare exposure rate at the cask surface. In the energy range that is significant for high level waste storage all coating materials possess very similar attenuation behavior. A specialty concrete, containing magnetite (Fe 3 O 4 ) and lead oxide (PbO), reduces the exposure rate at the outer surface of the overpack by several orders of magnitude. The higher-Z elements not present in ordinary concrete greatly increase attenuation of intermediate-energy gammas (0.4–1.0 MeV). The thin film coatings do not affect the shielding capabilities of the HLW packaging, as their total proposed thickness is nearly three orders of magnitude less than the mean free path (MFP) of the primary photons of interest.

  8. Problems related to the definition of the shielding of a large fast power reactor

    International Nuclear Information System (INIS)

    Moreau, J.

    Solutions for the shielding of a 1000 MW(e) power plant in the same technological line as Phenix are given. They have been evaluated with a monodimensional transport code. The choice is based on the comparison of their efficiency towards neutrons and on the consequences of their characteristics on the conception of the reactor tank. A few economical considerations give an idea of the influence of the choice in shielding on the cost of the power plant. At last the problem of the optimization possibilities is approached from the designer's point of view

  9. The quality control for biological-shield heavy concrete construction of nuclear power project

    International Nuclear Information System (INIS)

    Sun Hongjun; Ma Xinchao

    2012-01-01

    The paper introduces the function and characteristics of biological protective heavy-concrete, and its main application scope and role in Fangjiashan nuclear power project. From the aspects of raw material selection, mixing ratio test, heavy concrete production, the paper discusses the main control points of heavy concrete construction process, points out the basic characteristics of heavy concrete construction, and put forward measures to prevent density non-uniformity during heavy concrete construction and to control slump during transportation. Results prove that reasonable construction process control can assure the engineering quality. (authors)

  10. 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)

  11. 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

  12. Temperature field in the bottom of concrete reactor vessel; Temperaturno polje u podu betonskog reaktorskog suda

    Energy Technology Data Exchange (ETDEWEB)

    Jovasevic, V; Tosic, D; Zaric, S; Maksimovic, Lj [Boris Kidric Institute of nuclear sciences, Vinca, Belgrade (Yugoslavia)

    1969-07-01

    This paper contains detailed scheme of reactor bottom vessel made of concrete and the results of calculated relevant temperature distribution. Method applied for calculation is described taking into account all relevant factors and assuming that thermal conductivity of concrete is homogeneous and independent of temperature.

  13. Fast reactor shield sensitivity studies for steel--sodium--iron systems

    International Nuclear Information System (INIS)

    Oblow, E.M.; Weisbin, C.R.

    1977-01-01

    A study was made of the adequacy of the current ENDF/B-IV sodium and iron neutron cross section data files for fast reactor shield design work. Experimental data from 21 fast reactor shield configurations containing large thicknesses of steel, sodium, and iron were analyzed with discrete ordinates calculations and sensitivity methods to assess the data files. This study represents the largest full-scale sensitivity analysis of benchmark quality experimental data to date. Included in the sensitivity studies were the results of the new cross section adjustment algorithms added to the FORSS code system. Conclusions were drawn about the need for more accurate data for sodium and iron elastic and discrete inelastic cross sections above 1 MeV and the values of the total cross section in the vicinity of important minima

  14. Studying the effect of nano lead compounds additives on the concrete shielding properties for γ-rays

    International Nuclear Information System (INIS)

    Hassan, H.E.; Badran, H.M.; Aydarous, A.; Sharshar, T.

    2015-01-01

    In the present work the effect of concrete incorporation with two types of nano-lead compounds on its γ-ray shielding characteristics is investigated. The concrete samples were prepared according to the local standards of building materials and doped by different percentages of PbO and PbTiO_3 nano powders which were prepared using co-precipitation and oxalate precursor techniques, respectively. In addition, commercial PbO_2 powder additive was used to check the effect of particle size on concrete attenuation properties. The phase composition and particle size of all the lead-oxide additives were confirmed by XRD and TEM imaging. The γ-rays attenuation coefficients were measured as a function of the additive percentage of lead compounds for γ-ray energies of 662, 1173 and 1332 keV using "1"3"7Cs and "6"0Co sources. The microstructure changes occurred in the concrete samples doped with Pb compounds additives were probed using the positron annihilation spectroscopy (PAS) and the results were compared with that for normal concrete. The obtained data revealed that the overall defect density of the investigated samples, as seen by the positrons, decreases with increasing the nano-PbO contents which is in agreement with the determined values of the samples apparent densities. It was found that the γ-ray attenuation coefficient of concrete doped by nano-PbO is improved. The results are explained in the view of the fine structure enhanced modification and its impact on the γ-ray interaction probability at different energies.

  15. Minimum weight design of prestressed concrete reactor pressure vessels

    International Nuclear Information System (INIS)

    Boes, R.

    1975-01-01

    A method of non-linear programming for the minimization of the volume of rotationally symmetric prestressed concrete reactor pressure vessels is presented. It is assumed that the inner shape, the loads and the degree of prestressing are prescribed, whereas the outer shape is to be detemined. Prestressing includes rotational and vertical tension. The objective function minimizes the weight of the PCRV. The constrained minimization problem is converted into an unconstrained problem by the addition of interior penalty functions to the objective function. The minimum is determined by the variable metric method (Davidson-Fletcher-Powell), using both values and derivatives of the modified objective function. The one-dimensional search is approximated by a method of Kund. Optimization variables are scaled. The method is applied to a pressure vessel like for THTR. It is found that the thickness of the cylindrical wall may be reduced considerably for the load cases considered in the optimization. The thickness of the cover is reduced slightly. The largest reduction in wall thickness occurs at the junction of wall and cover. (Auth.)

  16. Determination of boron in Jabroc wood used as a shielding material in nuclear reactors

    International Nuclear Information System (INIS)

    Kamble, Granthali S.; Manisha, V.; Venkatesh, K.

    2015-01-01

    Jabroc are non-impregnated, densified wood laminates developed commercially for a wide range of industrial applications. Jabroc can be used with other neutron shielding materials such as Lead to form complex shielding structures. Its relative light weight and cleanliness in handling are additional features that make it a suitable candidate for the standard design of neutron shielding equipment. Jabroc can also be impregnated with Boron up to a maximum of 4% to be used in areas where Gamma radiation produced on Neutron capture reaches unacceptable dose rates. Boron impregnated Jabroc wood finds application in TAPS 3 and 4 as a shielding material for the Ion Chambers and the Horizontal Flux Units (HFU). The shielding property of this material is optimized by incorporating requisite amount of boron in wood. Boron content in this material has to be determined accurately prior to its use in the nuclear reactors. In this work a method was standardized to determine boron in Jabroc wood samples to check for conformance to specifications. The wood sample flakes were wetted with saturated barium hydroxide solution and dries under IR. The sample was ashed in a muffle furnace at 600℃ for 2 h

  17. 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)

  18. TIBER II/ETR [Engineering Test Reactor] nuclear shielding and optional tritium breeding system: An overview

    International Nuclear Information System (INIS)

    Lee, J.D.; Sawan, M.

    1987-01-01

    TIBER II, the Tokamak Ignition/Burn Experimental Reactor II, is a design concept developed as the US candidate for an International Engineering Test Reactor (ETR). An important objective of this design is to minimize cost by minimizing major radius while providing a wall loading greater than 1.0 MW/m2 and a total fluence greater than 3.0 MWY/m2 needed for blanket module testing. The shielding required for the superconducting TF coils is an important element in setting TIBER II's 3.0m major radius. 6 refs., 1 fig., 1 tab

  19. Comparison of some lead and non-lead based glass systems, standard shielding concretes and commercial window glasses in terms of shielding parameters in the energy region of 1 keV-100 GeV: A comparative study

    International Nuclear Information System (INIS)

    Kurudirek, Murat; Ozdemir, Yueksel; Simsek, Onder; Durak, Ridvan

    2010-01-01

    The effective atomic numbers, Z eff of some glass systems with and without Pb have been calculated in the energy region of 1 keV-100 GeV including the K absorption edges of high Z elements present in the glass. Also, these glass systems have been compared with some standard shielding concretes and commercial window glasses in terms of mean free paths and total mass attenuation coefficients in the continuous energy range. Comparisons with experiments were also provided wherever possible for glasses. It has been observed that the glass systems without Pb have higher values of Z eff than that of Pb based glasses at some high energy regions even if they have lower mean atomic numbers than Pb based glasses. When compared with some standard shielding concretes and commercial window glasses, generally it has been shown that the given glass systems have superior properties than concretes and window glasses with respect to the radiation-shielding properties, thus confirming the availability of using these glasses as substitutes for some shielding concretes and commercial window glasses to improve radiation-shielding properties in the continuous energy region.

  20. FFTF reactor-characterization program: gamma-ray measurements and shield characterization

    International Nuclear Information System (INIS)

    Bunch, W.L.; Moore, F.S. Jr.

    1983-02-01

    A series of experiments is to be made during the acceptance test program of the Fast Flux Test Facility (FFTF) to measure the gamma ray characteristics of the Fast Test Reactor (FTR) and to establish the performance characteristics of the reactor shield. These measurements are a part of the FFTF Reactor Characterization Program (RCP). Detailed plans have been developed for these experiments. During the initial phase of the Characteristics Program, which will be carried out in the In-Reactor Thimble (IRT), both active and passive measurement methods will be employed to obtain as much information concerning the gamma ray environment as is practical. More limited active gamma ray measurements also will be made in the Vibration Open Test Assembly (VOTA)

  1. Radiochemical analysis of concrete samples for decommission of nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Zapata-Garcia, Daniel; Wershofen, Herbert [Physikalisch-Technische Bundesanstalt (PTB), Bundesallee 100 38116, Braunschweig (Germany); Larijani, Cyrus; Sobrino-Petrirena, Maitane; Garcia-Miranda, Maria; Jerome, Simon M. [National Physical Laboratory (NPL), Hampton Road, Teddington, Middlesex, TW11 0LW (United Kingdom)

    2014-07-01

    Decommissioning of the oldest nuclear power reactors are some of the most challenging technological legacy issues many countries will face in forthcoming years, as many power reactors reach the end of their design lives. Decommissioning of nuclear reactors generates large amounts of waste that need to be classified according to their radioactive content. Approximately 10 % of the contaminated material ends up in different repositories (depending on their level of contamination) while the rest is decontaminated, measured and released into the environment or sent for recycling. Such classification needs to be done accurately in order to ensure that both the personnel involved in decommissioning and the population at large are not needlessly exposed to radiation or radioactive material and to minimise the environmental impact of such work. However, too conservative classification strategies should not be applied, in order to make proper use of radioactive waste repositories since space is limited and the full process must be cost-effective. Implicit in decommissioning and classification of waste is the need to analyse large amounts of material which usually combine a complex matrix with a non-homogeneous distribution of the radionuclides. Because the costs involved are large, it is possible to make great savings by the adoption of best available practices, such as the use of validated methods for on-site measurements and simultaneous determination of more than one radionuclide whenever possible. The work we present deals with the development and the validation of a procedure for the simultaneous determination of {sup 241}Am, plutonium isotopes, uranium isotopes and {sup 90}Sr in concrete samples. Samples are firstly ground and fused with LiBO{sub 2} and Li{sub 2}B{sub 4}O{sub 7}. After dissolution of the fused sample, silicate and alkaline elements are removed followed by radiochemical separation of the target radionuclides using extraction chromatography. Measurement

  2. Prestressed concrete reactor vessel for the HHT-670 MW(e) demonstration plant. Pt.1. Design of the multi-cavity prestressed concrete reactor vessel with warm liner

    International Nuclear Information System (INIS)

    Lafitte, R.; Marchand, J.D.

    1979-01-01

    The design studies and tests described in this paper were undertaken as part of ''PROJECT HHT'', a German-Swiss joint effort for the development of high-temperature helium cooled reactors with direct-cycle turbine. The prestressed concrete reactor pressure vessel encloses the core of the reactor itself, the heat exchangers (coolers and recuperators), the helium turbine, the main helium circuit, all nuclear and thermal equipment, and auxiliary reactor cooling equipment. In order to make the liner accessible for inspection, no thermal insulation is provided between the coolant and the liner. The temperature of the helium in contact with the liner is limited to 200 0 C, under all normal operation conditions of the reactor. In the HHT reactor pressure vessel, the resisting structure is protected thermally by a layer of warm concrete between the liner and the structural prestressed concrete. The main features of this pressure vessel are the marked pressure differences in the cavities during normal operation, and the use of warm liner. The objectives of the reference design were chiefly related to the sizing up of the main structure, taking into account the modifications to be expected in the material characteristics as a result of the high temperatures developed

  3. The study on the mechanical characteristics of concrete of nuclear reactor containment structure

    International Nuclear Information System (INIS)

    Jung, W. S.; Kwon, K. J.; Cho, M. S.; Song, Y. C.

    2000-01-01

    Reactor containment structure of nuclear power plant designed by prestressed concrete causes time-dependent prestress loss due to the mechanical characteristics of concrete. Prestress loss strongly affects to the safety factor of structure under the circumstances of designing, construction and inspection. Thus, this study is to investigate the mechanical characteristics of reactor containment concrete structure of Yonggwang No. 5 and 6. In this study, the compressive strength, modulus of elasticity, poisson's ratio and creep test followed by ASTM code are performed to investigate the mechanical characteristics of concrete made by V type cement. Additionally, since creep causes more time-dependent prestress loss than the other, the measurement value from the creep test is compared with the results from the creep prediction equations by KSCE, JSCE, Hansen, ACI and CEB-FIP model for the effective application. Hereafter, the results of this study may enable to assist the calculation effective stress considering time-dependent prestress loss of the prestressed concrete structures

  4. Nuclear design of the blanket/shield system for a Tokamak Experimental Power Reactor

    International Nuclear Information System (INIS)

    Abdou, M.A.

    1976-01-01

    The various options and trade-offs in the nuclear design of the blanket/shield for a Tokamak Experimental Power Reactor (TEPR) are investigated. The TEPR size and cost are particularly sensitive to the blanket/shield thickness, Δ/sub BS/, on the inner side of the torus. Radition damage to the components of the superconducting magnet and refrigeration power requirements set lower limits on Δ/sub BS/. These limits are developed in terms of TEPR design parameters such as the wall loading, duty cycle, and frequency of magnet anneals. The study of the nuclear performance of various material compositions shows that mixtures of tungsten, or tantalum, or stainless-steel alloys and boron carbide require the smallest Δ/sub BS/ for a given attenuation. This Δ/sub BS/ has to be doubled if the low induced activation materials graphite and aluminum are used. The space problems are greatly eased in the Argonne National Laboratory ANL-TEPR reference design by using two separate segments of the blanket/shield. The inner segment occupies the region of the high magnetic field, uses very efficient attenuators (tungsten- or tantalum- or stainless-steel-boron carbide mixtures), and is only 1 m thick. The outer blanket/shield is 131 cm and consists of an optimized composition of stainless steel and boron carbide. For the design parameters of 0.2 MW/m 2 neutron wall loading and 50 percent duty cycle, the reactor components can operate satisfactorily up to (a) 10 yr for the stainless-steel first wall, (b) 10 yr for the superconductor composite after which magnet warmup becomes necessary, and (c) 30 yr for the Mylar insulation. Nuclear heat generation rates in the blanket/shield and magnet are well within the practical limits for heat removal

  5. Containers, particularly prestressed concrete pressure vessels for nuclear reactor plants

    International Nuclear Information System (INIS)

    Schoening, J.; Schwiers, H.G.; Mitterbacher, P.

    1986-01-01

    Pressure and temperature changes act on the liner, which cause differential expansion between the liner and the prestressed concrete. So that there will be no overload or damage to the liner, its anchoring or the concrete structure, cutouts are provided in the concrete at deflection positions of the steel cladding, connections and penetrations. These cut-outs are filled with inserts made of elastic or plastic material. (DG) [de

  6. Improvements at the biological shielding of BNCT research facility in the IEA-R1 reactor

    International Nuclear Information System (INIS)

    Souza, Gregorio Soares de

    2011-01-01

    The technique of neutron capture in boron is a promising technique in cancer treatment, it uses the high LET particles from the reaction 10 B (n, α) 7 Li to destroy cancer cells.The development of this technique began in the mid-'50s and even today it is the object of study and research in various centers around the world, Brazil has built a facility that aims to conduct research in BNCT, this facility is located next to irradiation channel number three at the research nuclear reactor IEA-R1 and has a biological shielding designed to meet the radiation protection standards. This biological shielding was developed to allow them to conduct experiments with the reactor at maximum power, so it is not necessary to turn on and off the reactor to irradiate samples. However, when the channel is opened for experiments the background radiation in the experiments salon increases and this background variation makes it impossible to perform measurements in a neutron diffraction research that utilizes the irradiation channel number six. This study aims to further improve the shielding in order to minimize the variation of background making it possible to perform the research facility in BNCT without interfering with the action of the research group of the irradiation channel number six. To reach this purpose, the code MCNP5, dosimeters and activation detectors were used to plan improvements in the biological shielding. It was calculated with the help of the code an improvement that can reduce the average heat flow in 71.2% ± 13 and verified experimentally a mean reduce of 70 ± 9% in dose due to thermal neutrons. (author)

  7. The application of semianalytic method for calculating the thickness of biological shields of nuclear reactors. Part 2. Attenuation of gamma rays. An example of shield's thickness calculation

    International Nuclear Information System (INIS)

    Lukaszek, W.; Kucypera, S.

    1982-01-01

    The semianalytic method was used for calculating the attenuation of gamma rays and the thickness of biological shield of graphite moderated reactor. A short description of computer code as well as the exemplary results of calculations are given. (A.S.)

  8. Shielding research in France

    Energy Technology Data Exchange (ETDEWEB)

    Lafore, P

    1964-10-01

    Shielding research as an independent subject in France dates from 1956. The importance of these studies has been reflected in the contribution which they have made to power reactor design and in the resultant savings in expenditure for civil engineering and machinery for the removal of mobile shields. The Reactor Shielding Research Division numbers approximately 60 persons and uses several experimental facilities. These include: NAIADE I, installed near the ZOE reactor and operating with a natural uranium slab 2 cm thick (an effective diameter of 60 cm is the one most commonly used); the TRITON pool-type reactor, mainly used in shielding studies, includes an active-water loop, by means of which the secondary shields required for light-water reactors can be studied; core, NEREIDE, which is situated near a 2 m x 2 m aluminium window enables a large neutron source to be placed in a compartment without water in which large-scale mock-ups can be mounted for the study, in particular, of neutron diffusion in large cavities, and of reactor shielding of greater thickness than that in NAIADE I; SAMES 600 keV accelerator is used for monoenergetic neutron studies. Instrumentation studies are an important part of the work, mainly in the measurement of fast neutrons and their spectra by activation detectors. Of late, attention has been directed towards the use of (n, n') (rhodium) reactions and of heavy detectors for low-flux measurements. The simultaneous use of a large number of detectors poses automation problems. With our installation we can count 16 detectors simultaneously. Neutron spectrum studies are conducted with nuclear emulsions and a lithium-6 semiconductor spectrometer. As to the materials used, the research carried out in France involves chiefly graphite, iron and concrete at various temperatures up to 800 deg C. Different compounds, borated and non-borated and of densities up to between 1 and 9 are under consideration. Problems connected with applications are

  9. Roles of concrete technology for containment of radioactive contaminants

    International Nuclear Information System (INIS)

    Kitsutaka, Yoshinori; Imamoto, Keiichi

    2014-01-01

    A large amount of radioactive materials was emitted in the environment by the reactor accident at Fukushima Daiichi Nuclear Power Plant. Nuclear debris still remains in the reactor container. An investigative committee was organized in Japan Concrete Institute to study on the containment of radioactive materials and the safe utilization of concrete materials. We have investigated the effect of the hydrogen explosion upon the property of concrete and the transfer of materials into the concrete. We also present the outline of the advice made by Japan Concrete Institute about technologies on the concrete materials for the waterproofing in buildings and for water-shielding walls. (J.P.N.)

  10. Mass attenuation coefficients of X-rays in different barite concrete used in radiation protection as shielding against ionizing radiation

    International Nuclear Information System (INIS)

    Almeida, A. T. Jr.; Nogueira, M.S.; Santos, M.A.P.; Campos, L.L.; Araújo, F. G. S.

    2015-01-01

    The attenuation coefficient depends on the incident photon energy and the nature of the materials. In order to minimize exposure to individuals. Barite concrete has been largely used as a shielding material in installations housing gamma radiation sources as well as X-ray generating equipment. This study was conducted to evaluate the efficacy of different mixtures of barite concrete for shielding in diagnostic X-ray rooms. The mass attenuation coefficient (μ/ρ). The mass attenuation coefficients have been measured by employing the CdTe detector model XR-100T. The distance between the source and the exposed surface of all samples was measured by SSD light indicator of machine which was 350 cm. The slope of the linear plot of the intensity transmitted versus specimen thickness would yield the attenuation coefficient. The mass attenuation coefficients (μ/ρ) were compared with the tabulations based upon the results of the XCOM program. The rectangular barite concrete blocks in different thicknesses from were used for the radiation attenuation test. The experimental values were compared with theoretical values WinXcom. The plots of the logarithm of transmitted intensity versus specimen thickness were linear for all the samples and the µ/ρ was obtained from the plots by linear regression over the 25%-2% transmission range, under good geometrical condition. There is a good agreement between theoretical and experimental values, within the 9%. In fact over the entire transmission range of 25-2% the experimental and theoretical values agree well for both the energies. (authors)

  11. Shielding design of highly activated sample storage at reactor TRIGA PUSPATI

    International Nuclear Information System (INIS)

    Naim Syauqi Hamzah; Julia Abdul Karim; Mohamad Hairie Rabir; Muhd Husamuddin Abdul Khalil; Mohd Amin Sharifuldin Salleh

    2010-01-01

    Radiation protection has always been one of the most important things considered in Reaktor Triga PUSPATI (RTP) management. Currently, demands on sample activation were increased from variety of applicant in different research field area. Radiological hazard may occur if the samples evaluation done were misjudge or miscalculated. At present, there is no appropriate storage for highly activated samples. For that purpose, special irradiated samples storage box should be provided in order to segregate highly activated samples that produce high dose level and typical activated samples that produce lower dose level (1 - 2 mR/ hr). In this study, thickness required by common shielding material such as lead and concrete to reduce highly activated radiotracer sample (potassium bromide) with initial exposure dose of 5 R/ hr to background level (0.05 mR/ hr) were determined. Analyses were done using several methods including conventional shielding equation, half value layer calculation and Micro shield computer code. Design of new irradiated samples storage box for RTP that capable to contain high level gamma radioactivity were then proposed. (author)

  12. Economic aspect comparison between steel plate reinforced concrete and reinforced concrete technique in reactor containment wall construction

    International Nuclear Information System (INIS)

    Yuliastuti; Sriyana

    2008-01-01

    Construction costs of nuclear power plant were high due to the construction delays, regulatory delays, redesign requirement, and difficulties in construction management. Based on US DOE (United States Department of Energy) study in 2004, there were thirteen advanced construction technologies which were potential to reduce the construction time of nuclear power plant. Among these technologies was the application of steel-plate reinforced concrete (SC) on reactor containment construction. The conventional reinforced concrete (RC) technique were built in place and require more time to remove framework since the external form is temporary. Meanwhile, the SC technique offered a more efficient way to placing concrete by using a permanent external form made of steel. The objective of this study was to calculate construction duration and economic comparison between RC and SC technique. The result of this study showed that SC technique could reduce the construction time by 60% and 29,7% cost reduced compare to the RC technique. (author)

  13. 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.

  14. 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.

  15. Design of the segment structure and coolant ducts for a fusion reactor blanket and shield

    International Nuclear Information System (INIS)

    Briaris, D.A.; Stanbridge, J.R.

    1978-05-01

    An outline design and analysis of a support structure for the replaceable first wall of a helium cooled fusion reactor blanket has been undertaken. The proposed structure supports all the segment gravitational loads with maximum deflections limited to < 10 mm, and is itself supported off the outer shield by a simple vee-in-groove arrangement. It is a feature of the design that the coaxial coolant pipes and the segment structure operate at the same temperature, making it possible for them to be integrated, thereby avoiding the necessity for pipe bellows. The requirements of cooling the inner arm of the structure and increasing the major radius of the torus by approximately = 0.5 m, have been identified as problems associated with the 'horseshoe' shaped structure applicable to the reactor with divertor. For a ring structure, i.e. reactor without divertor, these problems do not arise. (author)

  16. The characteristics of the prestressed concrete reactor vessel of the HHT demonstration plant

    International Nuclear Information System (INIS)

    Schoening, J.; Schwiers, H.G.

    1979-01-01

    The paper concentrates on the design studies of the HTGR prestressed concrete reactor vessel (PCRV) for the HHT Demonstration Plant. The multi-cavity reactor pressure vessel accommodates all components carrying primary gas, including heat exchangers and gas turbine. For reasons of economics and availability of the reactor plant, generic requirements are made for the PCRV. A short description of the power plant is also presented

  17. 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)

  18. Design of a management information system for the Shielding Experimental Reactor ageing management

    International Nuclear Information System (INIS)

    He Jie; Xu Xianhong

    2010-01-01

    The problem of nuclear reactor ageing is a topic of increasing importance in nuclear safety recent years. Ageing management is usually implemented for reactors maintenance. In the practice, a large number of data and records need to be processed. However, there are few professional software applications that aid reactor ageing management, especially for research reactors. This paper introduces the design of a new web-based management information system (MIS), named the Shielding Experimental Reactor Ageing Management Information System (SERAMIS). It is an auxiliary means that helps to collect data, keep records, and retrieve information for a research reactor ageing management. The Java2 Enterprise Edition (J2EE) and network database techniques, such as three-tiered model, Model-View-Controller architecture, transaction-oriented operations, and JavaScript techniques, are used in the development of this system. The functionalities of the application cover periodic safety review (PSR), regulatory references, data inspection, and SSCs classification according to ageing management methodology. Data and examples are presented to demonstrate the functionalities. For future work, techniques of data mining will be employed to support decision-making.

  19. Design of a management information system for the Shielding Experimental Reactor ageing management

    Energy Technology Data Exchange (ETDEWEB)

    He Jie, E-mail: hejiejoe@163.co [Institute of Nuclear and New Energy Technology, Tsinghua University, Beijing 100084 (China); Xu Xianhong [Institute of Nuclear and New Energy Technology, Tsinghua University, Beijing 100084 (China)

    2010-01-15

    The problem of nuclear reactor ageing is a topic of increasing importance in nuclear safety recent years. Ageing management is usually implemented for reactors maintenance. In the practice, a large number of data and records need to be processed. However, there are few professional software applications that aid reactor ageing management, especially for research reactors. This paper introduces the design of a new web-based management information system (MIS), named the Shielding Experimental Reactor Ageing Management Information System (SERAMIS). It is an auxiliary means that helps to collect data, keep records, and retrieve information for a research reactor ageing management. The Java2 Enterprise Edition (J2EE) and network database techniques, such as three-tiered model, Model-View-Controller architecture, transaction-oriented operations, and JavaScript techniques, are used in the development of this system. The functionalities of the application cover periodic safety review (PSR), regulatory references, data inspection, and SSCs classification according to ageing management methodology. Data and examples are presented to demonstrate the functionalities. For future work, techniques of data mining will be employed to support decision-making.

  20. Development of a shielded ion microprobe analyzer (SIMA) and its application to fast reactor fuel elements

    International Nuclear Information System (INIS)

    Yuji, E.; Junji, K.; Sadamu, Y.; Toshiyuki, I.

    1983-01-01

    A shielded ion microprobe analyzer for elemental and isotopic analyses of irradiated fast reactor fuel and fuel component has been developed and installed in an alpha-gamma hot cell. Radiation shielding of the equipment ensures the radiation dose of -7 C/kg) for 5 Ci (1.85 x 10 11 Bq) of a 60 Co source. Hot samples can be automatically transferred from the cell to the sample chamber of the analyzer. Contamination inside the equipment through sputtering of the radioactive materials can be reduced with a special device. Distribution and migration of fission products, such as 137 Cs, 138 Ba, and 90 Sr, and of fissile materials, such as 235 U and 239 Pu in irradiated mixed-oxide fuel, and isotopic ratios of the elements can be obtained very precisely and quickly

  1. Development of a shielded ion microprobe analyzer (SIMA) and its application to fast reactor fuel elements

    International Nuclear Information System (INIS)

    Enokido, Y.; Itaki, T.; Komatsu, J.; Yamanouchi, S.

    1983-01-01

    A shielded ion microprobe analyzer for elemental and isotopic analyses of irradiated fast reactor fuel and fuel component has been developed and installed in an alpha-gamma hot cell. Radiation shielding of the equipment ensures the radiation dose of -7 C/kg) for 5 Ci (1.85 X 10 11 Bq) of a 60 Co source. Hot samples can be automatically transferred from the cell to the sample chamber of the analyzer. Contamination inside the equipment through sputtering of the radioactive materials can be reduced with a special device. Distribution and migration of fission products, such as 137 Cs, 138 Ba, and 90 Sr, and of fissile materials, such as 235 U and 239 Pu in irradiated mixed-oxide fuel, and isotopic ratios of the elements can be obtained very precisely and quickly

  2. 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

  3. Comparison of In-Vessel Shielding Design Concepts between Sodium-cooled Fast Burner Reactor and the Sodium-cooled Fast Breeder Reactor

    International Nuclear Information System (INIS)

    Yun, Sunghwan; Kim, Sang Ji

    2015-01-01

    In this study, quantities of in-vessel shields were derived and compared each other based on the replaceable shield assembly concept for both of the breeder and burner SFRs. Korean Prototype Gen-IV Sodium-cooled Fast Reactor (PGSFR) like SFR was used as the reference reactor and calculation method reported in the reference was used for shielding analysis. In this paper, characteristics of in-vessel shielding design were studied for the burner SFR and breeder SFR based on the replaceable shield assembly concept. An in-vessel shield to prevent secondary sodium activation (SSA) in the intermediate heat exchangers (IHXs) is one of the most important structures for the pool type Sodium-cooled Fast Reactor (SFR). In our previous work, two in-vessel shielding design concepts were compared each other for the burner SFR. However, a number of SFRs have been designed and operated with the breeder concept, in which axial and radial blankets were loaded for fuel breeding, during the past several decades. Since axial and radial blanket plays a role of neutron shield, comparison of required in-vessel shield amount between the breeder and burner SFRs may be an interesting work for SFR designer. Due to the blanket, the breeder SFR showed better performance in axial neutron shielding. Hence, 10.1 m diameter reactor vessel satisfied the design limit of SSA at the IHXs. In case of the burner SFR, due to more significant axial fast neutron leakage, 10.6 m diameter reactor vessel was required to satisfy the design limit of SSA at the IHXs. Although more efficient axial shied such as a mixture of ZrH 2 and B 4 C can improve shielding performance of the burner SFR, additional fabrication difficulty may mitigate the advantage of improved shielding performance. Therefore, it can be concluded that the breeder SFR has better characteristic in invessel shielding design to prevent SSA at the IHXs than the burner SFR in the pool-type reactor

  4. Laser cutting of concretes with various ballasts

    International Nuclear Information System (INIS)

    Hamasaki, Masanobu; Katsumura, Munehide; Utsumi, Hiroaki

    1985-01-01

    The biological shield concrete and the radiation shield concrete which construct a part of the atomic reactor must be demolished with the decommissioning of the atomic reactor plants. In a case, the demolition using a laser is expected as one of excellent method for the decommissioning of these radioactive concretes. The fundamental cuttings of a mortar, the concretes with andesite, lime stone and gray wacke as ballast and a concrete reinforced with mild steel rods were therefore carried out using a 5 kW output CO 2 laser. As the results of experiment, it was cleared that cutting results varied with ballasts, 100 mm thick reinforced concrete could be cut, safety was high because few dross and few fume were produced. (author)

  5. Strain components of nuclear-reactor-type concretes during first heat cycle

    International Nuclear Information System (INIS)

    Khoury, G.A.

    1995-01-01

    Strains of three advanced-gas-cooled-reactor-type nuclear reactor concretes were measured during the first heat cycle and their relative thermal stability determined. It was possible to isolate for the first time the shrinkage component for the period during heating. Predictions of the residual strains for the loaded specimens can be made by simple superposition of creep and shrinkage components up to a certain critical temperature, which for basalt concrete is about 500 C and for limestone concrete is about 200-300 C. Above the critical temperature, an expansive ''cracking'' strain component is present. It is shown that the strain behaviour of concrete provides a sensitive indication of its thermal stability during heating and subsequent cooling. (orig.)

  6. Bibliography, subject index, and author index of the literature examined by the Radiation Shielding Information Center (Reactor and Weapons Radiation Shielding). [1973--1976

    Energy Technology Data Exchange (ETDEWEB)

    1978-01-01

    An indexed bibliography is presented of literature selected by the Radiation Shielding Information Center since the previous volume was published in 1974 in the area of radiation transport and shielding against radiation from nuclear reactors, x-ray machines, radioisotopes, nuclear weapons (including fallout), and low-energy accelerators (e.g., neutron generators). In addition to lists of literature titles by subject categories (accessions 3501-4950), author and keyword indexes are given. Most of the literature selected for Vol. V was published in the years 1973 to 1976.

  7. Bibliography, subject index, and author index of the literature examined by the radiation shielding information center. Volume 6. Reactor and weapons radiation shielding

    International Nuclear Information System (INIS)

    1980-05-01

    An indexed bibliography is presented of literature selected by the Radiation Shielding Information Center since the previous volume was published in 1978 in the area of radiation transport and shielding against radiation from nuclear reactors, x-ray machines, radioisotopes, nuclear weapons (including fallout), and low energy accelerators (e.g., neutron generators). The bibliography was typeset from data processed by computer from magnetic tape files. In addition to lists of literature titles by subject categories (accessions 4951-6200), an author index is given

  8. Bibliography, subject index, and author index of the literature examined by the Radiation Shielding Information Center (Reactor and Weapons Radiation Shielding)

    International Nuclear Information System (INIS)

    1978-01-01

    An indexed bibliography is presented of literature selected by the Radiation Shielding Information Center since the previous volume was published in 1974 in the area of radiation transport and shielding against radiation from nuclear reactors, x-ray machines, radioisotopes, nuclear weapons (including fallout), and low-energy accelerators (e.g., neutron generators). In addition to lists of literature titles by subject categories (accessions 3501-4950), author and keyword indexes are given. Most of the literature selected for Vol. V was published in the years 1973 to 1976

  9. 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

  10. Experimental verification of creep analyses for prestressed concrete reactor vessels

    International Nuclear Information System (INIS)

    Aoyagi, Y.; Abe, H.; Ohnuma, H.

    1977-01-01

    The authors proposed a new method of creep analysis based on the theory of strain hardening, which assumes that accumulated creep at a given time influences the creep after that. This method was applied to calculate step-by-step the behaviors of uniaxial creep of concrete under variable temperatures and stresses, creep in reinforced concrete specimens and the behaviors of prestressed concrete beams under themal gradients. The experimental and calculated results agreed fairly well. Further, this method was incorporated in the finite element creep analysis for the prestressed concrete hollow cylinder and the full scale model. The calculated strain changes with time pursued closely those obtained by experiments. The above led to the conclusion that from the viewpoint of both accuracy and computation time the strain hardening method proposed by the authors may be judged advantageous for practical usages

  11. Aircraft impact on nuclear power plants concrete structures

    International Nuclear Information System (INIS)

    Coombs, R.F.; Barbosa, L.C.B.; Santos, S.H.C.

    1980-01-01

    A summary about the procedures for the analysis of aircraft on concrete structures, aiming to emphasize the aspects related to the nuclear power plants safety, is presented. The impact force is determined by the Riera model. The effect of this impact force on the concrete structures is presented, showing the advantages to use nonlinear behaviour in the concrete submitted to short loads. The simplifications used are shown through a verification example of the nuclear reactor concrete shielding. (E.G.) [pt

  12. Neutron multiplication and shielding problems in pressurized water reactor spent fuel shipping casks

    International Nuclear Information System (INIS)

    Devillers, C.; Blum, P.

    1977-01-01

    To evaluate the degree of accuracy of computational methods used in the shield design of spent fuel shipping casks, comparisons have been made between biological dose-rate calculations and measurements at the surface of a cask carrying three pressurized water reactor fuel assemblies. Neutron dose-rate measurements made with the fuel-carrying region successively wet and dry are also used to derive an experimental value of the k/sub eff/ of the wet fuel assemblies. Results obtained by this method are shown to be consistent with criticality calculations, taking into account fuel depletion

  13. Fault current limiter-predominantly resistive behavior of a BSCCO shielded-core reactor

    International Nuclear Information System (INIS)

    Ennis, M. G.; Tobin, T. J.; Cha, Y. S.; Hull, J. R.

    2000-01-01

    Tests were conducted to determine the electrical and magnetic characteristics of a superconductor shielded core reactor (SSCR). The results show that a closed-core SSCR is predominantly a resistive device and an open-core SSCR is a hybrid resistive/inductive device. The open-core SSCR appears to dissipate less than the closed-core SSCR. However, the impedance of the open-core SSCR is less than that of the closed-core SSCR. Magnetic and thermal diffusion are believed to be the mechanism that facilitates the penetration of the superconductor tube under fault conditions

  14. Formulary for neutron propagation in sodium-steel media for the fast reactor shields

    International Nuclear Information System (INIS)

    Bouteau, F.; Caumette, P.; Khairallah, A.; Oceraies, Y.; Devillers, C.

    1975-01-01

    The simplified calculational tool (''formulary'') for neutron propagation in the shields of fast reactors, being developed at CEA, has two objectives: to reduce the cost of the major part of design calculations, without a significant loss of accuracy; to facilitate the adjustment of the calculational tool with the results of the program of integral propagation experiments, which is conducted in parallel with the development of the calculational method. The version 0 (i.e. before any adjustment) of the formulary and a first test of its validity as compared to the results of integral measurements are presented [fr

  15. The 1500 MW fast breeder reactor the double envelope-vessel anchored in concrete

    International Nuclear Information System (INIS)

    Bolvin, M.

    1981-01-01

    This paper givers an account of EDF investigations to reduce the investment costs of the 1500 MW Fast Reactor (RNR 1500) without prejudice to the safety requirements. It deals with the double envelope-vessel, designed to minimize radiation consequences in the event of accidental leakage in the main vessel. In the Fast Reactors in operation (PHOENIX), under construction (CRYS-MALVILLE), and under project (NR 1500), the double envelope-steel vessel hangs down from the upper part of the reactor block, its weight being approximately 300 t. In the new design, the vessel is fixed into the concrete which supports the main vessel, by means of steel anchors. A thermal insulation isolates it from the main vessel. The installation of coils in the concrete, next to the lining, allows for water circulation to cool the concrete. (orig./GL)

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

    International Nuclear Information System (INIS)

    Akki, T.S.; Benayad, S.A.; Megahid, R.M.

    1992-01-01

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

  17. Nuclear reactor installation

    International Nuclear Information System (INIS)

    Keller, W.

    1976-01-01

    A nuclear reactor installation includes a pressurized-water coolant reactor vessel and a concrete biological shield surrounding this vessel. The shield forms a space between it and the vessel large enough to permit rapid escape of the pressurized-water coolant therefrom in the event the vessel ruptures. Struts extend radially between the vessel and shield for a distance permitting normal radial thermal movement of the vessel, while containing the vessel in the event it ruptures, the struts being interspaced from each other to permit rapid escape of the pressurized-water coolant from the space between the shield and the vessel

  18. Constitutive models for concrete and finite element analysis of prestressed concrete reactor vessels

    International Nuclear Information System (INIS)

    Smith, P.D.; Anderson, C.A.

    1977-01-01

    Two constitutive models for concrete are discussed. For short-term loads, the orthotropic variable modulus model is described, and for long-term loads a viscoelastic model utilizing a Dirichlet series approximation for the creep compliance function is summarized. The orthotropic variable modulus model is demonstrated in an analysis of a PCRV head with penetrations. The viscoelastic model is illustrated with a simulation of a prestressed concrete cylinder subject to non-uniform temperatures

  19. Importance of self-shielding for improving sensitivity coefficients in light water nuclear reactors

    International Nuclear Information System (INIS)

    Foad, Basma; Takeda, Toshikazu

    2014-01-01

    Highlights: • A new method has been developed for calculating sensitivity coefficients. • This method is based on the use of infinite dilution cross-sections instead of effective cross-sections. • The change of self-shielding factor due to cross-section perturbation has been considered. • SRAC and SAINT codes are used for calculating improved sensitivities, while MCNP code has been used for verification. - Abstract: In order to perform sensitivity analyzes in light water reactors where self-shielding effect becomes important, a new method has been developed for calculating sensitivity coefficient of core characteristics relative to the infinite dilution cross-sections instead of the effective cross-sections. This method considers the change of the self-shielding factor due to cross-section perturbation for different nuclides and reactions. SRAC and SAINT codes are used to calculate the improved sensitivity; while the accuracy of the present method has been verified by MCNP code and good agreement has been found

  20. Burst shield for a pressurized nuclear-reactor core and method of operating same

    International Nuclear Information System (INIS)

    Beine, B.; Schilling, F.

    1976-01-01

    A pressurized nuclear-reactor core stands on a base up from which extends a cylindrical side wall formed of a plurality of hollow iron castings held together by circumferential and longitudinal prestressed elements. A cylindrical space between this shield and the core serves for inspection of the core and is normally filled with cast-iron segmental slabs so that if the core bursts pieces thrown out do not acquire any dangerous kinetic energy before engaging the burst shield. The top of the shield is removably secured to the side so that it can be moved out of the way periodically for removal of the filler slabs and inspection of the core. An anchor on the upper end of each longitudinal prestressing element bears against a sleeve pressing against the uppermost side element, and a nut engageable with this anchor is engageable down over the top to hold it in place, removal of this nut leaving the element prestressed in the side wall. 11 claims, 16 drawing figures

  1. 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/

  2. Aging management program of the reactor building concrete at Point Lepreau Generating Station

    Directory of Open Access Journals (Sweden)

    Gendron T.

    2011-04-01

    Full Text Available In order for New Brunswick Power Nuclear (NBPN to control the risks of degradation of the concrete reactor building at the Point Lepreau Generating Station (PLGS the development of an aging management plan (AMP was initiated. The intention of this plan was to determine the requirements for specific structural components of concrete of the reactor building that require regular inspection and maintenance to ensure the safe and reliable operation of the plant. The document is currently in draft form and presents an integrated methodology for the application of an AMP for the concrete of the reactor building. The current AMP addresses the reactor building structure and various components, such as joint sealant and liners that are integral to the structure. It does not include internal components housed within the structure. This paper provides background information regarding the document developed and the strategy developed to manage potential degradation of the concrete of the reactor building, as well as specific programs and preventive and corrective maintenance activities initiated.

  3. Aging management program of the reactor building concrete at Point Lepreau Generating Station

    Science.gov (United States)

    Aldea, C.-M.; Shenton, B.; Demerchant, M. M.; Gendron, T.

    2011-04-01

    In order for New Brunswick Power Nuclear (NBPN) to control the risks of degradation of the concrete reactor building at the Point Lepreau Generating Station (PLGS) the development of an aging management plan (AMP) was initiated. The intention of this plan was to determine the requirements for specific structural components of concrete of the reactor building that require regular inspection and maintenance to ensure the safe and reliable operation of the plant. The document is currently in draft form and presents an integrated methodology for the application of an AMP for the concrete of the reactor building. The current AMP addresses the reactor building structure and various components, such as joint sealant and liners that are integral to the structure. It does not include internal components housed within the structure. This paper provides background information regarding the document developed and the strategy developed to manage potential degradation of the concrete of the reactor building, as well as specific programs and preventive and corrective maintenance activities initiated.

  4. Effect of density deviations of concrete on its attenuation efficiency

    International Nuclear Information System (INIS)

    Szymendera, L.; Wincel, K.; Blociszewski, S.; Kordyasz, D.; Sobolewska, I.

    In the work, the influence of concrete density deviation on shield thickness and total dose ratio outside the reactor shield, has--on the basis of numerical analysis--been considered. It has been noticed the possibility of introducing flexible corrections--without additional shielding calculation--to the design thickness of the shield. It has been also found that in common cases of shield design, where any necessity of minimizing the shield thickness does not exist, the tendency to minimize the value of this deviation is hardly substantiable

  5. Methods of preventing fast breeder reactor shield plug from adhesion of sodium

    International Nuclear Information System (INIS)

    Hashiguchi, Koh; Hara, Johji; Nei, Hiromichi; Daiku, Motoichi; Wagatsuma, Kenji

    1980-01-01

    The shield plug, which is located at the upper part of a reactor vessel of a sodium-cooled fast breeder reactor, is composed of a rotating and a stationary plug. Fuel exchange is performed easily by the rotation of the rotating plug. The vapor or mist of sodium evaporated from liquid sodium deposits on the gap surfaces of the rotating and stationary plugs and is solidified or changed into a solid reactant. If such condition continues for a long period, harmful effects are exerted on the fuel exchange operation. In order to develop methods of preventing the sodium deposition, investigation was made on the phenomenon of sodium deposition. By the use of the testing equipment simulating the shield plug, deposition tests and specimen measurements were made for different gap width test section size and condition. On the basis of the effects of these parameters clarified by experiments, the effectiveness of three kinds of mechanism for preventing sodium deposition were investigated experimentally. In addition, by using a thermo-siphon analogical model, analysis was performed to deduce experimental equations for sodium deposition. (author)

  6. Gas and water permeability of concrete for reactor buildings small specimens

    International Nuclear Information System (INIS)

    Mills, R.H.

    1986-03-01

    The effect on permeability of artifical aging by drying shrinkage and by freeze-thaw was determined by observing mass transfer of gas and water under a pressure gradient. It was found that damage due to freeze-thaw was negligible but that cracking around aggregate caused by drying shrinkage resulted in significantly increased permeability to both gas and water. The absence of freeze-thaw damage was attributed to self-dessication. Since the concrete was not exposed to an external source of water, the chemical reaction was sustained by consumption of mixing water. The resulting air voids were, apparently, sufficient to absorb expansive pressures due to ice formation. The response to lateral prestress was different for cracked and uncracked concrete. Although, in all cases, increased prestress resulted in reduced leakage, the effect was stronger in cracked concrete. Mean pore diameter as determined by gas diffusion was not, however, substantially affected because the leakage in cracked concrete remained very low. Reinforcing steel did not have a great influence on permeability of small specimens. Gas transmission through concrete was strongly influenced by moisture content. Free moisture constituted a barrier to gas flow, acting as a virtual solid. This is important since aging of concrete results in reduced free moisture. Ultrasonic pulse velocity appeared to vary with moisture content and porosity of concrete in the same way as gas permeability and gave promise of being effective for in-situ monitoring of concrete in reactor buildings

  7. Rectification of leak from upper aluminium thermal shield cooling water inlet line of Cirus reactor

    International Nuclear Information System (INIS)

    Bhatnagar, Anil; Joshi, N.S.; Kharpate, A.V.; Marik, S.K.

    2006-01-01

    During 1994, a small water leak was observed from the upper aluminium thermal shield of Cirus reactor. Detailed investigations revealed that the leakage was from the weld joint of one of the 1 1/4 inch NB Sch. 80 coolant inlet pipes connected to the upper aluminium thermal shield. The location of the leak was identified by monitoring the stabilised water level in the vertical inlet pipe under stagnant condition. The exact location was identified by installing an inflatable seal arrangement inside the leaky pipe and inflating the seal at different elevations to isolate the leaky location and ensuring that the leak was completely stopped. This location was about 15 feet below the operating floor of the reactor. The pipe was visually inspected with the help of a fibre-scope to assess the condition of the inner surface. Eddy current testing was also carried out for volumetric examination. This revealed one more localised flaw on the outer surface little above the leaky joint. A hollow plug, with expandable rings, having C-shaped cross section at both the ends and a straight portion in the middle to cover the defective region, was developed and qualified in a mock-up station after extensive trials. In view of the site constraints, a flexible hollow link assembly was engineered, for installing the plug remotely. The inner surface of the pipe was cleaned using an emery brush and a deburring tool. The plug was then installed covering the leak area and the rings were expanded by remote tightening. The shield was hydro-tested satisfactorily. (author)

  8. Effect of Flyash Addition on Mechanical and Gamma Radiation Shielding Properties of Concrete

    Directory of Open Access Journals (Sweden)

    Kanwaldeep Singh

    2014-01-01

    Full Text Available Six concrete mixtures were prepared with 0%, 20%, 30%, 40%, 50%, and 60% of flyash replacing the cement content and having constant water to cement ratio. The testing specimens were casted and their mechanical parameters were tested experimentally in accordance with the Indian standards. Results of mechanical parameters show their improvement with age of the specimens and results of radiation parameters show no significant effect of flyash substitution on mass attenuation coefficient.

  9. Study of measurement method of tritium induced in concrete of high-energy proton accelerator facilities

    International Nuclear Information System (INIS)

    Ohtsuka, N.; Ishihama, S.; Kunifuda, T.; Hayasaka, N.; Miura, T.

    2001-01-01

    Various long-loved radionuclides, 3 H, 7 Be, 22 Na, 51 Cr, 54 Mn, 56 Co, 57 Co, 60 Co, 134 Cs, 152 Eu and 154 Eu, have been produced in the shielding concrete of high energy proton accelerator facility through both nuclear spallation reactions and thermal neutron capture reactions of concrete elements, during machine operation. Tritium is the most important nuclide from the radiation protection. There were, however, few measurements of tritium concentration induced in the shielding concrete. In this study, the conditions of measurement method of tritium concentration induced in shielding concrete have been investigated using the activated shielding concrete of the 12 GeV proton beam-line tunnel at KEK and the standard rock (JG-1) irradiated of thermal neutron at the reactor. And the depth profiles of tritium induced in the shielding concrete of slow extracted proton beam line at KEK were determined using this method. (author)

  10. Radiation effects in concrete for nuclear power plants – Part I: Quantification of radiation exposure and radiation effects

    International Nuclear Information System (INIS)

    Field, K.G.; Remec, I.; Pape, Y. Le

    2015-01-01

    Highlights: • Neutron and gamma rays fields in concrete biological shield are calculated. • An extensive database on irradiated concrete properties has been collected. • Concrete mechanical properties decrease beyond 1.0 × 10 19 n/cm 2 fluence. • Loss of properties appears correlated with radiation induced-aggregate swelling. • Commercial reactor bio-shield may experience long-term irradiation damage. - Abstract: A large fraction of light water reactor (LWR) construction utilizes concrete, including safety-related structures such as the biological shielding and containment building. Concrete is an inherently complex material, with the properties of concrete structures changing over their lifetime due to the intrinsic nature of concrete and influences from local environment. As concrete structures within LWRs age, the total neutron fluence exposure of the components, in particular the biological shield, can increase to levels where deleterious effects are introduced as a result of neutron irradiation. This work summarizes the current state of the art on irradiated concrete, including a review of the current literature and estimates the total neutron fluence expected in biological shields in typical LWR configurations. It was found a first-order mechanism for loss of mechanical properties of irradiated concrete is due to radiation-induced swelling of aggregates, which leads to volumetric expansion of the concrete. This phenomena is estimated to occur near the end of life of biological shield components in LWRs based on calculations of estimated peak neutron fluence in the shield after 80 years of operation

  11. Material characteristics and construction methods for a typical research reactor concrete containment in Iran

    International Nuclear Information System (INIS)

    Ebrahimia, Mahsa; Suha, Kune Y.; Eghbalic, Rahman; Jahan, Farzaneh Asadi malek

    2012-01-01

    Generally selecting an appropriate material and also construction style for a concrete containment due to its function and special geometry play an important role in applicability and also construction cost and duration decrease in a research reactor (RR) project. The reactor containment enclosing the reactor vessel comprises physical barriers reflecting the safety design and construction codes, regulations and standards so as to prevent the community and the environment from uncontrolled release of radioactive materials. It is the third and the last barrier against radioactivity release. It protects the reactor vessel from such external events as earthquake and aircraft crash as well. Thus, it should be designed and constructed in such a manner as to withstand dead and live loads, ground and seismic loads, missiles and aircraft loads, and thermal and shrinkage loads. This study aims to present a construction method for concrete containment of a typical RR in Iran. The work also presents an acceptable characteristic for concrete and reinforcing re bar of a typical concrete containment. The current study has evaluated the various types of the RR containments. The most proper type was selected in accordance with the current knowledge and technology of Iran

  12. Material characteristics and construction methods for a typical research reactor concrete containment in Iran

    Energy Technology Data Exchange (ETDEWEB)

    Ebrahimia, Mahsa; Suha, Kune Y. [Seoul National Univ., Seoul (Korea, Republic of); Eghbalic, Rahman; Jahan, Farzaneh Asadi malek [School of Architecture and Urbanism, Qazvin (Iran, Islamic Republic of)

    2012-10-15

    Generally selecting an appropriate material and also construction style for a concrete containment due to its function and special geometry play an important role in applicability and also construction cost and duration decrease in a research reactor (RR) project. The reactor containment enclosing the reactor vessel comprises physical barriers reflecting the safety design and construction codes, regulations and standards so as to prevent the community and the environment from uncontrolled release of radioactive materials. It is the third and the last barrier against radioactivity release. It protects the reactor vessel from such external events as earthquake and aircraft crash as well. Thus, it should be designed and constructed in such a manner as to withstand dead and live loads, ground and seismic loads, missiles and aircraft loads, and thermal and shrinkage loads. This study aims to present a construction method for concrete containment of a typical RR in Iran. The work also presents an acceptable characteristic for concrete and reinforcing re bar of a typical concrete containment. The current study has evaluated the various types of the RR containments. The most proper type was selected in accordance with the current knowledge and technology of Iran.

  13. 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.)

  14. Multi-objective optimization of a compact pressurized water nuclear reactor computational model for biological shielding design using innovative materials

    Energy Technology Data Exchange (ETDEWEB)

    Tunes, M.A., E-mail: matheus.tunes@usp.br [Department of Metallurgical and Materials Engineering, Escola Politécnica da Universidade de São Paulo, Av. Prof. Mello Moraes, 2463 – CEP 05508 – 030 São Paulo (Brazil); Oliveira, C.R.E. de, E-mail: cassiano@unm.edu [Department of Nuclear Engineering, The University of New Mexico, Farris Engineering Center, 221, Albuquerque, NM 87131-1070 (United States); Schön, C.G., E-mail: schoen@usp.br [Department of Metallurgical and Materials Engineering, Escola Politécnica da Universidade de São Paulo, Av. Prof. Mello Moraes, 2463 – CEP 05508 – 030 São Paulo (Brazil)

    2017-03-15

    Highlights: • Use of two n-γ transport codes leads to optimized model of compact nuclear reactor. • It was possible to safely reduce both weight and volume of the biological shielding. • Best configuration obtained by using new composites for both γ and n attenuation. - Abstract: The aim of the present work is to develop a computational model of a compact pressurized water nuclear reactor (PWR) to investigate the use of innovative materials to enhance the biological shielding effectiveness. Two radiation transport codes were used: the first one – MCNP – for the PWR design and the GEM/EVENT to simulate (in a 1D slab) the behavior of several materials and shielding thickness on gamma and neutron radiation. Additionally MATLAB Optimization Toolbox was used to provide new geometric configurations of the slab aiming at reducing the volume and weight of the walls by means of a cost/objective function. It is demonstrated in the reactor model that the dose rate outside biological shielding has been reduced by one order of magnitude for the optimized model compared with the initial configuration. Volume and weight of the shielding walls were also reduced. The results indicated that one-dimensional deterministic code to reach an optimized geometry and test materials, combined with a three-dimensional model of a compact nuclear reactor in a stochastic code, is a fast and efficient procedure to test shielding performance and optimization before the experimental assessment. A major outcome of this research is that composite materials (ECOMASS 2150TU96) may replace (with advantages) traditional shielding materials without jeopardizing the nuclear power plant safety assurance.

  15. Neutron shielding and constructional characteristics of a new type concrete and from borated clinker

    International Nuclear Information System (INIS)

    Cakaloz, T.

    1979-07-01

    A boron containing cement, which can be used as nuclear shielding material, is produced at pilot plant scale applying two different methods. In the first method, the raw mixture of a normal portland cement is mixed with pre-calcined colemanite, a calcium borate mineral, and clinkerized in a rotary kiln (borated-clinker). In the second method, the colemanite is mixed with an admixture, which contains mainly limestone and marl, and burnt in the rotary kiln to obtain a borated-lime composite. The borated-lime composite is then added to the normal portland cement clinker up to 2% B 2 O 3 content for shielding purpose. The results have shown that the borated-clinker contained untolerable amount of free lime resulting in a decrease in compressive strength. The addition of the borated-lime composite to the normal portland cement clinker up to 1% B 2 O 3 content did not alter the setting time and the volume expansion properties. The reduction in the compressive strength was found to be tolerable, however, the decrease in the bending strength was 20% lower than that of permissible value. On the other hand, the increase in B 2 O 3 content of the mixture improved the neutron absorptivity resulting in an increase in total cross-section about 7 times for 1% B 2 O 3 without changing the gamma absorption value

  16. Mobile concrete solidification systems for power reactor waste

    International Nuclear Information System (INIS)

    Tchemitcheff, E.; Bordas, Y.

    1990-01-01

    In late 1988 SGN received an order from Electricite de France (EDF) for the construction of a mobile concrete solidification system to process secondary system resins generated by the P'4 and N4 series PWR power plants in France. This order was placed in view of SGN's experience with low- and medium-level radioactive waste treatment and conditioning over a period of almost 20 years. In addition to the construction of fixed waste processing facilities using more conventional technologies, SGN has been involved in application of the mobile system concept to the bituminization process in the United States, which led to the construction and commissioning of two transportable systems in collaboration with its American licensee US Ecology. It has also conducted large-scale R ampersand D on LLW/MLW concrete solidification, particularly for ion exchange resins. 5 figs

  17. The permeability of concrete for reactor containment vessels

    International Nuclear Information System (INIS)

    Mills, R.H.

    1983-07-01

    Review of the literature pertaining to water, water vapour and gas transmission through concrete revealed conflicting views on the mechanisms involved and the influence of mix design parameters such as initial porosities and water/cement ratio. Consideration of the effects of ageing and of construction defects in field concrete were totally neglected in published work. Permeability data from three published papers were compared with permeability calculated according to Powers. The ratio of calculated to observed permeability varied from 40 x 10 -3 to 860 x 10 -3 for one group: from 0.17 x 10 3 to 8.6 x 10 3 in the second; and from 24 x 10 3 to 142 x 10 3 for the third. There were therefore wide discrepancies within each group of data and between groups. A bibliography was prepared and an exploratory experimental programme was mounted to determine the relative importance of key parameters such as cement type, porosity and water/cement ratio. Contrary to frequently cited references it was found that permeability of concrete was not significantly influenced by water/cement ratio when the starting porosity was constant. If water/cement ratio was held constant, however, the permeability was strongly influenced by starting porosity. It was also found that with constant water/cement ratio permeability increased with cement content. The value of fly ash and blast furnace slag in partial substitution for Portland cement is neglected in the literature but it is important since such substitutions alleviate alkali-silicate reactions. Permeability of concrete was significantly decreased by partial substitution of Portland cement with fly ash but there was no benefit in the use of blast furnace slag

  18. Thermal effects, creep and nonlinear responde of concrete reactor vessels

    International Nuclear Information System (INIS)

    Bazant, Z.P.

    1978-01-01

    A new mathematical model for prediction of pore pressure and moisture transfer in concrete heated well beyond 100 0 C is outlined. The salient features of the model are:(1) the hypothesis taht the pore space available to capillary water grows with increasing temperature as well as increasing pressure in excess of saturation pressure, and (2) the hypothesis that moisture permeability increases by two orders of magnitude when passing 100 0 C. Permaability below 100 0 C is controlled by migration of adsorbed water through gel-pore sized necks on passages through the material; these necks are lost above 100 0 C and viscosity then governs. The driving force of moisture transfer may be considered as the gradient of pore pressure, which is defined as pressure of vapor rather than liquid water if concrete is not saturated. Thermodynamic properties of water may be used to determine sorption isotherms in saturated concrete. The theory is the necessary first step in rationally predicting thermal stresses and deformations, and assessing the danger of explosive spalling. However, analysis of creep and nonlinear triaxial behavior is also needed for this purpose. A brief review of recent achievements in these subjects is also given. (Author)

  19. 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)

  20. 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

  1. Conception of thermonuclear reactor with a shielding layer of the first wall

    International Nuclear Information System (INIS)

    Marin, S.V.

    1979-01-01

    Considered is the way of the shielding of the first wall of a thermonuclear reactor by the layer of ISSEC (Internal spectral shifter and Energy Converter). It is a constructive non-power element placed between a plasma and the first wall, and intended for the softening of the spectrum and intensity reduction of particle fluxes falling on the first wall. Results of neutron-physical calculations of the UWMAK-type reactor blanket (in the S 4 -P 3 approximation) are presented. While comparing five materials (C, Mo, Nb, V,W) by the rate of radiation damage formation, gas production, radioactivity level and energy output in the blanket with the 316 stainless steel first wall, it is obvious that the conception of ISSEC permits to prolong the service period of the first wall. Construction elements should be then in the same irradiation conditions as those in fast reactors. Molybdenum has been taken as the best ISSEC material. It reduces the number of displaced atoms of the first wall by 20% and decreases helium production by about 100%, increases energy output in the blanket by 15-18%. However, graphite is advantageous, while comparing it to molybdenum in values of residual energy output, radioactivity level, costs and manufacture simplicity. One problem stays unsolved, which is connected with chemical sputtering of graphite at the formation of C 2 H 2 in the high temperature range. So it is hard to prefer any material now

  2. Demonstration test on manufacturing 200 l drum inner shielding material for recycling of reactor operating metal scrap

    International Nuclear Information System (INIS)

    Umemura, A.; Kimura, K.; Ueno, H.

    1993-01-01

    Low-level reactor wastes should be safely recycled considering those resource values, the reduction of waste disposal volume and environmental effects. The reasonable recycling system of reactor operating metal scrap has been studied and it was concluded that the 200 liter drum inner shielding material is a very promising product for recycling within the nuclear industry. The drum inner shielding material does not require high quality and so it is expected to be easily manufactured by melting and casting from roughly sorted scrap metals. This means that the economical scrap metal recycling system can be achieved by introducing it. Furthermore its use will ensure safety because of being contained in a drum. In order to realize this recycling system with the drum inner shielding material, the demonstration test program is being conducted. The construction of the test facility, which consists of a melting and refining furnace, a casting apparatus, a machining apparatus etc., was finishing in September, 1992

  3. Discrete nodal integral transport-theory method for multidimensional reactor physics and shielding calculations

    International Nuclear Information System (INIS)

    Lawrence, R.D.; Dorning, J.J.

    1980-01-01

    A coarse-mesh discrete nodal integral transport theory method has been developed for the efficient numerical solution of multidimensional transport problems of interest in reactor physics and shielding applications. The method, which is the discrete transport theory analogue and logical extension of the nodal Green's function method previously developed for multidimensional neutron diffusion problems, utilizes the same transverse integration procedure to reduce the multidimensional equations to coupled one-dimensional equations. This is followed by the conversion of the differential equations to local, one-dimensional, in-node integral equations by integrating back along neutron flight paths. One-dimensional and two-dimensional transport theory test problems have been systematically studied to verify the superior computational efficiency of the new method

  4. ITER [International Thermonuclear Experimental Reactor] shield and blanket work package report

    International Nuclear Information System (INIS)

    1988-06-01

    This report summarizes nuclear-related work in support of the US effort for the International Thermonuclear Experimental Reactor (ITER) Study. The purpose of this work was to prepare for the first international ITER workshop devoted to defining a basic ITER concept that will serve as a basis for an indepth conceptual design activity over the next 2-1/2 years. Primary tasks carried out during the past year included: design improvements of the inboard shield developed for the TIBER concept, scoping studies of a variety of tritium breeding blanket options, development of necessary design guidelines and evaluation criteria for the blanket options, further safety considerations related to nuclear components and issues regarding structural materials for an ITER device. 44 refs., 31 figs., 29 tabs

  5. Calculation of neutron and gamma ray energy spectra for fusion reactor shield design: comparison with experiment

    International Nuclear Information System (INIS)

    Santoro, R.T.; Alsmiller, R.G. Jr.; Barnes, J.M.; Chapman, G.T.

    1980-08-01

    Integral experiments that measure the transport of approx. 14 MeV D-T neutrons through laminated slabs of proposed fusion reactor shield materials have been carried out. Measured and calculated neutron and gamma ray energy spectra are compared as a function of the thickness and composition of stainless steel type 304, borated polyethylene, and Hevimet (a tungsten alloy), and as a function of detector position behind these materials. The measured data were obtained using a NE-213 liquid scintillator using pulse-shape discrimination methods to resolve neutron and gamma ray pulse height data and spectral unfolding methods to convert these data to energy spectra. The calculated data were obtained using two-dimensional discrete ordinates radiation transport methods in a complex calculational network that takes into account the energy-angle dependence of the D-T neutrons and the nonphysical anomalies of the S/sub n/ method

  6. Temperature distribution due to the heat generation in nuclear reactor shielding

    International Nuclear Information System (INIS)

    Torres, L.M.R.

    1985-01-01

    A study is performed for calculating nuclear heating due to the interaction of neutrons and gamma-rays with matter. Modifications were implemented in the ANISN and DOT 3.5 codes, that solve the transport equation using the discrete ordinate method, in one two-dimensions respectively, to include nuclear heating calculations in these codes. In order to determine the temperature distribution, using the finite difference method, a numerical model was developed for solving the heat conduction equation in one-dimension, in plane, cylindrical and spherical geometries, and in two-dimensions, X-Y and R-Z geometries. Based on these models, computer programs were developed for calculating the temperature distribution. Tests and applications of the implemented modifications were performed in problems of nuclear heating and temperature distribution due to radiation energy deposition in fission and fusion reactor shields. (Author) [pt

  7. 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)

  8. Decontamination and concrete core sampling by teleoperated robot at Fukushima Daiichi reactor buildings

    International Nuclear Information System (INIS)

    Watanabe, Masaru; Onitsuka, Hironori; Shimonabe, Noriaki; Fujita, Jun; Matsumura, Takumi; Okumura, Atsushi

    2015-01-01

    For decommissioning of Fukushima daiichi nuclear power station, reduction of the dose equivalent rates inside the reactor buildings is an important issue. Concrete core sampling from the buildings to investigate the contamination is necessary for study about effective decontamination. However, dose rate inside the reactor buildings is very high. For example, dose rate of 1st floor on the Unit 1 is 1.2 - 1820 [mSv / h], the Unit 2 is 2.5 - 220 [mSv / h] and Unit 3 is 2.2 - 4780 [mSv / h]. So it is difficult for workers to work long hours. Therefore, a teleoperated robot, named 'MHI-MEISTeR (Mitsubishi Heavy Industries - Maintenance Equipment Integrated System of Telecontrol Robot)', has been developed to conduct operations like concrete core samples from the reactor buildings. Actually, some concrete core samples from Fukushima daiichi were taken by MHI-MEISTeR. In addition, MHI-MEISTeR is designed as a versatile robot, and so it can conduct suction / blast decontamination works as well as concrete core sampling. The above operations were performed by MHI-MEISTeR in Fukushima daiichi nuclear power station. (author)

  9. Design and analysis of concrete reactor vessels: New developments, problems and trends

    International Nuclear Information System (INIS)

    Bazant, Z.P.

    1984-01-01

    This lecture reviews new developments in analysis and design of prestressed concrete reactor vessels (PCRV). After a brief assessment of the current status and experience, the advantages, disadvantages, and especially the safety features of PCRV, are discussed. Attention is then focused on the design of penetrations and openings, and on the design for high-temperature resistance - areas in which further developments are needed. Various possible designs for high-temperature exposure of concrete in a hypothetical accident are analyzed. Considered are not only PCRVs for gas-cooled reactors (GCR), but also guard vessels for liquid metal fast breeder reactors (LMFBR), for which designs mitigating the adverse effects of molten sodium, molten steel, and core melt are surveyed. Realistic analysis of the problems requires further development in the knowledge of material behavior and its mathematical modeling. Recent advances in the modeling of high-temperature response of concrete, including pore water transfer, pore pressure, creep and shrinkage are outlined. This is followed by a discussion of new developments in the analysis of cracking of concrete, where the need of switching from stress criteria to energy criteria for fracture is emphasized. The lecture concludes with a brief discussion of long-time behavior, the effect of aging, and probabilistic analysis of creep. (orig.)

  10. Benchmark shielding calculations for the NEACRP [Nuclear Energy Agency-Committee on Reactor Physics] Working Group on shielding assessment of transportation packages

    International Nuclear Information System (INIS)

    Broadhead, B.L.; Brady, M.C.; Parks, C.V.

    1990-11-01

    In 1985, the Nuclear Energy Agency-Committee on Reactor Physics (NEACRP) established a working group on shielding assessment of transportation packages. Following the initial distribution of a set of six problems, discussions were held at the Organization for Economic Cooperation and Development (OECD) Headquarters in Paris, France, in June/July 1986, May 1988, and February/March 1990. The US contribution to the working group is documented in this report. The results from this effort permit the evaluation of a number of approximations and effects that must be considered in a typical shielding analysis of a transportation cask. Among the effects reported here are the performance of multiple cross-section sets, the comparison of several source generation codes, and multidimensional versus one-dimensional (1-D) analyses. 18 refs., 16 figs., 33 tabs

  11. Concrete

    DEFF Research Database (Denmark)

    2015-01-01

    Concrete is a component of coherent transition between a concrete base and a wooden construction. The structure is based on a quantity of investigations of the design possibilities that arise when combining digital fabrication tools and material capacities. Through tangible experiments the project...... specific for this to happen. And the knowledge and intention behind the drawing becomes specialised through the understanding of the fabrication processes and their affect on the materials.The structure Concrete is a result of a multi-angled kerf series in ash wood and a concrete base. The ash wood is cut...... using a 5-axis CNC router with a thin saw blade attached. The programming of the machining results in variations of kerfs that lets the ash wood twist into unique shapes.The shapes of the revolving ash ribbons continue into the concrete creating a cohesive shape. The form for the concrete itself is made...

  12. Study and implementation of the CADIS methodology to research reactor shielding design

    International Nuclear Information System (INIS)

    Souza, Gregorio S.; Shorto, Julian M.B.; Santos, Adimir dos

    2013-01-01

    The Consistent Adjoint Driven Importance Sampling (CADIS) is a methodology that basically uses source biasing and a mesh-based importance map. Therefore, to make the best use of an importance map, the map must be consistent with the source biasing. To achieve this consistency, a Sn calculation could be made to improve the importance map and the computational performance. The MAVRIC (Monaco with Automated Variance Reduction using Importance Calculations) code does that and this work intends to study the code options to generate the importance map. A pool type 10 MW research reactor was designed in a simple way just to study the prompt gamma rays penetration in the concrete and therefore study the CADIS methodology applied to point detectors and mesh tallies. By keeping constant the simulation time and the CPU (Central Processing Unit) power a significant improvement was achieved in the relative errors for the point detectors and for the mesh tally. (author)

  13. CHANGE IN DEFORMATION PROPERTIES MODELING OF CONCRETE IN PROTECTIVE STRUCTURES OF NUCLEAR REACTOR BY IONIZING RADIATION

    Directory of Open Access Journals (Sweden)

    E. K. Agakhanov

    2016-01-01

    Full Text Available The necessity of studying the effect impact of elementary particles impact on the strength and deformation materials properties used in protective constructions nuclear reactors and reactor technology has been stipulated. A nuclear reactor pressure vessel from prestressed concrete, combining the functions of biological protection is to be considered. The neutron flux problem distribution in the pressure vessel of a nuclear reactor has been solved. The solution is made in axisymmetric with the finite element method using a flat triangular finite element. Computing has been conducted in Matlab package. The comparison with the results has been obtained using the finite difference method, as well as the graphs of changes under the influence of radiation exposure and the elastic modulus of concrete radiation deformations have been constructed. The proposed method allows to simulate changes in the deformation properties of concrete under the influence of neutron irradiation. Results of the study can be used in the calculation of stress-strain state of structures, taking into account indirect heterogeneity caused by the physical fields influence.

  14. INTOR radiation shielding for personnel access

    International Nuclear Information System (INIS)

    Gohar, Y.; Abdou, M.

    1981-01-01

    The INTOR reactor shield system consists of the blanket, bulk shield, penetration shield, component shield, and biological shield. The bulk shield consists of two parts: (a) the inboard shield; and (b) the outboard shield. The distinction between the different components of the shield system is essential to satisfy the different design constraints and achieve various objectives

  15. Borated concrete for ZPPR fuel storage

    International Nuclear Information System (INIS)

    Gasidlo, J.M.

    1985-01-01

    Fuel handling at the Zero Power Plutonium Reactor (ZPPR) led to two requirements for storage of ZPPR fuel: a low neutron multiplication and shielded storage to minimize personnel doses. Boron-poisoned concrete was chosen as the storge medium with boron frit as the poisoning agent. The calculated effects of water content and boron concentration led to specifying a concrete with a water content that was higher than ordinary concrete. The finite size of the boron frit particles caused concern about reduced effectiveness due to self-shielding. The self-shielding was evaluated using optical path lengths for spheres and tabulated self-shielding for slabs. The results showed that the finite-sized particles were at least 80% as effective as infinitely dilute absorption. Neutron and gamma dose rates measured in the vault verified that personnel could work in the vault on a regular basis without exceeding personnel dose limits. 4 refs., 3 figs., 7 tabs

  16. The need to pressure test prestressed concrete reactor vessels

    International Nuclear Information System (INIS)

    Forgie, J.H.; Holland, J.A.

    1983-01-01

    In the period when PCRV were relatively unproven, proof pressure testing provided a useful demonstration of vessel integritiy and a confirmation of model testing and of analysis. No failures have occurred during concrete vessel tests in the UK or in the subsequent operational life of the vessels and much has been learned of their behaviour in service. The paper examines the advantages and disadvantages of proof testing PCRV in the light of the above increased knowledge of vessel performance. The paper draws attention to certain hypothetical loading cases that could be more onerous than the proof test and suggests that pressure testing could itself cause unnecessarily high loading to parts of the vessel. Always recognising the safety considerations and demonstrations of such are of prime importance, the authors suggest that a lower pressure level could be adopted without loss of original intent. In addition some ground rules are suggested as to cases where proof testing could be omitted. (orig./HP)

  17. Bonded or Unbonded Technologies for Nuclear Reactor Prestressed Concrete Containments

    International Nuclear Information System (INIS)

    Abrishami, Homayoun; Tcherner, Julia; Barre, Francis; Borgerhoff, Michael; Bumann, Urs; Calonius, Kim; Courtois, Alexis; Debattista, Jean-Marc; Gallitre, Etienne; Isard, Cedric; Elison, Oscar; Graves, Herman; Sircar, Madhumita; Huerta, Alejandro; White, Andrew; ); Jackson, Paul; Kjellin, Daniel; Lillhoek, Sofia; Louhivirta, Jari; Myllymaeki, Jukka; Vaelikangas, Pekka; Martin, Jose; Nakano, Makio; Puttonen, Jari; Rambach, Jean-Mathieu; Tarallo, Francois; Smith, Leslie; Stepan, Jan; Touret, Jean-Pierre; Varpasuo, Pentti

    2015-01-01

    OECD/NEA/CSNI Working Group on Integrity and Ageing of Components and Structures (WGIAGE) has the main mission to advance the current understanding of those aspects relevant to ensuring the integrity of structures, systems and components under design and beyond design loads, to provide guidance in choosing the optimal ways of dealing with challenges to the integrity of operating as well as new nuclear power plants, and to make use of an integrated approach to design, safety and plant life management. The work related to the risks of the loss of pre-stressing force in concrete structures has been in high priority during the activities of the concrete sub-group of WGIAGE. Therefore, the CAPS of WGIAGE: Study on post-tensioning methodologies in containments, was approved by CSNI in June 2009. In this study the two post-tensioning methodologies: bonded and un-bonded methods and their technological features are analysed. In the bonded technology, the tendon cannot slide in its duct due to the cement grouting which is injected after tensioning. In the un-bonded technology, the tendon can slide inside its duct, the corrosion protection is given by grease, wax or dry air. A key point concerning the assessment of durability and safety of prestressed concrete containments is the technology chosen for tendon protection: bonded with cement grout or un-bonded and protected by grease or soft products. The mechanical behaviour of the containment is directly influenced by the adherence of the tendons to the concrete, locally and under high stresses in case of severe accident. The bonded or un-bonded tendons of post-tensioned concrete containment of the Nuclear Power Plants have the major role of containment (balance of the pressure effect during design basis and beyond design accident). Many difficulties around the design, the construction and the in service inspection are related to the tendons. The main goal of the CAPS work was to clarify the consequences and necessary

  18. Modelling dielectric-constant values of concrete: an aid to shielding effectiveness prediction and ground-penetrating radar wave technique interpretation

    International Nuclear Information System (INIS)

    Bourdi, Taoufik; Rhazi, Jamal Eddine; Ballivy, Gérard; Boone, François

    2012-01-01

    A number of efficient and diverse mathematical methods have been used to model electromagnetic wave propagation. Each of these methods possesses a set of key elements which eases its understanding. However, the modelling of the propagation in concrete becomes impossible without modelling its electrical properties. In addition to experimental measurements; material theoretical and empirical models can be useful to investigate the behaviour of concrete's electrical properties with respect to frequency, moisture content (MC) or other factors. These models can be used in different fields of civil engineering such as (1) electromagnetic compatibility which predicts the shielding effectiveness (SE) of a concrete structure against external electromagnetic waves and (2) in non-destructive testing to predict the radar wave reflected on a concrete slab. This paper presents a comparison between the Jonscher model and the Debye models which is suitable to represent the dielectric properties of concrete, although dielectric and conduction losses are taken into consideration in these models. The Jonscher model gives values of permittivity, SE and radar wave reflected in a very good agreement with those given by experimental measurements and this for different MCs. Compared with other models, the Jonscher model is very effective and is the most appropriate to represent the electric properties of concrete.

  19. Gas and water permeability of concrete for reactor buildings--prototype scale specimens

    International Nuclear Information System (INIS)

    Mills, R.H.

    1987-02-01

    The permeability testing was performed on four concrete cylinders, 0.25 m in diameter and 2 m long, modelling the wall-thickness of reactor containment structures on the prototype scale. Tests were performed on the cylinders before and after artificial induction of longitudinal cracks, intented to model defects developing after some period of adverse service conditions. Permeability increased greatly with the introduction of longitudinal cracks in the concrete, and was also affected by moisture content and casting direction. The influence of reinforcing steel could not be resolved within the bounds of experimental variability. Ultrasound measurements were taken on each cylinder before and after cracking, and a correlation between increased permeability and lowered Ultrasonic Pulse Velocity was observed. Ultrasonic Pulse Velocity measurements thus show promise as a means of continuous monitoring of the integrity of the concrete barrier in service

  20. Computational methodology for the Oak Ridge Research Reactor (ORR) and Bulk Shielding Reactor (BSR): cross-section and validation. Volume 1

    International Nuclear Information System (INIS)

    Miller, L.F.; Williams, M.L.

    1986-03-01

    A neutronics library suitable for low-enrichment uranium (LEU) and high-enrichment uranium (HEU) fueled cores for both the Oak Ridge Research Reactor (ORR) and the Bulk Shielding Reactor (BSR) is documented herein. The library is obtained from version V of the Evaluated Nuclear Data File (ENDF/B-V) and contains 223 nuclides weighted over a variety of region-dependent neutron spectra. Self-shielding and zone-weighting effects are incorporated with 227-group calculations for several reactor-core configurations. Libraries are archived for both transport and diffusion theory seven-group calculations. Complete listings of processing details are included so that libraries with different specifications can be easily obtained. Results from validation calculations indicate that the neutronics libraries obtained from this effort are suitable for neutronics computations for the ORR and BSR. 12 refs., 5 figs., 15 tabs

  1. 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.)

  2. Long-Term Behaviors of the OPC Concrete with Fly-ash and Type V Concrete Applied on Reactor Containment Building

    International Nuclear Information System (INIS)

    Yoon, Eui Sik; Lee, Hee Taik; Paek, Yong Lak; Park, Young Soo

    2010-01-01

    The prestressed concrete has been used extensively in the construction of Reactor Containment Buildings (RCBs) in Korea in order to strengthen the RCBs and at the same time, prevent the release of radiation due to the Design Basis Accident and Design Basis Earthquake. It is well known that the prestressed concrete loses its prestressing force over the age, and the shrinkage and creep of the concrete significantly contributes to these long term prestressing losses. In this study, an evaluations of long term behaviors of the concrete such as creep and shrinkage have been performed for two types of concretes : Ordinary Portland Cement containing fly-ash used for the Shin- Kori 1 and 2 NPP and Type V cement used for the Ul- Chin 5 and 6 NPP

  3. Long-Term Behaviors of the OPC Concrete with Fly-ash and Type V Concrete Applied on Reactor Containment Building

    Energy Technology Data Exchange (ETDEWEB)

    Yoon, Eui Sik; Lee, Hee Taik; Paek, Yong Lak [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of); Park, Young Soo [Korea Hydro and Nuclear Power Co., Busan (Korea, Republic of)

    2010-10-15

    The prestressed concrete has been used extensively in the construction of Reactor Containment Buildings (RCBs) in Korea in order to strengthen the RCBs and at the same time, prevent the release of radiation due to the Design Basis Accident and Design Basis Earthquake. It is well known that the prestressed concrete loses its prestressing force over the age, and the shrinkage and creep of the concrete significantly contributes to these long term prestressing losses. In this study, an evaluations of long term behaviors of the concrete such as creep and shrinkage have been performed for two types of concretes : Ordinary Portland Cement containing fly-ash used for the Shin- Kori 1 and 2 NPP and Type V cement used for the Ul- Chin 5 and 6 NPP

  4. To the problem of reinforced concrete reactor vessel design and calculation

    International Nuclear Information System (INIS)

    Kirillov, A.P.; Artem'ev, V.P.; Bogopol'skij, V.G.; Nikolaev, Yu.B.; Paushkin, A.G.

    1980-01-01

    Modern methods for calculating reactor vessels of prestressed reinforced concrete are analyzed. It is shown that during the stage of technical and economical substantiation of reactor vessel structure for determining its stressed-deformed state engineering methods of calculation must be used, in particular, fragmentation method, method of rings and plates, and during the stages of contract and detail designs - method of finite elements and dynamic relaxation method. It is concluded that when solving cyclic symmetrical problems as well as asymmetrical problems, calculational algorithms for axis-symmetrical distributions of stresses in the vessel with provision for elastic properties of structural material may be used

  5. Analytical model for shear strength of end slabs of prestressed concrete nuclear reactor vessels

    International Nuclear Information System (INIS)

    Abdulrahman, H.O.; Sozen, M.A.; Schnobrich, W.C.

    1979-04-01

    The results are presented of an investigation of the behavior and strength of flat end slabs of cylindrical prestressed concrete nuclear reactor vessels. The investigation included tests of ten small-scale pressure vessels and development of a nonlinear finite-element model to simulate the deformation response and strength of the end slabs. Because earlier experimental studies had shown that the flexural strength of the end slab could be calculated using intelligible procedures, the emphasis of this investigation was on shear strength

  6. Cylindrical reinforced-concrete pressure vessel for nuclear reactors

    International Nuclear Information System (INIS)

    Vaessen, F.

    1975-01-01

    The cylindrical pressure vessel has got a wall and an isolating layer composed of blocks of heat-resistant concrete or of ceramic material. The side of the isolating layer facing the interior of the presssure vessel is coated by a liner made of metallic material. In cold state and without internal pressure, the radius of this liner is smaller by a differential amount than that of the isolating layer. By means of radially displaceable fixing elements consisting of an anchoring tube and a holding tube inserted in it, the liner can be made to rest against the isolating layer. This occurs if the pressure vessel is brought to operational temperature. The anchoring tube is attached to the isolating layer whereas the displaceable holding tube is connected with the liner. The possible relative travelling distance of these two elements is equal to the difference of length of the two radii. In addition, the liner may consist of single parts connected with each other through compensating flanges. There may also be additional springs arranged between the isolating layer and the liner. (DG/PB) [de

  7. Steel fibre concrete, a safer material for reactor construction. A general theory for rupture prediction

    International Nuclear Information System (INIS)

    Rammant, J.P.; Van Laethem, L.; Backx, E.

    1977-01-01

    The effect of steel fibre reinforcement on the mechanical behavior of concrete reactor structures is studied. It is shown that this material leads to a higher safety factor for highly stressed concrete structures like prestressed concrete pressure vessels. The reinforcement of concrete with short steel fibres results clearly in a fundamental change of the material properties. The study comprises basic experiments, the elaboration of an expression of the material laws, the development of a general computer program and the comparison of computational results with more elaborate experiments. Basic experimental work is conducted to determine the material characteristics of the fibre reinforced concrete. It is shown how the fibre reinforcement mechanism is translated into mathematical formulae by expressing the principal characteristics as matrix relationships. These relationships describe the elasto-plastic behavior and the cracked behavior. Probabilistic principles are used to express to fibre efficiency, such that a general stress-strain relationship is incorporated in a subsequent computer program. A general finite element program is developed which includes the new matrix relationships, the pull-out of fibres and the general stress-strain equations. A nonlinear calculation method gives the propagation of the distributed cracks with increasing load untill failure of the structure. Similarly, thermal cycling conditions are accounted for. For example the crack propagation in a fibre reinforced beam was measured by the photostress coating technique: the comparison with the computed crack propagation reveals an excellent agreement. Other comparative studies on simple structural parts are also reported

  8. Analysis and application of prestressed concrete reactor vessels for LMFBR containment

    International Nuclear Information System (INIS)

    Marchertas, A.H.; Fistedis, S.H.; Bazant, Z.P.; Belytschko, T.B.

    1978-01-01

    An analytical model of a prestressed concrete reactor vessel (PCRV) for LMFBR and the associated finite element computer code, involving an explicit time integration procedure, is described. The model is axisymmetric and includes simulations of the tensile cracking of concrete, the reinforcement, and a prestressing capability. The tensile cracking of concrete and the steel reinforcement are both modeled as continuously distributed within the finite element. The stresses in the reinforcement and concrete are computed separately and combined to give an overall stress state of the composite material. Attention is given to the fact that cracks do not form instantaneously, but develop gradually. Thus, after crack initiation the normal stress is reduced to zero gradually as a function of time. Residual shear resistance of cracks due to aggregate interlock is also taken into account. Prestressing of the PCRV is modeled by special structural members which represent an averaged prestressing layer equivalent to an axisymmetric shell. The internal prestressing members are superimposed over the reinforced concrete body of the PCRV; they are permitted to stretch and slide in a predetermined path, simulating the actual tendons. The validity of the code is examined by comparison with experimental data. (Auth.)

  9. Heavy metal oxide glasses as gamma rays shielding material

    International Nuclear Information System (INIS)

    Kaur, Preet; Singh, Devinder; Singh, Tejbir

    2016-01-01

    The gamma rays shielding parameters for heavy metal oxide glasses and concrete samples are comparable. However, the transparent nature of glasses provides additional feature to visualize inside the shielding material. Hence, different researchers had contributed in computing/measuring different shielding parameters for different configurations of heavy metal oxide glass systems. In the present work, a detailed study on different heavy metal (_5_6Ba, _6_4Gd, _8_2Pb, _8_3Bi) oxide glasses has been presented on the basis of different gamma rays shielding parameters as reported by different researchers in the recent years. It has been observed that among the selected heavy metal oxide glass systems, Bismuth based glasses provide better gamma rays shielding. Hence, Bismuth based glasses can be better substitute to concrete walls at nuclear reactor sites and nuclear labs.

  10. Heavy metal oxide glasses as gamma rays shielding material

    Energy Technology Data Exchange (ETDEWEB)

    Kaur, Preet; Singh, Devinder; Singh, Tejbir, E-mail: dr.tejbir@gmail.com

    2016-10-15

    The gamma rays shielding parameters for heavy metal oxide glasses and concrete samples are comparable. However, the transparent nature of glasses provides additional feature to visualize inside the shielding material. Hence, different researchers had contributed in computing/measuring different shielding parameters for different configurations of heavy metal oxide glass systems. In the present work, a detailed study on different heavy metal ({sub 56}Ba, {sub 64}Gd, {sub 82}Pb, {sub 83}Bi) oxide glasses has been presented on the basis of different gamma rays shielding parameters as reported by different researchers in the recent years. It has been observed that among the selected heavy metal oxide glass systems, Bismuth based glasses provide better gamma rays shielding. Hence, Bismuth based glasses can be better substitute to concrete walls at nuclear reactor sites and nuclear labs.

  11. 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)

  12. Concrete

    OpenAIRE

    Kruse Aagaard, Anders

    2015-01-01

    Concrete is a component of coherent transition between a concrete base and a wooden construction. The structure is based on a quantity of investigations of the design possibilities that arise when combining digital fabrication tools and material capacities.Through tangible experiments the project discusses materiality and digitally controlled fabrications tools as direct expansions of the architect’s digital drawing and workflow. The project sees this expansion as an opportunity to connect th...

  13. The Use of Prestressed Concrete Vessels in the French Power Reactor Programme

    International Nuclear Information System (INIS)

    Conte, F.; Dambrine, C.; Gaussot, D.

    1963-01-01

    This paper deals with the use of pre-stressed concrete for the G2 and G3 reactors at Marcoule and for the EDF3 reactor now under construction at Chinon. The first two reactors have been operating at power since 1959 and 1960 respectively. Messrs. Conte and Dambrine discuss the problems that arose during construction of the vessels for G2 and G3 and also deal with the experience gained in operation - experience which suggests that they are extremely safe- Work on the EDF3 vessel, begun at Chinon in the second half of 1961, is still under way and should be finished towards the end of 1963. Mr. Gaussot discusses the reasons for choosing this type of vessel, the results of calculations and mock-up tests, and the problems presented by the construction itself. A number of studies have been devoted to the future prospects of prestressed concrete structures for reactors. It would seem that working pressures could be increased, if desired, and, in any case, that dimensions could be considerably enlarged, thus offering the chance of integral-type solutions. (author) [fr

  14. New developments in resonant mixture self-shielding treatment with Apollo code and application to Jules Horowitz reactor core calculation

    International Nuclear Information System (INIS)

    Coste-Delclaux, M.; Aggery, A.; Huot, N.

    2005-01-01

    APOLLO2 is a modular multigroup transport code developed by Cea in Saclay. Until last year, 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. Last year, 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 calculations of a simplified Jules Horowitz reactor using a Monte-Carlo code (TRIPOLI4) as a reference and APOLLO2 in its standard and improved versions, show that, as far as the effective multiplication factor is concerned, the mixture treatment does not bring an improvement, because the new treatment suppresses compensation between the reaction rate discrepancies. The discrepancy of 300 pcm that appears with the reference calculation is in accordance with the technical specifications of the Jules Horowitz reactor

  15. Nuclear reactor plant with a gas-cooled nuclear reactor situated in a cylindrical prestressed concrete pressure vessel

    International Nuclear Information System (INIS)

    Becker, G.; Elter, C.; Fritz, R.; Rautenberg, J.; Schoening, J.; Stracke, W.

    1986-01-01

    A simplified construction of the nuclear reactor plant with a guarantee of great safety is achieved by the auxiliary heat exhangers, which remove the post-shutdown heat in fault situations, being arranged in the wellknown way in pairs above one another in a vertical shaft. The associated auxiliary blowers are situated at the top for the upper auxiliary heat exchangers and at the bottom for the lower auxiliary heat exchangers. The cold gas is taken from the lower auxiliary blowers through a parallel gas pipe laid in concrete, which enters the vertical shaft concerned in the area of the cold gas pipe. (orig./HP) [de

  16. Tower Shielding Reactor II design and operation report. Vol. 3. Assembling and testing of the control mechanism assembly

    International Nuclear Information System (INIS)

    Ward, D.R.; Holland, L.B.

    1979-09-01

    The mechanisms that are operated to control the reactivity of the Tower Shielding Reactor II(TSR-II) are mounted on a Control Mechanism Housing (CMH) that is centered inside the reactor core. The information required to procure, fabricate, inspect, and assemble a CMH is contained in the ORNL engineering drawings listed in the appropriate sections. The components are fabricated and inspected from these drawings in accordance with a Quality Assurance Plan and a Manufacturing Plan. The material in this report describes the acceptance and performance tests of CMH subassemblies used ty the Tower Shielding Facility (TSF) staff but it can also be used by personnel fabricating the components. This information which was developed and used before the advent of the formalized QA Program and Manufacturing Plans evolved during the fabrication and testing of the first five CMHs

  17. Study of the radiation scattered and produced by concrete shielding of radiotherapy rooms and its effects on equivalent doses in patients' organs

    International Nuclear Information System (INIS)

    Braga, K.L.; Rebello, W.F.; Andrade, E.R.; Gavazza, S.; Medeiros, M.P.C.; Mendes, R.M.S.; Gomes, R.G.; Silva, M.G.; Thalhofer, J.L.; Silva, A.X.; Santos, R.F.G.

    2015-01-01

    Within a radiotherapy room, in addition to the primary beam, there is also secondary radiation due to the leakage of the accelerator head and the radiation scattering from room objects, patient and even the room's shielding itself, which is projected to protect external individuals disregarding its effects on the patient. This work aims to study the effect of concrete shielding wall over the patient, taking into account its contribution on equivalent doses. The MCNPX code was used to model the linear accelerator Varian 2100/2300 C/D operating at 18MeV, with MAX phantom representing the patient undergoing radiotherapy treatment for prostate cancer following Brazilian Institute of Cancer four-fields radiation application protocol (0°, 90°, 180° and 270°). Firstly, the treatment was patterned within a standard radiotherapy room, calculating the equivalent doses on patient's organs individually. In a second step, this treatment was modeled withdrawing the walls, floor and ceiling from the radiotherapy room, and then the equivalent doses calculated again. Comparing these results, it was found that the concrete has an average shielding contribution of around 20% in the equivalent dose on the patient's organs. (author)

  18. 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.

  19. The application of external vibration monitoring to reactors with concrete pressure vessels

    International Nuclear Information System (INIS)

    Hammill, W.J.

    1979-01-01

    The application of external vibration monitoring techniques to advanced gas cooled reactors (AGR) which have concrete pressure vessels is considered. A monitoring system for a particular AGR coolant circuit structure is developed, whose primary objective is to detect impacting of two components, although the detection of forced vibration response is also considered. Experimental results from instrumented components in the reactor and data from rig tests on full size units have been used together with a mathematical model of some elements of the transmission path in order to establish its dynamic characteristics and relate internal component vibration to externally measured signals. The application of external vibration monitoring to the external detection of the forced vibration response of an internal reactor assembly and the remote monitoring of circulator sound output is discussed. (author)

  20. Design and manufacturing concrete cells for shielding and storing radioactive semi liquid waste (resin) from MPR-GAS

    International Nuclear Information System (INIS)

    Pudjijanto-MS; Bahdir-Johan

    2003-01-01

    Semi liquid or quasi solid waste on Multipurpose Reactor G.A. Siwabessy (MPR-GAS) produced from operating resin rinsing systems and resin disposal systems during changes insert trap resin. Volume of the disposal resin waste in the filter mixed-bed per operation rinsing period are approx. 1.00 m 3 (in the Primary Cooling Water Treatment System) with activity ∼ 18.6 Ci/m 3 (0.688 TBq/m 3 ), 0.50 m 3 (in the Radioisotope Storage Pool Water Treatment System) with activity approx ∼ 0.162 Ci/m 3 (5.99 x 10 3 MBq/m 3 ) and 0.50 m 3 (in the Interim Spent Fuel Storage Pool Water Treatment System) with activity ∼ 0.162 Ci/m 3 (5.99 x 10 3 MBq/m 3 ) respectively. On the discharging and unloading, the gross radioactivity concentration of the resin waste loaded in the disposal resin waste tank are approx. 10 Ci/m 3 (0.37 TBq/m 3 ). After 6 months delayed, this activity is still 0.32 Ci/m 3 (11.84 GBq/m 3 ). Based on this data, some concrete cells to storage resin waste as semi liquid or quasi solid waste produced continuously by MPR-GAS installation has been designed and manufactured eternally

  1. Assessment of models for steam release from concrete and implications for modeling corium behavior in reactor cavities

    International Nuclear Information System (INIS)

    Washington, K.E.; Carroll, D.E.

    1988-01-01

    Models for concrete outgassing have been developed and incorporated into a developmental version of the CONTAIN code for the assessment of corium behavior in reactor cavities. The resultant code, referred to as CONTAIN/OR in order to distinguish it from the released version of CONTAIN, has the capability to model transient heat conduction and concrete outgassing in core-concrete interaction problems. This study focused on validation and assessment of the outgassing model through comparisons with other concrete response codes. In general, the model is not mechanistic; however, there are certain important processes and feedback effects that are treated rigorously. The CONTAIN outgassing model was compared against two mechanistic concrete response codes (USINT and SLAM). Gas release and temperature profile predictions for several concrete thicknesses and heating rates were performed with acceptable agreement seen in each case. The model was also applied to predict corium behavior in a reactor cavity for a hypothetical severe accident scenario. In this calculation, gases evolving from the concrete during nonablating periods fueled exothermic Zr chemical reactions in the corium. Higher corium temperatures and more concrete ablation were observed when compared with that seen when concrete outgassing was neglected. Even though this result depends somewhat upon the makeup of the corium sources and the concrete type in the cavity, it does show that concrete outgassing can be important in the modeling of corium behavior in reactor cavities. In particular, the need to expand the traditional role of CORCON from steady-state ablation to the consideration of more transient events is clearly evident as a result of this work. 5 refs., 11 figs., 1 tab

  2. Calculation of concrete shielding wall thickness for 450kVp X-ray tube with MCNP simulation and result comparison with half value layer method calculation

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Sang Heon; Lee, Eun Joong; Kim, Chan Kyu; Cho, Gyu Seong [Dept. of Nuclear and Quantum Engineering, KAIST, Daejeon (Korea, Republic of); Hur, Sam Suk [Sam Yong Inspection Engineering Co., Ltd., Seoul (Korea, Republic of)

    2016-11-15

    Radiation generating devices must be properly shielded for their safe application. Although institutes such as US National Bureau of Standards and National Council on Radiation Protection and Measurements (NCRP) have provided guidelines for shielding X-ray tube of various purposes, industry people tend to rely on 'Half Value Layer (HVL) method' which requires relatively simple calculation compared to the case of those guidelines. The method is based on the fact that the intensity, dose, and air kerma of narrow beam incident on shielding wall decreases by about half as the beam penetrates the HVL thickness of the wall. One can adjust shielding wall thickness to satisfy outside wall dose or air kerma requirements with this calculation. However, this may not always be the case because 1) The strict definition of HVL deals with only Intensity, 2) The situation is different when the beam is not 'narrow'; the beam quality inside the wall is distorted and related changes on outside wall dose or air kerma such as buildup effect occurs. Therefore, sometimes more careful research should be done in order to verify the effect of shielding specific radiation generating device. High energy X-ray tubes which is operated at the voltage above 400 kV that are used for 'heavy' nondestructive inspection is an example. People have less experience in running and shielding such device than in the case of widely-used low energy X-ray tubes operated at the voltage below 300 kV. In this study, Air Kerma value per week, outside concrete shielding wall of various thickness surrounding 450 kVp X-ray tube were calculated using MCNP simulation with the aid of Geometry Splitting method which is a famous Variance Reduction technique. The comparison between simulated result, HVL method result, and NCRP Report 147 safety goal 0.02 mGy wk-1 on Air Kerma for the place where the public are free to pass showed that concrete wall of thickness 80 cm is needed to achieve the

  3. Calculation of concrete shielding wall thickness for 450kVp X-ray tube with MCNP simulation and result comparison with half value layer method calculation

    International Nuclear Information System (INIS)

    Lee, Sang Heon; Lee, Eun Joong; Kim, Chan Kyu; Cho, Gyu Seong; Hur, Sam Suk

    2016-01-01

    Radiation generating devices must be properly shielded for their safe application. Although institutes such as US National Bureau of Standards and National Council on Radiation Protection and Measurements (NCRP) have provided guidelines for shielding X-ray tube of various purposes, industry people tend to rely on 'Half Value Layer (HVL) method' which requires relatively simple calculation compared to the case of those guidelines. The method is based on the fact that the intensity, dose, and air kerma of narrow beam incident on shielding wall decreases by about half as the beam penetrates the HVL thickness of the wall. One can adjust shielding wall thickness to satisfy outside wall dose or air kerma requirements with this calculation. However, this may not always be the case because 1) The strict definition of HVL deals with only Intensity, 2) The situation is different when the beam is not 'narrow'; the beam quality inside the wall is distorted and related changes on outside wall dose or air kerma such as buildup effect occurs. Therefore, sometimes more careful research should be done in order to verify the effect of shielding specific radiation generating device. High energy X-ray tubes which is operated at the voltage above 400 kV that are used for 'heavy' nondestructive inspection is an example. People have less experience in running and shielding such device than in the case of widely-used low energy X-ray tubes operated at the voltage below 300 kV. In this study, Air Kerma value per week, outside concrete shielding wall of various thickness surrounding 450 kVp X-ray tube were calculated using MCNP simulation with the aid of Geometry Splitting method which is a famous Variance Reduction technique. The comparison between simulated result, HVL method result, and NCRP Report 147 safety goal 0.02 mGy wk-1 on Air Kerma for the place where the public are free to pass showed that concrete wall of thickness 80 cm is needed to achieve the safety goal

  4. Design and analysis of reactor containment of steel-concrete composite laminated shell

    International Nuclear Information System (INIS)

    Ichikawa, K.

    1977-01-01

    Reinforced and prestressed concrete containments for reactors have been developed in order to avoid the difficulties of welding of steel containments encountered as their capacities have become large: growing thickness of steel shells gave rise to the requirement of stress relief at the construction sites. However, these concrete vessels also seem to face another difficulty: the lack of shearing resistance capacity. In order to improve the shearing resistance capacity of the containment vessel, while avoiding the difficulty of welding, a new scheme of containment consisting of steel-concrete laminated shell is being developed. In the main part of a cylindrical vessel, the shell consists of two layers of thin steel plates located at the inner and outer surfaces, and a layer of concrete core into which both the steel plates are anchored. In order to validate the feasibility and safety of this new design, the results of analysis on the basis of up-to-date design loads are presented. The results of model tests in 1:30 scale are also reported. (Auth.)

  5. A comparison of elastic-plastic and variable modulus-cracking constitutive models for prestressed concrete reactor vessels

    International Nuclear Information System (INIS)

    Anderson, C.A.; Smith, P.D.

    1979-01-01

    Numerical prediction of the behavior of prestressed concrete reactor vessels (PCRVs) under static, dynamic and long term loadings is complicated by the currently ill-defined behavior of concrete under stress and the three-dimensional nature of PCRVs. Which constitutive model most closely approximates the behavior of concrete in PCRVs under load has not yet been decided. Many equations for accurately modeling the three-dimensional behavior of PCRVs tax the capability of a most up-to-date computing system. The main purpose of this paper is to compare the characteristics of two constitutive models which have been proposed for concrete, variable modulus cracking model and elastic-plastic model. Moreover, the behavior of typical concrete structures was compared, the materials of which obey these constitutive laws. The response to internal pressure of PCRV structure, the constitutive models for concrete, the test problems using a thick-walled concrete ring and a rectangular concrete plate, and the analysis of an axisymmetric concrete pressure vessel PV-26 using the variable modulus cracking model of the ADINA code are explained. The variable modulus cracking model can predict the behavior of reinforced concrete structures well into the range of nonlinear behavior. (Kako, I.)

  6. Development of fast reactor containment safety analysis code, CONTAIN-LMR. (3) Improvement of sodium-concrete reaction model

    International Nuclear Information System (INIS)

    Kawaguchi, Munemichi; Doi, Daisuke; Seino, Hiroshi; Miyahara, Shinya

    2015-01-01

    A computer code, CONTAIN-LMR, is an integrated analysis tool to predict the consequence of severe accident in a liquid metal fast reactor. Because a sodium-concrete reaction behavior is one of the most important phenomena in the accident, a Sodium-Limestone Concrete Ablation Model (SLAM) has been developed and installed into the original CONTAIN code at Sandia National Laboratories (SNL) in the U.S. The SLAM treats chemical reaction kinetics between the sodium and the concrete compositions mechanistically using a three-region model, containing a pool (sodium and reaction debris) region, a dry (boundary layer (B/L) and dehydrated concrete) region, and a wet (hydrated concrete) region, the application is limited to the reaction between sodium and limestone concrete. In order to apply SLAM to the reaction between sodium and siliceous concrete which is an ordinary structural concrete in Japan, the chemical reaction kinetics model has been improved to consider the new chemical reactions between sodium and silicon dioxide. The improved model was validated to analyze a series of sodium-concrete experiments which were conducted in Japan Atomic Energy Agency (JAEA). It has been found that relatively good agreement between calculation and experimental results is obtained and the CONTAIN-LMR code has been validated with regard to the sodium-concrete reaction phenomena. (author)

  7. Time varying stress in ligaments of perforated plates with reference to prestressed concrete reactor vessels

    International Nuclear Information System (INIS)

    Stefanou, G.D.

    1978-01-01

    The work described herein relates to the prediction of stresses in materials which exhibit time varying strains with particular reference to the ligaments of perforated circular concrete slabs, subjected to long-term radial prestress and uniform elevated temperature. The perforations are reinforced with steel liners and arranged in a square central lattice symmetrical about two orthogonal axes. Special reference is made to the distribution of stress in the standpipe region of prestressed concrete cylindrical pressure or containment vessels for gas cooled reactors. In order to assess the stress distribution around the perforated zone of a circular slab, a method of analysis was developed by the author, based on the ''Equivalent Elastic Modulus'' of the perforated zone and the ''Effective Modulus Method'', utilizing experimental data obtained from tests performed on model specimens. The object of this paper is to extend the above method of analysis into the perforated region, and assess the long-term stresses in the ligaments. The proposed method is accomplished by an application of the Finite Element Method for the elastic plane stress case. Comparisons of experimental results and theoretical predictions by the proposed method, and other analytical methods are made for a series of perforated concrete slabs subjected to radial in-plane loading: 10,342 kN/m 2 (1,5000 psi), and uniform elevated temperature of 80 0 C. The investigation, though in general terms, could be applied to the perforated region of cylindrical pressure vessels for nuclear reactors. Finally the paper describes briefly in Appendix 3 a direct solution procedure for calculating time dependent stresses in concrete structures based on the principles of variational calculus. Analytical predictions obtained by the proposed method which is a step-by-step analysis, are compared with the variational principle method. (author)

  8. Concrete works for Hamaoka No. 1 nuclear power plant

    International Nuclear Information System (INIS)

    Horiuchi, Minoru; Sugihara, Kazuo; Iwasawa, Jiro.

    1975-01-01

    Various aspects of concrete works performed for the reactor building of Hamaoka No.1 plant are reviewed. Control building and waste disposal building were all together combined with the reactor building in order to improve safety against earthquakes. Special consideration was given for the quality control of concrete works by establishing quality control committee, making quality control manual and by performing daily examination and monthly report. The quality and various materials of concrete used are described. The composition of concrete used for various parts of the building is also listed. Detailed description is made regarding the concrete placing for foundation mat, under a containment vessel, and the construction of air gaps and the placing of shielding concrete around the containment vessel. Curves representing the temperature history of concrete at various points are presented. As for testing, the items of test, methods of measurement, and the results of these test and measurement are presented in detail. (Aoki, K.)

  9. The moisture conditions of nuclear reactor concrete containment walls - an example for a BWR reactor

    Energy Technology Data Exchange (ETDEWEB)

    Nilsson, L.O.; Johansson, P. [Lund Institute of Technology, Laboratory of Building Materials, PO Box 118, 221 00 Lund (Sweden)

    2006-07-01

    A method is presented on how to quantify the moisture conditions of nuclear concrete containment walls. The method is based on first quantifying the boundary conditions at the outer and inner surfaces and then describing the moisture fixation and moisture transport within the concrete wall. The temperature and humidity conditions of the outdoor air and of the air close to the wall surfaces are monitored for a period of time and the vapour contents in the different points are compared. From the differences between the vapour contents the sources of moisture are identified and quantified. The previous and future climatic conditions are then predicted. An example is given for the conditions in the containment walls at Barsebaeck nuclear power plant, where moisture measurements have been performed in situ and on samples taken from the walls. (authors)

  10. Extensive neutronic sensitivity-uncertainty analysis of a fusion reactor shielding blanket

    International Nuclear Information System (INIS)

    Hogenbirk, A.

    1994-01-01

    In this paper the results are presented of an extensive neutronic sensitivity-uncertainty study performed for the design of a shielding blanket for a next-step fusion reactor, such as ITER. A code system was used, which was developed at ECN Petten. The uncertainty in an important response parameter, the neutron heating in the inboard superconducting coils, was evaluated. Neutron transport calculations in the 100 neutron group GAM-II structure were performed using the code ANISN. For the sensitivity and uncertainty calculations the code SUSD was used. Uncertainties due to cross-section uncertainties were taken into account as well as uncertainties due to uncertainties in energy and angular distributions of scattered neutrons (SED and SAD uncertainties, respectively). The subject of direct-term uncertainties (i.e. uncertainties due to uncertainties in the kerma factors of the superconducting coils) is briefly touched upon. It is shown that SAD uncertainties, which have been largely neglected until now, contribute significantly to the total uncertainty. Moreover, the contribution of direct-term uncertainties may be large. The total uncertainty in the neutron heating, only due to Fe cross-sections, amounts to approximately 25%, which is rather large. However, uncertainty data are scarce and the data may very well be conservative. It is shown in this paper that with the code system used, sensitivity and uncertainty calculations can be performed in a straightforward way. Therefore, it is suggested that emphasis is now put on the generation of realistic, reliable covariance data for cross-sections as well as for angular and energy distributions. ((orig.))

  11. Design of an integral missile shield in integrated head assembly for pressurized water reactor at commercial nuclear plants

    International Nuclear Information System (INIS)

    Baliga, Ravi; Watts, Tom Neal; Kamath, Harish

    2015-01-01

    In ICONE22, the authors presented the Integrated Head Assembly (IHA) design concept implemented at Callaway Nuclear Power Plant in Missouri, USA. The IHA concept is implemented to reduce the outage duration and the associated radiation exposure to the workers by reducing critical path time during Plant Refueling Outage. One of the head area components in the IHA is a steel missile shield designed to protect the Control Rod Drive Mechanism (CRDM) assembly from damaging other safety-related components in the vicinity in the Containment. Per Federally implemented General Design Criteria for commercial nuclear plants in the USA, the design of Nuclear Steam Supply System (NSSS) must provide protection from the damages caused by a postulated event of CRDM housing units and their associated parts disengaging from the reactor vessel assembly. This event is considered as a Loss of Coolant Accident (LOCA) and assumes that once the CRDM housing unit and their associated parts disengage from the reactor vessel internals assembly, they travel upward by the water jet with the following sequence of events: Per Reference 1, the drive shaft and control rod cluster are forced out of the reactor core by the differential pressure across the drive shaft with the assumption that the drive shaft and control rod cluster, latched together, are fully inserted when the accident occurs. After the travel, the rod cluster control spider will impact the lower side of the upper support plate inside the reactor vessel fracturing the flexure arms in the joint freeing the drive shaft from the control rod cluster. The control rod cluster is stopped by the upper support plate and will remain below the upper support plate during this accident. However, the drive shaft will continue to accelerate in the upward direction until it is stopped by a safety feature in the IHA. The integral missile shield as a safety feature in the IHA is designed to stop the CRDM drive shaft from moving further up in the

  12. A prestressed concrete pressure vessel for helium high temperature reactor system

    International Nuclear Information System (INIS)

    Horner, R.M.W.; Hodzic, A.

    1976-01-01

    A novel prestressed concrete pressure vessel has been developed to provide the primary containment for a fully integrated system comprising a high temperature nuclear reactor, three horizontally mounted helium turbines, associated heat exchangers and inter-connecting ducts. The design and analysis of the pressure vessel is described. Factors affecting the final choice of layout are discussed, and earlier development work seeking to resolve the conflicting requirements of the structural, mechanical, and system engineers outlined. Proposals to increase the present output of about 1000 MW of electrical power to over 3000 MW, by incorporating four turbines in a single pressure vessel are presented. (author)

  13. New containment concept for light-water reactors: double shield without barrier seal

    International Nuclear Information System (INIS)

    Costaz, J.L.

    1975-01-01

    A new containment system has been developed by EDF for its future PWR power plants (1350 MW). It comprises one internal prestressed concrete enclosure without barrier and one external reinforced concrete enclosure. The research and development program includes concrete seal tests, the detection of leaks on structures, the performance of structures under projectile input. These tests as well as the study of the radiological effects of accidental loss of primary coolant show that this type of containment marks a definite step from the point of view of safety, as compared with the solution prestressed concrete plus metallic barrier [fr

  14. The status of shielding research at Tajoura research center

    International Nuclear Information System (INIS)

    El-Bakkoush, F.A.

    2005-01-01

    This paper gives a description to the shielding research activities which have been carried-out at the radiation shielding group ,Tajoura Research Center. This includes the design of different types of concrete shields made from local aggregates which have suitable radiation attenuation properties. These include, Ordinary Concrete(with density p = 2.3 ton/m3) heavy weight concrete (with density p =3.6 ton/m3) and heat resistant concrete with aggregates having bound- in water. Investigation have been carried -out by measuring the neutron and gamma-rays spectra which have been transmitted through barriers having different thickness. These were performed using a collimated beam of reactor neutrons and gamma-ray transmitted from the horizontal channel no 1 of Tajoura-Research reactor with 10 MW Max ape rating power. The transmitted fast neutron and gamma spectra were measured by neutron-gamma spectrometer employing NE-213 liquid organic scintillater. Discrimination of against undesired pulses of neutrons or gamma-ray was achieved by a pulse shape discrimination method based on differences in the shape of the decay part of the emitted pulses. The obtained results are presented in the form of displayed neutron and gamma spectra measured behind different thickness of the investigated concrete shield. These spectra were used to derive the macroscopic cross section for at different energy for material under investigation

  15. The results of the CCI-3 reactor material experiment investigating 2-D core-concrete interaction and debris coolability with a siliceous concrete crucible

    International Nuclear Information System (INIS)

    Farmer, M.T.; Basu, S.

    2006-01-01

    The OECD-sponsored Melt Coolability and Concrete Interaction (MCCI) program is conducting reactor material experiments and associated analysis with the objectives of resolving the ex-vessel debris coolability issue, and to address remaining uncertainties related to long-term two-dimensional molten core-concrete interactions under both wet and dry cavity conditions. Achievement of these two objectives will demonstrate the efficacy of severe accident management guidelines for existing plants and provide the technical basis for better containment designs for future plants. Despite years of international research, there are remaining uncertainties in the models that evaluate the lateral vs. axial power split during core-concrete interaction because of a lack of truly two-dimensional experiment data. As a result, there are differences in the 2-D cavity erosion predicted by codes such as MELCOR, WECHSL, and COSACO. In the continuing effort to bridge this data gap, the third in a series of large scale Core-Concrete Interaction experiments (CCI-3) has been conducted as part of the MCCI program. This test involved the interaction of a 375 kg core-oxide melt within a two-dimensional siliceous concrete crucible. The initial phase of the test was conducted under dry conditions. After a predetermined ablation depth was reached, the cavity was flooded to obtain data on the coolability of a core melt after core-concrete interaction has progressed for some time. This paper provides a summary description of the test facility and an overview of test results

  16. Latest developments in prestressed concrete vessels for gas-cooled reactors

    International Nuclear Information System (INIS)

    Ople, F.S. Jr.

    1979-01-01

    This paper is an update of the design development of prestressed concrete vessels, commonly referred to as 'PCRVs' starting with the first single-cavity PCRV for the Fort St. Vrain Nuclear Generating Station to the latest multi-cavity PCRV configurations being utilized as the primary reactor vessels for both the High Temperature Gas-Cooled Reactor (HTGR) and the Gas-Cooled Fast Breeder Reactor (GCFR) in the U.S.A. The complexity of PCRV design varies not only due to the type of vessel configuration (single versus multi-cavity) but also on the application to the specific type of reactor concept. PCRV technology as applied to the Steam Cycle HTGR is fairly well established; however, some significant technical complexities are associated with PCRV design for the Gas Turbine HTGR and the GCFR. For the Gas Turbine HTGR, for instance, the fluid dynamics of the turbo-machinery cause multi-pressure conditions to exist in various portions of the power conversion loops during operation. This condition complicates the design approach and the proof test specification for the PCRV. The geometric configuration of the multi-cavity PCRV is also more complex due to the introduction of large horizontal cylindrical cavities (housing the turbo/machines for the Gas Turbine HTGR and circulators for the GCFR) in addition to the vertical cylindrical cavities for the core and heat exchangers. Because of this complex geometry, it becomes difficult to achieve an optimum prestressing arrangement for the PCRV. Other novel features of the multi-cavity PCRV resulting from the continuing design optimization effort are the incorporation of an asymmetric (offset core) configuration and the use of large vessel cavity/penetration concrete closures directly held down by prestressing tendons for both economic and safety reasons. (orig.)

  17. 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.

  18. 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

  19. Meeting 'Prestressed-concrete reactor pressure vessels', 13th and 14th october 1975, Berlin

    International Nuclear Information System (INIS)

    Schickert, G.

    1976-01-01

    Influence of radioactive radiation on the mechanical properties of concrete; behaviour of concrete in short-time testing under multiaxial mechanical stresses; behaviour of concrete in long-time testing under multiaxial mechanical stresses at higher temperatures; temperature stress of concrete; strength formation of concrete; steel fiber concrete. (LH) [de

  20. An assessment of fuel freezing and drainage phenomena in a reactor shield plug following a core disruptive accident

    International Nuclear Information System (INIS)

    El-Genk, M.; Cronenberg, A.W.

    1978-01-01

    An important problem related to the assessment of the recriticality potential for an LMFBR following a core disruptive accident is an understanding of the freezing phenomena of molten fuel on a cold structure which may prevent fuel dispersal and sunsequent shutdown. Transient analytical freezing and drainage calculations have been applied to molten UO 2 travel through the rather cold lower shield plug of the Clinch River Breeder Reactor (CRBR). The successive approximation technique is used to obtain a solution of the non-linear freezing problem, where such effects as heat generation, viscous heat dissipation, temperature dependent thermophysical properties and a convective boundary condition at the solidification front have been incorporated into the present analytical formulation. Results indicate that previous steady-state analysis overestimate the rate of frozen layer build-up by about a factor of two. However, of primary importance is the driving force for drainage and the diameter of the shield plug flow channel. (Auth.)

  1. 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)

  2. 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)

  3. Design criteria for prestressed concrete pressure vessels for high temperature reactors

    International Nuclear Information System (INIS)

    Schimmelpfennig, K.

    1991-01-01

    This paper summarizes the work on design criteria for concrete structures of Prestressed Concrete Reactor Vessels (PCRVs), which has been carried out since 1984 by a couple of competent institutions. After some basic considerations on the safety demands on PCRVs, especially their Prestressed Concrete Structure (PCS), and the consequences for an elevated level of quality to be ensured by the design criteria, an impression is given, first, by what means a higher quality standard is gained with respect to selection of materials and specification of material data in comparison to the usual building industry and what kind of criteria on this behalf should be fixed in a PCRV code. As a further quality increasing feature, the specific demands on design analysis as practised according to the present state of science and as to be treated within a code are discussed. This concerns analyses for steady state and transient temperatures as well as stress and strain analyses for service and ultimate load conditions. It is outlined to what degree calculation models should be detailed, which includes statements about admissible idealizations. As a central topic the question is discussed in what way the ultimate load capacity has to be evaluated, thereby presenting results of some investigations pointing out the conditions under which the design is determined by the different kinds of ultimate load conditions. Finally, some reflections on the demands on monitoring the PCS behaviour during its lifetime and on several questions still to be answered in this field are expressed. (orig.)

  4. High-temperature gas-cooled reactor technology development program. Annual progress report for period ending December 31, 1980

    International Nuclear Information System (INIS)

    1981-08-01

    Research activities are described concerning HTGR chemistry; fueled graphite development; prestressed concrete pressure vessel development; structural materials; HTGR graphite studies; HTR core evaluation; reactor physics; shielding; application and project assessments; and HTR Core Flow Test Loop studies

  5. High-temperature gas-cooled reactor technology development program. Annual progress report for period ending December 31, 1980

    Energy Technology Data Exchange (ETDEWEB)

    1981-08-01

    Research activities are described concerning HTGR chemistry; fueled graphite development; prestressed concrete pressure vessel development; structural materials; HTGR graphite studies; HTR core evaluation; reactor physics; shielding; application and project assessments; and HTR Core Flow Test Loop studies.

  6. Safety assessment of a multicavity prestressed concrete reactor vessel with hot liner

    Energy Technology Data Exchange (ETDEWEB)

    Lafitte, R.; Marchand, J. D. [Bonnard et Gardel, Ingenieurs-Conseil, Lausanne (Switzerland)

    1981-01-15

    The prestressed concrete reactor vessel of the high temperature reactor with helium turbine project differs from those realized up to this day by the important number of cavities, by the different cavity pressures and by a liner in contact with hot gas. For the cases of operating conditions, the computations can be based on an identical pressure in all the cavities. The overdimensioning of the vessel which results is not a determining factor at this stage of the project. The possible loss of leaktightness of the liner can introduce gas pressure into the walls of the vessel. The great thickness of the walls makes it impossible to withstand the resulting forces with prestressing in offering sufficient safety factor against collapse. It is thus important to design a drainage network largely dimensioned. The warm liner appears at this stage of the project too highly stressed by fatigue at the singularity points (ducts between cavities, angles). A solution is proposed which limits the variations of thermal stresses by using a steel with low coefficient of thermal expansion. The cavity closures, which are numerous and some with large dimensions are an important aspect of the vessel safety. A solution of reinforced concrete shell with independent liner is proposed.

  7. Safety assessment of a multicavity prestressed concrete reactor vessel with hot liner

    International Nuclear Information System (INIS)

    Lafitte, R.; Marchand, J.D.

    1981-01-01

    The prestressed concrete reactor vessel of the high temperature reactor with helium turbine project differs from those realized up to this day by the important number of cavities, by the different cavity pressures and by a liner in contact with hot gas. For the cases of operating conditions, the computations can be based on an identical pressure in all the cavities. The overdimensioning of the vessel which results is not a determining factor at this stage of the project. The possible loss of leaktightness of the liner can introduce gas pressure into the walls of the vessel. The great thickness of the walls makes it impossible to withstand the resulting forces with prestressing in offering sufficient safety factor against collapse. It is thus important to design a drainage network largely dimensioned. The warm liner appears at this stage of the project too highly stressed by fatigue at the singularity points (ducts between cavities, angles). A solution is proposed which limits the variations of thermal stresses by using a steel with low coefficient of thermal expansion. The cavity closures, which are numerous and some with large dimensions are an important aspect of the vessel safety. A solution of reinforced concrete shell with independent liner is proposed

  8. Magnetic field shielding system in a tokamak experimental power reactor (EPR): concept and calculations

    International Nuclear Information System (INIS)

    Peng, Y.K.M.; Marcus, F.B.; Dory, R.A.; Moore, J.R.

    1975-01-01

    A poloidal magnetic field shielding system is proposed for a tokamak EPR. This coil system minimizes the pulsed poloidal field that intersects the TF (toroidal field) coils and hence reduces the risk of superconductor quenching and structural failure of the coils. Based on an idealized shielding model, we have determined the configurations for the OH (ohmic heating), the S-VF (shield-vertical field), and the T-VF (trimming-vertical field) coils in a typical tokamak EPR. It is found that the pulsed poloidal field strength is greatly reduced in the TF coil region. The overall requirement in stored plasma and vertical field energy is also substantially reduced when compared with conventional EPR designs. Use of this field shielding system is expected to enhance reliability of the superconducting TF coils in a tokamak EPR

  9. The design of bonded reinforcement for thermal stresses in prestressed concrete reactor vessels

    International Nuclear Information System (INIS)

    Kotulla, B.; Hansson, V.

    1977-01-01

    This paper deals with examples of thermal loadings where instationary growth of tensile zones and redistribution of stresses by cracking are of importance. Temperatures produce, in addition to prestressing and internal pressure, the most important stresses in a prestressed concrete reactor pressure vessel. Characteristic of thermal stresses is that they are influenced to a large extent by creep of concrete and that they influence stress redistributions by temperature dependent creep data. Computations show that during the first instationary heating process of the vessel stresses are reduced by creep effects to about fifty percent of the values of the stationary elastic case at the hot face. With a following cooling, creep effects are generally much less, so this case may produce tensile stresses on the internal face of the wall which lead to cracking of the concrete. Tensile stresses first occur due to the instationary growth of the temperature field in a narrow zone near the liner. If outside this zone compressive stresses exist due to prestressing then crack spreading is limited and restraint by the parts of the wall under compression causes crack distribution even without reinforcement in this zone. Growth of cracks with the instationary spreading of tensile zones according to temperature development was calculated. These calculations take into account discrete cracks, reinforcement and different assumptions for tensile strength. Reinforcement of small diameter near the surface has the best influence on crack spacing. Calculations show that for the stationary state of cooling the forces in the reinforcement may be as low as twenty to thirty percent of the tensile force not taking into account cracking of the concrete

  10. FOCUS, Neutron Transport System for Complex Geometry Reactor Core and Shielding Problems by Monte-Carlo

    International Nuclear Information System (INIS)

    Hoogenboom, J.E.

    1980-01-01

    1 - Description of problem or function: FOCUS enables the calculation of any quantity related to neutron transport in reactor or shielding problems, but was especially designed to calculate differential quantities, such as point values at one or more of the space, energy, direction and time variables of quantities like neutron flux, detector response, reaction rate, etc. or averages of such quantities over a small volume of the phase space. Different types of problems can be treated: systems with a fixed neutron source which may be a mono-directional source located out- side the system, and Eigen function problems in which the neutron source distribution is given by the (unknown) fundamental mode Eigen function distribution. Using Monte Carlo methods complex 3- dimensional geometries and detailed cross section information can be treated. Cross section data are derived from ENDF/B, with anisotropic scattering and discrete or continuous inelastic scattering taken into account. Energy is treated as a continuous variable and time dependence may also be included. 2 - Method of solution: A transformed form of the adjoint Boltzmann equation in integral representation is solved for the space, energy, direction and time variables by Monte Carlo methods. Adjoint particles are defined with properties in some respects contrary to those of neutrons. Adjoint particle histories are constructed from which estimates are obtained of the desired quantity. Adjoint cross sections are defined with which the nuclide and reaction type are selected in a collision. The energy after a collision is selected from adjoint energy distributions calculated together with the adjoint cross sections in advance of the actual Monte Carlo calculation. For multiplying systems successive generations of adjoint particles are obtained which will die out for subcritical systems with a fixed neutron source and will be kept approximately stationary for Eigen function problems. Completely arbitrary problems can

  11. Neutronics model of the bulk shielding reactor (BSR): validation by comparison of calculations with the experimental measurements

    International Nuclear Information System (INIS)

    Johnson, J.O.; Miller, L.F.; Kam, F.B.K.

    1981-05-01

    A neutronics model for the Oak Ridge National Laboratory Bulk Shielding Reactor (ORNL-SAR) was developed and verified by experimental measurements. A cross-section library was generated from the 218 group Master Library using the AMPX Block Code system. A series of one-, two-, and three-dimensional neutronics calculations were performed utilizing both transport and diffusion theory. Spectral comparison was made with 58 Ni(n,p) reaction. The results of the comparison between the calculational model and other experimental measurements showed agreement within 10% and therefore the model was determined to be adequate for calculating the neutron fluence for future irradiation experiments in the ORNL-BSR

  12. Neutronics model of the bulk shielding reactor (BSR): validation by comparison of calculations with the experimental measurements

    Energy Technology Data Exchange (ETDEWEB)

    Johnson, J.O.; Miller, L.F.; Kam, F.B.K.

    1981-05-01

    A neutronics model for the Oak Ridge National Laboratory Bulk Shielding Reactor (ORNL-SAR) was developed and verified by experimental measurements. A cross-section library was generated from the 218 group Master Library using the AMPX Block Code system. A series of one-, two-, and three-dimensional neutronics calculations were performed utilizing both transport and diffusion theory. Spectral comparison was made with /sup 58/Ni(n,p) reaction. The results of the comparison between the calculational model and other experimental measurements showed agreement within 10% and therefore the model was determined to be adequate for calculating the neutron fluence for future irradiation experiments in the ORNL-BSR.

  13. Effect of high temperature or fire on heavy weight concrete properties

    International Nuclear Information System (INIS)

    Sakr, K.; EL-Hakim, E.

    2005-01-01

    Temperature plays an important role in the use of concrete for shielding nuclear reactors. In the present work, the effect of different durations (1, 2 and 3 h) of high temperatures (250, 500, 750 and 950 deg. C) on the physical, mechanical and radiation properties of heavy concrete was studied. The effect of fire fitting systems on concrete properties was investigated. Results showed that ilmenite concrete had the highest density, modulus of elasticity and lowest absorption percent, and it had also higher values of compressive, tensile, bending and bonding strengths than gravel or baryte concrete. Ilmenite concrete showed the highest attenuation of transmitted gamma rays. Firing (heating) exposure time was inversely proportional to mechanical properties of all types of concrete. Ilmenite concrete was more resistant to elevated temperature. Foam or air proved to be better than water as a cooling system in concrete structure exposed to high temperature because water leads to a big damage in concrete properties

  14. A 1055 ft/sec impact test of a two foot diameter model nuclear reactor containment system without fracture

    Science.gov (United States)

    Puthoff, R. L.

    1972-01-01

    A study to determine the feasibility of containing the fission products of a mobile reactor in the event of an impact is presented. The model simulated the reactor core, energy absorbing gamma shielding, neutron shielding and the containment vessel. It was impacted against an 18,000 pound reinforced concrete block at 1055 ft/sec. The model was significantly deformed and the concrete block demolished. No leaks were detected nor were any cracks observed in the model after impact.

  15. 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.

  16. Development and application of a material law for steel-fibre-reinforced concrete with regard to its use for pre-stressed concrete reactor vessels

    International Nuclear Information System (INIS)

    Schimmelpfennig, K.; Borgerhoff, M.

    1995-01-01

    On the basis of the evaluation of many publications on the mechanical behaviour of steel fibre reinforced concrete (SFRC) and on the results of experiments using an SFRC especially developed for pre-stressed concrete reactor vessels (PCRVs), a material law for SFRC including general multiaxial stress conditions has been developed. From fibre pull-out tests described in the literature and by use of the experimental results, relations describing the capable tensile stress in SFRC after cracking, as a function of crack width, have been derived. There is a significant increase in the biaxial compressive strength of SFRC compared with plain concrete. The improved behaviour under multiaxial stress conditions, with one of the principal stresses being tensile, is outlined in comparison with different formulations of failure envelopes of plain concrete. For the purpose of verifying the material law implemented in the computer program used, analyses have been carried out for experiments with SFRC beams. After some modification concerning the shear behaviour, load-displacement curves and realistic crack propagations which correspond well have been obtained. In the stand-tube area in the centre of a PCRV top cap the use of SFRC is advantageous because of the difficulties concerning the arrangement of reinforcement in the concrete between the tubes. (orig.)

  17. Development of a new measurement method for fast breeder reactor fuel burnup using a shielded ion microprobe analyzer

    International Nuclear Information System (INIS)

    Mizuno, M.; Enokido, Y.; Itaki, T.; Kono, K.; Unno, I.; Yamanouchi, S.

    1985-01-01

    A new method of burnup measurement using a shielded ion microprobe analyzer (SIMA) has been developed. The method is based on the isotope analysis of uranium, plutonium, and fission products in irradiated mixed oxide fuel by means of secondary ion mass spectrometry (SIMS). Fourteen samples irradiated in the Japanese experimental fast reactor JOYO were examined. The maximum local burnup of JOYO MK-I core fuels was about5.1 at. %. The axial burnup distribution of the fuel pin was in good agreement with that of the sibling pin in the same subassembly, measured by surface ionization mass spectrometry, which requires the chemical separation of fission products and heavy metals. The new method facilitates the rapid and accurate measurement of fast breeder reactor fuel burnup without human radiation exposure during sample preparation and analysis

  18. Shielding computations for solution transfer lines from Analytical Lab to process cells of Demonstration Fast Reactor Plant (DFRP)

    International Nuclear Information System (INIS)

    Baskar, S.; Jose, M.T.; Baskaran, R.; Venkatraman, B.

    2018-01-01

    The diluted virgin solutions (both aqueous and organic) and aqueous analytical waste generated from experimental analysis of process solutions, pertaining to Fast Breeder Test Reactor (FBTR) and Prototype Fast Breeder Reactor (PFBR), in glove boxes of active analytical Laboratory (AAL) are pumped back to the process cells through a pipe in pipe arrangement. There are 6 transfer lines (Length 15-32 m), 2 for each type of transfer. The transfer lines passes through the area inside the AAL and also the operating area. Hence it is required to compute the necessary radial shielding requirement around the lines to limit the dose rates in both the areas to the permissible values as per the regulatory requirement

  19. Accuracy evaluation of the current data and method applied to shielding design of the Fusion Experimental Reactor (FER)

    International Nuclear Information System (INIS)

    Mori, Seiji; Kobayashi, Takeshi; Seki, Yasushi

    1988-06-01

    Shielding benchmarking study of the current data and method applied to the Fusion Experimental Reactor (FER) was performed. First, neutron and gamma ray fluxes were calculated by the one-dimensional S N code using various cross section libraries and the continuous energy Monte Carlo code. The results were compared in terms of the S N /MC ratio. The worst ratios are about 0.5 and 0.25 for neutron flux and gamma ray flux, respectively. Next, the analytical calculations of the iron sphere transmission experiment of 14 MeV neutrons were performed to examine the accuracy of cross section data of iron, which is the most important material of shield. The E/C ratio is larger than 2 even if the continuous energy Monte Carlo code was used. Thirdly, the influence of geometrical representation of the shield was investigated by comparing the homogeneous model and the heterogeneous model (alternating layers of SS316 and water). As a result, it was made clear that the homogeneous model underestimates neutron flux by a factor of 2. Finally, the necessity of benchmark experiment and improvement of cross section library was pointed out as the further R and D issues. (author)

  20. Radiation shielding

    International Nuclear Information System (INIS)

    Yue, D.D.

    1979-01-01

    Details are given of a cylindrical electric penetration assembly for carrying instrumentation leads, used in monitoring the performance of a nuclear reactor, through the containment wall of the reactor. Effective yet economical shielding protection against both fast neutron and high-energy gamma radiation is provided. Adequate spacing within the assembly allows excessive heat to be efficiently dissipated and means of monitoring all potential radiation and gas leakage paths are provided. (UK)

  1. 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)

  2. Shielding and activation calculations around the reactor core for the MYRRHA ADS design

    Science.gov (United States)

    Ferrari, Anna; Mueller, Stefan; Konheiser, J.; Castelliti, D.; Sarotto, M.; Stankovskiy, A.

    2017-09-01

    In the frame of the FP7 European project MAXSIMA, an extensive simulation study has been done to assess the main shielding problems in view of the construction of the MYRRHA accelerator-driven system at SCK·CEN in Mol (Belgium). An innovative method based on the combined use of the two state-of-the-art Monte Carlo codes MCNPX and FLUKA has been used, with the goal to characterize complex, realistic neutron fields around the core barrel, to be used as source terms in detailed analyses of the radiation fields due to the system in operation, and of the coupled residual radiation. The main results of the shielding analysis are presented, as well as the construction of an activation database of all the key structural materials. The results evidenced a powerful way to analyse the shielding and activation problems, with direct and clear implications on the design solutions.

  3. The Effect of Various Waste Materials' Contents on the Attenuation Level of Anti-Radiation Shielding Concrete.

    Science.gov (United States)

    Azeez, Ali Basheer; Mohammed, Kahtan S; Abdullah, Mohd Mustafa Al Bakri; Hussin, Kamarudin; Sandu, Andrei Victor; Razak, Rafiza Abdul

    2013-10-23

    Samples of concrete contain various waste materials, such as iron particulates, steel balls of used ball bearings and slags from steel industry were assessed for their anti-radiation attenuation coefficient properties. The attenuation measurements were performed using gamma spectrometer of NaI (Tl) detector. The utilized radiation sources comprised 137 Cs and ⁶⁰Co radioactive elements with photon energies of 0.662 MeV for 137 Cs and two energy levels of 1.17 and 1.33 MeV for the ⁶⁰Co. Likewise the mean free paths for the tested samples were obtained. The aim of this work is to investigate the effect of the waste loading rates and the particulate dispersive manner within the concrete matrix on the attenuation coefficients. The maximum linear attenuation coefficient (μ) was attained for concrete incorporates iron filling wastes of 30 wt %. They were of 1.12 ± 1.31×10 -3 for 137 Cs and 0.92 ± 1.57 × 10 -3 for ⁶⁰Co. Substantial improvement in attenuation performance by 20%-25% was achieved for concrete samples incorporate iron fillings as opposed to that of steel ball samples at different (5%-30%) loading rates. The steel balls and the steel slags gave much inferior values. The microstructure, concrete-metal composite density, the homogeneity and particulate dispersion were examined and evaluated using different metallographic, microscopic and measurement facilities.

  4. The Effect of Various Waste Materials’ Contents on the Attenuation Level of Anti-Radiation Shielding Concrete

    Science.gov (United States)

    Azeez, Ali Basheer; Mohammed, Kahtan S.; Abdullah, Mohd Mustafa Al Bakri; Hussin, Kamarudin; Sandu, Andrei Victor; Razak, Rafiza Abdul

    2013-01-01

    Samples of concrete contain various waste materials, such as iron particulates, steel balls of used ball bearings and slags from steel industry were assessed for their anti-radiation attenuation coefficient properties. The attenuation measurements were performed using gamma spectrometer of NaI (Tl) detector. The utilized radiation sources comprised 137Cs and 60Co radioactive elements with photon energies of 0.662 MeV for 137Cs and two energy levels of 1.17 and 1.33 MeV for the 60Co. Likewise the mean free paths for the tested samples were obtained. The aim of this work is to investigate the effect of the waste loading rates and the particulate dispersive manner within the concrete matrix on the attenuation coefficients. The maximum linear attenuation coefficient (μ) was attained for concrete incorporates iron filling wastes of 30 wt %. They were of 1.12 ± 1.31×10−3 for 137Cs and 0.92 ± 1.57 × 10−3 for 60Co. Substantial improvement in attenuation performance by 20%–25% was achieved for concrete samples incorporate iron fillings as opposed to that of steel ball samples at different (5%–30%) loading rates. The steel balls and the steel slags gave much inferior values. The microstructure, concrete-metal composite density, the homogeneity and particulate dispersion were examined and evaluated using different metallographic, microscopic and measurement facilities. PMID:28788363

  5. Monte Carlo analysis of the effects of a blanket-shield penetration on the performance of a tokamak fusion reactor

    International Nuclear Information System (INIS)

    Santoro, R.T.; Tang, J.S.; Alsmiller, R.G. Jr.; Barnes, J.M.

    1977-05-01

    Adjoint Monte Carlo calculations have been carried out using the three-dimensional radiation transport code, MORSE, to estimate the nuclear heating and radiation damage in the toroidal field (TF) coils adjacent to a 28 x 68 cm 2 rectangular neutral beam injector duct that passes through the blanket and shield of a D-T burning Tokamak reactor. The plasma region, blanket, shield, and TF coils were represented in cylindrical geometry using the same dimensions and compositions as those of the Experimental Power Reactor. The radiation transport was accomplished using coupled 35-group neutron, 21-group gamma-ray cross sections obtained by collapsing the DLC-37 cross-section library. Nuclear heating and radiation damage rates were estimated using the latest available nuclear response functions. The presence of the neutral beam injector duct leads to increases in the nuclear heating rates in the TF coils ranging from a factor of 3 to a factor of 196 depending on the location. Increases in the radiation damage also result in the TF coils. The atomic displacement rates increase from factors of 2 to 138 and the hydrogen and helium gas production rates increase from factors of 11 to 7600 and from 15 to 9700, respectively

  6. Application of micro blasting technique to demolition of biological shield wall of reactor building

    International Nuclear Information System (INIS)

    Kontani, Osamu

    2016-01-01

    Although heavily reinforced concrete structures in nuclear facilities could be dismantled effectively with the controlled blasting technique, the noises and vibrations caused by blasting were matters of concern. Recently, in the building replacement at urban areas, there are increasing cases of demolition of large reinforced concrete members below ground. Instead of applying heavy weight breakers, the micro blasting technique that enables to lower noises and vibrations has developed and applied to demolition work of large reinforced concrete members in urban areas. In this report, the features of the micro blasting technique is reviewed by comparing with existing controlled blasting technique and its applicability to demolition work of nuclear facilities is investigated. The results of those study find that it is confirmed that the micro blasting technique could be applicable to large reinforced concrete structures in nuclear facilities because of the low levels of noises and vibrations. However, it is recommended to perform mock-up tests to confirm the demolition efficiency and levels of the noises and vibrations since the rebar used in nuclear facilities is larger in diameters and in rebar ratio compared with ordinary reinforced concrete structures. (author)

  7. Analysis of core-concrete interaction event with flooding for the Advanced Neutron Source reactor

    International Nuclear Information System (INIS)

    Kim, S.H.; Taleyarkhan, R.P.; Georgevich, V.; Navarro-Valenti, S.

    1993-01-01

    This paper discusses salient aspects of the methodology, assumptions, and modeling of various features related to estimation of source terms from an accident involving a molten core-concrete interaction event (with and without flooding) in the Advanced Neutron Source (ANS) reactor at the Oak Ridge National Laboratory. Various containment configurations are considered for this postulated severe accident. Several design features (such as rupture disks) are examined to study containment response during this severe accident. Also, thermal-hydraulic response of the containment and radionuclide transport and retention in the containment are studied. The results are described as transient variations of source terms, which are then used for studying off-site radiological consequences and health effects for the support of the Conceptual Safety Analysis Report for ANS. The results are also to be used to examine the effectiveness of subpile room flooding during this type of severe accident

  8. Experience in surveillance of the prestress of concrete reactor vessels in Wylfa nuclear power station

    International Nuclear Information System (INIS)

    Dawson, P.; Paton, A.A.; Walsh, S.R.

    1989-01-01

    This paper describes experience gained in the in-service surveillance of the prestressing system for the prestressed concrete reactor vessels (PCRVs) at Wylfa nuclear power station. The paper gives details of results for the prestressing system obtained from the statutory in-service inspection program of the PCRVs. The program includes a detailed examination of a selection of prestressing tendon anchorages, anchorage load checks using a lift-off technique on a one percent sample of tendons and corrosion inspection of samples of prestressing strand and determination of their mechanical properties. The results obtained from the above in-service inspections have shown that the prestressing system continues to function within its design limits

  9. Summary of ORNL high-temperature gas-cooled reactor program

    International Nuclear Information System (INIS)

    Kasten, P.R.

    1981-01-01

    Oak Ridge National Laboratory (ORNL) efforts on the High-Temperature Gas-Cooled Reactor (HTGR) Program have been on HTGR fuel development, fission product and coolant chemistry, prestressed concrete reactor vessel (PCRV) studies, materials studies, graphite development, reactor physics and shielding studies, application assessments and evaluations and selected component testing

  10. Analyses of iron and concrete shielding experiments at JAEA/TIARA with JENDL/HE-2007, ENDF/B-VII.1 and FENDL-3.0

    International Nuclear Information System (INIS)

    Konno, Chikara; Ochiai, Kentaro; Sato, Satoshi; Ohta, Masayuki

    2015-01-01

    IAEA released a new Fusion Evaluated Nuclear Data Library, FENDL-3.0, in 2012. FENDL-3.0 extends the neutron energy range from 20 MeV to greater than 60 MeV. Now there is increasing interest in nuclear data above 20 MeV. Thus we have analyzed the iron and concrete shielding experiments with the 40 and 65 MeV neutron sources at TIARA in Japan Atomic Energy Agency with the latest high-energy nuclear data libraries, JENDL/HE-2007, ENDF/B-VII.1 and FENDL-3.0. The Monte Carlo code MCNP-5 and ACE files of JENDL/HE-2007, ENDF/B-VII.1 and FENDL-3.0, which are supplied from JAEA, LANL and IAEA, respectively, were used for this analysis. The collimated neutron beam and test shields were modeled in the analysis. The measured source neutron data were adopted in the analysis. The followings are found out from the results; (1) Iron experiments: The calculation result with FENDL-3.0 agrees with the measured one best. That with JENDL/HE-2007 fairly agrees with the measured one. On the contrary that with ENDF/B-VII.1 drastically overestimates the measured one. It is confirmed that this overestimation is due to the smaller non-elastic scattering data of "5"6Fe in ENDF/B-VII.1. (2) Concrete experiments: The calculation result with ENDL/HE-2007 agrees with the measured one best, while those with FENDL-3.0 and ENDF/B-VII.1 drastically overestimate the measured one. It is confirmed that this overestimation is due to both the larger elastic and smaller non-elastic scattering data of "1"6O in FENDL-3.0 and ENDF/B-VII.1.

  11. The Results of the CCI-3 Reactor Material Experiment Investigating 2-D Core-Concrete Interaction and Debris Coolability with a Siliceous Concrete Crucible

    International Nuclear Information System (INIS)

    Farmer, M.T.; Lomperski, S.; Basu, S.

    2006-01-01

    The OECD-sponsored Melt Coolability and Concrete Interaction (MCCI) program conducted reactor materials experiments and associated analysis to achieve the following two objectives: 1) resolve the ex-vessel debris coolability issue, and 2) address remaining uncertainties related to long-term two-dimensional molten core-concrete interactions under both wet and dry cavity conditions. Achievement of these two objectives will demonstrate the efficacy of severe accident management guidelines for existing plants, and provide the technical basis for better containment designs of future plants. With respect to the second objective, there are remaining uncertainties in the models that evaluate the lateral vs. axial power split during core-concrete interaction because of a lack of truly two-dimensional experiment data. As a result, there are differences in the 2-D cavity erosion profiles predicted by codes such as WECHSL, COSACO, TOLBIAC, MEDICIS, and MELCOR. In the continuing effort to bridge this data gap, the third in a series of large scale Core-Concrete Interaction experiments (CCI-3) has been conducted as part of the MCCI program. This test investigated the long-term interaction of a 375 kg core-oxide melt within a two-dimensional siliceous concrete crucible. The initial phase of the test was conducted under dry conditions. After a predetermined time interval, the cavity was flooded with water to obtain data on the coolability of a core melt after core-concrete interaction has progressed for some time. This paper provides a description of the facility and an overview of results from this test. (authors)

  12. Dosimetry and radiation shielding at the RA reactor, Annual report 1975, Annex 5

    International Nuclear Information System (INIS)

    Ninkovic, M.

    1976-01-01

    In the working environment at the RA reactor, the level of gamma radiation is measured continuously by the built-in stationary system. According to the needs, measurement are done in the reactor hall every day. The level of gamma radiation is measured separately in typical points when the reactor is operated at nominal power and during intervals between two operating campaigns. The level of neutron radiation is measured according to the needs by means of a mobile spherical neutron detector. These measurements are done in the reactor hall around the horizontal experimental channels. Measured values of neutron radiation are three times lower than the relevant levels of gamma radiation [sr

  13. Streaming experiment of gamma-ray obliquely incident on concrete shield wall with straight cylindrical ducts and verification of single scattering code

    International Nuclear Information System (INIS)

    Yamaji, Akio; Saito, Tetsuo.

    1988-01-01

    To investigate a proximity effect of ducts on shield performance against γ radiation, an experiment was performed at JRR-4 by entering the γ-ray beam into a concrete shield wall of 100 cm-thickness with 3 or 5 straight cylindrical ducts of radius of 4.45 cm placed in a straight line or crosswise at interval of 8.9 cm. The dose rates were measured using digital dosimeters on a horizontal line 20 cm apart from the rear of the wall with 0, 1, 3 and 5 ducts, and with the incident angles of 0deg, 7deg, 14deg and 20deg, respectively. The dose rate distributions depended on the number of ducts and the incident angle, and the dose rate ratios of with-three-ducts to no-duct distributed within 3.6∼12, 1.3∼5.0 and 1.1∼4.3, for the incident angles of 7deg, 14deg and 20deg, while those of with-single-duct to no-duct within 1.2∼7.1, 1.1∼2.7 and 1.0∼1.9, respectively. The experiment was analyzed using a multigroup single scattering code G33YSN able to deal with the geometry of the ducts exactly. For each incident angle, the calculation agreed with the experiment within a factor of 2. (author)

  14. Thermohydraulics in a high-temperature gas-cooled reactor prestressed-concrete reactor vessel during unrestricted core-heatup accidents

    International Nuclear Information System (INIS)

    Kroeger, P.G.; Colman, J.; Araj, K.

    1983-01-01

    The hypothetical accident considered for siting considerations in High Temperature Gas-Cooled Reactors (HTGR) is the so called Unrestricted Core Heatup Accident (UCHA), in which all forced circulation is lost at initiation, and none of the auxillary cooling loops can be started. The result is a gradual slow core heatup, extending over days. Whether the liner cooling system (LCS) operates during this time is of crucial importance. If it does not, the resulting concrete decomposition of the prestressed concrete reactor vessel (PCRV) will ultimately cause containment building (CB) failure after about 6 to 10 days. The primary objective of the work described here was to establish for such accident conditions the core temperatures and approximate fuel failure rates, to check for potential thermal barrier failures, and to follow the PCRV concrete temperatures, as well as PCRV gas releases from concrete decomposition. The work was done for the General Atomic Corporation Base Line Zero reactor of 2240 MW(t). Most results apply at least qualitatively also to other large HTGR steam cycle designs

  15. Nuclear reactor

    International Nuclear Information System (INIS)

    Jungmann, A.

    1975-01-01

    Between a PWR's reactor pressure vessel made of steel and the biological shield made of concrete there is a gap. This gap is filled up with a heat insulation facting the reactor pressure vessel, for example with insulating concrete segments jacketed with sheet steel and with an additional layer. This layer serves for smooth absorption of compressive forces originating in radial direction from the reactor pressure vessel. It consists of cylinder-segment shaped bricks made of on situ concrete, for instance. The bricks have cooling agent ports in one or several rows which run parallel to the wall of the pressure vessel and in alignment with superposed bricks. Between the layer of bricks and the biological shield or rather the heat insulation, there are joints which are filled, however, with injected mortar. That guarantees a smooth series of connected components resistant tom compression. Besides, a slip foil can be set between the heat insulation and the joining joint filled with mortar for the reduction of the friction at thermal expansions. (TK) [de

  16. Problem Oriented Neutron-Gamma Cross Sections Libraries for WWER-440 and WWER-1000 Shielding and Reactor Vessel Dosimetry Application

    International Nuclear Information System (INIS)

    Belousov, S.; Antonov, S.; Ilieva, K.

    1997-01-01

    The 47 neutron and 20 gamma group libraries BGL-440 and BGL-1000 for the shielding and reactor vessel dosimetry application have been generated for WWER-440 and WWER-1000 by collapsing the VITAMIN-B6 library (199 neutron and 42 gamma groups on the base of ENDF/B-6). The first parts of the libraries for neutron-gamma transport calculation, BGL-440-1 (150 nuclides) and BGL-1000-1 (140 nuclides), have been generated by a modified version of SAS1X control module of the SCALE system. The appropriate zone-average neutron flux had been used for these sub-libraries collapsing. The BGL-440-2 and BGL-1000-2 sub-libraries consist of cross sections for all 120 nuclides of VITAMIN-B6, for calculation of the transport through non-reactor materials of dosimeters, capsules, specimens which may be placed in the cavity behind the reactor vessel. The neutron spectrum just beyond the RPV had been used for this collapsing. As the first test the comparative calculations of the neutron flux on/behind the WWER-1000 reactor vessel have been realised using the libraries BGL-1000 and BUGLE, intended for the American PWR reactors. The integral neutron flux values by BGL-1000 and BUGLE differ by 3% onto the vessel, and 5% behind the vessel. This result shows that the calculations of the neutron flux responses for the WWER vessel surveillance, especially in locations behind the WWER vessel have to be done by the appropriate BGL library. Key words: neutron transport, multigroup neutron cross section libraries

  17. Status of radiation shield design for liquid metal fast breeder reactor spent fuel shipping cask application

    International Nuclear Information System (INIS)

    Dupree, S.A.; Rack, H.J.

    1976-09-01

    Neutron and gamma-ray transport calculations in one-dimensional cylindrical geometry have been performed on a trial reference LMFBR spent-fuel shipping cask that could transport one CRBR subassembly. In the study it was assumed that a layer of depleted U and a layer of neutron shielding materials were sandwiched between 5.08-cm-thick (2-in.) layers of stainless steel. The thicknesses of the internal layers were adjusted until a balanced dose rate (50 percent neuton and 50 percent gamma-ray) of 5 mrem/hr was achieved at a point 1.83 m (6 ft) from the cask surface. Neutron-shield materials considered were LiH, Be, B 4 C, DiH 2 . 5 , and C (graphite). Of these materials, LiH provided the smallest, lightest, and least expensive cask; however, its use would be contigent on expansion of production facilities for LiH and development of a canning or cladding procedure. The B 4 C shielded cask would offer the best alternative if the designs were limited to those using currently available materials

  18. Methods for calculating radiation attenuation in shields

    Energy Technology Data Exchange (ETDEWEB)

    Butler, J; Bueneman, D; Etemad, A; Lafore, P; Moncassoli, A M; Penkuhn, H; Shindo, M; Stoces, B

    1964-10-01

    In recent years the development of high-speed digital computers of large capacity has revolutionized the field of reactor shield design. For compact special-purpose reactor shields, Monte-Carlo codes in two- and three dimensional geometries are now available for the proper treatment of both the neutron and gamma- ray problems. Furthermore, techniques are being developed for the theoretical optimization of minimum-weight shield configurations for this type of reactor system. In the design of land-based power reactors, on the other hand, there is a strong incentive to reduce the capital cost of the plant, and economic considerations are also relevant to reactors designed for merchant ship propulsion. In this context simple methods are needed which are economic in their data input and computing time requirements and which, at the same time, are sufficiently accurate for design work. In general the computing time required for Monte-Carlo calculations in complex geometry is excessive for routine design calculations and the capacity of the present codes is inadequate for the proper treatment of large reactor shield systems in three dimensions. In these circumstances a wide range of simpler techniques are currently being employed for design calculations. The methods of calculation for neutrons in reactor shields fall naturally into four categories: Multigroup diffusion theory; Multigroup diffusion with removal sources; Transport codes; and Monte Carlo methods. In spite of the numerous Monte- Carlo techniques which are available for penetration and back scattering, serious problems are still encountered in practice with the scattering of gamma rays from walls of buildings which contain critical facilities and also concrete-lined discharge shafts containing irradiated fuel elements. The considerable volume of data in the unclassified literature on the solution of problems of this type in civil defence work appears not to have been evaluated for reactor shield design. In

  19. 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)

  20. 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)

  1. Structural evaluation of the Shippingport Reactor Pressure Vessel and Neutron Shield Tank package for impact and puncture loads

    International Nuclear Information System (INIS)

    Fischer, L.E.; Chou, C.K.; Lo, T.; Schwartz, M.W.

    1988-06-01

    A structural evaluation of Shippingport Reactor Pressure Vessel and Neutron Shield Tank package for impact and puncture loads under the normal and hypothetical accident conditions of 10 CFR 71 was performed. Component performance criteria for the Shippingport package and the corresponding structural acceptance criteria for these components were developed based on a review of the package geometry, the planned transport environment, and the external radiation standards and dispersal limits of 10 CFR 71. The evaluation was performed using structural analysis methods. A demonstration combining simplified model tests and nonlinear finite element analyses was made to substantiate the structural analysis methods used to evaluate the Shippingport package. The package was analyzed and the results indicate that the package meets external radiation standards and release limits of 10 CFR 71. 13 refs., 50 figs., 19 tabs

  2. Self-consistent Analysis of a Blanket and Shielding of a Fusion Reactor Concept

    International Nuclear Information System (INIS)

    Kim, Suk Kwon; Hong, B. G.; Lee, D. W.; Kim, D. H.; Lee, Y. O.

    2008-01-01

    To develop the concept of a DEMO reactor, a tokamak reactor system analysis code has been developed at KAERI (Korea Atomic Energy Research Institute). The system analysis code incorporates prospects of the development of plasma physics and the technologies in a simple mathematical model and it helps to develop the concept of a fusion reactor and to identify the necessary R and D areas for a realization of the concept. In the system code, a plant power balance equation and a plasma power balance equation are solved to find plant parameters which satisfy the plasma physics and technology constraints, simultaneously. The outcome of the system analysis is to identify which areas of plasma physics and technologies and to what extent they should be developed for a realization of given fusion reactor concepts

  3. 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)

  4. Criticality and shielding calculations for containers in dry of spent fuel of TRIGA Mark III reactor of ININ

    International Nuclear Information System (INIS)

    Barranco R, F.

    2015-01-01

    In this thesis criticality and shielding calculations to evaluate the design of a container of dry storage of spent nuclear fuel generated in research reactors were made. The design of such container was originally proposed by Argentina and Brazil, and the Instituto Nacional de Investigaciones Nucleares (ININ) of Mexico. Additionally, it is proposed to modify the design of this container to store spent fuel 120 that are currently in the pool of TRIGA Mark III reactor, the Nuclear Center of Mexico and calculations and analyzes are made to verify that the settlement of these fuel elements is subcritical limits and dose rates to workers and the general public are not exceeded. These calculations are part of the design criteria for security protection systems in dry storage system (Dss for its acronym in English) proposed by the Nuclear Regulatory Commission (NRC) of the United States. To carry out these calculations simulation codes of Monte Carlo particle transport as MCNPX and MCNP5 were used. The initial design (design 1) 78 intended to store spent fuel with a maximum of 115. The ININ has 120 fuel elements and spent 3 control rods (currently stored in the reactor pool). This leads to the construction of two containers of the original design, but for economic reasons was decided to modify (design 2) to store in a single container. Criticality calculations are performed to 78, 115 and fresh fuel elements 124 within the container, to the two arrangements described in Chapter 4, modeling the three-dimensional geometry assuming normal operating conditions and accident. These calculations are focused to demonstrate that the container will remain subcritical, that is, that the effective multiplication factor is less than 1, in particular not greater than 0.95 (as per specified by the NRC). Spent fuel 78 and 124 within the container, both gamma radiation to neutron shielding calculations for only two cases were simulated. First actinides and fission products generated

  5. 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

  6. Integral data testing of JENDL-3.2 for fusion reactor and shielding applications

    International Nuclear Information System (INIS)

    Oyama, Yukio

    1995-01-01

    Integral data testing of JENDL-3.2 is being performed in the activities of two working groups of the Japanese Nuclear Data Committee. The continuous and group-wise libraries prepared from JENDL-3.2 are planned to be tested by the working groups. In this paper, the continuous library FSXLIB-J3R2 processed from JENDL-3.2 for MCNP was tested for fission and fusion neutrons using data of integral experiments and compared to the results of JENDL-3.1. The results of integral data testing of JENDL-3.2 for fusion and shielding application are reviewed. (author)

  7. 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.

  8. Shielding calculations for the TFTR neutral beam injectors

    International Nuclear Information System (INIS)

    Santoro, R.T.; Lillie, R.A.; Alsmiller, R.G. Jr.; Barnes, J.M.

    1979-07-01

    Two-dimensional discrete ordinates calculations have been performed to determine the location and thickness of concrete shielding around the Tokamak Fusion Test Reactor (TFTR) neutral beam injectors. Two sets of calculations were performed: one to determine the dose equivalent rate on the roof and walls of the test cell building when no injectors are present, and one to determine the contribution to the dose equivalent rate at these locations from radiation streaming through the injection duct. Shielding the side and rear of the neutral beam injector with 0.305 and 0.61 m of concrete, respectively, and lining the inside of the test cell wall with an additional layer of concrete having a thickness of 0.305 m and a height above the axis of deuteron injection of 3.10 m are sufficient to maintain the biological dose equivalent rate outside the test cell to approx. 1 mrem/DT pulse

  9. An international survey of in-service inspection experience with prestressed concrete pressure vessels and containments for nuclear reactors

    International Nuclear Information System (INIS)

    1982-04-01

    An international survey is presented of experience obtained from the in-service surveillance of prestressed concrete pressure vessels and containments for nuclear reactors. Some information on other prestressed concrete structures is also given. Experience has been gained during the working life of such structures in Western Europe and the USA over the years since 1967. For each country a summary is given of the nuclear programme, national standards and Codes of Practice, and the detailed in-service inspection programme. Reports are then given of the actual experience obtained from the inspection programme and the methods of measurement, examination and reporting employed in each country. A comprehensive bibliography of over 100 references is included. The appendices contain information on nuclear power stations which are operating, under construction or planned worldwide and which employ either prestressed concrete pressure vessels or containments. (U.K.)

  10. New dose-mortality data based on 3-D radiation shielding calculation for concrete buildings at Nagasaki

    International Nuclear Information System (INIS)

    Rhoades, W.A.; Childs, R.L.; Ingersoll, D.T.

    1988-01-01

    The analysis of radiation doses received during the World War II attack on Nagasaki provides an important source of biochemical information. More than 40 years after the war, it has been possible to make a satisfactory calculation of the doses to personnel inside reinforced concrete buildings by use of a 3-dimensional discrete ordinates code, TORT. The results were used to deduce a new value of the LD50 parameter that is in good agreement with traditional values. The new discrete ordinates software appears to have potential application to conventional radiation transport calculations as well. 9 refs., 3 figs., 2 tabs

  11. Evaluation of alternative methods of simulating asymmetric bulk heating in fusion reactor blanket/shield components

    International Nuclear Information System (INIS)

    Deis, G.A.; Longhurst, G.R.; Miller, L.G.; Wadkins, R.P.; Wessol, D.E.

    1981-10-01

    As a part of Phase O, Test Program Element-II of the Office of Fusion Energy First Wall/Blanket/Shield Engineering Test Program, a study was conducted by EG and G Idaho, Inc., to identify, characterize, and recommend alternative approaches for simulating fusion bulk heating in blanket/shield components. This is the report on that effort. Since the usefulness of any simulation approach depends upon the particular experiment considered, classes of problem types (thermal-hydraulic, thermomechanical, etc.) and material types (structure, solid breeder, etc.) are developed. The evaluation of the various simulation approaches is performed for the various significant combinations of problem class and material class. The simulation approaches considered are discrete-source heating, direct resistance, electromagnetic induction, microwave heating, and nuclear heating. From the evaluations performed for each experiment type, discrete - source heating emerges as a good approach for bulk heating simulation in thermal - hydraulics experiments, and nuclear heating appears to be a good approach in experiments addressing thermomechanics and combined thermal-hydraulic/thermomechanics

  12. Radiation physics and shielding codes and analyses applied to design-assist and safety analyses of CANDUR and ACRTM reactors

    International Nuclear Information System (INIS)

    Aydogdu, K.; Boss, C. R.

    2006-01-01

    This paper discusses the radiation physics and shielding codes and analyses applied in the design of CANDU and ACR reactors. The focus is on the types of analyses undertaken rather than the inputs supplied to the engineering disciplines. Nevertheless, the discussion does show how these analyses contribute to the engineering design. Analyses in radiation physics and shielding can be categorized as either design-assist or safety and licensing (accident) analyses. Many of the analyses undertaken are designated 'design-assist' where the analyses are used to generate recommendations that directly influence plant design. These recommendations are directed at mitigating or reducing the radiation hazard of the nuclear power plant with engineered systems and components. Thus the analyses serve a primary safety function by ensuring the plant can be operated with acceptable radiation hazards to the workers and public. In addition to this role of design assist, radiation physics and shielding codes are also deployed in safety and licensing assessments of the consequences of radioactive releases of gaseous and liquid effluents during normal operation and gaseous effluents following accidents. In the latter category, the final consequences of accident sequences, expressed in terms of radiation dose to members of the public, and inputs to accident analysis, e.g., decay heat in fuel following a loss-of-coolant accident, are also calculated. Another role of the analyses is to demonstrate that the design of the plant satisfies the principle of ALARA (as low as reasonably achievable) radiation doses. This principle is applied throughout the design process to minimize worker and public doses. The principle of ALARA is an inherent part of all design-assist recommendations and safety and licensing assessments. The main focus of an ALARA exercise at the design stage is to minimize the radiation hazards at the source. This exploits material selection and impurity specifications and relies

  13. 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.

  14. Monte carlo calculation of the neutron effective dose rate at the outer surface of the biological shield of HTR-10 reactor

    International Nuclear Information System (INIS)

    Remetti, Romolo; Andreoli, Giulio; Keshishian, Silvina

    2012-01-01

    Highlights: ► We deal with HTR-10, that is a helium-cooled graphite-moderated pebble bed reactor. ► We carried out Monte Carlo simulation of the core by MCNP5. ► Extensive use of MCNP5 variance reduction methods has been done. ► We calculated the trend of neutron flux within the biological shield. ► We calculated neutron effective dose at the outer surface of biological shield. - Abstract: Research on experimental reactors, such as HTR-10, provide useful data about potentialities of very high temperature gas-cooled reactors (VHTR). The latter is today rated as one of the six nuclear reactor types involved in the Generation-IV International Forum (GIF) Initiative. In this study, the MCNP5 code has been employed to evaluate the neutron radiation trend vs. the biological shield's thickness and to calculate the neutron effective dose rate at the outer surface. The reactor's geometry has been completely modeled by means of lattices and universes provided by MCNP, even though some approximations were required. Monte Carlo calculations have been performed by means of a simple PC and, as a consequence, in order to obtain acceptable run times, it was made an extensive recourse to variance reduction methods.

  15. Study and installation of concrete shielding in the civil engineering of nuclear construction (1960); Etude et mise en place des betons de protection dans le genie civil des ouvrages nucleaires (1960)

    Energy Technology Data Exchange (ETDEWEB)

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

    1960-07-01

    The object of this report is to give technical information about high density concretes which have become very important for radiation biological shielding. The most generally used heavy aggregates (barytes, ilmenite, ferrophosphorus, limonite, magnetite and iron punching) to make these concretes are investigated from the point of view prospecting and physical and chemical characteristics. At first, a general survey of shielding concretes is made involving the study of components, mixing and placing methods, then, a detailed investigation of some high density concretes: barytes concrete, with incorporation of iron punching or iron shot, ferrophosphorus concrete, ilmenite concrete and magnetite concrete, more particularly with regard to grading and mix proportions and testing process. To put this survey in concrete form, two practical designs are described such as they have been carried out at the Saclay Nuclear Station. Specifications are given for diverse concretes and for making the proton-synchrotron 'Saturne' shielding blocks. (author) [French] Ce rapport a pour objet de donner des precisions techniques au sujet des betons a haute densite qui ont pris une grande importance pour la protection biologique contre les rayonnements. Les agregats lourds les plus couramment utilises (barytine, ilmenite, ferrophosphore, limonite, magnetite et riblons) pour la fabrication de ces betons, sont examines du point de vue prospection et caracteristiques physiques et chimiques. On procede d'abord a une etude generale des betons de protection comprenant l'etude des constituants, de la confection et de la mise en place, ensuite, a un examen detaille de quelques betons a haute densite: betons a base de barytine, avec incorporation de riblons ou de grenaille de fonte, betons au ferrophosphore, a base d'ilmenite ou de magnetite, notamment en ce qui concerne la granulometrie, la composition, le dosage et les processus d'essais. Pour concretiser ces etudes, deux applications pratiques

  16. Study and installation of concrete shielding in the civil engineering of nuclear construction (1960); Etude et mise en place des betons de protection dans le genie civil des ouvrages nucleaires (1960)

    Energy Technology Data Exchange (ETDEWEB)

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

    1960-07-01

    The object of this report is to give technical information about high density concretes which have become very important for radiation biological shielding. The most generally used heavy aggregates (barytes, ilmenite, ferrophosphorus, limonite, magnetite and iron punching) to make these concretes are investigated from the point of view prospecting and physical and chemical characteristics. At first, a general survey of shielding concretes is made involving the study of components, mixing and placing methods, then, a detailed investigation of some high density concretes: barytes concrete, with incorporation of iron punching or iron shot, ferrophosphorus concrete, ilmenite concrete and magnetite concrete, more particularly with regard to grading and mix proportions and testing process. To put this survey in concrete form, two practical designs are described such as they have been carried out at the Saclay Nuclear Station. Specifications are given for diverse concretes and for making the proton-synchrotron 'Saturne' shielding blocks. (author) [French] Ce rapport a pour objet de donner des precisions techniques au sujet des betons a haute densite qui ont pris une grande importance pour la protection biologique contre les rayonnements. Les agregats lourds les plus couramment utilises (barytine, ilmenite, ferrophosphore, limonite, magnetite et riblons) pour la fabrication de ces betons, sont examines du point de vue prospection et caracteristiques physiques et chimiques. On procede d'abord a une etude generale des betons de protection comprenant l'etude des constituants, de la confection et de la mise en place, ensuite, a un examen detaille de quelques betons a haute densite: betons a base de barytine, avec incorporation de riblons ou de grenaille de fonte, betons au ferrophosphore, a base d'ilmenite ou de magnetite, notamment en ce qui concerne la granulometrie, la composition, le dosage et les processus d'essais. Pour concretiser ces

  17. 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

  18. The SGR Multipurpose - Generation IV - Transportable Cogeneration Nuclear Reactor with Innovative Shielding

    International Nuclear Information System (INIS)

    Pahladsingh, R.R.

    2002-01-01

    Deregulation and liberalization are changing the global energy-markets. At the same time innovative technologies are introduced in the electricity industry; often as a requirement from the upcoming Digital Society. Energy solutions for the future are more seen as a mix of energy-sources for generation-, transmission- and distribution energy-services. The Internet Energy-web based 'Virtual' enterprises are coming up and will gradually change our society. It the fast changing world we have to realize that there will be less time to look for the adequate solutions to anticipate on global developments and the way they will influence our own societies. Global population may reach 9 billion people by 2030; this will put tremendous pressure on energy-, water- and food supply in the global economy. It is time to think about some major issues as described below and come up with the right answers. These are needed on very short term to secure a humane global economic growth and the sustainable global environment. The DOE (Department of Energy - USA) has started the Generation IV initiative for the new generation of nuclear reactors that must lead to much better safety, economics and public acceptance the new reactors. The SGR (Simplified Gas-cooled Reactor) is being proposed as a Generation IV modular nuclear reactor, using graphite pebbles as fuel, whereby an attempt has been made to meet all the DOE requirements, to be used for future nuclear reactors. The focus in this paper is on the changing and emerging global energy-markets and shows some relevant criteria to the nuclear industry and how we can anticipate with improved and new designs towards the coming Digital Society. (author)

  19. Efficient hydrogen production using heat in neutron shield of fusion reactor

    International Nuclear Information System (INIS)

    Okano, Kunihiko; Asaoka, Yoshiyuki; Hiwatari, Ryouji; Yoshida, Tomoaki

    2001-01-01

    In future perspective of energy supply, a hydrogen energy cycle is expected to play an important role as a CO 2 free fuel for mobile or co-generation systems. Fusion power plants should offer advantages, compatibilities and/or synergistic effects with or in such future energy systems. In this paper, a comprehensive power station, in which a fusion plant is integrated with a hydrogen production plant, is proposed. A tenuous heat source in the outboard shield, which is unsuitable to produce high-pressure and high-temperature steam for efficient electric power generation, is used for the hydrogen production. This integrated system provides some synergistic effects and it would be advantageous over any independent use of each plant. (author)

  20. Performance of advanced self-shielding models in DRAGON Version4 on analysis of a high conversion light water reactor lattice

    International Nuclear Information System (INIS)

    Karthikeyan, Ramamoorthy; Hebert, Alain

    2008-01-01

    A high conversion light water reactor lattice has been analysed using the code DRAGON Version4. This analysis was performed to test the performance of the advanced self-shielding models incorporated in DRAGON Version4. The self-shielding models are broadly classified into two groups - 'equivalence in dilution' and 'subgroup approach'. Under the 'equivalence in dilution' approach we have analysed the generalized Stamm'ler model with and without Nordheim model and Riemann integration. These models have been analysed also using the Livolant-Jeanpierre normalization. Under the 'subgroup approach', we have analysed Statistical self-shielding model based on physical probability tables and Ribon extended self-shielding model based on mathematical probability tables. This analysis will help in understanding the performance of advanced self-shielding models for a lattice that is tight and has a large fraction of fissions happening in the resonance region. The nuclear data for the analysis was generated in-house. NJOY99.90 was used for generating libraries in DRAGLIB format for analysis using DRAGON and A Compact ENDF libraries for analysis using MCNP5. The evaluated datafiles were chosen based on the recommendations of the IAEA Co-ordinated Research Project on the WIMS Library Update Project. The reference solution for the problem was obtained using Monte Carlo code MCNP5. It was found that the Ribon extended self-shielding model based on mathematical probability tables using correlation model performed better than all other models