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Sample records for shielded special nuclear

  1. Nuclear shields

    Linares, R.C.; Nienart, L.F.; Toelcke, G.A.

    1976-01-01

    A process is described for preparing melt-processable nuclear shielding compositions from chloro-fluoro substituted ethylene polymers, particularly PCTFE and E-CTFE, containing 1 to 75 percent by weight of a gadolinium compound. 13 claims, no drawings

  2. Shields for nuclear reactors

    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)

  3. Nuclear data for radiation shielding

    Miyasaka, Shunichi; Takahashi, Hiroshi.

    1976-01-01

    The third shielding expert conference was convened in Paris in Oct. 1975 for exchanging informations about the sensitivity evaluation of nuclear data in shielding calculation and integral bench mark experiment. The requirements about nuclear data presented at present from the field of nuclear design do not reflect sufficiently the requirements of shielding design, therefore it was the object to gather the requirements about nuclear data from the field of shielding. The nuclides used for shielding are numerous, and the nuclear data on these isotopes are required. Some of them cannot be ignored as the source of secondary γ-ray or in view of the radioactivation of materials. The requirements for the nuclear data of neutrons in the field of shielding are those concerning the reaction cross sections producing secondary γ-ray, the reaction cross sections including the production of secondary neutrons, elastic scattering cross sections, and total cross sections. The topics in the Paris conference about neutron shielding data are described, such as the methodology of sensitivity evaluation, the standardization of group constant libraries, the bench mark experiment on iron and sodium, and the cross section of γ-ray production. In the shielding of nuclear fission reactors, the γ-ray production owing to nuclear fission reaction is also important. In (d, t) fusion reactors, high energy neutrons are generated, and high energy γ-ray is emitted through giant E1 resonance. (Kako, I.)

  4. Radiation shield for nuclear reactors

    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

  5. Design of radiation shields in nuclear reactor core

    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

  6. Nuclear steam generator tubesheet shield

    Nickerson, J.H.D.; Ruhe, A.

    1982-01-01

    The invention involves improvements to a nuclear steam generator of the type in which a plurality of U-shaped tubes are connected at opposite ends to a tubesheet and extend between inlet and outlet chambers, with the steam generator including an integral preheater zone adjacent to the downflow legs of the U-shaped tubes. The improvement is a thermal shield disposed adjacent to an upper face of the tubesheet within the preheater zone, the shield including ductile cladding material applied directly to the upper face of the tubesheet, with the downflow legs of the U-shaped tubes extending through the cladding into the tubesheet

  7. Nuclear reactor neutron shielding

    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.

  8. Isotope effects on nuclear shielding

    Hansen, P.E.

    1983-01-01

    This review concentrates upon empirical trends and practical uses of mostly secondary isotope effects, both of the intrinsic and equilibrium types. The text and the tables are arranged in the following fashion. The most 'popular' isotope effect is treated first, deuterium isotope effects on 13 C nuclear shielding, followed by deuterium on 1 H nuclear shieldings, etc. Focus is thus on the isotopes producing the effect rather than on the nuclei suffering the effect. After a brief treatment of each type of isotope effect, general trends are dealt with. Basic trends of intrinsic isotope effects such as additivity, solvent effects, temperature effects, steric effects, substituent effects and hyperconjugation are discussed. Uses of isotope effects for assignment purposes, in stereochemical studies, in hydrogen bonding and in isotopic tracer studies are dealt with. Kinetic studies, especially of phosphates, are frequently performed by utilizing isotope effects. In addition, equilibrium isotope effects are treated in great detail as these are felt to be new and very important and may lead to new uses of isotope effects. Techniques used to obtain isotope effects are briefly surveyed at the end of the chapter. (author)

  9. Concrete shielding for nuclear ship 'Mutsu'

    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.)

  10. Concrete shielding for nuclear ship 'Mutsu'

    Nagase, Tetsuo; Nakajima, Tadao; Okumura, Tadahiko; Saito, Tetsuo

    1983-01-01

    The nuclear ship ''Mutsu'' was constructed in 1970 as the fourth in the world. On September 1, 1974, during the power raising test in the Pacific Ocean, radiation leak was detected. As the result of investigation, it was found that the cause was the fast neutrons streaming through the gap between the reactor pressure vessel and the primary shield. In order to repair the shielding facility, the Japan Nuclear Ship Research Development Agency carried out research and development and shielding design. It was decided to adopt serpentine concrete for the primary shield, which is the excellent moderator of fast neutrons even at high temperature, and heavy concrete for the secondary shield, which is effective for shielding both gamma ray and neutron beam. The repair of shielding was carried out in the Sasebo Shipyard, and completed in August, 1982. The outline of the repair work is reported. The weight increase was about 300 t. The conditions of the shielding design, the method of shielding analysis, the performance required for the shielding concrete, the preliminary experiment on heavy concrete and the construction works of serpentine concrete and heavy concrete are described. (Kako, I.)

  11. Improvements in or relating to nuclear shields

    Hawkins, R.J.; Riley, K.; Powell, C.

    1981-01-01

    A nuclear radiation shield comprises two pieces of steel held together edge to edge by a weld, the depth of which is less than the thickness of either of the edges. As the radiaion shielding effect of the weld will be less than the steel, an insert is bolted or welded over the weld. (U.K.)

  12. Special concrete shield selection using the analytic hierarchy process

    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

  13. Radiation shield for nuclear reactors

    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)

  14. Nuclear reactor shield including magnesium oxide

    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

  15. Shield structure for a nuclear reactor

    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.)

  16. Summary of Prometheus Radiation Shielding Nuclear Design Analyses , for information

    J. Stephens

    2006-01-01

    This report transmits a summary of radiation shielding nuclear design studies performed to support the Prometheus project. Together, the enclosures and references associated with this document describe NRPCT (KAPL and Bettis) shielding nuclear design analyses done for the project

  17. Nuclear data requirements for fusion reactor shielding

    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)

  18. Nuclear reactor cavity streaming shield

    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

  19. Methodology of shielding calculation for nuclear reactors

    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

  20. Development of radiation shielding standards in the American Nuclear Society

    Trubey, D.K.

    1975-11-01

    The American Nuclear Society (ANS) is a standards-writing organization-member of the American National Standards Institute (ANSI). The ANS Standards Committee has a subcommittee denoted ANS-6, Shielding, whose charge is to establish standards in connection with radiation protection and shielding, to provide shielding information to other standards writing groups, and to prepare recommended sets of shielding data and test problems. This paper is a progress report of this subcommittee

  1. Radiation shielding activities at the OECD/Nuclear Energy Agency

    Sartori, Enrico; Vaz, Pedro

    2000-01-01

    The OECD Nuclear Energy Agency (NEA) has devoted considerable effort over the years to radiation shielding issues. The issues are addressed through international working groups. These activities are carried out in close co-ordination and co-operation with the Radiation Safety Information Computational Center (RSICC). The areas of work include: basic nuclear data activities in support of radiation shielding, computer codes, shipping cask shielding applications, reactor pressure vessel dosimetry, shielding experiments database. The method of work includes organising international code comparison exercises and benchmark studies. Training courses on radiation shielding computer codes are organised regularly including hands-on experience in modelling skills. The scope of the activity covers mainly reactor shields and spent fuel transportation packages, but also fusion neutronics and in particular shielding of accelerators and irradiation facilities. (author)

  2. Shielding designs and tests of a new exclusive ship for transporting spent nuclear fuels

    Ono, M.; Ito, D.; Kitano, T.; Ueki, K.; Akiyama, H.; Obara, I.; Sanui, T.

    2000-01-01

    The Rokuei-Maru, a ship built specially for the transport of spent nuclear fuels in casks, was launched April in 1996. She is the first ship to comply with special Japanese regulations, KAISA 520, based on the INF code. DOT3.5 and MCNP-4A were used for the evaluation of dose equivalent rates of her shielding structures. On-board gamma-ray shielding tests were executed to confirm the effectiveness of the ship's shielding performance. The tests confirmed that effective shielding has been achieved and the dose equivalent rate in the accommodation and other inhabited spaces is sufficiently lower than the regulated limitations. This was achieved by employing the appropriate calculation methods and shielding materials. (author)

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

    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)

  4. Shield design development of nuclear propulsion merchant ship

    Tanaka, Yoshihisa

    1975-01-01

    Shielding design both in Japan and abroad for nuclear propulsion merchant ships is explained, with emphasis on the various technological problems having occurred in the shield design for one-body type and separate type LWRs as conceptual design. The following matters are described: the peculiarities of the design as compared with the case of land-based nuclear reactors, problems in the design standards of shielding, the present status and development of the design methods, and the instances of the design; thereby, the trends of shielding design are disclosed. The following matters are pointed out: Importance of the optimum design, of shielding, significance of radiation streaming through large voids, activation of the secondary water in built-in type steam generators, and the need of the guides for shield design. (Mori, K.)

  5. Self Shielding in Nuclear Fissile Assay Using LSDS

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

    2012-01-01

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

  6. MFTF-α+T end cell vacuum vessel and nuclear shield trade studies

    Kirchner, J.

    1984-01-01

    Three separate and distinct vacuum vessel and nuclear shield trade studies were performed in series. The studies are: vacuum topology, nuclear shield location and composition, and water bulk shield location and material selection

  7. Analysis of shield for the nuclear ship MUTSU

    Fuse, Takayoshi; Takeuchi, Kiyoshi; Yamaji, Akio

    1975-01-01

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

  8. QED Theory of the Nuclear Magnetic Shielding in Hydrogenlike Ions

    Yerokhin, V. A.; Pachucki, K.; Harman, Z.; Keitel, C. H.

    2011-01-01

    The shielding of the nuclear magnetic moment by the bound electron in hydrogenlike ions is calculated ab initio with inclusion of relativistic, nuclear, and quantum electrodynamics (QED) effects. The QED correction is evaluated to all orders in the nuclear binding strength parameter and, independently, to the first order in the expansion in this parameter. The results obtained lay the basis for the high-precision determination of nuclear magnetic dipole moments from measurements of the g factor of hydrogenlike ions.

  9. Space nuclear reactor shields for manned and unmanned applications

    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)

  10. Space nuclear reactor shields for manned and unmanned applications

    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

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

    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.

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

    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.)

  13. Nuclear shielding of openings in ITER Tokamak building

    Dammann, A., E-mail: alexis.dammann@iter.org [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Arumugam, A.P.; Beaudoin, V.; Beltran, D.; Benchikhoune, M.; Berruyer, F.; Cortes, P.; Gandini, F. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Ghirelli, N. [ASSYSTEM E.O.S, ZAC Saint Martin, 23, rue Benjamin Franklin, 84120 Pertuis (France); Gray, A.; Hurzlmeier, H.; Le Page, M. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Lemée, A. [SOGETI High Tech, 180 Rue René Descartes, 13851 Aix en Provence (France); Lentini, G.; Loughlin, M.; Mita, Y.; Patisson, L.; Rigoni, G.; Rathi, D.; Song, I. [ITER Organization, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France)

    2013-10-15

    Highlights: ► Establishment of a methodology to design shielded opening in external wall of the Tokamak building. ► Analysis of the shielding requirement, case by case, depending on the localization and the context. ► Implementation of an integrated solution for shielded opening. -- Abstract: The external walls of the Tokamak building, made of thick concrete, provide the nuclear shielding for operators working in adjacent buildings and for the environment. There are a series of openings to these external walls, devoted to ducts or pipes for ventilation, waveguides and transmission lines for heating systems and diagnostics, cooling pipes, cable trays or busbars. The shielding properties of the wall shall be preserved by adequate design of the openings in order not to affect the radiological zoning in adjacent areas. For some of them, shielding properties of the wall are not affected because the size of the network is quite small or the source is far from the opening. But for most of the openings, specific features shall be considered. Even if the approach is the same and the ways to shield can be standardized, specific analysis is requested in any case because the constraints are different.

  14. Non-combustible nuclear radiation shields with high hydrogen content

    Hall, W.C.; Peterson, J.M.

    1978-01-01

    The invention relates to compositions, methods of production, and uses of non-combustible nuclear radiation shields, with particular emphasis on those containing a high concentration of hydrogen atoms, especially effective for moderating neutron energy by elastic scatter, dispersed as a discontinuous phase in a continuous phase of a fire resistant matrix

  15. Smuggling special nuclear materials

    Lazaroiu, Gheorghe

    1999-01-01

    Ever since the collapse of the former Soviet Union reports have circulated with increasing frequency concerning attempts to smuggle materials from that country's civil and military nuclear programs. Such an increase obviously raises a number of concerns (outlined in the author's introduction), chief among which is the possibility that these materials might eventually fall into the hands of proliferant states or terrorist groups. The following issues are presented: significance of materials being smuggled; sources and smuggling routes; potential customers; international efforts to reduce nuclear smuggling; long-term disposition of fissile materials. (author)

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

    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

  17. Shielding behavior of multi-transformation phase change materials (MTPCM) against nuclear radiations

    Kumar, Ravindra; Goplani, Deepak; Kumar, Rohitash; Das, Mrinal Kumar; Kumar, Pramod; Jodha, Ajay Singh; Misra, Manoj; Khatri, P.K.

    2008-01-01

    In nuclear hardened structures and AFV's, special shielding materials are being used to provide protection from radiations generated in nuclear blast. However, in blast an intense heat pulse is also generated along with radiation. Currently used shield does not take care of this heat pulse. Defence Laboratory, Jodhpur has developed multi transformation phase change materials (MTPCM) based cool panels for passive moderation of temperature in severe desert heat. The MTPCM contains light nuclei of hydrogen, carbon and oxygen, and thus can absorb good amount of neutrons. MTPCM can also absorb intense heat pulse along with heat generated by secondary fires during blast as its latent heat (160-170 J/g) without significant rise in temperature (melting point 36-38 deg. C). Thus MTPCM can provide protection against both radiation as well as heat pulse generated in a nuclear blast along with its designed regular function of passively moderating temperature below 40 deg C during severe desert summer. A study has been undertaken to explore multiple applications of MTPCM panel. Protection factor provided by standard MTPCM panels against neutron and gamma radiations (both initial and fall out) were measured and results compared with PF provided by special lining pad currently being used in AFV's and field structures for nuclear protection. It is observed that MTPCM provides good PF (2.17) against neutron which is better than currently used shield pads (PFP%1.8). Present paper discusses results of this study. (author)

  18. Active neutron technique for detecting attempted special nuclear material diversion

    Smith, G.W.; Rice, L.G. III.

    1979-01-01

    The identification of special nuclear material (SNM) diversion is necessary if SNM inventory control is to be maintained at nuclear facilities. (Special nuclear materials are defined for this purpose as either 235 U of 239 Pu.) Direct SNM identification by the detection of natural decay or fission radiation is inadequate if the SNM is concealed by appropriate shielding. The active neutron interrogation technique described combines direct SNM identification by delayed fission neutron (DFN) detection with implied SNM detection by the identification of materials capable of shielding SNM from direct detection. This technique is being developed for application in an unattended material/equipment portal through which items such as electronic instruments, packages, tool boxes, etc., will pass. The volume of this portal will be 41-cm wide, 53-cm high and 76-cm deep. The objective of this technique is to identify an attempted diversion of at least 20 grams of SNM with a measurement time of 30 seconds

  19. Device for sealing and shielding a nuclear fuel storage tank

    Masaki, Gengo.

    1975-01-01

    Object: To provide a shield device for opening and closing a great opening in a relay-storage-tank within a hot cell for temporarily storing a nuclear fuel, in which the device is simplified in construction and which can perform the opening and closing operation in simple, positive and quick manner. Structure: A biological shield is positioned upwardly of an opening of a nuclear fuel storage tank to render an actuator inoperative. A sealing plate, which is pivotally supported by a plurality of support rod devices from the biological shield for parallel movement with respect to the biological shield, comes in contact with a resilient seal disposed along the entire peripheral edge of the opening to form an air-tight seal therebetween. In order to release the opening, the actuator is first actuated and the end of the sealing plate is horizontally pressed by a piston rod thereof. Then, the sealing plate is moved along the line depicted by the end of the support rod in the support rod devices and as a consequence, the plate is moved away from the resilient seal in the peripheral edge of the opening. When a driving device is actuated to travel the plate along the aforesaid line while maintaining the condition as described, the biological device moves along the guide. (Kamimura, M.)

  20. Technology development for special nuclear components

    Sanatkumar, A.

    1994-01-01

    One of the attractive features of Candu Pressurised Heavy Water Reactor design which influenced the decision to make it the foundation of our nuclear power programme, is that its main components (calandria, end shields, coolant channel components) are relatively simple - in comparison with reactor pressure vessel and associated components of Boiling Water Reactors or Pressurised Water Reactors - and considered to be within the scope of manufacture of developing countries. Over the last two decades, India has been very successful in technology development in many important and critical areas. We are now about to launch the construction of the first 500 MWe PHWR project at Tarapur. In this context, this paper focuses attention on some of the aspects relating to self-reliance in design, engineering and manufacture of these special components as currently perceived. (author). 3 refs

  1. Shielding modification and safety review on the nuclear ship 'Mutsu'

    Osanai, Masao

    1978-01-01

    The Japan Atomic Energy Commission (JAEC) called on the Japan Nuclear Ship Development Agency (JNSDA) for shielding modification and safety review on the nuclear ship ''Mutsu'', and JNSDA has conducted the research and development (R and D) to meet the request of JAEC for the above two items. Concerning the shield modification, the following matters are described: the study on the cause of radiation leakage which was concluded to the fast neutron streaming, the conceptual design for this modification, the mock up experiment for shielding utilizing JRR-4, the basic design following on the conceptual design, including the detailed drawings of the modified construction and the shielding analysis using RADHEAT-V3 code, and the relating experiments such as the heat transfer test of the primary shielding structure and the test of strength in stranding. As for the safety review, the survey of the troubles and the technical problems having been experienced in the light water reactor plants of land use, for example, fuel integrity, stress corrosion cracking and the leakage of steam generator tubes, the revision of the design so as to adapt to current safety standards and regulations, for example, in-service inspection, the setting of additional leak detectors in the primary cooling system, the modification of emergeney filters, etc., and the review of the design and construction corresponding to recent R and D works, such as re-evaluation of the core design, cooling capability of natural circulation, thermal stress analysis of main pipings, and the evaluation of ECCS performance are presented . (Nakai, Y.)

  2. Testing of the PELSHIE shielding code using Benchmark problems and other special shielding models

    Language, A.E.; Sartori, D.E.; De Beer, G.P.

    1981-08-01

    The PELSHIE shielding code for gamma rays from point and extended sources was written in 1971 and a revised version was published in October 1979. At Pelindaba the program is used extensively due to its flexibility and ease of use for a wide range of problems. The testing of PELSHIE results with the results of a range of models and so-called Benchmark problems is desirable to determine possible weaknesses in PELSHIE. Benchmark problems, experimental data, and shielding models, some of which were resolved by the discrete-ordinates method with the ANISN and DOT 3.5 codes, were used for the efficiency test. The description of the models followed the pattern of a classical shielding problem. After the intercomparison with six different models, the usefulness of the PELSHIE code was quantitatively determined [af

  3. γ-ray shielding behaviors of some nuclear engineering materials

    Mann, Kulwinder Singh

    2017-01-01

    The essential requirement of a material to be used for engineering purposes at nuclear establishments is its ability to attenuate the most penetrating ionizing radiations, gamma (γ)-rays. Mostly, high-Z materials such as heavy concrete, lead, mercury, and their mixtures or alloys have been used in the construction of nuclear establishments and thus termed as nuclear engineering materials (NEM). The NEM are classified into two categories, namely opaque and transparent, depending on their behavior towards the visible spectrum of EM waves. The majority of NEM are opaque. By contrast, various types of glass, which are transparent to visible light, are necessary at certain places in the nuclear establishments. In the present study, γ-ray shielding behaviors (GSB) of six glass samples (transparent NEM) were evaluated and compared with some opaque NEM in a wide range of energy (15 keV–15 MeV) and optical thickness (OT). The study was performed by computing various γ-ray shielding parameters (GSP) such as the mass attenuation coefficient, equivalent atomic number, and buildup factor. A self-designed and validated computer-program, the buildup factor-tool, was used for various computations. It has been established that some glass samples show good GSB, thus can safely be used in the construction of nuclear establishments in conjunction with the opaque NEM as well

  4. γ-ray shielding behaviors of some nuclear engineering materials

    Mann, Kulwinder Singh [Dept. of Physics, D.A.V. College, Punjab (India)

    2017-06-15

    The essential requirement of a material to be used for engineering purposes at nuclear establishments is its ability to attenuate the most penetrating ionizing radiations, gamma (γ)-rays. Mostly, high-Z materials such as heavy concrete, lead, mercury, and their mixtures or alloys have been used in the construction of nuclear establishments and thus termed as nuclear engineering materials (NEM). The NEM are classified into two categories, namely opaque and transparent, depending on their behavior towards the visible spectrum of EM waves. The majority of NEM are opaque. By contrast, various types of glass, which are transparent to visible light, are necessary at certain places in the nuclear establishments. In the present study, γ-ray shielding behaviors (GSB) of six glass samples (transparent NEM) were evaluated and compared with some opaque NEM in a wide range of energy (15 keV–15 MeV) and optical thickness (OT). The study was performed by computing various γ-ray shielding parameters (GSP) such as the mass attenuation coefficient, equivalent atomic number, and buildup factor. A self-designed and validated computer-program, the buildup factor-tool, was used for various computations. It has been established that some glass samples show good GSB, thus can safely be used in the construction of nuclear establishments in conjunction with the opaque NEM as well.

  5. γ-Ray Shielding Behaviors of Some Nuclear Engineering Materials

    Kulwinder Singh Mann

    2017-06-01

    Full Text Available The essential requirement of a material to be used for engineering purposes at nuclear establishments is its ability to attenuate the most penetrating ionizing radiations, gamma (γ-rays. Mostly, high-Z materials such as heavy concrete, lead, mercury, and their mixtures or alloys have been used in the construction of nuclear establishments and thus termed as nuclear engineering materials (NEM. The NEM are classified into two categories, namely opaque and transparent, depending on their behavior towards the visible spectrum of EM waves. The majority of NEM are opaque. By contrast, various types of glass, which are transparent to visible light, are necessary at certain places in the nuclear establishments. In the present study, γ-ray shielding behaviors (GSB of six glass samples (transparent NEM were evaluated and compared with some opaque NEM in a wide range of energy (15 keV–15 MeV and optical thickness (OT. The study was performed by computing various γ-ray shielding parameters (GSP such as the mass attenuation coefficient, equivalent atomic number, and buildup factor. A self-designed and validated computer-program, the buildup factor-tool, was used for various computations. It has been established that some glass samples show good GSB, thus can safely be used in the construction of nuclear establishments in conjunction with the opaque NEM as well.

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

    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)

  7. Special issue: the nuclear industry in Europe

    Anon.

    1982-01-01

    This special issue contains papers on the following topics: French nuclear policy; nuclear energy development in Europe; nuclear diversification; Alsthom-Atlantique in the nuclear field; 1981 nuclear electricity generation; EDF siting policy; the N4 model of the 1300 MW series; Creys-Malville; the nuclear industry in Europe; pumps in the nuclear industry [fr

  8. Special nuclear material simulation device

    Leckey, John H.; DeMint, Amy; Gooch, Jack; Hawk, Todd; Pickett, Chris A.; Blessinger, Chris; York, Robbie L.

    2014-08-12

    An apparatus for simulating special nuclear material is provided. The apparatus typically contains a small quantity of special nuclear material (SNM) in a configuration that simulates a much larger quantity of SNM. Generally the apparatus includes a spherical shell that is formed from an alloy containing a small quantity of highly enriched uranium. Also typically provided is a core of depleted uranium. A spacer, typically aluminum, may be used to separate the depleted uranium from the shell of uranium alloy. A cladding, typically made of titanium, is provided to seal the source. Methods are provided to simulate SNM for testing radiation monitoring portals. Typically the methods use at least one primary SNM spectral line and exclude at least one secondary SNM spectral line.

  9. Validation of nuclear models used in space radiation shielding applications

    Norman, Ryan B.; Blattnig, Steve R.

    2013-01-01

    A program of verification and validation has been undertaken to assess the applicability of models to space radiation shielding applications and to track progress as these models are developed over time. In this work, simple validation metrics applicable to testing both model accuracy and consistency with experimental data are developed. The developed metrics treat experimental measurement uncertainty as an interval and are therefore applicable to cases in which epistemic uncertainty dominates the experimental data. To demonstrate the applicability of the metrics, nuclear physics models used by NASA for space radiation shielding applications are compared to an experimental database consisting of over 3600 experimental cross sections. A cumulative uncertainty metric is applied to the question of overall model accuracy, while a metric based on the median uncertainty is used to analyze the models from the perspective of model development by examining subsets of the model parameter space.

  10. {sup 3}He detector analysis of some special shielding materials

    Avdic, S; Pesic, M [Boris Kidric, Institute of Nuclear Sciences, Beograd (Yugoslavia); Marinkovic, P [ETF Belgrade Univ. (Yugoslavia)

    1990-07-01

    The shielding properties of commercial materials of reactor Experiments, Inc. (R/X) were analyzed at the facility which includes bare heavy water experimental reactor RB with external neutron converter ENC, The fast neutron spectrum measurements in energy range from 1 MeV to 10 MeV was performed using ORTEC semiconductor neutron detector with He{sup 3} in diode coincidence arrangement. The neutron spectra have been evaluated from measured pulse-height distribution using numerical code HE3 for computation of detector efficiency in a collimated neutron beam. The neutron dose rates behind ENC with and without sample R/X material were determined using cubic spline interpolation routine for calculating the corresponding flux-dose rate conversion factors. Satisfactory shielding properties of the examined material in a fast neutron field in measurements and calculations are demonstrated. (author)

  11. Shielding estimation for nuclear medicine therapy ward: our experience

    Skopljak-Beganovic, A.; Kucukalic-Selimovic, E.; Beganovic, A.; Drljevic, A.

    2008-01-01

    Full text: The aim of this study was to calculate and estimate the shielding thickness for a new Nuclear Medicine Therapy Ward. Parameters available for shielding calculation were: ground plan of the ward, radionuclides planned for use, maximum administered activity of I-131, maximum delivered activity of I-131 to the ward per week, average time spent in the hospital after the treatment. The most hazardous and most commonly used radioisotope is I-131. The target dose that needs to be met for occupationally exposed workers is 0.3 mSv per year. There are several factors that could be changed in order to achieve this value: distance from the source, shielding thickness, angle of incidence, occupational and usage factors. The maximum dose rate at 1 meter from the thyroid gland of the patient was considered to be 100 mSv/h. The distances and incidence angles could not be changed since these vales were predetermined in the ground plan. Different usage and occupational factors were used for different rooms in the ward. We used occupational factor 1 for the bed and 1/6 for the bathroom, and usage factor 1 for nurses' room and patient room and 1/6 for the corridors, etc. The easiest way of calculating dose attenuation in material was by introducing the HVL and TVL for broad beams. TVL and HVL were taken from the graph.The results show that shielding thickness should be in the range of 3 mmPb for room doors to 30 mmPb for the wall adjacent to the nurse's office. Most of the walls are 20 mmPb thick. These values were calculated using conservative assumptions and are more then enough to protect staff, patients and public from external radiation. If the construction cannot support the weight of lead some rearrangements regarding patient positions could be made. (author)

  12. Shielding requirements for the transport of nuclear warhead components under decommissioning

    Hansen, L.F.

    1994-09-01

    The requirements to carry out accurate shielding calculations involved with the safe off-site transportation of packages containing nuclear warhead components, special assemblies and radioactive materials are discussed. The need for (a) detailed information on the geometry and material composition of the packaging and radioactive load, (b) accurate representation of the differential energy spectra (dN/dE) for the neutron and gamma spectra emitted by the radioactive materials enclosed in the packaging, (c) well-tested neutron and photon cross section libraries, (d) and accurate three-dimensional Monte Carlo transport codes are illustrated. A brief discussion of the need for reliable dose measurements is presented

  13. Shield nuclear design for the 5-kWe TE system

    Keshishian, V.

    1972-01-01

    The nuclear analysis of the 5-kW(e) reactor shield is presented. Calculation methods and optimization techniques used are presented. Borated stainless steel was selected for the gamma ray shield with tungsten alloy as an alternate. The total shield weight was calculated to be 355 lb. (U.S.)

  14. E-Alerts: Nuclear science and technology (radiation shielding, protection, and safety). E-mail newsletter

    1999-01-01

    Topics include: Shielding design, nuclear radiation transport properties of materials, decontamination; Container design and transportation requirements for radioactive materials; and Fallout shelters

  15. Use of multiple radiographic techniques for nuclear shielding development

    Day, S., E-mail: dayse@mcmaster.ca [McMaster Univ., McMaster Nuclear Reactor, Hamilton, Ontario (Canada)

    2016-01-15

    The McMaster Nuclear Reactor (MNR) is a medium-sized research reactor located on the campus of McMaster University in Hamilton, Ontario. The largest nuclear reactor on a Canadian university campus, MNR is an open-pool, light water research reactor, Equipped with both in-core and ex-core irradiation facilities, MNR serves research, education, and industrial needs while producing radioisotopes for medical applications. Presently the University is preparing for the installation of a positron beam line facility at one of the MNR beam ports. One of five such facilities worldwide, the McMaster Intense Positron Beam Facility (MIPBF) will provide orders of magnitude more positrons than a standard bench-top arrangement, making possible experiments not previously feasible. Funded by the Canadian Foundation for Innovation and the Ontario Research Foundation, the MIPBF project is a collaborative effort between positron groups at McMaster, Western, and York universities and researchers and staff at MNR, The project has involved design and fabrication of both the positron production and transport system and custom biological shielding. The MIPBF shielding fabrication is scheduled to be complete by fall 2015 followed shortly by installation and commissioning testing at MNR. Positron beam availability is planned for later in 2016. Applications include defect characterization and surface analysis of advanced engineering materials and fundamental science experiments on antimatter properties, The MIPBF will be the latest addition to the extensive array of materials-related facilities at McMaster University. (author)

  16. Use of multiple radiographic techniques for nuclear shielding development

    Day, S.

    2016-01-01

    The McMaster Nuclear Reactor (MNR) is a medium-sized research reactor located on the campus of McMaster University in Hamilton, Ontario. The largest nuclear reactor on a Canadian university campus, MNR is an open-pool, light water research reactor, Equipped with both in-core and ex-core irradiation facilities, MNR serves research, education, and industrial needs while producing radioisotopes for medical applications. Presently the University is preparing for the installation of a positron beam line facility at one of the MNR beam ports. One of five such facilities worldwide, the McMaster Intense Positron Beam Facility (MIPBF) will provide orders of magnitude more positrons than a standard bench-top arrangement, making possible experiments not previously feasible. Funded by the Canadian Foundation for Innovation and the Ontario Research Foundation, the MIPBF project is a collaborative effort between positron groups at McMaster, Western, and York universities and researchers and staff at MNR, The project has involved design and fabrication of both the positron production and transport system and custom biological shielding. The MIPBF shielding fabrication is scheduled to be complete by fall 2015 followed shortly by installation and commissioning testing at MNR. Positron beam availability is planned for later in 2016. Applications include defect characterization and surface analysis of advanced engineering materials and fundamental science experiments on antimatter properties, The MIPBF will be the latest addition to the extensive array of materials-related facilities at McMaster University. (author)

  17. Control of Nuclear Materials and Special Equipment (Nuclear Safety Regulations)

    Cizmek, A.; Prah, M.; Medakovic, S.; Ilijas, B.

    2008-01-01

    Based on Nuclear Safety Act (OG 173/03) the State Office for Nuclear Safety (SONS) in 2008 adopted beside Ordinance on performing nuclear activities (OG 74/06) and Ordinance on special conditions for individual activities to be performed by expert organizations which perform activities in the area of nuclear safety (OG 74/06) the new Ordinance on the control of nuclear material and special equipment (OG 15/08). Ordinance on the control of nuclear material and special equipment lays down the list of nuclear materials and special equipment as well as of nuclear activities covered by the system of control of production of special equipment and non-nuclear material, the procedure for notifying the intention to and filing the application for a license to carry out nuclear activities, and the format and contents of the forms for doing so. This Ordinance also lays down the manner in which nuclear material records have to be kept, the procedure for notifying the State administration organization (regulatory body) responsible for nuclear safety by the nuclear material user, and the keeping of registers of nuclear activities, nuclear material and special equipment by the State administration organization (regulatory body) responsible for nuclear safety, as well as the form and content of official nuclear safety inspector identification card and badge.(author)

  18. Safeguards for special nuclear materials

    Carlson, R.L.

    1979-12-01

    Safeguards, accountability, and nuclear materials are defined. The accuracy of measuring nuclear materials is discussed. The use of computers in nuclear materials accounting is described. Measures taken to physically protect nuclear materials are described

  19. Shielding considerations for advanced space nuclear reactor systems

    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

  20. Carbon-13 isotope effects on 199Hg nuclear shielding

    Sebald, Angelika; Wrackmeyer, Bernd

    1985-01-01

    Secondary 13 C/ 12 C isotope effects on 199 Hg nuclear shielding (Δdeltasup(i)( 199 Hg)) are of interest because of the unusual shift to high frequency which has been observed for a few alkyl mercury compounds. Continuing interest in the NMR parameters of mercury compounds prompted a search for the values Δdeltasup(i)( 199 Hg) in a greater variety of organomercurials. This should help to find out about the range of Δdeltasup(i)( 199 Hg) and to obtain a firmer basis for the discussion of the high-frequency shifts. The data and experimental conditions are given for chemical shifts delta 199 Hg, coupling constants sup(n)J( 199 Hg 13 C) and 13 C/ 12 C isotope shifts Δdeltasup(i)( 199 Hg) of fourteen 199 Hg organomercury compounds. The results are discussed. (author)

  1. Status of the ITER tokamak nuclear shielding and radiological protection design

    Leichtle, D., E-mail: dieter.leichtle@f4e.europa.eu [Fusion for Energy, Josep Pla 2, Barcelona 08019 (Spain); Chaffard, P.Y.; Izquierdo, J. [Fusion for Energy, Josep Pla 2, Barcelona 08019 (Spain); Juarez, R. [UNED, Juan del Rosal 12, Madrid 28040 (Spain); Pampin, R.; Portone, A. [Fusion for Energy, Josep Pla 2, Barcelona 08019 (Spain)

    2016-11-01

    Highlights: • Comprehensive review of design status of the ITER tokamak regarding nuclear shielding. • Investigation of shield design options and streaming mitigation measures. • Review of state-of-the-art shutdown dose rate analyses for selected port systems. - Abstract: Nuclear shielding of the ITER tokamak encompasses several systems and interfaces in a complex radiation environment. Therefore any shielding design has to involve a series of structures, systems and components in an integrated approach. This is evident for the complex ex-vessel radiation environment with streaming and leakage of plasma neutrons and subsequent activation of ex-vessel structures which give raise to excessive shutdown dose rates in accessible areas of the cryostat. The paper reviews recent nuclear analyses related to the performance of primary shields and highlights challenges toward an integrated nuclear shielding design. The general need of propagation of shielding requirements is highlighted in the context of radiation cross talk due to penetrations. Radiation streaming through gaps and penetrations is a key problem in any efficient shield design. The impact on the evolving radiation environment due to several design options along streaming paths such as port gaps, as well as their modeling for nuclear analysis, is presented. Implications regarding design integration and compliance with integrated shielding requirements and ALARA dose are finally given.

  2. Special monitoring in nuclear medicine

    Beltran, C.C.; Puerta, J.A.; Morales, J.

    2006-01-01

    Colombia counts with around 56 centers of Nuclear Medicine, 70 Nuclear Doctors and more of 100 Technologists in this area. The radioisotopes more used are the 131 I and the 99m Tc. The radiological surveillance singular in the country is carried out for external dosimetry, being the surveillance by incorporation of radioactive materials very sporadic in our media. Given the necessity to implement monitoring programs in the incorporation of radionuclides of the occupationally exposed personnel, in the routine practice them routine of Nuclear Medicine, it was implemented a pilot program of Special Monitoring with two centers of importance in the city of Medellin. This program it was carried out with the purpose of educating, to stimulate and to establish a program of reference monitoring with base in the National Program of Monitoring in the radionuclides Incorporation that serves like base for its application at level of all the services of Nuclear Medicine in the country. This monitoring type was carried out with the purpose of obtaining information on the work routine in these centers, form of manipulation and dosage of the radionuclides, as well as the administration to the patient. The application of the program was carried out to define the frequency of Monitoring and analysis technique for the implementation of a program of routine monitoring, following the recommendations of the International Commission of Radiological Protection. For their application methods of activity evaluation were used in urine and in 7 workers thyroid, of those which only two deserve an analysis because they presented important activities. The measures were carried out during one month, every day by means in urine samples and to the most critic case is practiced two thyroid measures, one in the middle of the period and another when concluding the monitoring. To the other guy is practiced an activity count in thyroid when concluding the monitoring period. The obtained result of the

  3. Nuclear magnetic shielding tensors of 207Pb2+ in Pb(NO3)2

    Lutz, O.; Nolle, A.

    1980-01-01

    The NMR signals of 207 Pb were observed in a single crystal of Pb(NO 3 ) 2 and could be assigned to the four different Pb 2+ sites by the dependence of the linewidths on the orientation. Four different nuclear magnetic shielding tensors with equal principal values but with different characteristic vectors could be determined. The symmetry of the shielding tensors is in agreement with the symmetry at the Pb 2+ sites. It is shown, that intermolecular contributions can not account for the anisotropy of the nuclear magnetic shielding, which is 3 0 / 00 of the isotropic absolute magnetic shielding. (orig.)

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

    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.

  5. A shielded measurement system for irradiated nuclear fuel measurements

    Mosby, W.R.; Aumeier, S.E.; Klann, R.T.

    1999-01-01

    The US Department of Energy (DOE) is driving a transition toward dry storage of irradiated nuclear fuel (INF), toward characterization of INF for final disposition, and toward resumption of measurement-based material control and accountability (MC and A) efforts for INF. For these reasons, the ability to efficiently acquire radiological measurements of INF in a dry environment is important. The DOE has recently developed a guidance document proposing MC and A requirements for INF. The intent of this document is to encourage the direct measurement of INF on inventory within DOE. The guidance document reinforces and clarifies existing material safeguards requirements as they pertain to INF. Validation of nuclear material contents of non-self-protecting INF must be accomplished by direct measurement, application of validated burnup codes using qualified initial fissile content, burnup data, and age or by other valid means. The fuel units must remain intact with readable identification numbers. INF may be subject to periodic inventories with visual item accountability checks. Quantitative measurements may provide greater assurance of the integrity of INF inventories at a lower cost and with less personnel exposure than visual item accountability checks. Currently, several different approaches are used to measure the radiological attributes of INF. Although these systems are useful for a wide variety of applications, there is currently no relatively inexpensive measurement system that is readily deployable for INF measurements for materials located in dry storage. The authors present the conceptual design of a shielded measurement system (SMS) that could be used for this purpose. The SMS consists of a shielded enclosure designed to house a collection of measurement systems to allow measurements on spent fuel outside of a hot cell. The phase 1 SMS will contain 3 He detectors and ionization chambers to allow for gross neutron and gamma-ray measurements. The phase 2 SMS

  6. Multiconfigurational self-consistent field calculations of nuclear shieldings using London atomic orbitals

    Ruud, Kenneth; Helgaker, Trygve; Kobayashi, Rika

    1994-01-01

    to corresponding individual gauges for localized orbitals (IGLO) results. The London results show better basis set convergence than IGLO, especially for heavier atoms. It is shown that the choice of active space is crucial for determination of accurate nuclear shielding constants.......Nuclear shielding calculations are presented for multiconfigurational self-consistent field wave functions using London atomic orbitals (gauge invariant atomic orbitals). Calculations of nuclear shieldings for eight molecules (H2O, H2S, CH4, N2, CO, HF, F2, and SO2) are presented and compared...

  7. Shielding tests for a new ship for the transport of spent nuclear fuels

    Ito, D.; Kitano, T.; Akiyama, H.; Ueki, K.; Sanui, T.

    1998-01-01

    a new ship for the transport of spent nuclear fuels which uses serpentine concrete as its major shielding material has been constructed. The shielding calculations are based on DOT3.5 code (CCC-276) and the DLC23). Experiments with Cf-252 and Co-60 sources were carried out to confirm the validity of this method of calculating the shielding effectiveness of serpentine concrete. In these experiments, neutron and gamma-ray dose equivalent rates were measured in various arrangements to simulate the shielding structures of the ship, the calculations for each arrangement were performed by this shielding calculation method. For both neutron and gamma-rays, the calculation results agreed with the experiments very well, confirming that this calculation method used in the ship's shielding design is valid. Two kinds of on-board gamma-ray shielding tests were performed to confirm the ship's actual shielding effectiveness. In one kind of test, gamma-ray dose equivalent rates were measured for each shielding wall using Co-60 sources. In the other kind of test, gamma-ray dose equivalent rates in the ship's accommodation area were measured when a strong Co-60 source was placed in a loaded shipping cask's position. In both gamma-ray shielding tests all measured dose equivalent rates were less than the calculated values, confirming that the ship's actual shielding is sufficient to meet safety requirements. (authors)

  8. Characterization of serpentine. A potential nuclear shielding material

    Sengupta, A.; Rajeswari, B.; Kadam, R.M.; Kshirsagar, R.J.

    2012-01-01

    The use of serpentine as a potential nuclear shielding material necessitates a chemical quality control of the samples before its use in reactors. With this in view, characterization of these mineral samples was carried out using inductively coupled plasma atomic emission spectrometry (ICP-AES) and Instrumental neutron activation analysis (INAA) methods. The analytical results obtained by both ICP-AES and NAA techniques were found to be comparable. Na, Cr, Co, Zn, and Cu were found to be present in all samples of Indian origin while Ga, Ag, Ni, and Cd were found to below the limits of detection. A comparison on the detection limits of elements of interest was also carried out by both the analytical techniques and found to be in good agreement. An infrared spectroscopic investigation was also carried out on all the mineral samples. Bands at 3,689 and 3,648 cm -1 were attributed to inner and outer hydroxyl stretching of Mg-OH, respectively. The weak and broad band centered around 3,416 cm -1 was assigned due to the stretching vibrations of the adsorbed water molecules while three bands at 1076, 1022 and 968 cm -1 were prescribed to the vibrations of the SiO 4 tetrahedra. (author)

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

    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)

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

    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

  11. Guideline on radiation protection requirements for ionizing radiation shielding in nuclear power plants

    1988-01-01

    The guideline which entered into force on 1 May 1988 stipulates the radiation protection requirements for shielding against ionizing radiation to be met in the design, construction, commissioning, operation, and decommissioning of nuclear power plants

  12. Estimation methods for special nuclear materials holdup

    Pillay, K.K.S.; Picard, R.R.

    1984-01-01

    The potential value of statistical models for the estimation of residual inventories of special nuclear materials was examined using holdup data from processing facilities and through controlled experiments. Although the measurement of hidden inventories of special nuclear materials in large facilities is a challenging task, reliable estimates of these inventories can be developed through a combination of good measurements and the use of statistical models. 7 references, 5 figures

  13. Shielding device for control rod in nuclear reactor

    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.)

  14. Radiation shielding properties of a novel cement–basalt mixture for nuclear energy applications

    Ipbüker, Cagatay; Nulk, Helena; Gulik, Volodymyr [University of Tartu, Institute of Physics (Estonia); Biland, Alex [HHK Technologies, Houston (United States); Tkaczyk, Alan Henry, E-mail: alan@ut.ee [University of Tartu, Institute of Physics (Estonia)

    2015-04-01

    Highlights: • Basalt fiber is a relatively cheap material that can be used as reinforcement. • Gamma-ray attenuation remains relatively stable with addition of basalt fiber. • Neutron attenuation remains relatively stable with addition of basalt fiber. • Cement–basalt mixture has a good potential for use in nuclear energy applications. - Abstract: The radiation shielding properties of a new proposed building material, a novel cement–basalt fiber mixture (CBM), are investigated. The authors analyze the possibility of this material to be a viable substitute to outgoing materials in nuclear energy applications, which will lead to a further sustained development of nuclear energy in the future. This computational study involves four types of concrete with various amounts of basalt fiber in them. The gamma-ray shielding characteristics of proposed CBM material are investigated with the help of WinXCom program, whereas the neutron shielding characteristics are computed by the Serpent code. For gamma-ray shielding, we find that the attenuation coefficients of concretes with basalt fibers are not notably influenced by the addition of fibers. For neutron shielding, additional basalt fiber in mixture presents negligible effect on neutron radiation shielding. With respect to radiation shielding, it can be concluded that basalt fibers have good potential as an addition to heavyweight concrete for nuclear energy applications.

  15. Radiation shielding properties of a novel cement–basalt mixture for nuclear energy applications

    Ipbüker, Cagatay; Nulk, Helena; Gulik, Volodymyr; Biland, Alex; Tkaczyk, Alan Henry

    2015-01-01

    Highlights: • Basalt fiber is a relatively cheap material that can be used as reinforcement. • Gamma-ray attenuation remains relatively stable with addition of basalt fiber. • Neutron attenuation remains relatively stable with addition of basalt fiber. • Cement–basalt mixture has a good potential for use in nuclear energy applications. - Abstract: The radiation shielding properties of a new proposed building material, a novel cement–basalt fiber mixture (CBM), are investigated. The authors analyze the possibility of this material to be a viable substitute to outgoing materials in nuclear energy applications, which will lead to a further sustained development of nuclear energy in the future. This computational study involves four types of concrete with various amounts of basalt fiber in them. The gamma-ray shielding characteristics of proposed CBM material are investigated with the help of WinXCom program, whereas the neutron shielding characteristics are computed by the Serpent code. For gamma-ray shielding, we find that the attenuation coefficients of concretes with basalt fibers are not notably influenced by the addition of fibers. For neutron shielding, additional basalt fiber in mixture presents negligible effect on neutron radiation shielding. With respect to radiation shielding, it can be concluded that basalt fibers have good potential as an addition to heavyweight concrete for nuclear energy applications

  16. Uncovering Special Nuclear Materials by Low-energy Nuclear Reaction Imaging.

    Rose, P B; Erickson, A S; Mayer, M; Nattress, J; Jovanovic, I

    2016-04-18

    Weapons-grade uranium and plutonium could be used as nuclear explosives with extreme destructive potential. The problem of their detection, especially in standard cargo containers during transit, has been described as "searching for a needle in a haystack" because of the inherently low rate of spontaneous emission of characteristic penetrating radiation and the ease of its shielding. Currently, the only practical approach for uncovering well-shielded special nuclear materials is by use of active interrogation using an external radiation source. However, the similarity of these materials to shielding and the required radiation doses that may exceed regulatory limits prevent this method from being widely used in practice. We introduce a low-dose active detection technique, referred to as low-energy nuclear reaction imaging, which exploits the physics of interactions of multi-MeV monoenergetic photons and neutrons to simultaneously measure the material's areal density and effective atomic number, while confirming the presence of fissionable materials by observing the beta-delayed neutron emission. For the first time, we demonstrate identification and imaging of uranium with this novel technique using a simple yet robust source, setting the stage for its wide adoption in security applications.

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

    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

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

    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)

  19. Uranium-lead shielding for nuclear material transportation systems

    Lusk, E.C.; Miller, N.E.; Basham, S.J. Jr.

    1978-01-01

    The basis for the selection of shielding materials for spent fuel shipping containers is described with comments concerning the favorable and unfavorable aspects of steel, lead, and depleted uranium. A concept for a new type of material made of depleted uranium and lead is described which capitalizes on the best cask shielding characteristics of both materials. This cask shielding is made by filling the shielding cavity with pieces of depleted uranium and then backfilling the interstitial voids with lead. The lead would be bonded to the uranium and also to the cask shells if desired. Shielding density approaching 80 percent of that of solid uranium could be achieved, while a density of 65 percent is readily obtainable. This material should overcome the problems of the effect of lead melting in the fire accident, high thermal gradients at uranium-stainless steel interfaces and at a major reduction in cost over that of a solid uranium shielded cask. A development program is described to obtain information on the properties of the composite material to aid in design analysis and licensing and to define the fabrication techniques

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

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

    2001-03-01

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

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

    Fujita, Mitsutane; Noda, Tetsuji

    2001-01-01

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

  2. Shield materials recommended for space power nuclear reactors

    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.

  3. Pilot tests for dismantling by blasting of the biological shield of a shut down nuclear power station

    Freund, H.U.

    1995-01-01

    Following free-field tests on concrete blocks the feasibility of explosive dismantling of the biological shield of nuclear power stations has been succesfully tested at the former hotsteam reaction in Karlstein/Main Germany. For this purpose a model shield of scale 1:2 was embedded into the reactor structure at which bore-hole blasting tests employing up to about 15 kg of explosive were performed. An elaborate measurement system allowed to receive detailed information on the blast side-effects: Special emphasis was focussed on the quantitative registration of the dynamic blast loads; data for the transfer of the dismantling method to the removal of real ractor structures were obtained. (orig.) [de

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

    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

  5. Shielding analysis method applied to nuclear ship 'MUTSU' and its evaluation based on experimental analyses

    Yamaji, Akio; Miyakoshi, Jun-ichi; Iwao, Yoshiaki; Tsubosaka, Akira; Saito, Tetsuo; Fujii, Takayoshi; Okumura, Yoshihiro; Suzuoki, Zenro; Kawakita, Takashi.

    1984-01-01

    Procedures of shielding analysis are described which were used for the shielding modification design of the Nuclear Ship ''MUTSU''. The calculations of the radiation distribution on board were made using Sn codes ANISN and TWOTRAN, a point kernel code QAD and a Monte Carlo code MORSE. The accuracies of these calculations were investigated through the analysis of various shielding experiments: the shield tank experiment of the Nuclear Ship ''Otto Hahn'', the shielding mock-up experiment for ''MUTSU'' performed in JRR-4, the shielding benchmark experiment using the 16 N radiation facility of AERE Harwell and the shielding effect experiment of the ship structure performed in the training ship ''Shintoku-Maru''. The values calculated by the ANISN agree with the data measured at ''Otto Hahn'' within a factor of 2 for fast neutrons and within a factor of 3 for epithermal and thermal neutrons. The γ-ray dose rates calculated by the QAD agree with the measured values within 30% for the analysis of the experiment in JRR-4. The design values for ''MUTSU'' were determined in consequence of these experimental analyses. (author)

  6. FCXSEC: multigroup cross-section libraries for nuclear fuel cycle shielding calculations

    Ford, W.E. III; Webster, C.C.; Diggs, B.R.; Pevey, R.E.; Croff, A.G.

    1980-05-01

    Starting with the pseudo-composition-independent VITAMIN-C cross-sectin library, composition-dependent fine-(171n-36γ) and broad-group (22n-21γ) self-shielded AMPX master, broad-group microscopic ANISN-formatted, and broad-group macroscopic ANISN-formatted cross-section libraries were generated to be used for nuclear fuel cycle shielding calculations. The specifications for the data and the procedure used to prepare the libraries are described

  7. Minimizing exposure in nuclear medicine through optimum use of shielding devices

    Rutherford, B.L.; King, S.H.; Erdman, M.C.; Miller, K.L.

    1991-01-01

    Exposure to radiation from nuclear medicine nuclides can be minimized through the use of various shielding devices. This paper reviews the dose reductions achieved through use of various syringe shields, lead aprons, leaded gloves, and several types of eyeglass lenses for 67 Ga, 99m Tc, 131 I and 201 Tl. The authors have found that combination of devices can best provide for minimizing doses

  8. Special Nuclear Material Gamma-Ray Signatures for Reachback Analysts

    Karpius, Peter Joseph [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Myers, Steven Charles [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2016-08-29

    These are slides on special nuclear material gamma-ray signatures for reachback analysts for an LSS Spectroscopy course. The closing thoughts for this presentation are the following: SNM materials have definite spectral signatures that should be readily recognizable to analysts in both bare and shielded configurations. One can estimate burnup of plutonium using certain pairs of peaks that are a few keV apart. In most cases, one cannot reliably estimate uranium enrichment in an analogous way to the estimation of plutonium burnup. The origin of the most intense peaks from some SNM items may be indirect and from ‘associated nuclides.' Indirect SNM signatures sometimes have commonalities with the natural gamma-ray background.

  9. Gamma spectrometric discrimination of special nuclear materials

    Dowdall, M.; Mattila, A.; Ramebaeck, H.; Aage, H.K.; Palsson, S.E.

    2012-12-01

    This report presents details pertaining to an exercise conducted as part of the NKS-B programme using synthetic gamma ray spectra to simulate the type of data that may be encountered in the interception of material potentially containing special nuclear materials. A range of scenarios were developed involving sources that may or may not contain special nuclear materials. Gamma spectral data was provided to participants as well as ancillary data and participants were asked, under time constraint, to determine whether or not the data was indicative of circumstances involving special nuclear materials. The situations varied such that different approaches were required in order to obtain the correct result in each context. In the majority of cases participants were able to correctly ascertain whether or not the situations involved special nuclear material. Although fulfilling the primary goal of the exercise, some participants were not in a position to correctly identify with certainty the material involved, Situations in which the smuggled material was being masked by another source proved to be the most challenging for participants. (Author)

  10. Gamma spectrometric discrimination of special nuclear materials

    Dowdall, M. [Norwegian Radiation Protection Authority (Norway); Mattila, A. [Radiation and Nuclear Safety Authority, Helsinki (Finland); Ramebaeck, H. [Swedish Defence Research Agency, Stockholm (Sweden); Aage, H.K. [Danish Emergency Management Agency, Birkeroed (Denmark); Palsson, S.E. [Icelandic Radiation Safety Authority, Reykjavik (Iceland)

    2012-12-15

    This report presents details pertaining to an exercise conducted as part of the NKS-B programme using synthetic gamma ray spectra to simulate the type of data that may be encountered in the interception of material potentially containing special nuclear materials. A range of scenarios were developed involving sources that may or may not contain special nuclear materials. Gamma spectral data was provided to participants as well as ancillary data and participants were asked, under time constraint, to determine whether or not the data was indicative of circumstances involving special nuclear materials. The situations varied such that different approaches were required in order to obtain the correct result in each context. In the majority of cases participants were able to correctly ascertain whether or not the situations involved special nuclear material. Although fulfilling the primary goal of the exercise, some participants were not in a position to correctly identify with certainty the material involved, Situations in which the smuggled material was being masked by another source proved to be the most challenging for participants. (Author)

  11. Shielding calculations for ships carrying irradiated nuclear fuel

    Dean, M.H.

    1985-01-01

    A number of ships have been constructed to carry irradiated fuel from Japan to the U.K. and France, for reprocessing. About 20 transport flasks may be carried on each voyage. Permanent shielding must be provided on the ships to ensure that no member of the crew receives an annual dose greater than a specified limit. As the fuel is of varying type and radiation history, and as flasks of differing designs are used, many shielding calculations are needed. There are a number of difficulties in making shielding calculations for the ships. The geometry is complex, dimensions are large and considerable air spaces are involved. The paper considers possible methods of calculation. The line-of-sight method is chosen for most of the calculations, for both γ-radiation and neutrons. The basic data which is used in the calculations is described. As the methods of calculation are somewhat approximate, it is necessary to provide confirmation that they are sufficiently accurate. Validation has been provided in two ways. First, measurements have been made on board one of the ships, Pacific Crane, and these have been checked against calculation. Second, a simplified model of the flasks and ship has been set up, and calculations checked against more sophisticated methods. Results of the validation checks are presented, and it is shown that adequate accuracy is achieved. (author)

  12. Optimized shielding calculation to the transport of 131I employed in nuclear medicine

    Sahyun, A.; Sordi, G.M.; Rodrigues, D.; Sanches, M.P.; Romero F, C.R.

    1996-01-01

    The objective of this paper is to present the basis for shielding calculation used in different situations that could occur during the transport of 131 I utilized in nuclear medicine for diagnostic and therapeutic purposes. The aim of these calculation is to optimize the shielding in order to satisfy the transport of radioactive material. These calculations were proposed for estimated activities around 1,85 GBq (50mCi), 3,7 GBq(100mCi) and 7,4 GBq(200mCi), considering the driver of the cargo company and his assistant as the critical group and the general people considered as effect of collective dose. The population density considered in the models is the one related to Sao Paulo city, because the transport is done by the highway across the city and the radioactive material is distributed from west to north and south, where the airports are located. This area ranges a perimeter of 40 km. For the collective dose calculation, it was considered a population dose of less than 1/100 of the annual limit dose for the public. Our main concern is related to the large volume of radioactive material that is transported per week, specially because 1/3 of this material has activities around 3,7 GBq (100mCi). During the calculations, we have figured out that the activities at the moment of transport are nearly 40% greater than the one related to the calibration date. As for the discrepancy of official alpha value of US$10000/man-Sv and the real value for our country of US$3000/man-Sv,a comparative study was performed. (authors). 3 refs., 2 figs., 2 tabs

  13. Shielding repair and comprehensive safety inspection of the nuclear-powered ship Mutsu

    Takada, Etsuo

    1982-01-01

    Eight years after the radiation leakage accident, the nuclear-powered ship Mutsu returned again to its home port Ominato. During the period, for four years, the n.s. Mutsu was subjected to shielding repair and comprehensive inspection at Sasebo port. In the future, the ship will start on experimental navigation after its functional and power-up tests. The works of shielding repair and the comprehensive inspection with subsequent repair are described in technical aspects. The basic policy of the repair was two points, i.e. the usage of shielding materials excellent in shielding capacity, less in radioactivation and enduring operating temperature, and structural strength resisting ship-hull acceleration, shock and vibration, enabling easy maintenance and inspection. Comprehensive inspection was made on not only machinery integrity but also the design itself. (Mori, K.)

  14. Evaluation of radiation shielding rate of lead aprons in nuclear medicine

    Han, Sang Hyun; Han, Beom Heui; Lee, Sang Ho; Hong, Dong Heui; Kim, Gi Jin

    2017-01-01

    Considering that the X-ray apron used in the department of radiology is also used in the department of nuclear medicine, the study aimed to analyze the shielding rate of the apron according to types of radioisotopes, thus γ ray energy, to investigate the protective effects. The radioisotopes used in the experiment were the top 5 nuclides in usage statistics "9"9"mTc, "1"8F, "1"3"1I, "1"2"3I, and "2"0"1Tl, and the aprons were lead equivalent 0.35 mmPb aprons currently under use in the department of nuclear medicine. As a result of experiments, average shielding rates of aprons were "9"9"mTc 31.59%, "2"0"1Tl 68.42%, and "1"2"3I 76.63%. When using an apron, the shielding rate of "1"3'1I actually resulted in average dose rate increase of 33.72%, and "1"8F showed an average shielding rate of –0.315%, showing there was almost no shielding effect. As a result, the radioisotopes with higher shielding rate of apron was in the descending order of "1"2"3I, "2"0"1Tl, "9"9"mTc, "1"8F, "1"3"1I. Currently, aprons used in the nuclear medicine laboratory are general X-ray aprons, and it is thought that it is not appropriate for nuclear medicine environment that utilizes γ rays. Therefore, development of nuclear medicine exclusive aprons suitable for the characteristics of radioisotopes is required in consideration of effective radiation protection and work efficiency of radiation workers

  15. Evaluation of radiation shielding rate of lead aprons in nuclear medicine

    Han, Sang Hyun; Han, Beom Heui; Lee, Sang Ho [Dept. of Radiological Science, Seonam University, Asan (Korea, Republic of); Hong, Dong Heui [Dept. of Radiological Science, Far East University, Eumseong (Korea, Republic of); Kim, Gi Jin [Dept. of Nuclear Medicine, Konyang University Hospital, Daejeon (Korea, Republic of)

    2017-03-15

    Considering that the X-ray apron used in the department of radiology is also used in the department of nuclear medicine, the study aimed to analyze the shielding rate of the apron according to types of radioisotopes, thus γ ray energy, to investigate the protective effects. The radioisotopes used in the experiment were the top 5 nuclides in usage statistics {sup 99m}Tc, {sup 18}F, {sup 131}I, {sup 123}I, and {sup 201}Tl, and the aprons were lead equivalent 0.35 mmPb aprons currently under use in the department of nuclear medicine. As a result of experiments, average shielding rates of aprons were {sup 99m}Tc 31.59%, {sup 201}Tl 68.42%, and {sup 123}I 76.63%. When using an apron, the shielding rate of {sup 13}'1I actually resulted in average dose rate increase of 33.72%, and {sup 18}F showed an average shielding rate of –0.315%, showing there was almost no shielding effect. As a result, the radioisotopes with higher shielding rate of apron was in the descending order of {sup 123}I, {sup 201}Tl, {sup 99m}Tc, {sup 18}F, {sup 131}I. Currently, aprons used in the nuclear medicine laboratory are general X-ray aprons, and it is thought that it is not appropriate for nuclear medicine environment that utilizes γ rays. Therefore, development of nuclear medicine exclusive aprons suitable for the characteristics of radioisotopes is required in consideration of effective radiation protection and work efficiency of radiation workers.

  16. Radiation shielding analysis of a special linear accelerator for electron beam and X-ray.

    Kang, W G; Pyo, S H; Alkhuraiji, T S; Han, B S; Kang, C M

    2017-01-01

    The King AbdulAziz City for Science & Technology in the Kingdom of Saudi Arabia plans to build a 10 MeV, 15 kW linear accelerator (LINAC) for electron beam and X-ray. The accelerator will be supplied by EB Tech, Republic of Korea, and the design and construction of the accelerator building will be conducted in the cooperation with EB Tech. This report presents the shielding analysis of the accelerator building using the Monte Carlo N-Particle Transport Code (MCNP). In order to improve the accuracy in estimating deep radiation penetration and to reduce computation time, various variance reduction techniques, including the weight window (WW) method, the deterministic transport (DXTRAN) spheres were considered. Radiation levels were estimated at selected locations in the shielding facility running MCNP6 for particle histories up to 1.0×10+8. The final results indicated that the calculated doses at all selected detector locations met the dose requirement of 50 mSv/yr, which is the United State Nuclear Regulatory Commission (U.S. NRC) requirement.

  17. Shielding calculations for ships carrying irradiated nuclear fuel

    Burstall, R.F.; Dean, M.H.

    1983-01-01

    A number of ships have been constructed to carry irradiated fuel from Japan to the UK and France, for reprocessing. About twenty transport flasks may be carried on each voyage. Permanent shielding must be provided on the ships to ensure that no member of the crew receives an annual dose rate greater than a specified limit. As the fuel is of varying type and radiation history, and as flasks of differing designs are used, many calculations are needed. There are a number of difficulties in making shielding calculations for the ships. The geometry is complex, dimensions are large, and considerable air spaces are involved. The paper considers possible methods of calculation. The line-of-sight method is chosen for most of the calculations, for both gamma radiation and neutrons. The basic data which is used in the calculations is described. As the methods of calculation are somewhat approximate, it is necessary to provide confirmation that they are sufficiently accurate. Validation has been provided in two ways. First, measurements have been made on board the ships, and these have been checked against calculation. Second, a simplified model of the flasks and ship has been set up, and calculations checked against more sophisticated methods. Results of the validation checks are presented, and it is shown that adequate accuracy is achieved. (author)

  18. Detection of special nuclear materials with the associate particle technique

    Carasco, Cédric; Deyglun, Clément; Pérot, Bertrand; Eléon, Cyrille; Normand, Stéphane; Sannié, Guillaume; Boudergui, Karim; Corre, Gwenolé; Konzdrasovs, Vladimir; Pras, Philippe

    2013-01-01

    In the frame of the French trans-governmental R and D program against chemical, biological, radiological, nuclear and explosives (CBRN-E) threats, CEA is studying the detection of Special Nuclear Materials (SNM) by neutron interrogation with fast neutrons produced by an associated particle sealed tube neutron generator. The deuterium-tritium fusion reaction produces an alpha particle and a 14 MeV neutron almost back to back, allowing tagging neutron emission both in time and direction with an alpha particle position-sensitive sensor embedded in the generator. Fission prompt neutrons and gamma rays induced by tagged neutrons which are tagged by an alpha particle are detected in coincidence with plastic scintillators. This paper presents numerical simulations performed with the MCNP-PoliMi Monte Carlo computer code and with post processing software developed with the ROOT data analysis package. False coincidences due to neutron and photon scattering between adjacent detectors (cross talk) are filtered out to increase the selectivity between nuclear and benign materials. Accidental coincidences, which are not correlated to an alpha particle, are also taken into account in the numerical model, as well as counting statistics, and the time-energy resolution of the data acquisition system. Such realistic calculations show that relevant quantities of SNM (few kg) can be distinguished from cargo and shielding materials in 10 min acquisitions. First laboratory tests of the system under development in CEA laboratories are also presented.

  19. Early test facilities and analytic methods for radiation shielding: Proceedings

    Ingersoll, D.T.; Ingersoll, J.K.

    1992-11-01

    This report represents a compilation of eight papers presented at the 1992 American Nuclear Society/European Nuclear Society International Meeting. The meeting is of special significance since it commemorates the fiftieth anniversary of the first controlled nuclear chain reaction. The papers contained in this report were presented in a special session organized by the Radiation Protection and Shielding Division in keeping with the historical theme of the meeting. The paper titles are good indicators of their content and are: (1) The origin of radiation shielding research: The Oak Ridge experience, (2) Shielding research at the hanford site, (3) Aircraft shielding experiments at General Dynamics Fort Worth, 1950-1962, (4) Where have the neutrons gone?, a history of the tower shielding facility, (5) History and evolution of buildup factors, (6) Early shielding research at Bettis atomic power laboratory, (7) UK reactor shielding: then and now, (8) A very personal view of the development of radiation shielding theory

  20. Analytical model for relativistic corrections to the nuclear magnetic shielding constant in atoms

    Romero, Rodolfo H. [Facultad de Ciencias Exactas, Universidad Nacional del Nordeste, Avenida Libertad 5500 (3400), Corrientes (Argentina)]. E-mail: rhromero@exa.unne.edu.ar; Gomez, Sergio S. [Facultad de Ciencias Exactas, Universidad Nacional del Nordeste, Avenida Libertad 5500 (3400), Corrientes (Argentina)

    2006-04-24

    We present a simple analytical model for calculating and rationalizing the main relativistic corrections to the nuclear magnetic shielding constant in atoms. It provides good estimates for those corrections and their trends, in reasonable agreement with accurate four-component calculations and perturbation methods. The origin of the effects in deep core atomic orbitals is manifestly shown.

  1. Analytical model for relativistic corrections to the nuclear magnetic shielding constant in atoms

    Romero, Rodolfo H.; Gomez, Sergio S.

    2006-01-01

    We present a simple analytical model for calculating and rationalizing the main relativistic corrections to the nuclear magnetic shielding constant in atoms. It provides good estimates for those corrections and their trends, in reasonable agreement with accurate four-component calculations and perturbation methods. The origin of the effects in deep core atomic orbitals is manifestly shown

  2. Nuclear data needs for fast breeder reactor shielding

    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

  3. Rattling nucleons: New developments in active interrogation of special nuclear material

    Runkle, Robert C.; Chichester, David L.; Thompson, Scott J.

    2012-01-01

    Active interrogation is a vigorous area of research and development due to its promise of offering detection and characterization capabilities of special nuclear material in environments where passive detection fails. The primary value added by active methods is the capability to penetrate shielding - special nuclear material itself, incidental materials, or intentional shielding - and advocates hope that active interrogation will provide a solution to the problem of detecting shielded uranium, which is at present the greatest obstacle to interdiction efforts. The technique also provides a unique benefit for quantifying nuclear material in high background-radiation environments, an area important for nuclear material safeguards and material accountancy. Progress has been made in the field of active interrogation on several fronts, most notably in the arenas of source development, systems integration, and the integration and exploitation of multiple fission and non-fission signatures. But penetration of interrogating radiation often comes at a cost, not only in terms of finance and dose but also in terms of induced backgrounds, system complexity, and extended measurement times (including set up and acquisition). These costs make the calculus for deciding to implement active interrogation more subtle than may be apparent. The purpose of this review is thus to examine existing interrogation methods, compare and contrast their attributes and limitations, and identify missions where active interrogation may hold the most promise.

  4. 10 CFR 74.51 - Nuclear material control and accounting for strategic special nuclear material.

    2010-01-01

    ... 10 Energy 2 2010-01-01 2010-01-01 false Nuclear material control and accounting for strategic special nuclear material. 74.51 Section 74.51 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) MATERIAL CONTROL AND ACCOUNTING OF SPECIAL NUCLEAR MATERIAL Formula Quantities of Strategic Special Nuclear...

  5. Some political issues related to future special nuclear materials production

    Peaslee, A.T. Jr.

    1981-08-01

    The Federal Government must take action to assure the future adequate supply of special nuclear materials for nuclear weapons. Existing statutes permit the construction of advanced defense production reactors and the reprocessing of commercial spent fuel for the production of special materials. Such actions would not only benefit the US nuclear reactor manufacturers, but also the US electric utilities that use nuclear reactors

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

    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

  7. Development of Point Kernel Shielding Analysis Computer Program Implementing Recent Nuclear Data and Graphic User Interfaces

    Kang, Sang Ho; Lee, Seung Gi; Chung, Chan Young; Lee, Choon Sik; Lee, Jai Ki

    2001-01-01

    In order to comply with revised national regulationson radiological protection and to implement recent nuclear data and dose conversion factors, KOPEC developed a new point kernel gamma and beta ray shielding analysis computer program. This new code, named VisualShield, adopted mass attenuation coefficient and buildup factors from recent ANSI/ANS standards and flux-to-dose conversion factors from the International Commission on Radiological Protection (ICRP) Publication 74 for estimation of effective/equivalent dose recommended in ICRP 60. VisualShield utilizes graphical user interfaces and 3-D visualization of the geometric configuration for preparing input data sets and analyzing results, which leads users to error free processing with visual effects. Code validation and data analysis were performed by comparing the results of various calculations to the data outputs of previous programs such as MCNP 4B, ISOSHLD-II, QAD-CGGP, etc

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

    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

  9. Preface: Special Topic on Nuclear Quantum Effects.

    Tuckerman, Mark; Ceperley, David

    2018-03-14

    Although the observable universe strictly obeys the laws of quantum mechanics, in many instances, a classical description that either ignores quantum effects entirely or accounts for them at a very crude level is sufficient to describe a wide variety of phenomena. However, when this approximation breaks down, as is often the case for processes involving light nuclei, a full quantum treatment becomes indispensable. This Special Topic in The Journal of Chemical Physics showcases recent advances in our understanding of nuclear quantum effects in condensed phases as well as novel algorithmic developments and applications that have enhanced the capability to study these effects.

  10. Preface: Special Topic on Nuclear Quantum Effects

    Tuckerman, Mark; Ceperley, David

    2018-03-01

    Although the observable universe strictly obeys the laws of quantum mechanics, in many instances, a classical description that either ignores quantum effects entirely or accounts for them at a very crude level is sufficient to describe a wide variety of phenomena. However, when this approximation breaks down, as is often the case for processes involving light nuclei, a full quantum treatment becomes indispensable. This Special Topic in The Journal of Chemical Physics showcases recent advances in our understanding of nuclear quantum effects in condensed phases as well as novel algorithmic developments and applications that have enhanced the capability to study these effects.

  11. Special nuclear material inventory sampling plans

    Vaccaro, H.; Goldman, A.

    1987-01-01

    Since their introduction in 1942, sampling inspection procedures have been common quality assurance practice. The U.S. Department of Energy (DOE) supports such sampling of special nuclear materials inventories. The DOE Order 5630.7 states, Operations Offices may develop and use statistically valid sampling plans appropriate for their site-specific needs. The benefits for nuclear facilities operations include reduced worker exposure and reduced work load. Improved procedures have been developed for obtaining statistically valid sampling plans that maximize these benefits. The double sampling concept is described and the resulting sample sizes for double sample plans are compared with other plans. An algorithm is given for finding optimal double sampling plans that assist in choosing the appropriate detection and false alarm probabilities for various sampling plans

  12. Calculation of shielding and radiation doses for PET/CT nuclear medicine facility

    Mollah, A.S.; Muraduzzaman, S.M.

    2011-01-01

    Positron emission tomography (PET) is a new modality that is gaining use in nuclear medicine. The use of PET and computed tomography (CT) has grown dramatically. Because of the high energy of the annihilation radiation (511 keV), shielding requirements are an important consideration in the design of a PET or PET/CT imaging facility. The goal of nuclear medicine and PET facility shielding design is to keep doses to workers and the public as low as reasonably achievable (ALARA). Design involves: 1. Calculation of doses to occupants of the facility and adjacent regions based on projected layouts, protocols and workflows, and 2. Reduction of doses to ALARA through adjustment of the aforementioned parameters. The radiological evaluation of a PET/CT facility consists of the assessment of the annual effective dose both to workers occupationally exposed, and to members of the public. This assessment takes into account the radionuclides involved, the facility features, the working procedures, the expected number of patients per year, and so on. The objective of the study was to evaluate shielding requirements for a PET/CT to be installed in the department of nuclear medicine of Bangladesh Atomic Energy Commission (BAEC). Minimizing shielding would result in a possible reduction of structural as well as financial burden. Formulas and attenuation coefficients following the basic AAPM guidelines were used to calculate un-attenuated radiation through shielding materials. Doses to all points on the floor plan are calculated based primarily on the AAPM guidelines and include consideration of broad beam attenuation and radionuclide energy and decay. The analysis presented is useful for both, facility designers and regulators. (author)

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

    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

  14. Bonderenko self-shielded cross sections and multiband parameters derived from the LLL Evaluated-Nuclear-Data Library (ENDL)

    Cullen, D.E.

    1978-01-01

    Bonderenko self-shielded cross sections and multiband parameters from the Lawrence Livermore Laboratory Evaluated-Nuclear-Data Library (ENDL) as of July 4, 1978 are presented. These data include total, elastic, capture, and fission cross sections in the TART 175 group structure. Multiband parameters are listed. Bonderenko self-shielded cross section and the multiband parameters are presented on microfiche

  15. Nuclear Rocket Test Facility Decommissioning Including Controlled Explosive Demolition of a Neutron-Activated Shield Wall

    Michael Kruzic

    2007-01-01

    Located in Area 25 of the Nevada Test Site, the Test Cell A Facility was used in the 1960s for the testing of nuclear rocket engines, as part of the Nuclear Rocket Development Program. The facility was decontaminated and decommissioned (D and D) in 2005 using the Streamlined Approach For Environmental Restoration (SAFER) process, under the Federal Facilities Agreement and Consent Order (FFACO). Utilities and process piping were verified void of contents, hazardous materials were removed, concrete with removable contamination decontaminated, large sections mechanically demolished, and the remaining five-foot, five-inch thick radiologically-activated reinforced concrete shield wall demolished using open-air controlled explosive demolition (CED). CED of the shield wall was closely monitored and resulted in no radiological exposure or atmospheric release

  16. Quantum shielding effects on the Gamow penetration factor for nuclear fusion reaction in quantum plasmas

    Lee, Myoung-Jae; Jung, Young-Dae

    2017-01-01

    The quantum shielding effects on the nuclear fusion reaction process are investigated in quantum plasmas. The closed expression of the classical turning point for the Gamow penetration factor in quantum plasmas is obtained by the Lambert W-function. The closed expressions of the Gamow penetration factor and the cross section for the nuclear fusion reaction in quantum plasmas are obtained as functions of the plasmon energy and the relative kinetic energy by using the effective interaction potential with the WKB analysis. It is shown that the influence of quantum screening suppresses the Sommerfeld reaction factor. It is also shown that the Gamow penetration factor increases with an increase of the plasmon energy. It is also shown that the quantum shielding effect enhances the deuterium formation by the proton-proton reaction in quantum plasmas. In addition, it is found that the energy dependences on the reaction cross section and the Gamow penetration factor are more significant in high plasmon-energy domains.

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

    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)

  18. Evaluation of serpentine ore as a nuclear shielding material using gas chromatographic techniques

    Singh, B.N.; Unnikrishnan, E.K.; Kumar, Sangita D.

    2007-01-01

    Serpentine ore mixed with cement has been recognized as a candidate shielding material for use in nuclear reactors because of its many desirable properties. Therefore the assessment of serpentine ore for release of volatile gases during exposure to elevated temperatures, irradiation and changes in chemical composition, is essential. The present paper deals with the studies on the serpentine ores using gas chromatography and combustion gas chromatographic techniques. (author)

  19. Development of point Kernel radiation shielding analysis computer program implementing recent nuclear data and graphic user interfaces

    Kang, S.; Lee, S.; Chung, C.

    2002-01-01

    There is an increasing demand for safe and efficient use of radiation and radioactive work activity along with shielding analysis as a result the number of nuclear and conventional facilities using radiation or radioisotope rises. Most Korean industries and research institutes including Korea Power Engineering Company (KOPEC) have been using foreign computer programs for radiation shielding analysis. Korean nuclear regulations have introduced new laws regarding the dose limits and radiological guides as prescribed in the ICRP 60. Thus, the radiation facilities should be designed and operated to comply with these new regulations. In addition, the previous point kernel shielding computer code utilizes antiquated nuclear data (mass attenuation coefficient, buildup factor, etc) which were developed in 1950∼1960. Subsequently, the various nuclear data such mass attenuation coefficient, buildup factor, etc. have been updated during the past few decades. KOPEC's strategic directive is to become a self-sufficient and independent nuclear design technology company, thus KOPEC decided to develop a new radiation shielding computer program that included the latest regulatory requirements and updated nuclear data. This new code was designed by KOPEC with developmental cooperation with Hanyang University, Department of Nuclear Engineering. VisualShield is designed with a graphical user interface to allow even users unfamiliar to radiation shielding theory to proficiently prepare input data sets and analyzing output results

  20. Concrete radiation shielding

    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)

  1. Three-dimensional coupled Monte Carlo-discrete ordinates computational scheme for shielding calculations of large and complex nuclear facilities

    Chen, Y.; Fischer, U.

    2005-01-01

    Shielding calculations of advanced nuclear facilities such as accelerator based neutron sources or fusion devices of the tokamak type are complicated due to their complex geometries and their large dimensions, including bulk shields of several meters thickness. While the complexity of the geometry in the shielding calculation can be hardly handled by the discrete ordinates method, the deep penetration of radiation through bulk shields is a severe challenge for the Monte Carlo particle transport technique. This work proposes a dedicated computational scheme for coupled Monte Carlo-Discrete Ordinates transport calculations to handle this kind of shielding problems. The Monte Carlo technique is used to simulate the particle generation and transport in the target region with both complex geometry and reaction physics, and the discrete ordinates method is used to treat the deep penetration problem in the bulk shield. The coupling scheme has been implemented in a program system by loosely integrating the Monte Carlo transport code MCNP, the three-dimensional discrete ordinates code TORT and a newly developed coupling interface program for mapping process. Test calculations were performed with comparison to MCNP solutions. Satisfactory agreements were obtained between these two approaches. The program system has been chosen to treat the complicated shielding problem of the accelerator-based IFMIF neutron source. The successful application demonstrates that coupling scheme with the program system is a useful computational tool for the shielding analysis of complex and large nuclear facilities. (authors)

  2. Shields calculations for teletherapy equipment. Regulatory approach of the National Center of Nuclear Safety

    Fuente P, A. de la; Dumenigo G, C.; Quevedo G, J.R.; Lopez F, Y.

    2006-01-01

    The evaluation of applications of construction licenses for the new services of radiotherapy has occupied a significant space in the activity developed by the National Center of Nuclear Safety (CNSN) in the last 2 years. Presently work the experiences of the authors in the evaluation of the required shield for the local where cobalt therapy equipment and lineal accelerators of medical use are used its are exposed, the practical problems detected are approached during the application of the methodologies recommended in both cases and its are discussed which have been the suppositions of items accepted by the Regulatory Authority for the realization of these shield calculations. The accumulated experience allows to assure that the realistic application of the item data and the rational use of the engineering logic makes possible to design local for radiotherapy equipment that fulfill the established dose restrictions in the in use legislation in Cuba, without it implies an excessive expense of construction materials. (Author)

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

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

    2017-03-15

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

  4. Current status and improvement of the nuclear physics experiment course for speciality of nuclear physics and nuclear technology

    Qu Guopu; Guo Lanying

    1999-01-01

    The author reviews the current status of the nuclear physics experiment course for speciality of nuclear physics and nuclear technology in higher education and expresses author's views on the future improvement of the nuclear physics experiment course

  5. Specialty glass development for radiation shielding windows and nuclear waste immobilization

    Mandal, S.; Ghorui, S.; Roy Chowdhury, A.; Sen, R.; Chakraborty, A.K.; Sen, S.; Maiti, H.S.

    2015-01-01

    The technology of two important varieties of specialty glasses, namely high density Radiation Shielding Window (RSW) glass and specialty glass beads of borosilicate composition have been successfully developed in CGCRI with an aim to meet the countries requirement. Radiation Shielding Windows used in nuclear installations, are viewing devices, which allow direct viewing into radioactive areas while still providing adequate protection to the operating personnel. The glass blocks are stabilized against damage from radiation by introducing cerium in definite proportions. Considering the essentially of developing an indigenous technology to make the country self-sufficient for this critical item, CGCRI has taken up a major programme to develop high lead containing glasses required for RSWs under a MoD with BARC. On the other hand, the specialty glass bead of specific composition and properties is a critical material required for management of radioactive waste in a closed nuclear fuel cycle that is followed by India. During reprocessing of the spent nuclear fuel, high level radio-active liquid waste (HLW) is produced containing unwanted radio isotopes some of which remain radioactive for thousands of years. The need is to immobilize them within a molecular structure so that they will not come out and be released to the ambience and thereby needs to be resolved if nuclear power is to make a significant contribution to the country's power requirement. Borosilicate glass has emerged as the material of choice for immobilization due to its unique random network structure

  6. Relativistic heavy-atom effects on heavy-atom nuclear shieldings

    Lantto, Perttu; Romero, Rodolfo H.; Gómez, Sergio S.; Aucar, Gustavo A.; Vaara, Juha

    2006-11-01

    The principal relativistic heavy-atom effects on the nuclear magnetic resonance (NMR) shielding tensor of the heavy atom itself (HAHA effects) are calculated using ab initio methods at the level of the Breit-Pauli Hamiltonian. This is the first systematic study of the main HAHA effects on nuclear shielding and chemical shift by perturbational relativistic approach. The dependence of the HAHA effects on the chemical environment of the heavy atom is investigated for the closed-shell X2+, X4+, XH2, and XH3- (X =Si-Pb) as well as X3+, XH3, and XF3 (X =P-Bi) systems. Fully relativistic Dirac-Hartree-Fock calculations are carried out for comparison. It is necessary in the Breit-Pauli approach to include the second-order magnetic-field-dependent spin-orbit (SO) shielding contribution as it is the larger SO term in XH3-, XH3, and XF3, and is equally large in XH2 as the conventional, third-order field-independent spin-orbit contribution. Considering the chemical shift, the third-order SO mechanism contributes two-thirds of the difference of ˜1500ppm between BiH3 and BiF3. The second-order SO mechanism and the numerically largest relativistic effect, which arises from the cross-term contribution of the Fermi contact hyperfine interaction and the relativistically modified spin-Zeeman interaction (FC/SZ-KE), are isotropic and practically independent of electron correlation effects as well as the chemical environment of the heavy atom. The third-order SO terms depend on these factors and contribute both to heavy-atom shielding anisotropy and NMR chemical shifts. While a qualitative picture of heavy-atom chemical shifts is already obtained at the nonrelativistic level of theory, reliable shifts may be expected after including the third-order SO contributions only, especially when calculations are carried out at correlated level. The FC/SZ-KE contribution to shielding is almost completely produced in the s orbitals of the heavy atom, with values diminishing with the principal

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

    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. Evaluation of nuclear data for radiation shielding by model calculations and international co-operation aspects

    Canetta, E.; Maino, G.; Menapace, E.

    2001-01-01

    The matter is reviewed, also following previous discussions at ICRS-9, concerning evaluation and related theoretical activities on nuclear data for radiation shielding within the framework of international co-operation initiatives, according to recognised needs and priorities. Both cross-section data.- for reactions induced by neutrons and photons - and nuclear structure data have been considered. In this context, main contributions and typical results are presented from theoretical and evaluation activities at the ENEA Applied Physics Division, especially concerning neutron induced reaction data up to 20 MeV and photonuclear reaction data such as photon absorption and (gamma,n) cross-sections. Relevant aspects of algebraic nuclear models and of evaporation and pre-equilibrium models are discussed. (authors)

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

    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

  10. Considerations about the licensing process of special nuclear industrial facilities

    Talarico, M.A., E-mail: talaricomarco@hotmail.com [Marinha do Brasil, Rio de Janeiro, RJ (Brazil). Coordenacao do Porgrama de Submarino com Propulsao Nuclear; Melo, P.F. Frutuoso e [Coordenacao dos Programas de Pos-Graduacao em Engenharia (COPPE/UFRJ), Rio de Janeiro, RJ (Brazil). Programa de Engenharia Nuclear

    2015-07-01

    This paper brings a discussion about the challenges involved in the development of a new kind of nuclear facility in Brazil, a naval base for nuclear submarines, with attention to the licensing process and considerations about the risk-informed decision making application to the licensing process. Initially, a model of such a naval base, called in this work, special industrial facility, is proposed, with its systems and respective sets of basic requirements, in order to make it possible the accomplishment of the special industrial facility support function to the nuclear submarine. A discussion about current challenges to overcome in this project is presented: the challenges due to the new characteristics of this type of nuclear facility; existence of several interfaces between the special industrial facilities systems and nuclear submarine systems in design activities; lack of specific regulation in Brazil to allow the licensing process of special industrial facilities by the nuclear safety authority; and comments about the lack of information from reference nuclear facilities, as is the case with nuclear power reactors (for example, the German Grafenrheinfeld nuclear plant is the reference plant for the Brazilian Angra 2 nuclear plant). Finally, in view of these challenges, an analysis method of special industrial facility operational scenarios to assist the licensing process is proposed. Also, considerations about the application of risk-informed decision making to the special industrial facility activity and licensing process in Brazil are presented. (author)

  11. Considerations about the licensing process of special nuclear industrial facilities

    Talarico, M.A.; Melo, P.F. Frutuoso e

    2015-01-01

    This paper brings a discussion about the challenges involved in the development of a new kind of nuclear facility in Brazil, a naval base for nuclear submarines, with attention to the licensing process and considerations about the risk-informed decision making application to the licensing process. Initially, a model of such a naval base, called in this work, special industrial facility, is proposed, with its systems and respective sets of basic requirements, in order to make it possible the accomplishment of the special industrial facility support function to the nuclear submarine. A discussion about current challenges to overcome in this project is presented: the challenges due to the new characteristics of this type of nuclear facility; existence of several interfaces between the special industrial facilities systems and nuclear submarine systems in design activities; lack of specific regulation in Brazil to allow the licensing process of special industrial facilities by the nuclear safety authority; and comments about the lack of information from reference nuclear facilities, as is the case with nuclear power reactors (for example, the German Grafenrheinfeld nuclear plant is the reference plant for the Brazilian Angra 2 nuclear plant). Finally, in view of these challenges, an analysis method of special industrial facility operational scenarios to assist the licensing process is proposed. Also, considerations about the application of risk-informed decision making to the special industrial facility activity and licensing process in Brazil are presented. (author)

  12. Estimation of Shielding Thickness for a Prototype Department of Energy National Spent Nuclear Fuel Program Transport Cask

    SANCHEZ,LAWRENCE C.; MCCONNELL,PAUL E.

    2000-07-01

    Preliminary shielding calculations were performed for a prototype National Spent Nuclear Fuel Program (NSNFP) transport cask. This analysis is intended for use in the selection of cask shield material type and preliminary estimate of shielding thickness. The radiation source term was modeled as cobalt-60 with radiation exposure strength of 100,000 R/hr. Cobalt-60 was chosen as a surrogate source because it simultaneous emits two high-energy gammas, 1.17 MeV and 1.33 MeV. This gamma spectrum is considered to be large enough that it will upper bound the spectra of all the various spent nuclear fuels types currently expected to be shipped within the prototype cask. Point-kernel shielding calculations were performed for a wide range of shielding thickness of lead and depleted uranium material. The computational results were compared to three shielding limits: 200 mrem/hr dose rate limit at the cask surface, 50 mR/hr exposure rate limit at one meter from the cask surface, and 10 mrem/hr limit dose rate at two meters from the cask surface. The results obtained in this study indicated that a shielding thickness of 13 cm is required for depleted uranium and 21 cm for lead in order to satisfy all three shielding requirements without taking credit for stainless steel liners. The system analysis also indicated that required shielding thicknesses are strongly dependent upon the gamma energy spectrum from the radiation source term. This later finding means that shielding material thickness, and hence cask weight, can be significantly reduced if the radiation source term can be shown to have a softer, lower energy, gamma energy spectrum than that due to cobalt-60.

  13. The design study of the JT-60SU device. No.8. Nuclear shielding and safety design

    Miya, Naoyuki; Kikuchi, Mitsuru; Ushigusa, Kenkichi [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment] [and others

    1998-03-01

    Results of nuclear shielding design study and safety analysis for the steady-state tokamak device JT-60SU are described. D-T operation (option) for two years is adopted in addition to ten years operation using deuterium. Design work has been done in accordance with general laws for radioisotopes handling in Japan as a guideline of safety evaluation, which is applied to the operation of present JT-60U device. Optimization of the shielding design for the device structure including vacuum vessel has been presented to meet with allowable limits of biological shielding determined in advance. It is shown that JT-60SU can be operated safely in the present JT-60 experimental building. It is planed to use 100g/year of tritium in D-T operation phase. A concept of multiple -barrier system is applied to the facility design to prevent propagation of tritium, in which the torus hall and the tritium removal room provide the tertiary confinement. From the design of atmosphere detritiation system for accidental tritium release, it is shown that tritium concentration level can be reduced to the allowable level after two weeks with reasonable compact size components. Safety assessment related to activation of coolant/air, and atmospheric tritium effluents are discussed. (author)

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

    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

  15. Can active proton interrogation find shielded nuclear threats at human-safe radiation levels?

    Liew, Seth Van, E-mail: vanliew@gmail.com

    2017-05-21

    A new method of low-dose proton radiography is presented. The system is composed of an 800 MeV proton source, bending magnets, and compact detectors, and is designed for drive-through cargo scanning. The system has been simulated using GEANT4. Material identification algorithms and pixel sorting methods are presented that allow the system to perform imaging at doses low enough to scan passenger vehicles and people. Results are presented on imaging efficacy of various materials and cluttered cargoes. The identification of shielded nuclear materials at human-safe doses has been demonstrated.

  16. IAEA to hold special session on nuclear terrorism

    2001-01-01

    Experts from around the world are meeting at the IAEA this week for an international symposium on nuclear safeguards, verification, and security. A special session on 2 November focuses on the issue of combating nuclear terrorism. The Special Session, which will bring together experts on nuclear terrorism from around the world, will deal with the following issues: The Psychology of terrorism; Intelligence, police and border protection; Guarding nuclear reactors and material from terrorists and thieves; The threat of nuclear terrorism: Nuclear weapons or other nuclear explosive devices; The threat of nuclear terrorism: Intentional dispersal of radioactive material - Sabotage of fixed installations or transport systems; The Legal Framework: Treaties and Conventions, Laws; Regulations and Codes of Practice; IAEA Nuclear Security and Safety Programmes

  17. High-fidelity MCNP modeling of a D-T neutron generator for active interrogation of special nuclear material

    Katalenich, Jeff; Flaska, Marek; Pozzi, Sara A.; Hartman, Michael R.

    2011-01-01

    Fast and robust methods for interrogation of special nuclear material (SNM) are of interest to many agencies and institutions in the United States. It is well known that passive interrogation methods are typically sufficient for plutonium identification because of a relatively high neutron production rate from 240 Pu . On the other hand, identification of shielded uranium requires active methods using neutron or photon sources . Deuterium-deuterium (2.45 MeV) and deuterium-tritium (14.1 MeV) neutron-generator sources have been previously tested and proven to be relatively reliable instruments for active interrogation of nuclear materials . In addition, the newest generators of this type are small enough for applications requiring portable interrogation systems. Active interrogation techniques using high-energy neutrons are being investigated as a method to detect hidden SNM in shielded containers . Due to the thickness of some containers, penetrating radiation such as high-energy neutrons can provide a potential means of probing shielded SNM. In an effort to develop the capability to assess the signal seen from various forms of shielded nuclear materials, University of Michigan Neutron Science Laboratory's D-T neutron generator and its shielding were accurately modeled in MCNP. The generator, while operating at nominal power, produces approximately 1x10 10 neutrons/s, a source intensity which requires a large amount of shielding to minimize the dose rates around the generator. For this reason, the existing shielding completely encompasses the generator and does not include beam ports. Therefore, several MCNP simulations were performed to estimate the yield of uncollided 14.1-MeV neutrons from the generator for active interrogation experiments. Beam port diameters of 5, 10, 15, 20, and 25 cm were modeled to assess the resulting neutron fluxes. The neutron flux outside the beam ports was estimated to be approximately 2x10 4 n/cm 2 s.

  18. Nuclear Rocket Facility Decommissioning Project: Controlled Explosive Demolition of Neutron-Activated Shield Wall

    Michael R, Kruzic

    2008-01-01

    Located in Area 25 of the Nevada Test Site (NTS), the Test Cell A (TCA) Facility (Figure 1) was used in the early to mid-1960s for testing of nuclear rocket engines, as part of the Nuclear Rocket Development Program, to further space travel. Nuclear rocket testing resulted in the activation of materials around the reactors and the release of fission products and fuel particles. The TCA facility, known as Corrective Action Unit 115, was decontaminated and decommissioned (D and D) from December 2004 to July 2005 using the Streamlined Approach for Environmental Restoration (SAFER) process, under the Federal Facility Agreement and Consent Order. The SAFER process allows environmental remediation and facility closure activities (i.e., decommissioning) to occur simultaneously, provided technical decisions are made by an experienced decision maker within the site conceptual site model. Facility closure involved a seven-step decommissioning strategy. First, preliminary investigation activities were performed, including review of process knowledge documentation, targeted facility radiological and hazardous material surveys, concrete core drilling and analysis, shield wall radiological characterization, and discrete sampling, which proved to be very useful and cost-effective in subsequent decommissioning planning and execution and worker safety. Second, site setup and mobilization of equipment and personnel were completed. Third, early removal of hazardous materials, including asbestos, lead, cadmium, and oil, was performed ensuring worker safety during more invasive demolition activities. Process piping was to be verified void of contents. Electrical systems were de-energized and other systems were rendered free of residual energy. Fourth, areas of high radiological contamination were decontaminated using multiple methods. Contamination levels varied across the facility. Fixed beta/gamma contamination levels ranged up to 2 million disintegrations per minute (dpm)/100

  19. Nuclear Rocket Facility Decommissioning Project: Controlled Explosive Demolition of Neutron-Activated Shield Wall

    Michael R. Kruzic

    2008-06-01

    Located in Area 25 of the Nevada Test Site (NTS), the Test Cell A (TCA) Facility (Figure 1) was used in the early to mid-1960s for testing of nuclear rocket engines, as part of the Nuclear Rocket Development Program, to further space travel. Nuclear rocket testing resulted in the activation of materials around the reactors and the release of fission products and fuel particles. The TCA facility, known as Corrective Action Unit 115, was decontaminated and decommissioned (D&D) from December 2004 to July 2005 using the Streamlined Approach for Environmental Restoration (SAFER) process, under the Federal Facility Agreement and Consent Order. The SAFER process allows environmental remediation and facility closure activities (i.e., decommissioning) to occur simultaneously, provided technical decisions are made by an experienced decision maker within the site conceptual site model. Facility closure involved a seven-step decommissioning strategy. First, preliminary investigation activities were performed, including review of process knowledge documentation, targeted facility radiological and hazardous material surveys, concrete core drilling and analysis, shield wall radiological characterization, and discrete sampling, which proved to be very useful and cost-effective in subsequent decommissioning planning and execution and worker safety. Second, site setup and mobilization of equipment and personnel were completed. Third, early removal of hazardous materials, including asbestos, lead, cadmium, and oil, was performed ensuring worker safety during more invasive demolition activities. Process piping was to be verified void of contents. Electrical systems were de-energized and other systems were rendered free of residual energy. Fourth, areas of high radiological contamination were decontaminated using multiple methods. Contamination levels varied across the facility. Fixed beta/gamma contamination levels ranged up to 2 million disintegrations per minute (dpm)/100

  20. Do nuclear engineering educators have a special responsibility

    Weinberg, A.M.

    1977-01-01

    Each 1000 MW(e) reactor in equilibrium contains 15 x 10 9 Ci of radioactivity. To handle this material safety requires an extremely high level of expertise and commitment - in many respects, an expertise that goes beyond what is demanded of any other technology. If one grants that nuclear engineering is more demanding than other engineering because the price of failure is greater, one must ask how can we inculcate into the coming generations of nuclear engineers a full sense of the responsibility they bear in practising their profession. Clearly a first requirement is that all elements of the nuclear community -utility executives, equipment engineers, operating engineers, nuclear engineers, administrators - must recognize and accept the idea that nuclear energy is something special, and that therefore its practitioners must be special. This sense must be instilled into young nuclear engineers during their education. A special responsibility therefore devolves upon nuclear engineering educators: first, to recognize the special character of their profession, and second, to convey this sense to their students. The possibility of institutionalizing this sense of responsibility by establishing a nuclear Hippocratic Oath or special canon of ethics for nuclear engineers ought to be discussed within the nuclear community. (author)

  1. Special monitoring in nuclear medicine; Monitoreo especial en medicina nuclear

    Beltran, C.C.; Puerta, J.A.; Morales, J. [Asociacion Colombiana de Proteccion Radiologica (Colombia)]. e-mail: ccbeltra@gmail.com

    2006-07-01

    Colombia counts with around 56 centers of Nuclear Medicine, 70 Nuclear Doctors and more of 100 Technologists in this area. The radioisotopes more used are the {sup 131} I and the {sup 99m} Tc. The radiological surveillance singular in the country is carried out for external dosimetry, being the surveillance by incorporation of radioactive materials very sporadic in our media. Given the necessity to implement monitoring programs in the incorporation of radionuclides of the occupationally exposed personnel, in the routine practice them routine of Nuclear Medicine, it was implemented a pilot program of Special Monitoring with two centers of importance in the city of Medellin. This program it was carried out with the purpose of educating, to stimulate and to establish a program of reference monitoring with base in the National Program of Monitoring in the radionuclides Incorporation that serves like base for its application at level of all the services of Nuclear Medicine in the country. This monitoring type was carried out with the purpose of obtaining information on the work routine in these centers, form of manipulation and dosage of the radionuclides, as well as the administration to the patient. The application of the program was carried out to define the frequency of Monitoring and analysis technique for the implementation of a program of routine monitoring, following the recommendations of the International Commission of Radiological Protection. For their application methods of activity evaluation were used in urine and in 7 workers thyroid, of those which only two deserve an analysis because they presented important activities. The measures were carried out during one month, every day by means in urine samples and to the most critic case is practiced two thyroid measures, one in the middle of the period and another when concluding the monitoring. To the other guy is practiced an activity count in thyroid when concluding the monitoring period. The obtained

  2. Modifier cation effects on (29)Si nuclear shielding anisotropies in silicate glasses.

    Baltisberger, Jay H; Florian, Pierre; Keeler, Eric G; Phyo, Pyae A; Sanders, Kevin J; Grandinetti, Philip J

    2016-07-01

    We have examined variations in the (29)Si nuclear shielding tensor parameters of SiO4 tetrahedra in a series of seven alkali and alkaline earth silicate glass compositions, Cs2O·4.81 SiO2, Rb2O·3.96 SiO2, Rb2O·2.25 SiO2, K2O·4.48 SiO2, Na2O·4.74 SiO2, BaO·2.64 SiO2, and SrO·2.36 SiO2, using natural abundance (29)Si two-dimensional magic-angle flipping (MAF) experiments. Our analyses of these 2D spectra reveal a linear dependence of the (29)Si nuclear shielding anisotropy of Q((3)) sites on the Si-non-bridging oxygen bond length, which in turn depends on the cation potential and coordination of modifier cations to the non-bridging oxygen. We also demonstrate how a combination of Cu(2+) as a paramagnetic dopant combined with echo train acquisition can reduce the total experiment time of (29)Si 2D NMR measurements by two orders of magnitude, enabling higher throughput 2D NMR studies of glass structure. Copyright © 2016 Elsevier Inc. All rights reserved.

  3. Modifier cation effects on 29Si nuclear shielding anisotropies in silicate glasses

    Baltisberger, Jay H.; Florian, Pierre; Keeler, Eric G.; Phyo, Pyae A.; Sanders, Kevin J.; Grandinetti, Philip J.

    2016-07-01

    We have examined variations in the 29Si nuclear shielding tensor parameters of SiO4 tetrahedra in a series of seven alkali and alkaline earth silicate glass compositions, Cs2O · 4.81 SiO2, Rb2O · 3.96 SiO2, Rb2O · 2.25 SiO2, K2O · 4.48 SiO2, Na2O · 4.74 SiO2, BaO · 2.64 SiO2, and SrO · 2.36 SiO2, using natural abundance 29Si two-dimensional magic-angle flipping (MAF) experiments. Our analyses of these 2D spectra reveal a linear dependence of the 29Si nuclear shielding anisotropy of Q(3) sites on the Si-non-bridging oxygen bond length, which in turn depends on the cation potential and coordination of modifier cations to the non-bridging oxygen. We also demonstrate how a combination of Cu2+ as a paramagnetic dopant combined with echo train acquisition can reduce the total experiment time of 29Si 2D NMR measurements by two orders of magnitude, enabling higher throughput 2D NMR studies of glass structure.

  4. 78 FR 38739 - Special Nuclear Material Control and Accounting Systems for Nuclear Power Plants

    2013-06-27

    ... Systems for Nuclear Power Plants AGENCY: Nuclear Regulatory Commission. ACTION: Regulatory guide; issuance... Guide (RG) 5.29, ``Special Nuclear Material Control and Accounting Systems for Nuclear Power Plants... material control and accounting. This guide applies to all nuclear power plants. ADDRESSES: Please refer to...

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

    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

  6. A special radiation shielding for the radiotherapy of a pregnant patient; Eine Spezialabschirmung fuer die Strahlentherapie von Schwangeren

    Buchgeister, M.; Mondry, A. [Bereich Medizinische Physik, Universitaetsklinik fuer Radioonkologie, Tuebingen (Germany); Spillner, P.; Paulsen, F.; Belka, C.; Bamberg, M. [Strahlentherapie, Universitaetsklinik fuer Radioonkologie, Tuebingen (Germany)

    2008-02-15

    Purpose: measurement of the radiation dose of different radiotherapy techniques with a phantom at the position of the uterus of a pregnant patient with and without a special radiation shielding. Material and methods: a special radiation shielding for the radiation therapy of a pregnant patient was constructed in the shape of a tunnel over the abdomen from 1 cm thick lead to reduce scatter radiation to the uterus and the fetus therein. The reduction of the scatter radiation to the lower abdomen was measured for three typical cases (cerebrum with lateral opposed fields, 6-MV photons; tangential fields for mamma irradiation, 6 MV; and anteroposterior-posteroanterior [AP-PA] opposed fields of 15 MV for treatment of the mediastinum) at an anthropomorphic Alderson phantom with thermoluminescence dosimeters in different depths and with an ionization chamber positioned in an RW3 solid water phantom. In the case of lateral opposed fields, a movable lead wall was additionally positioned next to the accelerator's head to reduce the scatter radiation from this source. Results: depending on the geometry of the radiation fields and on the photon energy, a reduction of the dose to the lower abdomen averaged over the depths of 6, 9, and 12 cm from 16% (15 MV, mediastinum case) to 51% (6 MV, cerebrum with additional lead wall) was achieved. The absolute scattered dose with shielding in place for a 2-Gy fraction dose results to 3.85 mGy and 0.27 mGy, respectively. Conclusion: national and international recommendations on the radiation dose to a fetus of a pregnant patient state limits of 200 mSv (DGMP report no. 7) and 100 mGy (ICRP 84), below which an abortion should not be considered and above which an indication for termination of the pregnancy could be given, respectively. The dose to the fetus can be kept below these limits with the shielding described in this work. Therefore, a radiation therapy of a pregnant patient is possible when these special precautions to reduce

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

    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.

  8. Studies for calculations of the thicknesses of shielding necessary for implementation of a nuclear medicine service with PET-CT

    Nascimento, Erika M.; Lopes Filho, Ferdinand J.; Souza, Milena Thays B. de; Aragao Filho, Geraldo L.

    2013-01-01

    The thickness of shielding for controlled and surrounding areas must be considered in terms of the PET-CT equipment installation. However, for a project to install a PET-CT requires the participation of technologists, engineers and architects so that together we can meet the requirements of radiological protection at low cost, enabling its installation and maintenance in nuclear medical centers and hospitals. The objective of this paper is to describe the calculations needed to shield a nuclear medicine center that will install a PET-CT equipment and describe how the participation of other professionals can contribute to a lower cost

  9. Report on strategic special nuclear material inventory differences

    1977-08-01

    Information concerning accounting for significant quantities of strategic special nuclear material (SSNM) in ERDA facilities is reported. Inventory difference data are provided for fiscal year 1976 for ERDA and ERDA contractor facilities that possessed significant quantities of SSNM

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

    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.)

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

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

    2016-09-01

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

  12. Oscillatory vapour shielding of liquid metal walls in nuclear fusion devices.

    van Eden, G G; Kvon, V; van de Sanden, M C M; Morgan, T W

    2017-08-04

    Providing an efficacious plasma facing surface between the extreme plasma heat exhaust and the structural materials of nuclear fusion devices is a major challenge on the road to electricity production by fusion power plants. The performance of solid plasma facing surfaces may become critically reduced over time due to progressing damage accumulation. Liquid metals, however, are now gaining interest in solving the challenge of extreme heat flux hitting the reactor walls. A key advantage of liquid metals is the use of vapour shielding to reduce the plasma exhaust. Here we demonstrate that this phenomenon is oscillatory by nature. The dynamics of a Sn vapour cloud are investigated by exposing liquid Sn targets to H and He plasmas at heat fluxes greater than 5 MW m -2 . The observations indicate the presence of a dynamic equilibrium between the plasma and liquid target ruled by recombinatory processes in the plasma, leading to an approximately stable surface temperature.Vapour shielding is one of the interesting mechanisms for reducing the heat load to plasma facing components in fusion reactors. Here the authors report on the observation of a dynamic equilibrium between the plasma and the divertor liquid Sn surface leading to an overall stable surface temperature.

  13. Ring thermal shield piping modification at Pickering Nuclear Generating Station 'A' Unit 1

    Brown, R.; Cobanoglu, M.M.

    1995-01-01

    Each of the four Pickering Nuclear Generating Station A (PNGSA) CANDU units was constructed with its reactor and dump tank surrounded by a concrete Calandria Vault (CV). The Ring Thermal Shield (RTS) system at PNGSA units is a water cooled structure with internal cooling channels with the purpose of attenuating excessive heat flux from the calandria shell to the end shield rings and adjoining concrete (Figure 1). In newer CANDU units the reactor calandria vessel is surrounded by a large water filled shield tank which eliminates the requirement for the RTS system. The RTS structures are situated in the space between the calandria and the vault walls. Each RTS is assembled from eight flat sided carbon steel segments, tilted towards the calandria and supported from the end shield rings. Cooling water to the RTS is supplied by carbon steel cooling pipes with a portion of the pipe run embedded in the vault walls. Flow through each RTS is divided into two independent circuits, having an inlet and an outlet cooling line. There are four locations of RTS inlet and outlet cooling lines. The inlet lines are located at the bottom and the outlet lines at the top of the RTS. The 'L' shaped section of RTS inlet and outlet cooling lines, from the RTS waterbox to the start of embedded portion at the concrete wall, had become defective due to corrosion induced by excessive Moisture levels in the calandria vaults. An on-line leak sealing capability was developed and placed in service in all four PNGSA units. However, a leak found during the 1994 Unit 1 outage was too large,to seal with the current capability, forcing Ontario Hydro (OH) to develop a method to replace the corroded pipes. The repair project was subject to some lofty performance targets. All tools had to be able to withstand dose rates of up to 3000 Rem/hour. These tools, along with procedures and personnel had to successfully repair the RTS system within 6 months otherwise a costly outage extension would result. This

  14. Shielding benchmark problems

    Tanaka, Shun-ichi; Sasamoto, Nobuo; Oka, Yoshiaki; Kawai, Masayoshi; Nakazawa, Masaharu.

    1978-09-01

    Shielding benchmark problems were prepared by the Working Group of Assessment of Shielding Experiments in the Research Comittee on Shielding Design of the Atomic Energy Society of Japan, and compiled by the Shielding Laboratory of Japan Atomic Energy Research Institute. Twenty-one kinds of shielding benchmark problems are presented for evaluating the calculational algorithm and the accuracy of computer codes based on the discrete ordinates method and the Monte Carlo method and for evaluating the nuclear data used in the codes. (author)

  15. Investigation of Active Interrogation Techniques to Detect Special Nuclear Material in Maritime Environments

    Miller, Thomas Martin; Patton, Bruce W.

    2010-01-01

    The detection and interdiction of special nuclear material (SNM) is still a high-priority focus area for many organizations around the world. One method that is commonly considered a leading candidate in the detection of SNM is active interrogation (AI). AI is different from its close relative, passive interrogation, in that an active source is used to enhance or create a detectable signal (usually fission) from SNM, particularly in shielded scenarios or scenarios where the SNM has a low activity. The use of AI thus makes the detection of SNM easier or, in some scenarios, even enables previously impossible detection. In this work the signal from prompt neutrons and photons as well as delayed neutrons and photons will be combined, as is typically done in AI. In previous work AI has been evaluated experimentally and computationally. However, for the purposes of this work, past scenarios are considered lightly shielded and tightly coupled spatially. At most, the previous work interrogated the contents of one standard cargo container (2.44 x 2.60 x 6.10 m) and the source and detector were both within a few meters of the object being interrogated. A few examples of this type of previous work can be found in references 1 and 2. Obviously, more heavily shielded AI scenarios will require larger source intensities, larger detector surface areas (larger detectors or more detectors), greater detector efficiencies, longer count times, or some combination of these.

  16. Special nuclear material information, security classification guidance. Instruction

    Flickinger, A.

    1982-01-01

    The Instruction reissues DoD Instruction 5210.67, July 5, 1979, and provides security classification guidance for information concerning significant quantities of special nuclear material, other than that contained in nuclear weapons and that used in the production of energy in the reactor plant of nuclear-powered ships. Security classification guidance for these data in the latter two applications is contained in Joint DoE/DoD Nuclear Weapons Classification Guide and Joint DoE/DoD Classification Guide for the Naval Nuclear Propulsion Program

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

    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

  18. Special requirements for bolting material for nuclear and other special applications

    Anon.

    1981-01-01

    A specification that provides special requirements for bolting material for use in nuclear and other special applications is presented. The requirements of the specification are supplemental to the requirements of the basic material specifications and they include tempering, welding, elimination of surface defects, certification and identification, quality assurance and various examination methods

  19. Radiation shield ring assembly and method of disassembling components of a nuclear steam generator using such assembly

    Meuschke, R.E.; Wolfe, D.L.

    1982-01-01

    This invention relates to an apparatus and a method for cutting, within a shielding confinement, the irradiated components of a nuclear steam generator to reduce such components to a size to permit their subsequent removal from the containment structure of the generator

  20. 10 CFR 74.41 - Nuclear material control and accounting for special nuclear material of moderate strategic...

    2010-01-01

    ... 10 Energy 2 2010-01-01 2010-01-01 false Nuclear material control and accounting for special nuclear material of moderate strategic significance. 74.41 Section 74.41 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) MATERIAL CONTROL AND ACCOUNTING OF SPECIAL NUCLEAR MATERIAL Special Nuclear Material...

  1. Annual report on strategic special nuclear material inventory differences

    1991-01-01

    This report of unclassified Inventory Difference (ID's) covers the twelve months from April 1, 1989 through March 31, 1990 for all key Department of Energy (DOE) and DOE contractor operated facilities possessing strategic special nuclear material. Classified information is not included in this report. This classified information includes data for the Rocky Flats and Y-12 nuclear weapons production facilities or facilities under ID investigation. However, classified ID data from such facilities receive the same scrutiny and analyses was the included data. The data in this report have been prepared and reviewed by DOE contractors, field offices, and Headquarters. When necessary, special investigations have been performed in addition to these reviews: This ID data, explanations, reviews, and any additional special investigations, together with the absence of physical indications of any theft attempt, support a finding that no theft or diversion of significant quantities of strategic special nuclear material has occurred in DOE facilities during the twelve-month period covered by this report

  2. REACTOR SHIELD

    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.

  3. Special nuclear material inventory sampling plans

    Vaccaro, H.S.; Goldman, A.S.

    1987-01-01

    This paper presents improved procedures for obtaining statistically valid sampling plans for nuclear facilities. The double sampling concept and methods for developing optimal double sampling plans are described. An algorithm is described that is satisfactory for finding optimal double sampling plans and choosing appropriate detection and false alarm probabilities

  4. About the Scythian Shields

    About the Scythian Shields

    2017-10-01

    Full Text Available Shields played major role in the armament system of the Scythians. Made from organic materials, they are poorly traced on the materials of archaeological excavations. Besides, scaly surface of shields was often perceived in practice as the remnants of the scaly armor. E. V. Chernenko was able to discern the difference between shields’ scaly plates and armor scales. The top edge of the scales was bent inwards, and shield plates had a wire fixation. These observations let significantly increase the number of shields, found in the burial complexes of the Scythians. The comparison of archaeological materials and the images of Scythian warriors allow distinguishing the main forms of Scythian shields. All shields are divided into fencing shields and cover shields. The fencing shields include round wooden shields, reinforced with bronze sheet, and round moon-shaped shields with a notch at the top, with a metal scaly surface. They came to the Scythians under the Greek influence and are known in the monuments of the 4th century BC. Oval shields with scaly surface (back cover shields were used by the Scythian cavalry. They protected the rider in case of frontal attack, and moved back in case of maneuver or closein fighting. Scythian battle tactics were based on rapid approaching the enemy and throwing spears and further rapid withdrawal. Spears stuck in the shields of enemies, forcing them to drop the shields, uncover, and in this stage of the battle the archers attacked the disorganized ranks of the enemy. That was followed by the stage of close fight. Oval form of a wooden shield with leather covering was used by the Scythian infantry and spearmen. Rectangular shields, including wooden shields and the shields pleached from rods, represented a special category. The top of such shield was made of wood, and a pleached pad on leather basis was attached to it. This shield could be a reliable protection from arrows, but it could not protect against javelins

  5. Special feature article-very urgent nuclear energy personnel training

    Saito, Shinzo; Tsujikura, Yonezo; Kawahara, Akira

    2007-01-01

    Securing human resources is important for the sustainable development of research, development and utilization of nuclear energy. However, concerns have been raised over the maintenance of human resources due to the decline of public and private investment in research and development of nuclear energy in recent years. To that end, it is essential for the workplace in the field of nuclear energy to be engaging. This special feature article introduced the Government's fund program supporting universities and vocational colleges to develop human resources in the area of nuclear energy. Electric utilities, nuclear industries, nuclear safety regulators and related academia presented respective present status and issues of nuclear energy personnel training with some expectations to the program to secure human resources with professional qualifications. (T. Tanaka)

  6. Construction of special structures for nuclear power projects

    Raghavan, N.

    2003-01-01

    Construction is a very important stage in the course of realization of Nuclear Power Projects and as much care has be devoted to this stage as to the planning and engineering stages. While the setting up of nuclear power projects used to take over seven years in the past, the time period has now been considerably reduced to about five years with advancements in construction engineering, project management and design techniques, on the basis of new initiatives from the owner agency, Nuclear Power Corporation of India. In this article, the constructional aspects of the specialized structures for nuclear power generation are looked into. (author)

  7. Radiation shielding and criticality safety assessment for KN-12 spent nuclear fuel transport cask

    Kim, Jong Kyung; Shin, Chang Ho; Kim, Gi Hwan [Hanyang Univ., Seoul (Korea, Republic of)

    2001-08-15

    Because SNFs involve TRU (Transuranium), fission products, and fissile materials, they are highly radioactive and also have a possibility to be critical. Therefore, radiation shielding and criticality safety for transport casks containing the SNFs should be guaranteed through reliable valuation procedure. IAEA safety standard series No ST-1 recommends regulation for safe transportation of the SNFs by transport casks, and United States is carrying out it according to the regulation guide, 10 CFR parts 71 and 72. Present research objective is to evaluate the KN-12 spent nuclear fuel transport cask that is designed for transportation of up to 12 assemblies and is standby status for being licensed in accordance with Korea Atomic Energy Act. Both radiation shielding and criticality analysis using the accurate Monte Carlo transport code, MCNP-4B are carried out for the KN-12 SNF cask as a benchmark calculation. Source terms for radiation shielding calculation are obtained using ORIGEN-S computer code. In this work, for normal transport conditions, the results from MCNP-4B shows the maximum dose rate of 0.557 mSv/hr at the side surface. And the maximum dose rate of 0.0871 mSv/hr was resulted at the 2 m distance from the cask. The level of calculated dose rate is 27.9% of the limit at the cask surface, 87.1% at 2 m from the cask surface for normal transport condition. For hypothetical accident conditions, the maximum rate of 2.5144 mSv/hr was resulted at the 1 m distance from the cask and this level is 25.1% of the limit for hypothetical accident conditions. In criticality calculations using MCNP-4B, the k{sub eff} values yielded for 5.0 w/o U-235 enriched fresh fuel are 0.92098 {+-} 0.00065. This result confirms subcritical condition of the KN-12 SNF cask and gives 96.95% of recommendations for criticality safety evaluation by US NRC these results will be useful as a basis for approval for the KN-12 SNF cask.

  8. Microchannel plate special nuclear materials sensor

    Feller, W.B.; White, P.L.; White, P.B.; Siegmund, O.H.W.; Martin, A.P.; Vallerga, J.V.

    2011-01-01

    Nova Scientific Inc., is developing for the Domestic Nuclear Detection Office (DNDO SBIR no. HSHQDC-08-C-00190), a solid-state, high-efficiency neutron detection alternative to 3 He gas tubes, using neutron-sensitive microchannel plates (MCPs) containing 10 B and/or Gd. This work directly supports DNDO development of technologies designed to detect and interdict nuclear weapons or illicit nuclear materials. Neutron-sensitized MCPs have been shown theoretically and more recently experimentally, to be capable of thermal neutron detection efficiencies equivalent to 3 He gas tubes. Although typical solid-state neutron detectors typically have an intrinsic gamma sensitivity orders of magnitude higher than that of 3 He gas detectors, we dramatically reduce gamma sensitivity by combining a novel electronic coincidence rejection scheme, employing a separate but enveloping gamma scintillator. This has already resulted in a measured gamma rejection ratio equal to a small 3 He tube, without in principle sacrificing neutron detection efficiency. Ongoing improvements to the MCP performance as well as the coincidence counting geometry will be described. Repeated testing and validation with a 252 Cf source has been underway throughout the Phase II SBIR program, with ongoing comparisons to a small commercial 3 He gas tube. Finally, further component improvements and efforts toward integration maturity are underway, with the goal of establishing functional prototypes for SNM field testing.

  9. Survey of radiation protection, radiation transport, and shielding information needs of the nuclear power industry. Final report

    Maskewitz, B.F.; Trubey, D.K.; Roussin, R.W.; McGill, B.L.

    1976-04-01

    The Radiation Shielding Information Center (RSIC) is engaged in a program to seek out, organize, and disseminate information in the area of radiation transport, shielding, and radiation protection. This information consists of published literature, nuclear data, and computer codes and advanced analytical techniques required by ERDA, its contractors, and the nuclear power industry to improve radiation analysis and computing capability. Information generated in this effort becomes a part of the RSIC collection and/or data base. The purpose of this report on project 219-1 is to document the results of the survey of information and computer code needs of the nuclear power industry in the area of radiation analysis and protection

  10. Survey of radiation protection, radiation transport, and shielding information needs of the nuclear power industry. Final report

    Maskewitz, B.F.; Trubey, D.K.; Roussin, R.W.; McGill, B.L.

    1976-04-01

    The Radiation Shielding Information Center (RSIC) is engaged in a program to seek out, organize, and disseminate information in the area of radiation transport, shielding, and radiation protection. This information consists of published literature, nuclear data, and computer codes and advanced analytical techniques required by ERDA, its contractors, and the nuclear power industry to improve radiation analysis and computing capability. Information generated in this effort becomes a part of the RSIC collection and/or data base. The purpose of this report on project 219-1 is to document the results of the survey of information and computer code needs of the nuclear power industry in the area of radiation analysis and protection.

  11. Construction of nuclear special information service

    Oh, Jeong Hoon; Kim, Tae Whan; Lee, Kyu Jeong; Yi, Ji Ho; Chun, Young Choon; Yoo, Jae Bok; Yoo, An Na

    2009-02-01

    Domestic INIS project has carried out various activities on supporting a decision-making for INIS Secretariat, exchanges of the statistical information between INIS and the country,and technical assistance for domestic end-users using INIS database. Based on the construction of INIS database sent by member states, the data published in the country has been gathered, collected, and inputted to INIS database according to the INIS reference series. Using the INIS output data, it has provided domestic users with searching INIS CD-ROM DB, INIS online database, INIS SDI service, and non-conventional literature delivery services. INIS2 DB Host site in Korea has serviced users of domestic and INIS member countries. In order to maintain the same data as Vienna center, the data update process has been performed. Also, publicity information activities were performed by many ways. To construct digital information infrastructure, we changed web server of NUCLIS21 and upgraded information management manager system and constructed full text DB on research and technical reports. Also we collected web DB, digital journals and so on and made an effort on operation of knowledge management system and research documents management. We have inputted over 3,000 records per year since 2002 and the input amount this year has reached 3,738 records. In order to input the comprehensive domestic publication related to nuclear energy, and rise in position of the national center, it is necessary to continue efforts and support budgets. We expect the INIS2 DB Host site will make a contribution to the improvement of productivity in the nuclear energy research as well as the diffusion of information about nuclear energy. We provided users with stable services changing web server of NUCLIS21. Also we contributed to the improvement of librarians' productivity by upgrading information management manager system and provided users with services of web DB, digital journals and so on

  12. Construction of nuclear special information service

    Oh, Jeong Hoon; Kim, Tae Whan; Yi, Ji Ho; Chun, Young Choon; Yoo, Jae Bok; Yoo, An Na; Choi, Heon Soo

    2012-01-01

    Domestic INIS project has carried out various activities on supporting a decision-making for INIS Secretariat, exchanges of the statistical information between INIS and the country, and technical assistance for domestic end-users using INIS database. Based on the construction of INIS database sent by member states, the data published in the country has been gathered, collected, and inputted to INIS database according to the INIS reference series. Using the INIS output data, it has provided domestic users with searching INIS CD-ROM DB, INIS online database, INIS SDI service, and non-conventional literature delivery services. INIS2 DB Host site in Korea has serviced users of domestic and INIS member countries. In order to maintain the same data as Vienna center, the data update process has been performed. Also, publicity information activities were performed by many ways. To construct digital information infrastructure, we changed web server of KORNIS21 and we reconstruct digital library system as separation of KAERI network system. We constructed patent trend analysis system and new SDI service system. Also we collected Web DB, digital journals and so on and made an effort on operation of knowledge management system and research documents management. We have inputted over 4,000 records per year since 2009 and the input amount this year has reached 4,284 records. In order to input the comprehensive domestic publication related to nuclear energy, and rise in position of the national center, it is necessary to continue efforts and support budgets. We expect the INIS2 DB Host site will make a contribution to the improvement of productivity in the nuclear energy research as well as the diffusion of information about nuclear energy. We provided users with stable services changing web server of KORNIS21. Also we contributed to the improvement of researcher's productivity by constructing patent trend analysis system, new SDI service system and provided users with services

  13. Build of virtual instrument laboratory related to nuclear species specialized

    Shan Jian; Zhao Guizhi; Zhao Xiuliang; Tang Lingzhi

    2009-01-01

    As rapid development of specialized related to nuclear science,the requirement of laboratory construct is analyzed in this article at first, One total conceive, One scheme deploy soft and hardware,three concrete characteristics targets and five different phases of put in practice of virtual instrument laboratory of specialized related to nuclear science are suggest in the paper,the concrete hardware structure and the headway of build of virtual instrument laboratory are described,and the first step effect is introduced.Lastly,the forward target and the further deliberateness that the virtual instrument laboratory construct are set forth in the thesis. (authors)

  14. Activities of the ANS special committee on nuclear nonproliferation

    Buckner, M.R. [Westinghouse Savannah River Co., Aiken, SC (United States); Sanders, T.L. [Sandia National Labs., Albuquerque, NM (United States)

    2001-07-01

    The American Nuclear Society (ANS) Special Committee on Nuclear Nonproliferation (SCNN) believes that to reverse current trends, U.S. policy must revisit the fundamental premise of Atoms for Peace: A collaborative nuclear enterprise enhances rather than diminishes national security. To accomplish this, the U.S. Government must develop an integrated policy on energy, nuclear technology, and national security. The policy must recognize that these are interrelated and that an integrated policy will require substantial investments in nuclear research and development and in nuclear education. This paper describes the current activities of the SCNN to heighten awareness of nonproliferation issues for decision makers and ANS members, and alert them to the need for action to resolve these concerns. (author)

  15. Activities of the ANS special committee on nuclear nonproliferation

    Buckner, M.R.; Sanders, T.L.

    2001-01-01

    The American Nuclear Society (ANS) Special Committee on Nuclear Nonproliferation (SCNN) believes that to reverse current trends, U.S. policy must revisit the fundamental premise of Atoms for Peace: A collaborative nuclear enterprise enhances rather than diminishes national security. To accomplish this, the U.S. Government must develop an integrated policy on energy, nuclear technology, and national security. The policy must recognize that these are interrelated and that an integrated policy will require substantial investments in nuclear research and development and in nuclear education. This paper describes the current activities of the SCNN to heighten awareness of nonproliferation issues for decision makers and ANS members, and alert them to the need for action to resolve these concerns. (author)

  16. Physical protection of facilities and special nuclear materials in france

    Jeanpierre, G.

    1980-01-01

    Physical protection of nuclear facilities and special nuclear materials is subject in France to a national governmental regulation which provides for the basic principles to be taken into account and the minimal level of protection deemed necessary. But the responsibility of implementation is left to the facility management and the resulting decentralization allows for maximum efficiency. All safeguards measures comply with the commitments taken at the international level by the French government

  17. Semiannual report on strategic special nuclear material inventory differences

    1987-07-01

    This twentieth periodic semiannual report of unclassified Inventory Differences (ID's) covers the second six months of fiscal year 1986 (April 1, 1986, through September 30, 1986) for all key Department of Energy (DOE) and DOE contractor operated facilities possessing strategic special nuclear materials. Data for the Rocky Flats and Y-12 nuclear weapons production facilities are not included in the report in order to protect classified nuclear weapons information; however, classified ID data from these facilities receive the same scrutiny and analyses as the unclassified data

  18. 10 CFR 70.11 - Persons using special nuclear material under certain Department of Energy and Nuclear Regulatory...

    2010-01-01

    ... 10 Energy 2 2010-01-01 2010-01-01 false Persons using special nuclear material under certain Department of Energy and Nuclear Regulatory Commission contracts. 70.11 Section 70.11 Energy NUCLEAR... using special nuclear material under certain Department of Energy and Nuclear Regulatory Commission...

  19. Radiation shielding lead shield

    Dei, Shoichi.

    1991-01-01

    The present invention concerns lead shields for radiation shielding. Shield boxes are disposed so as to surround a pipeline through which radioactive liquids, mists or like other objects are passed. Flanges are formed to each of the end edges of the shield boxes and the shield boxes are connected to each other by the flanges. Upon installation, empty shield boxes not charged with lead particles and iron plate shields are secured at first at the periphery of the pipeline. Then, lead particles are charged into the shield boxes. This attains a state as if lead plate corresponding to the depth of the box is disposed. Accordingly, operations for installation, dismantling and restoration can be conducted in an empty state with reduced weight to facilitate the operations. (I.S.)

  20. Active and Passive Diagnostic Signatures of Special Nuclear Materials

    Myers, William L. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Karpius, Peter Joseph [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Myers, Steven Charles [Los Alamos National Lab. (LANL), Los Alamos, NM (United States)

    2017-05-26

    An overview will be given discussing signatures associated with special nuclear materials acquired using both active and passive diagnostic techniques. Examples of how technology advancements have helped improve diagnostic capabilities to meet the challenges of today’s applications will be discussed.

  1. Calculations of atomic magnetic nuclear shielding constants based on the two-component normalized elimination of the small component method

    Yoshizawa, Terutaka; Zou, Wenli; Cremer, Dieter

    2017-04-01

    A new method for calculating nuclear magnetic resonance shielding constants of relativistic atoms based on the two-component (2c), spin-orbit coupling including Dirac-exact NESC (Normalized Elimination of the Small Component) approach is developed where each term of the diamagnetic and paramagnetic contribution to the isotropic shielding constant σi s o is expressed in terms of analytical energy derivatives with regard to the magnetic field B and the nuclear magnetic moment 𝝁 . The picture change caused by renormalization of the wave function is correctly described. 2c-NESC/HF (Hartree-Fock) results for the σiso values of 13 atoms with a closed shell ground state reveal a deviation from 4c-DHF (Dirac-HF) values by 0.01%-0.76%. Since the 2-electron part is effectively calculated using a modified screened nuclear shielding approach, the calculation is efficient and based on a series of matrix manipulations scaling with (2M)3 (M: number of basis functions).

  2. Organ dose assessment of nuclear medicine practitioners using L-block shielding device for handing diagnostic radioisotopes

    Kang, Se Sik; Kim, Jung Hoon [Dep. of Radiological Science, College of Health Science, Catholic University of Pusan, Busan (Korea, Republic of); Cho, Yong In [Dept. of Diagnostic Radiology, Dongnam Institute of Radiological and Medical Science, Busan (Korea, Republic of)

    2017-03-15

    In the case of nuclear medicine practitioners in medical institutions, a wide range of exposure dose to individual workers can be found, depending on the type of source, the amount of radioactivity, and the use of shielding devices in handling radioactive isotopes. In this regard, this study evaluated the organ dose on practitioners as well as the dose reduction effect of the L-block shielding device in handling the diagnostic radiation source through the simulation based on the Monte Carlo method. As a result, the distribution of organ dose was found to be higher as the position of the radiation source was closer to the handling position of a practitioner, and the effective dose distribution was different according to the ICRP tissue weight. Furthermore, the dose reduction effect according to the L-block thickness tended to decrease, which showed the exponential distribution, as the shielding thickness increased. The dose reduction effect according to each radiation source showed a low shielding effect in proportion to the emitted gamma ray energy level.

  3. Advanced methodologies of evaluating the radiation sources and ionising radiation shieldings for reducing the irradiation in nuclear field personnel

    Pantazi, D.; Mateescu, S.; Stanciu, M.

    2003-01-01

    One of the technical measures of protection against ionizing radiations is the radiation shielding. The process of implementing modern and efficient methods of evaluating the radiation shielding implies advanced calculation methods. That means using from simpler 1-D or 2-D computing codes such as MicroShield or QAD up to systems of codes such as SCALE (containing several independent modules) or the Monte Carlo multipurpose and many particles, MCNP, transport code. The main objective of this work is to present the Monte Carlo based evaluation of the dose rates from the CANDU type spent fuel all along the path of its handling up to intermediate storage. These values will be then compared with the values obtained from calculations with different computing programs. To obtain this objective two problems were approached: - establishing geometrical models according to the definition used by MCNP code so that the characteristics of CANDU type nuclear fuel are taking into account; - checking the validity of the proposed models by comparing the MCNP results with those obtained with other computing codes specific for shielding evaluation and radiation dose calculation

  4. Assessment of specialized educational programs for licensed nuclear reactor operators

    Melber, B.D.; Saari, L.M.; White, A.S.; Geisendorfer, C.L.; Huenefeld, J.C.

    1986-02-01

    This report assesses the job-relatedness of specialized educational programs for licensed nuclear reactor operators. The approach used involved systematically comparing the curriculum of specialized educational programs for college credit, to academic knowledge identified as necessary for carrying out the jobs of licenses reactor operators. A sample of eight programs, including A.S. degree, B.S. degree, and coursework programs were studied. Subject matter experts in the field of nuclear operations curriculum and training determined the extent to which individual program curricula covered the identified job-related academic knowledge. The major conclusions of the report are: There is a great deal of variation among individual programs, ranging from coverage of 15% to 65% of the job-related academic knowledge. Four schools cover at least half, and four schools cover less than one-third of this knowledge content; There is no systematic difference in the job-relatedness of the different types of specialized educational programs, A.S. degree, B.S. degree, and coursework; and Traditional B.S. degree programs in nuclear engineering cover as much job-related knowledge (about one-half of this knowledge content) as most of the specialized educational programs

  5. Considerations for sampling nuclear materials for SNM accounting measurements. Special nuclear material accountability report

    Brouns, R.J.; Roberts, F.P.; Upson, U.L.

    1978-05-01

    This report presents principles and guidelines for sampling nuclear materials to measure chemical and isotopic content of the material. Development of sampling plans and procedures that maintain the random and systematic errors of sampling within acceptable limits for SNM(Special Nuclear Materials) accounting purposes are emphasized

  6. Calculation of binary magnetic properties and potential energy curve in xenon dimer: second virial coefficient of (129)Xe nuclear shielding.

    Hanni, Matti; Lantto, Perttu; Runeberg, Nino; Jokisaari, Jukka; Vaara, Juha

    2004-09-22

    Quantum chemical calculations of the nuclear shielding tensor, the nuclear quadrupole coupling tensor, and the spin-rotation tensor are reported for the Xe dimer using ab initio quantum chemical methods. The binary chemical shift delta, the anisotropy of the shielding tensor Delta sigma, the nuclear quadrupole coupling tensor component along the internuclear axis chi( parallel ), and the spin-rotation constant C( perpendicular ) are presented as a function of internuclear distance. The basis set superposition error is approximately corrected for by using the counterpoise correction (CP) method. Electron correlation effects are systematically studied via the Hartree-Fock, complete active space self-consistent field, second-order Møller-Plesset many-body perturbation, and coupled-cluster singles and doubles (CCSD) theories, the last one without and with noniterative triples, at the nonrelativistic all-electron level. We also report a high-quality theoretical interatomic potential for the Xe dimer, gained using the relativistic effective potential/core polarization potential scheme. These calculations used valence basis set of cc-pVQZ quality supplemented with a set of midbond functions. The second virial coefficient of Xe nuclear shielding, which is probably the experimentally best-characterized intermolecular interaction effect in nuclear magnetic resonance spectroscopy, is computed as a function of temperature, and compared to experiment and earlier theoretical results. The best results for the second virial coefficient, obtained using the CCSD(CP) binary chemical shift curve and either our best theoretical potential or the empirical potentials from the literature, are in good agreement with experiment. Zero-point vibrational corrections of delta, Delta sigma, chi (parallel), and C (perpendicular) in the nu=0, J=0 rovibrational ground state of the xenon dimer are also reported.

  7. Modeling most likely pathways for smuggling radioactive and special nuclear materials on a worldwide multimodal transportation network

    Saeger, Kevin J [Los Alamos National Laboratory; Cuellar, Leticia [Los Alamos National Laboratory

    2010-01-01

    Nuclear weapons proliferation is an existing and growing worldwide problem. To help with devising strategies and supporting decisions to interdict the transport of nuclear material, we developed the Pathway Analysis, Threat Response and Interdiction Options Tool (PATRIOT) that provides an analytical approach for evaluating the probability that an adversary smuggling radioactive or special nuclear material will be detected during transit. We incorporate a global, multi-modal transportation network, explicit representation of designed and serendipitous detection opportunities, and multiple threat devices, material types, and shielding levels. This paper presents the general structure of PATRIOT, and focuses on the theoretical framework used to model the reliabilities of all network components that are used to predict the most likely pathways to the target.

  8. Modeling most likely pathways for smuggling radioactive and special nuclear materials on a worldwide multi-modal transportation network

    Saeger, Kevin J [Los Alamos National Laboratory; Cuellar, Leticia [Los Alamos National Laboratory

    2010-10-28

    Nuclear weapons proliferation is an existing and growing worldwide problem. To help with devising strategies and supporting decisions to interdict the transport of nuclear material, we developed the Pathway Analysis, Threat Response and Interdiction Options Tool (PATRIOT) that provides an analytical approach for evaluating the probability that an adversary smuggling radioactive or special nuclear material will be detected during transit. We incorporate a global, multi-modal transportation network, explicit representation of designed and serendipitous detection opportunities, and multiple threat devices, material types, and shielding levels. This paper presents the general structure of PATRIOT, all focuses on the theoretical framework used to model the reliabilities of all network components that are used to predict the most likely pathways to the target.

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

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

    2007-07-01

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

  10. Development of the production of special steels for nuclear industries

    Vieillard-Baron, B.

    1977-01-01

    The development of electro-nuclear industries has a powerful impact on the production of special steels, although the quantity of material applied to the non-conventional parts of nuclear power plants is quite small as compared to the total production requirements in this industrial field. Evolution bears on the product research, development and testing methods, on the technical and marketing services - in particular the establishment of quality control teams and assurance manuals - and the implementation of high performance production equipments. Manufacturing must however take place under normal work load and productivity conditions of production tools, and thus ensure a satisfactory profitability on investments entailed [fr

  11. Safety catching device for pipes in missile shielding cylinders of nuclear power plants

    Hering, S.; Doll, B.

    1976-01-01

    The safety catching device consists of a steel wire passed in U-shape around the pipe to be caught and supported by two anchor ties embedded in the concrete of the missile shielding cylinder. This flexible catching device is to cause the energy released in case of a pipe rupture to be absorbed and no dangerous bending shesses to be transferred to the walls of the missile shielding cylinder. (UWI) [de

  12. Radiation shielding

    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)

  13. Assessment of radiation shielding materials for protection of space crews using CR-39 plastic nuclear track detector

    DeWitt, J.M.; Benton, E.R.; Uchihori, Y.; Yasuda, N.; Benton, E.V.; Frank, A.L.

    2009-01-01

    A significant obstacle to long duration human space exploration such as the establishment of a permanent base on the surface of the Moon or a human mission to Mars is the risk posed by prolonged exposure to space radiation. In order to keep mission costs at acceptable levels while simultaneously minimizing the risk from radiation to space crew health and safety, a judicious use of optimized shielding materials will be required. We have undertaken a comprehensive study using CR-39 plastic nuclear track detector (PNTD) to characterize the radiation shielding properties of a range of materials-both common baseline materials such as Al and polyethylene, and novel multifunctional materials such as carbon composites-at heavy ion accelerators. The study consists of analyzing CR-39 PNTD exposed in front of and behind shielding targets of varying composition and at a number of depths (target thicknesses) relevant to the development and testing of materials for space radiation shielding. Most targets consist of 10 cm x 10 cm slabs of solid materials ranging in thickness from 1 to >30 g/cm 2 . Exposures have been made to beams of C, O, Ne, Si, Ar, and Fe at energies ranging from 290 MeV/amu to 1 GeV/amu at the National Institute of Radiological Sciences HIMAC and the NASA Space Radiation Laboratory (NSRL) at Brookhaven National Laboratory. Analysis of the exposed detectors yields LET spectrum, dose, and dose equivalent as functions of target depth and composition, and incident heavy ion charge, energy, and fluence. Efforts are currently underway to properly weigh and combine these results into a single quantitative estimate of a material's ability to shield space crews from the interplanetary galactic cosmic ray flux.

  14. An innovative method for on-power radiometry of end-shields of nuclear power plants

    Kumar, Gaurav; Gupta, Pankaj; Nawal, Shriram; Gautam, Mahesh; Kakkar, Aman Deep; Yadav, Umed

    2012-01-01

    Every lndian PHWR reactor calandria is sandwiched within a pair of shield on either side. These shields are perpendicular to the coaxial axis of calandria and are called end-shields. These provide shielding from leakage radiation from reactor core in escaping out to Fuelling Machine vault, thereby significantly reducing the dose rates in the vaults. This has got a direct impact on radiation field in accessible areas. By maintaining low dose rates in accessible areas, the individual and collective doses of radiation workers can be effectively controlled well within the stipulated limits. Thus, it is of utmost importance to ensure adequacy of shielding provided by end-shields. In this context, a limited radiometry exercise is executed after filling of end-shields with steel balls and prior to their installation at designated place. This exercise provides limited inputs along the periphery of end-shield due to limited strength of radiation source, its handling provisions and dose constraints to the individual. In order to ascertain an in-depth analysis of shielding adequacy on-power, different methodologies have been adopted and have certain limitation in precisely pinpointing the affected area/location besides limitation on number of locations that can be monitored at a single stretch. To overcome these important anomalies, a computer based setup has been indigenously designed. The setup essentially comprises of a radiation monitor with wide energy, measuring, temperature and humidity range; a custom designed 25 m long compatible cable with suitable connectors; a laptop with additional cooling arrangement; a configurable interfacing software; thermal shielding for the detector and tying/fixing provisions. The radiation monitor after being properly shielded for thermal impacts is installed on the head of Fuelling Machine. It is connected through long cable to a laptop kept at Fuelling Machine service area with due cooling provisions (as temperature in the area will

  15. Methods of Verification, Accountability and Control of Special Nuclear Material

    Stewart, J.E.

    1999-01-01

    This session demonstrates nondestructive assay (NDA) measurement, surveillance and analysis technology required to protect, control and account (MPC and A) for special nuclear materials (SNM) in sealed containers. These measurements, observations and analyses comprise state-of-the art, strengthened, SNM safeguards systems. Staff member specialists, actively involved in research, development, training and implementation worldwide, will present six NDA verification systems and two software tools for integration and analysis of facility MPC and A data

  16. Shielding requirements for particle bed propulsion systems

    Gruneisen, S. J.

    1991-06-01

    Nuclear Thermal Propulsion systems present unique challenges in reliability and safety. Due to the radiation incident upon all components of the propulsion system, shielding must be used to keep nuclear heating in the materials within limits; in addition, electronic control systems must be protected. This report analyzes the nuclear heating due to the radiation and the shielding required to meet the established criteria while also minimizing the shield mass. Heating rates were determined in a 2000 MWt Particle Bed Reactor (PBR) system for all materials in the interstage region, between the reactor vessel and the propellant tank, with special emphasis on meeting the silicon dose criteria. Using a Lithium Hydride/Tungsten shield, the optimum shield design was found to be: 50 cm LiH/2 cm W on the axial reflector in the reactor vessel and 50 cm LiH/2 cm W in a collar extension of the inside shield outside of the pressure vessel. Within these parameters, the radiation doses in all of the components in the interstage and lower tank regions would be within acceptable limits for mission requirements.

  17. Shielding benchmark problems, (2)

    Tanaka, Shun-ichi; Sasamoto, Nobuo; Oka, Yoshiaki; Shin, Kazuo; Tada, Keiko.

    1980-02-01

    Shielding benchmark problems prepared by Working Group of Assessment of Shielding Experiments in the Research Committee on Shielding Design in the Atomic Energy Society of Japan were compiled by Shielding Laboratory in Japan Atomic Energy Research Institute. Fourteen shielding benchmark problems are presented newly in addition to twenty-one problems proposed already, for evaluating the calculational algorithm and accuracy of computer codes based on discrete ordinates method and Monte Carlo method and for evaluating the nuclear data used in codes. The present benchmark problems are principally for investigating the backscattering and the streaming of neutrons and gamma rays in two- and three-dimensional configurations. (author)

  18. Report from the Special Committee on Fukushima Nuclear Accident

    Ozawa, Mamoru

    2012-01-01

    The Special Committee on Fukushima Nuclear Accident was established in April 2011 under the Heat Transfer Society of Japan (HTSJ) and discussed (1) how had evolved heat transfer research in progress of nuclear technology, (2) role of expert group in the area of heat transfer academy and technology and (3) energy prospect in Japan after the Fukushima nuclear accident. This report was described by the chairman of the special committee summarizing one year discussions as (1) background of heat transfer research progress, (2) progression of Fukushima Daiichi Nuclear Power Plant accident, (3) energy problem in Japan after the Fukushima accident and (4) social role of the HTSJ. This HTSJ was a unique, nonprofit association in Japan of the people engaged in heat transfers research or in various engineering aspects related to heat transfer, which meant interdisciplinary or common platform of heat transfer as elementary technologies. Such actual complex problems could be discussed in the HTSJ from an overlooking viewpoint in order for the HTSJ to play a social role. (T. Tanaka)

  19. Neutron interrogation system using high gamma ray signature to detect contraband special nuclear materials in cargo

    Slaughter, Dennis R [Oakland, CA; Pohl, Bertram A [Berkeley, CA; Dougan, Arden D [San Ramon, CA; Bernstein, Adam [Palo Alto, CA; Prussin, Stanley G [Kensington, CA; Norman, Eric B [Oakland, CA

    2008-04-15

    A system for inspecting cargo for the presence of special nuclear material. The cargo is irradiated with neutrons. The neutrons produce fission products in the special nuclear material which generate gamma rays. The gamma rays are detecting indicating the presence of the special nuclear material.

  20. Nuclear analysis and shielding optimisation in support of the ITER In-Vessel Viewing System design

    Turner, Andrew; Pampin, Raul; Loughlin, M.J.; Ghani, Zamir; Hurst, Gemma; Lo Bue, Alessandro; Mangham, Samuel; Puiu, Adrian; Zheng, Shanliang

    2014-01-01

    The In-Vessel Viewing System (IVVS) units proposed for ITER are deployed to perform in-vessel examination. During plasma operations, the IVVS is located beyond the vacuum vessel, with shielding blocks envisaged to protect components from neutron damage and reduce shutdown dose rate (SDR) levels. Analyses were conducted to determine the effectiveness of several shielding configurations. The neutron response of the system was assessed using global variance reduction techniques and a surface source, and shutdown dose rate calculations were undertaken using MCR2S. Unshielded, the absorbed dose to piezoelectric motors (PZT) was found to be below stable limits, however activation of the primary closure plate (PCP) was prohibitively high. A scenario with shielding blocks at probe level showed significantly reduced PCP contact dose rate, however still marginally exceeded port cell requirements. The addition of shielding blocks at the bioshield plug demonstrated PCP contact dose rates below project requirements. SDR levels in contact with the isolated IVVS cartridge were found to marginally exceed the hands-on maintenance limit. For engineering feasibility, shielding blocks at bioshield level are to be avoided, however the port cell SDR field requires further consideration. In addition, alternative low-activation steels are being considered for the IVVS cartridge

  1. Nuclear analysis and shielding optimisation in support of the ITER In-Vessel Viewing System design

    Turner, Andrew, E-mail: andrew.turner@ccfe.ac.uk [EURATOM/CCFE Fusion Association, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Pampin, Raul [F4E Fusion for Energy, Josep Pla 2, Torres Diagonal Litoral B3, 08019 Barcelona (Spain); Loughlin, M.J. [ITER Organisation, Route de Vinon sur Verdon, 13115 Saint Paul Lez Durance (France); Ghani, Zamir; Hurst, Gemma [EURATOM/CCFE Fusion Association, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Lo Bue, Alessandro [F4E Fusion for Energy, Josep Pla 2, Torres Diagonal Litoral B3, 08019 Barcelona (Spain); Mangham, Samuel [EURATOM/CCFE Fusion Association, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom); Puiu, Adrian [F4E Fusion for Energy, Josep Pla 2, Torres Diagonal Litoral B3, 08019 Barcelona (Spain); Zheng, Shanliang [EURATOM/CCFE Fusion Association, Culham Science Centre, Abingdon, Oxon OX14 3DB (United Kingdom)

    2014-10-15

    The In-Vessel Viewing System (IVVS) units proposed for ITER are deployed to perform in-vessel examination. During plasma operations, the IVVS is located beyond the vacuum vessel, with shielding blocks envisaged to protect components from neutron damage and reduce shutdown dose rate (SDR) levels. Analyses were conducted to determine the effectiveness of several shielding configurations. The neutron response of the system was assessed using global variance reduction techniques and a surface source, and shutdown dose rate calculations were undertaken using MCR2S. Unshielded, the absorbed dose to piezoelectric motors (PZT) was found to be below stable limits, however activation of the primary closure plate (PCP) was prohibitively high. A scenario with shielding blocks at probe level showed significantly reduced PCP contact dose rate, however still marginally exceeded port cell requirements. The addition of shielding blocks at the bioshield plug demonstrated PCP contact dose rates below project requirements. SDR levels in contact with the isolated IVVS cartridge were found to marginally exceed the hands-on maintenance limit. For engineering feasibility, shielding blocks at bioshield level are to be avoided, however the port cell SDR field requires further consideration. In addition, alternative low-activation steels are being considered for the IVVS cartridge.

  2. Safety catching device for pipe lines in missile shielding cylinders of nuclear power plants

    Hering, S.; Doll, B.

    1975-01-01

    The safety catching device for pipes in the missile shielding cylinders consists of a flexible steel cable surrounding the pipe in a distance in U-shape. The arrester cable - which works as a spring and is freely movable in all directions - is attached to the cylinder wall. For this, the ends of the cable are primarily fastened to anchor boxes which are then inserted in a stay tube with the same axis as the cable ends. The anchor boxes are fastened to the outer wall of the missile shielding cylinder by anchor bolts and holding plates. (DG/AK) [de

  3. Theoretical Study of H/D Isotope Effects on Nuclear Magnetic Shieldings Using an ab initio Multi-Component Molecular Orbital Method

    Masanori Tachikawa

    2013-05-01

    Full Text Available We have theoretically analyzed the nuclear quantum effect on the nuclear magnetic shieldings for the intramolecular hydrogen-bonded systems of σ-hydroxy acyl aromatic species using the gauge-including atomic orbital technique combined with our multi-component density functional theory. The effect of H/D quantum nature for geometry and nuclear magnetic shielding changes are analyzed. Our study clearly demonstrated that the geometrical changes of hydrogen-bonds induced by H/D isotope effect (called geometrical isotope effect: GIE is the dominant factor of deuterium isotope effect on 13C chemical shift.

  4. Nuclear Magnetic Shielding of Monoboranes: Calculation and Assessment of B-11 NMR Chemical Shifts in Planar BX3 and in Tetrahedral [BX4](-) Systems

    Macháček, Jan; Bühl, M.; Fanfrlík, Jindřich; Hnyk, Drahomír

    2017-01-01

    Roč. 121, č. 50 (2017), s. 9631-9637 ISSN 1089-5639 R&D Projects: GA ČR(CZ) GA17-08045S Institutional support: RVO:61388980 ; RVO:61388963 Keywords : Electrostatic potentials * Nonrelativistic * Nuclear magnetic shieldings Subject RIV: CA - Inorganic Chemistry OBOR OECD: Inorganic and nuclear chemistry Impact factor: 2.847, year: 2016

  5. Present and future problems of radiation shielding for maritime transport of nuclear spent fuels

    Ueki, K.; Nariyama, N.; Ohashi, A.

    2000-01-01

    The transport of spent fuels with casks began in September 1999 by the exclusive spent fuel transport vessel the 'Rokuei Maru'. The casks have been transported to the reprocessing plant at Rokkasho-village in Aomori Prefecture. The 'Rokuei Maru' is approximately 100 m-length, 16.5 m-width and 3,000 gross-tons. The 20 NFT casks can be loaded into 5 holds. At the present time, the NFT casks can carry spent fuels of up to 44,000 MWD/MTU. Serpentine concrete is employed as a neutron shields in the hatch covers, the bulkheads, and the house front of the accommodations except the wheelhouse. Polyethylene covers the side walls in each hold. The neutron shielding ability of serpentine concrete and polyethylene was investigated by a shielding experiment using a 252 Cf-neutron source. The shielding experiment was analyzed with the Monte Carlo code MCNP 4B. In the near future, on-board experiment will be carried out to measure the dose-equivalent rate distributions in the 'Rokuei Maru' and the measured data and the Monte Carlo analysis of it will establish the radiation safety of the ship. (author)

  6. A special purpose simulation language for nuclear power plants

    Saphier, D.

    1980-01-01

    A special purpose block-oriented simulation language, 'The Dynamic Simulator for Nuclear Power Plants' (DSNP), was developed at Argonne National Laboratory. The dominant feature of DSNP is the ability to transform a power plant flowchart or block diagram directly into a simulation program. The user is required to recognize the symbolic DSNP statements for the appropriate physical component, and list these statements in a logical sequence according to the flow of physical properties in the simulated power plant. At present most of the component models in DSNP are of the lumped parameter type. Although DSNP is a special purpose simulation language, it also has all the features of a general purpose simulation language, and in particular a powerful macro processor. The use of DSNP is demonstrated by a sample problem simulating a reactor startup accident. (Auth.)

  7. Design of an integral missile shield in integrated head assembly for pressurized water reactor at commercial nuclear plants

    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

  8. Westinghouse Hanford Company special nuclear material vault storage study

    Borisch, R.R.

    1996-01-01

    Category 1 and 2 Special Nuclear Materials (SNM) require storage in vault or vault type rooms as specified in DOE orders 5633.3A and 6430.1A. All category 1 and 2 SNM in dry storage on the Hanford site that is managed by Westinghouse Hanford Co (WHC) is located in the 200 West Area at Plutonium Finishing Plant (PFP) facilities. This document provides current and projected SNM vault inventories in terms of storage space filled and forecasts available space for possible future storage needs

  9. Effect of special features of nuclear power plants

    Scharf, H.

    1986-01-01

    Special features of nuclear power plants are reported with the Muelheim-Kaerlich pressurized water reactor as the reference plant. This nuclear reactor uses 'Once Through Steam Generators (OTSG)' with 'Integrated Economizer' to provide the turbine with superheated steam. The implementation of OTSG allows to operate the plant with constant steam pressure over the entire power range, and with constant main coolant temperature over a power range from 15% power to 100% power. Control of the plant during power operation is provided by the 'Integrated Control System', which simultaneously sends signals to the plant's subsystems reactor, OTSG, and turbine to get optimum response of the plant during power transients. The characteristics of this 'Integrated Control System' and its different modes of operation are presented. (orig./GL)

  10. JENDL special purpose data files and related nuclear data

    Iijima, Shungo

    1989-01-01

    The objectives of JENDL Special Purpose Data Files under development are the applications of nuclear data to the evaluation of the fuel cycle, nuclear activation, and radiation damage. The files in plan consist of 9 types of data, viz., the actinide cross sections, the decay data, the activation cross sections, the (α,n) cross sections, the photo-reaction cross sections, the dosimetry cross sections, the gas production cross sections, the primary knock-on atom spectra and KERMA factors, and the data for standard. The status of the compilation and the evaluation of these data are briefly reviewed. In particular, the features of the data required for the evaluation of the activation cross sections, (α,n) cross sections, photo-reaction cross sections, and PKA data are discussed in some detail. The need for the realistic definition of the scope of the work is emphasized. (author)

  11. Accelerator shielding benchmark problems

    Hirayama, H.; Ban, S.; Nakamura, T.

    1993-01-01

    Accelerator shielding benchmark problems prepared by Working Group of Accelerator Shielding in the Research Committee on Radiation Behavior in the Atomic Energy Society of Japan were compiled by Radiation Safety Control Center of National Laboratory for High Energy Physics. Twenty-five accelerator shielding benchmark problems are presented for evaluating the calculational algorithm, the accuracy of computer codes and the nuclear data used in codes. (author)

  12. On evaluated nuclear data for beta-delayed gamma rays following of special nuclear materials

    Mencarini, Leonardo de H.; Caldeira, Alexandre D., E-mail: mencarini@ieav.cta.b, E-mail: alexdc@ieav.cta.b [Instituto de Estudos Avancados (IEAv/CTA), Sao Jose dos Campos, SP (Brazil)

    2011-07-01

    In this paper, a new type of information available in ENDF is discussed. During a consistency check of the evaluated nuclear data library ENDF/B-VII.0 performed at the Nuclear Data Subdivision of the Institute for Advanced Studies, the size of the files for some materials drew the attention of one of the authors. Almost 94 % of all available information for these special nuclear materials is used to represent the beta-delayed gamma rays following fission. This is the first time this information is included in an ENDF version. (author)

  13. On evaluated nuclear data for beta-delayed gamma rays following of special nuclear materials

    Mencarini, Leonardo de H.; Caldeira, Alexandre D.

    2011-01-01

    In this paper, a new type of information available in ENDF is discussed. During a consistency check of the evaluated nuclear data library ENDF/B-VII.0 performed at the Nuclear Data Subdivision of the Institute for Advanced Studies, the size of the files for some materials drew the attention of one of the authors. Almost 94 % of all available information for these special nuclear materials is used to represent the beta-delayed gamma rays following fission. This is the first time this information is included in an ENDF version. (author)

  14. Radiation shielding at interim storage facility for CANDU-type nuclear spent fuel

    Mateescu, S.; Radu, M. Pantazi D.; Stanciu, M.

    1997-01-01

    Technical measures in radiological protection are taken in the interim storage facility design to ensure that, during normal operation, exposures of workers and members of public to ionizing radiation are limited to levels lower than regulatory limits. The spent fuel storage design provides for radiation exposure to be as low as reasonable achievable (ALARA principles). The evaluation of radiation shields includes the most conservative provisions: - all locations which may contain spent fuel are full; - the spent fuel has reached the maximum burnup; - the post irradiation cooling period should be the minimum reasonable; - equipment for handling contains the maximum amount of spent fuel. Radiation shields should ensure that external radiation fields do not exceed limits accepted by the Regulatory Body Module. The evaluation has been performed with two computer codes, QAD-5K and MICROSHIELD-4. (authors)

  15. Gauge origin independent calculations of nuclear magnetic shieldings in relativistic four-component theory

    Ilias, Miroslav; Saue, Trond; Enevoldsen, Thomas

    2009-01-01

    The use of perturbation-dependent London atomic orbitals, also called gauge including atomic orbitals, has proven efficient for calculations of NMR shielding constants and other magnetic properties in the nonrelativistic framework. In this paper, the theory of London atomic orbitals for NMR...... calculates the diamagnetic contribution as an expectation value, leads to significant errors and is not recommended. (C) 2009 American Institute of Physics. [doi:10.1063/1.3240198]...

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

    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

  17. Effects of neutron source ratio on nuclear characteristics of D-D fusion reactor blankets and shields

    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)

  18. Clinical Training of Medical Physicists Specializing in Nuclear Medicine

    2011-01-01

    The application of radiation in human health, for both diagnosis and treatment of disease, is an important component of the work of the IAEA. The responsibility for the increasingly technical aspects of this work is undertaken by the medical physicist. To ensure good practice in this vital area, structured clinical training programmes are required to complement academic learning. This publication is intended to be a guide to the practical implementation of such a programme for nuclear medicine. There is a general and growing awareness that radiation medicine is increasingly dependent on well trained medical physicists who are based in a clinical setting. However an analysis of the availability of medical physicists indicates a large shortfall of qualified and capable professionals. This is particularly evident in developing countries. While strategies to increase educational opportunities are critical to such countries, the need for guidance on structured clinical training was recognized by the members of the Regional Cooperative Agreement for Research, Development and Training related to Nuclear Science and Technology (RCA) for the Asia-Pacific region. Consequently, a technical cooperation regional project (RAS6038) under the RCA programme was formulated to address this need in this region by developing suitable material and establishing its viability. Development of a clinical training guide for medical physicists specialising in nuclear medicine was started in 2009 with the appointment of a core drafting committee of regional and international experts. The publication drew on the experience of clinical training in Australia, Croatia and Sweden and was moderated by physicists working in the Asian region. The present publication follows the approach of earlier IAEA publications in the Training Course Series, specifically Nos 37 and 47, Clinical Training of Medical Physicists Specializing in Radiation Oncology and Clinical Training of Medical Physicists

  19. Plans of reorganization of USA nuclear military complex and provision of military program by special nuclear materials

    Semenovskaya, I.V.

    1996-01-01

    Consideration is given to plans and implementation of the program of reorganization of USA nuclear military complex, related with conducted reduction of nuclear arsenal after concluding the Strategic Nuclear Armament Reduction Treaty. Particular attention is paid to problems of satisfying short-term and long-term requirements in special nuclear materials and in tritium in particular

  20. 75 FR 2163 - Constellation Energy; Notice of Docketing of Special Nuclear Material License SNM-2505 Amendment...

    2010-01-14

    ... NUCLEAR REGULATORY COMMISSION [Docket No. 72-8; NRC-2010-0011] Constellation Energy; Notice of Docketing of Special Nuclear Material License SNM-2505 Amendment Application for the Calvert Cliffs... Constellation Energy (Constellation) to amend its Special Nuclear Material License No. SNM-2505, under the...

  1. SP-100 GES/NAT radiation shielding systems design and development testing

    Disney, R.K.; Kulikowski, H.D.; McGinnis, C.A.; Reese, J.C.; Thomas, K.; Wiltshire, F.

    1991-01-01

    Advanced Energy Systems (AES) of Westinghouse Electric Corporation is under subcontract to the General Electric Company to supply nuclear radiation shielding components for the SP-100 Ground Engineering System (GES) Nuclear Assembly Test to be conducted at Westinghouse Hanford Company at Richland, Washington. The radiation shielding components are integral to the Nuclear Assembly Test (NAT) assembly and include prototypic and non-prototypic radiation shielding components which provide prototypic test conditions for the SP-100 reactor subsystem and reactor control subsystem components during the GES/NAT operations. W-AES is designing three radiation shield components for the NAT assembly; a prototypic Generic Flight System (GFS) shield, the Lower Internal Facility Shield (LIFS), and the Upper Internal Facility Shield (UIFS). This paper describes the design approach and development testing to support the design, fabrication, and assembly of these three shield components for use within the vacuum vessel of the GES/NAT. The GES/NAT shields must be designed to operate in a high vacuum which simulates space operations. The GFS shield and LIFS must provide prototypic radiation/thermal environments and mechanical interfaces for reactor system components. The NAT shields, in combination with the test facility shielding, must provide adequate radiation attenuation for overall test operations. Special design considerations account for the ground test facility effects on the prototypic GFS shield. Validation of the GFS shield design and performance will be based on detailed Monte Carlo analyses and developmental testing of design features. Full scale prototype testing of the shield subsystems is not planned

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

    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

  3. Shielded cells transfer automation

    Fisher, J.J.

    1984-01-01

    Nuclear waste from shielded cells is removed, packaged, and transferred manually in many nuclear facilities. Radiation exposure is absorbed by operators during these operations and limited only through procedural controls. Technological advances in automation using robotics have allowed a production waste removal operation to be automated to reduce radiation exposure. The robotic system bags waste containers out of glove box and transfers them to a shielded container. Operators control the system outside the system work area via television cameras. 9 figures

  4. The nuclear shield in the 'thirty-year war' of physicists against ignorant criticism of modern physical theories

    Vizgin, Vladimir P

    1999-01-01

    This article deals with the almost 'thirty-year war' led by physicists against the authorities' incompetent philosophical and ideological interference with science. The 'war' is shown to have been related to the history of Soviet nuclear weapons. Theoretical milestones of 20th century physics, to wit, theory of relativity and quantum mechanics, suffered endless 'attacks on philosophical grounds'. The theories were proclaimed idealistic as well as unduly abstract and out of touch with practice; their authors and followers were labelled 'physical idealists', and later, in the 1940s and 1950s, even 'cosmopolitans without kith or kin'. Meanwhile, quantum and relativistic theories, as is widely known, had become the basis of nuclear physics and of the means of studying the atomic nucleus (charged particle accelerators, for instance). The two theories thus served, to a great extent, as a basis for both peaceful and military uses of nuclear energy, made possible by the discovery of uranium nuclear fission under the action of neutrons. In the first part, the article recounts how prominent physicists led the way to resisting philosophical and ideological pressure and standing up for relativity, quantum theories and nuclear physics, thus enabling the launch of the atomic project. The second part contains extensive material proving the point that physicists effectively used the 'nuclear shield' in the 1940s and 1950s against the 'philosophical-cosmopolitan' pressure, indeed saving physics from a tragic fate as that of biology at the Academy of Agricultural Sciences (VASKhNIL) session in 1948. (from the history of physics)

  5. Special Nuclear Material Portal Monitoring at the Nevada Test Site

    DeAnn Long; Michael Murphy

    2008-01-01

    Prior to April 2007, acceptance and performance testing of the various Special Nuclear Material (SNM) monitoring devices at the Nevada Test Site (NTS) was performed by the Radiological Health Instrumentation department. Calibration and performance testing on the PM-700 personnel portal monitor was performed, but there was no test program for the VM-250 vehicle portal monitor. The handheld SNM monitors, the TSA model 470B, were being calibrated annually, but there was no performance test program. In April of 2007, the Material Control and Accountability Manager volunteered to take over performance testing of all SNM portal monitors at NTS in order to strengthen the program and meet U.S. Department of Energy Order requirements. This paper will discuss the following activities associated with developing a performance testing program: changing the culture, learning the systems, developing and implementing procedures, troubleshooting and repair, validating the process, physical control of equipment, acquisition of new systems, and implementing the performance test program

  6. Special nuclear material radiation monitors for the 1980's

    Fehlau, P.E.

    1985-01-01

    During the two decades that automatic gamma-radiation monitors have been applied to detecting special nuclear material (SNM), little attention has been devoted to how well the monitors perform in plant environments. Visits to 11 DOE facilities revealed poor information flow between developers, manufacturers, and maintainers of SNM radiation monitors. To help users achieve best performance from their monitors or select new ones, Los Alamos National Laboratory developed a hand-held monitor user's guide, calibration manuals for some commercial SNM pedestrian monitors, and an application guide for SNM pedestrian monitors. In addition, Los Alamos evaluated new commercial SNM monitors, considered whether to apply neutron detection to SNM monitoring, and investigated the problem of operating gamma-ray SNM monitors in variable plutonium gamma-radiation fields. As a result, the performance of existing SNM monitors will improve and alternative monitoring methods will become commerciallly available during the 1980s. 9 refs., 6 figs., 1 tab

  7. Detecting special nuclear material using muon-induced neutron emission

    Guardincerri, Elena; Bacon, Jeffrey; Borozdin, Konstantin; Matthew Durham, J.; Fabritius II, Joseph [Los Alamos National Laboratory, Los Alamos, NM 87545 (United States); Hecht, Adam [University of New Mexico, Albuquerque, NM 87131 (United States); Milner, Edward C. [Southern Methodist University, Dallas, TX 75205 (United States); Miyadera, Haruo; Morris, Christopher L. [Los Alamos National Laboratory, Los Alamos, NM 87545 (United States); Perry, John [Los Alamos National Laboratory, Los Alamos, NM 87545 (United States); University of New Mexico, Albuquerque, NM 87131 (United States); Poulson, Daniel [Los Alamos National Laboratory, Los Alamos, NM 87545 (United States)

    2015-07-21

    The penetrating ability of cosmic ray muons makes them an attractive probe for imaging dense materials. Here, we describe experimental results from a new technique that uses neutrons generated by cosmic-ray muons to identify the presence of special nuclear material (SNM). Neutrons emitted from SNM are used to tag muon-induced fission events in actinides and laminography is used to form images of the stopping material. This technique allows the imaging of SNM-bearing objects tagged using muon tracking detectors located above or to the side of the objects, and may have potential applications in warhead verification scenarios. During the experiment described here we did not attempt to distinguish the type or grade of the SNM.

  8. Criticality and shielding calculations of an interim dry storage system for the spent fuel from Atucha I Nuclear Power Plant

    Silva, M

    2006-01-01

    The Atucha I Nuclear Power Plant (CNA-I) has enough room to store its spent fuel (SF) in damp in its two pool houses until the middle of 2015.Before that date there is the need to have an interim dry storage system for spent fuel that would make possible to empty at least one of the pools, whether to keep the plant operating if its useful life is extended, or to be able to empty the reactor core in case of decommissioning.Nucleolectrica Argentina S.A. (NA-SA) and the Comision Nacional de Energia Atomica (CNEA), due to their joint responsibility in the management of the SF, have proposed interim dry storage systems.These systems have to be evaluated in order to choose one of them by the end of 2006.In this work the Monte Carlo code MCNP was used to make the criticality and shielding calculations corresponding to the model proposed by CNEA.This model suggests the store of sealed containers with 36 or 37 SF in concrete modules.Each one of the containers is filled in the pool houses and transported to the module in a transference cask with lead walls.The results of the criticality calculations indicates that the solutions of SF proposed have widely fulfilled the requirements of subcriticality, even in supposed extreme accidental situations.Regarding the transference cask, the SF dose rate estimations allow us to make a feedback for the design aiming to the geometry and shielding improvements.Regarding the store modules, thicknesses ranges of concrete walls are suggested in order to fulfill the dose requirements stated by the Autoridad Regulatoria Nuclear Argentina [es

  9. 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

    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.)

  10. Spin-orbit ZORA and four-component Dirac-Coulomb estimation of relativistic corrections to isotropic nuclear shieldings and chemical shifts of noble gas dimers

    Jankowska, Marzena; Kupka, Teobald; Stobiński, Leszek

    2016-01-01

    Hartree-Fock and density functional theory with the hybrid B3LYP and general gradient KT2 exchange-correlation functionals were used for non-relativistic and relativistic nuclear magnetic shielding calculations of helium, neon, argon, krypton and xenon dimers and free atoms. Relativistic...

  11. Development of a pencil-type single shield graphite quasi-adiabatic calorimeter and comparison of its performance with a double-shield graphite calorimeter for the measurement of nuclear heat deposition rate in a fusion environment

    Joneja, O.P.; Rosselet, M.; Ligou, J.; Gardel, P.

    1995-01-01

    Recently, heat deposition rate measurements were reported that used a quasi-adiabatic double-shield graphite calorimeter. It was found that for a better understanding of nuclear heating due to incident radiation, having a calorimeter that could be conveniently moved axially and radially inside large material blocks would be advisable. Here, a simpler design, based on three elements, i.e., core, jacket, and shield is conceived. The fabrication and testing details are presented, and the performance of the current calorimeter is compared with a double-shield calorimeter under similar conditions. Such a system is found to be extremely sensitive and can be employed successfully at the LOTUS facility for future nuclear heat deposition rate measurements in large blocks of materials. The current design paves the way for the convenient testing of a large amount of kerma factor data required for constructing future fusion machines. The same configuration with minor changes can be extended to most of the fusion materials of interest. The core of the new calorimeter measures 11 mm in diameter and height and has overall dimensions of 24 mm in diameter and 180 mm in height. The response of the calorimeter is measured by placing it in front of the Haefely neutron generator. 12 refs., 16 figs., 9 tabs

  12. The Shield or the Sword? Revisiting the Debate on Racial Disproportionality in Special Education and Implications for School Psychologists

    Sullivan, Amanda L.; Proctor, Sherrie L.

    2016-01-01

    Scholars in special education and school psychology are engaged in renewed debate about the disproportionate representation of students from culturally and linguistically diverse backgrounds in special education following research and commentaries challenging long held assumptions that many students are inappropriately identified with special…

  13. Shielding practice

    Sauermann, P.F.

    1985-08-01

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

  14. Spin-orbit ZORA and four-component Dirac-Coulomb estimation of relativistic corrections to isotropic nuclear shieldings and chemical shifts of noble gas dimers.

    Jankowska, Marzena; Kupka, Teobald; Stobiński, Leszek; Faber, Rasmus; Lacerda, Evanildo G; Sauer, Stephan P A

    2016-02-05

    Hartree-Fock and density functional theory with the hybrid B3LYP and general gradient KT2 exchange-correlation functionals were used for nonrelativistic and relativistic nuclear magnetic shielding calculations of helium, neon, argon, krypton, and xenon dimers and free atoms. Relativistic corrections were calculated with the scalar and spin-orbit zeroth-order regular approximation Hamiltonian in combination with the large Slater-type basis set QZ4P as well as with the four-component Dirac-Coulomb Hamiltonian using Dyall's acv4z basis sets. The relativistic corrections to the nuclear magnetic shieldings and chemical shifts are combined with nonrelativistic coupled cluster singles and doubles with noniterative triple excitations [CCSD(T)] calculations using the very large polarization-consistent basis sets aug-pcSseg-4 for He, Ne and Ar, aug-pcSseg-3 for Kr, and the AQZP basis set for Xe. For the dimers also, zero-point vibrational (ZPV) corrections are obtained at the CCSD(T) level with the same basis sets were added. Best estimates of the dimer chemical shifts are generated from these nuclear magnetic shieldings and the relative importance of electron correlation, ZPV, and relativistic corrections for the shieldings and chemical shifts is analyzed. © 2015 Wiley Periodicals, Inc.

  15. Special Radiation Protection Precautions in Therapeutic Nuclear Medicine

    Stefanoyiannis, A. P.; Gerogiannis, J.

    2010-01-01

    Therapeutic Nuclear Medicine concerns the administration of appropriate amounts of radioactivity of certain isotopes, in order to achieve internal localized irradiation of neoplasmatic cells. Due to the increased level and the specific isotope characteristics of administered radioactivity, special Radiation Protection precautions must be taken. This study addresses such issues, based on national as well as international legislation and guidelines. Application of the principle of optimization is of outmost importance and is based on individual dose planning. The decision about the release of Nuclear Medicine patients after therapy is determined on an individual basis, taking into account patients' pattern of contact with other people, their age and that of persons in the home environment, in addition to other factors. Estimation of the absorbed dose given to the treated organ is based on uptake measurements and other biokinetic data, as well as on the mass of the treated tissue or organ. Concerning pregnant women, the rule of thumb is that they should not be treated, unless the radionuclide therapy is required to save their lives. In that case, the potential absorbed dose and risk to the foetus should be estimated and conveyed to the patient. After radionuclide therapy, a female should be advised to avoid pregnancy for the period of time depending on the specific radionuclide. This is to ensure that the dose to a conceptus/foetus would probably not exceed 1 mGy (the member of the public dose limit). The radiation risk for relatives and caregivers is small and unlikely to exceed the legal dose constraints during the period of the patient's treatment. Solid waste from the patient's stay in hospital is a different matter, and is normally incinerated or held for a period until radioactive decay brings the activity to an acceptable level.

  16. Estimation methods for process holdup of special nuclear materials

    Pillay, K.K.S.; Picard, R.R.; Marshall, R.S.

    1984-06-01

    The US Nuclear Regulatory Commission sponsored a research study at the Los Alamos National Laboratory to explore the possibilities of developing statistical estimation methods for materials holdup at highly enriched uranium (HEU)-processing facilities. Attempts at using historical holdup data from processing facilities and selected holdup measurements at two operating facilities confirmed the need for high-quality data and reasonable control over process parameters in developing statistical models for holdup estimations. A major effort was therefore directed at conducting large-scale experiments to demonstrate the value of statistical estimation models from experimentally measured data of good quality. Using data from these experiments, we developed statistical models to estimate residual inventories of uranium in large process equipment and facilities. Some of the important findings of this investigation are the following: prediction models for the residual holdup of special nuclear material (SNM) can be developed from good-quality historical data on holdup; holdup data from several of the equipment used at HEU-processing facilities, such as air filters, ductwork, calciners, dissolvers, pumps, pipes, and pipe fittings, readily lend themselves to statistical modeling of holdup; holdup profiles of process equipment such as glove boxes, precipitators, and rotary drum filters can change with time; therefore, good estimation of residual inventories in these types of equipment requires several measurements at the time of inventory; although measurement of residual holdup of SNM in large facilities is a challenging task, reasonable estimates of the hidden inventories of holdup to meet the regulatory requirements can be accomplished through a combination of good measurements and the use of statistical models. 44 references, 62 figures, 43 tables

  17. 10 CFR 70.20a - General license to possess special nuclear material for transport.

    2010-01-01

    ... 10 Energy 2 2010-01-01 2010-01-01 false General license to possess special nuclear material for transport. 70.20a Section 70.20a Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) DOMESTIC LICENSING OF... transport. (a) A general license is issued to any person to possess formula quantities of strategic special...

  18. Simulations and imaging algorithm development for a cosmic ray muon tomography system for the detection of special nuclear material in transport containers

    Jewett, C.; Anghel, V.N.P.; Armitage, J.; Boudjemline, K.; Botte, J.; Bryman, D.; Bueno, J.; Charles, E.; Cousins, T.; Didsbury, R.; Erhardt, L.; Erlandson, A.; Gallant, G.; Jason, A.; Jonkmans, G.; Liu, Z.; McCall, M.; Noel, S.; Oakham, F.G.; Ong, D.; Stocki, T.; Thompson, M.; Waller, D.

    2011-01-01

    The Cosmic Ray Inspection and Passive Tomography (CRIPT) collaboration is developing a cosmic ray muon tomography system to identify Special Nuclear Materials (SNM) in cargo containers. In order to gauge the viability of the technique, and to determine the best detector type, GEANT4 was used to simulate the passage of cosmic ray muons through a cargo container. The scattering density estimation (SDE) algorithm was developed and tested with data from these simulations to determine how well it could reconstruct the interior of a container. The simulation results revealed the ability of cosmic ray muon tomography techniques to image spheres of lead-shielded Special Nuclear Materials (SNM), such as uranium or plutonium, in a cargo container, containing a cargo of granite slabs. (author)

  19. Simulations and imaging algorithm development for a cosmic ray muon tomography system for the detection of special nuclear material in transport containers

    Jewett, C.; Anghel, V.N.P. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada); Armitage, J.; Boudjemline, K.; Botte, J. [Carleton Univ., Dept. of Physics, Ottawa, Ontario (Canada); Bryman, D. [Advanced Applied Physics Solutions, Vancouver, British Columbia (Canada); Univ. of British Columbia, Vancouver, British Columbia (Canada); Bueno, J. [Advanced Applied Physics Solutions, Vancouver, British Columbia (Canada); Charles, E. [Canada Border Services Agency, Ottawa, Ontario (Canada); Cousins, T. [International Safety Research, Ottawa, Ontario (Canada); Didsbury, R. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada); Erhardt, L. [Defence Research and Development Canada, Ottawa, Ontario (Canada); Erlandson, A. [Carleton Univ., Dept. of Physics, Ottawa, Ontario (Canada); Gallant, G. [Canada Border Services Agency, Ottawa, Ontario (Canada); Jason, A. [Los Alamos National Laboratory, Los Alamos (United States); Jonkmans, G. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada); Liu, Z. [Advanced Applied Physics Solutions, Vancouver, British Columbia (Canada); Univ. of British Columbia, Vancouver, British Columbia (Canada); McCall, M.; Noel, S. [International Safety Research, Ottawa, Ontario (Canada); Oakham, F.G. [Carleton Univ., Dept. of Physics, Ottawa, Ontario (Canada); TRIUMF, Vancouver, British Columbia, (Canada); Ong, D.; Stocki, T. [Health Canada, Ottawa, Ontario (Canada); Thompson, M. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada); Waller, D. [Defence Research and Development Canada, Ottawa, Ontario (Canada)

    2011-07-01

    The Cosmic Ray Inspection and Passive Tomography (CRIPT) collaboration is developing a cosmic ray muon tomography system to identify Special Nuclear Materials (SNM) in cargo containers. In order to gauge the viability of the technique, and to determine the best detector type, GEANT4 was used to simulate the passage of cosmic ray muons through a cargo container. The scattering density estimation (SDE) algorithm was developed and tested with data from these simulations to determine how well it could reconstruct the interior of a container. The simulation results revealed the ability of cosmic ray muon tomography techniques to image spheres of lead-shielded Special Nuclear Materials (SNM), such as uranium or plutonium, in a cargo container, containing a cargo of granite slabs. (author)

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

    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.)

  1. Superheated emulsions for the detection of special nuclear material

    D’Errico, Francesco; Di Fulvio, Angela

    2011-01-01

    A novel solution for the detection and smuggling interdiction of special nuclear materials is presented here consisting of large detector modules which contain superheated emulsions and which are readout with an optical approach. The detectors can be produced to be fully sensitive to prompt fission neutrons and totally insensitive to the interrogation beam, whether X-rays or neutrons below a chosen energy threshold. Therefore, the detectors are able to operate while the selected interrogation beam is on and they will only pick up the signal from fission neutrons. A position-sensitive readout mechanism is used in our design, relying on the scattering of light by neutron-induced bubbles. A beam of coherent light crosses the active area of the detector, and local variations in scattered light due to the presence of bubbles are detected in real time by arrays of silicon planar photodiodes affixed along the whole length of the detector. The system may offer a variety of advantages compared to current approaches, such as the possibility of simultaneous irradiation and detection, i.e. a 100% duty cycle, without requiring complex signal analysis, and high signal-to-noise ratio, minimizing costly nuisance alarms, thanks to its inherent insensitivity to photons.

  2. Internal transfers of special nuclear material - March 1975

    Anon.

    1976-01-01

    Paragraph 70.51(e) of 10 CFR Part 70 requires, with certain exceptions stated in the rule, that each licensee authorized to possess more than one effective kilogram of special nuclear material (SNM) maintain certain procedures. These procedures are to include: (1) records of the quantities of SNM added to or removed from the process; (2) documentation of all transfers of SNM between material-balance areas to show the identity and quantity of SNM transferred; (3) requirements for authorized signatures on each document used to record the transfer of SNM between material-balance areas; and (4) means for control of and accounting for internal transfer documents. Paragraph 70.58(e) requires licensees to establish, maintain, and follow a system for measuring the SNM transferred between material-balance areas and item-control areas. Paragraph 70.58(f) requires that licensees have a program that evaluates and controls the quality of their measurement system. Additionally, all licensees authorized to possess SNM must comply with paragraph 70.51(b) of 10 CFR Part 70. That rule requires licensees to keep records showing, among other things, the inventory of all SNM in their possession and its location. This guide sets forth acceptable methods for controlling and documenting transfers of SNM within a plant site in order to meet the requirements listed above

  3. Tamper and radiation resistant instrumentation for safeguarding special nuclear materials

    Parsons, B.B.; Wells, J.L.

    1977-01-01

    A tamper-resistant liquid level/accountability instrumentation system for safeguards use has been developed and tested. The tests demonstrate the accuracy of liquid level measurement using TDR (Time Domain Reflectometry) techniques and the accuracy of differential pressure and temperature measurements utilizing a custom designed liquid level sensor probe. The calibrated liquid level, differential pressure, and temperature data provide sufficient information to accurately determine volume, density, and specific gravity. Test solutions used include ordinary tap water, diluted nitric acid in varying concentrations, and diluted uranium trioxide also in varying concentrations. System operations and preliminary test results conducted at the General Electric Midwest Fuel Recovery Plant and the National Bureau of Standards, respectively, suggest that the system will provide the safeguards inspector with an additional tool for real-time independent verification of normal operations and special nuclear materials accountancy data for chemical reprocessing plants. This paper discusses the system design concepts, including a brief description of the tamper and radiation resistant features, the preliminary test results, and the significance of the work

  4. Special Nuclear Material Portal Monitoring at the Nevada Test Site

    Mike Murphy

    2008-01-01

    In the past, acceptance and performance testing of the various Special Nuclear Material (SNM) monitoring devices at the Nevada Test Site has been performed by the Radiological Health Instrumentation Department. Calibration and performance tests on the PM-700 personnel portal monitor were performed but there was no test program for the VM-250 vehicle portal monitor because it had never been put into service. The handheld SNM monitors, the TSA model 470B, were being calibrated annually, but there was no program in place to test them quarterly. In April of 2007, the Material Control and Accountability (MC and A) Manager at the time decided that the program needed to be strengthened and MC and A took over performance testing of all SNM portal monitoring equipment. This paper will discuss the following activities associated with creating a performance testing program: changing the culture, learning the systems, writing procedures, troubleshooting/repairing, validating the process, control of equipment, acquisition of new systems, and running the program

  5. Selected topics in special nuclear materials safeguard system design

    King, L.L.; Thatcher, C.D.; Clarke, J.D.; Rodriguez, M.P.

    1991-01-01

    During the past two decades the improvements in circuit integration have given rise to many new applications in digital processing technology by continuously reducing the unit cost of processing power. Along with this increase in processing power a corresponding decrease in circuit volume has been achieved. Progress has been so swift that new classes of applications become feasible every 2 or 3 years. This is especially true in the application of proven new technology to special nuclear materials (SNM) safeguard systems. Several areas of application were investigated in establishing the performance requirements for the SNM safeguard system. These included the improvements in material control and accountability and surveillance by using multiple sensors to continuously monitor SNM inventory within the selected value(s); establishing a system architecture to provide capabilities needed for present and future performance requirements; and limiting operating manpower exposure to radiation. This paper describes two selected topics in the application of state-of-the-art, well-proven technology to SNM safeguard system design

  6. Special Nuclear Material Detection with a Water Cherenkov based Detector

    Sweany, M.; Bernstein, A.; Bowden, N.; Dazeley, S.; Svoboda, R.

    2008-01-01

    Fission events from Special Nuclear Material (SNM), such as highly enriched uranium or plutonium, produce a number of neutrons and high energy gamma-rays. Assuming the neutron multiplicity is approximately Poissonian with an average of 2 to 3, the observation of time correlations between these particles from a cargo container would constitute a robust signature of the presence of SNM inside. However, in order to be sensitive to the multiplicity, one would require a high total efficiency. There are two approaches to maximize the total efficiency; maximizing the detector efficiency or maximizing the detector solid angle coverage. The advanced detector group at LLNL is investigating one way to maximize the detector size. We are designing and building a water Cerenkov based gamma and neutron detector for the purpose of developing an efficient and cost effective way to deploy a large solid angle car wash style detector. We report on our progress in constructing a larger detector and also present preliminary results from our prototype detector that indicates detection of neutrons

  7. Oscillatory vapour shielding of liquid metal walls in nuclear fusion devices

    van Eden, G.G.; Kvon, V.; Van De Sanden, M.C.M.; Morgan, T.W.

    2017-01-01

    Providing an efficacious plasma facing surface between the extreme plasma heat exhaust and the structural materials of nuclear fusion devices is a major challenge on the road to electricity production by fusion power plants. The performance of solid plasma facing surfaces may become critically

  8. Electromagnetic shielding

    Tzeng, Wen-Shian V.

    1991-01-01

    Electromagnetic interference (EMI) shielding materials are well known in the art in forms such as gaskets, caulking compounds, adhesives, coatings and the like for a variety of EMI shielding purposes. In the past, where high shielding performance is necessary, EMI shielding has tended to use silver particles or silver coated copper particles dispersed in a resin binder. More recently, aluminum core silver coated particles have been used to reduce costs while maintaining good electrical and physical properties. (author). 8 figs

  9. Application of gas shielded arc welding and submerged arc welding for fabrication of nuclear reactor vessels

    Gehani, M.L.; Rodrigues, W.D.

    1976-01-01

    The remarkable progress made in the development of knowhow and expertise in the manufacture of equipment for nuclear power plants in India is outlined. Some of the specific advances made in the application of higher efficiency weld processes for fabrication of nuclear reactor vessels and the higher level of quality attained are discussed in detail. Modifications and developments in submerged arc, gas tungsten arc and gas metal arc processes for welding of Calandria which have been a highly challenging and rewarding experience are discussed. Future scope for making the gas metal arc process more economical by using various gas-mixes like Agron + Oxygen, Argon + Carbon Dioxide, Argon + Nitrogen (for Copper Alloys) etc., in various proportions are outlined. Quality and dimensional control exercised in these jobs of high precision are highlighted. (K.B.)

  10. The SGR Multipurpose - Generation IV - Transportable Cogeneration Nuclear Reactor with Innovative Shielding

    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)

  11. Special nuclear materials cutoff exercise: Issues and lessons learned. Volume 3

    Libby, R.A.; Segal, J.E.; Stanbro, W.D.; Davis, C.

    1995-08-01

    This document is appendices D-J for the Special Nuclear Materials Cutoff Exercise: Issues and Lessons Learned. Included are discussions of the US IAEA Treaty, safeguard regulations for nuclear materials, issue sheets for the PUREX process, and the LANL follow up activity for reprocessing nuclear materials.

  12. Special nuclear materials cutoff exercise: Issues and lessons learned. Volume 3

    Libby, R.A.; Segal, J.E.; Stanbro, W.D.; Davis, C.

    1995-08-01

    This document is appendices D-J for the Special Nuclear Materials Cutoff Exercise: Issues and Lessons Learned. Included are discussions of the US IAEA Treaty, safeguard regulations for nuclear materials, issue sheets for the PUREX process, and the LANL follow up activity for reprocessing nuclear materials

  13. Nuclear Rocket Facility Decommissioning Project: Controlled Explosive Demolition of Neutron-Activated Shield Wall

    Michael R. Kruzic

    2007-01-01

    Located in Area 25 of the Nevada Test Site (NTS), the Test Cell A (TCA) Facility was used in the early to mid-1960s for the testing of nuclear rocket engines, as part of the Nuclear Rocket Development Program, to further space travel. Nuclear rocket testing resulted in the activation of materials around the reactors and the release of fission products and fuel particles in the immediate area. Identified as Corrective Action Unit 115, the TCA facility was decontaminated and decommissioned (D and D) from December 2004 to July 2005 using the Streamlined Approach for Environmental Restoration (SAFER) process, under the ''Federal Facility Agreement and Consent Order''. The SAFER process allows environmental remediation and facility closure activities (i.e., decommissioning) to occur simultaneously provided technical decisions are made by an experienced decision maker within the site conceptual site model, identified in the Data Quality Objective process. Facility closure involved a seven-step decommissioning strategy. Key lessons learned from the project included: (1) Targeted preliminary investigation activities provided a more solid technical approach, reduced surprises and scope creep, and made the working environment safer for the D and D worker. (2) Early identification of risks and uncertainties provided opportunities for risk management and mitigation planning to address challenges and unanticipated conditions. (3) Team reviews provided an excellent mechanism to consider all aspects of the task, integrated safety into activity performance, increase team unity and ''buy-in'' and promoted innovative and time saving ideas. (4) Development of CED protocols ensured safety and control. (5) The same proven D and D strategy is now being employed on the larger ''sister'' facility, Test Cell C

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

    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

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

    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)

  16. Special safety requirements applied to Brazilian nuclear power plant

    Lepecki, W.P.S.; Hamel, H.J.E.; Koenig, N.; Vieira, P.C.R.; Fritzsche, J.C.

    1981-01-01

    Some safety aspects of the Angra 2 and 3 nuclear power plants are presented. An analysis of the civil and mechanical project of these nuclear power plant having in view a safety analysis is done. (E.G.) [pt

  17. Nuclear energy: the real cost. A special report

    Spencer, K.; Marshall, R.; Sweet, C.; Prior, M.; Welsh, I.; Bunyard, P.; Goldsmith, E.; Hildyard, N.; Jeffery, J.W. (Committee for the Study of the Economics of Nuclear Electricity, Camelford (UK))

    1981-12-01

    This report on the discussions within a small group of academics falls under the headings: chairman's foreword; summary and recommendations; the government's nuclear power programme and its implications; the CEGB's planning record; the past performance of Britain's nuclear power stations - a guide for the future (query); nuclear power -early uncertainties; historic costs - 'the fraud inherent in all inflationary finance'; current cost accounting; fuel costs - coal stays steady, nuclear rises; net effective cost and the rationale for nuclear power; reinterpreting net effective costs; other considerations; conclusions and recommendations; references.

  18. A study of thermal, structural and shielding safety analysis for dry storage of spent nuclear fuel

    Shin, S. H. [Kyungpook Nationl Univ., Daegu (Korea, Republic of)

    1997-03-15

    As a replaced method for MRS, the dry storage has been intensively developed by the advanced countries of nuclear power technology. Currently, the domestic technology for the dry storage is also under development. In the present study, the developed technical standards for USNRC and its operation are summarized. Futhermore, the SAR for VECTRA's NUHOMES satisfied with DOE and NRC's requirements is inversely analyzed and combined with both USNRC's regulatory guide and LLNL's SARS. In the safety analysis of a dry storage, the principal design criteria which identifies the structural and mechanical safety criteria is investigated. Based on the design criteria, hypothetical accident analysis as well as off-normal operation analysis are investigated.

  19. Estimation of isotropic nuclear magnetic shieldings in the CCSD(T) and MP2 complete basis set limit using affordable correlation calculations

    Kupka, Teobald; Stachów, Michal; Kaminsky, Jakub

    2013-01-01

    , estimated from calculations with the family of polarizationconsistent pcS-n basis sets is reported. This dependence was also supported by inspection of profiles of deviation between CBS estimated nuclear shieldings and obtained with significantly smaller basis sets pcS-2 and aug-cc-pVTZ-J for the selected......A linear correlation between isotropic nuclear magnetic shielding constants for seven model molecules (CH2O, H2O, HF, F2, HCN, SiH4 and H2S) calculated with 37 methods (34 density functionals, RHF, MP2 and CCSD(T) ), with affordable pcS-2 basis set and corresponding complete basis set results...... set of 37 calculation methods. It was possible to formulate a practical approach of estimating the values of isotropic nuclear magnetic shielding constants at the CCSD(T)/CBS and MP2/CBS levels from affordable CCSD(T)/pcS-2, MP2/pcS-2 and DFT/CBS calculations with pcS-n basis sets. The proposed method...

  20. Shielding plugs

    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.)

  1. Simple Design Method for Magnetic Shield Room(<Special Issue>The 20th MAGDA Conference in Pacific Asia (MAGDA2011))

    Keisuke, FUJISAKI; Masahiro, FUJIKURA; Jirou, MINO; Nippon Steel Corporation:Toyota Technological Institute; Nippon Steel Corporation; Nippon Steel Engineering Co., Ltd

    2012-01-01

    As a primary evaluation of the magnetic shield room design, a simple design method is proposed here. The steel sheet number of the magnetic shield room is designed so as to make the distributed magnetic flux density from the exciting coil catch the magnetic shield body. The proposed method is applied to a full-scale magnetic shield room and the leakage magnetic flux density is evaluated by numerical calculation. Though it introduces a large steel number of the magnetic shield body, the leakag...

  2. Structural response of nuclear containment shield buildings with unanticipated construction openings

    Mac Namara, Sinead Caitriona

    As Nuclear Power Plants age many require steam generator replacement. There is a nickel alloy in the steam generator tubes that is susceptible to stress cracking and although these cracks can be sealed the generator becomes uneconomical without 10%-15% of the tubes. The steam generator in a typical nuclear power plant is housed in the containment structure next to the reactor. The equipment hatch is not big enough to facilitate steam generator replacement, thus construction openings in the dome of the containment structure are required. To date the structural consequences of construction openings in the dome have not been examined. This thesis examines the effects of such openings. The prototype concrete dome is made up of a 2 ft thick dome atop 3 ft thick and 170 ft high cylindrical walls (radius 65.5 ft) with a tension ring 15 ft high and 8 ft thick in between. The dome of the building is cast in two layers; a lower 9 inch layer that serves as the formwork for an upper 15 inch layer. The weight of the dome is carried in axial compression along the hoops and meridians of the dome. The first finite element model uses shell elements and considers two limiting load cases; where the two layers act as one, and where the lower layer carries the weight of both. The openings interrupt the hoops and meridians and the weight of the dome must be redistributed around the openings. Without openings, the stresses due to dead load in the structure are very low when compared to the material strength. The impact of the openings is increased compression stresses near the opening. The maximum stresses are approximately four times larger than in the original structure. These results are confirmed by the second model which is made from layers of solid elements. This model shows a significant difference between the compression on the top surface of the dome, in the affected areas, and that on the bottom surface, leading to shear stresses. These shear stresses are largest around the

  3. Effect of nuclear shielding in collision of positive charged helium ions with helium atoms

    Ghavaminia, Hoda; Ghavaminia, Shirin

    2018-03-01

    Differential in angle and absolute cross sections in energy of the scattered particles are obtained for single charge exchange in ^3He^+-^4He collisions by means of the four body boundary-corrected first Born approximation (CB1-4B). The quantum-mechanical post and prior transition amplitudes are derived in terms of two-dimensional real integrals in the case of the prior form and five-dimensional quadratures for the post form. The effect of the dynamic electron correlation through the complete perturbation potential and the nuclear-screening influence of the passive electrons on the electron capture process is investigated. The results obtained in the CB1-4B method are compared with the available experimental data. For differential cross sections, the present results are in better agreement with experimental data than other theoretical data at extreme forward scattering angles. The integral cross sections are in excellent agreement with the experiment. Also, total cross sections for single electron capture, has been investigated using the classical trajectory Monte Carlo method. The present calculated results are found to be in an excellent agreement with the experimental data.

  4. Defense Special Weapons Agency Advisory Panel on the Nuclear Weapon Effects Program

    1998-01-01

    We performed the audit in response to allegations made to the Defense Hotline concerning conflicts of interest among members of the Defense Special Weapons Agency Advisory Panel on the Nuclear Weapon Effects Program...

  5. Shielding analysis in the design phase of the new Emergency Operation Facility for Tihange Nuclear Power Plant

    Genard, Gilles; Portal, Romain; Bouchat, Virginie; Vanderperre, Serge

    2017-09-01

    In the framework of the design studies for a new Emergency Operation Facility (EOF) for Tihange NPP, radiation protection analyses are needed to comply with effective dose rate criteria. In this aim, the shielding performed by, at the one hand, external walls and roof of the building and, on the other hand, internal walls, has been sized by means of MicroShield calculations. This paper explains how the calculations for external walls, doors, roof, floor and internal walls are made. The assumptions on the source terms and on the source geometry as well as the way the shielding is determined and the results of the sizing are presented.

  6. Shielding analysis in the design phase of the new Emergency Operation Facility for Tihange Nuclear Power Plant

    Genard Gilles

    2017-01-01

    Full Text Available In the framework of the design studies for a new Emergency Operation Facility (EOF for Tihange NPP, radiation protection analyses are needed to comply with effective dose rate criteria. In this aim, the shielding performed by, at the one hand, external walls and roof of the building and, on the other hand, internal walls, has been sized by means of MicroShield calculations. This paper explains how the calculations for external walls, doors, roof, floor and internal walls are made. The assumptions on the source terms and on the source geometry as well as the way the shielding is determined and the results of the sizing are presented.

  7. Nuclear Magnetic Shielding Constants from Quantum Mechanical/Molecular Mechanical Calculations Using Polarizable Embedding: Role of the Embedding Potential

    Steinmann, Casper; Olsen, Jógvan Magnus Haugaard; Kongsted, Jacob

    2014-01-01

    We present NMR shielding constants obtained through quantum mechanical/molecular mechanical (QM/MM) embedding calculations. Contrary to previous reports, we show that a relatively small QM region is sufficient, provided that a high-quality embedding potential is used. The calculated averaged NMR...... shielding constants of both acrolein and acetone solvated in water are based on a number of snapshots extracted from classical molecular dynamics simulations. We focus on the carbonyl chromophore in both molecules, which shows large solvation effects, and we study the convergence of shielding constants...

  8. Radiation protection/shield design

    Disney, R.K.

    1977-01-01

    Radiation protection/shielding design of a nuclear facility requires a coordinated effort of many engineering disciplines to meet the requirements imposed by regulations. In the following discussion, the system approach to Clinch River Breeder Reactor Plant (CRBRP) radiation protection will be described, and the program developed to implement this approach will be defined. In addition, the principal shielding design problems of LMFBR nuclear reactor systems will be discussed in realtion to LWR nuclear reactor system shielding designs. The methodology used to analyze these problems in the U.S. LMFBR program, the resultant design solutions, and the experimental verification of these designs and/or methods will be discussed. (orig.) [de

  9. Improved design for vibration-proof platinum RTD in special position of nuclear power plant

    Liu Zhuo; Ma Jinna; Wu Bin

    2014-01-01

    In accordance with the actual situation for the vibration of violence in a special position of nuclear power plant, an improved design for platinum RTD was proposed. The structure design is verified to meet the measure requirement in the special position. (authors)

  10. Investigations of active interrogation techniques to detect special nuclear material in maritime environments: Standoff interrogation of small- and medium-sized cargo ships

    Miller, Thomas M.; Patton, Bruce W.; Grogan, Brandon R.; Henkel, James J.; Murphy, Brian D.; Johnson, Jeffrey O.; Mihalczo, John T.

    2013-01-01

    In this work, several active interrogation (AI) sources are evaluated to determine their usefulness in detecting the presence of special nuclear material (SNM) in fishing trawlers, small cargo transport ships, and luxury yachts at large standoff distances from the AI source and detector. This evaluation is performed via computational analysis applying Monte Carlo methods with advanced variance reduction techniques. The goal is to determine the AI source strength required to detect the presence of SNM. The general conclusion of this study is that AI is not reliable when SNM is heavily shielded and not tightly coupled geometrically with the source and detector, to the point that AI should not be considered a via interrogation option in these scenarios. More specifically, when SNM is shielded by hydrogenous material large AI source strengths are required if detection is based on neutrons, which is not surprising. However, if the SNM is shielded by high-Z material the required AI source strengths are not significantly different if detection is based on neutrons or photons, which is somewhat surprising. Furthermore, some of the required AI source strengths that were calculated are very large. These results coupled with the realities of two ships moving independently at sea and other assumptions made during this analysis make the use of standoff AI in the maritime environment impractical

  11. Investigations of active interrogation techniques to detect special nuclear material in maritime environments: Standoff interrogation of small- and medium-sized cargo ships

    Miller, Thomas M., E-mail: millertm@ornl.gov; Patton, Bruce W.; Grogan, Brandon R.; Henkel, James J.; Murphy, Brian D.; Johnson, Jeffrey O.; Mihalczo, John T.

    2013-12-01

    In this work, several active interrogation (AI) sources are evaluated to determine their usefulness in detecting the presence of special nuclear material (SNM) in fishing trawlers, small cargo transport ships, and luxury yachts at large standoff distances from the AI source and detector. This evaluation is performed via computational analysis applying Monte Carlo methods with advanced variance reduction techniques. The goal is to determine the AI source strength required to detect the presence of SNM. The general conclusion of this study is that AI is not reliable when SNM is heavily shielded and not tightly coupled geometrically with the source and detector, to the point that AI should not be considered a via interrogation option in these scenarios. More specifically, when SNM is shielded by hydrogenous material large AI source strengths are required if detection is based on neutrons, which is not surprising. However, if the SNM is shielded by high-Z material the required AI source strengths are not significantly different if detection is based on neutrons or photons, which is somewhat surprising. Furthermore, some of the required AI source strengths that were calculated are very large. These results coupled with the realities of two ships moving independently at sea and other assumptions made during this analysis make the use of standoff AI in the maritime environment impractical.

  12. Accounting systems for special nuclear material control. Technical report

    Korstad, P.A.

    1980-05-01

    Nuclear material accounting systems were examined and compared to financial double-entry accounting systems. Effective nuclear material accounting systems have been designed using the principles of double-entry financial accounting. The modified double-entry systems presently employed are acceptable if they provide adequate control over the recording and summarizing of transactions. Strong internal controls, based on principles of financial accounting, can help protect nuclear materials and produce accurate, reliable accounting data. An electronic data processing system can more accurately maintain large volumes of data and provide management with more current, reliable information

  13. 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

    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)

  14. 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

    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)

  15. Shielding analysis in the design phase of the new Emergency Operation Facility for Tihange Nuclear Power Plant

    Genard Gilles; Portal Romain; Bouchat Virginie; Vanderperre Serge

    2017-01-01

    In the framework of the design studies for a new Emergency Operation Facility (EOF) for Tihange NPP, radiation protection analyses are needed to comply with effective dose rate criteria. In this aim, the shielding performed by, at the one hand, external walls and roof of the building and, on the other hand, internal walls, has been sized by means of MicroShield calculations. This paper explains how the calculations for external walls, doors, roof, floor and internal walls are made. The assump...

  16. Real-Time Characterization of Special Nuclear Materials

    Walston, Sean [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Candy, Jim [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Chambers, Dave [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Chandrasekaran, Hema [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Snyderman, Neal [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

    2015-09-04

    When confronting an item that may contain nuclear material, it is urgently necessary to determine its characteristics. Our goal is to provide accurate information with high-con dence as rapidly as possible.

  17. Shielding container

    Darling, K.A.M.

    1981-01-01

    A shielding container incorporates a dense shield, for example of depleted uranium, cast around a tubular member of curvilinear configuration for accommodating a radiation source capsule. A lining for the tubular member, in the form of a close-coiled flexible guide, provides easy replaceability to counter wear while the container is in service. Container life is extended, and maintenance costs are reduced. (author)

  18. Absolute NMR shielding scales and nuclear spin–rotation constants in {sup 175}LuX and {sup 197}AuX (X = {sup 19}F, {sup 35}Cl, {sup 79}Br and {sup 127}I)

    Demissie, Taye B., E-mail: taye.b.demissie@uit.no; Komorovsky, Stanislav; Repisky, Michal; Ruud, Kenneth [Centre for Theoretical and Computational Chemistry, Department of Chemistry, UiT The Arctic University of Norway, N-9037 Tromsø (Norway); Jaszuński, Michał [Institute of Organic Chemistry, Polish Academy of Sciences, Kasprzaka 44, 01 224 Warszawa (Poland)

    2015-10-28

    We present nuclear spin–rotation constants, absolute nuclear magnetic resonance (NMR) shielding constants, and shielding spans of all the nuclei in {sup 175}LuX and {sup 197}AuX (X = {sup 19}F, {sup 35}Cl, {sup 79}Br, {sup 127}I), calculated using coupled-cluster singles-and-doubles with a perturbative triples (CCSD(T)) correction theory, four-component relativistic density functional theory (relativistic DFT), and non-relativistic DFT. The total nuclear spin–rotation constants determined by adding the relativistic corrections obtained from DFT calculations to the CCSD(T) values are in general in agreement with available experimental data, indicating that the computational approach followed in this study allows us to predict reliable results for the unknown spin–rotation constants in these molecules. The total NMR absolute shielding constants are determined for all the nuclei following the same approach as that applied for the nuclear spin–rotation constants. In most of the molecules, relativistic effects significantly change the computed shielding constants, demonstrating that straightforward application of the non-relativistic formula relating the electronic contribution to the nuclear spin–rotation constants and the paramagnetic contribution to the shielding constants does not yield correct results. We also analyze the origin of the unusually large absolute shielding constant and its relativistic correction of gold in AuF compared to the other gold monohalides.

  19. Legislation for the countermeasures on special issues of nuclear safety regulations

    Cho, Byung Sun; Lee, Mo Sung; Chung, Gum Chun; Kim, Heon Jin; Oh, Ho Chul

    2004-02-01

    Since the present nuclear safety regulation has some legal problems that refer to special issues and contents of regulatory provisions, this report has preformed research on the legal basic theory of nuclear safety regulation to solve the problems. In addition, this report analyzed the problems of each provisions and suggested the revision drafts on the basis of analyzing problems and the undergoing theory of nuclear safety regulation

  20. Legislation for the countermeasures on special issues of nuclear safety regulations

    Cho, Byung Sun; Lee, Mo Sung; Chung, Gum Chun; Kim, Heon Jin; Oh, Ho Chul [Chongju Univ., Cheongju (Korea, Republic of)

    2004-02-15

    Since the present nuclear safety regulation has some legal problems that refer to special issues and contents of regulatory provisions, this report has preformed research on the legal basic theory of nuclear safety regulation to solve the problems. In addition, this report analyzed the problems of each provisions and suggested the revision drafts on the basis of analyzing problems and the undergoing theory of nuclear safety regulation.

  1. Experience with a servo-hydraulic mechanical testing machine installed in a new shielded active facility at Windscale Nuclear Power Development Laboratories

    Garlick, A.; Hindmarch, P.; Gravenor, J.G.; Rhodes, D.

    1982-03-01

    An Instron model 1273 servo-hydraulic machine has been installed within a lead-shielded cell at Windscale in order to provide a facility capable of performing a wide range of mechanical tests on nuclear reactor structural materials and fuel assembly components. This particular type of machine was chosen because it has design features associated with the load frame, location of the actuator and adjustment and clamping of the cross-head that are especially well suited to remote operation within a shielded cell. The design of the testing facility is described and the programmes of work that have been completed over the past 11/2 years of operation are reviewed. (author)

  2. A method for assay of special nuclear material in high level liquid waste streams

    Venkata Subramani, C.R.; Swaminathan, K.; Asuvathraman, R.; Kutty, K.V.G.

    2003-01-01

    The assay of special nuclear material in the high level liquid waste streams assumes importance as this is the first stage in the extraction cycle and considerable losses of plutonium could occur here. This stream contains all the fission products as also the minor actinides and hence normal nuclear techniques cannot be used without prior separation of the special nuclear material. This paper presents the preliminary results carried out using wavelength dispersive x-ray fluorescence as part of the developmental efforts to assay SNM in these streams by instrumental techniques. (author)

  3. Special role of scientists as citizens in a nuclear age

    Masperi, L.

    1999-01-01

    We are living in an age dominated by science and technology. There are probably at present more scientists alive than all those that passed away through the history of humankind. It is certain that science and technology have the possibility of solving the problems of society. However, these problems have not so far been satisfactorily treated and some scientifically ordered society which might be envisaged should hurt the common humanitarian feeling. These are probably the reasons for which a part of the public opinion has lost its trust in science and many people try to find a relief for their anguish in esoteric religions. It seems therefore necessary for a change of attitude on the part of scientists to restore the positive consideration from the society and to be able to contribute to the future evolution of humankind according to a peaceful and harmonious pattern. This short essay will start describing the historical search for solutions through science, will continue by attempting to define the values which should he added to science in the present time, and will end with possible recommendations for scientists in their connection with society, with particular emphasis on the nuclear issue. All that will be considered here refers mainly to natural sciences. Trying to find which are the attitudes of scientists that could contribute to the benefit of society, one may start with the need of feeling love for the scientific achievements they are able to make. Concerning the specific nuclear issue, both civilian and military applications must be considered. The military applications of nuclear energy should be completely prohibited. Scientists may play a relevant role in the elimination of the nuclear weapon possibility in regions of threshold states or with undeclared arsenals. On the way to nuclear-weapon-free world, it will be crucial to convince all or some of the nuclear powers to dismantle their nuclear arsenals. Scientists may make contribution to the

  4. The Thermos program for nuclear reactors specialized in district heating

    Lerouge, B.

    1976-01-01

    Many studies have been made in France on the use of nuclear heat for district heating. After a brief account of the problems raised by the use of thermal waste from big nuclear power stations, the quantitative and qualitative needs of heating networks are analyzed and the Thermos project described. This is a very robust reactor of the pool type, with an output of 100MW, supplying low-pressure water at 100 deg C. The advantages from the aspects of safety and economy are described, and the present state of the project and its possible developments summarized [fr

  5. Introduction to Special Edition (of the Journal of Nuclear Materials Management) on Reducing the Threat from Radioactive Materials

    Mladineo, Stephen V.

    2007-01-01

    Introductory article for special edition of the JOURNAL OF NUCLEAR MATERIALS MANAGEMENT outlining the Institute of Nuclear Materials Management Nonproliferation and Arms Control Technical Division. In particular the International Nuclear and Radiological Security Standing Committee and its initial focus covering four topical areas--Radiological Threat Reduction, Nuclear Smuggling and Illicit Trafficking, Countering Nuclear Terrorism, and Radiological Terrorism Consequence Management

  6. A Stochastic Imaging Technique for Spatio-Spectral Characterization of Special Nuclear Material

    Hamel, Michael C.

    Radiation imaging is advantageous for detecting, locating and characterizing special nuclear material (SNM) in complex environments. A dual-particle imager (DPI) has been designed that is capable of detecting gamma-ray and neutron signatures from shielded SNM. The system combines liquid organic and NaI(Tl) scintillators to form a combined Compton and neutron scatter camera. Effective image reconstruction of detected particles is a crucial component for maximizing the performance of the system; however, a key deficiency exists in the widely used list-mode maximum-likelihood estimation-maximization (MLEM) image reconstruction technique. The steady-state solution produced by this iterative method will have poor quality compared to solutions produced with fewer iterations. A stopping condition is required to achieve a better solution but these conditions fail to achieve maximum image quality. Stochastic origin ensembles (SOE) imaging is a good candidate to address this problem as it uses Markov chain Monte Carlo to reach a stochastic steady-state solution that has image quality comparable to the best MLEM solution. The application of SOE to the DPI is presented in this work. SOE was originally applied in medical imaging applications with no mechanism to isolate spectral information based on location. This capability is critical for non-proliferation applications as complex radiation environments with multiple sources are often encountered. This dissertation extends the SOE algorithm to produce spatially dependent spectra and presents experimental result showing that the technique was effective for isolating a 4.1-kg mass of weapons grade plutonium (WGPu) when other neutron and gamma-ray sources were present. This work also demonstrates the DPI as an effective tool for localizing and characterizing highly enriched uranium (HEU). A series of experiments were performed with the DPI using a deuterium-deuterium (DD) and deuterium-tritium (DT) neutron generator, as well as

  7. Shielding Benchmark Computational Analysis

    Hunter, H.T.; Slater, C.O.; Holland, L.B.; Tracz, G.; Marshall, W.J.; Parsons, J.L.

    2000-01-01

    Over the past several decades, nuclear science has relied on experimental research to verify and validate information about shielding nuclear radiation for a variety of applications. These benchmarks are compared with results from computer code models and are useful for the development of more accurate cross-section libraries, computer code development of radiation transport modeling, and building accurate tests for miniature shielding mockups of new nuclear facilities. When documenting measurements, one must describe many parts of the experimental results to allow a complete computational analysis. Both old and new benchmark experiments, by any definition, must provide a sound basis for modeling more complex geometries required for quality assurance and cost savings in nuclear project development. Benchmarks may involve one or many materials and thicknesses, types of sources, and measurement techniques. In this paper the benchmark experiments of varying complexity are chosen to study the transport properties of some popular materials and thicknesses. These were analyzed using three-dimensional (3-D) models and continuous energy libraries of MCNP4B2, a Monte Carlo code developed at Los Alamos National Laboratory, New Mexico. A shielding benchmark library provided the experimental data and allowed a wide range of choices for source, geometry, and measurement data. The experimental data had often been used in previous analyses by reputable groups such as the Cross Section Evaluation Working Group (CSEWG) and the Organization for Economic Cooperation and Development/Nuclear Energy Agency Nuclear Science Committee (OECD/NEANSC)

  8. Advanced BorobondTM Shields for Nuclear Materials Containment and BorobondTM Immobilization of Volatile Fission Products - Final CRADA Report

    Wagh, Arun S.

    2016-01-01

    Borobond is a company-proprietary material developed by the CRADA partner in collaboration with Argonne, and is based on Argonne's Ceramicrete technology. It is being used by DOE for nuclear materials safe storage, and Boron Products, LLC is the manufacturer and supplier of Borobond. The major objective of this project was to produce a more versatile composition of this material and find new applications. Major target applications were use for nuclear radiation shields, such as in dry storage casks; use in immobilization of most difficult waste streams, such as Hanford K-Basin waste; use for soluble and volatile fission products, such as Cs, Tc, Sr, and I; and use for corrosion and fire protection applications in nuclear facilities.

  9. DECOMMISSIONING OF SHIELDED FACILITIES AT WINFRITH USED FOR POST IRRADIATION EXAMINATION OF NUCLEAR FUELS and OTHER ACTIVE ITEMS

    Miller, K.D.; Parkinson, S.J.; Cornell, R.M.; Staples, A.T.

    2003-01-01

    This paper describes the approaches used in the clearing, cleaning, decontamination and decommissioning of a very large suite of seven concrete shielded caves and other facilities used by UKAEA at Winfrith Technology Centre, England over a period of about 30 years for the postirradiation examination (PIE) of a wide range of nuclear fuels and other very active components. The basic construction of the facilities will first be described, setting the scene for the major challenges that 1970s' thinking posed for decommissioning engineers. The tendency then to use large and heavy items of equipment supported upon massive steel bench structures produced a series of major problems that had to be overcome. The means of solving these problems by utilization of relatively simple and inexpensive equipment will be described. Later, a further set of challenges was experienced to decontaminate the interior surfaces to allow man entries to be undertaken at acceptable dose rates. The paper will describe the types of tooling used and the range of complementary techniques that were employed to steadily reduce the dose rates down to acceptable levels. Some explanations will also be given for the creation of realistic dose budgets and the methods of recording and continuously assessing the progress against these budgets throughout the project. Some final considerations are given to the commercial approaches to be adopted throughout this major project by the decommissioning engineers. Particular emphasis will be given to the selection of equipment and techniques that are effective so that the whole process can be carried out in a cost-effective and timely manner. The paper also provides brief complementary information obtained during the decommissioning of a plutonium-contaminated facility used for a range of semi-experimental purposes in the late 1970s. The main objective here was to remove the alpha contamination in such a manner that the volume of Plutonium Contaminated Materials (P

  10. 10 CFR 73.25 - Performance capabilities for physical protection of strategic special nuclear material in transit.

    2010-01-01

    ... strategic special nuclear material in transit. 73.25 Section 73.25 Energy NUCLEAR REGULATORY COMMISSION... Transit § 73.25 Performance capabilities for physical protection of strategic special nuclear material in transit. (a) To meet the general performance objective and requirements of § 73.20 an in-transit physical...

  11. Nuclear waste disposal: achieving adequate financing - special study

    Quasebarth, M.V.

    1984-01-01

    An analysis by the Congressional Budget Office (CBO) evaluates whether the current one mill fee now charged to nuclear-electricity consumers will adequately finance the waste disposal program. The CBO found that, if the fee is adjusted annually for inflation, it should provide enough revenues to cover all program costs under all nuclear growth forecasts. If the fee is unchanged, however, the fees will be inadequate if inflation exceeds 3% annually. The report suggests two alternatives for fee revision, but makes no recommendations. The alternatives are to increase the fee only at specific intervals or to automatically adjust the fee through indexation. The report examines the effect of delaying the program, cost overruns, and alternative inflation rate and interest rate assumptions. 3 figures, 12 tables

  12. Electromagnetic shield

    Miller, J.S.

    1987-01-01

    An electromagnetic shield is described comprising: closed, electrically-conductive rings, each having an open center; and binder means for arranging the rings in a predetermined, fixed relationship relative to each other, the so-arranged rings and binder means defining an outer surface; wherein electromagnetic energy received by the shield from a source adjacent its outer surface induces an electrical current to flow in a predetermined direction adjacent and parallel to the outer surface, through the rings; and wherein each ring is configured to cause source-induced alternating current flowing through the portion of the ring closest to the outer surface to electromagnetically induce an oppositely-directed current in the portion of the ring furthest from the surface, such oppositely-directed current bucking any source-induced current in the latter ring portion and thus reducing the magnitude of current flowing through it, whereby the electromagnetic shielding effected by the shield is enhanced

  13. Neutron shieldings

    Tarutani, Kohei

    1979-01-01

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

  14. EEC energy policy guidelines with special reference to nuclear energy

    Brunner, G.

    1978-01-01

    The possibility of a new oil boycott is continuing threat to the European countries: a new oil crisis would put to a severe test not only the economic but also the social system within the community. In the major part of the community power stations cannot be supplied by solar energy and coal fired power stations present a significant environmental burden so that Europe cannot forgo, under any circumstances, the further development of nuclear energy. (orig.) [de

  15. Nuclear and energy. Special issue on the Fukushima power plant

    2011-01-01

    This issue analyses the first consequences of the Fukushima accident at the world level, i.e. impacts which are either already noticeable or predictable. A first article proposes a portrait of Japan (its historical relationship with nature, the cultural education, the role of its bureaucracy, the Japanese business and political worlds) and evokes the nuclear safety organization at the institutional level. It also evokes the different companies involved in nuclear energy production. The second article discusses and comments the environmental and radiological impact of the accident (protection of the inhabitants, environment monitoring, comparison with Chernobyl, main steps of degradation of the reactors, releases in the sea, total release assessment, soil contamination, food contamination, radiation protection). A third article discusses the international impact, notably for the existing or projected power plants in different countries, in terms of public opinion, and with respect to negotiations on climate. The fourth article discusses the reactions of different countries possessing nuclear reactors. The last article questions the replacement of the lost production (that of Fukushima and maybe another power plant) by renewable energies

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

    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...

  17. 10 CFR 74.31 - Nuclear material control and accounting for special nuclear material of low strategic significance.

    2010-01-01

    ... and maintain a measurement system which assures that all quantities in the material accounting records...) In each inventory period, control total material control and accounting measurement uncertainty so... 10 Energy 2 2010-01-01 2010-01-01 false Nuclear material control and accounting for special...

  18. Real-time inventory system for special nuclear material

    Kuckertz, T.H.; Ethridge, C.D.; Nicholson, N.

    1979-01-01

    This paper describes a special purpose peripheral device for a minicomputer that can monitor SNM in vault storage in real time and give timely indication of any tampering with the material. This device, called a shelf monitor, is designed around a single-chip microcomputer, and can be manufactured in quantity for about $100. A typical system of shelf monitors controlled by a minicomputer is described. The minicomputer is used to acquire data associated with the weight and gamma activity of the sample of SNM under observation. Significant deviations in the weight and gamma activity are cause for a tampering alarm.

  19. Real-time inventory system for special nuclear material

    Kuckertz, T.H.; Ethridge, C.D.; Nicholson, N.

    1979-01-01

    This paper describes a special purpose peripheral device for a minicomputer that can monitor SNM in vault storage in real time and give timely indication of any tampering with the material. This device, called a shelf monitor, is designed around a single-chip microcomputer, and can be manufactured in quantity for about $100. A typical system of shelf monitors controlled by a minicomputer is described. The minicomputer is used to acquire data associated with the weight and gamma activity of the sample of SNM under observation. Significant deviations in the weight and gamma activity are cause for a tampering alarm

  20. Microcomputer based shelf system to monitor special nuclear materials in storage

    Nicholson, N.; Kuckertz, T.H.; Ethridge, C.D.

    1980-01-01

    Diversion of special nuclear material has become a matter of grave concern in recent years. Large quantities of this material are kept in long-term storage and must be inventoried periodically, resulting in a time-consuming activity that exposes personnel to additional radiation. A system that provides continuous surveillance of stored special nuclear materials has been developed. A shelf monitor has been designed using a single component microcomputer to collect data from a Geiger Muller tube that monitors gamma emissions and a scale that monitors the total weight of the special nuclear material and its container. A network of these shelf monitors reports their acquired data to a minicomputer for analysis and storage. Because a large number of these monitors is likely to be needed in most storage facilities, one objective of this program has been to develop a low cost but reliable monitor

  1. Semi-annual report on strategic special nuclear material inventory differences

    1981-08-01

    This eighth periodic semiannual report of inventory differences covers the second six months of fiscal year 1980 (April 1, 1980, through September 30, 1980), for the Department of Energy (DOE) and DOE contractor facilities possessing significant quantities of strategic special nuclear material. Strategic special nuclear material is plutonium or uranium-233 or uranium-235 in material whose uranium-235 content is 20 percent or greater (known as highly enriched uranium). A significant quantity is either 2 kilograms of plutonium or uranium-233 or 5 kilograms of uranium-235 in highly enriched uranium or the appropriate weighted combination. All Inventory Differences reported here have been analyzed, investigated when necessary, and resolved. These data and explanations, together with the absences of physical indications of any theft attempt, support a finding that during this period no theft or diversion of strategic special nuclear material has occurred

  2. Land contamination. Technical guidance on special sites: nuclear sites

    Hill, M.; Steeds, J.; Slade, N.

    2001-01-01

    Part IIA of the Environmental Protection Act 1990 sets out a regulatory regime for the identification and remediation of land where contamination is causing unacceptable risks to defined receptors. The Environment Agency has a number of regulatory roles under this regime. Where land is designated as a Special Site, as defined in the Contaminated Land (England) Regulations 2000, the Agency will act as the enforcing authority. It is expected that a similar regime will be introduced in Wales during 2001, but the reader should check whether definitions of Special Sites in the Welsh regulations are the same as in the English ones. The Environment Agency's approach to carrying out its regulatory responsibilities is set out in its Part RA Process Documentation,, available on the Agency website (www.environment-agency. gov.uk). This documentation sets out how the Agency intends to carry out its responsibilities under Part IIA of the Environmental Protection Act 1990, which came into force in England on 1 April 2000

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

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

    1976-01-01

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

  4. Study and application of the ANISN and DOT 3.5 codes to problems in nuclear radiation shielding

    Otto, A.C.

    1983-01-01

    The application of the Sn transport codes ANISN and DOT 3.5 to problems in radiation shielding is reviewed. In addition, a large array of codes involved in radiation shielding calculations is described and applied in this work. The ANISN and DOT 3.5 codes solve the multigroup transport equation in plane, cylindrical and spherical geometries, the first in one dimension and the second in two dimensions, by using the Sn approximation and were designed to solve coupled neutron-photon transport problems commonly found in reactor shielding calculations. In this work the numerical methods used in these codes are reviewed and their basic application to deep-penetration and void problems is discussed. Benchmark problems are solved by employing the array of codes previously mentioned. In particular, the ability of the ISOFLUXO program coupled to the DOT 3.5 code of mapping contours of regions with approximately the same scalar fluxes is illustrated, showing that they can be efficiently used in shielding analysis. (Author) [pt

  5. Vulnerability Analysis Considerations for the Transportation of Special Nuclear Material

    Nicholson, Lary G.; Purvis, James W.

    1999-01-01

    The vulnerability analysis methodology developed for fixed nuclear material sites has proven to be extremely effective in assessing associated transportation issues. The basic methods and techniques used are directly applicable to conducting a transportation vulnerability analysis. The purpose of this paper is to illustrate that the same physical protection elements (detection, delay, and response) are present, although the response force plays a dominant role in preventing the theft or sabotage of material. Transportation systems are continuously exposed to the general public whereas the fixed site location by its very nature restricts general public access

  6. Cover gases in nuclear reactors with special reference to argon

    Jose, C.J.; Shah, G.C.; Prabhu, L.H.; Vartak, D.G.

    1975-01-01

    The report describes the specifications of an ideal cover gas for the smooth operation of a nuclear reactor. The advantages of using helium as cover gas, the sources of impurities in helium cover gas and the methods of purification of helium are given in detail. Various problems associated with the use of argon as cover gas and methods to purify and decontaminate argon cover gas are discussed on the basis of experimental data collected. A laboratory model of the system which can be used to evaluate the performance of the gas purification adsorbents is also described. (author)

  7. Special report: the Reagan administration, Congress, and nuclear power

    Nichols, T.A.; Bink, B.H.

    1982-01-01

    One year ago, the Reagan administration set out to invigorate the light-water-reactor industry; to reactivate the breeder-reactor program; to encourage private reprocessing; and to expedite waste management. The authors feel that, to this end, Congress and the administration are on the brink of major regulatory reforms, a sweeping nuclear-waste-disposal program, and tightening guidelines for burial of low-level waste, all of which are vital to the industry's continued health. This report covers the status of their efforts

  8. Reactor head shielding apparatus

    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

  9. Summaries of special research project on nuclear fusion 1980

    Uchida, Taijiro

    1981-09-01

    This is a report of the research project entitled ''Nuclear fusion'', supported by the grant in aid for fusion research from the Ministry of Education in the fiscal year 1980. The research project was started in April, 1980, and comprises the following seventeen subjects of nuclear fusion research. 1) Heavy irradiation effects, 2) plasma-wall interaction, 3) neutronics, 4) welding engineering, 5) science and technology of tritium, 6) biological effects of tritium, 7) diagnostics of high temperature plasma, 8) new lasers, 9) fundamentals of plasma heating, 10) high efficiency energy conversion, 11) theory and computer simulation, 12) superconducting materials, 13) fundamental phenomena of superconductivity, 14) magnet technology, 15) heat transfer and structural engineering, 16) system design, and 17) resources and assessment of fusion energy. 43 summaries concerning reactor materials and plasma-wall interaction, 29 summaries concerning the science, technology and biological effects of tritium, 41 summaries concerning the fundamentals of reactor plasma control, 15 summaries concerning the technology of superconducting magnets, and 14 summaries concerning the design of fusion reactors and its evaluation are collected in this report, and their results and progress can be known. (Kako, I.)

  10. Improving Industry-Relevant Nuclear-Knowledge Development through Special Partnerships

    Cilliers, A.

    2016-01-01

    Full text: South African Network for Nuclear Education Science and Technology (SAN NEST) has the objective to develop the nuclear education system in South Africa to a point where suitably qualified and experienced nuclear personnel employed by nuclear science and technology programmes in South Africa are predominantly produced by the South African education system. This is done to strengthen the nuclear science and technology education programmes to better meet future demands in terms of quality, capacity and relevance. To ensure sustainable relevance, it is important to develop special partnerships with industry. This paper describes unique partnerships that were developed with nuclear industry partners. The success of these partnerships has ensured more industry partners to embrace the model which has proven to develop relevant knowledge, support research and provide innovative solutions for industry. (author

  11. Shielding structure analysis for LSDS facility

    Choi, Hong Yeop; Kim, Jeong Dong; Lee, Yong Deok; Kim, Ho Dong [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    The nuclear material (Pyro, Spent nuclear fuel) itself and the target material to generate neutrons is the LSDS system for isotopic fissile assay release of high intensity neutron and gamma rays. This research was performed to shield from various strong radiation. A shielding evaluation was carried out with a facilities model of LSDS system. The MCNPX 2.5 code was used and a shielding evaluation was performed for the shielding structure and location. The radiation dose based on the hole structure and location of the wall was evaluated. The shielding evaluation was performed to satisfy the safety standard for a normal person (1 μSv/h) and to use enough interior space. The MCNPX2.5 code was used and a dose evaluation was performed for the location of the shielding material, shielding structure, and hole structure. The evaluation result differs according to the shielding material location. The dose rate was small when the shielding material was positioned at the center. The dose evaluation result regarding the location of the shielding material was applied to the facility and the shielding thickness was determined (In 50 cm + Borax 5 cm + Out 45cm). In the existing hole structure, the radiation leak is higher than the standard. A hole structure model to prevent leakage of radiation was proposed. The general public dose limit was satisfied when using the concrete reinforcement and a zigzag structure. The shielding result will be of help to the facility shielding optimization.

  12. Shielding structure analysis for LSDS facility

    Choi, Hong Yeop; Kim, Jeong Dong; Lee, Yong Deok; Kim, Ho Dong

    2014-01-01

    The nuclear material (Pyro, Spent nuclear fuel) itself and the target material to generate neutrons is the LSDS system for isotopic fissile assay release of high intensity neutron and gamma rays. This research was performed to shield from various strong radiation. A shielding evaluation was carried out with a facilities model of LSDS system. The MCNPX 2.5 code was used and a shielding evaluation was performed for the shielding structure and location. The radiation dose based on the hole structure and location of the wall was evaluated. The shielding evaluation was performed to satisfy the safety standard for a normal person (1 μSv/h) and to use enough interior space. The MCNPX2.5 code was used and a dose evaluation was performed for the location of the shielding material, shielding structure, and hole structure. The evaluation result differs according to the shielding material location. The dose rate was small when the shielding material was positioned at the center. The dose evaluation result regarding the location of the shielding material was applied to the facility and the shielding thickness was determined (In 50 cm + Borax 5 cm + Out 45cm). In the existing hole structure, the radiation leak is higher than the standard. A hole structure model to prevent leakage of radiation was proposed. The general public dose limit was satisfied when using the concrete reinforcement and a zigzag structure. The shielding result will be of help to the facility shielding optimization

  13. The SNM Scanner: A Non-invasive Protocol for Effective Monitoring of Special Nuclear Material Inventories

    Lanier, R G; Dauffy, L S; Hodge, A M

    2004-01-01

    We suggest a system of monitoring special nuclear material inventories which uses simple mathematical techniques to compare the gross features of emitted gamma-ray spectra. In this report we develop the techniques necessary to make such spectral comparisons and describe their application. We also apply these ideas and develop inventory confirmation results using a room-temperature CdTe detector in a real nuclear-material inventory environment

  14. Intelligent data-acquisition instrumentation for special nuclear material assay data analysis

    Ethridge, C.D.

    1980-01-01

    The Detection, Surveillance, Verification, and Recovery Group of the Los Alamos Scientific Laboratory Energy Division/Nuclear Safeguards Programs is now utilizing intelligent data-acquisition instrumentation for assay data analysis of special nuclear material. The data acquisition and analysis are enabled by the incorporation of a number-crunching microprocessor sequenced by a single component microcomputer. Microcomputer firmware establishes the capability for processing the computation of several selected functions and also the ability of instrumentation self-diagnostics

  15. Dosimetry and shielding

    Farinelli, U.

    1977-01-01

    Today, reactor dosimetry and shielding have wide areas of overlap as concerns both problems and methods. Increased interchange of results and know-how would benefit both. The areas of common interest include calculational methods, sensitivity studies, theoretical and experimental benchmarks, cross sections and other nuclear data, multigroup libraries and procedures for their adjustment, experimental techniques and damage functions. This paper reviews the state-of-the-art and the latest development in each of these areas as far as shielding is concerned, and suggests a number of interactions that could be profitable for reactor dosimetry. Among them, re-evaluation of the potentialities of calculational methods (in view of the recent developments) in predicting radiation environments of interest; the application of sensitivity analysis to dosimetry problems; a common effort in the field of theoretical benchmarks; the use of the shielding one-material propagation experiments as reference spectra for detector cross sections; common standardization of the detector nuclear data used in both fields; the setting up of a common (or compatible) multigroup structure and library applicable to shielding, dosimetry and core physics; the exchange of information and experience in the fields of cross section errors, correlations and adjustment; and the intercomparison of experimental techniques

  16. Decree 2805 by means of which the National Accounting and Control of Basic Nuclear Materials and Special Fusionable Materials System, is established

    1979-01-01

    This Decree has for object to establish a National Accounting and Control of Basic Nuclear Materials and Special Fusionable Materials System, under the supervision of the National Council for the Nuclear Industry Development. Its aims are to account nuclear materials, to control nuclear activities, to preserve and control nuclear information, to keep technical relationship with specialized organizations, and to garant nuclear safeguards [es

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

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

    1976-01-01

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

  18. Special course for global nuclear human resource development in cooperation with Hitachi-GE nuclear energy in Tokyo institute of technology

    Ujita, H.; Futami, T.; Saito, M.; Murata, F.; Shimizu, M.

    2012-01-01

    Many Asian countries are willing to learn Japanese nuclear power plants experiences, and are interested in introducing nuclear power generation to meet their future energy demand. Special course for Global Nuclear Human Resource Development was established in April, 2011 in the Department of Nuclear Engineering at Graduate School of Tokyo Institute of Technology in cooperation with Hitachi-GE Nuclear Energy. Purpose of the special course is to develop global nuclear engineers and researchers not only in the Tokyo Institute of Technology but also in the educational institutes of Southeast Asian countries

  19. Radiation shielding

    Aitken, D.

    1979-01-01

    Shields for equipment in which ionising radiation is associated with high electrical gradients, for example X-ray tubes and particle accelerators, incorporate a radiation-absorbing metal, as such or as a compound, and are electrically non-conducting and can be placed in the high electrical gradient region of the equipment. Substances disclosed include dispersions of lead, tungsten, uranium or oxides of these in acrylics polyesters, PVC, ABS, polyamides, PTFE, epoxy resins, glass or ceramics. The material used may constitute an evacuable enclosure of the equipment or may be an external shield thereof. (U.K.)

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

    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. Special conditions for the application of coating materials in nuclear power plants

    Boetius, I.

    1980-01-01

    Proceeding from the special conditions for the application of coating materials in nuclear power plants the following factors influencing the decontamination of surface coatings are discussed from the point of view of radiation protection: abrasion resistance, waterproofness, mechanical and adhesion strength, and permeability. For practical use it is recommended to test the surface tightness of coatings with radiation-exposed specimens

  2. Semi-annual report on strategic special nuclear material inventory differences

    1983-07-01

    This report covers data for the period from April 1, 1982, through September 30, 1982, and includes accounting corrections for data from earlier periods. These data and explanations, together with the absences of physical indications of any theft attempt, support a finding that during this period no theft or diversion of strategic special nuclear material has occurred

  3. Shields calculations for teletherapy equipment. Regulatory approach of the National Center of Nuclear Safety; Calculos de blindajes para equipos de teleterapia, enfoque regulatorio del Centro Nacional de Seguridad Nuclear

    Fuente P, A. de la; Dumenigo G, C.; Quevedo G, J.R.; Lopez F, Y. [Centro Nacional de Seguridad Nuclear, Calle 28 No. 504 e/5 y 7 Ave. Miramar, Playa, Ciudad de la Habana (Cuba)]. e-mail: andres@orasen.co.cu

    2006-07-01

    The evaluation of applications of construction licenses for the new services of radiotherapy has occupied a significant space in the activity developed by the National Center of Nuclear Safety (CNSN) in the last 2 years. Presently work the experiences of the authors in the evaluation of the required shield for the local where cobalt therapy equipment and lineal accelerators of medical use are used its are exposed, the practical problems detected are approached during the application of the methodologies recommended in both cases and its are discussed which have been the suppositions of items accepted by the Regulatory Authority for the realization of these shield calculations. The accumulated experience allows to assure that the realistic application of the item data and the rational use of the engineering logic makes possible to design local for radiotherapy equipment that fulfill the established dose restrictions in the in use legislation in Cuba, without it implies an excessive expense of construction materials. (Author)

  4. Gonadal shield.

    Purdy, J A; Stiteler, R D; Glasgow, G P; Mill, W B

    1975-10-01

    A secondary gonadal shield for use in the pelvic irradiation of males was designed and built using material and apparatus available with the Cerrobend blocking system. The gonadal dose was reduced to approximately 1.5 to 2.5% of the given dose.

  5. Shielding research in France

    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

  6. Scale-4 and related modular systems for the evaluation of nuclear facilities and package design featuring criticality, shielding and transfer capabilities

    1994-01-01

    Nuclear industry, licensing and regulatory authorities need to be able to rely on good performance of computer codes and nuclear data used in calculations for design and operation of nuclear energy facilities. Given the international impact of a major nuclear accident, and the current crisis in public confidence, it is equally important that the methods, programs and data issued should be internationally accepted. The SCALE modular system has been developed and its capabilities extended during the last 15 years. The driving idea behind its development is that it should contain well established computer codes and data libraries, have an user friendly input format, combine and automate analyses requiring multiple computer codes or calculations into standard analytic sequences and to be well documented and publicly available. The fifth version called SCALE-4 has now been released through the Radiation Shielding Information Center (RSIC) to the OECD/NEA Data Bank. SCALE is now used worldwide. The NEA Data Bank alone has distributed more than one hundred copies of the different versions. The OECD/NEA Data Bank has been asked by its international management committee to hold a seminar with the purpose of exchanging information on the latest developments and experiences among code authors and users, to ensure that users have a correct understanding as to how SCALE should be used to model different problems, and to issue recommendations for further development and benchmarking

  7. Special feature article. Nuclear new age. Towards reform of laws and regulations

    Madarame, Haruki; Morokuzu, Muneo; Shiroyama, Hideaki; Nishiwaki, Yoshihiro; Marumo, Syunji; Suzuki, Takahiro; Hariyama, Hideo

    2007-01-01

    Since about half a century passes after the peaceful use of nuclear energy began in Japan, the safety laws and regulations of the nuclear energy becomes difficult to cope enough with the current situation without regulation structure changing. In March 2007, Tokyo University set up nuclear energy legislation study meeting' consisting of members from regulatory bodies, electric utilities, nuclear industries and others. The special feature introduces five opinions obtained through the argument in the meeting. As an example, Law for the Regulations of Nuclear Sources Material, Nuclear Fuel Material and Reactors is applied to basic design about the commercial reactors, but Electric Utilities Industry Law is applied to a detailed design, and different licensing standard is applied in each. Taking the adjustment of licensing standard with unifying these is necessary for efficiency. In addition, current law for the Regulations of Nuclear Sources Material, Nuclear Fuel Material and Reactors regulates according to the businesses such as reactor facilities, fuel fabrication plants and radioactive waste disposal. As for plural businesses, a low procedure and safety measures are demanded every each business. It is also necessary to include structure of the comprehensive licensing that assumed an enterprise running plural businesses. (T. Tanaka)

  8. Gas shielded metal arc welding with fusible electrode wire. First returns on experience and opportunities in nuclear maintenance and fabrication

    Huguet, Fr.; Joly, P.; Leconte, F.; Baritaux, S.; Prin, C.

    2013-06-01

    In a brief text and a Power Point Presentation, the authors report a return on experience for the implementation of two applications using gas shielded metal arc welding process (GMAW): the on-site welding of the final joint of steam generators, and the coating of a tubing flare. In the first case, the authors analyze not only the compliance with specified technical requirements, but also outline the need to support the process with new verification methods in real time, associated development and validation efforts, and organisational and decisional measures to guarantee a good implementation of the process on site. In the second case, they analyze the process ability to meet technical specifications requiring dilution control, a perfect reproducibility, as well a good control of the welding bath. The authors outline that these two applications which are both using the same term (gas shielded metal arc welding with fusible electrode wire), implement two different transfer regimes and processes. They also discuss operational constraints, and technical opportunities and constraints of fusible electrode wire

  9. Shielded container

    Fries, B.A.

    1978-01-01

    A shielded container for transportation of radioactive materials is disclosed in which leakage from the container is minimized due to constructional features including, inter alia, forming the container of a series of telescoping members having sliding fits between adjacent side walls and having at least two of the members including machine sealed lids and at least two of the elements including hand-tightenable caps

  10. Semi-annual report on strategic special nuclear material inventory differences

    1979-01-01

    This report provides and explains the generally small differences between the amounts of nuclear materials charged to DOE facilities and the amounts that could be physically inventoried. This report covers data for the period from October 1, 1977, through March 31, 1978, and includes accounting corrections for data from earlier periods. The data and explanations, together with the absence of physical indications of any theft attempt, support a finding that during this period no theft or diversion of a significant amount of strategic special nuclear material has occurred

  11. Technology development for radiation shielding analysis

    Ha, Jung Woo; Lee, Jae Kee; Kim, Jong Kyung

    1986-12-01

    Radiation shielding analysis in nuclear engineering fields is an important technology which is needed for the calculation of reactor shielding as well as radiation related safety problems in nuclear facilities. Moreover, the design technology required in high level radioactive waste management and disposal facilities is faced on serious problems with rapidly glowing nuclear industry development, and more advanced technology has to be developed for tomorrow. The main purpose of this study is therefore to build up the self supporting ability of technology development for the radiation shielding analysis in order to achieve successive development of nuclear industry. It is concluded that basic shielding calculations are possible to handle and analyze by using our current technology, but more advanced technology is still needed and has to be learned for the degree of accuracy in two-dimensional shielding calculation. (Author)

  12. Design experience: CRBRP radiation shielding

    Disney, R.K.; Chan, T.C.; Gallo, F.G.; Hedgecock, L.R.; McGinnis, C.A.; Wrights, G.N.

    1978-11-01

    The Clinch River Breeder Reactor Plant (CRBRP) is being designed as a fast breeder demonstration project in the U.S. Liquid Metal Fast Breeder Reactor (LMFBR) program. Radiation shielding design of the facility consists of a comprehensive design approach to assure compliance with design and government regulatory requirements. Studies conducted during the CRBRP design process involved the aspects of radiation shielding dealing with protection of components, systems, and personnel from radiation exposure. Achievement of feasible designs, while considering the mechanical, structural, nuclear, and thermal performance of the component or system, has required judicious trade-offs in radiation shielding performance. Specific design problems which have been addressed are in-vessel radial shielding to protect permanent core support structures, flux monitor system shielding to isolate flux monitoring systems for extraneous background sources, reactor vessel support shielding to allow personnel access to the closure head during full power operation, and primary heat transport system pipe chaseway shielding to limit intermediate heat transport system sodium system coolant activation. The shielding design solutions to these problems defined a need for prototypic or benchmark experiments to provide assurance of the predicted shielding performance of selected design solutions and the verification of design methodology. Design activities of CRBRP plant components an systems, which have the potential for radiation exposure of plant personnel during operation or maintenance, are controlled by a design review process related to radiation shielding. The program implements design objectives, design requirements, and cost/benefit guidelines to assure that radiation exposures will be ''as low as reasonably achievable''

  13. Neutron shielding material

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

    1980-01-01

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

  14. 78 FR 26090 - Content Specifications and Shielding Evaluations for Type B Transportation Packages

    2013-05-03

    ... NUCLEAR REGULATORY COMMISSION [NRC-2012-0270] Content Specifications and Shielding Evaluations for...) 2013-04, ``Content Specifications and Shielding Evaluations for Type B Transportation Packages.'' This... Packages for Radioactive Material,'' for the review of content specifications and shielding evaluations...

  15. Report on {open_quotes}audit of internal controls over special nuclear materials{close_quotes}

    NONE

    1996-04-01

    The Department of Energy (Department) is responsible for safeguarding a significant amount of plutonium, uranium-233 and enriched uranium - collectively referred to as special nuclear materials - stored in the United States. The Department`s office of Nonproliferation and National Security has overall management cognizance for developing policies for safeguarding these materials, while other Headquarters program offices have {open_quotes}landlord{close_quotes} responsibilities for the sites where the materials are stored, and the Department`s operations and field offices provide onsite management of contractor operations. The Department`s management and operating contractors, under the direction of the Department, safeguard and account for the special nuclear material stored at Department sites.

  16. Survey of special nuclear material vehicle monitors for domestic and international safeguards

    Fehlau, P.E.; Atwater, H.F.; Caldwell, J.T.; Shunk, E.R.

    1979-01-01

    Special nuclear materials vehicle monitors, including gateside vehicle monitors, hand-held personnel-vehicle monitors, and a new tunnel monitor concept for very large vehicles, are discussed. The results of a comparison of effectiveness of monitors for domestic application are presented. The results of calculations and small scale prototype measurements are given for a tunnel-like neutron monitor for monitoring at the perimeter of an enrichment plant subjected to International Safeguards

  17. Specialized medical sections for the treatment of radiation injuries from accidents in nuclear power plants

    Deanovic, Z.; Boranic, M.; Vitale, B.

    1980-01-01

    Presented is the organization of the final, highly specialized treatment (diagnostic and therapeutic) of persons that have been severely injured in a radiation or nuclear accident. In this organizational scheme, the leading idea was to group and establish suitable medical sections for the acceptance, diagnosic work-up, and treatment of radiation casualties, around a strong medical center in which the different specialists would be available

  18. Impact of nuclear fuel cycle centers on shipping special nuclear materials and wastes

    Blomeke, J.O.

    1975-01-01

    The impact of integrated nuclear fuel cycle facilities on the transportation sector appears from this admittedly rather narrow study to be of only marginal significance. However, there are other factors which must be taken into account such as nuclear safeguards, economics, and radiological, ecological, institutional, and sociological impacts. Unless more clear-cut advantages can be shown by on-going studies for some of these other considerations, the regimentation and control of industry that would result from the imposition of the integrated fuel cycle facility concept probably could not be justified. (auth)

  19. 2008 Special NSREC Issue of the IEEE Transactions on Nuclear Science Comments by the Editors

    Schwank, Jim; Buchner, Steve; Marshall, Paul; Duzellier, Sophie; Brown, Dennis; Poivey, Christian; Pease, Ron

    2008-12-01

    The December 2008 special issue of the IEEE Transactions on Nuclear Science contains selected papers from the 45th annual IEEE International Nuclear and Space Radiation Effects Conference (NSREC) held in Tucson, Arizona, July 14 - 18, 2008. Over 115 papers presented at the 2008 NSREC were submitted for consideration for this year's special issue. Those papers that appear in this special issue were able to successfully complete the review process before the deadline for the December issue. A few additional papers may appear in subsequent issues of the TRANSACTIONS. This publication is the premier archival journal for research on space and nuclear radiation effects in materials, devices, circuits, and systems. This distinction is the direct result of the conscientious efforts of both the authors, who present and document their work, and the reviewers, who selflessly volunteer their time and talent to help review the manuscripts. Each paper in this journal has been reviewed by experts selected by the editors for their expertise and knowledge of the particular subject areas.

  20. Radiation shielding for fusion reactors

    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)

  1. Normalization of shielding structure quality and the method of its studying

    Bychkov, Ya.A.; Lavdanskij, P.A.

    1987-01-01

    Method for evaluation of nuclear facility radiation shield quality is suggested. Indexes of shielding structure radiation efficiency and face efficiency are used as the shielding structure quality indexes. The first index is connected with radiation dose rate during personnel irradiation behind the shield, and the second one - with the stresses in shielding structure introduction of the indexes presented allows to evaluate objectively the quality of nuclear facility shielding structure quality design construction and operation and to economize labour and material resources

  2. Detecting special nuclear materials in suspect containers using high-energy gamma rays emitted by fission products

    Norman, Eric B [Oakland, CA; Prussin, Stanley G [Kensington, CA

    2009-01-27

    A method and a system for detecting the presence of special nuclear materials in a suspect container. The system and its method include irradiating the suspect container with a beam of neutrons, so as to induce a thermal fission in a portion of the special nuclear materials, detecting the gamma rays that are emitted from the fission products formed by the thermal fission, to produce a detector signal, comparing the detector signal with a threshold value to form a comparison, and detecting the presence of the special nuclear materials using the comparison.

  3. Commissioning of calorimeter in radiochemical laboratory for non-destructive assay of special nuclear materials

    Patra, S.; Mhatre, A.M.; Agarwal, C.; Chaudhury, S.; Pujari, P.K.

    2017-01-01

    Accounting of special nuclear materials (SNM) in every stages of nuclear fuel cycle is a necessity where one needs the quantitative estimation of SNM in variety of samples like sealed containers or finished products without altering its physical and chemical form. Non-destructive assay (NDA) techniques are capable of assaying such samples by the way of measuring passive/active neutrons/gamma rays or by the measurement of decay heat. Radiochemistry Division has been actively involved in the development and deployment of various NDA methodologies for meeting the demand of nuclear material accounting as and when required. Recently a radiometric calorimeter, developed by Reactor Control Division, E and I Group, BARC, has been installed in Lab C-33, Radiochemistry Division

  4. SNF shipping cask shielding analysis

    Johnson, J.O.; Pace, J.V. III.

    1996-01-01

    The Waste Management and Remedial Action Division has planned a modification sequence for storage facility 7827 in the Solid Waste Storage Area (SWSA). The modification cycle is: (1) modify an empty caisson, (2) transfer the spent nuclear fuel (SNF) of an occupied caisson to a hot cell in building 3525 for inspection and possible repackaging, and (3) return the package to the modified caisson in the SWSA. Although the SNF to be moved is in the solid form, it has different levels of activity. Thus, the following 5 shipping casks will be available for the task: the Loop Transport Carrier, the In- Pile Loop LITR HB-2 Carrier, the 6.5-inch HRLEL Carrier, the HFIR Hot Scrap Carrier, and the 10-inch ORR Experiment Removal Shield Cask. This report describes the shielding tasks for the 5 casks: determination of shielding characteristics, any streaming avenues, estimation of thermal limits, and shielding calculational uncertainty for use in the transportation plan

  5. 2012 Special NSREC Issue of the IEEE Transactions on Nuclear Science Comments by the Editors

    Schwank, Jim; Brown, Dennis; Girard, Sylvain; Gouker, Pascale; Gerardin, Simone; Quinn, Heather; Barnaby, Hugh

    2012-12-01

    The December 2012 special issue of the IEEE Transactions on Nuclear Science contains selected papers from the 49th annual IEEE International Nuclear and Space Radiation Effects Conference (NSREC) held July 16-20, 2012, in Miami, Florida USA. 95 papers presented at the 2012 NSREC were submitted for consideration for this year’s special issue. Those papers that appear in this special issue were able to successfully complete the review process before the deadline for the December issue. A few additional papers may appear in subsequent issues of the TRANSACTIONS. This publication is the premier archival journal for research on space and nuclear radiation effects in materials, devices, circuits, and systems. This distinction is the direct result of the conscientious efforts of both the authors, who present and document their work, and the reviewers, who selflessly volunteer their time and talent to help review the manuscripts. Each paper in this journal has been reviewed by experts selected by the editors for their expertise and knowledge of the particular subject areas. The peer review process for a typical technical journal generally takes six months to one year to complete. To publish this special issue of the IEEE Transactions on Nuclear Science (in December), the review process, from initial submission to final form, must be completed in about 10 weeks. Because of the short schedule, both the authors and reviewers are required to respond very quickly. The reviewers listed on the following pages contributed vitally to this quick-turn review process.We would like to express our sincere appreciation to each of them for accepting this difficult, but critical role in the process. To provide consistent reviews of papers throughout the year, the IEEE Transactions on Nuclear Science relies on a year-round editorial board that manages reviews for submissions throughout the year to the TRANSACTIONS in the area of radiation effects. The review process is managed by a Senior

  6. The development of a specialized in-vivo body counter for radiation monitoring in the nuclear industry

    1981-01-01

    The design, construction, calibration and testing of a prototype mobile counting system to measure natural uranium in the lungs of workers in the nuclear fuel industry is described. The measurement method is based on the detection of gamma rays emitted during the decay of the daughters of uranium. The mobile facility consisted of a shadow shield and phoswich detector system mounted in a medium-sized trunk. Care was taken to ensure undistorted visual and verbal contact between the subject and the operator. Tests were made to optimize the location and thickness of the lead shielding. A single axle truck was insulated and a heater was installed for winter operation. There was sufficient room in the box to include a change room and waiting area as well as the shadow shield and electronics rack. An extensive series of test measurements was made at a number of locations. Control group studies of males were carried out and a model was developed which correctly predicted the subject background in the uranium region of the gamma spectrum when no uranium was present. The mobile counter was calibrated using the Rando Phantom. On a field trip to Bancroft In February 1981 the shadow shield configuration was shown to provide adequate background reduction for accurate uranium-in-lung measurements. With repeated measurements, changes in an individual's lung burden as small as 2 mg uranium could be detected. The minimum detectable amount of uranium on the basis of one measurement was 5.4 mg

  7. Special duties of paints in nuclear environment in India with examples of application

    Singha Roy, P.K.; Subbaratnam, T.

    1978-01-01

    Any process involving nuclear interactions gives rise to radioactivity and radiation in various degrees apart from conventional effects like heat, pressure, etc. Vessels and structures containing such processes are subject to various degrees of radioactivity and contamination by liquid, gaseous or particulate radioactive materials coming in contact with the substrate. Thus, in any nuclear environment surfaces require protection against both radioactive contamination and normal corrosion forces. The protective coating should be able to resist the radiation field in which it is placed and also withstand decontamination process for removal of radioactive contamination from the surface as and when required. Radioactive contamination due to any radiochemical operation is objectionable for two reasons: (i) it constitutes a potential radiation hazard to occupants of the facility, and (ii) cross contamination jeoparadizes the validity of the experimental data. The difficulty experienced in decontaminating many materials used in construction of such nuclear or radiochemical facilities has therefore given emphasis to the search for suitable coating matetials. This paper deals with the two special duties expected of paints in such environment, viz., radiation resistance and decontaminability tests in this regard in BARC and examples of actual selections and applications in an Indian nuclear power station. Some areas requiring investigations and serious attention for industrial painting in general and nuclear painting work in particular have been also identified. (author)

  8. Neutronic reactor thermal shield

    Lowe, P.E.

    1976-01-01

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

  9. The UKB prescription and the heavy atom effects on the nuclear magnetic shielding of vicinal heavy atoms.

    Maldonado, Alejandro F; Aucar, Gustavo A

    2009-07-21

    Fully relativistic calculations of NMR magnetic shielding on XYH3 (X = C, Si, Ge and Sn; Y = Br, I), XHn (n = 1-4) molecular systems and noble gases performed with a fully relativistic polarization propagator formalism at the RPA level of approach are presented. The rate of convergence (size of basis set and time involved) for calculations with both kinetic balance prescriptions, RKB and UKB, were investigated. Calculations with UKB makes it feasible to obtain reliable results for two or more heavy-atom-containing molecules. For such XYH3 systems, the influence of heavy vicinal halogen atoms on sigma(X) is such that heavy atom effects on heavy atoms (vicinal plus their own effects or HAVHA + HAHA effects) amount to 30.50% for X = Sn and Y = I; being the HAHA effect of the order of 25%. So the vicinal effect alone is of the order of 5.5%. The vicinal heavy atom effect on light atoms (HALA effect) is of the order of 28% for X = C and Y = I. A similar behaviour, but of opposite sign, is observed for sigma(Y) for which sigmaR-NR (I; X = C) (HAHA effect) is around 27% and sigmaR-NR(I; X = Sn) (HAVHA + HAHA effects) is close to 21%. Its electronic origin is paramagnetic for halogen atoms but both dia- and paramagnetic for central atoms. The effect on two bond distant hydrogen atoms is such that the largest variation of sigma(H) within the same family of XYH3 molecules appears for X = Si and Y = I: around 20%. In this case sigma(H; X = Sn, Y = I) = 33.45 ppm and sigma(H; X = Sn, Y = H) = 27.82 ppm.

  10. Clinical Training of Medical Physicists Specializing in Nuclear Medicine (Spanish Edition); Capacitacion clinica de fisicos medicos especialistas en medicina nuclear

    NONE

    2013-07-15

    The application of radiation in human health, for both diagnosis and treatment of disease, is an important component of the work of the IAEA. The responsibility for the increasingly technical aspects of this work is undertaken by the medical physicist. To ensure good practice in this vital area, structured clinical training programmes are required to complement academic learning. This publication is intended to be a guide to the practical implementation of such a programme for nuclear medicine. There is a general and growing awareness that radiation medicine is increasingly dependent on well trained medical physicists who are based in a clinical setting. However an analysis of the availability of medical physicists indicates a large shortfall of qualified and capable professionals. This is particularly evident in developing countries. While strategies to increase educational opportunities are critical to such countries, the need for guidance on structured clinical training was recognized by the members of the Regional Cooperative Agreement for Research, Development and Training related to Nuclear Science and Technology (RCA) for the Asia-Pacific region. Consequently, a technical cooperation regional project (RAS6038) under the RCA programme was formulated to address this need in this region by developing suitable material and establishing its viability. Development of a clinical training guide for medical physicists specialising in nuclear medicine was started in 2009 with the appointment of a core drafting committee of regional and international experts. The publication drew on the experience of clinical training in Australia, Croatia and Sweden and was moderated by physicists working in the Asian region. The present publication follows the approach of earlier IAEA publications in the Training Course Series, specifically Nos 37 and 47, Clinical Training of Medical Physicists Specializing in Radiation Oncology and Clinical Training of Medical Physicists

  11. 2ΔH(D) and 1ΔN(D) isotope effects on nuclear shielding of ammonium ions in complexes with crown ethers and cryptands

    Hansen, P.E.; Hansen, Aa.E.; Lycka, A.; Buvari-Barcza, A.

    1993-01-01

    One-bond deuterium isotope effects on nitrogen nuclear shielding, 1 ΔN(D) and two-bond isotope effects at 1 H nuclear shielding, 2 ΔH(D), have been investigated in a series of inclusion complexes. The hosts comprise, SC-24, [2.2.2], [2.2.1], [2.2.1] D, [2.2] DD, K5, 18-crown-6 and 18-crown-6 tetracarboxylic acid. The structure of the host [2.2.1] is discussed based on isotope effects and 1 H chemical shifts, and an interesting exchange reaction is observed between CDCl 3 and ammonium ions in the [2.2.1] complex. The counterion dependence is shown to be zero for ammonium ions fully included in cages such as SC-24 and [2.2.2] and for 18-C-6(COOH) 4 . For cryptands and podands such as [2.2.1] and K5 a weak counterion dependence of 2 ΔH(D) opposite to that found for ammonium ions in water is observed, while 1 ΔN(D) and 15 N chemical shifts depend very strongly on the counterions, as also found for ammonium ions in water solution. The anilinium ion shows effects similar to those observed for the ammonium ion. 1 ΔN(D) isotope effects correlate well with δN and 2 ΔH(D) correlate with δNH. The correlation between 1 ΔN(D) and 2 ΔH(D) is different for ammonium ions in water and ions included in crowns, cryptands or podands. This shows that different mechanisms are operating. The 2 ΔH(D) isotope effects and δNH chemical shifts depend on the distance to the nearest acceptor (oxygen or nitrogen), and they also depend on the type of acceptor. 2 ΔH(D) isotope effects are found to be a good gauge of N ... N or N ... O distances of the inclusion complexes. The 2 ΔH(D) isotope effects vary with temperature, but the dramatic changes in the 1 H NMR spectra of the host are not reflected in the isotope effects. The one-bond couplings, 1 J(N,H), correlate with 2 ΔH(D). The variations are much larger for the cryptands than found in water solutions. (au) (50 refs.)

  12. Radiation shielding application of lead glass

    Nathuram, R.

    2017-01-01

    Nuclear medicine and radiotherapy centers equipped with high intensity X-ray or teletherapy sources use lead glasses as viewing windows to protect personal from radiation exposure. Lead is the main component of glass which is responsible for shielding against photons. It is therefore essential to check the shielding efficiency before they are put in use. This can be done by studying photon transmission through the lead glasses. The study of photon transmission in shielding materials has been an important subject in medical physics and is potential useful in the development of radiation shielding materials

  13. Radiation dose reduction by water shield

    Zeb, J.; Arshed, W.; Ahmad, S.S.

    2007-06-01

    This report is an operational manual of shielding software W-Shielder, developed at Health Physics Division (HPD), Pakistan Institute of Nuclear Science and Technology (PINSTECH), Pakistan Atomic Energy Commission. The software estimates shielding thickness for photons having their energy in the range 0.5 to 10 MeV. To compute the shield thickness, self absorption in the source has been neglected and the source has been assumed as a point source. Water is used as a shielding material in this software. The software is helpful in estimating the water thickness for safe handling, storage of gamma emitting radionuclide. (author)

  14. 10 CFR 73.73 - Requirement for advance notice and protection of export shipments of special nuclear material of...

    2010-01-01

    ... shipments of special nuclear material of low strategic significance. 73.73 Section 73.73 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) PHYSICAL PROTECTION OF PLANTS AND MATERIALS Records and Reports § 73.73... Incident Response, using any appropriate method listed in § 73.4; (2) Assure that the notification will be...

  15. Study of special challenges for NDT-methods on nuclear structures

    Maack, Stefan; Wiggenhauser, Herbert [BAM Bundesanstalt fuer Materialforschung und -pruefung, Berlin (Germany); Thunell, Bjoern [Scanscot Technology AB, Lund (Sweden)

    2015-07-01

    The special design of buildings, constructed for nuclear power plants is a particular challenge for the nondestructive testing in the building industry. In particular the major component thicknesses, the degree of reinforcement and surface coating systems make the application of NDT methods difficult. The studies first steps were undertaken to determine to which extent established applications of these techniques are useable in the field of infrastructure buildings. Methods have been evaluated that are already state of the art. So for example the ground penetrating radar was used for locating metallic mounting parts. Furthermore, the low-lying internal structure of the containment was investigated with the ultrasonic method.

  16. U.S. N.R.C. special safeguards study on nuclear material control and accounting

    Smith, G.D.

    1976-01-01

    In Feb. 1975, NRC directed that an effort be made to determine a safeguards program for Pu recycle. This paper summarizes results of individual contractor evaluations of upgrading material control and accounting concepts as applied to strategically important special nuclear material and describes staff interpretations of these results as applied to future high-throughput fuel-cycle facilities. Real-time material control, design for physical inventory, Pu isotopics control and calorimetry, and material control and accounting for highly enriched uranium fuel materials were the concepts studied. 1 table, 15 references

  17. Computer-based accountability system (Phase I) for special nuclear materials at Argonne-West

    Ingermanson, R.S.; Proctor, A.E.

    1982-05-01

    An automated accountability system for special nuclear materials (SNM) is under development at Argonne National Laboratory-West. Phase I of the development effort has established the following basic features of the system: a unique file organization allows rapid updating or retrieval of the status of various SNM, based on batch numbers, storage location, serial number, or other attributes. Access to the program is controlled by an interactive user interface that can be easily understood by operators who have had no prior background in electronic data processing. Extensive use of structured programming techniques make the software package easy to understand and to modify for specific applications. All routines are written in FORTRAN

  18. A magnet without a magnetic circuit, of high homogeneity, specially for nuclear magnetic resonance images

    Barjhoux, Yves.

    1981-01-01

    This invention concerns a high homogeneity, double access magnet without a magnetic circuit. It is specially adapted for nuclear magnetic resonance (N.M.R.) imagery. Another advantage worth stressing resides in the possibilities of NMR in biochemical analysis which will enable, for instance, cancerous tumours to be detected in vivo. In order to increase the NMR signal ratio over background noise, it is necessary to increase the homogeneity of the B 0 orientating magnetic field. This magnetic field must orientate the nuclear magnetic moments of the elementary particles which compose the body being examined and in particular the protons. It must therefore be relatively constant in intensity and direction in the entire domain of the examination [fr

  19. Glass fiber sensors for detecting special nuclear materials at portal and monitor stations

    Hull, C.D.; Seymour, R.; Crawford, T.; Bliss, M.; Craig, R.A.

    2001-01-01

    Nuclear Safeguards and Security Systems LLC (NucSafe) participated in the Illicit Trafficking Radiation Assessment Program (ITRAP) recently conducted by the Austrian Research Center, Seibersdorf (ARCS) for IAEA, INTERPOL, and the World Customs Organization (IAEA, in press). This presentation reviews ITRAP test results of NucSafe instrumentation. NucSafe produces stationary, mobile, and hand-held systems that use neutron and gamma ray sensors to detect Special Nuclear Materials (SNM). Neutron sensors are comprised of scintillating glass fibers (trade name 'PUMA' for Pu Materials Analysis), which provide several advantages over 3 He and 10 BF 3 tubes. PUMA 6 Li glass fiber sensors offer greater neutron sensitivity and dynamic counting range with significantly less microphonic susceptibility than tubes, while eliminating transport and operational hazards. PUMA sensors also cost less per active area than gas tubes, which is important since rapid neutron detection at passenger, freight, and vehicle portals require large sensor areas to provide the required sensitivity

  20. Semi-annual report on strategic special nuclear material inventory differences

    1984-07-01

    This fourteenth periodic semiannual report of Inventory Differences (ID) covers the last six months of fiscal year 1983 (April 1, 1983 through September 30, 1983), for the Department of Energy (DOE) and DOE contractor facilities possessing significant quantities of strategic special nuclear material (SSNM). Inventory Differences are expected in nuclear material processing and are not, in and of themselves, evidence of lost or stolen material. On the other hand, ID analysis provides valuable information on the effectiveness of the safeguards system's physical protection and material control measures as well as a check on the process controls and material management procedures. ID's outside safeguards control limits or involving a missing SSNM discrete item are investigated. If necessary, an operation may be shut down until an ID is resolved

  1. Shielding design for better plant availability

    Biro, G.G.

    1975-01-01

    Design methods are described for providing a shield system for nuclear power plants that will facilitate maintenance and inspection, increase overall plant availability, and ensure that man-rem exposures are as low as practicable

  2. Problems of the power plant shield optimization

    Abagyan, A.A.; Dubinin, A.A.; Zhuravlev, V.I.; Kurachenko, Yu.A.; Petrov, Eh.E.

    1981-01-01

    General approaches to the solution of problems on the nuclear power plant radiation shield optimization are considered. The requirements to the shield parameters are formulated in a form of restrictions on a number of functionals, determined by the solution of γ quantum and neutron transport equations or dimensional and weight characteristics of shield components. Functional determined by weight-dimensional parameters (shield cost, mass and thickness) and functionals, determined by radiation fields (equivalent dose rate, produced by neutrons and γ quanta, activation functional, radiation functional, heat flux, integral heat flux in a particular part of the shield volume, total energy flux through a particular shield surface are considered. The following methods of numerical solution of simplified optimization problems are discussed: semiempirical methods using radiation transport physical leaks, numerical solution of approximate transport equations, numerical solution of transport equations for the simplest configurations making possible to decrease essentially a number of variables in the problem. The conclusion is drawn that the attained level of investigations on the problem of nuclear power plant shield optimization gives the possibility to pass on at present to the solution of problems with a more detailed account of the real shield operating conditions (shield temperature field account, its strength and other characteristics) [ru

  3. Hot Cell Window Shielding Analysis Using MCNP

    Pope, Chad L.; Scates, Wade W.; Taylor, J. Todd

    2009-01-01

    The Idaho National Laboratory Materials and Fuels Complex nuclear facilities are undergoing a documented safety analysis upgrade. In conjunction with the upgrade effort, shielding analysis of the Fuel Conditioning Facility (FCF) hot cell windows has been conducted. This paper describes the shielding analysis methodology. Each 4-ft thick window uses nine glass slabs, an oil film between the slabs, numerous steel plates, and packed lead wool. Operations in the hot cell center on used nuclear fuel (UNF) processing. Prior to the shielding analysis, shield testing with a gamma ray source was conducted, and the windows were found to be very effective gamma shields. Despite these results, because the glass contained significant amounts of lead and little neutron absorbing material, some doubt lingered regarding the effectiveness of the windows in neutron shielding situations, such as during an accidental criticality. MCNP was selected as an analysis tool because it could model complicated geometry, and it could track gamma and neutron radiation. A bounding criticality source was developed based on the composition of the UNF. Additionally, a bounding gamma source was developed based on the fission product content of the UNF. Modeling the windows required field inspections and detailed examination of drawings and material specifications. Consistent with the shield testing results, MCNP results demonstrated that the shielding was very effective with respect to gamma radiation, and in addition, the analysis demonstrated that the shielding was also very effective during an accidental criticality.

  4. Special committee review of the Nuclear Regulatory Commission's severe accident risks report (NUREG--1150)

    Kouts, H.J.C.; Apostolakis, G.; Kastenberg, W.E.; Birkhofer, E.H.A.; Hoegberg, L.G.; LeSage, L.G.; Rasmussen, N.C.; Teague, H.J.; Taylor, J.J.

    1990-08-01

    In April 1989, the Nuclear Regulatory Commission's (NRC) Office of Nuclear Regulatory Research (RES) published a draft report ''Severe Accident Risks: An Assessment for Five US Nuclear Power Plants,'' NUREG-1150. This report updated, extended and improved upon the information presented in the 1974 ''Reactor Safety Study,'' WASH-1400. Because the information in NUREG-1150 will play a significant role in implementing the NRC's Severe Accident Policy, its quality and credibility are of critical importance. Accordingly, the Commission requested that the RES conduct a peer review of NUREG-1150 to ensure that the methods, safety insights and conclusions presented are appropriate and adequately reflect the current state of knowledge with respect to reactor safety. To this end, RES formed a special committee in June of 1989 under the provisions of the Federal Advisory Committee Act. The Committee, composed of a group of recognized national and international experts in nuclear reactor safety, was charged with preparing a report reflecting their review of NUREG-1150 with respect to the adequacy of the methods, data, analysis and conclusions it set forth. The report which precedes reflects the results of this peer review

  5. Shielding walls against ionizing radiation

    1993-05-01

    Hot-cell shielding walls consist of building blocks made of lead according to DIN 25407 part 1, and of special elements according to DIN 25407 part 2. Alpha-gamma cells can be built using elements for protective contamination boxes according to DIN 25480 part 1. This standards document intends to provide planning engineers, manufacturers, future users and the competent authorities and experts with a basis for the design of hot cells with lead shielding walls and the design of hot-cell equipment. (orig./HP) [de

  6. Shielded chamber control stations: an answer to radiological protection requirements

    Delaboudiniere, Catherine.

    1979-06-01

    This study, carried out at the radiometallurgy laboratory of the Fontenay-aux-Roses Nuclear Research Centre, is based on an ergonomic approach. It shows that the shielded chamber control station, designed for human protection against ionizing radiations, demands an effort of adaptation on the part of the operator. This station does in fact meet radiological protection obligations but would benefit from certain improvements aimed at reducing the secondary fatigue of manipulators already under the strain of adapting to the machine. Fields in which improvements and/or changes are desirable have been investigated with special reference to telemanipulators. New and better designed instruments will soon appear on the telemanipulator market [fr

  7. The ''nuclear car wash'': a scanner to detect illicit special nuclear material in cargo containers

    Slaughter, D. R.; Accatino, M. R.; Bernstein, A.; Dougan, A. D.; Hall, J. M.; Loshak, A.; Manatt, D. R.; Pohl, B. A.; Prussin, S. G.; Walling, R. S.; Weirup, D. L.

    2004-01-01

    There is an urgent need to improve the reliability of screening cargo containers for illicit nuclear material that may be hidden there for terrorist purposes. A screening system is described for detection of fissionable material hidden in maritime cargo containers. The system makes use of a low intensity neutron beam for producing fission; and the detection of the abundant high-energy γ rays emitted in the β-decay of short-lived fission products and β-delayed neutrons. The abundance of the delayed γ rays is almost an order of magnitude larger than that of the delayed neutrons normally used to detect fission and they are emitted on about the same time scale as the delayed neutrons, i.e., ∼1 min. The energy and temporal distributions of the delayed γ rays provide a unique signature of fission. Because of their high energy, these delayed γ rays penetrate loW--Z cargoes much more readily than the delayed neutrons. Coupled with their higher abundance, the signal from the delayed γ rays escaping from the container is predicted to be as much as six decades more intense than the delayed neutron signal, depending upon the type and thickness of the intervening cargo. The γ rays are detected in a large array of scintillators located along the sides of the container as it is moved through them. Measurements have confirmed the signal strength in somewhat idealized experiments and have also identified one interference when 14.5 MeV neutrons from the D, T reaction are used for the interrogation. The interference can be removed easily by the appropriate choice of the neutron source

  8. Important aspects of radiation shielding for fusion reactor tokamaks

    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

  9. Analysis of Shield Construction in Spherical Weathered Granite Development Area

    Cao, Quan; Li, Peigang; Gong, Shuhua

    2018-01-01

    The distribution of spherical weathered bodies (commonly known as "boulder") in the granite development area directly affects the shield construction of urban rail transit engineering. This paper is based on the case of shield construction of granite globular development area in Southern China area, the parameter control in shield machine selection and shield advancing during the shield tunneling in this special geological environment is analyzed. And it is suggested that shield machine should be selected for shield construction of granite spherical weathered zone. Driving speed, cutter torque, shield machine thrust, the amount of penetration and the speed of the cutter head of shield machine should be controlled when driving the boulder formation, in order to achieve smooth excavation and reduce the disturbance to the formation.

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

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

    1983-01-01

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

  11. Safety guide data on radiation shielding in a reprocessing facility

    Sekiguchi, Noboru; Naito, Yoshitaka

    1986-04-01

    In a reprocessing facility, various radiation sources are handled and have many geometrical conditions. To aim drawing up a safety guidebook on radiation shielding in order to evaluate shielding safety in a reprocessing facility with high reliability and reasonableness, JAERI trusted investigation on safety evaluation techniques of radiation shielding in a reprocessing facility to Nuclear Safety Research Association. This report is the collection of investigation results, and describes concept of shielding safety design principle, radiation sources in reprocessing facility and estimation of its strength, techniques of shielding calculations, and definite examples of shielding calculation in reprocessing facility. (author)

  12. A data communications systems for tamper-protected special nuclear materials (SNM) inventory management

    Hurkamp, A.C.

    1995-01-01

    The Department of Energy (D.O.E.) is responsible for the long term storage and protection of large quantities of Special Nuclear Material (SNM). This material is stored within individual containers located in vaults. Security measures are required to ensure that the SNM remains within the canisters where it is stored and that it is not disturbed in any manner. Conventional security and inventory techniques are manpower intensive and often require exposure to radiation hazards. The Purpose of this D.O.E sponsored project is to develop a cost effective system to monitor Special Nuclear Materials that, when fielded, would result in an extension of manual inventory cycles at a wide variety of SNM storage locations. The system consists of a computer, radio frequency interrogator, and individual miniaturized radio frequency transponders (tags) that are co-located with individual SNM containers. Each tag can perform SNM inventory, tamper alarm, and multiple sensor data transmission to the interrogator under the control of software designed by the user. SNM custodians can customize the system by their choice of packaging, software, and sensors. When implemented in accordance with current department of energy (D.O.E.) Policy on SNM inventory extension, the system has the potential to qualify for maximum extension times thus saving considerable resources through reduction of radiation exposure

  13. A novel method to assay special nuclear materials by measuring prompt neutrons from polarized photofission

    Mueller, J.M., E-mail: mueller@tunl.duke.edu [Triangle Universities Nuclear Laboratory, Durham, NC 27710 (United States); Department of Physics, Duke University, Durham, NC 27708 (United States); Ahmed, M.W. [Triangle Universities Nuclear Laboratory, Durham, NC 27710 (United States); Department of Physics, Duke University, Durham, NC 27708 (United States); Department of Mathematics and Physics, North Carolina Central University, Durham, NC 27707 (United States); Weller, H.R. [Triangle Universities Nuclear Laboratory, Durham, NC 27710 (United States); Department of Physics, Duke University, Durham, NC 27708 (United States)

    2014-08-01

    A novel method of measuring the enrichment of special nuclear material is presented. Recent photofission measurements using a linearly polarized γ-ray beam were performed on samples of {sup 232}Th, {sup 233,235,238}U, {sup 237}Np, and {sup 239,240}Pu. Prompt neutron polarization asymmetries, defined to be the difference in the prompt neutron yields parallel and perpendicular to the plane of beam polarization divided by their sum, were measured. It was discovered that the prompt neutron polarization asymmetries differed significantly depending on the sample. Prompt neutrons from photofission of even–even (non-fissile) targets had significant polarization asymmetries (∼0.2 to 0.5), while those from odd-A (generally fissile) targets had polarization asymmetries close to zero. This difference in the polarization asymmetries could be exploited to measure the fissile versus non-fissile content of special nuclear materials, and potentially to detect the presence of fissile material during active interrogation. The proposed technique, its expected performance, and its potential applicability are discussed.

  14. Physical protection of special nuclear materials in the Federal Republic of Germany

    Rossnagel, A.

    1987-01-01

    Measures to protect special nuclear materials in West Germany are based on a relatively likely average threat and not on the less-likely but maximum credible threat. The reason is to avoid the costs of maintaining a very high level of security that is seldom needed. The concept of delayed action, which divides the responsibility for security between private protection forces and local police forces, would be insufficient in the event of the maximum credible threat. In principle, sufficient security is possible only when the facilities and transports are constantly protected by a police force large enough to deal with the maximum credible threat without assistance from outside. They must also be adequately armed to avert all possible attackers, and they must be paid by the licensees. The threat of an insider(s) diverting special nuclear material can be addressed only by strengthening the scope and depth of the checks on applicants and employees and the permanent work controls. It is recognized, however, that such security measures may not be compatible with constitutional civil liberties

  15. A novel method to assay special nuclear materials by measuring prompt neutrons from polarized photofission

    Mueller, J.M.; Ahmed, M.W.; Weller, H.R.

    2014-01-01

    A novel method of measuring the enrichment of special nuclear material is presented. Recent photofission measurements using a linearly polarized γ-ray beam were performed on samples of 232 Th, 233,235,238 U, 237 Np, and 239,240 Pu. Prompt neutron polarization asymmetries, defined to be the difference in the prompt neutron yields parallel and perpendicular to the plane of beam polarization divided by their sum, were measured. It was discovered that the prompt neutron polarization asymmetries differed significantly depending on the sample. Prompt neutrons from photofission of even–even (non-fissile) targets had significant polarization asymmetries (∼0.2 to 0.5), while those from odd-A (generally fissile) targets had polarization asymmetries close to zero. This difference in the polarization asymmetries could be exploited to measure the fissile versus non-fissile content of special nuclear materials, and potentially to detect the presence of fissile material during active interrogation. The proposed technique, its expected performance, and its potential applicability are discussed

  16. Future global manpower shortages in nuclear industries with special reference to india including remedial measures

    Ghosh Hazra, G.S.

    2008-01-01

    -2050. Service sector in India accounts for about 50% of GDP which will continue to increase further and will provide more jobs and better paid jobs than core industries and there will be continued shift of choice of employment towards service sector creating deep gap of manpower resource requirement in basic and core industries. There are reports that some countries may have to abandon some future projects because of non availability skilled manpower in core industries. The installed capacity of nuclear power in India in the year 2052 will be about 200 G We from the present about 4 G We which will be a manifold increase. This will need about estimated 1,30,000 skilled manpower from the present about 12,000 persons in nuclear industries. Moreover, the need for competent persons in nuclear industries because of high safety requirements of nuclear installations will further add to the problem. The following short-term strategies to retain and attract new employees in nuclear industries may be envisaged amongst others: - Recruit employees prior to the departure of experienced technical staff to facilitate knowledge transfer in time. - Increase compensation and the number of higher level positions. - Increase permanent entry-level intake of skilled manpower taking into account historical turn-over rate. - Implement attractive student loan repayment programs by tying up with banks and financial institutions. - Implement well researched strategies and measures including reassessing the practical capacity which nations including India can achieve in power generation in future taking practical aspects of manpower shortage. - Implement advanced technology which requires lesser manpower. - Implement higher level of automation in nuclear industries. The paper aims to highlight the acute problems of future manpower shortages in nuclear industries globally with special reference to India and discusses some remedial measures which may be taken to address the issue. (author)

  17. Seismic proof test of shielding block walls

    Ohte, Yukio; Watanabe, Takahide; Watanabe, Hiroyuki; Maruyama, Kazuhide

    1989-01-01

    Most of the shielding block walls used for building nuclear facilities are built by dry process. When a nuclear facility is designed, seismic waves specific at each site are set as input seismic motions and they are adopted in the design. Therefore, it is necessary to assure safety of the shielding block walls for earthquake by performing anti-seismic experiments under the conditions at each site. In order to establish the normal form that can be applied to various seismic conditions in various areas, Shimizu Corp. made an actual-size test samples for the shielding block wall and confirmed the safety for earthquake and validity of normalization. (author)

  18. Contamination shield

    Bayer, W.; Pecornik, D.

    1982-01-01

    An acrylate resin is presented as contamination protection coating for components and instruments in nuclear facilities and for spent fuel transport containers. The resin is evaporated or sublimated at 130 0 C and can thus be removed easily from the protected component. The radioactive particles entrained during evaporation are retained by suitable filters. (TK) [de

  19. Interim report on the special research project 'exposure to environmental radiation due to nuclear facilities'

    1981-03-01

    This special research project was started in 1978 as five-year plan. The purposes are to clarify the aspect of radiation exposure in human bodies due to the radioactive substances brought into the environment regarding the utilization of atomic energy, its mechanism and various factors affecting it, and to contribute to the evaluation of exposure dose, the reduction of radiation exposure, the conditions of locating nuclear facilities and the improvement of the method of disposing radioactive wastes. In addition to the fields treated in the previous special research project, the experimental research concerning the metabolism of environmental radioactive nuclides in bodies, namely the problem of the peculiarity of radioactive nuclide kinetics in infants and fetuses different from adults and the possibility of causing the changes in the intake and metabolism of nuclides in foods by the difference in their states of existence, was newly included. Also the research concerning the method of evaluating the absorbed dose in human organs at the time of irradiation outside and inside bodies in a new subject. Accordingly, this special research project is composed of (1) the research concerning the radionuclide kinetics in the environment, (2) the research concerning the radionuclide kinetics in bodies, (3) the research concerning the measurement and evaluation of dose absorbed in internal organs due to environmental radiation, and (4) the research concerning the monitoring of low level environmental radiation. The results obtained so far are reported. (Kako, I.)

  20. Clinical Training of Medical Physicists Specializing in Nuclear Medicine (Spanish Edition)

    2013-01-01

    The application of radiation in human health, for both diagnosis and treatment of disease, is an important component of the work of the IAEA. The responsibility for the increasingly technical aspects of this work is undertaken by the medical physicist. To ensure good practice in this vital area, structured clinical training programmes are required to complement academic learning. This publication is intended to be a guide to the practical implementation of such a programme for nuclear medicine. There is a general and growing awareness that radiation medicine is increasingly dependent on well trained medical physicists who are based in a clinical setting. However an analysis of the availability of medical physicists indicates a large shortfall of qualified and capable professionals. This is particularly evident in developing countries. While strategies to increase educational opportunities are critical to such countries, the need for guidance on structured clinical training was recognized by the members of the Regional Cooperative Agreement for Research, Development and Training related to Nuclear Science and Technology (RCA) for the Asia-Pacific region. Consequently, a technical cooperation regional project (RAS6038) under the RCA programme was formulated to address this need in this region by developing suitable material and establishing its viability. Development of a clinical training guide for medical physicists specialising in nuclear medicine was started in 2009 with the appointment of a core drafting committee of regional and international experts. The publication drew on the experience of clinical training in Australia, Croatia and Sweden and was moderated by physicists working in the Asian region. The present publication follows the approach of earlier IAEA publications in the Training Course Series, specifically Nos 37 and 47, Clinical Training of Medical Physicists Specializing in Radiation Oncology and Clinical Training of Medical Physicists

  1. Clinical Training of Medical Physicists Specializing in Nuclear Medicine (French Edition)

    2012-01-01

    The application of radiation in human health, for both diagnosis and treatment of disease, is an important component of the work of the IAEA. The responsibility for the increasingly technical aspects of this work is undertaken by the medical physicist. To ensure good practice in this vital area, structured clinical training programmes are required to complement academic learning. This publication is intended to be a guide to the practical implementation of such a programme for nuclear medicine. There is a general and growing awareness that radiation medicine is increasingly dependent on well trained medical physicists who are based in a clinical setting. However an analysis of the availability of medical physicists indicates a large shortfall of qualified and capable professionals. This is particularly evident in developing countries. While strategies to increase educational opportunities are critical to such countries, the need for guidance on structured clinical training was recognized by the members of the Regional Cooperative Agreement for Research, Development and Training related to Nuclear Science and Technology (RCA) for the Asia-Pacific region. Consequently, a technical cooperation regional project (RAS6038) under the RCA programme was formulated to address this need in this region by developing suitable material and establishing its viability. Development of a clinical training guide for medical physicists specialising in nuclear medicine was started in 2009 with the appointment of a core drafting committee of regional and international experts. The publication drew on the experience of clinical training in Australia, Croatia and Sweden and was moderated by physicists working in the Asian region. The present publication follows the approach of earlier IAEA publications in the Training Course Series, specifically Nos 37 and 47, Clinical Training of Medical Physicists Specializing in Radiation Oncology and Clinical Training of Medical Physicists

  2. Proceedings of a meeting on radiation shielding and related topics

    1978-01-01

    This is a proceedings of a meeting on radiation shielding and related topics held on Feb. 22 and 23 in 1978 at Nuclear Engineering Research Laboratory of University of Tokyo. The reports includes the following items (1) studies on neutronics with accelerators (2) radiation damage (3) shielding design (4) radiation streaming (5) shielding experiments from a point of view of radiation measurements (6) shielding benchmark experiments (7) prospects on the study of neutronics. All items are written in Japanese. (auth.)

  3. Special problems of setting up nuclear medicine in a developing country

    Ganatra, R.D.

    1992-01-01

    There are some special problems in setting up nuclear medicine in a developing country. They can be briefly described in the form of the following general rules. 1) Impossible triangle. For the practice of nuclear medicine, three things are needed: Instrument, Radiopharmaceutical and a Patient. In a developing country, these three become three sides of an impossible triangle. When the radiopharmaceutical is available, the instrument may not be working; when the instrument is functioning, the radiopharmaceutical may not have been obtained from the foreign supplier; and when both are there, the patient might no longer be in the hospital. Three sides of this triangle never join to become a congruent whole. 2) Reverse square law. Further away one is from the source of supply of instruments and radiopharmaceuticals, the problems multiply by the square of this distance. 3) Future of nuclear medicine is tied to the electrical supply available in a developing country. These problems related to power supply are described in the Chapter on maintenance of instruments

  4. Sheathed electrical resistance heaters for nuclear or other specialized service - approved 1973

    Anon.

    1976-01-01

    This specification presents the requirements for cylindrical metal-sheathed, electrical resistance heaters with compacted mineral-oxide insulation for nuclear or other specialized service. The intended use of a sheathed heater in a specific nuclear or general application will require an evaluation by the purchaser of the compatibility of the heater assembly in the proposed application including the effects of the integrated proposed application including the effects of the integrated neutron flux, temperature, and atmosphere on the properties of the materials of construction. This specification does not include all possible specifications, standards, etc. for materials that may be used in sheathing, insulation, resistance wire, or conductors wire in nuclear environments. The requirements of this specification include only the austenitic stainless steels and nickel-based alloys for sheathing; magnesium oxide, aluminum oxide, beryllium oxide for insulation; and nickel-chromium or iron-chromium-aluminum heater elements with or without low-resistance connecting wires. The intent of this specification is to present the requirements for heaters capable of operating at sheath temperatures and heat fluxes that will limit the maximum internal heater-element temperature to 1050 0 C

  5. Special problems of setting up nuclear medicine in a developing country

    Ganatra, R D

    1993-12-31

    There are some special problems in setting up nuclear medicine in a developing country. They can be briefly described in the form of the following general rules. 1) Impossible triangle. For the practice of nuclear medicine, three things are needed: Instrument, Radiopharmaceutical and a Patient. In a developing country, these three become three sides of an impossible triangle. When the radiopharmaceutical is available, the instrument may not be working; when the instrument is functioning, the radiopharmaceutical may not have been obtained from the foreign supplier; and when both are there, the patient might no longer be in the hospital. Three sides of this triangle never join to become a congruent whole. 2) Reverse square law. Further away one is from the source of supply of instruments and radiopharmaceuticals, the problems multiply by the square of this distance. 3) Future of nuclear medicine is tied to the electrical supply available in a developing country. These problems related to power supply are described in the Chapter on maintenance of instruments

  6. Strategic special nuclear material Inventory Differences. Semiannual report, April 1-September 30, 1984

    1985-07-01

    This sixteenth periodic semiannual report of Inventory Differences (ID) covers the last six months of fiscal year 1984 (April 1, 1984, through September 30, 1984), for the Department of Energy (DOE) and DOE contractor facilities possessing significant quantities of strategic special nuclear material (SSNM). Inventory Differences are simply the differences between the amount of material shown in the accounting records and the amount of material reported in the physical inventory. These differences are generally due to errors in estimating material in unmeasurable form at the time of an inventory, unmeasurable holdup in equipment, measurement imprecisions, inaccuracies in initial determinations of SSNM produced or used in nuclear reactors, and inventory or bookkeeping errors. Both DOE and contractors operating DOE facilities carefully maintain, analyze, and investigate ID data. Inventory Differences are expected in nuclear material processing and are not, in and of themselves, evidence of lost or stolen material. On the other hand, ID analysis provides valuable information on the effectiveness of the safeguards system's physical protection and material control measures as well as a check on the process controls and material management procedures. ID's outside safeguards control limits or involving a missing SSNM discrete item are investigated. If necessary, an operation may be shut down until an ID is resolved

  7. Shipping container for nuclear fuels

    Housholder, W.R.; Greer, N.L.

    1976-01-01

    A container for nuclear materials is described wherein a specially and uniquely constructed pressure vessel and gamma shield assembly for holding the nuclear materials is provided in a housing, and wherein a positioning means extends between the housing and the assembly for spacing the same, insulation in the housing essentially filling the space between the assembly and housing, the insulation comprising beads, globules or the like of water encapsulated in plastic and which, in one important embodiment, contains neutron absorbing matter

  8. Radiation shielding for fission reactors

    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)

  9. Radiation shielding device

    Nakagawa, Takahiro; Yamagami, Makoto.

    1996-01-01

    A fixed shielding member made of a radiation shielding material is constituted in perpendicular to an opening formed on radiation shielding walls. The fixed shielding member has one side opened and has other side, the upper portion and the lower portion disposed in close contact with the radiation shielding walls. Movable shielding members made of a radiation shielding material are each disposed openably on both side of the fixed shielding member. The movable shielding member has a shaft as a fulcrum on one side thereof for connecting it to the radiation shielding walls. The other side has a handle attached for opening/closing the movable shielding member. Upon access of an operator, when each one of the movable shielding members is opened/closed on every time, leakage of linear or scattered radiation can be prevented. Even when both of the movable shielding members are opened simultaneously, the fixed shielding member and the movable shielding members form labyrinth to prevent leakage of linear radioactivity. (I.N.)

  10. Radiation portal monitor with {sup 10}B+ZnS(Ag) neutron detector performance for the detection of special nuclear materials

    Guzman G, K. A.; Gallego, E.; Lorente, A.; Ibanez F, S. [Universidad Politecnica de Madrid, Departamento de Ingenieria Energetica, ETSI Industriales, C. Jose Gutierrez Abascal 2, 28006 Madrid (Spain); Vega C, H. R. [Universidad Autonoma de Zacatecas, Unidad Academica de Estudios Nucleares, Cipres No. 10, Fracc. La Penuela, 98068 Zacatecas, Zac. (Mexico); Gonzalez, J. A. [Universidad Politecnica de Madrid, Laboratorio de Ingenieria Nuclear, ETSI Caminos, Canales y Puertos, C. Prof. Aranguren 3, 28040 Madrid (Spain); Mendez, R., E-mail: ingkarenguzman@gmail.com [Centro de Investigaciones Energeticas, Medioambientales y Tecnologicas, Laboratorio de Patrones Neutronicos, Av. Complutense 40, 28040 Madrid (Spain)

    2016-10-15

    In homeland security, neutron detection is used to prevent the smuggling of special nuclear materials. Thermal neutrons are normally detected with {sup 3}He proportional counters, in the radiation portal monitors, Rpms, however due to the {sup 3}He shortage new procedures are being studied. In this work Monte Carlo methods, using the MCNP6 code, have been used to study the neutron detection features of a {sup 10}B+ZnS(Ag) under real conditions inside of a Rpm. The performance for neutron detection was carried out for {sup 252}Cf, {sup 238}U and {sup 239}Pu under different conditions. In order to mimic an actual situation occurring at border areas, a sample of SNM sited inside a vehicle was simulated and the Rpm with {sup 10}B+ZnS(Ag) response was calculated. At 200 cm the {sup 10}B+ZnS(Ag) on Rpm response is close to 2.5 cps-ng {sup 252}Cf, when the {sup 252}Cf neutron source is shielded with 0.5 cm-thick lead and 2.5 cm-thick polyethylene fulfilling the ANSI recommendations. Three different geometries of neutron detectors of {sup 10}B+ZnS(Ag) in a neutron detection system in Rpm were modeled. Therefore, the {sup 10}B+ZnS(Ag) detectors are an innovative and viable replacement for the {sup 3}He detectors in the Rpm. (Author)

  11. Radiation portal monitor with "1"0B+ZnS(Ag) neutron detector performance for the detection of special nuclear materials

    Guzman G, K. A.; Gallego, E.; Lorente, A.; Ibanez F, S.; Vega C, H. R.; Gonzalez, J. A.; Mendez, R.

    2016-10-01

    In homeland security, neutron detection is used to prevent the smuggling of special nuclear materials. Thermal neutrons are normally detected with "3He proportional counters, in the radiation portal monitors, Rpms, however due to the "3He shortage new procedures are being studied. In this work Monte Carlo methods, using the MCNP6 code, have been used to study the neutron detection features of a "1"0B+ZnS(Ag) under real conditions inside of a Rpm. The performance for neutron detection was carried out for "2"5"2Cf, "2"3"8U and "2"3"9Pu under different conditions. In order to mimic an actual situation occurring at border areas, a sample of SNM sited inside a vehicle was simulated and the Rpm with "1"0B+ZnS(Ag) response was calculated. At 200 cm the "1"0B+ZnS(Ag) on Rpm response is close to 2.5 cps-ng "2"5"2Cf, when the "2"5"2Cf neutron source is shielded with 0.5 cm-thick lead and 2.5 cm-thick polyethylene fulfilling the ANSI recommendations. Three different geometries of neutron detectors of "1"0B+ZnS(Ag) in a neutron detection system in Rpm were modeled. Therefore, the "1"0B+ZnS(Ag) detectors are an innovative and viable replacement for the "3He detectors in the Rpm. (Author)

  12. An evaluation of some special techniques for nuclear waste disposal in space

    Mackay, J. S.

    1973-01-01

    A preliminary examination is reported of several special ways for space disposal of nuclear waste material which utilize the radioactive heat in the waste to assist in the propulsion for deep space trajectories. These include use of the wastes in a thermoelectric generator (RTG) which operates an electric propulsion device and a radioisotope - thermal thruster which uses hydrogen or ammonia as the propellant. These propulsive devices are compared to the space tug and the space tug/solar electric propulsion combination for disposal of waste on a solar system escape trajectory. Such comparisons indicate that the waste-RTG approach has considerable potential provided the combined specific mass of the waste container - RTG system does not exceed approximately 150 kg/kw sub e. Several exploratory numerical calculations have been made for high earth orbit and Earth escape destinations.

  13. Checking the special professional qualification of selected personnel of Czechoslovak nuclear power facilities

    Kovar, P.; Bahnova, V.

    1990-01-01

    The system of examinations of selected staff members of Czechoslovak nuclear power plants for their special professional quanlification is described in detail. This selected personnel includes secondary circuit operators, primary circuit operators, graduate shift leaders as well as reactor unit managers. Attention is paid to the structure, methodology, contents and criteria of evaluation of the written, oral and practical parts of the examination, which is sat for before the State Examination Commission. Based on the results of the examinations, the Czechoslovak Atomic Energy Commission grants, prolongs or cancels licenses for the particular functions. Over the period from 1985 to March 1989, 394 new licenses were issued, 93 licenses were prolonged and 4 were withdrawn. (Z.M.). 7 figs., 3 tabs., 3 refs

  14. Special issue on start of construction of the Ohma nuclear power plant of J-Power

    Takahashi, Taizo

    2008-01-01

    The Electric Power Development Co., Ltd. (J-Power) started construction of its Ohma nuclear power plant - a 1383 MWe Advanced Boiling Water Reactor (ABWR) - in Aomori prefecture on May 2008. The reactor of the Ohma plant will be the first to load MOX fuels in all of its reactor cores. It will be able to consume a quarter of all the recycled MOX fuels produced at Rokkasho reprocessing plant and hence make a major contribution to Japan's policy of recycling plutonium recovered from spent fuels. Special issue reviewed history and overview of the Ohma plant as well as its significance to enhance power generation portfolio in corresponding with national interests. Local governor's expectations and present state of the Ohma plant were also described. After preparation works, construction began on excavation of foundation for the service building, proceeding as always with safety the foremost priority during construction. (T. Tanaka)

  15. An activated barrier for protection of special nuclear materials in vital areas

    Timm, R.E.; Miranda, J.E.

    1984-01-01

    The Argonne National Laboratory and Sandia National Laboratory have recently installed an activated barrier, the Access Denial System (ADS) for the upgrade of safeguards of special nuclear materials. The technology of this system was developed in the late 70's by Sandia National Laboratory-Albuquerque. The installation was the first for the Department of Energy. Subsequently, two additional installations have been completed. The Access Denial System, combined with physical restraints, provide the system delay. The principal advantages of the activated barrier are: (1) it provides an order of magnitude improvement in delay over that of a fixed barrier, (2) it can be added to existing vital areas with a minimum of renovations, (3) existing operations are minimally impacted, and (4) health and safety risks are virtually nonexistent. Hardening of the vital areas using the ADS was accomplished in a costeffective manner

  16. Project plan remove special nuclear material from PFP project plutonium finishing plant; TOPICAL

    BARTLETT, W.D.

    1999-01-01

    This plan presents the overall objectives, description, justification and planning for the Plutonium Finishing Plant (PFP) Remove Special Nuclear Material (SNM) Materials. The intent of this plan is to describe how this project will be managed and integrated with other facility stabilization and deactivation activities. This plan supplements the overall integrated plan presented in the Plutonium Finishing Plant Integrated Project Management Plan (IPMP), HNF-3617,Rev. 0. This project plan is the top-level definitive project management document for PFP Remove SNM Materials project. It specifies the technical, schedule, requirements and the cost baselines to manage the execution of the Remove SNM Materials project. Any deviations to the document must be authorized through the appropriate change control process

  17. Evaluating safeguards effectiveness against protracted theft of special nuclear material by insiders

    Al-Ayat, R.A.; Sicherman, A.

    1991-01-01

    The new draft DOE Material Control and Accountability Order 5633.3 requires that facilities handling special nuclear material (SNM) evaluate their effectiveness against protracted theft of SNM. Protracted theft means repeated thefts of small quantities of material to accumulate a goal quantity. In this paper the authors discuss issues regarding the evaluation of safeguards and describe how we are augmenting the Analytic System and Software for Evaluating Safeguards and Security (ASSESS) to provide the user with a tool for evaluating effectiveness against protracted theft. Currently, the Insider module of ASSESS focuses on evaluating the timely detection of abrupt theft attempts by various types of single nonviolent insiders. In this paper we describe the approach we're implementing to augment ASSESS to handle various cases of protracted theft attempts

  18. Project plan remove special nuclear material from PFP project plutonium finishing plant

    BARTLETT, W.D.

    1999-01-01

    This plan presents the overall objectives, description, justification and planning for the Plutonium Finishing Plant (PFP) Remove Special Nuclear Material (SNM) Materials. The intent of this plan is to describe how this project will be managed and integrated with other facility stabilization and deactivation activities. This plan supplements the overall integrated plan presented in the Plutonium Finishing Plant Integrated Project Management Plan (IPMP), HNF-3617, Rev. 0. This project plan is the top-level definitive project management document for PFP Remove SNM Materials project. It specifies the technical, schedule, requirements and the cost baselines to manage the execution of the Remove SNM Materials project. Any deviations to the document must be authorized through the appropriate change control process

  19. Activated barrier for protection of special nuclear materials in vital areas

    Timm, R.E.; Miranda, J.E.; Reigle, D.L.; Valente, A.D.

    1984-01-01

    The Argonne National Laboratory and Sandia National Laboratory have recently installed an activated barrier, the Access Denial System (ADS) for the upgrade of safeguards of special nuclear materials. The technology of this system was developed in the late 70's by Sandia National Laboratory-Albuquerque. The installation was the first for the Department of Energy. Subsequently, two additional installations have been completed. The Access Denial System, combined with physical restraints, provide the system delay. The principal advantages of the activated barrier are: (1) it provides an order of magnitude improvement in delay over that of a fixed barrier, (2) it can be added to existing vital areas with a minimum of renovations, (3) existing operations are minimally impacted, and (4) health and safety risks are virtually nonexistent. Hardening of the vital areas using the ADS was accomplished in a cost-effective manner. 3 references, 1 figure, 1 table

  20. Nuclear data needs for fusion reactors

    Gohar, Y.

    1986-01-01

    The nuclear design of fusion components (e.g., first wall, blanket, shield, magnet, limiter, divertor, etc.) requires an accurate prediction of the radiation field, the radiation damage parameters, and the activation analysis. The fusion nucleonics for these tasks are reviewed with special attention to point out nuclear data needs and deficiencies which effect the design process. The main areas included in this review are tritium breeding analyses, nuclear heating calculations, radiation damage in reactor components, shield designs, and results of uncertainty analyses as applied to fusion reactor studies. Design choices and reactor parameters that impact the neutronics performance of the blanket are discussed with emphasis on the tritium breeding ratio. Nuclear data required for kerma factors, shielding analysis, and radiation damage are discussed. Improvements in the evaluated data libraries are described to overcome the existing problems. 84 refs., 11 figs., 9 tabs

  1. Preparation of a basic data base for shielding design. 4

    Nakao, Makoto; Takemura, Morio

    1999-03-01

    With use of a standard groupwise shielding design library JSSTDL produced from the latest evaluated nuclear data library JENDL-3.2, experimental analyses for the JASPER experiments were performed. In order to verify the new version of JSSTDL, whose cross sections of thermal energy region was updated, the polyethylene transmission experiments of the Special Materials Experiment was analysed again, and also zirconium transmission experiment of the Experiment was newly analysed. JSSTDL was applied to the analysis of neutron multiplicative region of the in-vessel fuel storage mockup configurations in the IVS Experiment. Also it was applied to the analyses of neutron streaming effect through the mockup of sodium window in B 4 C shield in the Flux Monitor Experiment and also the mockup of narrow gaps in thick concrete shield in the Gap streaming Experiment. The results were compared with those obtained by the same analysis method and input data using the JSDJ2 library that had been applied consistently to the JASPER experiment analyses. Although the analysis with the new version of JSSTDL resulted in a little reduction of overestimation in the polyethylene transmission configuration, the results obtained with JSSTDL are, in general, higher than those with JSDJ2 as had been found in analyses in preceding years for the Radial Shield Attenuation Experiment, the Axial Shield Experiment, the Intermediate Heat Exchanger Experiment and so on. Compilation of the input data necessary for future reanalyses of important configurations in JASPER experiments, that were selected at the first stage of this study, were continued and new data were added into the computer disk holding previously accumulated data. (author)

  2. Guide for the preparation of applications for special nuclear material licenses of less than critical mass quantities - July 1976

    Anon.

    1977-01-01

    This guide describes the type of information needed to evaluate an application for a specific license for receipt, possession, use, and transfer of special nuclear material. It is intended for applicants requesting authorization to possess and use up to 2000 grams of plutonium, total, in the form of sealed plutonium-beryllium neutron sources, and any special nuclear material in quantities and forms not sufficient to form a critical mass. The latter quantities are considered to be 350 grams of contained uranium-235, 200 grams of uranium-233, 200 grams of plutonium (in any form other than plutonium-beryllium neutron sources) or any combination of them

  3. The application of semianalytic method for calculating the thickness of biological shields of nuclear reactors. Part 1. Theoretical basis of a semianalytic method. Attenuation of neutrons' radiation

    Lukaszek, W.; Kucypera, S.

    1982-01-01

    The basis of a semianalytic method for calculating attenuation of rays (neutron, gamma) in material medium is described. The method was applied in determining the neutrons' flux density in one dimensional Cartesian geometry of the reflector and the shield. (author)

  4. Very small HTGR nuclear power plant concepts for special terrestrial applications

    McDonald, C.F.; Goodjohn, A.J.

    1983-01-01

    The role of the very small nuclear power plant, of a few megawatts capacity, is perceived to be for special applications where an energy source as required but the following prevail: 1) no indigenous fossil fuel source, in long transport distances that add substantially to the cost of oil, coal in gas, and 3) secure long-term power production for defense applications with freedom from fuel supply lines. A small High Temperature Gas-Cooled reactor (HTGR) plant could provide the total energy needs for 1) a military installation, 2) an island base of strategic significance, 3) an industrial community or 4) an urban area. The small HTGR is regarded as a fixed-base installation (as opposed to a mobile system). All of the major components would be factory fabricated and transported to the site where emphasis would be placed on minimizing the construction time. The very small HTGR plant, currently in an early stage of design definition, has the potential for meeting the unique needs of the small energy user in both the military and private sectors. The plant may find acceptance for specialized applications in the industrialized nations and to meet the energy needs of developing nations. Emphasis in the design has been placed on safety, simplicity and compactness

  5. Special Purpose Nuclear Reactor (5 MW) for Reliable Power at Remote Sites Assessment Report

    Sterbentz, James William [Idaho National Lab. (INL), Idaho Falls, ID (United States). Nuclear Science and Technology Division; Werner, James Elmer [Idaho National Lab. (INL), Idaho Falls, ID (United States). Nuclear Science and Technology Division; McKellar, Michael George [Idaho National Lab. (INL), Idaho Falls, ID (United States). Nuclear Science and Technology Division; Hummel, Andrew John [Idaho National Lab. (INL), Idaho Falls, ID (United States). Nuclear Science and Technology Division; Kennedy, John Charles [Idaho National Lab. (INL), Idaho Falls, ID (United States). Nuclear Science and Technology Division; Wright, Richard Neil [Idaho National Lab. (INL), Idaho Falls, ID (United States). Nuclear Science and Technology Division; Biersdorf, John Michael [Idaho National Lab. (INL), Idaho Falls, ID (United States). Nuclear Science and Technology Division

    2017-04-01

    The Phenomena Identification and Ranking Table (PIRT) technique was conducted on the Special Purpose Reactor nuclear plant design. The PIRT is a structured process to identify safety-relevant/safety-significant phenomena and assess the importance and knowledge base by ranking the phenomena. The Special Purpose Reactor is currently in the conceptual design stage. The candidate reactor has a solid monolithic stainless steel core with an array of heat pipes and fuel pellets embedded in the monolith. The heat pipes are used to remove heat from the core using simple, reliable, and well-characterized physics (capillarity, boiling, and condensation). In the initial design, one heat exchanger is used for the working fluid that produces energy, and a second heat exchanger is used to remove decay heat in emergency or shutdown conditions. In addition, a power conversion cycle such as an open-air Brayton system is available as an option for power conversion and process heat. This report summarizes and documents the process and scope of the four PIRT reviews, noting the major activities and conclusions. The identified phenomena, analyses, rationales, and associated ratings are presented along with a summary of the findings from the four individual PIRTs, namely (1) Reactor Accident and Normal Operations, (2) Heat Pipes, (3) Materials, and (4) Power Conversion. The PIRT reports for these four major system areas evaluated are attached as appendixes to this report and provide considerably more detail about each assessment as well as a more complete listing of the phenomena that were evaluated.

  6. Non-nuclear radiological emergencies. Special plan for radiological risk of the Valencian Community

    Rodríguez Rodrigo, I.; Piles Alepuz, I.; Peiró Juan, J.; Calvet Rodríguez, D.

    2015-01-01

    After the publication of the Radiological Hazard Basic Directive, Generalitat (the regional government in Valencian Community) initiated the edition of the pertinent Special Plan, with the objective to assemble the response of all the Security and Emergency Agencies, including the Armed Forces, in a radiological emergency affecting the territory of the Valencian Community, under a single hierarchy command. Being approved and homologated the Radiological Hazard Special Plan, Generalitat has undertaken the implementation process planned to finish in June 2015. Following the same process as other Plans, implementation is organized in a first informative stage, followed of a formative and training stage, and finishing with an activation exercise of the Plan. At the end of the process, is expected that every Agency will know their functions, the structure and organization in which the intervention takes place, the resources needed, and adapt their protocols to the Plan requirements. From the beginning, it has been essential working together with the Nuclear Safety Council, as is established in the agreement signed in order to collaborate in Planning, Preparedness and Response in Radiological Emergencies. [es

  7. Environmental-impact appraisal related to special nuclear materials. License No. SNM-696; Docket No. 70-734

    1983-06-01

    This Environmental Impact Appraisal is issued by the US Nuclear Regulatory Commission in response to an application by GA Technologies, Inc., (GA) for renewal of Special Nuclear Material (SNM) License No. SNM-696 covering plant operations at San Diego, California. The proposed action provides for continuing research, development, and production activities involving SNM, uranium enriched in the U-235 and U-233 isotopes, and plutonium

  8. Technical criteria for terminating or reducing domestic safeguards on low-grade special nuclear material

    Crawford, D.W.

    1996-01-01

    A graded table for terminating or reducing domestic safeguards has been developed for use by programs and facilities within the Department of Energy in decisions regarding the need for or levels of protection of low-grade nuclear materials. Contained in this table are technical criteria which can allow for complete removal of safeguards over many special nuclear material forms and concentrations of typical low-grade materials either currently located at generating or processing sites and materials which may arise from processing operations related to stabilization and disposition activities. In addition, these criteria include higher concentration levels which may warrant maintaining some level of (albeit reduced) security on low-grade materials while allowing reductions in materials control and accountability requirements. These reductions can range from complete removal of these materials from materials control and accountability requirements such as measurements, physical inventories and recordkeeping, to deferring these measurements and physical inventories until a time that either the material is removed from the site or resubmitted for processing. It is important to note that other conditions contained in current Departmental safeguards and security policy be met prior to safeguards termination or reduction

  9. Optimal beta-ray shielding thicknesses for different therapeutic radionuclides and shielding materials

    Cho, Yong In; Kim, Ja Mee; Kim, Jung Hoon

    2017-01-01

    To better understand the distribution of deposited energy of beta and gamma rays according to changes in shielding materials and thicknesses when radionuclides are used for therapeutic nuclear medicine, a simulation was conducted. The results showed that due to the physical characteristics of each therapeutic radionuclide, the thicknesses of shielding materials at which beta-ray shielding takes place varied. Additional analysis of the shielding of gamma ray was conducted for radionuclides that emit both beta and gamma rays simultaneously with results showing shielding effects proportional to the atomic number and density of the shielding materials. Also, analysis of bremsstrahlung emission after beta-ray interactions in the simulation revealed that the occurrence of bremsstrahlung was relatively lower than theoretically calculated and varied depending on different radionuclides. (authors)

  10. PC based temporary shielding administrative procedure (TSAP)

    Olsen, D.E.; Pederson, G.E.; Hamby, P.N.

    1995-01-01

    A completely new Administrative Procedure for temporary shielding was developed for use at Commonwealth Edison's six nuclear stations. This procedure promotes the use of shielding, and addresses industry requirements for the use and control of temporary shielding. The importance of an effective procedure has increased since more temporary shielding is being used as ALARA goals become more ambitious. To help implement the administrative procedure, a personal computer software program was written to incorporate the procedural requirements. This software incorporates the useability of a Windows graphical user interface with extensive help and database features. This combination of a comprehensive administrative procedure and user friendly software promotes the effective use and management of temporary shielding while ensuring that industry requirements are met

  11. A perturbation technique for shield weight minimization

    Watkins, E.F.; Greenspan, E.

    1993-01-01

    The radiation shield optimization code SWAN (Ref. 1) was originally developed for minimizing the thickness of a shield that will meet a given dose (or another) constraint or for extremizing a performance parameter of interest (e.g., maximizing energy multiplication or minimizing dose) while maintaining the shield volume constraint. The SWAN optimization process proved to be highly effective (e.g., see Refs. 2, 3, and 4). The purpose of this work is to investigate the applicability of the SWAN methodology to problems in which the weight rather than the volume is the relevant shield characteristic. Such problems are encountered in shield design for space nuclear power systems. The investigation is carried out using SWAN with the coupled neutron-photon cross-section library FLUNG (Ref. 5)

  12. PC based temporary shielding administrative procedure (TSAP)

    Olsen, D.E.; Pederson, G.E. [Sargent & Lundy, Chicago, IL (United States); Hamby, P.N. [Commonwealth Edison Co., Downers Grove, IL (United States)

    1995-03-01

    A completely new Administrative Procedure for temporary shielding was developed for use at Commonwealth Edison`s six nuclear stations. This procedure promotes the use of shielding, and addresses industry requirements for the use and control of temporary shielding. The importance of an effective procedure has increased since more temporary shielding is being used as ALARA goals become more ambitious. To help implement the administrative procedure, a personal computer software program was written to incorporate the procedural requirements. This software incorporates the useability of a Windows graphical user interface with extensive help and database features. This combination of a comprehensive administrative procedure and user friendly software promotes the effective use and management of temporary shielding while ensuring that industry requirements are met.

  13. Special nuclear materials cutoff exercise: Issues and lessons learned. Volume 1: Summary of exercise

    Libby, R.A.; Davis, C.; Segal, J.E.; Stanbro, W.D.

    1995-08-01

    In a September 1993 address to the United Nations General Assembly, President Clinton announced a new nonproliferation and export control policy that established a framework for US efforts to prevent the proliferation of weapons of mass destruction. The new policy proposed that the US undertake a comprehensive approach to the growing accumulation of fissile material. One of the key elements was for the US to support a special nuclear materials (SNM) multilateral convention prohibiting the production of highly enriched uranium (HEU) or plutonium for nuclear explosives purposes or outside of international safeguards. This policy is often referred to as the President's Cutoff Initiative or the Fissile Material Cutoff Treaty (FMCT). Because both the US Department of Energy (DOE) and foreign reprocessing facilities similar to PUREX will likely to be inspected under a FMCT, the DOE Office of Arms Control and Nonproliferation, Negotiations and Analysis Division (DOE/NN-41) tasked Pacific Northwest Laboratory (PNL) to perform an information gathering exercise, the PUREX Exercise, using the Plutonium-Uranium Extraction (PUREX) Plant located on the Hanford Site in Washington State. PUREX is a former production reactor fuel reprocessing plant currently undergoing a transition to a ''decontamination and decommissioning (D ampersand D) ready'' mode. The PUREX Exercise was conducted March 29--30, 1994, to examine aspects of the imposition of several possible cutoff regimes and to study verification of non-production of SNM for nuclear weapons purposes or outside of safeguards. A follow-up activity to further examine various additional verification regimes was held at Los Alamos National Laboratory (LANL) on May 10, 1994

  14. Special national report of the Slovak Republic compiled under the convention on nuclear safety. April 2012

    2012-04-01

    A Special safety report of the Slovak Republic in 2012 is presented. An account of activities carried out by the Nuclear Regulatory Authority of the Slovak Republic (UJD) is presented. These activities are reported under the headings: (0) Introduction; (0.1) Purpose of the report; (0.2) Brief description of the site characteristics and units; (1) Executive summary; (2) External events; (2.1) Seismic; (2.2) Flooding; (2.3) Extreme weather conditions; (3) Design issues; (3.1) Loss of electrical power; (3.2) Loss of the decay heat removal capability/ultimate heat sink; (3.3) Loss of the primary ultimate heat sink, combined with station black out (see stress tests specifications); (4) Severe accident management; (4.1) Organization and arrangements of the licensee to manage accidents; (4.2) Accident management measures in place at the various stages of a scenario of loss of the core cooling function; (4.3) Maintaining the containment integrity after occurrence of significant fuel damage (up to core meltdown) in the reactor core; (4.4) Accident management measures to restrict the radioactive releases; (5) National organizations (regulator, technical support organizations, operator, government); (5.1) Legislative and regulatory framework; (6) Emergency preparedness and response and post--accident management (off-site); (6.1) Implementation of legislation in the field of emergency preparedness; (7) International cooperation; (7.1) Conventions and communications; (7.2) Cooperation with the international organizations; (7.3) Providing feedback including occurrences at nuclear installations of other nuclear power plants abroad.

  15. Shielding calculational system for plutonium

    Zimmerman, M.G.; Thomsen, D.H.

    1975-08-01

    A computer calculational system has been developed and assembled specifically for calculating dose rates in AEC plutonium fabrication facilities. The system consists of two computer codes and all nuclear data necessary for calculation of neutron and gamma dose rates from plutonium. The codes include the multigroup version of the Battelle Monte Carlo code for solution of general neutron and gamma shielding problems and the PUSHLD code for solution of shielding problems where low energy gamma and x-rays are important. The nuclear data consists of built in neutron and gamma yields and spectra for various plutonium compounds, an automatic calculation of age effects and all cross-sections commonly used. Experimental correlations have been performed to verify portions of the calculational system. (23 tables, 7 figs, 16 refs) (U.S.)

  16. Recent trends in radiation shielding: a RSIC perspective

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

    1979-01-01

    The subject of radiation transport and shielding in the nuclear power industry is reviewed, and advances in the state of the art are described. These fall into the areas of computational methods, nuclear cross sections, industry practices, and standards. Computer codes and data available from the Radiation Shielding Information Center (RSIC) representing recent advances are also described

  17. Core-dependent and ligand-dependent relativistic corrections to the nuclear magnetic shieldings in MH4-n Y n (n = 0-4; M = Si, Ge, Sn, and Y = H, F, Cl, Br, I) model compounds.

    Maldonado, Alejandro F; Aucar, Gustavo A; Melo, Juan I

    2014-09-01

    The nuclear magnetic shieldings of Si, Ge, and Sn in MH(4-n) Y(n) (M = Si, Ge, Sn; Y = F, Cl, Br, I and n = 1-4) molecular systems are highly influenced by the substitution of one or more hydrogens by heavy-halogen atoms. We applied the linear response elimination of small components (LRESC) formalism to calculate those shieldings and learn whether including only a few of the leading relativistic correction terms is sufficient to be able to quantitatively reproduce the full relativistic value. It was observed that the nuclear magnetic shieldings change as the number of heavy halogen substituents and their weights vary, and the pattern of σ(M) generally does not exhibit the normal halogen dependence (NHD) behavior that can be seen in similar molecular systems containing carbon atoms. We also analyzed each relativistic correction afforded by the LRESC method and split them in two: core-dependent and ligand-dependent contributions; we then looked for the electronic mechanisms involved in the different relativistic effects and in the total relativistic value. Based on this analysis, we were able to study the electronic mechanism involved in a recently proposed relativistic effect, the "heavy atom effect on vicinal heavy atom" (HAVHA), in more detail. We found that the main electronic mechanism is the spin-orbit or σ p (T(3)) correction, although other corrections such as σ p (S(1)) and σ p (S(3)) are also important. Finally, we analyzed proton magnetic shieldings and found that, for molecules containing Sn as the central atom, σ(H) decreases as the number of heavy halogen substituents (of the same type: either F, Cl, or Br) increases, albeit at different rates for different halogens. σ(H) only increase as the number of halogen substituents increases if the halogen is iodine.

  18. Handbook of radiation shielding data

    Courtney, J.C.

    1976-07-01

    This handbook is a compilation of data on units, conversion factors, geometric considerations, sources of radiation, and the attenuation of photons, neutrons, and charged particles. It also includes related topics in health physics. Data are presented in tabular and graphical form with sufficient narrative for a least first-approximation solutions to a variety of problems in nuclear radiation protection. Members of the radiation shielding community contributed the information in this document from unclassified and uncopyrighted sources, as referenced

  19. 2008 annual nuclear technology conference: opting out of the use of nuclear power. German special approach leads into a dead end of energy policy. Conference report

    Anon.

    2008-01-01

    The President of the Deutsches Atomforum, Dr. Walter Hohlefelder, emphasized in his opening address at the 2008 Annual Nuclear Technology Conference in Hamburg that the German special approach to nuclear power utilization led straight into a dead end of energy policy. ''The outcome is foreseeable: The ambitious German goals of carbon dioxide reduction are missed, the competitiveness of the country is jeopardized, dependency on foreign energy imports rises,'' Dr. Hohlefelder stated. In view of the growing challenges in energy policy Germany had no alternative but to reassess nuclear power. The only outcome of this reappraisal could be extension of the life of nuclear power plants currently in operation. This was necessary also in order to avoid an impending gap in German electricity supply, Dr. Hohlefelder added. He invited all stakeholders to join in an open, unbiased dialog. Dr. Hohlefelder openly criticized the continued ban on research into the development of new reactors. ''A policy of this kind, a policy which bans thinking, is unacceptable in a technology-oriented, industrialized nation such as Germany.'' Nuclear power technology as a high-tech area was a unique achievement which had contributed to the prosperity of the country. The Annual Nuclear Technology Conference, which was held for the 39th time this year, is one of the biggest specialized conferences in the nuclear field with an attendance, this year, of approximately 1300 participants from more than twenty nations. (orig.)

  20. Handout on shielding calculation

    Heilbron Filho, P.F.L.

    1991-01-01

    In order to avoid the difficulties of the radioprotection supervisors in the tasks related to shielding calculations, is presented in this paper the basic concepts of shielding theory. It also includes exercises and examples. (author)

  1. Design of emergency shield

    Soliman, S.E.

    1993-01-01

    Manufacturing of an emergency movable shield in the hot laboratories center is urgently needed for the safety of personnel in case of accidents or spilling of radioactive materials. In this report, a full design for an emergency shield is presented and the corresponding dose rates behind the shield for different activities (from 1 mCi to 5 Ci) was calculated by using micro shield computer code. 4 figs., 1 tab

  2. 10 CFR 74.11 - Reports of loss or theft or attempted theft or unauthorized production of special nuclear material.

    2010-01-01

    ... 10 Energy 2 2010-01-01 2010-01-01 false Reports of loss or theft or attempted theft or... Requirements § 74.11 Reports of loss or theft or attempted theft or unauthorized production of special nuclear... plutonium shall notify the NRC Operations Center within 1 hour of discovery of any loss or theft or other...

  3. Several problems in accelerator shielding study

    Nakamura, Takashi; Hirayama, Hideo; Ban, Shuichi.

    1980-01-01

    Recently, the utilization of accelerators has increased rapidly, and the increase of accelerating energy and beam intensity is also remarkable. The studies on accelerator shielding have become important, because the amount of radiation emitted from accelerators increased, the regulation of the dose of environmental radiation was tightened, and the cost of constructing shielding rose. As the plans of constructing large accelerators have been made successively, the survey on the present state and the problems of the studies on accelerator shielding was carried out. Accelerators are classified into electron accelerators and proton accelerators in view of the studies on shielding. In order to start the studies on accelerator shielding, first, the preparation of the cross section data is indispensable. The cross sections for generating Bremsstrahlung, photonuclear reactions generating neutrons, generation of neutrons by hadrons, nuclear reaction of neutrons and generation of gamma-ray by hadrons are described. The generation of neutrons and gamma-ray as the problems of thick targets is explained. The shielding problems are complex and diversified, but in this paper, the studies on the shielding, by which basic data are obtainable, are taken up, such as beam damping and side wall shielding. As for residual radioactivity, main nuclides and the difference of residual radioactivity according to substances have been studied. (J.P.N.)

  4. Electromagnetically shielded building

    Takahashi, T.; Nakamura, M.; Yabana, Y.; Ishikawa, T.; Nagata, K.

    1992-01-01

    This invention relates to a building having an electromagnetic shield structure well-suited for application to an information network system utilizing electromagnetic waves, and more particularly to an electromagnetically shielded building for enhancing the electromagnetic shielding performance of an external wall. 6 figs

  5. Electromagnetically shielded building

    Takahashi, T; Nakamura, M; Yabana, Y; Ishikawa, T; Nagata, K

    1992-04-21

    This invention relates to a building having an electromagnetic shield structure well-suited for application to an information network system utilizing electromagnetic waves, and more particularly to an electromagnetically shielded building for enhancing the electromagnetic shielding performance of an external wall. 6 figs.

  6. ANS shielding standards for light-water reactors

    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

  7. Preparation of a basic data base for shielding design. 3

    Takemura, Morio

    1998-03-01

    With use of a standard groupwise shielding design library JSSTDL produced from the latest evaluated nuclear data library JENDL-3.2, experimental analyses for the In-Vessel Fuel Storage (IVS) Experiment and the Intermediate Heat Exchanger (IHX) Experiment were performed. The results were compared with those obtained by the same analysis method and input data using the JSDJ2 library that had been applied consistently to the JASPER experiment analyses. In general, the results obtained with JSSTDL are higher than those with JSDJ2 as were found in analyses in last two years for the Radial Shield Attenuation Experiment and the Special Materials Experiment and also the Axial Shield Experiment. The calculation-to-experiment ratios of the fast neutron flux just behind deep penetration in sodium were obtained first by these IVS and IHX experimental analyses with the JSSTDL library. However, it was confirmed not to be easy to evaluate the accuracy of sodium cross section because of its dependency on how to model the swelled sodium slabs and tanks. The analyses with JSSTDL library were verified by comparison with other analyses with another library based on JENDL-3.2. Compilation of the input data necessary for future reanalyses of important configurations in JASPER experiments, that were selected at the first stage of this study, were continued and new data were added into the computer disk holding previously accumulated data. (author)

  8. High ionization radiation field remote visualization device - shielding requirements

    Fernandez, Antonio P. Rodrigues; Omi, Nelson M.; Silveira, Carlos Gaia da; Calvo, Wilson A. Pajero

    2011-01-01

    The high activity sources manipulation hot-cells use special and very thick leaded glass windows. This window provides a single sight of what is being manipulated inside the hot-cell. The use of surveillance cameras would replace the leaded glass window, provide other sights and show more details of the manipulated pieces, using the zoom capacity. Online distant manipulation may be implemented, too. The limitation is their low ionizing radiation resistance. This low resistance also limited the useful time of robots made to explore or even fix problematic nuclear reactor core, industrial gamma irradiators and high radioactive leaks. This work is a part of the development of a high gamma field remote visualization device using commercial surveillance cameras. These cameras are cheap enough to be discarded after the use for some hours of use in an emergency application, some days or some months in routine applications. A radiation shield can be used but it cannot block the camera sight which is the shield weakness. Estimates of the camera and its electronics resistance may be made knowing each component behavior. This knowledge is also used to determine the optical sensor type and the lens material, too. A better approach will be obtained with the commercial cameras working inside a high gamma field, like the one inside of the IPEN Multipurpose Irradiator. The goal of this work is to establish the radiation shielding needed to extend the camera's useful time to hours, days or months, depending on the application needs. (author)

  9. Elrotherm shielding systems. New pioneering material composites; Elrotherm-Abschirmsysteme. Neue Zukunftsweisende Materialkompositionen

    Zika-Beyerlein, B [ElringKlinger (Germany). Geschaeftsbereich Abschirmtechnik

    2004-09-01

    Tightly packed engine compartments put special demands on thermal and acoustic shielding systems. With new material composites allowing for particularly thin-walled and light shielding parts, ElringKlinger is well equipped for the future. (orig.)

  10. Electromagnetic shielding formulae

    Dahlberg, E.

    1979-02-01

    This addendum to an earlier collection of electromagnetic shielding formulae (TRITA-EPP-75-27) contains simple transfer matrices suitable for calculating the quasistatic shielding efficiency for multiple transverse-field and axial-field cylindrical and spherical shields, as well as for estimating leakage fields from long coaxial cables and the normal-incidence transmission of a plane wave through a multiple plane shield. The differences and similarities between these cases are illustrated by means of equivalent circuits and transmission line analogies. The addendum also includes a discussion of a possible heuristic improvement of some shielding formulae. (author)

  11. The Active Muon Shield

    Bezshyiko, Iaroslava

    2016-01-01

    In the SHiP beam-dump of the order of 1011 muons will be produced per second. An active muon-shield is used to magnetically deflect these muons out of the acceptance of the spectrom- eter. This note describes how this shield is modelled and optimized. The SHiP spectrometer is being re-optimized using a conical decay-vessel, and utilizing the possibility to magnetize part of the beam-dump shielding iron. A shield adapted to these new conditions is presented which is significantly shorter and lighter than the shield used in the Technical Proposal (TP), while showing a similar performance.

  12. Self-shielding factors

    Kaul, D.C.

    1982-01-01

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

  13. Terminating Safeguards on Excess Special Nuclear Material: Defense TRU Waste Clean-up and Nonproliferation - 12426

    Hayes, Timothy [Los Alamos National Laboratory, Carlsbad Operations Group (United States); Nelson, Roger [Department Of Energy, Carlsbad Operations Office (United States)

    2012-07-01

    The Department of Energy (DOE) and the National Nuclear Security Administration (NNSA) manages defense nuclear material that has been determined to be excess to programmatic needs and declared waste. When these wastes contain plutonium, they almost always meet the definition of defense transuranic (TRU) waste and are thus eligible for disposal at the Waste Isolation Pilot Plant (WIPP). The DOE operates the WIPP in a manner that physical protections for attractiveness level D or higher special nuclear material (SNM) are not the normal operating condition. Therefore, there is currently a requirement to terminate safeguards before disposal of these wastes at the WIPP. Presented are the processes used to terminate safeguards, lessons learned during the termination process, and how these approaches might be useful for future defense TRU waste needing safeguards termination prior to shipment and disposal at the WIPP. Also described is a new criticality control container, which will increase the amount of fissile material that can be loaded per container, and how it will save significant taxpayer dollars. Retrieval, compliant packaging and shipment of retrievably stored legacy TRU waste has dominated disposal operations at WIPP since it began operations 12 years ago. But because most of this legacy waste has successfully been emplaced in WIPP, the TRU waste clean-up focus is turning to newly-generated TRU materials. A major component will be transuranic SNM, currently managed in safeguards-protected vaults around the weapons complex. As DOE and NNSA continue to consolidate and shrink the weapons complex footprint, it is expected that significant quantities of transuranic SNM will be declared surplus to the nation's needs. Safeguards termination of SNM varies due to the wide range of attractiveness level of the potential material that may be directly discarded as waste. To enhance the efficiency of shipping waste with high TRU fissile content to WIPP, DOE designed an

  14. PWR upper/lower internals shield

    Homyk, W.A. [Indian Point Station, Buchanan, NY (United States)

    1995-03-01

    During refueling of a nuclear power plant, the reactor upper internals must be removed from the reactor vessel to permit transfer of the fuel. The upper internals are stored in the flooded reactor cavity. Refueling personnel working in containment at a number of nuclear stations typically receive radiation exposure from a portion of the highly contaminated upper intervals package which extends above the normal water level of the refueling pool. This same issue exists with reactor lower internals withdrawn for inservice inspection activities. One solution to this problem is to provide adequate shielding of the unimmersed portion. The use of lead sheets or blankets for shielding of the protruding components would be time consuming and require more effort for installation since the shielding mass would need to be transported to a support structure over the refueling pool. A preferable approach is to use the existing shielding mass of the refueling pool water. A method of shielding was devised which would use a vacuum pump to draw refueling pool water into an inverted canister suspended over the upper internals to provide shielding from the normally exposed components. During the Spring 1993 refueling of Indian Point 2 (IP2), a prototype shield device was demonstrated. This shield consists of a cylindrical tank open at the bottom that is suspended over the refueling pool with I-beams. The lower lip of the tank is two feet below normal pool level. After installation, the air width of the natural shielding provided by the existing pool water. This paper describes the design, development, testing and demonstration of the prototype device.

  15. Dual Neutral Particle Beam Interrogation of Intermodal Shipping Containers for Special Nuclear Material

    Keith, Rodney Lyman

    Intermodal shipping containers entering the United States provide an avenue to smuggle unsecured or stolen special nuclear material (SNM). The only direct method fielded to indicate the presence of SNM is by passive photon/neutron radiation detection. Active interrogation using neutral particle beams to induce fission in SNM is a method under consideration. One by-product of fission is the creation of fragments that undergo radioactive decay over a time period on the order of tens of seconds after the initial event. The "delayed" gamma-rays emitted from these fragments over this period are considered a hallmark for the presence of SNM. A fundamental model is developed using homogenized cargos with a SNM target embedded at the center and computationally interrogated using simultaneous neutron and photon beams. Findings from analysis of the delayed gamma emissions from these experiments are intended to mitigate the effects of poor quality information about the composition and disposition of suspect cargo before examination in an active interrogation portal.

  16. Technique for detecting a small magnitude loss of special nuclear material

    Pike, D.H.; Chernick, M.R.; Downing, D.J.

    The detection of losses of special nuclear materials has been the subject of much research in recent years. The standard industry practice using ID/LEID will detect large magnitude losses. Time series techniques such as the Kalman Filter or CUSUM methods will detect small magnitude losses if they occur regularly over a sustained period of time. To date no technique has been proposed which adequately addresses the problem of detecting a small magnitude loss occurring in a single period. This paper proposes a method for detecting a small magnitude loss. The approach makes use of the influence function of Hempel. The influence function measures the effect of a single inventory difference on a group of statistics. An inventory difference for a period in which a loss occurs can be expected to produce an abnormality in the calculated statistics. This abnormality is measurable by the influence function. It is shown that a one period loss smaller in magnitude than the LEID can be detected using this approach

  17. Technical and economical prerequisites of special district-heating nuclear power plant development

    Baturov, B.B.; Boldyrev, V.M.; Losev, V.L.; Sigal, M.V.

    1983-01-01

    Results are presented of technical and economical analysis of advisability of constructing combined Nuclear Power and Heating Plants (NPHP) assuring the possibility of their location near the areas of heat power consumption in case of observing a given degree of radiation safety for population and personnel. Specific features determining the choice of turbine-driven units for such plants are analyzed. Conditions of competiveness of a specialized NPHP with alternative power units, NPHP based on the WWER-1000 reactor and district heating plants (NDHP), are determined. Analysis of design specifications of NPHP with two VK-500 reactor units and structures of capital investments in such a plant reveal that an increase in the total capital investments in the NPHP would not exceed 2% with account of the difference in costs for grid heaters and a corresponding change in dimensions of operation rooms as well as changes in costs of heat removal system (within one site) at the TK turbine replacement by the T turbine

  18. On screening for Special Nuclear Materials (SNMs) with X-ray diffraction

    Harding, G.

    2010-01-01

    A novel detection technique employing X-ray diffraction (XRD) to screen for Special Nuclear Materials (SNMs), in particular for uranium, has been recently proposed. It is based on the interesting fact that uranium (and incidentally, plutonium) has a non-cubic lattice structure, in contrast to all other non-SNM, high-density elements of the Periodic Table. The principle of this screening technique is briefly elucidated by comparing the XRD lines of uranium with those of lead, a material of high atomic number (Z) commonly found in container traffic. Several physical conditions that must be satisfied to enable XRD for SNM screening are considered. To achieve adequate penetration, both of suspicious high-Z materials and their containers, photon energies of 1 MeV and above must be employed. Implications from partial coherence theory for the XRD measurement geometry at such photon energies are presented. The question of multiple scatter degradation of the coherent scatter signal is addressed. Technological considerations relevant to performing XRD at 1 MeV, particularly regarding the radiation source and detector, are discussed. A novel secondary aperture scheme permitting high energy XRD is presented. It is concluded that the importance of the application and the prospect of its feasibility are sufficient to warrant experimental verification.

  19. Evaluating safeguards effectiveness against protracted theft of special nuclear material by insiders

    Al-Ayat, R.; Sicherman, A.

    1991-01-01

    The new draft DOE Material Control and Accountability Order 5633.3 requires that facilities handling special nuclear material (SNM) evaluate their effectiveness against provided theft of SNMProtracted theft means repeated thefts of small quantities of material to accumulate a goal quanfity. To evaluate the safeguards effectiveness against pro thefts, one must addresses several issues: (1) defining relevant time frames for various threat scenarios and delayed detection safeguards: (2) identifying which safeguards come into play more than once because of repeated adversary actions or because of periodic occurrence during the theft time frame (e.g., daily administrative check on presence of material): (3) considering whether the second and subsequent applications of safeguards are different in effectiveness from the first; (4)synthesizing how physical security, material control, and material accountability safeguards combine to provide protection against protracted theft scenarios. In this paper we discuss these issues and describe how we are augmenting the Analytic System and Software for Evaluating Safeguards and Security (ASSESS) to provide the user with a tool for evaluating effectiveness against protracted theft. Currently, the Insider module of ASSESS focuses on evaluating the ''timely'' detection of abrupt theft attempts by various types of single nonviolent insiders. In this paper we describe the approach we're implementing to augment ASSESS to handle various cases of protracted theft attempts

  20. Nuclear Waste Imaging and Spent Fuel Verification by Muon Tomography

    Jonkmans, G.; Anghel, V. N. P.; Jewett, C.; Thompson, M.

    2012-01-01

    This paper explores the use of cosmic ray muons to image the contents of shielded containers and detect high-Z special nuclear materials inside them. Cosmic ray muons are a naturally occurring form of radiation, are highly penetrating and exhibit large scattering angles on high Z materials. Specifically, we investigated how radiographic and tomographic techniques can be effective for non-invasive nuclear waste characterization and for nuclear material accountancy of spent fuel inside dry stor...

  1. Radiation shielding performance of some concrete

    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

  2. Reactor vessel head permanent shield

    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

  3. Some problems raised by the operation of large nuclear turbo-generator sets. Cooling of shielded conductors. Precautionary steps for their thermal dimensioning

    Nisol, R.

    1976-01-01

    The role of the shielded conductors as power dissipation feeder from the generator towards the network is recalled. Their natural cooling limits and the possibilities of forced cooling are examined. The known incidence of short-circuit currents upon the components of the generator-network connection is reported [fr

  4. Benchmark analyses for the ITER bulk shield experiment with EFF-3.0, -3.1 and FENDL-1, -2 nuclear cross-section data

    Fischer, U.; Wu, Y.; Hansen, W.; Richter, D.; Seidel, K.; Unholzer, S.

    1999-01-01

    The present article is part of the summary report on the Consultants' Meeting on the transport sublibrary of the Fusion Evaluated Data Library version 2.0. It reports on the comparison between benchmark experiments on a mock-up of the ITER inboard shield system at FNG, Frascati and Monte Carlo calculations, using different versions of the FENDL and EFF libraries

  5. Wake Shield Target Protection

    Valmianski, Emanuil I.; Petzoldt, Ronald W.; Alexander, Neil B.

    2003-01-01

    The heat flux from both gas convection and chamber radiation on a direct drive target must be limited to avoid target damage from excessive D-T temperature increase. One of the possibilities of protecting the target is a wake shield flying in front of the target. A shield will also reduce drag force on the target, thereby facilitating target tracking and position prediction. A Direct Simulation Monte Carlo (DSMC) code was used to calculate convection heat loads as boundary conditions input into ANSYS thermal calculations. These were used for studying the quality of target protection depending on various shapes of shields, target-shield distance, and protective properties of the shield moving relative to the target. The results show that the shield can reduce the convective heat flux by a factor of 2 to 5 depending on pressure, temperature, and velocity. The protective effect of a shield moving relative to the target is greater than the protective properties of a fixed shield. However, the protective effect of a shield moving under the drag force is not sufficient for bringing the heat load on the target down to the necessary limit. Some other ways of diminishing heat flux using a protective shield are discussed

  6. JENDL special purpose file

    Nakagawa, Tsuneo

    1995-01-01

    In JENDL-3,2, the data on all the reactions having significant cross section over the neutron energy from 0.01 meV to 20 MeV are given for 340 nuclides. The object range of application extends widely, such as the neutron engineering, shield and others of fast reactors, thermal neutron reactors and nuclear fusion reactors. This is a general purpose data file. On the contrary to this, the file in which only the data required for a specific application field are collected is called special purpose file. The file for dosimetry is a typical special purpose file. The Nuclear Data Center, Japan Atomic Energy Research Institute, is making ten kinds of JENDL special purpose files. The files, of which the working groups of Sigma Committee are in charge, are listed. As to the format of the files, ENDF format is used similarly to JENDL-3,2. Dosimetry file, activation cross section file, (α, n) reaction data file, fusion file, actinoid file, high energy data file, photonuclear data file, PKA/KERMA file, gas production cross section file and decay data file are described on their contents, the course of development and their verification. Dosimetry file and gas production cross section file have been completed already. As for the others, the expected time of completion is shown. When these files are completed, they are opened to the public. (K.I.)

  7. New facility shield design criteria

    Howell, W.P.

    1981-07-01

    The purpose of the criteria presented here is to provide standard guidance for the design of nuclear radiation shields thoughout new facilities. These criteria are required to assure a consistent and integrated design that can be operated safely and economically within the DOE standards. The scope of this report is confined to the consideration of radiation shielding for contained sources. The whole body dose limit established by the DOE applies to all doses which are generally distributed throughout the trunk of the body. Therefore, where the whole body is the critical organ for an internally deposited radionuclide, the whole body dose limit applies to the sum of doses received must assure control of the concentration of radionuclides in the building atmosphere and thereby limit the dose from internal sources

  8. Estimation of temperature distribution in a reactor shield

    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

  9. Special problems in nuclear instrumentation. Final report for period ending July 31, 1978

    Spokas, J.J.

    1979-01-01

    The development of special conducting plastics to meet the special needs of radiation dosimetry is reported. The most critical application of the special plastics is in the construction of ionization chambers which provide the best means of accurately measuring absorbed dose and exposure, low-noise signal cables, and electrometers for the ionization chambers are described

  10. Radiation shielding concrete

    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.)

  11. Radiation shielding method for pipes, etc

    Nagao, Tetsuya; Takahashi, Shuichi.

    1988-01-01

    Purpose: To constitute shielding walls of a dense structure around pipes and enable to reduce the wall thickness thereof upon periodical inspection, etc. for nuclear power plants. Constitution: For those portions of pipes requring shieldings, cylindrical vessels surrounding the portions are disposed and connected to a mercury supply system, a mercury discharge system and a freezing system for solidifying mercury. After charging mercury in a tank by way of a supply hose to the cylindrical vessels, the temperature of the mercury is lowered below the freezing point thereof to solidify the mercury while circulating cooling medium, to thereby form dense cylindrical radioactive-ray shielding walls. The specific gravity of mercury is greater than that of lead and, accordingly, the thickness of the shielding walls can be reduced as compared with the conventional wall thickness of the entire laminates. (Takahashi, M.)

  12. Nuclear power newsletter Vol. 3, no. 3, special issue, September 2006

    2006-09-01

    The topics presented in this newsletter are: IAEA's Contribution to Peaceful Use of Nuclear Power by Mr. Sinha; IAEA's Contribution to Peaceful Use of Nuclear Power by Mr. Tipping; Message from the Director of the Division of Nuclear Power; Nuclear power plant operating performance and life cycle management; Improving organizational performance; Coordination of INPRO; Technology development for advanced reactors; Support for non-electric applications of nuclear power; Planned meetings in 2006 and 2007; Division of Nuclear Power Web site link. The first two topics have been indexed separately

  13. Mobile Pit verification system design based on passive special nuclear material verification in weapons storage facilities

    Paul, J. N.; Chin, M. R.; Sjoden, G. E. [Nuclear and Radiological Engineering Program, George W. Woodruff School of Mechanical Engineering, Georgia Institute of Technology, 770 State St, Atlanta, GA 30332-0745 (United States)

    2013-07-01

    A mobile 'drive by' passive radiation detection system to be applied in special nuclear materials (SNM) storage facilities for validation and compliance purposes has been designed through the use of computational modeling and new radiation detection methods. This project was the result of work over a 1 year period to create optimal design specifications to include creation of 3D models using both Monte Carlo and deterministic codes to characterize the gamma and neutron leakage out each surface of SNM-bearing canisters. Results were compared and agreement was demonstrated between both models. Container leakages were then used to determine the expected reaction rates using transport theory in the detectors when placed at varying distances from the can. A 'typical' background signature was incorporated to determine the minimum signatures versus the probability of detection to evaluate moving source protocols with collimation. This established the criteria for verification of source presence and time gating at a given vehicle speed. New methods for the passive detection of SNM were employed and shown to give reliable identification of age and material for highly enriched uranium (HEU) and weapons grade plutonium (WGPu). The finalized 'Mobile Pit Verification System' (MPVS) design demonstrated that a 'drive-by' detection system, collimated and operating at nominally 2 mph, is capable of rapidly verifying each and every weapon pit stored in regularly spaced, shelved storage containers, using completely passive gamma and neutron signatures for HEU and WGPu. This system is ready for real evaluation to demonstrate passive total material accountability in storage facilities. (authors)

  14. Latest experiences in inspecting the inside of BWR vessel shields

    Alberdi, R.; Gonzalez, E.

    2001-07-01

    In the last few years, the owners of BWR nuclear power plants have been forced to address new fuel shield inspection requirements, TECNATOM has responded to this situation by launching the TEIDE projects, which include development of an inspection machine and the corresponding Non-Destructive Tests to examine the inside of this shield. With these projects, TECNATOM has performed more than 12 fuel shield inspections in different countries. This article describes the experience gained in the last three years. (Author)

  15. Evaluation of the shielding integrity of end-shields in PHWR type NPPs

    Sah, B.M.L.; Ramamirtham, B.; Kutty, B.S.

    1996-01-01

    In the new plants (Narora Atomic Power Plants (NAPP) onwards) relatively higher radiation fields exist on the north and south fuelling machine (FM) vault walls of the E1 100m accessible area passages. These fields were first noticed at NAPS-1 and subsequently at NAPS-2 and KAPS-1. Such surveys done at RAPS have indicated that the fields on these walls would come out to be quite low (only 1-2 mR/h) from sources other than that arising from 41 Ar contamination. RAPS/MAPS experience pointed to adequacy of shielding of the FM vault walls and sufficient overall shielding thickness of the end-shields. Further, radiometry tests of end-shields carried out at Kaiga and RAPP 3 and 4 indicated fairly satisfactory and uniform filling of balls. Hence, incomplete filling of water column of the end-shields due to any venting problem was suspected to be one possible reason for the observed high fields in NAPS and Kakrapar Atomic Power Station (KAPS). Since the presence of high radiation fields, both neutron and gamma, is of long-term concern, a special study/measurement of radiation levels on reactor face during high power operation was undertaken. In order to compare the shielding integrity of the older (RAPS/MAPS solid plate type shielding) and newer (NAPS/KAPS steel ball-filled type) end shields, these experiments were done at MAPS-2 and NAPS-2. (author). 2 refs., 2 tabs

  16. Department of Energy (DOE) system for the transportation of strategic quantities of special nuclear material (SQ SNM)

    Dickason, D.P.

    1978-01-01

    Since 1947 DOE and its predecessor agencies, AEC and ERDA, have moved nuclear materials by a variety of commercial and government transportation modes. In the late 1960's world-wide terrorism and other dissident activities prompted the then-AEC to review its procedures for safeguarding SNM. These reviews resulted in immediate and long-range programs for improvement of overall safeguards. Domestic transportation of completed nuclear weapons and SNM used in the weapons program was selected for special consideration. In the early 1970's AEC started the development of a Safe Secure Trailer (SST) to transport nuclear weapons and nuclear components and the development and installation of a high frequency (HF) communications system to assure continuous radio contact between selected highway and rail shipments and Headquarters, Albuquerque Operations (ALO). Late 1974 AEC directed ALO to develop a transportation system to extend weapons-level protection to all AEC SQ SNM shipments and to consolidate, manage, and operate this system. As of September 1976 all SQ SNM was being transported in the Safe Secure DOE (then ERDA) transportation safeguards system, composed of the following principal elements: (1) Transport equipment consisting of Safe Secure Trailers and specially modified towing tractors; Safe Secure Railcars and specially modified escort coaches; and specially designed highway escort vehicles. (2) An automated high-frequency digital radio system that enables continuous communications between the transporting equipment and central control. (3) A courier force that operates all transport equipment (except aircraft and rail power units) and mobile communications equipment; provides armed protection for shipments; and assures proper safety en route. (4) A central Headquarters staff that plans, executes, and controls shipments and directs, manages, and operates the system

  17. Performance evaluation of a commercially available heat flow calorimeter and applicability assessment for safeguarding special nuclear materials

    Bracken, D.S.; Biddle, R.; Rudy, C.

    1998-01-01

    The performance characteristics of a commercially available heat-flow calorimeter will be presented. The heat-flow sensors within the calorimeter are based on thermopile technology with a vendor-quoted sensitivity of 150 microV/mW. The calorimeter is a full-twin design to compensate for ambient temperature fluctuations. The efficacy of temperature fluctuation compensations will also be detailed. Finally, an assessment of design applicability to special nuclear materials control and accountability and safeguarding will be presented

  18. Status report of shielding investigation in Japan

    Shindo, M.

    1964-01-01

    The Japan Atomic Energy Research Institute (JAERI) was established in 1954, and immediately proceeded with the construction of a research reactor. The first symposium in Japan on nuclear energy was held in 1957. Most of the papers presented in the field of reactor shielding were limited to shielding materials and their fabrication. In the first stage of our investigations, our efforts were devoted to practical design studies of reactor shielding. As a result of these studies, it was found that the formulae at hand for calculations were inadequate, but at that time no electronic computer was available in Japan nor were theoretical calculations very actively undertaken. Problems on nuclear ship shielding had been investigated at the Ship Research Institute, since 1956 and many fruitful results had been obtained. About that time the Japan Atomic Industry Forum started activities and took the initiative in organizing shielding research. Research workers in the shipbuilding industry in particular have been seriously studying shielding problems. Few years after the first symposium, problems concerning more fundamental studies were treated by many research workers. Shielding experiments using radioisotopes were carried out and many fruitful results were obtained. They are described in the this paper. Medium size electronic computers became available in Japan, permitting a theoretical study group to make an active contribution. They produced some codes, and their results are also described in the following sections. This constituted the second stage of our investigations. A swimming-pool reactor, JRR-4 (Japan Research Reactor-4), has been under construction at JAERI since 1962 and will become critical in autumn 1964. After characteristic tests it will be a very powerful tool for the shielding investigations. This id the beginning of the third stage of investigations

  19. The function of specialized organization in work safety engineering for nuclear installations

    Salvatore, J.E.L.

    1989-01-01

    The attributions of Brazilian CNEN in the licensing procedures of any nuclear installation are discussed. It is shown that the work safety engineering and industrial safety constitute important functions for nuclear safety. (M.C.K.) [pt

  20. Shielding design to obtain compact marine reactor

    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)

  1. Mechanical design of the TIBER breeding shield

    Rathke, J.; Deutsch, L. (Grumman Corp., Bethpage, NY (USA). Space Systems Div.)

    1989-04-01

    TIBER features a segmented shield assembly that provides the nuclear shielding for the superconducting toroidal field coils. In addition to its primary function, the shield also provides tritium breeding through the use of water coolant that contains 16 wt% dissolved lithium nitrate. Because the TIBER reactor need not provide electrical power, the coolant is maintained at low pressure (0.2 MPa) and low temperature (75/sup 0/C). The shield is made in several segments to facilitate assembly and allow for replacement of high heat flux components (divertor blades). The segments are designated as inboard, outboard, upper, lower, and divertor modules. In total, there are 96 separate modules in the machine, consisting of six different types. The design features of the different modules vary primarily depending on the thickness of the shield in a given location. The very thick outboard shield has a breeding zone in the inboard portion of the module, with a shielding zone behind it. The breeding zone consists of a stainless steel casing filled with beryllium spheres. The shielding zone consists of the same casing filled with steel spheres. Both of these zones have lithiated water circulated throughout to provide cooling and breeding. In zones with minimal thickness, tungsten alloys are used to achieve the required shielding. These alloys are incoprorated in subassemblies utilizing stainless steel casings surrounding blocks of tungsten heavy metal alloy. These are infiltrated with lead on final assembly to form a thermally continuous panel. Several of these panels are then assembled into an outer stainless steel case to form an inboard module. These modules also use the lithiated coolant. The details of the design are presented and discussed. (orig.).

  2. INTOR radiation shielding for personnel access

    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

  3. Locating sensors for detecting source-to-target patterns of special nuclear material smuggling: a spatial information theoretic approach.

    Przybyla, Jay; Taylor, Jeffrey; Zhou, Xuesong

    2010-01-01

    In this paper, a spatial information-theoretic model is proposed to locate sensors for detecting source-to-target patterns of special nuclear material (SNM) smuggling. In order to ship the nuclear materials from a source location with SNM production to a target city, the smugglers must employ global and domestic logistics systems. This paper focuses on locating a limited set of fixed and mobile radiation sensors in a transportation network, with the intent to maximize the expected information gain and minimize the estimation error for the subsequent nuclear material detection stage. A Kalman filtering-based framework is adapted to assist the decision-maker in quantifying the network-wide information gain and SNM flow estimation accuracy.

  4. Standard guide for application of radiation monitors to the control and physical security of special nuclear material

    Anon.

    1989-01-01

    This guide briefly describes the state-of-the-art of radiation monitors for detecting special nuclear material (SNM) in order to establish the context in which to write performance standards for the monitors. This guide extracts information from technical documentation to provide information for selecting, calibrating, testing, and operating such radiation monitors when they are used for the control and protection of SNM. This guide offers an unobtrusive means of searching pedestrians, packages, and motor vehicles for concealed SNM as one part of a nuclear material control or security plan for nuclear materials. The radiation monitors can provide an efficient, sensitive, and reliable means of detecting the theft of small quantities of SNM while maintaining a low likelihood of nuisance alarms

  5. Standard guide for application of radiation monitors to the control and physical security of special nuclear material

    American Society for Testing and Materials. Philadelphia

    1999-01-01

    1.1 This guide briefly describes the state-of-the-art of radiation monitors for detecting special nuclear material (SNM) (see 3.1.11) in order to establish the context in which to write performance standards for the monitors. This guide extracts information from technical documentation to provide information for selecting, calibrating, testing, and operating such radiation monitors when they are used for the control and protection of SNM. This guide offers an unobtrusive means of searching pedestrians, packages, and motor vehicles for concealed SNM as one part of a nuclear material control or security plan for nuclear materials. The radiation monitors can provide an efficient, sensitive, and reliable means of detecting the theft of small quantities of SNM while maintaining a low likelihood of nuisance alarms. 1.2 Dependable operation of SNM radiation monitors rests on selecting appropriate monitors for the task, operating them in a hospitable environment, and conducting an effective program to test, calibrat...

  6. Locating Sensors for Detecting Source-to-Target Patterns of Special Nuclear Material Smuggling: A Spatial Information Theoretic Approach

    Xuesong Zhou

    2010-08-01

    Full Text Available In this paper, a spatial information-theoretic model is proposed to locate sensors for detecting source-to-target patterns of special nuclear material (SNM smuggling. In order to ship the nuclear materials from a source location with SNM production to a target city, the smugglers must employ global and domestic logistics systems. This paper focuses on locating a limited set of fixed and mobile radiation sensors in a transportation network, with the intent to maximize the expected information gain and minimize the estimation error for the subsequent nuclear material detection stage. A Kalman filtering-based framework is adapted to assist the decision-maker in quantifying the network-wide information gain and SNM flow estimation accuracy.

  7. Analysis and evaluation of a shielding for scum mensurations

    Viant Garrido, E.; Fernandez Rondon, M.; Torres Berdeguez, M.; Fernandez, M.; Perez Sanchez, D.; Falcon, L.

    1998-01-01

    In this work the evaluation of a shielding is presented developed by the Nuclear Technology Center of carry out scum mensurations in systems of mensuration gamma spectrometric. The evaluation of the capacity of attenuation of the shielding one carries out starting from the comparisons of the count rates with systems used by anteriority by the Laboratory of Environmental Radiological Surveillance

  8. Shielded regenerative neutron detector

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

    1978-01-01

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

  9. Radiation shielding plate

    Kobayashi, Torakichi; Sugawara, Takeo.

    1983-01-01

    Purpose: To reduce the weight and stabilize the configuration of a radiation shielding plate which is used in close contact with an object to be irradiated with radiation rays. Constitution: The radiation shielding plate comprises a substrate made of lead glass and a metallic lead coating on the surface of the substrate by means of plating, vapor deposition or the like. Apertures for permeating radiation rays are formed to the radiation shielding plate. Since the shielding plate is based on a lead glass plate, a sufficient mechanical strength can be obtained with a thinner structure as compared with the conventional plate made of metallic lead. Accordingly, if the shielding plate is disposed on a soft object to be irradiated with radiation rays, the object and the plate itself less deform to obtain a radiation irradiation pattern with distinct edges. (Moriyama, K.)

  10. Necessity of Internal Monitoring for Nuclear Medicine Staff in a Large Specialized Chinese Hospital.

    Wang, Hong-Bo; Zhang, Qing-Zhao; Zhang, Zhen; Hou, Chang-Song; Li, Wen-Liang; Yang, Hui; Sun, Quan-Fu

    2016-04-12

    This work intends to quantify the risk of internal contaminations in the nuclear medicine staff of one hospital in Henan province, China. For this purpose, the criteria proposed by the International Atomic Energy Agency (IAEA) to determine whether it is necessary to conduct internal individual monitoring was applied to all of the 18 nuclear medicine staff members who handled radionuclides. The activity of different radionuclides used during a whole calendar year and the protection measures adopted were collected for each staff member, and the decision as to whether nuclear medicine staff in the hospital should be subjected to internal monitoring was made on the basis of the criteria proposed by IAEA. It is concluded that for all 18 members of the nuclear medicine staff in the hospital, internal monitoring is required. Internal exposure received by nuclear medicine staff should not be ignored, and it is necessary to implement internal monitoring for nuclear medicine staff routinely.

  11. Transport of chemically bonded nuclear energy in a closed cycle with special consideration to energy disconnection

    Ossami, S.

    1976-01-01

    The article describes the utilisation of nuclear energy in the form of 'nuclear long-distance energy'. Heat produced by nuclear fission is bonded to a reversible chemical reaction (cracking gas) which release the heat again at the place of comsumption by catalytic transformation. The article deals in particular with the process of methane cracking/methanisation, the disconnection of the energy (heat) by the methanisation process and the decisive role of the methanisation catalyzers. (orig.) [de

  12. Legislation for the countermeasures on special issues of nuclear safety regulations

    Cho, Byung Sun; Lee, Mo Sung; Chung, Gum Chun; Kim, Hak Man; Oh, Ho Chul [Chongju Univ., Cheongju (Korea, Republic of)

    2003-02-15

    Since the present legal system on nuclear safety regulation has some problems that refer to contents of regulatory provisions, this mid-report has preformed research on the legal basic theory of nuclear safety regulation. And then secondly this report analyzed the problems of each provisions and suggested the revision drafts on the basis of analyzing problems and the undergoing theory of nuclear safety regulation. In order to interpret easily this report finally took the cases of judicial precedents on nuclear safety regulation in USA, Germany, Japan and Korea.

  13. The concept of nuclear-weapon-free zones, with special reference to East Asia

    Prawitz, J.

    1997-01-01

    The concept of nuclear-weapon-free zone, as it has developed from political deliberation since mid-1950s, today covers a spectrum of arrangements, geographically ranging from whole continent like Latin America to a corridor in central Europe, and functionally serving the purposes of preventing the spread of nuclear weapons and avoiding nuclear war. Among the nuclear-weapon-free zones proposed but not established, one is now being negotiated and two are subject to investigations published in official reports. These are the proposed zones in ASEAN area, The Middle East and Nordic Europe

  14. Legislation for the countermeasures on special issues of nuclear safety regulations

    Cho, Byung Sun; Lee, Mo Sung; Chung, Gum Chun; Kim, Hak Man; Oh, Ho Chul

    2003-02-01

    Since the present legal system on nuclear safety regulation has some problems that refer to contents of regulatory provisions, this mid-report has preformed research on the legal basic theory of nuclear safety regulation. And then secondly this report analyzed the problems of each provisions and suggested the revision drafts on the basis of analyzing problems and the undergoing theory of nuclear safety regulation. In order to interpret easily this report finally took the cases of judicial precedents on nuclear safety regulation in USA, Germany, Japan and Korea

  15. CNN's evaluation of the radiological monitoring special programs at the sites of nuclear installations

    Lareyne, O.; Marugan Tovar, I.; Martinez Moreno, M.; Sanz Alduan, M. T.

    2011-01-01

    In 2008, occurred in the Spanish plants various events related to the existence on the sites of the Spanish nuclear facilities or areas dotted with radioactive contamination outside the buildings, events that were reported to the Nuclear Safety Council (CSN) in accordance with the provisions of the regulation, as a result of the analysis of those events, the CSN issued in July 2008, a complementary technique INSTRUCTIONAL requiring the implementation of environmental monitoring programs in outdoor areas within sites nuclear tacilities. First, the CSN nuclear facilities referred to the programs and methodologies intended to perform such monitoring, and, in late 2009, reports of program results.

  16. Ligand field effects in the nuclear magnetic shielding of nitrogen-15 and cobalt-59 in bent nitrosyl complexes of cobalt(III)

    Duffin, P.A.; Larkworthy, L.F.; Mason, J.; Stephens, A.N.; Thompson, R.M.

    1987-01-01

    A range of square-pyramidal complexes of cobalt(III) with a bent apical nitrosyl ligand has been prepared and examined by 15 N and 59 Co NMR spectroscopy, in a study of nephelauxetic and spectrochemical effects at the metal and nitrogen nuclei in the bent Co-NO chromophore. The basal ligands in this comparison include dithiocarbamate, quadridentate Schiff base or porphine, and bis-chelating diamine or oximate, so as to give S 4 , S 2 N 2 , N 4 , OONN, or ONON coordination in the plane and a range of substituents in the chelate and phenylene rings. The shielding of both cobalt and nitrogen tends to decrease with decrease in the M(d) → π*(NO) back-bonding, as indicated by the MN and NO bond distances, the MNO angle and the NO stretching frequency. The shieldings decrease from sulfur to nitrogen to oxygen coligators and also with electron withdrawal by ring substituents (and vice versa), i.e. with decrease in the ligand field splitting and in the nephelauxetism of the coligands. These parallelisms of the cobalt and nitrogen shielding accord with the orbital theory that was developed to explain the bending of the MNO ligand and influences of the metal and coligands. Significant interdependence of spectrochemical and nephelauxetic effects at cobalt and nitrogen arises from the degree of overlap and similarity in energies of the frontier orbitals for the paramagnetic circulation at nitrogen [n(N) → π* (NO)] and at cobalt (d-d). 43 references, 5 figures, 2 tables

  17. Radiation protection and shielding standards for the 1980s

    Trubey, D.K.

    1982-01-01

    The American Nuclear Society (ANS) is a standards-writing organization member of the American National Standards Institute (ANSI). The ANS Standards Committee has a subcommittee denoted ANS-6, Radiation Protection and Shielding, whose charge 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. This paper is a progress report of this subcommittee. Significant progress has been made since the last comprehensive report to the Society

  18. Analysis of the emergency plan of Angra dos Reis Nuclear Power Plants: a critical view related to accessibility and mobility of people with special needs

    Correa, Luanda C.A.; Correa, Samanda C.A.; Santos, Geissa; Souza, Edmilson M.

    2014-01-01

    This work intends to make a critical analysis of the emergency plan of the Angra dos Reis Nuclear Power Plants related to appropriate transportation, accommodation and infrastructure for people with special needs

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

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

    2016-06-27

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

  20. On-site transportation and handling of uranium-233 special nuclear material: Preliminary hazards and accident analysis. Final

    Solack, T.; West, D.; Ullman, D.; Coppock, G.; Cox, C.

    1995-01-01

    U-233 Special Nuclear Material (SNM) currently stored at the T-Building Storage Areas A and B must be transported to the SW/R Tritium Complex for repackaging. This SNM is in the form of oxide powder contained in glass jars which in turn are contained in heat sealed double polyethylene bags. These doubled-bagged glass jars have been primarily stored in structural steel casks and birdcages for approximately 20 years. The three casks, eight birdcages, and one pail/pressure vessel will be loaded onto a transport truck and moved over an eight day period. The Preliminary Hazards and Accident Analysis for the on-site transportation and handling of Uranium-233 Special Nuclear Material, documented herein, was performed in accordance with the format and content guidance of DOE-STD-3009-94, Preparation Guide for US Department of Energy Nonreactor Nuclear Facility Safety Analysis Reports, dated July 1994, specifically Chapter Three, Hazard and Accident Analysis. The Preliminary Hazards Analysis involved detailed walkdowns of all areas of the U-233 SNM movement route, including the T-Building Storage Area A and B, T-Building truck tunnel, and the roadway route. Extensive discussions were held with operations personnel from the Nuclear Material Control Group, Nuclear Materials Accountability Group, EG and G Mound Security and the Material Handling Systems Transportation Group. Existing documentation related to the on-site transportation of hazardous materials, T-Building and SW/R Tritium Complex SARs, and emergency preparedness/response documentation were also reviewed and analyzed to identify and develop the complete spectrum of energy source hazards

  1. Introduction to This Special Issue of Georgia Social Science Journal on the Nuclear Arms Race.

    Totten, Sam

    1985-01-01

    Of the many serious problems now facing the world, the most important is threat of nuclear destruction. Social studies educators should provide ample classroom time for discussing and examining the facts concerning nuclear warfare, and they must make sure that various points of view are presented. (RM)

  2. 77 FR 22362 - Exemption Requests for Special Nuclear Material License SNM-362, Department of Commerce...

    2012-04-13

    ... Nuclear Material License SNM-362, Department of Commerce, Gaithersburg, MD AGENCY: Nuclear Regulatory... Commerce, National Institutes of Standards and Technology (NIST) in Gaithersburg, Maryland. NIST requested... within the Department of Commerce. The SNM license was renewed in 1979, 1985, 1991, and 1997. The current...

  3. Development of Neutron Shielding Material for Cask and Accelerator

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

    2008-01-01

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

  4. The Unsuspected Roles of Chemistry in Nuclear Power Plants: Special Chemical Technologies for Enhanced Safety and Increased Performance

    Sempere Belda, Luis

    2008-01-01

    The plant's chemists main responsibility is the establishment and monitoring of an adequate water chemistry to minimize corrosion and in PWRs, to control the neutron flux. But this is by no means the only way in which chemical applications contribute to the performance and safety of a NPP during its entire life: The use of special coatings and treatment protects the plant's components from aggressive environmental conditions. The chemical scale removal in steam generators improves the power output of aging plants, helping even to achieve permissions for NPP life extension. The use of special adhesives can replace welding in complicated or high-dose areas, even underwater. And chemical decontamination is used to remove activity from the components of the primary circuit prior to maintenance or replacement works in order to decrease the radiation exposure of the plant's personnel, employing revolutionary methods of waste minimization to limit the amount of generated radioactive waste to a minimum. The AREVA Group, in its pursue of excellence in all stages of the nuclear cycle, has devoted years of research and development to be able to provide the most advanced technological solutions in this field. The awareness of the existing possibilities will help present and future nuclear professionals, chemists and non-chemists alike, to benefit from the years of experience and continuous development in chemical technologies at the service of the nuclear industry. (authors)

  5. Educational Research Centre of the Joint Institute for Nuclear Research and students training on the 'Medical Physics' speciality

    Ivanova, S.P.; )

    2005-01-01

    The Educational Research Centre (ERC) of the Joint Institute for Nuclear Research is the place of joint activity of the JINR, Moscow State University (MSU) and Moscow Engineering Physical Institute (MEFI) on students training by a broadened circle of specialities with introduction of new educational forms. Active application of medical accelerator beams of the JINR Laboratory of Nuclear Beams becomes a reason for implementation of a new training chair in the MEFI on the JINR base - the Physical methods in applied studies in the medicine chair. For the 'medical physics' trend development in 2003 the workshop on discussion both curricula and teaching methodic by the speciality was held. One the Educational Research Centre main activities is both organization and conducting an international scientific schools and training courses. The International student School 'Nuclear-Physical Methods and Accelerators is the most popular and traditional. The principal aim of these schools and courses is familiarization of students and postgraduates with last achievement and and contemporary problems of applied medical physics. The school audience is a students and postgraduates of ERC, MSU, MEFI, and an institutes of Poland, Hungary, Slovakia, France, Czech and Bulgaria

  6. The Analysis of the System of special water purification of Beloyarskaya Nuclear Power Plant unit BN-800

    Valtseva, A. I.; Bibik, I. S.

    2017-11-01

    This article discusses how the latest system of special water purification KPF-30, designed specifically for the fourth power unit of Beloyarskaya nuclear power plant, which has a number of advantages over other water purification systems as chemical-physical and technical-economic, environmental, and other industrial indicators. The scheme covered in this article systems of special water purification involves the use of a hydrocyclone at the preliminary stage of water treatment, as a worthy alternative to ion-exchange filters, which can significantly reduce the volume of toxic waste. The world community implements the project of closing the nuclear fuel cycle, there is a need to improve the reliability of the equipment for safe processes and development of critical and supercritical parameters in the nuclear industry. Essentially, on operated NPP units, the only factor that can cost-effectively optimize to improve the reliability of equipment is the water chemistry. System KPF30 meets the principles and criteria of ecological safety, demonstrating the justification for reagent less method of water treatment on the main stages, in which no formation of toxic wastes, leading to irreversible consequences of environmental pollution and helps to conserve water.

  7. The Unsuspected Roles of Chemistry in Nuclear Power Plants: Special Chemical Technologies for Enhanced Safety and Increased Performance

    Sempere Belda, Luis [AREVA NP GmbH, An AREVA and SIEMENS Company, P.O. Box 1109, Erlangen (Germany)

    2008-07-01

    The plant's chemists main responsibility is the establishment and monitoring of an adequate water chemistry to minimize corrosion and in PWRs, to control the neutron flux. But this is by no means the only way in which chemical applications contribute to the performance and safety of a NPP during its entire life: The use of special coatings and treatment protects the plant's components from aggressive environmental conditions. The chemical scale removal in steam generators improves the power output of aging plants, helping even to achieve permissions for NPP life extension. The use of special adhesives can replace welding in complicated or high-dose areas, even underwater. And chemical decontamination is used to remove activity from the components of the primary circuit prior to maintenance or replacement works in order to decrease the radiation exposure of the plant's personnel, employing revolutionary methods of waste minimization to limit the amount of generated radioactive waste to a minimum. The AREVA Group, in its pursue of excellence in all stages of the nuclear cycle, has devoted years of research and development to be able to provide the most advanced technological solutions in this field. The awareness of the existing possibilities will help present and future nuclear professionals, chemists and non-chemists alike, to benefit from the years of experience and continuous development in chemical technologies at the service of the nuclear industry. (authors)

  8. HPGe detector shielding adjustment

    Trnkova, L.; Rulik, P.

    2008-01-01

    Low-level background shielding of HPGe detectors is used mainly for environmental samples with very low content of radionuclides. National Radiation Protection Institute (SURO) in Prague is equipped with 14 HPGe detectors with relative efficiency up to 150%. The detectors are placed in a room built from materials with low content of natural radionuclides and equipped with a double isolation of the floor against radon. Detectors themselves are placed in lead or steel shielding. Steel shielding with one of these detectors with relative efficiency of 100% was chosen to be rebuilt to achieve lower minimum detectable activity (MDA). Additional lead and copper shielding was built up inside the original steel shielding to reduce the volume of the inner space and filled with nitrogen by means of evaporating liquid nitrogen. The additional lead and copper shielding, consequent reduction of the inner volume and supply of evaporated nitrogen, caused a decrease of the background count and accordingly MDA values as well. The effect of nitrogen evaporation on the net areas of peaks belonging to radon daughters is significant. The enhanced shielding adjustment has the biggest influence in low energy range, what can be seen in collected data. MDA values in energy range from 30 keV to 400 keV decreased to 0.65-0.85 of original value, in energy range from 400 keV to 2 MeV they fell to 0.70-0.97 of original value. (authors)

  9. Shielding in experimental areas

    Stevens, A.; Tarnopolsky, G.; Thorndike, A.; White, S.

    1979-01-01

    The amount of shielding necessary to protect experimental detectors from various sources of background radiation is discussed. As illustrated an experiment has line of sight to sources extending approx. 90 m upstream from the intersection point. Packing a significant fraction of this space with shielding blocks would in general be unacceptable because primary access to the ring tunnel is from the experimental halls. (1) From basic machine design considerations and the inherent necessity to protect superconducting magnets it is expected that experimental areas in general will be cleaner than at any existing accelerator. (2) Even so, it will likely be necessary to have some shielding blocks available to protect experimental apparatus, and it may well be necessary to have a large amount of shielding available in the WAH. (3) Scraping will likely have some influence on all halls, and retractable apparatus may sometimes be necessary. (4) If access to any tunnel is needed to replace a magnet, one has 96 h (4 days) available to move shielding away to permit access without additional downtime. This (the amount of shielding one can shuffle about in 96 h) is a reasonable upper limit to shielding necessary in a hall

  10. Radiation shielding activities at IDOM

    Ordóñez, César Hueso; Gurpegui, Unai Cano; Valiente, Yelko Chento; Poveda, Imanol Zamora, E-mail: cesar.hueso@idom.com [IDOM, Consulting, Engineering and Architecture, S.A.U, Vizcaya (Spain)

    2017-07-01

    When human activities have to be performed under ionising radiation environments the safety of the workers must be guaranteed. Usually three principles are used to accomplish with ALARA (As Low As Reasonably Achievable) requirements: the more distance between the source term and the worker, the better; the less time spent to arrange any task, the better; and, once the previous principles are optimized should the exposure of the workers continues being above the regulatory limits, shielding has to be implemented. Through this paper some different examples of IDOM's shielding design activities are presented. Beginning with the gamma collimators for the Jules Horowitz Reactor, nuclear fuel's behaviour researching facility, where the beam path crosses the reactor's containment walls and is steered up to a gamma detector where the fuel spectrum is analysed and where the beam has to be attenuated several orders of magnitude in a short distance. Later it is shown IDOM’s approach for the shielding of the Emergency Control Management Center of Asociación Nuclear Ascó-Vandellòs-II NPPs, a bunker designed to withstand severe accident conditions and to support the involved staff during 30 days, considering the outside radioactive cloud and the inside source term that filtering units become as they filter the incoming air. And finally, a general approach to this kind of problems is presented, since the study of the source term considering all the possible contributions, passing through the material selection and the thicknesses calculation until the optimization of the materials. (author)

  11. Radiation shielding activities at IDOM

    Ordóñez, César Hueso; Gurpegui, Unai Cano; Valiente, Yelko Chento; Poveda, Imanol Zamora

    2017-01-01

    When human activities have to be performed under ionising radiation environments the safety of the workers must be guaranteed. Usually three principles are used to accomplish with ALARA (As Low As Reasonably Achievable) requirements: the more distance between the source term and the worker, the better; the less time spent to arrange any task, the better; and, once the previous principles are optimized should the exposure of the workers continues being above the regulatory limits, shielding has to be implemented. Through this paper some different examples of IDOM's shielding design activities are presented. Beginning with the gamma collimators for the Jules Horowitz Reactor, nuclear fuel's behaviour researching facility, where the beam path crosses the reactor's containment walls and is steered up to a gamma detector where the fuel spectrum is analysed and where the beam has to be attenuated several orders of magnitude in a short distance. Later it is shown IDOM’s approach for the shielding of the Emergency Control Management Center of Asociación Nuclear Ascó-Vandellòs-II NPPs, a bunker designed to withstand severe accident conditions and to support the involved staff during 30 days, considering the outside radioactive cloud and the inside source term that filtering units become as they filter the incoming air. And finally, a general approach to this kind of problems is presented, since the study of the source term considering all the possible contributions, passing through the material selection and the thicknesses calculation until the optimization of the materials. (author)

  12. Evaluation of Neutron shielding efficiency of Metal hydrides

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

    2012-05-15

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

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

    Kyllingstad, G.

    1977-01-01

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

  14. Emerging trends in forensic science with special emphasis on nuclear and radiochemistry

    Krishnamurthy, Rukmani

    2011-01-01

    Forensic science uses the basic principles of all physical and natural science and have evolved many domain of its owns, like Anthropometry, fingerprint, Foot print, ballistics, documentation, Forensic Biology and Serology, Forensic Chemistry, Nuclear forensic science, Forensic Physic, Toxicology, Odontology, Forensic DNA, Cyber Forensic, Forensic Psychology, Forensic engineering etc., which provides a fool prove scientific aid to criminal justice administration. Nuclear forensic science is a fairly young discipline and only a small number of laboratories are active practitioners. However, the number of incidents of illicit trafficking reported and furthermore, the threat of nuclear terrorism calls for preparedness and for effective tools providing hints on the origin of the material and thus on the perpetrator. The determination of characteristic parameters is subject to ongoing research and development work in a number of nuclear measurement laboratories. Parameters like isotopic composition, chemical impurities, age of the material, macroscopic parameters and microstructure provide clues on the origin and on the intended use of the material. Today, nuclear forensics has reached a high degree of maturity and it is highly relevant in the areas of non-proliferation and of nuclear security. Continued development activities and strengthened international cooperation will be of key importance for the perfection of the discipline of nuclear forensics

  15. Shielding member for thermonuclear device

    Onozuka, Masanori

    1997-06-30

    In a thermonuclear device for shielding fast neutrons by shielding members disposed in a shielding vessel (vacuum vessel and structures such as a blanket disposed in the vacuum vessel), the shielding member comprises a large number of shielding wires formed fine and short so as to have elasticity. The shielding wires are sealed in a shielding vessel together with water, and when the width of the shielding vessel is changed, the shielding wires follow after the change of the width while elastically deforming in the shielding vessel, so that great stress and deformation are not formed thereby enabling to improve reliability. In addition, the length, the diameter and the shape of each of the shielding wires can be selected in accordance with the shielding space of the shielding vessel. Even if the shape of the shielding vessel is complicated, the shielding wires can be inserted easily. Accordingly, the filling rate of the shielding members can be changed easily. It can be produced more easily compared with a conventional spherical pebbles. It can be produced more easily than existent spherical shielding pebbles thereby enabling to reduce the production cost. (N.H.)

  16. Onboard radiation shielding estimates for interplanetary manned missions

    Totemeier, A.; Jevremovic, T.; Hounshel, D.

    2004-01-01

    The main focus of space related shielding design is to protect operating systems, personnel and key structural components from outer space and onboard radiation. This paper summarizes the feasibility of a lightweight neutron radiation shield design for a nuclear powered, manned space vehicle. The Monte Carlo code MCNP5 is used to determine radiation transport characteristics of the different materials and find the optimized shield configuration. A phantom torso encased in air is used to determine a dose rate for a crew member on the ship. Calculation results indicate that onboard shield against neutron radiation coming from nuclear engine can be achieved with very little addition of weight to the space vehicle. The selection of materials and neutron transport analysis as presented in this paper are useful starting data to design shield against neutrons generated when high-energy particles from outer space interact with matter on the space vehicle. (authors)

  17. Heavy metal oxide glasses as gamma rays shielding material

    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.

  18. Heavy metal oxide glasses as gamma rays shielding material

    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.

  19. A special device used for measuring waste gas flow rate in the vent channel of Qinshan Nuclear Power Plant

    Zhang Yingjun; Zong Guifang; Shi Huaming; Yang Huimin; Jiang Yuana.

    1988-01-01

    A special Venturi-Pitot complex device is discribed which is used for measuring waste gas flow rate in the vent channel of Qinshan nuclear power plant. The device is located at the center of the channel. It can produce enlarged differential pressure signal under the condition of low gas velocity. And the flow resistance of this device is negligible. Experiments to determine the ratio of the velocity at the center of the channel to the average velocity were performed on a 1:12 test model. The special device was calibrated in a closed wind tunnel and its discharge coefficient was obtained. The uncertainty is ±3.5% and the nonlinearity is ±1.3%. The enlargement ratio and the discharge coefficient of the device are also deduced analytically on the basis of hydrodynamics theory

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

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

    1993-04-15

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