WorldWideScience

Sample records for reactor structural components

  1. Mechanical components: fabrication of major reactor structures

    International Nuclear Information System (INIS)

    Nicholson, S.

    1985-01-01

    The paper examines the validity of criticisms of quality assurance of mechanical plant and welded products within major reactor structures, taking into account experience gained on the AGR's. Various constructive recommendations are made aimed at furthering the objectives of quality assurance in the nuclear industry and making it more cost-effective. Current levels of quality related costs in the fabrication industry are provided as a basis for discussion. (U.K.)

  2. Surface modification method for reactor incore structural component

    International Nuclear Information System (INIS)

    Obata, Minoru; Sudo, Akira.

    1996-01-01

    A large number of metal or ceramic small spheres accelerated by pressurized air are collided against a surface of a reactor incore structures or a welded surface of the structural components, and then finishing is applied by polishing to form compression stresses on the surface. This can change residual stresses into compressive stress without increasing the strength of the surface. Accordingly, stress corrosion crackings of the incore structural components or welded portions thereof can be prevented thereby enabling to extend the working life of equipments. (T.M.)

  3. Statistical techniques for the identification of reactor component structural vibrations

    International Nuclear Information System (INIS)

    Kemeny, L.G.

    1975-01-01

    The identification, on-line and in near real-time, of the vibration frequencies, modes and amplitudes of selected key reactor structural components and the visual monitoring of these phenomena by nuclear power plant operating staff will serve to further the safety and control philosophy of nuclear systems and lead to design optimisation. The School of Nuclear Engineering has developed a data acquisition system for vibration detection and identification. The system is interfaced with the HIFAR research reactor of the Australian Atomic Energy Commission. The reactor serves to simulate noise and vibrational phenomena which might be pertinent in power reactor situations. The data acquisition system consists of a small computer interfaced with a digital correlator and a Fourier transform unit. An incremental tape recorder is utilised as a backing store and as a means of communication with other computers. A small analogue computer and an analogue statistical analyzer can be used in the pre and post computational analysis of signals which are received from neutron and gamma detectors, thermocouples, accelerometers, hydrophones and strain gauges. Investigations carried out to date include a study of the role of local and global pressure fields due to turbulence in coolant flow and pump impeller induced perturbations on (a) control absorbers, (B) fuel element and (c) coolant external circuit and core tank structure component vibrations. (Auth.)

  4. Residual stress improving method for reactor structural component and residual stress improving device therefor

    Energy Technology Data Exchange (ETDEWEB)

    Enomoto, Kunio; Otaka, Masahiro; Kurosawa, Koichi; Saito, Hideyo; Tsujimura, Hiroshi; Tamai, Yasukata; Urashiro, Keiichi; Mochizuki, Masato

    1996-09-03

    The present invention is applied to a BWR type reactor, in which a high speed jetting flow incorporating cavities is collided against the surface of reactor structural components to form residual compression stresses on the surface layer of the reactor structural components thereby improving the stresses on the surface. Namely, a water jetting means is inserted into the reactor container filled with reactor water. Purified water is pressurized by a pump and introduced to the water jetting means. The purified water jetted from the water jetting means and entraining cavities is abutted against the surface of the reactor structural components. With such procedures, since the purified water is introduced to the water jetting means by the pump, the pump is free from contamination of radioactive materials. As a result, maintenance and inspection for the pump can be facilitated. Further, since the purified water injection flow entraining cavities is abutted against the surface of the reactor structural components being in contact with reactor water, residual compression stresses are exerted on the surface of the reactor structural components. As a result, occurrence of stress corrosion crackings of reactor structural components is suppressed. (I.S.)

  5. Residual stress improving method for reactor structural component and residual stress improving device therefor

    International Nuclear Information System (INIS)

    Enomoto, Kunio; Otaka, Masahiro; Kurosawa, Koichi; Saito, Hideyo; Tsujimura, Hiroshi; Tamai, Yasukata; Urashiro, Keiichi; Mochizuki, Masato.

    1996-01-01

    The present invention is applied to a BWR type reactor, in which a high speed jetting flow incorporating cavities is collided against the surface of reactor structural components to form residual compression stresses on the surface layer of the reactor structural components thereby improving the stresses on the surface. Namely, a water jetting means is inserted into the reactor container filled with reactor water. Purified water is pressurized by a pump and introduced to the water jetting means. The purified water jetted from the water jetting means and entraining cavities is abutted against the surface of the reactor structural components. With such procedures, since the purified water is introduced to the water jetting means by the pump, the pump is free from contamination of radioactive materials. As a result, maintenance and inspection for the pump can be facilitated. Further, since the purified water injection flow entraining cavities is abutted against the surface of the reactor structural components being in contact with reactor water, residual compression stresses are exerted on the surface of the reactor structural components. As a result, occurrence of stress corrosion crackings of reactor structural components is suppressed. (I.S.)

  6. High-temperature-structural design and research and development for reactor system components

    International Nuclear Information System (INIS)

    Matsumura, Makoto; Hada, Mikio

    1985-01-01

    The design of reactor system components requires high-temperature-structural design guide with the consideration of the creep effect of materials related to research and development on structural design. The high-temperature-structural design guideline for the fast prototype reactor MONJU has been developed under the active leadership by Power Reactor and Nuclear Fuel Development Corporation and Toshiba has actively participated to this work with responsibility on in-vessel components, performing research and development programs. This paper reports the current status of high-temperature-structural-design-oriented research and development programs and development of analytical system including stress-evaluation program. (author)

  7. On the classification of structures, systems and components of nuclear research and test reactors

    International Nuclear Information System (INIS)

    Mattar Neto, Miguel

    2009-01-01

    The classification of structures, systems and components of nuclear reactors is a relevant issue related to their design because it is directly associated with their safety functions. There is an important statement regarding quality standards and records that says Structures, systems, and components important to safety shall be designed, fabricated, erected, and tested to quality standards commensurate with the importance of the safety functions to be performed. The definition of the codes, standards and technical requirements applied to the nuclear reactor design, fabrication, inspection and tests may be seen as the main result from this statement. There are well established guides to classify structures, systems and components for nuclear power reactors such as the Pressurized Water Reactors but one can not say the same for nuclear research and test reactors. The nuclear reactors safety functions are those required to the safe reactor operation, the safe reactor shutdown and continued safe conditions, the response to anticipated transients, the response to potential accidents and the control of radioactive material. So, it is proposed in this paper an approach to develop the classification of structures, systems and components of these reactors based on their intended safety functions in order to define the applicable set of codes, standards and technical requirements. (author)

  8. Structural integrity of water reactor pressure boundary components

    International Nuclear Information System (INIS)

    Loss, F.J.

    1977-01-01

    The dynamic fracture toughness was determined as a function of temperature for three-point bend specimens of A533-B, A508-2, and A302-B steels. Crack propagation rates at 288 0 C in a water reactor environment were determined for A533-B and A508-2. Radiation-induced degradation of notch toughness of reactor steels and welds was explored. The ''warm prestress'' occurring in a flawed reactor vessel following a LOCA and operation of ECCS was studied. 25 figures

  9. Structural integrity and management of aging in internal components of BWR reactors

    International Nuclear Information System (INIS)

    Arganis J, C.R.

    2004-01-01

    Presently work the bases to apply structural integrity and the handling of the aging of internal components of the pressure vessel of boiling water reactors of water are revised and is carried out an example of structural integrity in the horizontal welding H4 of the encircling one of the core of a reactor, taking data reported in the literature. It is also revised what is required to carry out the handling program or conduct of the aging (AMP). (Author)

  10. Safety classification of systems, structures, and components for pool-type research reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Tae Ryong [KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of)

    2016-08-15

    Structures, systems, and components (SSCs) important to safety of nuclear facilities shall be designed, fabricated, erected, and tested to quality standards commensurate with the importance of the safety functions. Although SSC classification guidelines for nuclear power plants have been well established and applied, those for research reactors have been only recently established by the International Atomic Energy Agency (IAEA). Korea has operated a pool-type research reactor (the High Flux Advanced Neutron Application Reactor) and has recently exported another pool-type reactor (Jordan Research and Training Reactor), which is being built in Jordan. Korea also has a plan to build one more pool-type reactor, the Kijang Research Reactor, in Kijang, Busan. The safety classification of SSCs for pool-type research reactors is proposed in this paper based on the IAEA methodology. The proposal recommends that the SSCs of pool-type research reactors be categorized and classified on basis of their safety functions and safety significance. Because the SSCs in pool-type research reactors are not the pressure-retaining components, codes and standards for design of the SSCs following the safety classification can be selected in a graded approach.

  11. Development of Structural Core Components for Breeder Reactors

    International Nuclear Information System (INIS)

    Saibaba, N.

    2013-01-01

    Core structural materials: • The desire is to have only fuel in the core, structural material form 25% of the total core: – To support and to retain the fuel in position; – Provide necessary ducts to make coolant flow through & transfer/remove heat. • For 500 MWe FBR with Oxide fuel (Peak Linear Power 450 W/cm), total fuel pins required in the core are of the order 39277 pins (both inner & outer core Fuel SA); • Considering 217 pins/Fuel SA there are 181 Fuel SA wrapper tubes • These structural materials see hostile core with max temperature and neutron flux

  12. Assessment of radiation fields from neutron irradiated structural components of the 40 MW research reactor CIRUS

    International Nuclear Information System (INIS)

    Sankaranarayanan, S.; Sharma, S.K.

    1993-01-01

    The paper summarizes the results of an assessment of the radiation fields from the long-lived neutron activation products (including the decay chain products) in the various structural components of the CIRUS reactor. Special attention is given for the analysis of neutron activation of impurity elements present in the materials of the structure. 16 refs, 4 figs, 4 tabs

  13. RCC-MRx: Design and construction rules for mechanical components in high-temperature structures, experimental reactors and fusion reactors

    International Nuclear Information System (INIS)

    2015-01-01

    The RCC-MRx code was developed for sodium-cooled fast reactors (SFR), research reactors (RR) and fusion reactors (FR-ITER). It provides the rules for designing and building mechanical components involved in areas subject to significant creep and/or significant irradiation. In particular, it incorporates an extensive range of materials (aluminum and zirconium alloys in response to the need for transparency to neutrons), sizing rules for thin shells and box structures, and new modern welding processes: electron beam, laser beam, diffusion and brazing. The RCC-MR code was used to design and build the prototype Fast Breeder Reactor (PFBR) developed by IGCAR in India and the ITER Vacuum Vessel. The RCC-Mx code is being used in the current construction of the RJH experimental reactor (Jules Horowitz reactor). The RCC-MRx code is serving as a reference for the design of the ASTRID project (Advanced Sodium Technological Reactor for Industrial Demonstration), for the design of the primary circuit in MYRRHA (Multi-purpose hybrid Research Reactor for High-tech Applications) and the design of the target station of the ESS project (European Spallation Source). Contents of the 2015 edition of the RCC-MRx code: Section I General provisions; Section II Additional requirements and special provisions; Section III Rules for nuclear installation mechanical components: Volume I: Design and construction rules: Volume A (RA): General provisions and entrance keys, Volume B (RB): Class 1 components and supports, Volume C (RC): Class 2 components and supports, Volume D (RD): Class 3 components and supports, Volume K (RK): Examination, handling or drive mechanisms, Volume L (RL): Irradiation devices, Volume Z (Ai): Technical appendices; Volume II: Materials; Volume III: Examinations methods; Volume IV: Welding; Volume V: Manufacturing operations; Volume VI: Probationary phase rules

  14. Thermal-structural response of EBR-II major components under reactor operational transients

    International Nuclear Information System (INIS)

    Chang, L.K.; Lee, M.J.

    1983-01-01

    Until recently, the LMFBR safety research has been focused primarily on severe but highly unlikely accident, such as hypothetical-core-disruptive accidents (HCDA's), and not enough attention has been given to accident prevention, which is less severe but more likely sequence. The objective of the EBR-II operational reliability testing (ORT) is to demonstrate that the reactor can be designed and operated to prevent accident. A series of mild duty cycles and overpower transients were designed for accident prevention tests. An assessment of the EBR-II major plant components has been performed to assure structural integrity of the reactor plant for the ORT program. In this paper, the thermal-structural response and structural evaluation of the reactor vessel, the reactor-vessel cover, the intermediate heat exchanger (IHX) and the superheater are presented

  15. State of the art seismic analysis for CANDU reactor structure components using condensation method

    Energy Technology Data Exchange (ETDEWEB)

    Soliman, S A; Ibraham, A M; Hodgson, S [Atomic Energy of Canada Ltd., Saskatoon, SK (Canada)

    1996-12-31

    The reactor structure assembly seismic analysis is a relatively complex process because of the intricate geometry with many different discontinuities, and due to the hydraulic attached mass which follows the structure during its vibration. In order to simulate reasonably accurate behaviour of the reactor structure assembly, detailed finite element models are generated and used for both modal and stress analysis. Guyan reduction condensation method was used in the analysis. The attached mass, which includes the fluid mass contained in the components plus the added mass which accounts for the inertia of the surrounding fluid entrained by the accelerating structure immersed in the fluid, was calculated and attached to the vibrating structures. The masses of the attached components, supported partly or totally by the assembly which includes piping, reactivity control units, end fittings, etc. are also considered in the analysis. (author). 4 refs., 6 tabs., 4 figs.

  16. Reactor component automatic grapple

    International Nuclear Information System (INIS)

    Greenaway, P.R.

    1982-01-01

    A grapple for handling nuclear reactor components in a medium such as liquid sodium which, upon proper seating and alignment of the grapple with the component as sensed by a mechanical logic integral to the grapple, automatically seizes the component. The mechanical logic system also precludes seizure in the absence of proper seating and alignment. (author)

  17. Three-dimensional fluid-structure interaction dynamics of a pool-reactor in-tank component

    International Nuclear Information System (INIS)

    Kulak, R.F.

    1979-01-01

    The safety evaluation of reactor-components often involves the analysis of various types of fluid/structural components interacting in three-dimensional space. For example, in the design of a pool-type reactor several vital in-tank components such as the primary pumps and the intermediate heat exchangers are contained within the primary tank. Typically, these components are suspended from the deck structure and largely submersed in the sodium pool. Because of this positioning these components are vulnerable to structural damage due to pressure wave propagation in the tank during a CDA. In order to assess the structural integrity of these components it is necessary to perform a dynamic analysis in three-dimensional space which accounts for the fluid-structure coupling. A model is developed which has many of the salient features of this fluid-structural component system

  18. Irradiation effects on the mechanical properties of aluminium and the structural integrity of aluminium reactor components

    International Nuclear Information System (INIS)

    Harrison, R.P.; McDonald, N.R.; Mitchell, D.R.G.; Hellier, A.K.; Stathers, P.A.; Carr, D.G.; Ripley, M.I.

    2000-01-01

    The results of micro-structural and mechanical property studies on aluminum after being exposed to large fluences of neutrons are presented. These property changes are of importance in determining the structural integrity of the Australian HIFAR reactor aluminium tank, which in turn determines the lifetime of the reactor. (author)

  19. Structural integrity and management of aging in internal components of BWR reactors; Integridad estructural y manejo del envejecimiento en componentes internos de reactores BWR

    Energy Technology Data Exchange (ETDEWEB)

    Arganis J, C.R. [Instituto Nacional de Investigaciones Nucleares, Km 36.5 Carretera Mexico, Toluca Salazar Edo. de Mexico (Mexico)]. E-mail: craj@nuclear.inin.mx

    2004-07-01

    Presently work the bases to apply structural integrity and the handling of the aging of internal components of the pressure vessel of boiling water reactors of water are revised and is carried out an example of structural integrity in the horizontal welding H4 of the encircling one of the core of a reactor, taking data reported in the literature. It is also revised what is required to carry out the handling program or conduct of the aging (AMP). (Author)

  20. Fundamental Understanding of Crack Growth in Structural Components of Generation IV Supercritical Light Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Iouri I. Balachov; Takao Kobayashi; Francis Tanzella; Indira Jayaweera; Palitha Jayaweera; Petri Kinnunen; Martin Bojinov; Timo Saario

    2004-11-17

    This work contributes to the design of safe and economical Generation-IV Super-Critical Water Reactors (SCWRs) by providing a basis for selecting structural materials to ensure the functionality of in-vessel components during the entire service life. During the second year of the project, we completed electrochemical characterization of the oxide film properties and investigation of crack initiation and propagation for candidate structural materials steels under supercritical conditions. We ranked candidate alloys against their susceptibility to environmentally assisted degradation based on the in situ data measure with an SRI-designed controlled distance electrochemistry (CDE) arrangement. A correlation between measurable oxide film properties and susceptibility of austenitic steels to environmentally assisted degradation was observed experimentally. One of the major practical results of the present work is the experimentally proven ability of the economical CDE technique to supply in situ data for ranking candidate structural materials for Generation-IV SCRs. A potential use of the CDE arrangement developed ar SRI for building in situ sensors monitoring water chemistry in the heat transport circuit of Generation-IV SCWRs was evaluated and proved to be feasible.

  1. An explication of the Graphite Structural Design Code of core components for the High Temperature Engineering Test Reactor

    International Nuclear Information System (INIS)

    Iyoku, Tatsuo; Ishihara, Masahiro; Toyota, Junji; Shiozawa, Shusaku

    1991-05-01

    The integrity evaluation of the core graphite components for the High Temperature Engineering Test Reactor (HTTR) will be carried out based upon the Graphite Structural Design Code for core components. In the application of this design code, it is necessary to make clear the basic concept to evaluate the integrity of core components of HTTR. Therefore, considering the detailed design of core graphite structures such as fuel graphite blocks, etc. of HTTR, this report explicates the design code in detail about the concepts of stress and fatigue limits, integrity evaluation method of oxidized graphite components and thermal irradiation stress analysis method etc. (author)

  2. Maintenance of Structures, Systems and Components of the RSG-GAS Reactor as an Implementation of BAPETEN Regulation No. 5 Year 2011

    International Nuclear Information System (INIS)

    Aep Saepudin Catur; Dede Solehudin Fauzi; Djunaidi

    2012-01-01

    Maintenance activities for structures, systems and components of the reactor, consider prerequisites for the operation of the non power reactor. This activity is intended to ensure that the structures, systems and components function properly. Implementation of reactor maintenance carried out starting from programs establishment, scheduling, maintenance accomplishment and maintenance report. This paper will describe the implementation of reactor maintenance non power as required by BAPETEN regulation no.5, year 2011. By understanding correctly of this regulation, it is expected that maintenance activity of structures, systems and components of the reactor can be successfully performed. The RSG-GAS reactor has implemented various types of reactor maintenance based on BAPETEN regulation no.5 year 2011 properly. As a result failure of the structures, systems and components of the reactor can be minimized then they can be kept reliable. (author)

  3. Structural integrity of water reactor pressure boundary components. Progress report ending 29 February 1976

    International Nuclear Information System (INIS)

    Loss, F.J.

    1976-01-01

    The report describes progress in the following areas: (a) fatigue crack propagation in reactor pressure vessel steels in an air environment, (b) dynamic fracture toughness of 1-in. (25-mm) and precracked Charpy-V bend specimens under impact loading, (c) postirradiation notch ductility and properties recovery in reactor vessel steels, (d) factors contributing to variable resistance of structural steels to radiation embrittlement, and (e) the initial program plan to investigate the phenomena of warm prestress and plastic net ligament in support of thermal shock studies

  4. XHM-1 alloy as a promising structural material for water-cooled fusion reactor components

    International Nuclear Information System (INIS)

    Solonin, M.I.; Alekseev, A.B.; Kazennov, Yu.I.; Khramtsov, V.F.; Kondrat'ev, V.P.; Krasina, T.A.; Rechitsky, V.N.; Stepankov, V.N.; Votinov, S.N.

    1996-01-01

    Experience gained in utilizing austenitic stainless steel components in water-cooled power reactors indicates that the main cause of their failure is the steel's propensity for corrosion cracking. In search of a material immune to this type of corrosion, different types of austenitic steels and chromium-nickel alloys were investigated and tested at VNIINM. This paper presents the results of studying physical and mechanical properties, irradiation and corrosion resistance in a water coolant at <350 C of the alloy XHM-1 as compared with austenitic stainless steels 00Cr16Ni15Mo3Nb, 00Cr20Ni25Nb and alloy 00Cr20Ni40Mo5Nb. Analysis of the results shows that, as distinct from the stainless steels studied, the XHM-1 alloy is completely immune to corrosion cracking (CC). Not a single induced damage was encountered within 50 to 350 C in water containing different amounts of chlorides and oxygen under tensile stresses up to the yield strength of the material. One more distinctive feature of the alloy compared to steels is that no change in the strength or total elongation is encountered in the alloy specimens irradiated to 32 dpa at 350 C. The XHM-1 alloy has adequate fabricability and high weldability characteristics. As far as its properties are concerned, the XHM-1 alloy is very promising as a material for water-cooled fusion reactor components. (orig.)

  5. Travelling cranes for heavy reactor component handling

    International Nuclear Information System (INIS)

    Champeil, M.

    1977-01-01

    Structure and operating machinery of two travelling cranes (600 t and 450 t) used in the Framatome factory for handling heavy reactor components are described. When coupled, these cranes can lift loads up to 1000 t [fr

  6. Reactor core structure

    International Nuclear Information System (INIS)

    Higashinakagawa, Emiko; Sato, Kanemitsu.

    1992-01-01

    Taking notice on the fact that Fe based alloys and Ni based alloys are corrosion resistant in a special atmosphere of a nuclear reactor, Fe or Ni based alloys are applied to reactor core structural components such as fuel cladding tubes, fuel channels, spacers, etc. On the other hand, the neutron absorption cross section of zirconium is 0.18 barn while that of iron is 2.52 barn and that of nickel is 4.6 barn, which amounts to 14 to 25 times compared with that of zirconium. Accordingly, if the reactor core structural components are constituted by the Fe or Ni based alloys, neutron economy is lowered. Since it is desirable that neutrons contribute to uranium fission with least absorption to the reactor core structural components, the reactor core structural components are constituted with the Fe or Ni based alloys of good corrosion resistance only at a portion in contact with reactor water, that is, at a surface portion, while the main body is constituted with zircalloy in the present invention. Accordingly, corrosion resistnace can be kept while keeping small neutron absorption cross section. (T.M.)

  7. Application of two component biodegradable carriers in a particle-fixed biofilm airlift suspension reactor: development and structure of biofilms.

    Science.gov (United States)

    Hille, Andrea; He, Mei; Ochmann, Clemens; Neu, Thomas R; Horn, Harald

    2009-01-01

    Two component biodegradable carriers for biofilm airlift suspension (BAS) reactors were investigated with respect to development of biofilm structure and oxygen transport inside the biofilm. The carriers were composed of PHB (polyhydroxybutyrate), which is easily degradable and PCL (caprolactone), which is less easily degradable by heterotrophic microorganisms. Cryosectioning combined with classical light microscopy and CLSM was used to identify the surface structure of the carrier material over a period of 250 days of biofilm cultivation in an airlift reactor. Pores of 50 to several hundred micrometers depth are formed due to the preferred degradation of PHB. Furthermore, microelectrode studies show the transport mechanism for different types of biofilm structures, which were generated under different substrate conditions. At high loading rates, the growth of a rather loosely structured biofilm with high penetration depths of oxygen was found. Strong changes of substrate concentration during fed-batch mode operation of the reactor enhance the growth of filamentous biofilms on the carriers. Mass transport in the outer regions of such biofilms was mainly driven by advection.

  8. Research and developments on nondestructive testing in fabrications of fast breeder reactor structural components in Japan

    International Nuclear Information System (INIS)

    Ebata, M.; Ooka, K.; Miyoshi, S.; Senda, T.

    1985-01-01

    Research and developments (R and D) have been conducted on the nondestructive testing techniques necessary for the construction of fast breeder reactor (FBR). Radiographic tests have been made on tube-tube plate welds and small-diameter tube welds, etc. Ultrasonic tests have been conducted on austenitic stainless steel welds. In the penetrant tests and magnetic particle tests, the investigations have been performed on the effects of various test factors on the test results

  9. A hierarchically structured identification- and classification method for vibration control of reactor components

    International Nuclear Information System (INIS)

    Saedtler, E.

    1981-01-01

    The method for controlling the vibrating behaviour of primary circuit components or for a general systems control is a combination of methods of the statistic systems theory, optimum filter theory, statistic decision theory and of the pattern recognition method. It is appropriate for automatic control of complex systems and stochastic events. (DG) [de

  10. Chemical decontamination of reactor components

    International Nuclear Information System (INIS)

    Riess, R.; Berthold, H.O.

    1977-08-01

    A solution for the decontamination of reactor components of the primary system was developed. This solution is a modification of the APAC- (Alkaline Permanganate Ammonium Citrate) system described in the literature. The most important advantage of the present solution over the APAC-method is that it does not induce any selective corrosion attack on materials like stainless steel (austenitic), Inconel 600 and Incoloy 800. (orig.) [de

  11. Manufacture of components for Canadian reactor programs

    International Nuclear Information System (INIS)

    Perry, L.P.

    Design features, especially those relating to calandrias, are pointed out for many CANDU-type reactors and the Taiwan research reactor. The special requirements shouldered by the Canadian suppliers of heavy reactor components are analyzed. (E.C.B.)

  12. Nonlinear seismic analysis of a reactor structure with impact between core components

    International Nuclear Information System (INIS)

    Hill, R.G.

    1975-01-01

    The seismic analysis of the FFTF-PIOTA (Fast Flux Test Facility-Postirradiation Open Test Assembly), subjected to a horizontal DBE (Design Base Earthquake) is presented. The PIOTA is the first in a set of open test assemblies to be designed for the FFTF. Employing the direct method of transient analysis, the governing differential equations describing the motion of the system are set up directly and are implicitly integrated numerically in time. A simple lumped-mass beam model of the FFTF which includes small clearances between core components is used as a ''driver'' for a fine mesh model of the PIOTA. The nonlinear forces due to the impact of the core components and their effect on the PIOTA are computed. 6 references

  13. Hierarchically structured identification and classification method for vibrational monitoring of reactor components

    International Nuclear Information System (INIS)

    Saedtler, E.

    1981-01-01

    The dissertation discusses: 1. Approximative filter algorithms for identification of systems and hierarchical structures. 2. Adaptive statistical pattern recognition and classification. 3. Parameter selection, extraction, and modelling for an automatic control system. 4. Design of a decision tree and an adaptive diagnostic system. (orig./RW) [de

  14. Implicit three-dimensional finite-element formulation for the nonlinear structural response of reactor components

    International Nuclear Information System (INIS)

    Kulak, R.F.; Belytschko, T.B.

    1975-09-01

    The formulation of a finite-element procedure for the implicit transient and static analysis of plate/shell type structures in three-dimensional space is described. The triangular plate/shell element can sustain both membrane and bending stresses. Both geometric and material nonlinearities can be treated, and an elastic-plastic material law has been incorporated. The formulation permits the element to undergo arbitrarily large rotations and translations; but, in its present form it is restricted to small strains. The discretized equations of motion are obtained by a stiffness method. An implicit integration algorithm based on trapezoidal integration formulas is used to integrate the discretized equations of motion in time. To ensure numerical stability, an iterative solution procedure with equilibrium checks is used

  15. Use of SiCf/SiC ceramic composites as structure material of a fusion reactor toroid internal components

    International Nuclear Information System (INIS)

    Aiello, G.

    2001-01-01

    The use of low neutron-induced activation structural materials seems necessary in order to improve safety in future fusion power reactors. Among them, SiC f /SiC composites appear as a very promising solution because of their low activation characteristics coupled with excellent mechanical properties at high temperatures. With the main objective of evaluating the limit of present-day composites, a tritium breeding blanket using SiC f /SiC as structural material (the TAURO blanket) has been developed in the last years by the Commissariat a l'Energie Atomique (CEA). The purpose of this thesis was to modify the available design tools (computer codes, design criteria), normally used for the analyses of metallic structures, in order to better take into account the mechanical behaviour of SiC f /SiC. Alter a preliminary improvement of the calculation methods, two main topics of study could be identified: the modelling of the mechanical behaviour of the composite and the assessment of appropriate design criteria. The different behavioural models available in literature were analysed in order to find the one that was the best suited to the specific problems met in the field of fusion power. The selected model was then implemented in the finite elements code CASTEM 2000 used within the CEA for the thermo-mechanical analyses of the TAURO blanket. For the design of the blanket, we proposed a new resistance criterion whose main advantage, with respect to the other examined, lies in the easiness of identification. The suggested solutions were then applied in the design studies of the TAURO blanket. We then could show that the use of appropriate calculation methodologies is necessary in order to achieve a correct design of the blanket and a more realistic estimate of the limits of present day composites. The obtained results can also be extended to all nuclear components making use of SiC f /SiC structures. (author) [fr

  16. Tritium in fusion reactor components

    International Nuclear Information System (INIS)

    Watson, J.S.; Fisher, P.W.; Talbot, J.B.

    1980-01-01

    When tritium is used in a fusion energy experiment or reactor, several implications affect and usually restrict the design and operation of the system and involve questions of containment, inventory, and radiation damage. Containment is expected to be particularly important both for high-temperature components and for those components that are prone to require frequent maintenance. Inventory is currently of major significance in cases where safety and environmental considerations limit the experiments to very low levels of tritium. Fewer inventory restrictions are expected as fusion experiments are placed in more-remote locations and as the fusion community gains experience with the use of tritium. However, the advent of power-producing experiments with high-duty cycle will again lead to serious difficulties based principally on tritium availability; cyclic operations with significant regeneration times are the principal problems

  17. Feasibility analysis of modified AL-6XN steel for structure component application in supercritical water-cooled reactor

    Institute of Scientific and Technical Information of China (English)

    Xinggang LI; Qingzhi YAN; Rong MA; Haoqiang WANG; Changchun GE

    2009-01-01

    Modified AL-6XN austenite steel was patterned after AL-6XN superaustenitic stainless steel by introducing microalloy elements such as zirconium and titanium in order to adapt to recrystallizing thermo-mechanical treatment and further improve crevice corrosion resistance. Modified AL-6XN exhibited comparable tensile strength, and superior plasticity and impact toughness to commercial AL-6XN steel. The effects of aging behavior on corrosion resistance and impact toughness were measured to evaluate the qualification of modified AL-6XN steel as an in-core component and cladding material in a supercritical water-cooled reactor. Attention should be paid to degradation in corrosion resistance and impact toughness after aging for 50 hours when modified AL-6XN steel is considered as one of the candidate materials for in-core components and cladding tubes in supercritical water-cooled reactors.

  18. Lubrication of nuclear reactor components

    International Nuclear Information System (INIS)

    Wild, E.; Mack, K.J.

    1978-01-01

    Safe operation of liquid metal cooled nuclear reactors requires a knowledge of the tribological behaviour of contacting components at high temperatures with slow relative movement at high frictional loads in a chemically aggressive environment. Experiments have been performed on various material combinations in liquid sodium and argon. Because of the small sliding movements, hydrodynamic lubrication is not expected and thus surface finish is an important factor. Tests have been performed on brushed, ground and lapped surfaces. Among the material combinations tested a CrC-coating on a 1.4961 stainless steel substrate performed well. Friction coefficients of 0.35-0.5 in argon and 0.1-1.2 in liquid sodium were recorded. (author)

  19. Component failure data base of TRIGA reactors

    International Nuclear Information System (INIS)

    Djuricic, M.

    2004-10-01

    This compilation provides failure data such as first criticality, component type description (reactor component, population, cumulative calendar time, cumulative operating time, demands, failure mode, failures, failure rate, failure probability) and specific information on each type of component of TRIGA Mark-II reactors in Austria, Bangladesh, Germany, Finland, Indonesia, Italy, Indonesia, Slovenia and Romania. (nevyjel)

  20. Hydrodynamic impact of reactor components - a review

    International Nuclear Information System (INIS)

    Krajcinovic, D.

    1977-01-01

    A variety of components belonging to a nuclear reactor are by virtue of their design exposed to a mass of fluid which is either in motion or can be set into motion under certain conditions. While the reactor is in its operational mode, the excitations of the structure by the fluid are generally of moderate intensities. In the case of a well designed component, these pressure fluctuations should not cause the failure of the structure. Problems of this type, generally known as vibrations of structures immersed into fluid (under either periodic or random excitations) have been studied in the past rather extensively. In an upset or emergency condition, a pressure pulse is usually generated and propagated through the fluid. While this hypothetical event is an occurrence of low probability the associated pressures are, as a rule, of intensities sufficiently large to cause extensive damage or even the failure of the component. This type of transient interaction problem is much less studied and the aim of this review is to offer a brief discussion of some of the more interesting results. (Auth.)

  1. Numerical analysis of magnetoelastic coupled buckling of fusion reactor components

    International Nuclear Information System (INIS)

    Demachi, K.; Yoshida, Y.; Miya, K.

    1994-01-01

    For a tokamak fusion reactor, it is one of the most important subjects to establish the structural design in which its components can stand for strong magnetic force induced by plasma disruption. A number of magnetostructural analysis of the fusion reactor components were done recently. However, in these researches the structural behavior was calculated based on the small deformation theory where the nonlinearity was neglected. But it is known that some kinds of structures easily exceed the geometrical nonlinearity. In this paper, the deflection and the magnetoelastic buckling load of fusion reactor components during plasma disruption were calculated

  2. TIMTEM - a digital program for the calculation of two-dimensional, non-linear temperature fields of reactor components of complex structure taking into account inhomogeneity and anisotropy

    International Nuclear Information System (INIS)

    Simon-Weidner, J.

    1975-05-01

    The digital program TIMTEM calculates twodimensional, nonlinear temperature fields of reactor components of complex structure; inhomogeneity and anisotropy are taken into account. Systems consisting of different materials and therefore having different temperature- and/or time-dependent material characteristics are allowed. Various local, time- and/or temperature-dependent boundary conditions can be considered, too, which may be locally different from each other or can be interconnected. (orig.) [de

  3. Proposed and existing passive and inherent safety-related structures, systems, and components (building blocks) for advanced light-water reactors

    International Nuclear Information System (INIS)

    Forsberg, C.W.; Moses, D.L.; Lewis, E.B.; Gibson, R.; Pearson, R.; Reich, W.J.; Murphy, G.A.; Staunton, R.H.; Kohn, W.E.

    1989-10-01

    A nuclear power plant is composed of many structures, systems, and components (SSCs). Examples include emergency core cooling systems, feedwater systems, and electrical systems. The design of a reactor consists of combining various SSCs (building blocks) into an integrated plant design. A new reactor design is the result of combining old SSCs in new ways or use of new SSCs. This report identifies, describes, and characterizes SSCs with passive and inherent features that can be used to assure safety in light-water reactors. Existing, proposed, and speculative technologies are described. The following approaches were used to identify the technologies: world technical literature searches, world patent searches, and discussions with universities, national laboratories and industrial vendors. 214 refs., 105 figs., 26 tabs

  4. Proposed and existing passive and inherent safety-related structures, systems, and components (building blocks) for advanced light-water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Forsberg, C.W.; Moses, D.L.; Lewis, E.B.; Gibson, R.; Pearson, R.; Reich, W.J.; Murphy, G.A.; Staunton, R.H.; Kohn, W.E.

    1989-10-01

    A nuclear power plant is composed of many structures, systems, and components (SSCs). Examples include emergency core cooling systems, feedwater systems, and electrical systems. The design of a reactor consists of combining various SSCs (building blocks) into an integrated plant design. A new reactor design is the result of combining old SSCs in new ways or use of new SSCs. This report identifies, describes, and characterizes SSCs with passive and inherent features that can be used to assure safety in light-water reactors. Existing, proposed, and speculative technologies are described. The following approaches were used to identify the technologies: world technical literature searches, world patent searches, and discussions with universities, national laboratories and industrial vendors. 214 refs., 105 figs., 26 tabs.

  5. Energy deposition in STARFIRE reactor components

    International Nuclear Information System (INIS)

    Gohar, Y.; Brooks, J.N.

    1985-04-01

    The energy deposition in the STARFIRE commercial tokamak reactor was calculated based on detailed models for the different reactor components. The heat deposition and the 14 MeV neutron flux poloidal distributions in the first wall were obtained. The poloidal surface heat load distribution in the first wall was calculated from the plasma radiation. The Monte Carlo method was used for the calculation to allow an accurate modeling for the reactor geometry

  6. International ENEA/ISMES/ENS specialist meeting on 'On-site experimental verification of the seismic behaviour of nuclear reactor structures and components'. Proceedings

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1988-07-01

    The seismic verification of nuclear plants is a subject of increasing interest in all the industrial countries, with respect to both the safety aspects and the impact of the seismic event on the design and the costs of a nuclear reactor. This topic is especially of great interest for a country like Italy, whose territory is unfortunately characterized by non - negligible seismicity: we remember, not too many years ago, the catastrophic earthquakes of Frioul and Irpinia, that caused thousands of dead people. The meeting aimed at establishing the state-of-the-art on on-site testing of nuclear reactors structures and components, with particular attention to experiences and research programmes concerning: methodologies of on-site tests and interpretation of the experimental data; seismic monitoring systems, recorded data, their use and interpretation; calibration and validation of numerical analyses. Six technical sessions were held, during which 23 high papers were presented and discussed, and six panel discussions were held (the importance of discussion was emphasized in the meeting). The technical contributions consisted of: an introduction paper, summarizing the seismic studies performed in Italy for PEC reactor and explaining the reasons why on-site tests had been performed on this reactor; 6 invited lectures, one for each of the countries that are more deeply involved in seismic analysis, providing the state-of-the-art on the topics of interest for the meeting; 16 contributed papers dealing with more specific technical items, related to the various countries and international organizations.

  7. International ENEA/ISMES/ENS specialist meeting on 'On-site experimental verification of the seismic behaviour of nuclear reactor structures and components'. Proceedings

    International Nuclear Information System (INIS)

    1988-01-01

    The seismic verification of nuclear plants is a subject of increasing interest in all the industrial countries, with respect to both the safety aspects and the impact of the seismic event on the design and the costs of a nuclear reactor. This topic is especially of great interest for a country like Italy, whose territory is unfortunately characterized by non - negligible seismicity: we remember, not too many years ago, the catastrophic earthquakes of Frioul and Irpinia, that caused thousands of dead people. The meeting aimed at establishing the state-of-the-art on on-site testing of nuclear reactors structures and components, with particular attention to experiences and research programmes concerning: methodologies of on-site tests and interpretation of the experimental data; seismic monitoring systems, recorded data, their use and interpretation; calibration and validation of numerical analyses. Six technical sessions were held, during which 23 high papers were presented and discussed, and six panel discussions were held (the importance of discussion was emphasized in the meeting). The technical contributions consisted of: an introduction paper, summarizing the seismic studies performed in Italy for PEC reactor and explaining the reasons why on-site tests had been performed on this reactor; 6 invited lectures, one for each of the countries that are more deeply involved in seismic analysis, providing the state-of-the-art on the topics of interest for the meeting; 16 contributed papers dealing with more specific technical items, related to the various countries and international organizations

  8. Component failures that lead to reactor scrams

    International Nuclear Information System (INIS)

    Burns, E.T.; Wilson, R.J.; Lim, E.Y.

    1980-04-01

    This report summarizes the operating experience scram data compiled from 35 operating US light water reactors (LWRs) to identify the principal components/systems related to reactor scrams. The data base utilized to identify the scram causes is developed from a EPRI-utility sponsored survey conducted by SAI coupled with recent data from the USNRC Gray Books. The reactor population considered in this evaluation is limited to 23 PWRs and 12 BWRs because of the limited scope of the program. The population includes all the US NSSS vendors. It is judged that this population accurately characterizes the component-related scrams in LWRs over the first 10 years of plant operation

  9. Cooling system for auxiliary reactor component

    International Nuclear Information System (INIS)

    Fujihira, Tomoko.

    1991-01-01

    A cooling system for auxiliary reactor components comprises three systems, that is, two systems of reactor component cooling water systems (RCCW systems) and a high pressure component cooling water system (HPCCW system). Connecting pipelines having partition valves are intervened each in a cooling water supply pipeline to an emmergency component of each of the RCCW systems, a cooling water return pipeline from the emmergency component of each of the RCCW systems, a cooling water supply pipeline to each of the emmergency components of one of the RCCW system and the HPCCW system and a cooling water return pipeline from each of the emmergency components of one of the RCCW system and the HPCCW system. With such constitution, cooling water can be supplied also to the emmergency components in the stand-by system upon periodical inspection or ISI, thereby enabling to improve the backup performance of the emmergency cooling system. (I.N.)

  10. Structural mechanics and reactor safety

    International Nuclear Information System (INIS)

    Brandes, K.

    1983-01-01

    Operational safety and reliability of nuclear power plants widely depend on the mechanical behaviour of their structural components and their resistance to the various and complex influences. Durability and consistency of structural components are determined by the kind of strain - during the life - and by environmental conditions. The Conferences on Structural Mechanics in Reactor Technology (SMiRT) are dedicated to the discussion of such questions. The 7th of these Conferences taking place in 2-year increments was held in Chicago in August 1983. The number of contributions again increased, the number of participants slightly decreased. There are some trends in this field worth mentioning, in particular the fact that experience from design and operation of nuclear power plants now available is more and more made use of, and that more and more attention is given the problems of fusion reactors. (orig./HP) [de

  11. Mechanical development for reliable reactor components

    International Nuclear Information System (INIS)

    Ross-Ross, P.A.; Metcalfe, R.

    1983-09-01

    The CANDU reactor has achieved worldwide distinction because of its reliable performance. To achieve this, special attention was given to the reliability and maintainability of components in the heavy water circuits. Development programs were initiated early in the history of the CANDU reactor to improve the effectiveness of pump seals, valves, and static seals because of unacceptable performance of the commercial equipment then available. As a result, pump seals with a five year life now appear achievable, and valves and static seals are no longer a significant concern in CANDU reactors. Increasing effort is being given remotely operated tools and fabrication systems for radioactive environments

  12. IAEA Coordinated Research Project on the Establishment of a Material Properties Database for Irradiated Core Structural Components for Continued Safe Operation and Lifetime Extension of Ageing Research Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Borio Di Tigliole, A.; Schaaf, Van Der; Barnea, Y.; Bradley, E.; Morris, C.; Rao, D. V. H. [Research Reactor Section, Vianna (Australia); Shokr, A. [Research Reactor Safety Section, Vienna (Australia); Zeman, A. [International Atomic Energy Agency, Vienna (Australia)

    2013-07-01

    Today more than 50% of operating Research Reactors (RRs) are over 45 years old. Thus, ageing management is one of the most important issues to face in order to ensure availability (including life extension), reliability and safe operation of these facilities for the future. Management of the ageing process requires, amongst others, the predictions for the behavior of structural materials of primary components subjected to irradiation such as reactor vessel and core support structures, many of which are extremely difficult or impossible to replace. In fact, age-related material degradation mechanisms resulted in high profile, unplanned and lengthy shutdowns and unique regulatory processes of relicensing the facilities in recent years. These could likely have been prevented by utilizing available data for the implementation of appropriate maintenance and surveillance programmes. This IAEA Coordinated Research Project (CRP) will provide an international forum to establish a material properties Database for irradiated core structural materials and components. It is expected that this Database will be used by research reactor operators and regulators to help predict ageing related degradation. This would be useful to minimize unpredicted outages due to ageing processes of primary components and to mitigate lengthy and costly shutdowns. The Database will be a compilation of data from RRs operators' inputs, comprehensive literature reviews and experimental data from RRs. Moreover, the CRP will specify further activities needed to be addressed in order to bridge the gaps in the new created Database, for potential follow-on activities. As per today, 13 Member States (MS) confirmed their agreement to contribute to the development of the Database, covering a wide number of materials and properties. The present publication incorporates two parts: the first part includes details on the pre-CRP Questionnaire, including the conclusions drawn from the answers received from

  13. IAEA Coordinated Research Project on the Establishment of a Material Properties Database for Irradiated Core Structural Components for Continued Safe Operation and Lifetime Extension of Ageing Research Reactors

    International Nuclear Information System (INIS)

    Borio Di Tigliole, A.; Schaaf, Van Der; Barnea, Y.; Bradley, E.; Morris, C.; Rao, D. V. H.; Shokr, A.; Zeman, A.

    2013-01-01

    Today more than 50% of operating Research Reactors (RRs) are over 45 years old. Thus, ageing management is one of the most important issues to face in order to ensure availability (including life extension), reliability and safe operation of these facilities for the future. Management of the ageing process requires, amongst others, the predictions for the behavior of structural materials of primary components subjected to irradiation such as reactor vessel and core support structures, many of which are extremely difficult or impossible to replace. In fact, age-related material degradation mechanisms resulted in high profile, unplanned and lengthy shutdowns and unique regulatory processes of relicensing the facilities in recent years. These could likely have been prevented by utilizing available data for the implementation of appropriate maintenance and surveillance programmes. This IAEA Coordinated Research Project (CRP) will provide an international forum to establish a material properties Database for irradiated core structural materials and components. It is expected that this Database will be used by research reactor operators and regulators to help predict ageing related degradation. This would be useful to minimize unpredicted outages due to ageing processes of primary components and to mitigate lengthy and costly shutdowns. The Database will be a compilation of data from RRs operators' inputs, comprehensive literature reviews and experimental data from RRs. Moreover, the CRP will specify further activities needed to be addressed in order to bridge the gaps in the new created Database, for potential follow-on activities. As per today, 13 Member States (MS) confirmed their agreement to contribute to the development of the Database, covering a wide number of materials and properties. The present publication incorporates two parts: the first part includes details on the pre-CRP Questionnaire, including the conclusions drawn from the answers received from the MS

  14. External vibrations measurement of reactor components

    Energy Technology Data Exchange (ETDEWEB)

    Rogers, S A [Nuclear Electric plc, Barnwood (United Kingdom); Sugden, J [Magnox Electric, Berkeley (United Kingdom)

    1997-12-31

    The paper outlines the use of External Vibration Monitoring for remote vibration assessment of internal reactor components. The main features of the technique are illustrated by a detailed examination of the specific application to the problem of Heysham 2 Fuel Plug Unit monitoring. (author). 6 figs.

  15. Modular core component support for nuclear reactor

    International Nuclear Information System (INIS)

    Finch, L.M.; Anthony, A.J.

    1975-01-01

    The core of a nuclear reactor is made up of a plurality of support modules for containing components such as fuel elements, reflectors and control rods. Each module includes a component support portion located above a grid plate in a low-pressure coolant zone and a coolant inlet portion disposed within a module receptacle which depends from the grid plate into a zone of high-pressure coolant. Coolant enters the module through aligned openings within the receptacle and module inlet portion and flows upward into contact with the core components. The modules are hydraulically balanced within the receptacles to prevent expulsion by the upward coolant forces. (U.S.)

  16. Reactor container structure

    International Nuclear Information System (INIS)

    Sato, Yoshimi; Fukuda, Yoshio.

    1993-01-01

    A main container of an FBR type reactor using liquid sodium as coolants is attached to a roof slug. The main container contains, as coolants, lower temperature sodium, and high temperature sodium above a reactor core and a partitioning plate. The main container has a structure comprising only longitudinal welded joints in parallel with axial direction in the vicinity of the liquid surface of high temperature sodium where a temperature gradient is steep and great thermal stresses are caused without disposing lateral welded joints in perpendicular to axial direction. Only the longitudinal welded joints having a great fatigue strength are thus disposed in the vicinity of the liquid surface of the high temperature sodium where axial thermal stresses are caused. This can improve reliability of strength at the welded portions of the main container against repeating thermal stresses caused in vicinity of the liquid surface of the main container from a view point of welding method. (I.N.)

  17. Project of integrity assessment of flawed components with structural discontinuity (IAF). Data book for estimation stress intensity factor. Surface crack on ICM housing for penetration in reactor vessel

    International Nuclear Information System (INIS)

    2012-12-01

    The project of Integrity Assessment of Flawed Components with Structural Discontinuity (IAF) was entrusted to Japan Power Engineering and Inspection Corporation (JAPEIC) from Nuclear and Industrial Safety Agency (NISA) and started from FY 2001. And then, it was taken over to Japan Nuclear Energy Safety Organization (JNES) which was established in October 2003 and carried out until FY 2007. In the IAF project, weld joints between nickel based alloys and low alloy steels around penetrations in reactor vessel, safe-end of nozzles and shroud supports were selected from among components and pipe arrangements in nuclear power plants, where high residual stresses were generated due to welding and complex structure. Residual stresses around of the weld joints were estimated by finite element analysis method (FEM) with a general modeling method, then the reasonability and the conservativeness was evaluated. In addition, for postulated surface crack of stress corrosion cracking (SCC), a simple calculation method of stress intensity factor (K) required to estimate the crack growth was proposed and the effectiveness was confirmed. JNES compiled results of the IAF project into Data Books of Residual Stress Analysis of Weld Joint, and Data Book of Simplified Stress Intensity Factor Calculation for Penetration of Reactor as typical Structure Discontinuity, respectively. Data Books of Residual Stress Analysis in Weld Joint. 1. Butt Weld Joint of Small Diameter Cylinder (4B Sch40) (JNES-RE-2012-0005), 2. Dissimilar Metal Weld Joint in Safe End (One-Side Groove Joint (JNES-RE-2012-0006), 3. Dissimilar Metal Weld Joint in Safe End (Large Diameter Both-Side Groove Joint) (JNES-RE-2012-0007), 4. Weld Joint around Penetrations in Reactor Vessel (Insert Joint) (JNES-RE-2012-0008), 5. Weld Joint in Shroud Support (H8, H9, H10 and H11 Welds) (JNES-RE-2012-0009), 6. Analysis Model of Dissimilar Metal Weld Joint Applied Post Weld Heat Treatment (PWHT) (JNES-RE-2012-0010). Data Book of

  18. Sodium removal from Hallam Reactor components

    International Nuclear Information System (INIS)

    Huntsman, L.K.; Meservey, R.H.

    1979-08-01

    This report discussed the removal of sodium from major components of the Hallam Nuclear Power Facility. This facility contained the experimental ractor used to test the feasibility of sodium coolant. The Idaho Operations Office of the Department of Energy assigned EG and G Idaho, Inc., the task of carrying out this decontamination and decommissioning program at the Idaho National Engineering Laboratory (INEL). Since their shipment to the INEL from Lincoln, Nebraska in 1968, the Hallam Reactor components had been stored in inert nitrogen to prevent the sodium in the components from reacting with moisture in the air. The procedure used to react the sodium in the components and to decontaminate them is discussed. Problems and unusual occurrences in the decontamination and decommissioning process are also reported

  19. Structural materials for fusion reactors

    International Nuclear Information System (INIS)

    Victoria, M.; Baluc, N.; Spaetig, P.

    2001-01-01

    In order to preserve the condition of an environmentally safe machine, present selection of materials for structural components of a fusion reactor is made not only on the basis of adequate mechanical properties, behavior under irradiation and compatibility with other materials and cooling media, but also on their radiological properties, i.e. activity, decay heat, radiotoxicity. These conditions strongly limit the number of materials available to a few families of alloys, generically known as low activation materials. We discuss the criteria for deciding on such materials, the alloys resulting from the application of the concept and the main issues and problems of their use in a fusion environment. (author)

  20. Design codes for gas cooled reactor components

    International Nuclear Information System (INIS)

    1990-12-01

    High-temperature gas-cooled reactor (HTGR) plants have been under development for about 30 years and experimental and prototype plants have been operated. The main line of development has been electricity generation based on the steam cycle. In addition the potential for high primary coolant temperature has resulted in research and development programmes for advanced applications including the direct cycle gas turbine and process heat applications. In order to compare results of the design techniques of various countries for high temperature reactor components, the IAEA established a Co-ordinated Research Programme (CRP) on Design Codes for Gas-Cooled Reactor Components. The Federal Republic of Germany, Japan, Switzerland and the USSR participated in this Co-ordinated Research Programme. Within the frame of this CRP a benchmark problem was established for the design of the hot steam header of the steam generator of an HTGR for electricity generation. This report presents the results of that effort. The publication also contains 5 reports presented by the participants. A separate abstract was prepared for each of these reports. Refs, figs and tabs

  1. Fabrication of high performance components for Indian nuclear reactors

    International Nuclear Information System (INIS)

    Jayaraj, R.N.

    2011-01-01

    Nuclear Fuel Complex (NFC), a Unit of the Department of Atomic Energy (DAE) has been engaged for well over three-and-half decades in the manufacture of fuels for Pressurized Heavy Water Reactors (PHWRs) and Boiling Water Reactors (BWRs). All the fuel assembly components, like, fuel clad tubes, end plugs, spacers, spacer grids etc. are also being manufactured at NFC in Zirconium alloy material. Apart from the regular production of these components and finished fuel assemblies, NFC has also been engaged in the production of Zirconium alloy reactor core structurals, like, pressure tubes, calandria tubes, garter springs and reactivity control mechanisms for PHWRs and square channels for BWRs. While all these structural components are produced through standardized flow sheets, there have been continuous innovations carried out in the processes to meet the ever increasing end-use characteristics laid down by the utilities. The paper enumerates various aspects of different technologies developed at NFC for the manufacture of high performance components for reactor applications

  2. Ageing management program for reactor components in HANARO

    International Nuclear Information System (INIS)

    Cho, Yeong-Garp; Wu, Sang-Ik; Lee, Jung-Hee; Ryu, Jeong-Soo; Park, Yong-Chul; Wu, Jong-Sup; Jun, Byung Jin

    2003-01-01

    The HANARO, an open-tank-in-pool type research reactor of 30MWth power in Korea, has operated for 8 years since its initial criticality in February of 1995. The reactor power has been gradually increased to 24 MWth through the service period. Therefore the reactor age is very young from the viewpoint of the ageing effect on the reactor structure and components by neutron irradiation considering the expected reactor lifetime. But, we have a few programs to manage the ageing from the aspect of design lifetime of reactor components. This paper summarizes the overall progress and plan for the ageing management for the reactor components including lifetime extension and design improvement, remote measurements and in-service inspections. The shutoff units and control absorber units have aged more rapidly than other structures or components because the number of rod drop cycles was higher than expected at the design stage. The system commissioning tests, periodic performance tests, and weekly operation for the stable supply of medical radioisotopes overriding the normal cycle operation have contributed to the high frequency of rod drop. Therefore, we have instituted a program to extend the lifetime of the shutoff units and the control absorber units. This program includes an endurance test to verify the performance for the extended number of drops and the management of shutdown methods to minimize the drop cycles for both the shutoff units and the control absorber units. The program also includes the design improvement of the damper mechanism of the control absorber units to reduce the impact force caused by rod drop. The inner shell of the reflector vessel surrounding the core is the most critical part from the viewpoint of neutron irradiation. The periodic measurement of the dimensional change in the vertical straightness of the inner shell is considered as one of the in-service inspections. We developed a few tools and verified the performance to measure the

  3. Aging and life extension of major light water reactor components

    International Nuclear Information System (INIS)

    Shah, V.N.; MacDonald, P.E.

    1993-01-01

    An understanding of the aging degradation of the major pressurized and boiling water reactor structures and components is given. The design and fabrication of each structure or component is briefly described followed by information on the associated stressors. Interactions between the design, materials and various stressors that cause aging degradation are reviewed. In many cases, aging degradation problems have occurred, and the plant experience to date is analyzed. The discussion summarize the available aging-related information and are supported with extensive references, including references to US Nuclear Regulatory Commission (USNRC) documents, Electric Power Research Institute reports, US and international conference proceedings and other publications

  4. Structural integrity aspects of reactor safety

    Indian Academy of Sciences (India)

    A large experimental programme supported the structural integrity demonstration. ... Categories in which the structures, systems and components (SSC) are .... One of the ways in which the decision to live with the defect can be aided is the .... The Advanced Heavy Water Reactor (AHWR) (figure 18) being designed by BARC ...

  5. Seismic behaviour of gas cooled reactor components

    International Nuclear Information System (INIS)

    1990-08-01

    On invitation of the French Government the Specialists' Meeting on the Seismic Behaviour of Gas-Cooled Reactor Components was held at Gif-sur-Yvette, 14-16 November 1989. This was the second Specialists' Meeting on the general subject of gas-cooled reactor seismic design. There were 27 participants from France, the Federal Republic of Germany, Israel, Japan, Spain, Switzerland, the United Kingdom, the Soviet Union, the United States, the CEC and IAEA took the opportunity to present and discuss a total of 16 papers reflecting the state of the art of gained experiences in the field of their seismic qualification approach, seismic analysis methods and of the capabilities of various facilities used to qualify components and verify analytical methods. Since the first meeting, the sophistication and expanded capabilities of both the seismic analytical methods and the test facilities are apparent. The two main methods for seismic analysis, the impedance method and the finite element method, have been computer-programmed in several countries with the capability of each of the codes dependent on the computer capability. The correlations between calculation and tests are dependent on input assumptions such as boundary conditions, soil parameters and various interactions between the soil, the buildings and the contained equipment. The ability to adjust these parameters and match experimental results with calculations was displayed in several of the papers. The expanded capability of some of the new test facilities was graphically displayed by the description of the SAMSON vibration test facility at Juelich, FRG, capable of dynamically testing specimens weighing up to 25 tonnes, and the TAMARIS facility at the CEA laboratories in Gif-sur-Yvette where the largest table is capable of testing specimens weighing up to 100 tonnes. The proceedings of this meeting contain all 16 presented papers. A separate abstract was prepared for each of these papers. Refs, figs and tabs

  6. Scale modeling flow-induced vibrations of reactor components

    International Nuclear Information System (INIS)

    Mulcahy, T.M.

    1982-06-01

    Similitude relationships currently employed in the design of flow-induced vibration scale-model tests of nuclear reactor components are reviewed. Emphasis is given to understanding the origins of the similitude parameters as a basis for discussion of the inevitable distortions which occur in design verification testing of entire reactor systems and in feature testing of individual component designs for the existence of detrimental flow-induced vibration mechanisms. Distortions of similitude parameters made in current test practice are enumerated and selected example tests are described. Also, limitations in the use of specific distortions in model designs are evaluated based on the current understanding of flow-induced vibration mechanisms and structural response

  7. Component nuclear containment structure

    International Nuclear Information System (INIS)

    Harstead, G.A.

    1979-01-01

    The invention described is intended for use primarily as a nuclear containment structure. Such structures are required to surround the nuclear steam supply system and to contain the effects of breaks in the nuclear steam supply system, or i.e. loss of coolant accidents. Nuclear containment structures are required to withstand internal pressure and temperatures which result from loss of coolant accidents, and to provide for radiation shielding during operation and during the loss of coolant accident, as well as to resist all other applied loads, such as earthquakes. The nuclear containment structure described herein is a composite nuclear containment structure, and is one which structurally combines two previous systems; namely, a steel vessel, and a lined concrete structure. The steel vessel provides strength to resist internal pressure and accommodate temperature increases, the lined concrete structure provides resistance to internal pressure by having a liner which will prevent leakage, and which is in contact with the concrete structure which provides the strength to resist the pressure

  8. Creep buckling problems in fast reactor components

    International Nuclear Information System (INIS)

    Ramesh, R.; Damodaran, S.P.; Chellapandi, P.; Chetal, S.C.; Bhoje, S.B.

    1995-01-01

    Creep buckling analyses for two important components of 500 M We Prototype Fast Breeder Reactor (PFBR), viz. Intermediate Heat Exchanger (IHX) and Inner Vessel (IV), are reported. The INCA code of CASTEM system is used for the large displacement elasto-plastic-creep analysis of IHX shell. As a first step, INCA is validated for a typical benchmark problem dealing with the creep buckling of a tube under external pressure. Prediction of INCA is also compared with the results obtained using Hoff's theory. For IV, considering the prohibitively high computational cost for the actual analysis, a simplified analysis which involves only large displacement elastoplastic buckling analysis is performed using isochronous stress strain curve approach. From both of these analysis is performed using isochronous stress strain curve approach. From both of these analysis, it has been inferred that creep buckling failure mode is not of great concern in the design of PFBR components. It has also been concluded from the analysis that Creep Cross Over Curve given in RCC-MR is applicable for creep buckling failure mode also. (author). 8 refs., 9 figs., 1 tab

  9. Lifetime analysis of fusion-reactor components

    International Nuclear Information System (INIS)

    Mattas, R.F.

    1983-01-01

    A one-dimensional computer code has been developed to examine the lifetime of first-wall and impurity-control components. The code incorporates the operating and design parameters, the material characteristics, and the appropriate failure criteria for the individual components. The major emphasis of the modelling effort has been to calculate the temperature-stress-strain-radiation effects history of a component so that the synergystic effects between sputtering erosion, swelling, creep, fatigue, and crack growth can be examined. The general forms of the property equations are the same for all materials in order to provide the greatest flexibility for materials selection in the code. The code is capable of determining the behavior of a plate, composed of either a single or dual material structure, that is either totally constrained or constrained from bending but not from expansion. The code has been utilized to analyze the first walls for FED/INTOR and DEMO

  10. Method to chemically decontaminate nuclear reactor components

    International Nuclear Information System (INIS)

    Bertholdt, H.O.

    1984-01-01

    The large decontamination of components of the primary circuit of activated corrosion products in the oxide layer of the structure materials firstly involves an approx. 1 hour oxidation treatment with alkali permanganate solution. Following intermediate rinsing with deionate, they are etched with an inhibited citrate-oxalate solution for 5-20 hours. This is followed by post-treatment with a citric acid/H2O2 solution containing suspended fiber particles. (orig./PW)

  11. Repair welding of fusion reactor components. Final technical report

    International Nuclear Information System (INIS)

    Chin, B.A.; Wang, C.A.

    1997-01-01

    The exposure of metallic materials, such as structural components of the first wall and blanket of a fusion reactor, to neutron irradiation will induce changes in both the material composition and microstructure. Along with these changes can come a corresponding deterioration in mechanical properties resulting in premature failure. It is, therefore, essential to expect that the repair and replacement of the degraded components will be necessary. Such repairs may require the joining of irradiated materials through the use of fusion welding processes. The present ITER (International Thermonuclear Experimental Reactor) conceptual design is anticipated to have about 5 km of longitudinal welds and ten thousand pipe butt welds in the blanket structure. A recent study by Buende et al. predict that a failure is most likely to occur in a weld. The study is based on data from other large structures, particularly nuclear reactors. The data used also appear to be consistent with the operating experience of the Fast Flux Test Facility (FFTF). This reactor has a fuel pin area comparable with the area of the ITER first wall and has experienced one unanticipated fuel pin failure after two years of operation. The repair of irradiated structures using fusion welding will be difficult due to the entrapped helium. Due to its extremely low solubility in metals, helium will diffuse and agglomerate to form helium bubbles after being trapped at point defects, dislocations, and grain boundaries. Welding of neutron-irradiated type 304 stainless steels has been reported with varying degree of heat-affected zone cracking (HAZ). The objectives of this study were to determine the threshold helium concentrations required to cause HAZ cracking and to investigate techniques that might be used to eliminate the HAZ cracking in welding of helium-containing materials

  12. NEUTRONIC REACTOR STRUCTURE

    Science.gov (United States)

    Weinberg, A.M.; Vernon, H.C.

    1961-05-30

    A neutronic reactor is described. It has a core consisting of natural uranium and heavy water and having a K-factor greater than unity which is surrounded by a reflector consisting of natural uranium and ordinary water having a Kfactor less than unity.

  13. Gap and impact of LMR [Liquid Metal Reactor] piping systems and reactor components

    International Nuclear Information System (INIS)

    Ma, D.C.; Gvildys, J.; Chang, Y.W.

    1987-01-01

    Because of high operation temperature, the LMR (Liquid Metal Reactor) plant is characterized by the thin-walled piping and components. Gaps are often present to allow free thermal expansion during normal plant operation. Under dynamic loadings, such as seismic excitation, if the relative displacement between the components exceeds the gap distance, impacts will occur. Since the components and piping become brittle over their design lifetime, impact is of important concern for it may lead to fractures of components and other serious effects. This paper deals with gap and impact problems in the LMR reactor components and piping systems. Emphasis is on the impacts due to seismic motion. Eight sections are contained in this paper. The gap and impact problems in LMR piping systems are described and a parametric study is performed on the effects of gap-induced support nonlinearity on the dynamics characteristics of the LMR piping systems. Gap and impact problems in the LMR reactor components are identified and their mathematical models are illustrated, and the gap and impact problems in the seismic reactor scram are discussed. The mathematical treatments of various impact models are also described. The uncertainties in the current seismic impact analyses of LMR components and structures are presented. An impact test on a 1/10-scale LMR thermal liner is described. The test results indicated that several clusters of natural modes can be excited by the impact force. The frequency content of the excited modes depends on the duration of the impact force; the shorter the duration, the higher the frequency content

  14. Canisters and nonfuel components at commercial nuclear reactors

    International Nuclear Information System (INIS)

    Gibbard, K.; Thorpe, J.; Moore, R.S.

    1995-01-01

    The Energy Information Administration of the U.S. Department of Energy (DOE) collects data annually from commercial nuclear power reactors via the Nuclear Fuel Data survey, Form RW-859. Over the past three years, the survey has collected data on the quantities and types of nonfuel components and on the quantities and contents of canisters in storage at reactor sites. This paper focuses on the annual changes in the data, specific implications of these changes, and lessons that should be applied to future revisions of the study. The total number of canisters reported by utilities for each year from 1986 to 1993 is listed. Changes in the quantities of nonfuel components report by General Reactors from 1992 to 1993 are also provided. Comparisons of canister and nonfuel components components data from year to year and from reactor to reactor point out that survey questions on these topics have been interpreted differently by reactor personnel

  15. Aging of metal components in US nuclear reactors

    International Nuclear Information System (INIS)

    Mayfield, M.E.; Strosnider, J.R.

    1998-01-01

    This paper presents an overview of the aging of metal components in U.S. Light Water Reactors. The types of degradation being experienced in components such as the pressure vessel, piping, reactor internals, and steam generators, and the programs being implemented to manage the degradation are discussed. (author)

  16. Status of ageing for reactor components in HANARO

    International Nuclear Information System (INIS)

    Cho, Yeong Garp; Wu, Jong Sup; Lee, Jung Hee; Ryu, Jeong Soo; Choung, Yun Hang; Jun, Byung Jin

    2005-01-01

    This paper summarizes the ageing status, the history of the performance and maintenance, and the ageing management program for the reactor structure, shutoff units and control absorber units of HANARO which has been operated for 10 years. From the ageing point of view, the results of the visual inspections of the core components, a deformation measurement of the core inner shell, and wear measurements of the fuel channels are described. The histories of the maintenance, performance and drop cycles were evaluated for the shutoff units and control absorber units. Also there is a summary for the lifetime extension program for the shutoff and control absorber units

  17. Overview moderator material for nuclear reactor components

    International Nuclear Information System (INIS)

    Mairing Manutu Pongtuluran; Hendra Prihatnadi

    2009-01-01

    In order for a reactor design is considered acceptable absolute technical requirement is fulfilled because the most important part of a reactor design. Safety considerations emphasis on the handling of radioactive substances emitted during the operation of a reactor and radioactive waste handling. Moderator material is a layer that interacts directly with neutrons split the nuclear fuel that will lead to changes in physical properties, nuclear properties, mechanical properties and chemical properties. Reviews moderator of this time is of the types of moderator is often used to meet the requirements as nuclear material. (author)

  18. Ageing investigation and upgrading of components/systems of Kartini research reactor

    Energy Technology Data Exchange (ETDEWEB)

    Syarip,; Setiawan, Widi [Yogyakarta Nuclear Research Centre, Yogyakarta (Indonesia)

    1998-10-01

    Kartini research reactor has been operated in good condition and has demonstrated successful operation for the past 18 years, utilized for: reactor kinetic and control studies, instrumentation tests, neutronic and thermohydraulic studies, routine neutron activation analysis, reactor safety studies, training for research reactor operators and supervisors, and reactor physics experiments. Several components of Kartini reactor use components from the abandoned IRT-2000 Project at Serpong and from Bandung Reactor Centre such as: reactor tank, reactor core, heat exchanger, motor blower for ventilation system, fuel elements, etc. To maintain a good operating performance and also for aging investigation purposes, the component failure data collection has been done. The method used is based on the Manual on Reliability Data Collection For Research Reactor PSAs, IAEA TECDOC 636, and analyzed by using Data Entry System (DES) computer code. Analysis result shows that the components/systems failure rate of Kartini reactor is around 1,5.10{sup -4} up to 2,8.10{sup -4} per hour, these values are within the ranges of the values indicated in IAEA TECDOC 478. Whereas from the analysis of irradiation history shows that the neutron fluence of fuel element with highest burn-up (2,05 gram U-235 in average) is around 1.04.10{sup 16} n Cm{sup -2} and this value is still far below its limiting value. Some reactor components/systems have been replaced and upgraded such as heat exchanger, instrumentation and control system (ICS), etc. The new reactor ICS was installed in 1994 which is designed as a distributed structure by using microprocessor based systems and bus system technology. The characteristic and operating performance of the new reactor ICS, as well as the operation history and improvement of the Kartini research reactor is presented. (J.P.N.)

  19. New facilities in Japan materials testing reactor for irradiation test of fusion reactor components

    International Nuclear Information System (INIS)

    Kawamura, H.; Sagawa, H.; Ishitsuka, E.; Sakamoto, N.; Niiho, T.

    1996-01-01

    The testing and evaluation of fusion reactor components, i.e. blanket, plasma facing components (divertor, etc.) and vacuum vessel with neutron irradiation is required for the design of fusion reactor components. Therefore, four new test facilities were developed in the Japan Materials Testing Reactor: an in-pile functional testing facility, a neutron multiplication test facility, an electron beam facility, and a re-weldability facility. The paper describes these facilities

  20. Verification of Remote Inspection Techniques for Reactor Internal Structures of Liquid Metal Reactor

    International Nuclear Information System (INIS)

    Joo, Young Sang; Lee, Jae Han

    2007-02-01

    The reactor internal structures and components of a liquid metal reactor (LMR) are submerged in hot sodium of reactor vessel. The division 3 of ASME code section XI specifies the visual inspection as major in-service inspection (ISI) methods of reactor internal structures and components. Reactor internals of LMR can not be visually examined due to opaque liquid sodium. The under-sodium viewing techniques using an ultrasonic wave should be applied for the visual inspection of reactor internals. Recently, an ultrasonic waveguide sensor with a strip plate has been developed for an application to the under-sodium inspection. In this study, visualization technique, ranging technique and monitoring technique have been suggested for the remote inspection of reactor internals by using the waveguide sensor. The feasibility of these remote inspection techniques using ultrasonic waveguide sensor has been evaluated by an experimental verification

  1. Verification of Remote Inspection Techniques for Reactor Internal Structures of Liquid Metal Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Joo, Young Sang; Lee, Jae Han

    2007-02-15

    The reactor internal structures and components of a liquid metal reactor (LMR) are submerged in hot sodium of reactor vessel. The division 3 of ASME code section XI specifies the visual inspection as major in-service inspection (ISI) methods of reactor internal structures and components. Reactor internals of LMR can not be visually examined due to opaque liquid sodium. The under-sodium viewing techniques using an ultrasonic wave should be applied for the visual inspection of reactor internals. Recently, an ultrasonic waveguide sensor with a strip plate has been developed for an application to the under-sodium inspection. In this study, visualization technique, ranging technique and monitoring technique have been suggested for the remote inspection of reactor internals by using the waveguide sensor. The feasibility of these remote inspection techniques using ultrasonic waveguide sensor has been evaluated by an experimental verification.

  2. Coatings for fast breeder reactor components

    International Nuclear Information System (INIS)

    Johnson, R.N.

    1984-04-01

    Several types of metallurgical coatings are used in the unique environments of the fast breeder reactor. Most of the coatings have been developed for tribological applications, but some also serve as corrosion barriers, diffusion barriers, or radionuclide traps. The materials that have consistently given the best performance as tribological coatings in the breeder reactor environments have been coatings based on chromium carbide, nickel aluminide, or Tribaloy 700 (a nickel-base hard-facing alloy). Other coatings that have been qualified for limited applications include chromium plating for low temperature galling protection and nickel plating for radionuclide trapping

  3. Optical inspections of research reactor tanks and tank components

    International Nuclear Information System (INIS)

    Boeck, H.; Hammer, J.

    1988-01-01

    By the end of 1987 worldwide there were 326 research reactors in operation, 276 of them operating more than 10 years, and 195 of them operating more than 20 years. The majority of these reactors are swimming-pool type or tank type reactors using aluminium as structural material. Although aluminium has prooven its excellent properties for reactor application in primary system, it is however subjected to various types of corrosion if it gets into contact with other materials such as mild steel in the presence of destilled water. This paper describes various methods of research reactor tank inspections, maintenance and repair possibilities. 9 figs. (Author)

  4. Reactor structure for FBR

    International Nuclear Information System (INIS)

    Yamazaki, Tsukasa.

    1986-01-01

    Purpose: To prevent deformation of equipment for FBR structure by inhibiting free convection generated at the roof slab through device. Constitution: The labyrinth is placed between the lower part of the roof slab and the lower one of the positioning flange, and then, convection-preventive wrinkle is provided for the side wall for the positioning flange against the roof slab side wall. The upper part of the positioning flange is fixed to the upper surface of the roof slab, the through-device flange is connected to the lower flange, and prevent generation of thermal stress. Thus, free convection at the through-device is prevented, and it has become possible to miniaturize the seal section of the intermediate heat exchanger and prevent galling of the circulating pump. The joint position of the positioning flange with the through-device flange can be shifted to the same height level of the roof slab, and the length under the hook of the overhead crance can be reduced. (Horiuchi, T.)

  5. Reactor component inventory system at FFTF

    International Nuclear Information System (INIS)

    Ordonez, C.R.; Redekopp, R.D.; Reed, E.A.

    1985-02-01

    A reliable inventory control system was developed at the Fast Flux Test Facility (FFTF) to keep track of the occupancy of 900 refueling facility locations, to compile historical data on the movement of each reactor assembly, and to simulate assembly moves. The simulate capability is valuable because it allows verification of documents before they are issued for use in the plant, and eliminates the possibility of planning illegal or impossible moves. The system is installed on a UNIVAC 1100 computer and is maintained using a data base management system by Sperry Univac called MAPPER

  6. CRBR reactor structures design. BRC meeting presentation

    International Nuclear Information System (INIS)

    Pennell, W.E.

    1975-01-01

    Some of the more important developments in LMFBR structures design technology are described and the application of the technology to design of the CRBR reactor components is illustrated. The LMFBR is both a high-temperature and a high-ΔT machine. High-temperature operation (up to 1100 0 F) requires that the designer consider the effects of thermal creep as a deformation mechanism and stress rupture as a failure mode. The large ΔT across the core coupled with a low core thermal inertia and the high conductivity of the sodium coolant combine to produce severe temperature gradients during a reactor scram. Structures designed to operate in this environment must be both light and stiff to minimize transient thermal stresses and prevent unacceptable flow-induced vibrations. Thermal shields may be required to protect the load-bearing structure. At CRBR core-component goal fluence levels, the predicted magnitude of core-component dimensional changes due to irradiation swelling and creep is very large compared with the more familiar dimensional changes associated with thermal expansion and thermal creep. The design of the core components, and in particular the core restraint system, is dominated by the need to accommodate the effects of irradiation swelling, creep and du []tility loss considerations. (auth)

  7. Common cause failures of reactor pressure components

    International Nuclear Information System (INIS)

    Mankamo, T.

    1978-01-01

    The common cause failure is defined as a multiple failure event due to a common cause. The existence of common failure causes may ruin the potential advantages of applying redundancy for reliability improvement. Examples relevant to large mechanical components are presented. Preventive measures against common cause failures, such as physical separation, equipment diversity, quality assurance, and feedback from experience are discussed. Despite the large number of potential interdependencies, the analysis of common cause failures can be done within the framework of conventional reliability analysis, utilizing, for example, the method of deriving minimal cut sets from a system fault tree. Tools for the description and evaluation of dependencies between components are discussed: these include the model of conditional failure causes that are common to many components, and evaluation of the reliability of redundant components subjected to a common load. (author)

  8. Fatigue evaluation in reactor vessel components

    International Nuclear Information System (INIS)

    Mattar Neto, Miguel; Miranda, Carlos A. de J.

    1994-01-01

    This paper presents a sequence of increasing complexity forms of evaluating fatigue damage of nuclear pressure vessel components caused by cycling loadings. Examples are included in order to illustrate such procedures. (author)

  9. Component mode synthesis in structural dynamics

    International Nuclear Information System (INIS)

    Reddy, G.R.; Vaze, K.K.; Kushwaha, H.S.

    1993-01-01

    In seismic analysis of Nuclear Reactor Structures and equipments eigen solution requires large computer time. Component mode synthesis is an efficient technique with which one can evaluate dynamic characteristics of a large structure with minimum computer time. Due to this reason it is possible to do a coupled analysis of structure and equipment which takes into account the interaction effects. Basically in this the method large size structure is divided into small substructures and dynamic characteristics of individual substructure are determined. The dynamic characteristics of entire structure are evaluated by synthesising the individual substructure characteristics. Component mode synthesis has been applied in this paper to the analysis of a tall heavy water upgrading tower. Use of fixed interface normal modes, constrained modes, attachment modes in the component mode synthesis using energy principle and using Ritz vectors have been discussed. The validity of this method is established by solving fixed-fixed beam and comparing the results obtained by conventional and classical method. The eigen value problem has been solved using simultaneous iteration method. (author)

  10. Development and verification test of integral reactor major components

    Energy Technology Data Exchange (ETDEWEB)

    Kim, J. I.; Kim, Y. W.; Kim, J. H. and others

    1999-03-01

    The conceptual designs for SG, MCP, CEDM to be installed in the integral reactor SMART were developed. Three-dimensional CAD models for the major components were developed to visualize the design concepts. Once-through helical steam generator was conceptually designed for SMART. Canned motor pump was adopted in the conceptual design of MCP. Linear pulse motor type and ballscrew type CEDM, which have fine control capabilities were studied for adoption in SMART. In parallel with the structural design, the electro-magnetic design was performed for the sizing motors and electro-magnet. Prototypes for the CEDM and MCP sub-assemblies were developed and tested to verify the performance. The impeller design procedure and the computer program to analyze the dynamic characteristics of MCP rotor shaft were developed. The design concepts of SG, MCP, CEDM were also invetigated for the fabricability.

  11. Development and verification test of integral reactor major components

    International Nuclear Information System (INIS)

    Kim, J. I.; Kim, Y. W.; Kim, J. H. and others

    1999-03-01

    The conceptual designs for SG, MCP, CEDM to be installed in the integral reactor SMART were developed. Three-dimensional CAD models for the major components were developed to visualize the design concepts. Once-through helical steam generator was conceptually designed for SMART. Canned motor pump was adopted in the conceptual design of MCP. Linear pulse motor type and ballscrew type CEDM, which have fine control capabilities were studied for adoption in SMART. In parallel with the structural design, the electro-magnetic design was performed for the sizing motors and electro-magnet. Prototypes for the CEDM and MCP sub-assemblies were developed and tested to verify the performance. The impeller design procedure and the computer program to analyze the dynamic characteristics of MCP rotor shaft were developed. The design concepts of SG, MCP, CEDM were also invetigated for the fabricability

  12. Structure of thermonuclear reactor wall

    International Nuclear Information System (INIS)

    Yamazaki, Seiichiro.

    1991-01-01

    In a thermonuclear reactor wall, there has been a worry that the brazing material is melted by high temperature heat and particle load, to peel off the joined portion and the protecting material is destroyed by temperature elevation, to expose the heat sink material. Then, in the reactor core structures of a thermonuclear reactor, such as a divertor plate comprising a protecting material made of carbon material and the heat sink material joined by brazing, a plate material made of a so-called refractory metal having a high atomic number such as tungsten, molybdenum or the alloy thereof is embedded or attached to an accurate position of the protecting material. This can prevent the brazing portion from destruction by escaping electrons generated upon occurrence of abnormality in the thermonuclear reactor, and peeling or destroy of the protecting material and the heat sink material. Sufficient characteristics of plasmas can always be maintained by disposing a material having a small atomic number, for example, carbon material, to the position facing to the plasmas. (N.H.)

  13. Evaluation of flow-induced vibration prediction techniques for in-reactor components

    International Nuclear Information System (INIS)

    Mulcahy, T.M.; Turula, P.

    1975-05-01

    Selected in-reactor components of a hydraulic and structural dynamic scale model of the U. S. Energy Research and Development Administration experimental Fast Test Reactor have been studied in an effort to develop and evaluate techniques for predicting vibration behavior of elastic structures exposed to a moving fluid. Existing analysis methods are used to compute the natural frequencies and modal shapes of submerged beam and shell type components. Component response is calculated, assuming as fluid forcing mechanisms both vortex shedding and random excitations characterized by the available hydraulic data. The free and force vibration response predictions are compared with extensive model flow and shaker test data. (U.S.)

  14. Structural dynamics in fast reactor accident analysis

    International Nuclear Information System (INIS)

    Fistedis, S.H.

    1975-01-01

    Analyses and codes are under development combining the hydrodynamics and solid mechanics (and more recently the bubble dynamics) phenomena to gage the stresses, strains, and deformations of important primary components, as well as the overall adequacy of primary and secondary containments. An arbitrary partition of the structural components treated evolves into (1) a core mechanics effort; and (2) a primary system and containment program. The primary system and containment program treats the structural response of components beyond the core, starting with the core barrel. Combined hydrodynamics-solid mechanics codes provide transient stresses and strains and final deformations for components such as the reactor vessel, reactor cover, cover holddown bolts, as well as the pulses for which the primary piping system is to be analyzed. Both, Lagrangian and Eulerian two-dimensional codes are under development, which provide greater accuracy and longer durations for the treatment of HCDA. The codes are being augmented with bubble migration capability pertaining to the latter stages of the HCDA, after slug impact. Recent developments involve the adaptation of the 2-D Eulerian primary system code to the 2-D elastic-plastic treatment of primary piping. Pulses are provided at the vessel-primary piping interfaces of the inlet and outlet nozzles, calculation includes the elbows and pressure drops along the components of the primary piping system. Recent improvements to the primary containment codes include introduction of bending strength in materials, Langrangian mesh regularization techniques, and treatment of energy absorbing materials for the slug impact. Another development involves the combination of a 2-D finite element code for the reactor cover with the hydrodynamic containment code

  15. Sodium components cleaning status in the Italian fast reactor program

    Energy Technology Data Exchange (ETDEWEB)

    De Luca, B [CNEN-RIT/MAT - Laboratorio Sviluppo Processi - C.S.N. Cassacia, Rome (Italy); Labanti, V [CNEN-DRV, Bologna (Italy); Mennucci, M [NIRA, Genoa (Italy)

    1978-08-01

    As a consequence of the Italian Fast Reactor Development, mainly aimed to the PEC project and to the participation in the French Superphenix project, it is of increasing importance to set up a reliable method for specific reactor components and related test loops. The first problem was the cleaning of the PEC fuelling machine. In order to perform the routine maintenance of the machine an alcohol cleaning method based on the use of 2-butoxyethanol-NN dimethylformamide mixture has been proposed.

  16. Nondestructive testing of nuclear reactor components integrity

    International Nuclear Information System (INIS)

    Mala, M.; Miklos, M.

    2011-01-01

    Nuclear energy must respond to current challenges in the energy market. The significant parameters are increase of the nuclear fuel price, closed fuel cycle, reduction and safe and the final disposal of high level radioactive waste. Nowadays, the discussions on suitable energy mix are taking place not only here in Czech Republic, but also in many other European countries. It is necessary to establish an appropriate ratio among the production of electricity from conventional, nuclear and renewable energy sources. Also, it is necessary to find ways how to streamline the economy, central part of the nuclear fuel cycle and thereby to increase the competitiveness of nuclear energy. This streamlining can be carried out by improving utilization of existing nuclear fuel with maintaining a high degree of nuclear facilities safety. Increasing operational reliability and safety together with increasing utilization of nuclear fuel place increasing demands on monitoring of changes during fuel burnup. The potential fuel assembly damages in light water reactors are prevented by the introduction of new procedures and programs of the fuel assembly monitoring. One of them is the Post Irradiation Inspection Program (PIIP) which is a good tool for monitoring of chemical regime impact on the fuel assembly cladding behavior. Main nondestructive techniques that are used at nuclear power plants for the fuel assembly integrity evaluation are ultrasonic measurements, eddy current measurements, radiographic testing, acoustic techniques and others. Ultrasonic system is usual tool for leak fuel rod evaluation and it is also used at Temelin NPP. Since 2009, Temelin NPP has cooperated with Research Center Rez Ltd in frame of PIIP program at both units WWER 1000. This program was established for US VVantage6 fuel assemblies and also it continues for Russian TVSA-T fuel assemblies. (author)

  17. Component event data bank fast breeder reactors

    International Nuclear Information System (INIS)

    Cambi, G.; Righini, R.; Sola, P.G.; Zappellini, G.

    1986-01-01

    ENEA, the Italian Committee for Research and Development of Nuclear Energy, has created a data bank expressly concerning FBRs. The structure of the bank is similar to previous CEDB for LWR, but has been fitted to the new requirements typical of FBRs and to the different conditions of the industry. (DG)

  18. UK fast reactor components - sodium removal decontamination and requalification

    International Nuclear Information System (INIS)

    Donaldson, D.M.; Bray, J.A.; Newson, I.H.

    1978-01-01

    Over the past two decades extensive experience on sodium removal techniques has been gained at the UKAEA's Dounreay Nuclear Establishment from both the Dounreay Fact Reactor (DFR) and the Prototype Fast Reactor (PFR). This experience has created confidence that complex components can be cleaned of sodium, maintenance or repair operations carried out, and the components successfully re-used. Part 2 of the paper, which describes recent operations associated with the PFR, demonstrates the background to these views. This past and continuing experience is being used in forming the basis of the plant to be provided for sodium removal, decontamination and requalification of components in the UK's future commercial fast reactors. Further improvements in techniques and in component designs can be expected in the course of the next few years. Consequently UK philosophy and approach with respect to maintenance and repair operations is sufficiently flexible to enable relevant improvements to be incorporated into the next scheduled fast reactor - the Commercial Demonstration Fast Reactor (CUR). This paper summarises the factors which are being taken into consideration in this continuously advancing field

  19. UK fast reactor components - sodium removal decontamination and requalification

    Energy Technology Data Exchange (ETDEWEB)

    Donaldson, D M [FRDD, UKAEA, Risley (United Kingdom); Bray, J A; Newson, I H [UKAEA, Dounreay Nuclear Power Establishment, Thurso (United Kingdom)

    1978-08-01

    Over the past two decades extensive experience on sodium removal techniques has been gained at the UKAEA's Dounreay Nuclear Establishment from both the Dounreay Fact Reactor (DFR) and the Prototype Fast Reactor (PFR). This experience has created confidence that complex components can be cleaned of sodium, maintenance or repair operations carried out, and the components successfully re-used. Part 2 of the paper, which describes recent operations associated with the PFR, demonstrates the background to these views. This past and continuing experience is being used in forming the basis of the plant to be provided for sodium removal, decontamination and requalification of components in the UK's future commercial fast reactors. Further improvements in techniques and in component designs can be expected in the course of the next few years. Consequently UK philosophy and approach with respect to maintenance and repair operations is sufficiently flexible to enable relevant improvements to be incorporated into the next scheduled fast reactor - the Commercial Demonstration Fast Reactor (CUR). This paper summarises the factors which are being taken into consideration in this continuously advancing field.

  20. Composites as structural materials in fusion reactors

    International Nuclear Information System (INIS)

    Megusar, J.

    1989-01-01

    In fusion reactors, materials are used under extreme conditions of temperature, stress, irradiation, and chemical environment. The absence of adequate materials will seriously impede the development of fusion reactors and might ultimately be one of the major difficulties. Some of the current materials problems can be solved by proper design features. For others, the solution will have to rely on materials development. A parallel and balanced effort between the research in plasma physics and fusion-related technology and in materials research is, therefore, the best strategy to ultimately achieve economic, safe, and environmentally acceptable fusion. The essential steps in developing composites for structural components of fusion reactors include optimization of mechanical properties followed by testing under fusion-reactor-relevant conditions. In optimizing the mechanical behavior of composite materials, a wealth of experience can be drawn from the research on ceramic matrix and metal matrix composite materials sponsored by the Department of Defense. The particular aspects of this research relevant to fusion materials development are methodology of the composite materials design and studies of new processing routes to develop composite materials with specific properties. Most notable examples are the synthesis of fibers, coatings, and ceramic materials in their final shapes form polymeric precursors and the infiltration of fibrous preforms by molten metals

  1. Reactor Structural Materials: Reactor Pressure Vessel Steels

    International Nuclear Information System (INIS)

    Chaouadi, R.

    2000-01-01

    The objectives of SCK-CEN's R and D programme on Rector Pressure Vessel (RPV) Steels are:(1) to complete the fracture toughness data bank of various reactor pressure vessel steels by using precracked Charpy specimens that were tested statically as well as dynamically; (2) to implement the enhanced surveillance approach in a user-friendly software; (3) to improve the existing reconstitution technology by reducing the input energy (short cycle welding) and modifying the stud geometry. Progress and achievements in 1999 are reported

  2. MULTIMETAL - Structural performance of multimetal component

    International Nuclear Information System (INIS)

    Keim, Elisabeth; Blasset, Sebastien; Tiete, Ralf; Gilles, Philippe; Karjalainen-Roikonen, Paeivi

    2012-01-01

    The main objectives of the project are: - Develop a standard for fracture resistance testing in multi-metal specimens; - Develop harmonized procedures for dissimilar metal welds brittle and ductile integrity assessment. The underlying aim of the project is to provide recommendations for a good practice approach for the integrity assessment (including testing) of dissimilar metal welds as part of overall integrity analyses and leak-before-break (LBB) procedures. In a nuclear power plant (NPP) a single metallic component may be fabricated from different materials. For example, reactor pressure vessel (RPV) components are mainly made of ferritic steel, whereas some of the connecting pipelines are fabricated from austenitic stainless steel. As a consequence, components made of different kind of steels need to be connected. Their connecting welds are called dissimilar metal welds (DMW). Despite extensive research in the past within the EURATOM Framework, e.g. the projects BIMET and ADIMEW, further work is needed to quantify the structural performance of DMWs. The first step of the project is to gather relevant information from field experience. Typical locations of DMWs in Western as well as Eastern type light water reactors (LWRs) will be identified together with their physical and metallurgical characteristics, as well as applicable structural integrity assessment methods. The collection of relevant field information including findings position (flaw) will be followed by computational structural integrity assessment analyses of DMWs for dedicated test configurations and real cases. These analyses will involve simple engineering methods and numerical analyses. The latter also involves the use of innovative micro-mechanical modelling approaches for ductile failure processes in order to augment existing numerical methods for structural integrity assessment of DMWs. Ageing related phenomena and realistic stress distributions in the weld area will be considered. The

  3. Structural analysis of nuclear components

    International Nuclear Information System (INIS)

    Ikonen, K.; Hyppoenen, P.; Mikkola, T.; Noro, H.; Raiko, H.; Salminen, P.; Talja, H.

    1983-05-01

    THe report describes the activities accomplished in the project 'Structural Analysis Project of Nuclear Power Plant Components' during the years 1974-1982 in the Nuclear Engineering Laboratory at the Technical Research Centre of Finland. The objective of the project has been to develop Finnish expertise in structural mechanics related to nuclear engineering. The report describes the starting point of the research work, the organization of the project and the research activities on various subareas. Further the work done with computer codes is described and also the problems which the developed expertise has been applied to. Finally, the diploma works, publications and work reports, which are mainly in Finnish, are listed to give a view of the content of the project. (author)

  4. Structural analysis of NPP components and structures

    International Nuclear Information System (INIS)

    Saarenheimo, A.; Keinaenen, H.; Talja, H.

    1998-01-01

    Capabilities for effective structural integrity assessment have been created and extended in several important cases. In the paper presented applications deal with pressurised thermal shock loading, PTS, and severe dynamic loading cases of containment, reinforced concrete structures and piping components. Hydrogen combustion within the containment is considered in some severe accident scenarios. Can a steel containment withstand the postulated hydrogen detonation loads and still maintain its integrity? This is the topic of Chapter 2. The following Chapter 3 deals with a reinforced concrete floor subjected to jet impingement caused by a postulated rupture of a near-by high-energy pipe and Chapter 4 deals with dynamic loading resistance of the pipe lines under postulated pressure transients due to water hammer. The reliability of the structural integrity analysing methods and capabilities which have been developed for application in NPP component assessment, shall be evaluated and verified. The resources available within the RATU2 programme alone cannot allow performing of the large scale experiments needed for that purpose. Thus, the verification of the PTS analysis capabilities has been conducted by participation in international co-operative programmes. Participation to the European Network for Evaluating Steel Components (NESC) is the topic of a parallel paper in this symposium. The results obtained in two other international programmes are summarised in Chapters 5 and 6 of this paper, where PTS tests with a model vessel and benchmark assessment of a RPV nozzle integrity are described. (author)

  5. Characteristics of outage radiation fields around various reactor components

    International Nuclear Information System (INIS)

    Verzilov, Y.; Husain, A.; Corbin, G.

    2008-01-01

    Full text: Activity monitoring surveys, consisting of gamma spectroscopy and dose rate measurements, of various CANDU station components such as the reactor face, feeder cabinet, steam generators and moderator heat exchangers are often performed during shutdown in order to trend the transport of activity around the primary heat transport and moderator systems. Recently, the increased dose expenditure for work such as feeder inspection and replacement in the reactor vault has also spurred interest in improved characterization of the reactor face fields to facilitate better ALARA decision making and hence a reduction in future dose expenditures. At present, planning for reactor face work is hampered by insufficient understanding of the relative contribution of the various components to the overall dose. In addition to the increased dose expenditure for work at the reactor face, maintenance work associated with horizontal flux detectors and liquid injection systems has also resulted in elevated dose expenditures. For instance at Darlington, radiation fields in the vicinity of horizontal flux detectors (HFD) and Liquid Injection Shutdown System (LISS) nozzle bellows are trending upwards with present contact fields being in the range 16-70 rem/h and working distance fields being in the range 100-500 mrem/h. This paper presents findings based on work currently being funded by the CANDU Owners Group. Measurements were performed at Ontario Power Generation's Pickering and Darlington nuclear stations. Specifically, the following are addressed: Characteristics of Reactor Vault Fields; Characteristics of Steam Generator Fields; Characteristics of Moderator Heat Exchanger Fields. Measurements in the reactor vault were performed at the reactor face, along the length of end fittings, along the length of feeders, at the bleed condenser and at the HFD and LISS nozzle bellows. Steam generator fields were characterized at various elevations above the tube sheet, with and without the

  6. Accounting sodium effect in calculation of strength of nuclear reactor components

    International Nuclear Information System (INIS)

    Nikitin, V.I.

    1981-01-01

    Accounting methods of liquid sodium effect on long-term strength and creep of structural materials of nuclear reactors are considered. The decrease of pearlite steel strength at the decarburization expense and the decrease of plasticity of austenitic steels at the expense of carburization are noted. The necessity to account thermal transfer of mass is shown. Values of safety factors are presented, they are recommended for the design of reactor component parts with the thickness not less than 1 mm [ru

  7. Development and application of a welding procedure for remote repair of Magnox reactor internal components

    International Nuclear Information System (INIS)

    Morgan-Warren, E.J.

    1988-01-01

    This paper summarises the development and application of an all-welding repair method for reinforcing magnox reactor internal components. The development was dominated by the necessity for remote operation and the environmental constraints, in particular the oxide covering on the steel reactor structure. The choice of welding process is described, together with the development of the procedure for remote operation. The quality assurance procedure, including the verification of the technique and monitoring of the repair operation, is discussed. (author)

  8. Replacement of core components in the Advanced Test Reactor

    International Nuclear Information System (INIS)

    Durney, J.L.; Croucher, D.W.

    1990-01-01

    The core internals of the Advanced Test Reactor are subjected to very high neutron fluences resulting in significant aging. The most irradiated components have been replaced on several occasions as a result of the neutron damage. The surveillance program to monitor the aging developed the needed criteria to establish replacement schedules and maximize the use of the reactor. The methods to complete the replacements with minimum radiation exposures to workers have been developed using the experience gained from each replacement. The original design of the reactor core and associated components allows replacements to be completed without special equipment. The plant has operated for about 20 years and is expected to continue operation for at least and additional 25 years. Aging evaluations are in progress to address additional replacements that may be needed during this period

  9. Replacement of core components in the Advanced Test Reactor

    International Nuclear Information System (INIS)

    Durney, J.L.; Croucher, D.W.

    1989-01-01

    The core internals of the Advanced Test Reactor are subjected to very high neutron fluences resulting in significant aging. The most irradiated components have been replaced on several occasions as a result of the neutron damage. The surveillance program to monitor the aging developed the needed criteria to establish replacement schedules and maximize the use of the reactor. Methods to complete the replacements with minimum radiation exposures to workers have been developed using the experience gained from each replacement. The original design of the reactor core and associated components allows replacements to be completed without special equipment. The plant has operated for about 20 years and will continue operation for perhaps another 20 years. Aging evaluations are in program to address additional replacements that may be needed during this extended time period. 3 figs

  10. The main objectives of lifetime management of reactor unit components

    International Nuclear Information System (INIS)

    Dragunov, Y.; Kurakov, Y.

    1998-01-01

    The main objectives of the work concerned with life management of reactor components in Russian Federation are as follows: development of regulations in the field of NPP components ageing and lifetime management; investigations of ageing processes; residual life evaluation taking into account the actual state of NPP systems, real loading conditions and number of load cycles, results of in-service inspections; development and implementation of measures for maintaining/enhancing the NPP safety

  11. Experimental Facilities for Performance Evaluation of Fast Reactor Components

    International Nuclear Information System (INIS)

    Chandramouli, S.; Kumar, V.A. Suresh; Shanmugavel, M.; Vijayakumar, G.; Vinod, V.; Noushad, I.B.; Babu, B.; Kumar, G. Padma; Nashine, B.K.; Rajan, K.K.

    2013-01-01

    Brief details about various experimental facilities catering to the testing and performance evaluation requirements of fast reactor components have been brought out. These facilities have been found to be immensely useful to continue research and development activities in the areas of component development and testing, sodium technology, thermal hydraulics and sodium instrumentation for the SFR’s. In addition new facilities which have been planned will be of great importance for the developmental activities related to future SFR’s

  12. Design of sodium cooled reactor systems and components for maintainability

    International Nuclear Information System (INIS)

    Carr, R.W.; Charnock, H.O.; McBride, J.P.

    1978-09-01

    Special maintenability problems associated with the design and operation of sodium cooled reactor plants are discussed. Some examples of both good and bad design practice are introduced from the design of the FFTF plant and other plants. Subjects include design for drainage, cleaning, decontamination, access, component removal, component disassembly and reassembly, remote tooling, jigs, fixtures, and design for minimizing radiation exposure of maintenance personnel. Check lists are included

  13. Measurement of secondary gamma-ray production cross sections of structural materials for fusion reactor. Extraction of discrete and continuum components

    International Nuclear Information System (INIS)

    Kondo, Tetsuo; Morotomi, Ryutaro; Nishio, Takashi; Murata, Isao; Takahashi, Akito

    2000-01-01

    A new method to deal with measured spectrum of secondary gamma-rays induced by D-T neutrons with Ge detector is proposed. Subtracting background components and discrete peaks from the raw secondary gamma-ray spectrum, the continuum component of secondary gamma-ray was successfully extracted. By using unfolding process, the continuum component of the secondary gamma-ray production cross section was derived. The measured cross section data obtained by this method are very useful for precise evaluation of secondary gamma-ray production cross sections. (author)

  14. Metal plutonium conversion to components of nuclear reactor fuel

    International Nuclear Information System (INIS)

    Subbotin, V.G.; Panov, A.V.; Mashirev, V.P.

    2000-01-01

    Capabilities of different technologies for plutonium conversion to the fuel components of nuclear reactors are studied. Advantages and shortcomings of aqueous and nonaqueous methods of plutonium treatment are shown. Proposals to combine and coordinate efforts of world scientific and technological community in solving problems concerning plutonium of energetic and weapon origin treatment were put forward. (authors)

  15. Chemical immobilization of fission products reactive with nuclear reactor components

    International Nuclear Information System (INIS)

    Grossman, L.N.; Kaznoff, A.I.; Clukey, H.V.

    1975-01-01

    This invention teaches a method of immobilizing deleterious fission products produced in nuclear fuel materials during nuclear fission chain reactions through the use of additives. The additives are disposed with the nuclear fuel materials in controlled quantities to form new compositions preventing attack of reactor components, especially nuclear fuel cld, by the deleterious fission products. (Patent Office Record)

  16. Metal plutonium conversion to components of nuclear reactor fuel

    Energy Technology Data Exchange (ETDEWEB)

    Subbotin, V.G.; Panov, A.V. [Russian Federal Nuclear Center, ALL-Russian Science and Research, Institute of Technical Physics, Snezhinsk (Russian Federation); Mashirev, V.P. [ALL-Russian Science and Research Institute of Chemical Technology, Moscow (Russian Federation)

    2000-07-01

    Capabilities of different technologies for plutonium conversion to the fuel components of nuclear reactors are studied. Advantages and shortcomings of aqueous and nonaqueous methods of plutonium treatment are shown. Proposals to combine and coordinate efforts of world scientific and technological community in solving problems concerning plutonium of energetic and weapon origin treatment were put forward. (authors)

  17. Aging assessment and mitigation for major LWR [light water reactor] components

    International Nuclear Information System (INIS)

    Shah, Y.N.; Ware, A.G.; Conley, D.A.; MacDonald, P.E.; Burns, J.J. Jr.

    1989-01-01

    This paper summarizes some of the results of the Aging Assessment and Mitigation Project sponsored by the US Nuclear Regulatory Commission (USNRC), Office of Nuclear Regulatory Research. The objective of the project is to develop an understanding of the aging degradation of the major light water reactor (LWR) structures and components and to develop methods for predicting the useful life of these components so that the impact of aging on the safe operation of nuclear power plants can be evaluated and addressed. The research effort consists of integrating, evaluating, and updating the available aging-related information. This paper discusses current accomplishments and summarizes the significant degradation processes active in two major components: pressurized water reactor pressurizer surge and spray lines and nozzles, and light water reactor primary coolant pumps. This paper also evaluates the effectiveness of the current inservice inspection programs and presents conclusions and recommendations related to aging of these two major components. 37 refs., 7 figs., 3 tabs

  18. Functional Generalized Structured Component Analysis.

    Science.gov (United States)

    Suk, Hye Won; Hwang, Heungsun

    2016-12-01

    An extension of Generalized Structured Component Analysis (GSCA), called Functional GSCA, is proposed to analyze functional data that are considered to arise from an underlying smooth curve varying over time or other continua. GSCA has been geared for the analysis of multivariate data. Accordingly, it cannot deal with functional data that often involve different measurement occasions across participants and a large number of measurement occasions that exceed the number of participants. Functional GSCA addresses these issues by integrating GSCA with spline basis function expansions that represent infinite-dimensional curves onto a finite-dimensional space. For parameter estimation, functional GSCA minimizes a penalized least squares criterion by using an alternating penalized least squares estimation algorithm. The usefulness of functional GSCA is illustrated with gait data.

  19. Generic component reliability data for research reactor PSA

    International Nuclear Information System (INIS)

    1997-02-01

    The purpose of this document is to provide reference generic component-reliability information for a variety of research reactor types. As noted in Section 2 and Table IV, component data accumulated over many years is in the database. It is expected that the report should provide representative data which will remain valid for a number of years. The database provides component failure rates on a time and/or demand related basis according to the operational modes of the components. No update of the database is presently planned. As a result of the implementation of data collection systems in the research reactors represented in these studies, updating of data from individual facilities could be made available by the contributing research reactor facilities themselves. As noted in Section 1.1, the report does not include a detailed discussion of information regarding component classification and reliability parameter definitions. The report does provide some insights and discussions regarding the practicalities of the data collection process and some guidelines for database usage. 9 refs, 7 tabs

  20. Examination of core components removed from CANDU reactors

    International Nuclear Information System (INIS)

    Cheadle, B.A.; Coleman, C.E.; Rodgers, D.K.; Davies, P.H.; Chow, C.K.; Griffiths, M.

    1988-11-01

    Components in the core of a nuclear reactor degrade because the environment is severe. For example, in CANDU reactors the pressure tubes must contend with the effects of hot pressurised water and damage by a flux of fast neutrons. To evaluate any deterioration of components and determine the cause of the occasional failure, we have developed a wide range of remote-handling techniques to examine radioactive materials. As well as pressure tubes, we have examined calandria tubes, garter springs, end fittings, liquid-zone control units and flux detectors. The results from these examinations have produced solutions to problems and continually provide information to help understand the processes that may limit the lifetime of a component

  1. Fabrication of nuclear ship reactor MRX model and study on inspection and maintenance of components

    International Nuclear Information System (INIS)

    Kasahara, Yoshiyuki; Nakazawa, Toshio; Kusunoki, Tsuyoshi; Takahashi, Hiroki; Yoritsune, Tsutomu.

    1997-10-01

    The MRX (Marine Reactor X) is an integral type small reactor adopting passive safety systems. As for an integral type reactor, primary system components are installed in the reactor vessel. It is therefor important to establish the appropriate procedure for construction, inspection and maintenance, dismauntling, etc., for all components in the reactor vessel as well as in the reactor containment, because inspection space is limited. To study these subjects, a one-fifth model of the MRX was fabricated and operation capabilities were studied. As a result of studies, the following results are obtained. (1) Manufacturing and installing problems of the reactor pressure vessel, the containment vessel and internal components are basically not abserved. (2) Heat transfer tube structures of the steam generator and the heat exchangers of emergency decay heat removal system and containment water cooler were not seen of any problem for fabrication. However, due consideration is required in the detailed design of supports of heat transfer tubes. (3) Further studies should be needed for designs of flange penetrations and leak countermeasures for pipes instrument cables. (4) Arrangements of equipments in the containment should be taken in consideration in detail because the space is narrow. (5) Further discussion is required for installation methods of instruments and cables. (author)

  2. Non-aqueous removal of sodium from reactor components

    Energy Technology Data Exchange (ETDEWEB)

    Welch, F H; Steele, O P [Rockwell International, Atomics International Division, Canoga Park (United States)

    1978-08-01

    Reactor components from sodium-cooled systems. whether radioactive or not, must have the sodium removed before they can be safely handled for 1) disposal, 2) examination and test, or 3) decontamination, repair, and requalification. In the latter two cases, the sodium must be removed in a manner which will not harm the component. and prevent future use. Two methods for sodium removal using non-aqueous techniques have been studied extensively in the U.S.A. in the past few years: the Alcohol Process, which uses a fully denatured ethanol to react away the sodium; and the Evaporative Process, which uses heat and vacuum to evaporate the sodium from the component.

  3. Non-aqueous removal of sodium from reactor components

    International Nuclear Information System (INIS)

    Welch, F.H.; Steele, O.P.

    1978-01-01

    Reactor components from sodium-cooled systems. whether radioactive or not, must have the sodium removed before they can be safely handled for 1) disposal, 2) examination and test, or 3) decontamination, repair, and requalification. In the latter two cases, the sodium must be removed in a manner which will not harm the component. and prevent future use. Two methods for sodium removal using non-aqueous techniques have been studied extensively in the U.S.A. in the past few years: the Alcohol Process, which uses a fully denatured ethanol to react away the sodium; and the Evaporative Process, which uses heat and vacuum to evaporate the sodium from the component

  4. Studies on components for a molten salt reactor

    International Nuclear Information System (INIS)

    Nejedly, M.; Matal, O.

    2003-01-01

    The aim is contribute to a design of selected components of molten salt reactors with fuel in the molten fluoride salt matrix. Molten salt reactors (MSRs) permit the utilization of plutonium and minor actinides as new nuclear fuel from a traditional nuclear power station with production of electric energy. Results of preliminary feasibility studies of an intermediate heat exchanger, a small power molten salt pump and a modular conception of a steam generator for a demonstration unit of the MSR (30 MW) are summarized. (author)

  5. Solutions against PWSCC in dissimilar welds cracks of reactor components

    International Nuclear Information System (INIS)

    Schlader, D.; Michaut, B.; Knapp, M

    2005-01-01

    This article provides a brief overview of the experience accumulated by Framatome ANP in the development and use of repair and mitigation techniques of the PWSCC in dissimilar welds cracks of reactor components. A brief description of the alternatives available to the industry for the solution of this problem for both PWR and BWR reactor types is also included. These solutions have been implemented many times by Framatome ANP in Europe and the US. The article also describes the way the know-how is shared among the different regions of the company in order to offer customer specific solutions. (Author)

  6. A plan for safety and integrity of research reactor components

    International Nuclear Information System (INIS)

    Moatty, Mona S. Abdel; Khattab, M.S.

    2013-01-01

    Highlights: ► A plan for in-service inspection of research reactor components is put. ► Section XI of the ASME Code requirements is applied. ► Components subjected to inspection and their classes are defined. ► Flaw evaluation and its acceptance–rejection criteria are reviewed. ► A plan of repair or replacement is prepared. -- Abstract: Safety and integrity of a research reactor that has been operated over 40 years requires frequent and thorough inspection of all the safety-related components of the facility. The need of increasing the safety is the need of improving the reliability of its systems. Diligent and extensive planning of in-service inspection (ISI) of all reactor components has been imposed for satisfying the most stringent safety requirements. The Safeguards Officer's responsibilities of Section XI of the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code ASME Code have been applied. These represent the most extensive and time-consuming part of ISI program, and identify the components subjected to inspection and testing, methods of component classification, inspection and testing techniques, acceptance/rejection criteria, and the responsibilities. The paper focuses on ISI planning requirements for welded systems such as vessels, piping, valve bodies, pump casings, and control rod-housing parts. The weld in integral attachments for piping, pumps, and valves are considered too. These are taken in consideration of safety class (1, 2, 3, etc.), reactor age, and weld type. The parts involve in the frequency of inspection, the examination requirements for each inspection, the examination method are included. Moreover the flaw evaluation, the plan of repair or replacement, and the qualification of nondestructive examination personnel are considered

  7. Challenges in design of zirconium alloy reactor components

    International Nuclear Information System (INIS)

    Kakodkar, Anil; Sinha, R.K.

    1992-01-01

    Zirconium alloy components used in core-internal assemblies of heavy water reactors have to be designed under constraints imposed by need to have minimum mass, limitations of fabrication, welding and joining techniques with this material, and unique mechanisms for degradation of the operating performance of these components. These constraints manifest as challenges for design and development when the size, shape and dimensions of the components and assemblies are unconventional or untried, or when one is aiming for maximization of service life of these components under severe operating conditions. A number of such challenges were successfully met during the development of core-internal components and assemblies of Dhruva reactor. Some of the then untried ideas which were developed and successfully implemented include use of electron beam welding, cold forming of hemispherical ends of reentrant cans, and a large variety of rolled joints of innovative designs. This experience provided the foundation for taking up and successfully completing several tasks relating to coolant channels, liquid poison channels and sparger channels for PHWRs and test sections for the in-pile loops of Dhruva reactor. For life prediction and safety assessment of coolant channels of PHWRs some analytical tools, notably, a computer code for prediction of creep limited life of coolant channels has been developed. Some of the future challenges include the development of easily replaceable coolant channels and also large diameter coolant channels for Advanced Heavy Water Reactor, and development of solutions to overcome deterioration of service life of coolant channels due to hydriding. (author). 5 refs., 13 figs., 1 tab

  8. A method for evaluation the activity of the reactor components

    International Nuclear Information System (INIS)

    Gugiu, E.D.; Roth, Cs.

    2003-01-01

    The ability to predict the radioactivity levels of the reactor components is an important aspect from waste management point of view, as well as from radioprotection purposes. A special case is represented by the research reactors where, one of the major contributions to the radioactivity inventory is due to the experimental devices involved in various research works during reactor life. Generally, aluminum and aluminum alloys are used in manufacturing these devices; as a result, the work presented in this paper is focused on the qualitative and quantitative analysis of the radioactive isotopes contained in these materials. A device used for silicon doping by neutron transmutation that was placed near TRIGA reactor core is investigated. The isotopic content of various samplings drawn from various points of the device was analyzed by gamma spectrometry using a HPGe detector. Computations, using the MCNP5 code, are also performed in order to evaluate the reaction rates for all the isotopes and their reactions. The Monte Carlo simulations are performed for a detailed geometry and material composition of the reactor core and the device. The Origen-S code is also used in order to evaluate the isotopic inventory and the activity values. A detailed analysis regarding the possibility to estimate by computations and/or by gamma spectrometry the activity values of the isotopes which are of interest for decommissioning is presented in the paper. (authors)

  9. Experimental study on hollow structural component by explosive welding

    Energy Technology Data Exchange (ETDEWEB)

    Duan, Mianjun, E-mail: dmjwl@163.com [PLA University of Science and Technology, Nanjing 210007 (China); Wei, Ling, E-mail: 386006087@qq.com [Tongda College, Nanjing University of Posts and Telecommunication, Nanjing 210007 (China); Hong, Jin [PLA University of Science and Technology, Nanjing 210007 (China); Ran, Hong [Southwestern Institute of Physics, Chengdu 610041 (China); Ma, Rui; Wang, Yaohua [PLA University of Science and Technology, Nanjing 210007 (China)

    2014-12-15

    Highlights: • This paper relates to a study on a thin double-layers hollow structural component by using an explosive welding technology. • This thin double-layer hollow structural component is an indispensable component required for certain core equipment of thermonuclear experimental reactor. • An adjusted explosive welding technology for manufacturing an inconel625 hollow structural component was developed which cannot be made by common technology. • The result shows that a metallurgical bonding was realized by the ribs and slabs of the hollow sheet. • The shearing strength of bonding interface exceeds that of the parent metal. - Abstract: A large thin-walled hollow structural component with sealed channels is required for the vacuum chamber of a thermonuclear experimental reactor, with inconel625 as its fabrication material. This hollow structural component is rarely manufactured by normal machining method, and its manufacture is also problematic in the field of explosive welding. With this in mind, we developed an adjusted explosive welding technology which involves a two-step design, setting and annealing technology. The joints were evaluated using optical microscope and scanning electron microscope, and a mechanical experiment was conducted, involving micro-hardness test, cold helium leak test and hydraulic pressure test. The results showed that a metallurgical bonding was realized by the ribs and slabs, and the shearing strength of the bonding interface exceeded that of the parent metal. Hence, the hollow structural component has a good comprehensive mechanical performance and sealing property.

  10. Experimental study on hollow structural component by explosive welding

    International Nuclear Information System (INIS)

    Duan, Mianjun; Wei, Ling; Hong, Jin; Ran, Hong; Ma, Rui; Wang, Yaohua

    2014-01-01

    Highlights: • This paper relates to a study on a thin double-layers hollow structural component by using an explosive welding technology. • This thin double-layer hollow structural component is an indispensable component required for certain core equipment of thermonuclear experimental reactor. • An adjusted explosive welding technology for manufacturing an inconel625 hollow structural component was developed which cannot be made by common technology. • The result shows that a metallurgical bonding was realized by the ribs and slabs of the hollow sheet. • The shearing strength of bonding interface exceeds that of the parent metal. - Abstract: A large thin-walled hollow structural component with sealed channels is required for the vacuum chamber of a thermonuclear experimental reactor, with inconel625 as its fabrication material. This hollow structural component is rarely manufactured by normal machining method, and its manufacture is also problematic in the field of explosive welding. With this in mind, we developed an adjusted explosive welding technology which involves a two-step design, setting and annealing technology. The joints were evaluated using optical microscope and scanning electron microscope, and a mechanical experiment was conducted, involving micro-hardness test, cold helium leak test and hydraulic pressure test. The results showed that a metallurgical bonding was realized by the ribs and slabs, and the shearing strength of the bonding interface exceeded that of the parent metal. Hence, the hollow structural component has a good comprehensive mechanical performance and sealing property

  11. Component vibration of VVER-reactors - diagnostics and modelling

    International Nuclear Information System (INIS)

    Altstadt, E.; Scheffler, M.; Weiss, F.P.

    1994-01-01

    The model comprises the whole primary circuit, including steam generators, loops, coolant pumps, main isolating valves and certainly the reactor pressure vessel and its internals. It was developed using the finite-element-code ANSYS. The model has a modular structure, so that various operational and assembling states can easily be considered. (orig./DG)

  12. Component and system simulation models for High Flux Isotope Reactor

    International Nuclear Information System (INIS)

    Sozer, A.

    1989-08-01

    Component models for the High Flux Isotope Reactor (HFIR) have been developed. The models are HFIR core, heat exchangers, pressurizer pumps, circulation pumps, letdown valves, primary head tank, generic transport delay (pipes), system pressure, loop pressure-flow balance, and decay heat. The models were written in FORTRAN and can be run on different computers, including IBM PCs, as they do not use any specific simulation languages such as ACSL or CSMP. 14 refs., 13 figs

  13. Structural materials issues for the next generation fission reactors

    Science.gov (United States)

    Chant, I.; Murty, K. L.

    2010-09-01

    Generation-IV reactor design concepts envisioned thus far cater to a common goal of providing safer, longer lasting, proliferation-resistant, and economically viable nuclear power plants. The foremost consideration in the successful development and deployment of Gen-W reactor systems is the performance and reliability issues involving structural materials for both in-core and out-of-core applications. The structural materials need to endure much higher temperatures, higher neutron doses, and extremely corrosive environments, which are beyond the experience of the current nuclear power plants. Materials under active consideration for use in different reactor components include various ferritic/martensitic steels, austenitic stainless steels, nickel-base superalloys, ceramics, composites, etc. This article addresses the material requirements for these advanced fission reactor types, specifically addressing structural materials issues depending on the specific application areas.

  14. Multi-component controllers in reactor physics optimality analysis

    International Nuclear Information System (INIS)

    Aldemir, T.

    1978-01-01

    An algorithm is developed for the optimality analysis of thermal reactor assemblies with multi-component control vectors. The neutronics of the system under consideration is assumed to be described by the two-group diffusion equations and constraints are imposed upon the state and control variables. It is shown that if the problem is such that the differential and algebraic equations describing the system can be cast into a linear form via a change of variables, the optimal control components are piecewise constant functions and the global optimal controller can be determined by investigating the properties of the influence functions. Two specific problems are solved utilizing this approach. A thermal reactor consisting of fuel, burnable poison and moderator is found to yield maximal power when the assembly consists of two poison zones and the power density is constant throughout the assembly. It is shown that certain variational relations have to be considered to maintain the activeness of the system equations as differential constraints. The problem of determining the maximum initial breeding ratio for a thermal reactor is solved by treating the fertile and fissile material absorption densities as controllers. The optimal core configurations are found to consist of three fuel zones for a bare assembly and two fuel zones for a reflected assembly. The optimum fissile material density is determined to be inversely proportional to the thermal flux

  15. Reactor materials program process water component failure probability

    International Nuclear Information System (INIS)

    Daugherty, W. L.

    1988-01-01

    The maximum rate loss of coolant accident for the Savannah River Production Reactors is presently specified as the abrupt double-ended guillotine break (DEGB) of a large process water pipe. This accident is not considered credible in light of the low applied stresses and the inherent ductility of the piping materials. The Reactor Materials Program was initiated to provide the technical basis for an alternate, credible maximum rate LOCA. The major thrust of this program is to develop an alternate worst case accident scenario by deterministic means. In addition, the probability of a DEGB is also being determined; to show that in addition to being mechanistically incredible, it is also highly improbable. The probability of a DEGB of the process water piping is evaluated in two parts: failure by direct means, and indirectly-induced failure. These two areas have been discussed in other reports. In addition, the frequency of a large bread (equivalent to a DEGB) in other process water system components is assessed. This report reviews the large break frequency for each component as well as the overall large break frequency for the reactor system

  16. Reactor control rod supporting structure

    International Nuclear Information System (INIS)

    Akimoto, Tokuzo; Miyata, Hiroshi.

    1984-01-01

    Purpose: To enable stable reactor core control even in extremely great vertical earthquakes, as well as under normal operation conditions in FBR type reactors. Constitution: Since a mechanism for converting the rotational movement of a control rod into vertical movement is placed at the upper portion of the reactor core at high temperature, the mechanism should cause fusion or like other danger after the elapse of a long period of time. In view of the above, the conversion mechanism is disposed to the lower portion of the reactor core at a lower temperature region. Further, the connection between the control rod and the control rod drive can be separated upon great vertical earthquakes. (Seki, T.)

  17. SIMODIS - a software package for simulating nuclear reactor components

    International Nuclear Information System (INIS)

    Guimaraes, Lamartine; Borges, Eduardo M.

    2000-01-01

    In this paper it is presented the initial development effort in building a nuclear reactor component simulation package. This package was developed to be used in the MATLAB simulation environment. It uses the graphical capabilities from MATLAB and the advantages of compiled languages, as for instance FORTRAN and C ++ . From the MATLAB it takes the facilities for better displaying the calculated results. From the compiled languages it takes processing speed. So far models from reactor core, UTSG and OTSG have been developed. Also, a series a user-friendly graphical interfaces have been developed for the above models. As a by product a set of water and sodium thermal and physical properties have been developed and may be used directly as a function from MATLAB, or by being called from a model, as part of its calculation process. The whole set was named SIMODIS, which stands for SIstema MODular Integrado de Simulacao. (author)

  18. Industrial structure at research reactor suppliers

    International Nuclear Information System (INIS)

    Roegler, H.-J.; Bogusch, E.; Friebe, T.

    2001-01-01

    Due to the recent joining of the forces of Framatome S. A. from France and the Nuclear Division of Siemens AG Power Generation (KWU) from Germany to a Joint Venture named Framatome Advanced Nuclear Power S.A.S., the issue of the necessary and of the optimal industrial structure for nuclear projects as a research reactor is, was discussed internally often and intensively. That discussion took place also in the other technical fields such as Services for NPPs but also in the field of interest here, i. e. Research Reactors. In summarizing the statements of this presentation one can about state that: Research Reactors are easier to build than NPPs, but not standardised; Research Reactors need a wide spectrum of skills and experiences; to design and build Research Reactors needs an experienced team especially in terms of management and interfaces; Research Reactors need background from built reference plants more than from operating plants; Research Reactors need knowledge of suitable experienced subsuppliers. Two more essential conclusions as industry involved in constructing and upgrading research reactors are: Research Reactors by far are more than a suitable core that generates a high neutron flux; every institution that designs and builds a Research Reactor lacks quality or causes safety problems, damages the reputation of the entire community

  19. Natural circulating passive cooling system for nuclear reactor containment structure

    Science.gov (United States)

    Gou, Perng-Fei; Wade, Gentry E.

    1990-01-01

    A passive cooling system for the contaminant structure of a nuclear reactor plant providing protection against overpressure within the containment attributable to inadvertent leakage or rupture of the system components. The cooling system utilizes natural convection for transferring heat imbalances and enables the discharge of irradiation free thermal energy to the atmosphere for heat disposal from the system.

  20. Passive cooling system for nuclear reactor containment structure

    Science.gov (United States)

    Gou, Perng-Fei; Wade, Gentry E.

    1989-01-01

    A passive cooling system for the contaminant structure of a nuclear reactor plant providing protection against overpressure within the containment attributable to inadvertent leakage or rupture of the system components. The cooling system utilizes natural convection for transferring heat imbalances and enables the discharge of irradiation free thermal energy to the atmosphere for heat disposal from the system.

  1. An Assessment of Remote Visual Methods to Detect Cracking in Reactor Components

    International Nuclear Information System (INIS)

    Cumblidge, Stephen E.; Anderson, Michael T.; Doctor, Steven R.; Simonen, Fredric A.; Elliot, Anthony J.

    2008-01-01

    Recently, the U.S. nuclear industry has proposed replacing current volumetric and/or surface examinations of certain components in commercial nuclear power plants, as required by the American Society of Mechanical Engineers Boiler and Pressure Vessel Code Section XI, Inservice Inspection of Nuclear Power Plant Components, with a simpler visual testing (VT) method. The advantages of VT are that these tests generally involve much less radiation exposure and time to perform the examination than do volumetric examinations such as ultrasonic testing. The issues relative to the reliability of VT in determining the structural integrity of reactor components were examined. Some piping and pressure vessel components in a nuclear power station are examined using VT as they are either in high radiation fields or component geometry precludes the use of ultrasonic testing (UT) methodology. Remote VT with radiation-hardened video systems has been used by nuclear utilities to find cracks in pressure vessel cladding in pressurized water reactors, core shrouds in boiling water reactors, and to investigate leaks in piping and reactor components. These visual tests are performed using a wide variety of procedures and equipment. The techniques for remote VT use submersible closed-circuit video cameras to examine reactor components and welds. PNNL conducted a parametric study that examined the important variables influencing the effectiveness of a remote visual test. Tested variables included lighting techniques, camera resolution, camera movement, and magnification. PNNL also conducted a limited laboratory test using a commercial visual testing camera system to experimentally determine the ability of the camera system to detect cracks of various widths under ideal conditions. The results of these studies and their implications are presented in this paper

  2. Laser-Based Maintenance and Repair Technologies for Reactor Components

    International Nuclear Information System (INIS)

    Masaki Yoda; Naruhiko Mukai; Makoto Ochiai; Masataka Tamura; Satoshi Okada; Katsuhiko Sato; Motohiko Kimura; Yuji Sano; Noboru Saito; Seishi Shima; Tetsuo Yamamoto

    2004-01-01

    Toshiba has developed various laser-based maintenance and repair technologies and applied them to existing nuclear power plants. Laser-based technology is considered to be the best tool for remote processing in nuclear power plants, and particularly so for the maintenance and repair of reactor core components. Accessibility could be drastically improved by a simple handling system owing to the absence of reactive force against laser irradiation and the flexible optical fiber. For the preventive maintenance, laser peening (LP) technology was developed and applied to reactor components in operating BWR plants. LP is a novel process to improve residual stress from tensile to compressive on material surface layer by irradiating focused high-power laser pulses in water. We have developed a fiber-delivered LP system as a preventive maintenance measure against stress corrosion cracking (SCC). Laser ultrasonic testing (LUT) has a great potential to be applied to the remote inspection of reactor components. Laser-induced surface acoustic wave (SAW) inspection system was developed using a compact probe with a multi-mode optical fiber and an interferometer. The developed system successfully detected a micro slit of 0.5 mm depth on weld metal and heat-affected zone (HAZ). An artificial SCC was also detected by the system. We are developing a new LP system combined with LUT to treat the inner surface of bottom-mounted instruments (BMI) of PWR plants. Underwater laser seal welding (LSW) technology was also developed to apply surface crack. LSW is expected to isolate the crack tip from corrosive water environment and to stop the propagation of the crack. Rapid heating and cooling of the process minimize the heat effect, which extends the applicability to neutron-irradiated material. This paper describes recent advances in the development and application of such laser-based technologies. (authors)

  3. Component failures at pressurized water reactors. Final report

    International Nuclear Information System (INIS)

    Reisinger, M.F.

    1980-10-01

    Objectives of this study were to identify those systems having major impact on safety and availability (i.e. to identify those systems and components whose failures have historically caused the greatest number of challenges to the reactor protective systems and which have resulted in greatest loss of electric generation time). These problems were identified for engineering solutions and recommendations made for areas and programs where research and development should be concentrated. The program was conducted in three major phases: Data Analysis, Engineering Evaluation, Cost Benefit Analysis

  4. NSSS Component Control System Design of Integral Reactor

    International Nuclear Information System (INIS)

    Lee, Joon Koo; Kwon, Ho Je; Jeong, Kwong Il; Park, Heui Youn; Koo, In Soo

    2005-01-01

    MMIS(Man Machine Interface System) of an integral reactor is composed of a Control Room, Plant Protection System, Control System and Monitoring System which are related with the overall plant operation. MMIS is being developed with a new design concept and digital technology to reduce the Human Factor Error and improve the systems' safety, reliability and availability. And CCS(component control system) is also being developed with a new design concept and digital hardware technology A fully digitalized system and design concept are introduced in the NSSS CCS

  5. Aging management of major light water reactor components

    International Nuclear Information System (INIS)

    Shah, V.N.; Sinha, U.P.; Ware, A.G.

    1992-01-01

    Review of technical literature and field experience has identified stress corrosion cracking as one of the major degradation mechanisms for the major light water reactor components. Three of the stress corrosion cracking mechanisms of current concern are (a) primary water stress corrosion cracking (PWSCC) in pressurized water reactors, and (b) intergranular stress corrosion cracking (IGSCC) and (c) irradiation-assisted stress corrosion cracking (IASCC) in boiling water reactors. Effective aging management of stress corrosion cracking mechanisms includes evaluation of interactions between design, materials, stressors, and environment; identification and ranking of susceptible sites; reliable inspection of any damage; assessment of damage rate; mitigation of damage; and repair and replacement using corrosion-resistant materials. Management of PWSCC includes use of lower operating temperatures, reduction in residual tensile stresses, development of reliable inspection techniques, and use of Alloy 690 as replacement material. Management of IGSCC of nozzle and attachment welds includes use of Alloy 82 as weld material, and potential use of hydrogen water chemistry. Management of IASCC also includes potential use of hydrogen water chemistry

  6. Dynamic loads on reactor vessel components by low pressure waves

    International Nuclear Information System (INIS)

    Benkert, J.; Mika, C.; Stegemann, D.; Valero, M.

    1978-01-01

    Starting from the conservation theorems for mass and impulses the code DRUWE has been developed enabling the calculation of dynamic loads of the reactor shell on the basis of simplified assumptions for the first period shortly after rupture. According to the RSK-guidelines it can be assumed that the whole weld size is opened within 15 msec. This time-dependent opening of the fractured plane can be taken into account in the computer program. The calculation is composed in a way that for a reactor shell devided into cross and angle sections the local, chronological pressure and strength curves, the total dynamic load as well as the moments acting on the fastenings of the reactor shell can be calculated. As input data only geometrical details concerning the concept of the pressure vessel and its components as well as the effective subcooling of the fluid are needed. By means of several parameters the program can be operated in a way that the results are available in form of listings or diagrams, respectively, but also as card pile for further examinations, e.g. strength analysis. (orig./RW) [de

  7. Development of guidelines for inelastic analysis in design of fast reactor components

    International Nuclear Information System (INIS)

    Nakamura, Kyotada; Kasahara, Naoto; Morishita, Masaki; Shibamoto, Hiroshi; Inoue, Kazuhiko; Nakayama, Yasunari

    2008-01-01

    The interim guidelines for the application of inelastic analysis to design of fast reactor components were developed. These guidelines are referred from 'Elevated Temperature Structural Design Guide for Commercialized Fast Reactor (FDS)'. The basic policies of the guidelines are more rational predictions compared with elastic analysis approach and a guarantee of conservative results for design conditions. The guidelines recommend two kinds of constitutive equations to estimate strains conservatively. They also provide the methods for modeling load histories and estimating fatigue and creep damage based on the results of inelastic analysis. The guidelines were applied to typical design examples and their results were summarized as exemplars to support users

  8. Proceedings [of the] symposium on zirconium alloys for reactor components

    International Nuclear Information System (INIS)

    1992-01-01

    A two day symposium on zirconium alloys for reactor components (ZARC-91) was organised during 12-13, 1991. There were 6 invited talks and 43 contributed papers in 10 technical sessions. This symposium, took stock of the progress achieved in the development, design, fabrication and quality assurance of zirconium alloy components and emphasized the R and D efforts required for meeting the challenges posed by the rapid growth of nuclear power in our country. Topics like physical metallurgy, corrosion and irradiation behaviour, and in-service inspection were also covered. The proceedings/papers are arranged under the headings: (1)invited talks, (2)fabrication, (3)design requirement, (4)quality assurance, (5)irradiation damage and PIE, (6)corrosion and hydriding, and (7)in-service inspection. (N.B.). refs., figs., tabs

  9. Review of leakage-flow-induced vibrations of reactor components

    International Nuclear Information System (INIS)

    Mulcahy, T.M.

    1983-05-01

    The primary-coolant flow paths of a reactor system are usually subject to close scrutiny in a design review to identify potential flow-induced vibration sources. However, secondary-flow paths through narrow gaps in component supports, which parallel the primary-flow path, occasionally are the excitation source for significant vibrations even though the secondary-flow rates are orders of magnitude smaller than the primary-flow rate. These so-called leakage flow problems are reviewed here to identify design features and excitation sources that should be avoided. Also, design rules of thumb are formulated that can be employed to guide a design, but quantitative prediction of component response is found to require scale-model testing

  10. Evaluation of aging degradation of structural components

    International Nuclear Information System (INIS)

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

    1992-03-01

    Irradiation embrittlement of the neutron shield tank (NST) A212 Grade B steel from the Shippingport reactor, as well as thermal embrittlement of CF-8 cast stainless steel components from the Shippingport and KRB reactors, has been characterized. Increases in Charpy transition temperature (CTT), yield stress, and hardness of the NST material in the low-temperature low-flux environment are consistent with the test reactor data for irradiations at 8 n/cm 2 ·s at the low operating temperature of the Shippingport NST, i.e., 55 degrees C. This suggest that radiation damage in Shippingport NST and HFIR surveillance samples may be different because of the neutron spectra and/or Cu and Ni content of the two materials. Cast stainless steel components show relatively modest decreases in fracture toughness and Charpy-impact properties and a small increase in tensile strength. Correlations for estimating mechanical properties of cast stainless steels predict accurate or slightly conservative values for Charpy-impact energy, tensile flow stress, fracture toughness J-R curve, and J IC of the materials. The kinetics of thermal embrittlement and degree of embrittlement at saturation, i.e., the minimum impact energy achieved after long-term aging, were established from materials that were aged further in the laboratory. The results were consistent with the estimates. The correlations successfully predict the mechanical properties of the Ringhals 2 reactor hot- and crossover-leg elbows (CF-8M steel) after service of ∼15 y

  11. Draft of standard for graphite core components in high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Shibata, Taiju; Sawa, Kazuhiro; Eto, Motokuni; Kunimoto, Eiji; Shiozawa, Shusaku; Oku, Tatsuo; Maruyama, Tadashi

    2010-01-01

    For the design of the graphite components in the High Temperature Engineering Test Reactor (HTTR), the graphite structural design code for the HTTR etc. were applied. However, general standard systems for the High Temperature Gas-cooled Reactor (HTGR) have not been established yet. The authors had studied on the technical issues which is necessary for the establishment of a general standard system for the graphite components in the HTGR. The results of the study were documented and discussed at a 'Special committee on research on preparation for codes for graphite components in HTGR' at Atomic Energy Society of Japan (AESJ). As a result, 'Draft of Standard for Graphite Core Components in High Temperature Gas-cooled Reactor.' was established. In the draft standard, the graphite components are classified three categories (A, B and C) in the standpoints of safety functions and possibility of replacement. For the components in the each class, design standard, material and product standards, and in-service inspection and maintenance standard are determined. As an appendix of the design standard, the graphical expressions of material property data of 1G-110 graphite as a function of fast neutron fluence are expressed. The graphical expressions were determined through the interpolation and extrapolation of the irradiated data. (author)

  12. Managerial improvement efforts after finding unreported cracks in reactor components

    International Nuclear Information System (INIS)

    Kawamura, S.

    2006-01-01

    In 2002 TEPCO found that there were unreported cracks in reactor components, of which inspection records had been falsified. Stress Corrosion Cracking indications found in Core Shrouds and Primary Loop Re-circulation pipes at some plants were removed from the inspection records and not reported to the regulators. Top management of TEPCO took the responsibility and resigned, and recovery was started under the leadership of new management team. First of all, behavioral standards were reconstituted to strongly support safety-first value. Ethics education was introduced and corporate ethics committee was organized with participation of external experts. Independent assessment organization was established to enhance quality assurance. Information became more transparent through Non-conformance Control Program. As for the material management, prevention and mitigation programs for the Stress Corrosion Cracking of reactor components were re-established. In addition to the above immediate recovery actions, long term improvement initiatives have also been launched and driven by our aspiration to excellence in safe operation of nuclear power plants. Vision and core values were set to align the people. Organizational learning was enhanced by benchmark studies, better systematic use of operational experience, self-assessment and external assessment. Based on these foundation blocks and with strong sponsorship from the top management, work processes were analyzed and improved by Peer Groups. (author)

  13. Optimization of structures and components

    CERN Document Server

    Muñoz-Rojas, Pablo Andrés

    2013-01-01

    Written by an international group of active researchers in the field, this volume presents innovative formulations and applied procedures for sensitivity analysis and structural design optimization. Eight chapters discuss subjects ranging from recent developments in the determination and application of topological gradients, to the use of evolutionary algorithms and meta-models to solve practical engineering problems. With such a comprehensive set of contributions, the book is a valuable source of information for graduate students and researchers entering or working in the matter.

  14. Vibration-damping structure for reactor building

    International Nuclear Information System (INIS)

    Kuno, Toshio; Iba, Chikara; Tanaka, Hideki; Kageyama, Mitsuru

    1998-01-01

    In a damping structure of a reactor building, an inner concrete body and a reactor container are connected by way of a vibration absorbing member. As the vibration absorbing member, springs or dampers are used. The inner concrete body and the reactor container each having weight and inherent frequency different from each other are opposed displaceably by way of the vibration absorbing member thereby enabling to reduce seismic input and reduce shearing force at least at leg portions. Accordingly, seismic loads are reduced to increase the grounding rate of the base thereby enabling to satisfy an allowable value. Therefore, it is not necessary to strengthen the inner concrete body and the reactor container excessively, the amount of reinforcing rods can be reduced, and the amount of a portion of the base buried to the ground can be reduced thereby enabling to constitute the reactor building easily. (N.H.)

  15. Assessment of structural materials inside the reactor pool of the Dalat research reactor

    International Nuclear Information System (INIS)

    Nguyen Nhi Dien; Luong Ba Vien; Nguyen Minh Tuan; Trang Cao Su

    2010-01-01

    Originally the Dalat Nuclear Research Reactor (DNRR) was a 250-kW TRIGA MARK II reactor, started building from early 1960s and achieved the first criticality on February 26, 1963. During the 1982-1984 period, the reactor was reconstructed and upgraded to 500kW, and restarted operation on March 20, 1984. From the original TRIGA reactor, only the pool liner, beam ports, thermal columns, and graphite reflector have been remained. The structural materials of pool liner and other components of TRIGA were made of aluminum alloy 6061 and aluminum cladding fuel assemblies. Some other parts, such as reactor core, irradiation rotary rack around the core, vertical irradiation facilities, etc. were replaced by the former Soviet Union's design with structural materials of aluminum alloy CAV-1. WWR-M2 fuel assemblies of U-Al alloy 36% and 19.75% 235 U enrichment and aluminum cladding have been used. In its original version, the reactor was setting upon an all-welded aluminum frame supported by four legs attached to the bottom of the pool. After the modification made, the new core is now suspended from the top of the pool liner by means of three aluminum concentric cylindrical shells. The upper one has a diameter of 1.9m, a length of 3.5m and a thickness of 10mm. This shell prevents from any visual access to the upper part of the pool liner, but is provided with some holes to facilitate water circulation in the 4cm gap between itself and the reactor pool liner. The lower cylindrical shells act as an extracting well for water circulation. As reactor has been operated at low power of 500 kW, it was no any problem with degradation of core structural materials due to neutron irradiation and thermal heat, but there are some ageing issues with aluminum liner and other structures (for example, corrosion of tightening-up steel bolt in the fourth beam port and flood of neutron detector housing) inside the reactor pool. In this report, the authors give an overview and assessment of

  16. Seismic qualification of safety class components in non-reactor nuclear facilities at Hanford site

    International Nuclear Information System (INIS)

    Ocoma, E.C.

    1989-01-01

    This paper presents the methods used during the walkdowns to compile as-built structural information to seismically qualify or verify the seismic adequacy of safety class components in the Plutonium Finishing Plant complex. The Plutonium finishing Plant is a non-reactor nuclear facility built during the 1950's and was designed to the Uniform Building Code criteria for both seismic and wind events. This facility is located at the US Department of Energy Hanford Site near Richland, Washington

  17. Life management for a non replaceable structure: the reactor building

    International Nuclear Information System (INIS)

    Torres, V.; Francia, L.

    1998-01-01

    Phase 1 of UNESA N.P.P. Lifetime Management Project identified and ranked important components, relative to plant life management. The list showed the Reactor Containment Structure in the third position, and thirteen concrete structures were among the list top twenty. Since the Reactor Containment Building, together with the Reactor Vessel, is the only non-replaceable plant component, and has a big impact on the plant technical life, there is an increasing interest on understanding its behavior to maintain structural integrity. This paper presents: a) IAEA (International Atomic Energy Agency) Coordinated Research Program experiences and studies. Under this Program, international experts address the most frequent degradation mechanisms affecting the containment building. b) IAEA Aging Management Program adapted to our plants. The paper addresses the aging mechanisms affecting the concrete structures, reinforcing steel and prestress systems as well as the aging management programs and the mitigation and control methods. Finally, this paper presents a new module called STRUCTURES, included in phase 2 of the above mentioned project, which will monitor and document the different aging mechanisms and management programs described in item b) regarding the Reactor Containment Building (concrete liner, post stressing system, anchor elements). This module will also support the Maintenance Rule related practices. (Author)

  18. On elastic structural elements for nuclear reactors

    International Nuclear Information System (INIS)

    Povolo, F.

    1978-03-01

    The in-pile stress-relaxation behaviour of materials usually employed for the elastic structural elements, in nuclear reactors, is critically reviewed and the results are compared with those obtained in commercial zirconium alloys irradiated under similar conditions. Finally, it is shown that, under certain conditions, some zirconium alloys may be used as an alternative material for these structural elements. (orig.) [de

  19. Component vibration of VVER-reactors - diagnostics and modelling

    International Nuclear Information System (INIS)

    Altstadt, E.; Scheffler, M.; Weiss, F.-P.

    1995-01-01

    Flow induced vibrations of reactor pressure vessel (RPV) internals (control element and core barrel motions) at VVER-440 reactors have led to the development of dedicated methods for on-line monitoring. These methods need a certain developed stage of the faults to be detected. To achieve a real sensitive early detection of mechanical faults of RPV internals, a theoretical vibration model was developed based on finite elements. The model comprises the whole primary circuit including the steam generators (SG). By means of that model all eigenfrequencies up to 30 Hz and the corresponding mode shapes were calculated for the normal vibration behaviour. Moreover the shift of eigenfrequencies and of amplitudes due to the degradation or to the failure of internal clamping and spring elements could be investigated, showing that a recognition of such degradations even inside the RPV is possible by pure excore vibration measurements. A true diagnostic, that is the identification of the failed component, might become possible because different faults influence different and well separated eigenfrequencies. (author)

  20. Some applications of capacitance technology in nuclear reactor components inspections

    International Nuclear Information System (INIS)

    Walton, H.

    1985-01-01

    The paper considers application of a capacitance measuring system that has overcome many of the original contraints, such as sensitivity to cable length, induced electric field and high acoustic noise, and illustrates the ease of use with examples of proven capability in severe environments of high temperature or high radiation. The Capacitance Displacement Transducer (CDT) measuring principle was originally developed as a working technique during the early years of full-scale, on-load refuelling trials performed in the Windscale Civil Advanced Gas-Cooled Reactor (CAGR) test rig where it was necessary to measure the vibrational behaviour of fuel components in simulated reactor conditions. At that time, 1968-1969, no instrumentation existed that would measure displacement in the range 0 to 100 mms to an accuracy of 25x10 -3 mms, without physical contact, at temperatures of 600 0 C in high velocity gas, in high acoustic noise fields of 150 db's over cable lengths approaching 100 metres. The principles incorporated in the CDT overcome all these problems. The advantages inherent in this system have been extended to metrology applications in more recent years by the further development of the electronics to enable linear displacement measurement to be obtained between two capacitance plates whose separation varies, either by plate movement or by surface irregularity. This principle has been used to good effect in novel applications associated with the inspection of nominally inaccessible internal tube surfaces

  1. Computation of the mechanical behaviour of nuclear reactor components

    International Nuclear Information System (INIS)

    Brosi, S.; Niffenegger, M.; Roesel, R.; Reichlin, K.; Duijvestijn, A.

    1994-01-01

    A possible limiting factor of the service life of a reactor is the mechanical load carrying margin, i.e. the excess of the load carrying capacity over the actual loading, of the central, heavy section components. This margin decreases during service but, for safety reasons, may not fall below a critical value. Therefore, it is essential to check and to control continuously the factors which cause the decrease. The reasons for the decrease are shown at length and in detail in an example relating to the test which almost achieved failure of a pipe emanating from a reactor pressure vessel, weakened by an artificial crack and undergoing a water-hammer loading. The latter was caused by a sudden valve closure supposed to follow upon a break far downstream. The computational and experimental difficulties associated with the simultaneous occurrence of an extreme weakening and an extreme loading in an already rather complicated geometry are explained. It is concluded that available computational tools and present know-how are sufficient to simulate the behaviour under such conditions as would prevail in normal service, and even to analyse departures from them, as long as not all difficulties arise simultaneously. (author) figs., tabs., refs

  2. Canisters and nonfuel components at commercial nuclear reactors

    International Nuclear Information System (INIS)

    Gibbard, K.; Disbrow, J.

    1994-01-01

    This paper discusses detailed data on canisters and nonfuel components (NFC) at US commercial nuclear power reactors. A wide variety of NFC have been reported on the Form RW-859, open-quotes Nuclear Fuel Dataclose quotes survey. They may have been integral with an assembly, noncanistered in baskets, destined for disposal as low-level radioactive waste, or stored in canisters. Similarly, data on the family of canistered spent nuclear fuel (SNF) in storage pools was compiled. Approximately 85 percent of the 40,194 pieces of nonfuel assembly (NFA) hardware reported were integral with an assembly. This represents data submitted by 95 of the 107 reactors in 10 generic assembly classes. In addition, a total of 286 canisters have been reported as being in storage pools as of December 31, 1992. However, an additional 264 open baskets were also reported to contain miscellaneous SNF and nonfuel materials, garbage and debris. All of these 286 canisters meet the dimensional envelope requirements specified for disposal for open-quotes standard fuelclose quotes under the Standard Contract for Disposal of Spent Nuclear Fuel and/or High-Level Radioactive Waste (10 CFR 961); most of the baskets do not

  3. Preloading of bolted connections in nuclear reactor component supports

    Energy Technology Data Exchange (ETDEWEB)

    Yahr, G T

    1984-10-01

    A number of failures of threaded fasteners in nuclear reactor component supports have been reported. Many of those failures were attributed to stress corrosion cracking. This report discusses how stress corrosion cracking can be avoided in bolting by controlling the maximum bolt preloads so that the sustained stresses in the bolts are below the level required to cause stress corrosion cracking. This is a basic departure from ordinary bolted joint design where the only limits on preload are on the minimum preload. Emphasis is placed on the importance of detailed analysis to determine the acceptable range of preload and the selection of a method for measuring the preload that is sufficiently accurate to ensure that the preload is actually within the acceptable range. Procedures for determining acceptable preload range are given, and the accuracy of various methods of measuring preload is discussed.

  4. Preloading of bolted connections in nuclear reactor component supports

    International Nuclear Information System (INIS)

    Yahr, G.T.

    1984-10-01

    A number of failures of threaded fasteners in nuclear reactor component supports have been reported. Many of those failures were attributed to stress corrosion cracking. This report discusses how stress corrosion cracking can be avoided in bolting by controlling the maximum bolt preloads so that the sustained stresses in the bolts are below the level required to cause stress corrosion cracking. This is a basic departure from ordinary bolted joint design where the only limits on preload are on the minimum preload. Emphasis is placed on the importance of detailed analysis to determine the acceptable range of preload and the selection of a method for measuring the preload that is sufficiently accurate to ensure that the preload is actually within the acceptable range. Procedures for determining acceptable preload range are given, and the accuracy of various methods of measuring preload is discussed

  5. Locking mechanism for in-vessel components of tokamak reactor

    International Nuclear Information System (INIS)

    Nishio, S.; Shimizu, K.; Koizumi, K.; Tada, E.

    1992-01-01

    The locking and unlocking mechanism for in-vessel replaceable components such as blanket modules, is one of the most critical issues of the tokamak fusion reactor, since the sufficient stiffness against the enormous electromagnetic loads and the easy replaceability are required. In this paper, the authors decide that a caulking cotter joint is worth initiating the R and D from veiwpoints of an effective use of space, a replaceability, a removability of nuclear heating, and a reliability. In this approach, the cotter driving (thrusting and plucking) mechanism is a critical technology. A flexible tube concept has been developed as the driving mechanism, where the stroke and driving force are obtained by a fat shape by the hydraulic pressure. The original normal tube is subjected to the working percentage of more than several hundreds percentage (from thickness of 1.2 mm to 0.2 mm) for plastically forming the flexible tube

  6. International feedback experience on the cutting of reactor internal components

    International Nuclear Information System (INIS)

    Boucau, J.

    2014-01-01

    Westinghouse capitalizes more than 30 years of experience in the cutting of internal components of reactor and their packaging in view of their storage. Westinghouse has developed and validated different methods for cutting: plasma torch cutting, high pressure abrasive water jet cutting, electric discharge cutting and mechanical cutting. A long feedback experience has enabled Westinghouse to list the pros and cons of each cutting technology. The plasma torch cutting is fast but rises dosimetry concerns linked to the control of the cuttings and the clarity of water. Abrasive water jet cutting requires the installation of costly safety devices and of an equipment for filtering water but this technology allows accurate cuttings in hard-to-reach zones. Mechanical cutting is the most favourable technology in terms of wastes generation and of the clarity of water but the cutting speed is low. (A.C.)

  7. Electrochemical machining - manufacturing of turbine and reactor components

    International Nuclear Information System (INIS)

    Otto, K.

    1987-01-01

    Electrochemical machining is a shaping process for metallic workpieces with complex geometries. Using an electrode corresponding to the negative of the desired shape, the material to be removed is dissolved anodically at erosion rates of up to 10 mm/min. The reproducible shape accuracy lies between 0,02 and 0,08 mm, depending on the machining problem. Surface finishes of less than 18 μm are attained. The hardness of the material has no influence on the metal removal process. The workpiece is not subjected to any thermal stressing during machining. The process is well suited for quantity production of complex parts and is used inter alia for turbine blades and components for nuclear reactors. (orig.) [de

  8. High flux testing reactor Petten. Replacement of the reactor vessel and connected components. Overall report

    International Nuclear Information System (INIS)

    Chrysochoides, N.G.; Cundy, M.R.; Von der Hardt, P.; Husmann, K.; Swanenburg de Veye, R.J.; Tas, A.

    1985-01-01

    The project of replacing the HFR originated in 1974 when results of several research programmes confirmed severe neutron embrittlement of aluminium alloys suggesting a limited life of the existing facility. This report contains the detailed chronology of events concerning preparation and execution of the replacement. After a 14 months' outage the reactor resumed routine operation on 14th February, 1985. At the end of several years of planning and preparation the reconstruction proceded in the following steps: unloading of the old core, decay of short-lived radioactivity in December 1983, removal of the old tank and of its peripheral equipment in January-February 1984, segmentation and waste disposal of the removed components in March-April, decontamination of the pools, bottom penetration overhauling in May-June, installation of the new tank and other new components in July-September, testing and commissioning, including minor modifications in October-December, and, trials runs and start-up preparation in January-February 1985. The new HFR Petten features increased and improved experimental facilities. Among others the obsolete thermal columns was replaced by two high flux beam tubes. Moreover the new plant has been designed for future increases of reactor power and neutron fluxes. For the next three to four years the reactor has to cope with a large irradiation programme, claiming its capacity to nearly 100%

  9. Material Challenges For Plasma Facing Components in Future Fusion Reactors

    International Nuclear Information System (INIS)

    Linke, J; Pintsuk, G.; Rödig, M.

    2013-01-01

    Increasing attention is directed towards thermonuclear fusion as a possible future energy source. Major advantages of this energy conversion technology are the almost inexhaustible resources and the option to produce energy without CO2-emissions. However, in the most advanced field of magnetic plasma confinement a number of technological challenges have to be met. In particular high-temperature resistant and plasma compatible materials have to be developed and qualified which are able to withstand the extreme environments in a commercial thermonuclear power reactor. The plasma facing materials (PFMs) and components (PFCs) in such fusion devices, i.e. the first wall (FW), the limiters and the divertor, are strongly affected by the plasma wall interaction processes and the applied intense thermal loads during plasma operation. On the one hand, these mechanisms have a strong influence on the plasma performance; on the other hand, they have major impact on the lifetime of the plasma facing armour. In present-day and next step devices the resulting thermal steady state heat loads to the first wall remain below 1 MWm-2; the limiters and the divertor are expected to be exposed to power densities being at least one order of magnitude above the FW-level, i.e. up to 20 MWm-2 for next step tokamaks such as ITER or DEMO. These requirements are responsible for high demands on the selection of qualified PFMs and heat sink materials as well as reliable fabrication processes for actively cooled plasma facing components. The technical solutions which are considered today are mainly based on the PFMs beryllium, carbon or tungsten joined to copper alloys or stainless steel heat sinks. In addition to the above mentioned quasi-stationary heat loads, short transient thermal pulses with deposited energy densities up to several tens of MJm-2 are a serious concern for next step tokamak devices. The most frequent events are so-called Edge Localized Modes (type I ELMs) and plasma disruptions

  10. UK fast reactor components. Sodium removal decontamination and requalification

    International Nuclear Information System (INIS)

    Donaldson, D.M.; Bray, J.A.; Newson, I.H.

    1978-01-01

    Extensive experience gained at the U.K.A.E.A. Dounreay Nuclear Power Development Establishment is being applied to form the basis of the plant to be provided for sodium removal, decontamination, and requalification of components in future commercial fast reactors. In the first part of a three part paper, the factors to be taken into account, showing the UK philosophy and approach to maintenance and repair operations are discussed. In the second part, PFR facilities for sodium removal and decontamination are described and some examples are given of cleaning components such as pumps, charge machine, cold trap baskets, and steam generator units. Similar facilities at DFR are briefly described. In the third part of the paper a short description is given of the Harwell mass transfer loop, currently used to study the deposition of activated stainless steel corrosion products. Decontamination method for pipework specimens cut from the loop are described and results of first screening tests of various chemical decontaminants are presented. (U.K.)

  11. Structural integrity evaluation of PWR nuclear reactor pressure vessels

    International Nuclear Information System (INIS)

    Cruz, Julio R.B.; Mattar Neto, Miguel

    1999-01-01

    The reactor pressure vessel (RPV) is the most important structural component of a PWR nuclear power plant. It contains the reactor core and is the main component of the primary system pressure boundary, the system responsible for removing the heat generated by the nuclear reactions. It is considered not replaceable and, therefore, its lifetime is a key element to define the plant life as a whole. Three critical issues related to the reliability of the RPV structural integrity come out by reason of the radiation damage imposed to the vessel material during operation. These issues concern the definition of pressure versus temperature limits for reactor heatup and cooldown, pressurized thermal shock evaluation and assessment of reactor vessels with low upper shelf Charpy impact energy levels. This work aims to present the major aspects related to these topics. The requirements for preventing fracture of the RPV are reviewed as well as the available technology for assessing the safety margins. For each mentioned problem, the several steps for structural integrity evaluation are described and the analysis methods are discussed. (author)

  12. Structural mechanics of nuclear plant components

    International Nuclear Information System (INIS)

    Roche, R.

    1986-10-01

    Sound structural analysis are needed for designing safe and reliable components, hence his play is very important in nuclear industry. This report is a provisional writing on the good practice in structural mechanics. Emphasis is put on non elastic analysis, damage appraisal, fatigue, fracture mechanics and also on elevated temperature behaviour [fr

  13. Structural integrity monitoring of critical components in nuclear facilities

    International Nuclear Information System (INIS)

    Roth, Maria; Constantinescu, Dan Mihai; Brad, Sebastian; Ducu, Catalin; Malinovschi, Viorel

    2007-01-01

    Full text: The paper presents the results obtained as part of the Project 'Integrated Network for Structural Integrity Monitoring of Critical Components in Nuclear Facilities', RIMIS, a research work underway within the framework of the Ministry of Education and Research Programme 'Research of Excellence'. The main objective of the Project is to constitute a network integrating the national R and D institutes with preoccupations in the structural integrity assessment of critical components in the nuclear facilities operating in Romania, in order to elaborate a specific procedure for this field. The degradation mechanisms of the structural materials used in the CANDU type reactors, operated by Unit 1 and Unit 2 at Cernavoda (pressure tubes, fuel elements sheaths, steam generator tubing) and in the nuclear facilities relating to reactors of this type as, for instance, the Hydrogen Isotopes Separation facility, will be investigated. The development of a flexible procedure will offer the opportunity to extend the applications to other structural materials used in the nuclear field and in the non-nuclear fields as well, in cooperation with other institutes involved in the developed network. The expected results of the project will allow the integration of the network developed at national level in the structures of similar networks operating within the EU, the enhancement of the scientific importance of Romanian R and D organizations as well as the increase of our country's contribution in solving the major issues of the nuclear field. (authors)

  14. Shield structure for a nuclear reactor

    International Nuclear Information System (INIS)

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

    1979-01-01

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

  15. Activities in the Czech Republic for reactor pressure components lifetime management

    International Nuclear Information System (INIS)

    Brumovsky, M.

    1994-01-01

    The following activities in the Czech republic for reactor pressure components lifetime management are described: upgrading and safety assurance of nuclear power plants (NPP) with reactors of WWER-440/V-230 type, safety assurance of NPPs with reactors of WWER-440/V-213, lifetime management programme of NPPs with WWER-440/V-213 reactors, preparation of start-up of NPPs with WWER-1000/V-320 reactors, preparation of guides for lifetime as well as defect allowability evaluation in main components of primary and secondary circuits. 3 figs

  16. Structured Performance Analysis for Component Based Systems

    OpenAIRE

    Salmi , N.; Moreaux , Patrice; Ioualalen , M.

    2012-01-01

    International audience; The Component Based System (CBS) paradigm is now largely used to design software systems. In addition, performance and behavioural analysis remains a required step for the design and the construction of efficient systems. This is especially the case of CBS, which involve interconnected components running concurrent processes. % This paper proposes a compositional method for modeling and structured performance analysis of CBS. Modeling is based on Stochastic Well-formed...

  17. The Thermal-hydraulic Analysis for the Aging Effect of the Component in CANDU-6 Reactor

    International Nuclear Information System (INIS)

    Bae, Jun Ho; Jung, Jong Yeob

    2014-01-01

    CANDU reactor consists of a lot of components, including pressure tube, reactor pump, steam generator, feeder pipe, and so on. These components become to have the aging characteristics as the reactor operates for a long time. The aging phenomena of these components lead to the change of operating parameters, and it finally results to the decrease of the operating safety margin. Actually, due to the aging characteristics of components, CANDU reactor power plant has the operating license for the duration of 30 years and the plant regularly check the plant operating state in the overhaul period. As the reactor experiences the aging, the reactor operators should reduce the reactor power level in order to keep the minimum safety margin, and it results to the deficit of economical profit. Therefore, in order to establish the safety margin for the aged reactor, the aging characteristics for components should be analyzed and the effect of aging of components on the operating parameter should be studied. In this study, the aging characteristics of components are analyzed and revealed how the aging of components affects to the operating parameter by using NUCIRC code. Finally, by scrutinizing the effect of operating parameter on the operating safety margin, the effect of aging of components on the safety margin has been revealed

  18. The adhesive bonding of beryllium structural components

    International Nuclear Information System (INIS)

    Fullerton-Batten, R.C.

    1977-01-01

    Where service conditions permit, adhesive bonding is a highly recommendable, reliable means of joining beryllium structural parts. Several important programs have successfully used adhesive bonding for joining structural and non-structural beryllium components. Adhesive bonding minimizes stress concentrations associated with other joining techniques and considerably improves fatigue resistance. In addition, no degradation of base metal properties occur. In many instances, structural joints can be fabricated more cheaply by adhesive bonding or in combination with adhesive bonding than by any other method used alone. An evaluation program on structural adhesive bonding of beryllium sheet components is described. A suitable surface pretreatment for beryllium adherends prior to bonding is given. Tensile shear strength and fatigue properties of FM 1000 and FM 123-5 adhesive bonded joints are reviewed and compared with data obtained from riveted joints of similar geometry. (author)

  19. System, structure, and component evaluation for life-cycle management

    International Nuclear Information System (INIS)

    Hanley, N.E.; Banerjee, A.K.; Woods, P.B.; Perrin, J.S.; Marian, F.A.

    1992-01-01

    In recent years, many nuclear organizations and utilities have studied the possibility of extending the service life of nuclear power plants beyond the original license period. From these studies, recommendations have resulted for maintaining the option of future decisions concerning operating license renewal. Several of the recommendations are considered beneficial to the management and operation of nuclear plants in meeting many of their near-term goals. In 1986, Public Service Electric and Gas (PSE and G) concluded that a full-scale nuclear plant license renewal program for their Salem 1 and 2 and Hope Creek nuclear stations was not cost-effective at that time. Rather, it would be better served if the nuclear plant life extension (PLEX) option were maintained for future consideration. To help plan for the life extension option, a strategic 5-yr life cycle management (LCM) program was begun. In support of the LCM program, evaluations for the following Salem structures and components were performed: (1) intake structures, (2) reactor vessel support, (3) containment liner, and (4) containment structure (below grade). This paper discusses the systems, structures, and components (SSC) evaluation methodology and, as an example, discusses the evaluation performed for reactor vessel support

  20. Generalized structured component analysis a component-based approach to structural equation modeling

    CERN Document Server

    Hwang, Heungsun

    2014-01-01

    Winner of the 2015 Sugiyama Meiko Award (Publication Award) of the Behaviormetric Society of Japan Developed by the authors, generalized structured component analysis is an alternative to two longstanding approaches to structural equation modeling: covariance structure analysis and partial least squares path modeling. Generalized structured component analysis allows researchers to evaluate the adequacy of a model as a whole, compare a model to alternative specifications, and conduct complex analyses in a straightforward manner. Generalized Structured Component Analysis: A Component-Based Approach to Structural Equation Modeling provides a detailed account of this novel statistical methodology and its various extensions. The authors present the theoretical underpinnings of generalized structured component analysis and demonstrate how it can be applied to various empirical examples. The book enables quantitative methodologists, applied researchers, and practitioners to grasp the basic concepts behind this new a...

  1. Welding of structural components and vessels

    International Nuclear Information System (INIS)

    1989-01-01

    'Welding of structural components and vessels' was chosen as the guiding topic for the 17th special conference in Munich so that current problems of this important area of application for welding engineering could be discussed in detail. The following topics were in the focus of the discussions: developments in steel, steel production and steel processing, reports on the practical application of welding in the manufacture of containers and pipes, quality assurance, product liability, safety considerations regarding creep-stressed components, problems of welding in large structures. 7 of the total number of 12 contributions were recorded separately for the data base ENERGY. (orig./MM) [de

  2. Data base formation for important components of reactor TRIGA MARK II

    International Nuclear Information System (INIS)

    Jordan, R.; Mavko, B.; Kozuh, M.

    1992-01-01

    The paper represents specific data base formation for reactor TRIGA MARK II in Podgorica. Reactor operation data from year 1985 to 1990 were collected. Two groups of collected data were formed. The first group includes components data and the second group covers data of reactor scrams. Time related and demand related models were used for data evaluation. Parameters were estimated by classical method. Similar data bases are useful everywhere where components unavailabilities may have severe drawback. (author) [sl

  3. GENERIC, COMPONENT FAILURE DATA BASE FOR LIGHT WATER AND LIQUID SODIUM REACTOR PRAs

    Energy Technology Data Exchange (ETDEWEB)

    S. A. Eide; S. V. Chmielewski; T. D. Swantz

    1990-02-01

    A comprehensive generic component failure data base has been developed for light water and liquid sodium reactor probabilistic risk assessments (PRAs) . The Nuclear Computerized Library for Assessing Reactor Reliability (NUCLARR) and the Centralized Reliability Data Organization (CREDO) data bases were used to generate component failure rates . Using this approach, most of the failure rates are based on actual plant data rather than existing estimates .

  4. Thermodynamic characterization of salt components for Molten Salt Reactor fuel

    NARCIS (Netherlands)

    Capelli, E.

    2016-01-01

    The Molten Salt Reactor (MSR) is a promising future nuclear fission reactor technology with excellent performance in terms of safety and reliability, sustainability, proliferation resistance and economics. For the design and safety assessment of this concept, it is extremely important to have a

  5. A discussion about simplified methodologies for failure assessment of nuclear reactor components

    International Nuclear Information System (INIS)

    Cruz, J.R.B.; Andrade, A.H.P. de; Landes, J.D.

    1996-01-01

    Failure of nuclear reactor components like pressure vessels and piping must be avoided for all phases of reactor operation. Especially severe loading conditions come from postulated accident scenarios during which the integrity of the component is required. The use of Fracture Mechanics concepts to investigate the mechanical behavior of flawed structures in the non-linear regime is a complex subject due to the fact that the crack driving force (expressed in terms of J or CTOD) is not /only a function of the cracked geometry, but depends also on the plastic flow properties of the material. Since the numerical solutions by the finite element method are expensive and time consuming, the existence of simplified engineering procedures is of great relevance. These allow a ready identification of the main parameters affecting the crack driving force, and permit a fast and simple evaluation of the structural integrity of the cracked component. This paper presents an overview of the major simplified ductile fracture methodologies that have been proposed in the literature trying to point out their similarities, strong points and negative aspects. Once the best characteristics of each method are identified, they could then be combined to develop a single methodology, one that would be both easy to use and capable of making accurate failure predictions

  6. Vibration of fusion reactor components with magnetic damping

    Energy Technology Data Exchange (ETDEWEB)

    D’Amico, Gabriele; Portone, Alfredo [Fusion for Energy – Torres Diagonal Litoral B3 – c/Josep Plá n.2, Barcelona (Spain); Rubinacci, Guglielmo [Department of Electrical Eng. and Information Technologies, Università di Napoli Federico II, Via Claudio, 21, 80125 Napoli (Italy); Testoni, Pietro, E-mail: pietro.testoni@f4e.europa.eu [Fusion for Energy – Torres Diagonal Litoral B3 – c/Josep Plá n.2, Barcelona (Spain)

    2016-11-01

    The aim of this paper is to assess the importance of the magnetic damping in the dynamic response of the main plasma facing components of fusion machines, under the strong Lorentz forces due to Vertical Displacement Events. The additional eddy currents due to the vibration of the conducting structures give rise to volume loads acting as damping forces, a kind of viscous damping, being these additional loads proportional to the vibration speed. This effect could play an important role when assessing, for instance, the inertial loads associated to VV movements in case of VDEs. In this paper, we present the results of a novel numerical formulation, in which the field equations are solved by adopting a very effective fully 3D integral formulation, not limited to the analysis of thin shell structures, as already successfully done in several approaches previously published.

  7. Space-time reactor kinetics for heterogeneous reactor structure

    Energy Technology Data Exchange (ETDEWEB)

    Raisic, N [Boris Kidric Institute of nuclear sciences Vinca, Belgrade (Yugoslavia)

    1969-11-15

    An attempt is made to formulate time dependent diffusion equation based on Feinberg-Galanin theory in the from analogue to the classical reactor kinetic equation. Parameters of these equations could be calculated using the existing codes for static reactor calculation based on the heterogeneous reactor theory. The obtained kinetic equation could be analogues in form to the nodal kinetic equation. Space-time distribution of neutron flux in the reactor can be obtained by solving these equations using standard methods.

  8. Romanian network for structural integrity assessment of nuclear components

    International Nuclear Information System (INIS)

    Roth, Maria; Constantinescu, Dan Mihai; Brad, Sebastian; Ducu, Catalin

    2008-01-01

    Full text: Based of the Romanian option to develop and operate nuclear facilities, using as model the networks created at European level and taking into account the international importance of the structural integrity assessments for lifetime extension of the nuclear components, a national Project started since 2005 in the framework of the National Program 'Research of Excellence', Modulus I 2006-2008, managed by the Ministry of Education and Research. Entitled 'Integrated Network for Structural Integrity Monitoring of Critical Components in Nuclear Facilities', with the acronym RIMIS, the Project had two main objectives: - to elaborate a procedure applicable to the structural integrity assessment of the critical components used in Romanian nuclear facilities; - to integrate the national networking in a similar one, at European level, to enhance the scientific significance of Romanian R and D organizations as well as to increase the contribution to solving one of the major issue of the nuclear field. The paper aimed to present the activities performed in the Romanian institutes, involved in the Project, the final results obtained as part of the R and D activities, including experimental, theoretical and modeling ones regarding structural integrity assessment of nuclear components employed in CANDU type reactors. Also the activity carried out in the framework of the NULIFE network, created at European level of the FP6 Program and sustained by the RIMIS network will be described. (authors)

  9. Estimation of radioactivity in structural materials of ETRR-1 reactor

    Energy Technology Data Exchange (ETDEWEB)

    Imam, M [National Center for Nuclear Safety and Radiation Control Atomic Energy Authority, Cairo (Egypt)

    1997-12-31

    Precise knowledge of the thermal neutron flux in the different structural materials of a reactor is necessary to estimate the radioactive inventory in these materials that are needed in any decommissioning study of the reactor. ETRR-1 is a research reactor that went critical on 2/1691. In spite of this long age of the reactor, the effective operation time of this reactor is very short since the reactor was shutdown for long periods. Because of this long age one may think of reactor decommissioning. For this purpose, the radioactivity of the reactor structural materials was estimated. Apart from the reactor core, the important structural materials in the ETRR-1 are the reactor tank, shielding concrete, and the graphite thermal column. The thermal neutron flux was determined by the monte Carlo method in these materials and the isotope inventory and the radioactivity were calculated by the international code ORIGEN-JR. 1 fig.

  10. Residual strength evaluation of concrete structural components ...

    Indian Academy of Sciences (India)

    This paper presents methodologies for residual strength evaluation of concrete structural components using linear elastic and nonlinear fracture mechanics principles. The effect of cohesive forces due to aggregate bridging has been represented mathematically by employing tension softening models. Various tension ...

  11. Protective guide structure for reactor control rod

    International Nuclear Information System (INIS)

    Ban, Minoru; Umeda, Kenji; Kubo, Noboru; Ito, Tomohiro.

    1996-01-01

    The present invention provides an improved protective guide structure for control rods, which does not cause swirling of coolants and resonance even though a slit is formed on a protective tube which surrounds a control rod element in a PWR type reactor. Namely, a reactor control rod is constituted with elongated control elements collectively bundled in the form of a cluster. The protective guide structure protectively guides the collected constituent at the upper portion of a reactor container. The protective structure comprises a plurality of protective tubes each having a C-shaped cross section disposed in parallel for receiving control rod elements individually in which the corners of the opening of the cross section of the protective tube are chamfered to an appropriate configuration. With such a constitution, even if coolant flows in a circumferential direction along the protective tubes surrounding the control rod elements, no shearing stream is caused to the coolants flow since the corners of the cross sectional opening (slit) of the tube are chamfered. Accordingly, occurrence of swirlings can be suppressed. (I.S.)

  12. Structural performance of a graphite blanket in fusion reactors

    International Nuclear Information System (INIS)

    Wolfer, W.G.; Watson, R.D.

    1978-01-01

    Irradiation of graphite in a fusion reactor causes dimensional changes, enhanced creep, and changes in elastic properties and fracture strength. Temperature and flux gradients through the graphite blanket structure produce differential distortions and stress gradients. An inelastic stress analysis procedure is described which treats these variations of the graphite properties in a consistent manner as dictated by physical models for the radiation effects. Furthermore, the procedure follows the evolution of the stress and fracture strength distributions during the reactor operation as well as for possible shutdowns at any time. The lifetime of the graphite structure can be determined based on the failure criterion that the stress at any location exceeds one-half of the fracture strength. This procedure is applied to the most critical component of the blanket module in the SOLASE design

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

    International Nuclear Information System (INIS)

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

    2009-07-01

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

  14. Reactor water spontaneous circulation structure in reactor pressure vessel

    International Nuclear Information System (INIS)

    Takahashi, Kazumi

    1998-01-01

    The gap between the inner wall of a reactor pressure vessel of a BWR type reactor and a reactor core shroud forms a down comer in which reactor water flows downwardly. A feedwater jacket to which feedwater at low temperature is supplied is disposed at the outer circumference of the pressure vessel just below a gas/water separator. The reactor water at the outer circumferential portion just below the air/water separator is cooled by the feedwater jacket, and the feedwater after cooling is supplied to the feedwater entrance disposed below the feedwater jacket by way of a feedwater introduction line to supply the feedwater to the lower portion of the down comer. This can cool the reactor water in the down comer to increase the reactor water density in the down comer thereby forming strong downward flows and promote the recycling of the reactor water as a whole. With such procedures, the reactor water can be recycled stably only by the difference of the specific gravity of the reactor water without using an internal pump. In addition, the increase of the height of the pressure vessel can be suppressed. (I.N.)

  15. Transactions of the 10th international conference on structural mechanics in reactor technology

    International Nuclear Information System (INIS)

    Hadjian, A.H.

    1989-01-01

    This book covers all aspects of engineering mechanics pertaining to mechanical and structural components and the relevant systems in nuclear reactors. Subjects covered include: theoretical developments in structural mechanics, loading conditions, behavior of materials, fluid mechanics, operating experience, accident sequences, and calculational procedures. Problems of structural mechanics analysis are focused within the general context of the design, reliability, and safety of nuclear reactors. Operating plant performance and life extension, waste repository technology and regulatory research have been formalized as distinct Divisions

  16. Probabilistic approaches to life prediction of nuclear plant structural components

    International Nuclear Information System (INIS)

    Villain, B.; Pitner, P.; Procaccia, H.

    1996-01-01

    In the last decade there has been an increasing interest at EDF in developing and applying probabilistic methods for a variety of purposes. In the field of structural integrity and reliability they are used to evaluate the effect of deterioration due to aging mechanisms, mainly on major passive structural components such as steam generators, pressure vessels and piping in nuclear plants. Because there can be numerous uncertainties involved in a assessment of the performance of these structural components, probabilistic methods. The benefits of a probabilistic approach are the clear treatment of uncertainly and the possibility to perform sensitivity studies from which it is possible to identify and quantify the effect of key factors and mitigative actions. They thus provide information to support effective decisions to optimize In-Service Inspection planning and maintenance strategies and for realistic lifetime prediction or reassessment. The purpose of the paper is to discuss and illustrate the methods available at EDF for probabilistic component life prediction. This includes a presentation of software tools in classical, Bayesian and structural reliability, and an application on two case studies (steam generator tube bundle, reactor pressure vessel). (authors)

  17. Probabilistic approaches to life prediction of nuclear plant structural components

    International Nuclear Information System (INIS)

    Villain, B.; Pitner, P.; Procaccia, H.

    1996-01-01

    In the last decade there has been an increasing interest at EDF in developing and applying probabilistic methods for a variety of purposes. In the field of structural integrity and reliability they are used to evaluate the effect of deterioration due to aging mechanisms, mainly on major passive structural components such as steam generators, pressure vessels and piping in nuclear plants. Because there can be numerous uncertainties involved in an assessment of the performance of these structural components, probabilistic methods provide an attractive alternative or supplement to more conventional deterministic methods. The benefits of a probabilistic approach are the clear treatment of uncertainty and the possibility to perform sensitivity studies from which it is possible to identify and quantify the effect of key factors and mitigative actions. They thus provide information to support effective decisions to optimize In-Service Inspection planning and maintenance strategies and for realistic lifetime prediction or reassessment. The purpose of the paper is to discuss and illustrate the methods available at EDF for probabilistic component life prediction. This includes a presentation of software tools in classical, Bayesian and structural reliability, and an application on two case studies (steam generator tube bundle, reactor pressure vessel)

  18. Analysis methods for structure reliability of piping components

    International Nuclear Information System (INIS)

    Schimpfke, T.; Grebner, H.; Sievers, J.

    2004-01-01

    In the frame of the German reactor safety research program of the Federal Ministry of Economics and Labour (BMWA) GRS has started to develop an analysis code named PROST (PRObabilistic STructure analysis) for estimating the leak and break probabilities of piping systems in nuclear power plants. The long-term objective of this development is to provide failure probabilities of passive components for probabilistic safety analysis of nuclear power plants. Up to now the code can be used for calculating fatigue problems. The paper mentions the main capabilities and theoretical background of the present PROST development and presents some of the results of a benchmark analysis in the frame of the European project NURBIM (Nuclear Risk Based Inspection Methodologies for Passive Components). (orig.)

  19. Modeling accelerator structures and RF components

    International Nuclear Information System (INIS)

    Ko, K., Ng, C.K.; Herrmannsfeldt, W.B.

    1993-03-01

    Computer modeling has become an integral part of the design and analysis of accelerator structures RF components. Sophisticated 3D codes, powerful workstations and timely theory support all contributed to this development. We will describe our modeling experience with these resources and discuss their impact on ongoing work at SLAC. Specific examples from R ampersand D on a future linear collide and a proposed e + e - storage ring will be included

  20. Structured automated code checking through structural components and systems engineering

    NARCIS (Netherlands)

    Coenders, J.L.; Rolvink, A.

    2014-01-01

    This paper presents a proposal to employ the design computing methodology proposed as StructuralComponents (Rolvink et al [6] and van de Weerd et al [7]) as a method to perform a digital verification process to fulfil the requirements related to structural design and engineering as part of a

  1. A study on the establishment of component/equipment performance criteria considering Heavy Water Reactor characteristics

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Keun Sun; Kwon, Young Chul; Lee, Min Kyu; Lee, Yun Soo [Sunmoon Univ., Asan (Korea, Republic of); Chang, Seong Hoong; Ryo, Chang Hyun [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of); Kim, Soong Pyung; Hwnag, Jung Rye; Chung, Chul Kee [Chosun Univ., Gwangju (Korea, Republic of)

    2002-03-15

    Foreign and domestic technology trends, regulatory requirements, design and researches for heavy water reactors are analyzed. Safety design guides of Canada industry and regulatory documents and consultative documents of Canada regulatory agency are reviewed. Applicability of MOST guidance 16 Revision 'guidance for technical criteria of nuclear reactor facility' is reviewed. Specific performance criteria are established for components and facilities for heavy water reactor.

  2. Creep/fatigue damage prediction of fast reactor components using shakedown methods

    International Nuclear Information System (INIS)

    Buckthorpe, D.E.

    1997-01-01

    The present status of the shakedown method is reviewed, the application of the shakedown based principles to complex hardening and creep behaviour is described and justified and the prediction of damage against design criteria outlined. Comparisons are made with full inelastic analysis solutions where these are available and against damage assessments using elastic and inelastic design code methods. Current and future developments of the method are described including a summary of the advances made in the development of the post process ADAPT, which has enabled the method to be applied to complex geometry features and loading cases. The paper includes a review of applications of the method to typical Fast Reactor structural example cases within the primary and secondary circuits. For the primary circuit this includes structures such as the large diameter internal shells which are surrounded by hot sodium and subject to slow and rapid thermal transient loadings. One specific case is the damage assessment associated with thermal stratifications within sodium and the effects of moving sodium surfaces arising from reactor trip conditions. Other structures covered are geometric features within components such as the Above Core structure and Intermediate Heat Exchanger. For the secondary circuit the method has been applied to alternative and more complex forms of geometry namely thick section tubeplates of the Steam Generator and a typical secondary circuit piping run. Both of these applications are in an early stage of development but are expected to show significant advantages with respect to creep and fatigue damage estimation compared with existing code methods. The principle application of the method to design has so far been focused on Austenitic Stainless steel components however current work shows some significant benefits may be possible from the application of the method to structures made from Ferritic steels such as Modified 9Cr 1Mo. This aspect is briefly

  3. Facility Configuration Study of the High Temperature Gas-Cooled Reactor Component Test Facility

    Energy Technology Data Exchange (ETDEWEB)

    S. L. Austad; L. E. Guillen; D. S. Ferguson; B. L. Blakely; D. M. Pace; D. Lopez; J. D. Zolynski; B. L. Cowley; V. J. Balls; E.A. Harvego, P.E.; C.W. McKnight, P.E.; R.S. Stewart; B.D. Christensen

    2008-04-01

    A test facility, referred to as the High Temperature Gas-Cooled Reactor Component Test Facility or CTF, will be sited at Idaho National Laboratory for the purposes of supporting development of high temperature gas thermal-hydraulic technologies (helium, helium-Nitrogen, CO2, etc.) as applied in heat transport and heat transfer applications in High Temperature Gas-Cooled Reactors. Such applications include, but are not limited to: primary coolant; secondary coolant; intermediate, secondary, and tertiary heat transfer; and demonstration of processes requiring high temperatures such as hydrogen production. The facility will initially support completion of the Next Generation Nuclear Plant. It will secondarily be open for use by the full range of suppliers, end-users, facilitators, government laboratories, and others in the domestic and international community supporting the development and application of High Temperature Gas-Cooled Reactor technology. This pre-conceptual facility configuration study, which forms the basis for a cost estimate to support CTF scoping and planning, accomplishes the following objectives: • Identifies pre-conceptual design requirements • Develops test loop equipment schematics and layout • Identifies space allocations for each of the facility functions, as required • Develops a pre-conceptual site layout including transportation, parking and support structures, and railway systems • Identifies pre-conceptual utility and support system needs • Establishes pre-conceptual electrical one-line drawings and schedule for development of power needs.

  4. Facility Configuration Study of the High Temperature Gas-Cooled Reactor Component Test Facility

    International Nuclear Information System (INIS)

    S. L. Austad; L. E. Guillen; D. S. Ferguson; B. L. Blakely; D. M. Pace; D. Lopez; J. D. Zolynski; B. L. Cowley; V. J. Balls; E.A. Harvego, P.E.; C.W. McKnight, P.E.; R.S. Stewart; B.D. Christensen

    2008-01-01

    A test facility, referred to as the High Temperature Gas-Cooled Reactor Component Test Facility or CTF, will be sited at Idaho National Laboratory for the purposes of supporting development of high temperature gas thermal-hydraulic technologies (helium, helium-Nitrogen, CO2, etc.) as applied in heat transport and heat transfer applications in High Temperature Gas-Cooled Reactors. Such applications include, but are not limited to: primary coolant; secondary coolant; intermediate, secondary, and tertiary heat transfer; and demonstration of processes requiring high temperatures such as hydrogen production. The facility will initially support completion of the Next Generation Nuclear Plant. It will secondarily be open for use by the full range of suppliers, end-users, facilitators, government laboratories, and others in the domestic and international community supporting the development and application of High Temperature Gas-Cooled Reactor technology. This pre-conceptual facility configuration study, which forms the basis for a cost estimate to support CTF scoping and planning, accomplishes the following objectives: (1) Identifies pre-conceptual design requirements; (2) Develops test loop equipment schematics and layout; (3) Identifies space allocations for each of the facility functions, as required; (4) Develops a pre-conceptual site layout including transportation, parking and support structures, and railway systems; (5) Identifies pre-conceptual utility and support system needs; and (6) Establishes pre-conceptual electrical one-line drawings and schedule for development of power needs

  5. Annual report on the state of RB reactor components and equipment, december 1999

    International Nuclear Information System (INIS)

    Milosevic, M.

    1999-12-01

    According to the performed analysis, it is considered that the RB reactor can be operated safely until the existing control and safety systems could be maintained in satisfactory operable state. Failures of heavy water circulation system valves which may cause decreased availability but no accident. During 1998 the reactor lattice was changed 13 times, meaning that experiments were done with 13 configurations of the reactor core. Total reactor operation amounted to 84 Wh with 40 start-ups (attained criticality levels). This report contains 4 Annexes, detailed description of the state of reactor equipment in 1999, reactor operation nd utilization data, plan for regular annual maintenance and refurbishment of reactor equipment and plan for minimum needed resources for regular maintenance of the components and equipment in the forthcoming year

  6. Annual report on the state of RB reactor components and equipment, december 1998

    International Nuclear Information System (INIS)

    Milosevic, M.

    1998-12-01

    According to the performed analysis, it is considered that the RB reactor can be operated safely until the existing control and safety systems could be maintained in satisfactory operable state. Failures of heavy water circulation system valves which may cause decreased availability but no accident. During 1998 the reactor lattice was changed 7 times, meaning that experiments were done with 7 configurations of the reactor core. Total reactor operation amounted to 177.5 Wh with 40 start-ups (attained criticality levels). This report contains 4 Annexes, detailed description of the state of reactor equipment in 1998, reactor operation nd utilization data, plan for regular annual maintenance and refurbishment of reactor equipment and plan for minimum needed resources for regular maintenance of the components and equipment in the forthcoming year

  7. Packaging of structural health monitoring components

    Science.gov (United States)

    Kessler, Seth S.; Spearing, S. Mark; Shi, Yong; Dunn, Christopher T.

    2004-07-01

    Structural Health Monitoring (SHM) technologies have the potential to realize economic benefits in a broad range of commercial and defense markets. Previous research conducted by Metis Design and MIT has demonstrated the ability of Lamb waves methods to provide reliable information regarding the presence, location and type of damage in composite specimens. The present NSF funded program was aimed to study manufacturing, packaging and interface concepts for critical SHM components. The intention is to be able to cheaply manufacture robust actuating/sensing devices, and isolate them from harsh operating environments including natural, mechanical, or electrical extremes. Currently the issues related to SHM system durability have remained undressed. During the course of this research several sets of test devices were fabricated and packaged to protect the piezoelectric component assemblies for robust operation. These assemblies were then tested in hot and wet conditions, as well as in electrically noisy environments. Future work will aim to package the other supporting components such as the battery and wireless chip, as well as integrating all of these components together for operation. SHM technology will enable the reduction or complete elimination of scheduled inspections, and will allow condition-based maintenance for increased reliability and reduced overall life-cycle costs.

  8. Study on the Safety Classification Criteria of Mechanical Systems and Components for Open Pool-Type Research Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Belal, Al Momani [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of); Jo, Jong Chull [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2013-10-15

    This paper describes a new compromised safety classification approach based on the comparative study of the different practices in safety classification of mechanical systems and components of open pool-type RRs, which have been adopted by several developed countries in the nuclear power area. It is hoped that the proposed safety classification criteria will be used to develop a harmonized consensus international standard. Different safety classification criteria for systems, structures, and components (SSCs) of nuclear reactors are used among the countries that export or import nuclear reactor technology, which may make the nuclear technology trade and exchange difficult. Thus, such various different approaches of safety classification need to be compromised to establish a global standard. This article proposes practicable optimized criteria for safety classification of SSCs for open pool-type research reactors (RRs)

  9. Study on the Safety Classification Criteria of Mechanical Systems and Components for Open Pool-Type Research Reactors

    International Nuclear Information System (INIS)

    Belal, Al Momani; Jo, Jong Chull

    2013-01-01

    This paper describes a new compromised safety classification approach based on the comparative study of the different practices in safety classification of mechanical systems and components of open pool-type RRs, which have been adopted by several developed countries in the nuclear power area. It is hoped that the proposed safety classification criteria will be used to develop a harmonized consensus international standard. Different safety classification criteria for systems, structures, and components (SSCs) of nuclear reactors are used among the countries that export or import nuclear reactor technology, which may make the nuclear technology trade and exchange difficult. Thus, such various different approaches of safety classification need to be compromised to establish a global standard. This article proposes practicable optimized criteria for safety classification of SSCs for open pool-type research reactors (RRs)

  10. Constitutive modeling and finite element procedure development for stress analysis of prismatic high temperature gas cooled reactor graphite core components

    International Nuclear Information System (INIS)

    Mohanty, Subhasish; Majumdar, Saurindranath; Srinivasan, Makuteswara

    2013-01-01

    Highlights: • Finite element procedure developed for stress analysis of HTGR graphite component. • Realistic fluence profile and reflector brick shape considered for the simulation. • Also realistic H-451 grade material properties considered for simulation. • Typical outer reflector of a GT-MHR type reactor considered for numerical study. • Based on the simulation results replacement of graphite bricks can be scheduled. -- Abstract: High temperature gas cooled reactors, such as prismatic and pebble bed reactors, are increasingly becoming popular because of their inherent safety, high temperature process heat output, and high efficiency in nuclear power generation. In prismatic reactors, hexagonal graphite bricks are used as reflectors and fuel bricks. In the reactor environment, graphite bricks experience high temperature and neutron dose. This leads to dimensional changes (swelling and or shrinkage) of these bricks. Irradiation dimensional changes may affect the structural integrity of the individual bricks as well as of the overall core. The present paper presents a generic procedure for stress analysis of prismatic core graphite components using graphite reflector as an example. The procedure is demonstrated through commercially available ABAQUS finite element software using the option of user material subroutine (UMAT). This paper considers General Atomics Gas Turbine-Modular Helium Reactor (GT-MHR) as a bench mark design to perform the time integrated stress analysis of a typical reflector brick considering realistic geometry, flux distribution and realistic irradiation material properties of transversely isotropic H-451 grade graphite

  11. Constitutive modeling and finite element procedure development for stress analysis of prismatic high temperature gas cooled reactor graphite core components

    Energy Technology Data Exchange (ETDEWEB)

    Mohanty, Subhasish, E-mail: smohanty@anl.gov [Argonne National Laboratory, South Cass Avenue, Argonne, IL 60439 (United States); Majumdar, Saurindranath [Argonne National Laboratory, South Cass Avenue, Argonne, IL 60439 (United States); Srinivasan, Makuteswara [U.S. Nuclear Regulatory Commission, Washington, DC 20555 (United States)

    2013-07-15

    Highlights: • Finite element procedure developed for stress analysis of HTGR graphite component. • Realistic fluence profile and reflector brick shape considered for the simulation. • Also realistic H-451 grade material properties considered for simulation. • Typical outer reflector of a GT-MHR type reactor considered for numerical study. • Based on the simulation results replacement of graphite bricks can be scheduled. -- Abstract: High temperature gas cooled reactors, such as prismatic and pebble bed reactors, are increasingly becoming popular because of their inherent safety, high temperature process heat output, and high efficiency in nuclear power generation. In prismatic reactors, hexagonal graphite bricks are used as reflectors and fuel bricks. In the reactor environment, graphite bricks experience high temperature and neutron dose. This leads to dimensional changes (swelling and or shrinkage) of these bricks. Irradiation dimensional changes may affect the structural integrity of the individual bricks as well as of the overall core. The present paper presents a generic procedure for stress analysis of prismatic core graphite components using graphite reflector as an example. The procedure is demonstrated through commercially available ABAQUS finite element software using the option of user material subroutine (UMAT). This paper considers General Atomics Gas Turbine-Modular Helium Reactor (GT-MHR) as a bench mark design to perform the time integrated stress analysis of a typical reflector brick considering realistic geometry, flux distribution and realistic irradiation material properties of transversely isotropic H-451 grade graphite.

  12. Activities in the Czech Republic for reactor pressure components lifetime management

    International Nuclear Information System (INIS)

    Brumovsky, M.

    1995-01-01

    Preparation of a system of regulatory guides for life assessment of main pressure components, for inspection qualification and demonstration programmes are outlined. Lifetime management programme for NPP with WWER-440/V-213 reactors is described. Figs and tabs

  13. Proceedings of the international conference on irradiation behaviour of metallic materials for fast reactor core components

    International Nuclear Information System (INIS)

    Poirier, J.; Dupouy, J.M.

    Radiation effects on metals or alloys used in fast reactor core components are examined in the papers presented at this conference, the accent being put on swelling and irradiation creep of steels and nickel alloys

  14. Surface erosion of fusion reactor components due to radiation blistering and neutron sputtering

    International Nuclear Information System (INIS)

    Das, S.K.; Kaminsky, M.

    1975-01-01

    Radiation blistering and neutron sputtering can lead to the surface erosion of fusion reactor components exposed to plasma radiations. Recent studies of methods to reduce the surface erosion caused by these processes are discussed

  15. Code qualification of structural materials for AFCI advanced recycling reactors.

    Energy Technology Data Exchange (ETDEWEB)

    Natesan, K.; Li, M.; Majumdar, S.; Nanstad, R.K.; Sham, T.-L. (Nuclear Engineering Division); (ORNL)

    2012-05-31

    This report summarizes the further findings from the assessments of current status and future needs in code qualification and licensing of reference structural materials and new advanced alloys for advanced recycling reactors (ARRs) in support of Advanced Fuel Cycle Initiative (AFCI). The work is a combined effort between Argonne National Laboratory (ANL) and Oak Ridge National Laboratory (ORNL) with ANL as the technical lead, as part of Advanced Structural Materials Program for AFCI Reactor Campaign. The report is the second deliverable in FY08 (M505011401) under the work package 'Advanced Materials Code Qualification'. The overall objective of the Advanced Materials Code Qualification project is to evaluate key requirements for the ASME Code qualification and the Nuclear Regulatory Commission (NRC) approval of structural materials in support of the design and licensing of the ARR. Advanced materials are a critical element in the development of sodium reactor technologies. Enhanced materials performance not only improves safety margins and provides design flexibility, but also is essential for the economics of future advanced sodium reactors. Code qualification and licensing of advanced materials are prominent needs for developing and implementing advanced sodium reactor technologies. Nuclear structural component design in the U.S. must comply with the ASME Boiler and Pressure Vessel Code Section III (Rules for Construction of Nuclear Facility Components) and the NRC grants the operational license. As the ARR will operate at higher temperatures than the current light water reactors (LWRs), the design of elevated-temperature components must comply with ASME Subsection NH (Class 1 Components in Elevated Temperature Service). However, the NRC has not approved the use of Subsection NH for reactor components, and this puts additional burdens on materials qualification of the ARR. In the past licensing review for the Clinch River Breeder Reactor Project (CRBRP

  16. An experience of cleaning and decontamination of the BN-350 reactor components

    International Nuclear Information System (INIS)

    Vasilenko, K.T.; Kochetkov, L.A.; Arkhipov, V.M.; Baklushin, R.P.; Gorlov, A.I.; Kiselev, G.V.; Rezinkin, P.S.; Samarkin, A.A.; Tverdovsky, N.D.

    1978-01-01

    In the course of start-up, adjustment and operation of the BN-350 reactor there arose a need for cleaning from sodium and decontamination of primary and secondary equipment components. Design schemes of the systems provided for this purpose as well as those specially designed for cleaning of steam generator evaporators are considered. Technological processes of cleaning and decontamination for some reactor components (removable parts of circulating pumps, evaporators, valves) are described, the results are presented. (author)

  17. Overview of European Community (Activity 3) work on materials properties of fast reactor structural materials

    International Nuclear Information System (INIS)

    Wood, D.S.

    The Fast Reactor Coordinating Committee set up in 1974 the Working Group Codes and Standards, and organized its work into four main activities: Manufacturing standards, Structural analysis, Materials and Classification of components. The main purpose of materials activity is to compare and contrast existing national specifications and associated properties relevant to structural materials in fast reactors. Funds are available on a yearly basis for tasks to be carried out through Study Contracts. At present about four Study Contract Reports are prepared each year

  18. Component-Level Prognostics Health Management Framework for Passive Components - Advanced Reactor Technology Milestone: M2AT-15PN2301043

    Energy Technology Data Exchange (ETDEWEB)

    Ramuhalli, Pradeep; Roy, Surajit; Hirt, Evelyn H.; Prowant, Matthew S.; Pitman, Stan G.; Tucker, Joseph C.; Dib, Gerges; Pardini, Allan F.

    2015-06-19

    This report describes research results to date in support of the integration and demonstration of diagnostics technologies for prototypical advanced reactor passive components (to establish condition indices for monitoring) with model-based prognostics methods. Achieving this objective will necessitate addressing several of the research gaps and technical needs described in previous technical reports in this series.

  19. Reactor Structure Materials: Corrosion of Reactor Core Internals

    International Nuclear Information System (INIS)

    Van Dyck, S.

    2000-01-01

    The objectives of SCK-CEN's R and D programme on the corrosion of reactor core internals are: (1) to gain mechanistic insight into the Irradition Assisted Stress Corrosion Cracking (IASCC) phenomenon by studying the influence of separate parameters in well controlled experiments; (2) to develop and validate a predictive capability on IASCC by model description and (3) to define and validate countermeasures and monitoring techniques for application in reactors. Progress and achievements in 1999 are described

  20. Overview of fast reactor structural materials programme in India

    International Nuclear Information System (INIS)

    Rodriguez, P.; Paranjpe, S.R.; Chetal, S.C.; Mannan, S.L.; Ray, S.K.; Seetharaman, V.; Srinivasan, G.

    The fast reactor structural materials activities in India comprise of the programme on the materials for the Fast Breeder Test Reactor (FBTR), the construction of which is nearing completion, and the programme on the candidate materials for the Prototype Fast Breeder Reactor (PFBR) which is now in the design stage. For the materials in use in FBTR, the main thrust has been towards detailed evaluation and documentation of long term (creep) properties of type 316 stainless steel base material in air. For the PFBR the philosophy has been to identify the candidate materials and to evolve a wider scope for the testing and evaluation programmes. The major structural component is identified as variants of type 304 stainless steel and the programmes undertaken include study of low cycle fatigue properties and environmental effects on creep and stress rupture properties. Evaluations of aging embrittlement of type 316 stainless steel base material and weldments are also in progress. The paper lists the testing programmes identified for adoption in the near future. These include creep-fatigue damage studies and fracture mechanics studies on weldments for type 304 stainless steel and testing programme on 2.25 Cr-1 Mo and 9 Cr-1 Mo steels, the identified candidate materials for steam generators. The development efforts also include a comprehensive programme on inelastic analysis procedure. (author)

  1. Development of components for the gas-cooled fast breeder reactor program

    International Nuclear Information System (INIS)

    Dee, J.B.; Macken, T.

    1977-01-01

    The gas-cooled fast breeder reactor (GCFR) component development program is based on an extension of high temperature gas-cooled reactor (HTGR) component technology; therefore, the GCFR development program is addressed primarily to components which differ in design and requirements from HTGR components. The principal differences in primary system components are due to the increase in helium coolant pressure level, which benefits system size and efficiency in the GCFR, and differences in the reactor internals and fuel handling systems due to the use of the compact metal-clad core. The purpose of this paper is to present an overview of the principal component design differences between the GCFR and HTGR and the consequent influences of these differences on GCFR component development programs. Development program plans are discussed and include those for the prestressed concrete reactor vessel (PCRV), the main helium circulator and its supporting systems, the steam generators, the reactor thermal shielding, and the fuel handling system. Facility requirements to support these development programs are also discussed. Studies to date show that GCFR component development continues to appear to be incremental in nature, and the required tests are adaptations of related HTGR test programs. (Auth.)

  2. Proceedings of 18th international conference on structural mechanics in reactor technology

    International Nuclear Information System (INIS)

    2005-07-01

    The 18th International Conference on Structural Mechanics in Reactor Technology was held on August 7-12, 2005 in Beijing, China, and Sponsored by International Association for Structural Mechanics in Reactor Technology, Chinese Nuclear Society, Chinese Society of Theoretical and Applied Mechanics, and Tsinghua University. 486 abstracts are Collected. The contents includes: opening, plenary and keynote presentations; computational mechanics; fuel and core structures; aging, life extension, and license renewal; design methods and rules for components; fracture mechanics; concrete material, containment and other structures; analysis and design for dynamic and extreme loads; seismic analysis, design and qualification; structural reliability and probabilistic safety assessment (PSA); operation, inspection and maintenance; severe accident management and structural evaluation; advanced reactors and generation IV reactors; decommissioning of nuclear facilities and waste management.

  3. KfK, Institute for Reactor Components. Results of research and development activities in 1989

    International Nuclear Information System (INIS)

    1990-03-01

    R and D activities at IRB (Institut fuer Reaktorbauelemente - Institute for Reactor Components) are dedicated to thermodynamics and fluid dynamics. Emphasis is on the design of nuclear reactor and fusion reactor components. Environmental engineering was added recently. Most activities are applications-oriented. Fundamental investigations focus on energy research and energy technology. The activities are carried out in the framework of different projects (PKF/nuclear fusion, PSF/nuclear safety, PSU/pollution control). Points of main effort are the development of basic liquid-metal-cooled blanket solutions, investigations on natural convection in reactor tanks, and the cooling properties of future containments for pressurized water reactors in the case of nuclear fusion accidents. (orig./GL) [de

  4. A study on the fault diagnostic techniques for reactor internal structures using neutron noise analysis

    International Nuclear Information System (INIS)

    Kim, Tae Ryong; Jeong, Seong Ho; Park, Jin Ho; Park, Jin Suk

    1994-08-01

    The unfavorable phenomena, such as flow induced vibration and aging process in reactor internals, cause degradation of structural integrity and may result in loosing some mechanical binding components which might impact other equipments and components or cause flow blockage. Since these malfunctions and potential failures change reactor noise signal, it is necessary to analyze reactor noise signal for early fault diagnosis in the point of few of safety and plant economics. The objectives of this study are to establish fault diagnostic and TS(thermal shield), and to develop a data acquisition and signal processing software system. In the first year of this study, an analysis technique for the reactor internal vibration using the reactor noise was proposed. With the technique proposed and the reactor noise signals (ex-core neutron and acceleration), the dynamic characteristics of Ulchin-1 reactor internals were obtained, and compared with those of Tricastin-1 which is the prototype of Ulchin-1. In the second year, a PC-based expert system for reactor internals fault diagnosis is developed, which included data acquisition, signal processing, feature extraction function, and represented diagnostic knowledge by the IF-THEN rule. To know the effect of the faults, the reactor internals of Ulchin-1 is modeled using FEM and simulated with an artificial defect given in the hold-down spring. Trend in the dynamic characteristics of reactor internals is also observed during one fuel cycle to know the effect of boron concentration. 100 figs, 7 tabs, 18 refs. (Author)

  5. Dose rates modeling of pressurized water reactor primary loop components with SCALE6.0

    International Nuclear Information System (INIS)

    Matijević, Mario; Pevec, Dubravko; Trontl, Krešimir

    2015-01-01

    Highlights: • Shielding analysis of the typical PWR primary loop components was performed. • FW-CADIS methodology was thoroughly investigated using SCALE6.0 code package. • Versatile ability of SCALE6.0/FW-CADIS for deep penetration models was proved. • The adjoint source with focus on specific material can improve MC modeling. - Abstract: The SCALE6.0 simulation model of a typical PWR primary loop components for effective dose rates calculation based on hybrid deterministic–stochastic methodology was created. The criticality sequence CSAS6/KENO-VI of the SCALE6.0 code package, which includes KENO-VI Monte Carlo code, was used for criticality calculations, while neutron and gamma dose rates distributions were determined by MAVRIC/Monaco shielding sequence. A detailed model of a combinatorial geometry, materials and characteristics of a generic two loop PWR facility is based on best available input data. The sources of ionizing radiation in PWR primary loop components included neutrons and photons originating from critical core and photons from activated coolant in two primary loops. Detailed calculations of the reactor pressure vessel and the upper reactor head have been performed. The efficiency of particle transport for obtaining global Monte Carlo dose rates was further examined and quantified with a flexible adjoint source positioning in phase-space. It was demonstrated that generation of an accurate importance map (VR parameters) is a paramount step which enabled obtaining Monaco dose rates with fairly uniform uncertainties. Computer memory consumption by the S N part of hybrid methodology represents main obstacle when using meshes with large number of cells together with high S N /P N parameters. Detailed voxelization (homogenization) process in Denovo together with high S N /P N parameters is essential for precise VR parameters generation which will result in optimized MC distributions. Shielding calculations were also performed for the reduced PWR

  6. AGING MANAGMENT OF REACTOR COOLANT SYSTEM MECHANICAL COMPONENTS FOR LICENSE RENEWAL

    International Nuclear Information System (INIS)

    SUBUDHI, M.; MORANTE, R.; LEE, A.D.

    2002-01-01

    The reactor coolant system (RCS) mechanical components that require an aging management review for license renewal include the primary loop piping and associated connections to other support systems, reactor vessel, reactor vessel internals, pressurizer. steam generators, reactor coolant pumps, and all other inter-connected piping, pipe fittings, valves, and bolting. All major RCS components are located inside the reactor building. Based on the evaluation findings of recently submitted license renewal applications for pressurized water reactors, this paper presents the plant programs and/or activities proposed by the applicants to manage the effects of aging. These programs and/or activities provide reasonable assurance that the intended function(s) of these mechanical components will be maintained for the period of extended operation. The license renewal application includes identification of RCS subcomponents that are within the scope of license renewal and are vulnerable to age-related degradation when exposed to environmental and operational conditions. determination of the effects of aging on their intended safety functions. and implementation of the aging management programs and/or activities including both current and new programs. Industry-wide operating experience, including generic communication by the NRC, is part of the aging management review for the RCS components. In addition, this paper discusses time-limited aging analyses associated with neutron embrittlement of the reactor vessel beltline region and thermal fatigue

  7. Positron annihilation studies on structural materials for nuclear reactors

    International Nuclear Information System (INIS)

    Rajaraman, R.; Amarendra, G.; Sundar, C.S.

    2012-01-01

    Structural steels for nuclear reactors have renewed interest owing to the future advanced fission reactor design with increased burn-up goals as well as for fusion reactor applications. While modified austenitic steels continue to be the main cladding materials for fast breeder reactors, Ferritic/martensitic steels and oxide dispersion strengthened ferritic steels are the candidate materials for future reactors applications in India. Sensitivity and selectivity of positron annihilation spectroscopy to open volume type defects and nano clusters have been extensively utilized in studying reactor materials. We have recently reviewed the application of positron techniques to reactor structural steels. In this talk, we will present successful application of positron annihilation spectroscopy to probe various structural materials such as D9, ferritic/martensitic, oxide dispersion strengthened (ODS) steels and related model alloys, highlighting our recent studies. (author)

  8. Physical characteristics of non-fuel assembly reactor components

    International Nuclear Information System (INIS)

    Hawkes, E.C.

    1994-09-01

    The primary objective of this report is to enhance the utility of the Characteristics Data Base (CDB). This has been accomplished by providing a pictorial representation of the principal non-fuel assembly (NFA) components along with a tabular summary of key information about each type of component. This report is intended for use as an adjunct to the CDB. Toward this end, the report may be used either as a complement to the detailed descriptions in the CDB, or as a stand-alone document that acts as an illustrated abstract of the CDB. Line drawings of major NFA components are included. Data not provided in the CDB are also included. Summary descriptions of each component are given in tabular format

  9. Development of large components for the fusion reactor vacuum circuits

    International Nuclear Information System (INIS)

    Perinic, D.; Lorrain, C.

    1986-06-01

    The Commission of the European Communities appointed in mid-1983 the Centre d'Etudes Nucleaires de Saclay and the Kernforschungszentrum Karlsruhe GmbH to investigate whether large vacuum components for use in the fusion machine can be built. The following individual targets have been defined for studies under this project: - Elaboration of technical specifications for large components. - Investigation of the feasibility. - Specification of the tests required and planning of a testing facility. The plasma chamber pumping system is essentially concerned

  10. Structural analysis of the reactor pool for the RRRP

    International Nuclear Information System (INIS)

    Alberro, J.G.; Abbate, A.D.

    2005-01-01

    The purpose of the present document is to describe the structural design of the Reactor Pool relevant to the RRRP (Replacement Research Reactor Project) for the Australian Nuclear Science and Technology Organisation. The structural analysis required coordinated design, engineering, analysis, and fabrication efforts. The pool has been designed, manufactured, and inspected following as guideline the ASME Boiler and Pressure Vessel Code, which defines the requirements for the pool to withstand hydrostatic and mechanical forces, ensuring its integrity throughout its lifetime. Standard off-the-shelf finite element programs (Nastran and Ansys codes) were used to evaluate the pool and further qualify the design and its construction. Both global and local effect analyses were carried out. The global analysis covers the structural integrity of the pool wall (6 mm thick) considering the different load states acting on it, namely hydrostatic pressure, thermal expansion, and seismic event. The local analysis evaluates the structural behaviour of the pool at specific points resulting from the interaction among components. It is confirmed that maximum stresses and displacements fall below the allowable values required by the ASME Boiler and Pressure Vessel Code. The water pressure analysis was validated by means of a hydrostatic test. (authors)

  11. Diagnostic Technology Development for Core Internal Structure in CANDU reactor

    International Nuclear Information System (INIS)

    Jung, Hyun Kyu; Cheong, Y. M.; Lee, Y. S. and others

    2005-04-01

    Degradation of critical components of nuclear power plants has become important as the operating years of plants increase. The necessity of degradation study including measurement and monitoring technology has increased continuously. Because the fuel channels and the neighboring sensing tubes and control rods are particularly one of the critical components in CANDU nuclear plant, they are treated as a major research target in order to counteract the possible problems and establish the counterplan for the CANDU reactor safety improvement. To ensure the core structure integrity in CANDU nuclear plant, the following 2 research tasks were performed: Development of NDE technologies for the gap measurement between the fuel channels and LIN tubes. Development of vibration monitoring technology of the fuel channels and sensing tubes. The technologies developed in this study could contribute to the nuclear safety and estimation of the remaining life of operating CANDU nuclear power plants

  12. Graphites and composites irradiations for gas cooled reactor core structures

    International Nuclear Information System (INIS)

    Van der Laan, J.G.; Vreeling, J.A.; Buckthorpe, D.E.; Reed, J.

    2008-01-01

    Full text of publication follows. Material investigations are undertaken as part of the European Commission 6. Framework Programme for helium-cooled fission reactors under development like HTR, VHTR, GCFR. The work comprises a range of activities, from (pre-)qualification to screening of newly designed materials. The High Flux Reactor at Petten is the main test bed for the irradiation test programmes of the HTRM/M1, RAPHAEL and ExtreMat Integrated Projects. These projects are supported by the European Commission 5. and 6. Framework Programmes. To a large extent they form the European contribution to the Generation-IV International Forum. NRG is also performing a Materials Test Reactor project to support British Energy in preparing extended operation of their Advanced Gas-cooled Reactors (AGR). Irradiations of commercial and developmental graphite grades for HTR core structures are undertaken in the range of 650 to 950 deg C, with a view to get data on physical and mechanical properties that enable engineering design. Various C- and SiC-based composite materials are considered for support structures or specific components like control rods. Irradiation test matrices are chosen to cover commercial materials, and to provide insight on the behaviour of various fibre and matrix types, and the effects of architecture and manufacturing process. The programme is connected with modelling activities to support data trending, and improve understanding of the material behaviour and micro-structural evolution. The irradiation programme involves products from a large variety of industrial and research partners, and there is strong interaction with other high technology areas with extreme environments like space, electronics and fusion. The project on AGR core structures graphite focuses on the effects of high dose neutron irradiation and simultaneous radiolytic oxidation in a range of 350 to 450 deg C. It is aimed to provide data on graphite properties into the parameter space

  13. The Capabilities and Limitation of Remote Visual Methods to Detect Service-Induced Cracks in Reactor Components

    International Nuclear Information System (INIS)

    Cumblidge, Stephen E.; Doctor, Steven R.; Anderson, Michael T.

    2006-01-01

    Since 1977, the U.S. Nuclear Regulatory Commission (NRC) Office of Nuclear Regulatory Research has funded a multiyear program at the Pacific Northwest National Laboratory (PNNL) to evaluate the reliability and accuracy of nondestructive evaluation (NDE) techniques employed for inservice inspection (ISI). Recently, the U.S. nuclear industry proposed replacing current volumetric and/or surface examinations of certain components in commercial nuclear power plants, as required by ASME Boiler and Pressure Vessel Code Section XI, with a simpler visual testing (VT) method. The advantages of VT are that these tests generally involve much less radiation exposure and examination times than do volumetric examinations such as ultrasonic testing (UT). However, for industry to justify supplementing volumetric methods with VT, and analysis of pertinent issues is needed to support the reliability of VT in determining the structural integrity of reactor components. As piping and pressure vessel components in a nuclear power station are generally underwater and in high radiation field, they need to be examined by VT from a distance with radiation-hardened video systems. Remote visual testing has been used by nuclear utilities to find cracks in pressure vessel cladding in pressurized water reactors, for shrouds in boiling water reactors, and to investigate leaks in piping and reactor components. These visual tests are performed using a wide variety of procedures and equipment. The techniques for remote visual testing use submersible closed-circuit video cameras to examine reactor components and welds. PNNL has conducted a parametric study that examines the important variables that affect the effectiveness of a remote visual test. Tested variables include lighting techniques, camera resolution, camera movement, and magnification. PNNL has also conducted a laboratory test using a commercial visual testing camera system to experimentally determine the ability of the camera system to

  14. Fracture mechanics and fatigue evaluation of nuclear reactor components

    International Nuclear Information System (INIS)

    Mattar Neto, Miguel; Andrade, Arnaldo H.P. de; Maneschy, Eduardo

    1995-01-01

    This paper presents a theoretical study available in the available literature for evaluation the environmental effects on the lifetime of nuclear power plant components. The author's motivation is to provide some technical tools to identify what research development could be done in this area

  15. An Analysis of Testing Requirements for Fluoride Salt Cooled High Temperature Reactor Components

    Energy Technology Data Exchange (ETDEWEB)

    Holcomb, David Eugene [ORNL; Cetiner, Sacit M [ORNL; Flanagan, George F [ORNL; Peretz, Fred J [ORNL; Yoder Jr, Graydon L [ORNL

    2009-11-01

    This report provides guidance on the component testing necessary during the next phase of fluoride salt-cooled high temperature reactor (FHR) development. In particular, the report identifies and describes the reactor component performance and reliability requirements, provides an overview of what information is necessary to provide assurance that components will adequately achieve the requirements, and then provides guidance on how the required performance information can efficiently be obtained. The report includes a system description of a representative test scale FHR reactor. The reactor parameters presented in this report should only be considered as placeholder values until an FHR test scale reactor design is completed. The report focus is bounded at the interface between and the reactor primary coolant salt and the fuel and the gas supply and return to the Brayton cycle power conversion system. The analysis is limited to component level testing and does not address system level testing issues. Further, the report is oriented as a bottom-up testing requirements analysis as opposed to a having a top-down facility description focus.

  16. Suitability of Co as an alloy material for components of the primary circuit of HTR reactors

    International Nuclear Information System (INIS)

    Iniotakis, N.

    1977-02-01

    For high temperature reactors it is of interest if Co-alloys could be used for the different components of the primary cooling circuit. It has been investigated in detail to what amount the Co-60 created by neutron activation of Co-59 contained in the material of the components could possibly contribute to the contamination of the primary cooling circuit of the reactor. The result of these investigations is compared with the contamination of the cooling circuit by fission and activation products like Co-137, Cs-134, Ag-11om etc. For pebble bed reactors with an OTTO-type fuel management it could be shown that there is no limitation for the use of cobalt in alloys for materials of the components in the primary cooling circuit. The only boundary condition is that the local Thermal Flux at the position of the components should be less than phisub(th) 7 n/cm 2 . sec. (orig.) [de

  17. Methods and means of the radioisotope flaw detection of the nuclear power reactors components

    International Nuclear Information System (INIS)

    Dekopov, A.S.; Majorov, A.N.; Firsov, V.G.

    1979-01-01

    Methods and means are considered for the radioisotopic flaw detection of the nuclear reactors pressure vessels and structural components of the reactor circuit. Methods of control are described as in the technological process of fabrication of the power reactors assemblies as during the systematic-preventive repair of the nuclear power station equipment during exploitation. Methodological base is given of the technology of radiation control of welded joints of the pressure vessel branch piper of the WWER-440 and WWER-1000 reactors in the process of assembling and exploitation and joining pipes with the pipe-plate of the steamgenerator in the process of fabrication. Methods of the radioisotope flaw detection in the process of exploitation take into consideration the influence of the radioisotope background, and ensure obtaining of the demanded by the rules of control, sensitivity. Methods of control of welded joints of the steamgenerator of nuclear power plants are based on the simultaneous examination of all joints with application of the shaped radiographic plate-holders. Special gamma-flaw-detection equipment is developed for control of the welded joints of the main branch-pipes. Design peculiarities are given of the installation for flaw detection. These installations are equipped with the system for emergency return of the radiation source into the storage position from the position for exposure. They have automatic exposure-meters for determination of the exposure time. Successfull exploitation of such installations in the Finland during assembling equipment for the nuclear reactor of the nuclear power plant ''Loviisa-1'' and in the USSR on the Novovoronezh nuclear power plant has shown possibility for detection of flaws having dimensions about 1% of the equipment used. For control of welded joints of pipes with pipe-plates at the steam generators, portable flaw-detectors are used. Sensitivity of these flaw-detectors towards detection of the wire standards has

  18. Thermodynamic characterization of salt components for the Molten Salt Reactor Fuel - 15573

    International Nuclear Information System (INIS)

    Capelli, E.; Konings, R.J.M.; Benes, A.

    2015-01-01

    Molten fluoride salts are considered as primary candidates for nuclear fuel in the Molten Salt Reactor (MSR), one of the 6 generation IV nuclear reactor designs. In order to determine the safety limits and to access the properties of the potential fuel mixtures, thermodynamic studies are very important. This study is a combination of experimental work and thermodynamic modelling and focusses on the fluoride systems with alkaline and alkaline earth fluorides as matrix and ThF 4 , UF 4 and PuF 3 as fertile and fissile materials. The purification of the single components was considered as essential first step for the study of more complex systems and ternary phase diagrams were described using Differential Scanning Calorimetry (DSC) and drop calorimetry, which are used to measure phase transitions, enthalpy of mixing and heat capacity. In addition to the calorimetric techniques, Knudsen Effusion Mass Spectrometry (KEMS) and X-ray Diffraction (XRD) were used to collect data on vapour pressure and crystal structure of fluorides. The results are then coupled with thermodynamic modelling using the Calphad method for the assessment of the phase diagrams. A thermodynamic database describing the most important systems for MSR application has been developed and it has been used to optimize the fuel composition in view of the relevant properties such as melting temperature. A reliable database of thermodynamic properties of fluoride salts has been generated. It includes the key systems for the MSR fuel and it is very useful to predict the properties of the fuel

  19. Reactor structure and superconducting magnet system of ITER

    International Nuclear Information System (INIS)

    Tada, Eisuke; Yoshida, Kiyoshi; Shibanuma, Kiyoshi; Okuno, Kiyoshi; Tsuji, Hiroshi; Shimamoto, Susumu

    1993-01-01

    Fusion Experimental Reactors are one of the major steps toward realization of the fusion energy and the key objective are to demonstrate the scientific and technological feasibility prior to the Demo Fusion Reactor. ITER (International Thermonuclear Experimental Reactor) is one of experimental reactors and the conceptual design has been completed by the united efforts of USA, USSR, EC and Japan. In parallel with the conceptual design, key technology development in various areas has being conducted. This paper describes the overall design concepts and the latest technological achievements of the ITER reactor structure and superconducting magnet system. (author)

  20. Preliminary conceptual design and analysis on KALIMER reactor structures

    International Nuclear Information System (INIS)

    Kim, Jong Bum

    1996-10-01

    The objectives of this study are to perform preliminary conceptual design and structural analyses for KALIMER (Korea Advanced Liquid Metal Reactor) reactor structures to assess the design feasibility and to identify detailed analysis requirements. KALIMER thermal hydraulic system analysis results and neutronic analysis results are not available at present, only-limited preliminary structural analyses have been performed with the assumptions on the thermal loads. The responses of reactor vessel and reactor internal structures were based on the temperature difference of core inlet and outlet and on engineering judgments. Thermal stresses from the assumed temperatures were calculated using ANSYS code through parametric finite element heat transfer and elastic stress analyses. While, based on the results of preliminary conceptual design and structural analyses, the ASME Code limits for the reactor structures were satisfied for the pressure boundary, the needs for inelastic analyses were indicated for evaluation of design adequacy of the support barrel and the thermal liner. To reduce thermal striping effects in the bottom are of UIS due to up-flowing sodium form reactor core, installation of Inconel-718 liner to the bottom area was proposed, and to mitigate thermal shock loads, additional stainless steel liner was also suggested. The design feasibilities of these were validated through simplified preliminary analyses. In conceptual design phase, the implementation of these results will be made for the design of the reactor structures and the reactor internal structures in conjunction with the thermal hydraulic, neutronic, and seismic analyses results. 4 tabs., 24 figs., 4 refs. (Author)

  1. Annual report on the state of RB reactor components and equipment, December 1997

    International Nuclear Information System (INIS)

    Milosevic, M.

    1997-12-01

    According to the performed analysis, it is considered that the RB reactor can be operated safely until the existing control and safety systems could be maintained in satisfactory operable state. Failures of heavy water circulation system valves which may cause decreased availability but no accident. During 1997 the reactor lattice was not changed due to application of the coupled fast-thermal core HERBE. Total reactor operation amounted to 69.5 Wh with 66 start-ups (attained criticality levels). This report contains 4 Annexes, detailed description of the state of reactor equipment, plan for forming new HERBE core, plan for regular annual maintenance of the reactor, and plan for minimum needed resources for regular maintenance of the components and equipment in the forthcoming year

  2. Catalyst support structure, catalyst including the structure, reactor including a catalyst, and methods of forming same

    Science.gov (United States)

    Van Norman, Staci A.; Aston, Victoria J.; Weimer, Alan W.

    2017-05-09

    Structures, catalysts, and reactors suitable for use for a variety of applications, including gas-to-liquid and coal-to-liquid processes and methods of forming the structures, catalysts, and reactors are disclosed. The catalyst material can be deposited onto an inner wall of a microtubular reactor and/or onto porous tungsten support structures using atomic layer deposition techniques.

  3. Calandria cooling structure in pressure tube reactor

    International Nuclear Information System (INIS)

    Hyugaji, Takenori; Sasada, Yasuhiro.

    1976-01-01

    Purpose: To contrive the structure of a heavy water distributing device in a pressure tube reactor thereby to reduce the variation in the cooling function thereof due to the welding deformation and installation error. Constitution: A heating water distributing plate is provided at the lower part of the upper tubular plate of a calandria tank to form a heavy water distributing chamber between both plates and a plurality of calandria tubes. Heavy water which has flowed in the upper part of the heavy water distributing plate from the heavy water inlet nozzle flows down through gaps formed around the calandria tubes, whereby the cooling of the calandria tank and the calandria tubes is carried out. In the above described calandria cooling structure, a heavy water distributing plate support is provided to secure the heavy water distributing plate and torus-shaped heavy water distributing rings are fixed to holes formed in the heavy water distributing plate penetrating through the calandria tubes thereby to form torus-shaped heavy water outlet ports each having a space. (Seki, T.)

  4. Manufacture of heavy reactor components with particular considerations to quality assurance

    International Nuclear Information System (INIS)

    Kreppel, H.; Clausmeyer, H.

    1980-01-01

    The use of adequate quality assurance measures is one of the most important prerequisites for the manufacture of reactor components. Nature and extent of the quality assurance system at present adopted in the Federal Republic of Germany are illustrated, using the manufacture of a reactor pressure vessel as an example. The system comprises quality organization, planning of all quality assurance measures, quality surveillance through all stages of manufacture and documentation of quality attained. (orig.)

  5. Manufacture of heavy reactor components with particular consideration to quality assurance

    International Nuclear Information System (INIS)

    Clausmeyer, H.; Kreppel, H.

    1977-01-01

    The use of adequate quality assurance measures is one of the most important prerequisites for the manufacture of reactor components. Nature and extent of the quality assurance system at present adopted in the Federal Republic of Germany are illustrated, using the manufacture of a reactor pressure vessel as an example. The system comprises quality organization, planning of all quality assurance measures, quality surveillance through all stages of manufacture and documentation of quality attained. (orig.) [de

  6. Manufacture of heavy reactor components with particular consideration to quality assurance

    International Nuclear Information System (INIS)

    Kreppel, H.; Clausmeyer, H.

    1981-01-01

    The use of adequate quality assurance measures is one of the most important prerequisites for the manufacture of reactor components. Nature and extent of the quality assurance system at present adopted in the Federal Republic of Germany are illustrated, using the manufacture of a reactor pressure vessel as an example. The system comprises quality organization, planning of all quality assurance measures, quality surveillance through all stages of manufacture and documentation of quality attained. (orig.)

  7. Component and Technology Development for Advanced Liquid Metal Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Anderson, Mark [Univ. of Wisconsin, Madison, WI (United States)

    2017-01-30

    The following report details the significant developments to Sodium Fast Reactor (SFR) technologies made throughout the course of this funding. This report will begin with an overview of the sodium loop and the improvements made over the course of this research to make it a more advanced and capable facility. These improvements have much to do with oxygen control and diagnostics. Thus a detailed report of advancements with respect to the cold trap, plugging meter, vanadium equilibration loop, and electrochemical oxygen sensor is included. Further analysis of the university’s moving magnet pump was performed and included in a section of this report. A continuous electrical resistance based level sensor was built and tested in the sodium with favorable results. Materials testing was done on diffusion bonded samples of metal and the results are presented here as well. A significant portion of this work went into the development of optical fiber temperature sensors which could be deployed in an SFR environment. Thus, a section of this report presents the work done to develop an encapsulation method for these fibers inside of a stainless steel capillary tube. High temperature testing was then done on the optical fiber ex situ in a furnace. Thermal response time was also explored with the optical fiber temperature sensors. Finally these optical fibers were deployed successfully in a sodium environment for data acquisition. As a test of the sodium deployable optical fiber temperature sensors they were installed in a sub-loop of the sodium facility which was constructed to promote the thermal striping effect in sodium. The optical fibers performed exceptionally well, yielding unprecedented 2 dimensional temperature profiles with good temporal resolution. Finally, this thermal striping loop was used to perform cross correlation velocimetry successfully over a wide range of flow rates.

  8. Liquid metal systems development: reactor vessel support structure evaluation

    International Nuclear Information System (INIS)

    McEdwards, J.A.

    1981-01-01

    Results of an evaluation of support structures for the reactor vessel are reported. The U ring, box ring, integral ring, tee ring and tangential beam supports were investigated. The U ring is the recommended vessel support structure configuration

  9. 10 CFR 50.69 - Risk-informed categorization and treatment of structures, systems and components for nuclear...

    Science.gov (United States)

    2010-01-01

    ..., systems and components for nuclear power reactors. (a) Definitions. Risk-Informed Safety Class (RISC)-1... functions. Risk-Informed Safety Class (RISC)-2 structures, systems and components (SSCs) means nonsafety-related SSCs that perform safety significant functions. Risk-Informed Safety Class (RISC)-3 structures...

  10. Application of HOLOSAFT for nondestructive testing of reactor components

    International Nuclear Information System (INIS)

    Schmitz, V.; Mueller, W.; Schaefer, G.; Graeber, B.; Hoppstaedter, K.

    1985-01-01

    The aim of the project was to develop a superimposed ultrasonic test process, or to combine existing ones, so that a classification and three dimensional representation of defects is made possible. Two analytic test processes - ultrasonic holography and SAFT (synthetic aperture focussing technique) are combined, using identical hardware components and developing common software packages to create an imaging process called HOLOSAFT. The high possible lateral resolution of ultrasonic holography parallel to the test sample surface is used, together with the high possible axial resolution of the SAFT process at right angles to the surface, in order to make measurement of defects possible in three coordinate directions. The development of the process is described in detail, where, based on physical-mathematical bases, the equipment and software developed for pulse echo and tandem arrangements are discussed. The possible resolution is examined in laboratory experiments as a function of the test head diameter, the picture is examined as a function of the aperture length and the picture quality is examined as a function of the ultrasonic devices and defect orientation. Other chapters are concerned with measuring the defect depth, the determination of inclined positions, multi-angle sounding and examination of components with curved surfaces. The results show the great capacity for analysis of the HOLOSAFT process and its suitability for application in nuclear power stations. (orig./HP) [de

  11. Flaw assessment procedure for high temperature reactor components

    International Nuclear Information System (INIS)

    Ainsworth, R.A.; Takahashi, Y.

    1990-01-01

    An interim high-temperature flaw assessment procedure is described. This is a result of a collaborative effort between Electric Power Research Institute in the USA, Central Research Institute of Electric Power Industry in Japan, and Nuclear Electric plc in the UK. The procedure addresses preexisting defects subject to creep-fatigue loading conditions. Laws employed to calculate the crack growth per cycle are defined in terms of fracture mechanics parameters and constants related to the component material. The crack growth laws may be integrated to calculate the remaining life of a component or to predict the amount of crack extension in a given period. Fatigue and creep crack growth per cycle are calculated separately, and the total crack extension is taken as the simple sum of the two contributions. An interaction between the two propagation modes is accounted for in the material properties in the separate calculations. In producing the procedure, limitations of the approach have been identified. Some of these limitations are to be addressed in an extension of the current collaborative program. 20 refs

  12. Ageing Management of the Kinshasa Trico II Research Reactor Components and Structures. A Case Study of the 5 Tonne Overhead Travelling Crane and the Ventilation System Inlet and Outlet

    Energy Technology Data Exchange (ETDEWEB)

    Kombele, D. G.; Kankunku-K, P.; Mwamba, V. L.; Kobakozete, J. I.; Kiamana, M. M.; Lukibanza, J. W.; Kalala, A. T.; Mfinda, D. M.; Bilulu, T. L.; Ilunga, S. T. [Reactor Service Technical Department, Regional Nuclear Research Centre of Kinshasa (CREN-K), Atomic Energy General Commission (CGEA), Kinshasa (Congo, The Democratic Republic of the)

    2014-08-15

    The cable isolation sheath of the overhead traveling crane became fragile and brittle on a length of more than two meters. This degradation of the sheath has been caused by the ageing of the cable and the effects of heat, with changeable ambient temperatures in the reactor hall combined with the Joule effect, in relation to the cyclic use of the crane. This ageing effect was discovered when a failure occurred at the end of a nearly completed routine operation of displacing slightly radioactive spent ion exchange resins into waste storage. During the return of the hoist to its rest position, a severe short circuit happened between the cable and the mass of the hoist motor support, followed by a strong detonation that produced sparks and immobilized the overall system. A visual examination of the cable showed a change of its physical properties as mentioned above. A further investigation showed that two master contacts of the hoist were also burned. The bridge crane was inspected and certified in January 2010 by a competent authority eleven months before the event. On the other hand, the reactor ventilation system started presenting its limits in the late 1990s after more than 17 years of operation. Many failures occurred in the extraction motors and the air conditioning system, causing both a temperature increase and a lack of negative pressure in the reactor hall. The crane problem was solved after replacing the damaged cable and the two burned contacts. A review of the overall status of the bridge crane by the licensing authority is scheduled before the end of 2011. The paper also describes steps related to the renewal of the air inlet system and the restoration of negative pressure in the reactor hall. (author)

  13. Recent developments for fast reactor structural design standard (FDS)

    International Nuclear Information System (INIS)

    Kasahara, N.; Nakamuria, K.; Morishita, M.; Shibamoto, H.; Nagashima, H.; Inoue, K.

    2005-01-01

    For realization of reliable and economical fast reactor (FR) plants, Japan Nuclear Cycle Development Institute(JNC) and Japan Atomic Power Company(JAPC) are cooperating on 'Feasibility Study on Commercialized FR Cycle Systems'. To certify the design concepts through evaluation of their structural integrity, the research and development of 'Elevated Temperature Structural Design Guide for Commercialized Fast Reactor (FDS)' is recognized as an essential theme. FDS focuses on particular failure modes of FRs such as ratchet deformation and creep fatigue damages due to cyclic thermal loads. To evaluate these modes, three main developments are in progress. One is 'Refinement of Failure Criteria' for particular modes of FRs. Next is development of 'Guidelines for Inelastic Design Analysis' in order to predict elastic plastic and creep behaviors. Furthermore, efforts are being made toward preparing 'Guidelines for Thermal Load Modeling' for FR component design where thermal loads are dominant. These studies were performed under the sponsorship of the Ministry of Economy, Trade and Industry of Japanese government. (authors)

  14. Exchange method for reactor inner structural member

    Energy Technology Data Exchange (ETDEWEB)

    Tsujimura, Hiroshi; Kurosawa, Koichi; Ono, Shigeki; Uozumi, Hiroto; Takada, Ko; Watanabe, Yoshio; Ito, Masato; Yoshie, Yutaka [Hitachi Ltd., Tokyo (Japan); Nihei, Ken-ichi

    1996-09-13

    A dryer and a shroud head are removed from the inside of a reactor pressure vessel (RPV) of a BWR type reactor, and they are stacked in a dryer and steam separator pool (DSP). Next, fuel assemblies, fuel support fittings, control rods and control rode guide tubes are successively removed and stored in an exclusive storage vessel. Then, guide rods are removed by cutting and temporarily placed in the DSP. Then, an upper lattice plate and a reactor core support plate are successively removed and temporarily placed in the DSP. Reactor core spray pipes are removed by cutting and temporarily placed in the DSP. Then, a shroud support cylinder is cut, and the shroud is removed and temporarily placed in the DSP. Subsequently, reactor water is drained, and a reactor core shroud to which the upper lattice plate and the reactor core support plate are previously disposed is suspended in the RPV, and the existent shroud support cylinder and the new reactor core shroud are welded. (I.N.)

  15. Development of plasma facing components for fusion experimental reactors

    Energy Technology Data Exchange (ETDEWEB)

    Onozuka, M.; Fujiya, Y.; Inoue, M.; Morimoto, M. [Mitsubishi Heavy Industries, Ltd., Yokohama (Japan)

    1995-12-31

    The divertor structure and fabrication process have been investigated, including the structures of the divertor elements and support, fundamental brazing techniques, brazing of large divertor tiles and fabrication method of large divertor modules. Using direct brazing, a partial divertor module with large CFC tiles was fabricated and tested. It was shown that the model had sufficient structural integrity against thermal shocks of {approximately}17MW/m{sup 2} {times} 4 sec for up to 1,600 times. A fabrication technique for large and complex-shaped divertor module has been developed and successfully applied to a 1m-long linear and 0.8m-long curved divertor modules. In addition, preliminary investigation of direct brazing of beryllium to the copper substrate has been conducted. It was found that the bending strength of the bonded materials was around 40 MPa. Furthermore, boron coating on the CFC and Mo has been examined. Using the boron ion implantation technique, boron ions were implanted to the CFC and Mo plates prior to the boron atoms deposition. The samples fabricated with this method were found to have a sufficient thermal shock resistance.

  16. Nuclear Reactor Component Code CUPID-I: Numerical Scheme and Preliminary Assessment Results

    International Nuclear Information System (INIS)

    Cho, Hyoung Kyu; Jeong, Jae Jun; Park, Ik Kyu; Kim, Jong Tae; Yoon, Han Young

    2007-12-01

    A component scale thermal hydraulic analysis code, CUPID (Component Unstructured Program for Interfacial Dynamics), is being developed for the analysis of components of a nuclear reactor, such as reactor vessel, steam generator, containment, etc. It adopted three-dimensional, transient, two phase and three-field model. In order to develop the numerical schemes for the three-field model, various numerical schemes have been examined including the SMAC, semi-implicit ICE, SIMPLE, Row Scheme and so on. Among them, the ICE scheme for the three-field model was presented in the present report. The CUPID code is utilizing unstructured mesh for the simulation of complicated geometries of the nuclear reactor components. The conventional ICE scheme that was applied to RELAP5 and COBRA-TF, therefore, were modified for the application to the unstructured mesh. Preliminary calculations for the unstructured semi-implicit ICE scheme have been conducted for a verification of the numerical method from a qualitative point of view. The preliminary calculation results showed that the present numerical scheme is robust and efficient for the prediction of phase changes and flow transitions due to a boiling and a flashing. These calculation results also showed the strong coupling between the pressure and void fraction changes. Thus, it is believed that the semi-implicit ICE scheme can be utilized for transient two-phase flows in a component of a nuclear reactor

  17. Nuclear Reactor Component Code CUPID-I: Numerical Scheme and Preliminary Assessment Results

    Energy Technology Data Exchange (ETDEWEB)

    Cho, Hyoung Kyu; Jeong, Jae Jun; Park, Ik Kyu; Kim, Jong Tae; Yoon, Han Young

    2007-12-15

    A component scale thermal hydraulic analysis code, CUPID (Component Unstructured Program for Interfacial Dynamics), is being developed for the analysis of components of a nuclear reactor, such as reactor vessel, steam generator, containment, etc. It adopted three-dimensional, transient, two phase and three-field model. In order to develop the numerical schemes for the three-field model, various numerical schemes have been examined including the SMAC, semi-implicit ICE, SIMPLE, Row Scheme and so on. Among them, the ICE scheme for the three-field model was presented in the present report. The CUPID code is utilizing unstructured mesh for the simulation of complicated geometries of the nuclear reactor components. The conventional ICE scheme that was applied to RELAP5 and COBRA-TF, therefore, were modified for the application to the unstructured mesh. Preliminary calculations for the unstructured semi-implicit ICE scheme have been conducted for a verification of the numerical method from a qualitative point of view. The preliminary calculation results showed that the present numerical scheme is robust and efficient for the prediction of phase changes and flow transitions due to a boiling and a flashing. These calculation results also showed the strong coupling between the pressure and void fraction changes. Thus, it is believed that the semi-implicit ICE scheme can be utilized for transient two-phase flows in a component of a nuclear reactor.

  18. Conceptual design of nuclear fusion power reactor DREAM. Reactor structures and remote maintenance

    International Nuclear Information System (INIS)

    Nishio, Satoshi; Seki, Yasushi; Ueda, Shuzo; Kurihara, Ryoichi; Adachi, Junichi; Yamazaki, Seiichiro; Hashimoto, Toshiyuki.

    1997-01-01

    Nuclear fusion reactors are required to be able to compete another energy sources in economy, reliability, safety and environmental integrity for commercial use. In the DREAM (DRastically EAsy Maintenance) reactor, a very low activated material of SiC/SiC composite has been introduced for the structural material, a reactor configuration for very easy maintenance and the helium gas of a high temperature for the cooling system, and hence DREAM has been proven to be very attractively as the commercial power reactor due to the high availability and efficiency of the plant and minimization of radioactive wastes. (author)

  19. Space Fission Reactor Structural Materials: Choices Past, Present and Future

    International Nuclear Information System (INIS)

    Busby, Jeremy T.; Leonard, Keith J.

    2007-01-01

    Nuclear powered spacecraft will enable missions well beyond the capabilities of current chemical, radioisotope thermal generator and solar technologies. The use of fission reactors for space applications has been considered for over 50 years, although, structural material performance has often limited the potential performance of space reactors. Space fission reactors are an extremely harsh environment for structural materials with high temperatures, high neutron fields, potential contact with liquid metals, and the need for up to 15-20 year reliability with no inspection or preventative maintenance. Many different materials have been proposed as structural materials. While all materials meet many of the requirements for space reactor service, none satisfy all of them. However, continued development and testing may resolve these issues and provide qualified materials for space fission reactors.

  20. Nuclear heat source component design considerations for HTGR process heat reactor plant concept

    International Nuclear Information System (INIS)

    McDonald, C.F.; Kapich, D.; King, J.H.; Venkatesh, M.C.

    1982-05-01

    The coupling of a high-temperature gas-cooled reactor (HTGR) and a chemical process facility has the potential for long-term synthetic fuel production (i.e., oil, gasoline, aviation fuel, hydrogen, etc) using coal as the carbon source. Studies are in progress to exploit the high-temperature capability of an advanced HTGR variant for nuclear process heat. The process heat plant discussed in this paper has a 1170-MW(t) reactor as the heat source and the concept is based on indirect reforming, i.e., the high-temperature nuclear thermal energy is transported [via an intermediate heat exchanger (IHX)] to the externally located process plant by a secondary helium transport loop. Emphasis is placed on design considerations for the major nuclear heat source (NHS) components, and discussions are presented for the reactor core, prestressed concrete reactor vessel (PCRV), rotating machinery, and heat exchangers

  1. Transient temperature response of in-vessel components due to pulsed operation in tokamak fusion experimental reactor (FER)

    International Nuclear Information System (INIS)

    Minato, Akio; Tone, Tatsuzo

    1985-12-01

    A transient temperature response of the in-vessel components (first wall, blanket, divertor/limiter and shielding) surrounding plasma in Tokamak Fusion Experimental Reactor (FER) has been analysed. Transient heat load during start up/shut down and pulsed operation cycles causes the transient temperature response in those components. The fatigue lifetime of those components significantly depends upon the resulting cyclic thermal stress. The burn time affects the temperature control in the solid breeder (Li 2 O) and also affects the thermo-mechanical design of the blanket and shielding which are constructed with thick structure. In this report, results of the transient temperature response obtained by the heat transfer and conduction analyses for various pulsed operation scenarios (start up, shut down, burn and dwell times) have been investigated in view of thermo-mechanical design of the in-vessel components. (author)

  2. Specialists' meeting on heat exchanging components of gas-cooled reactors

    International Nuclear Information System (INIS)

    1984-01-01

    The objective of the Meeting sponsored by IAEA was to provide a forum for the exchange and discussion of technical information related to heat exchanging and heat conducting components for gas-cooled reactors. The technical part of the meeting covered eight subjects: Heat exchanging components for process heat applications, design and requirements, and research and development programs; Status of the design and construction of intermediate He/He exchangers; Design, construction and performance of steam generators; Metallic materials and design codes; Design and construction of valves and hot gas ducts; Description of component test facilities and test results; Manufacturing of heat exchanging components

  3. Specialists' meeting on heat exchanging components of gas-cooled reactors

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1984-07-01

    The objective of the Meeting sponsored by IAEA was to provide a forum for the exchange and discussion of technical information related to heat exchanging and heat conducting components for gas-cooled reactors. The technical part of the meeting covered eight subjects: Heat exchanging components for process heat applications, design and requirements, and research and development programs; Status of the design and construction of intermediate He/He exchangers; Design, construction and performance of steam generators; Metallic materials and design codes; Design and construction of valves and hot gas ducts; Description of component test facilities and test results; Manufacturing of heat exchanging components.

  4. Conceptual design of blanket structures for fusion experimental reactor (FER)

    International Nuclear Information System (INIS)

    1984-03-01

    Conceptual design study for in-vessel components including tritium breeding blanket of FER has been carried out. The objective of this study is to obtain the engineering and technological data for selecting the reactor concept and for its construction by investigating fully and broadly. The design work covers in-vessel components (such as tritium breeding blanket, first wall, shield, divertor and blanket test module), remote handling system and tritium system. The designs of those components and systems are accomplished in consideration of their accomodation to whole reactor system and problems for furthur study are clarified. (author)

  5. Nuclear reactor having an inflatable vessel closure seal structure

    International Nuclear Information System (INIS)

    1980-01-01

    An improved type of closure head seal for the rotatable plugs of the reactor vessel of a liquid metal fast breeder reactor is described. The seal prevents the release of radioactive particles while allowing the plug to be rotated without major manipulation of the seal structure. (UK)

  6. Operation experience at the Neuherberg Research Reactor (FRN) with several modifications of reactor components

    Energy Technology Data Exchange (ETDEWEB)

    Demmeler, M; Rau, G [Gesellschaft fuer Strahlen- und Umweltforschung mbH, Neuherberg (Germany)

    1974-07-01

    Since the first full power operation in September 1972 up till now (Dec. 1973) the TRIGA Mark III reactor FRN has run more than 500 MWh in steady state operation and has been pulsed for 265 times. During startup experiments, neutron- and gamma-flux mapping has been performed with special technical devices in the core and in several irradiation positions, mainly in the thermal column and in the exposure room. Furthermore reactivity values of each fuel element have been measured at full power of 1 MW, thus enabling a more accurate burnup calculation. Troubles with the rotary specimen rack occurred at power rates above 280 kW; here, the lazy susan stuck, caused by thermal stress. Thus it will be replaced by a hydraulic-operated type, which has been developed at the TRIGA reactor Heidelberg. In order to increase irradiation capacity, a new core configuration has been set up a few months ago, replacing several fuel-reflector-elements by irradiation tubes within the grid-plate positions E-22, G-2, G-17 and G-36. Four additional fuel elements had to be inserted to compensate for the resulting reactivity losses. The original plan of regaining sufficient excess-reactivity by inserting a fuel element in grid-plate position A-l failed because of local boiling in the center of the core by 1 MW-operation. Experiments at the reactor started with the begin of routine-operation in September 1973. Up till now, a total of 450 neutron- and gamma- irradiations have been performed, mainly for neutron-activations. (author)

  7. Residual strength evaluation of concrete structural components ...

    Indian Academy of Sciences (India)

    fundamental material parameters that can be determined for use in design or evaluation. ... of plain and reinforced concrete beams using fracture mechanics principles. Design equations ... components accounting for tension softening effect.

  8. On the major ductile fracture methodologies for failure assessment of nuclear reactor components

    International Nuclear Information System (INIS)

    Cruz, Julio R.B.; Andrade, Arnaldo H.P. de; Landes, John D.

    1996-01-01

    In structures like nuclear reactor components there is a special concern with the loads that may occur under postulated accident conditions. These loads can cause the stresses to go well beyond the linear elastic limits, requiring the use of ductile fracture mechanics methods to the prediction of the structure behavior. Since the use of numerical methods to apply EPFM concepts is expensive and time consuming, the existence of analytical engineering procedures are of great relevance. The lack of precision in detail, as compared with numerical nonlinear analyses, is compensated by the possibility of quick failure assessments. This is a determinant factor in situations where a systematic evaluation of a large range of geometries and loading conditions is necessary, like in thr application of the Leak-Before-Break (LBB) concept on nuclear piping. This paper outlines four ductile fracture analytical methods, pointing out positive and negative aspects of each one. The objective is to take advantage of this critical review to conceive a new methodology, one that would gather strong points of the major existent methods and would try to eliminate some of their drawbacks. (author)

  9. Application of the regulations on pressurized components or light water reactor primary coolant circuits

    International Nuclear Information System (INIS)

    Barthelemy, F.; Menjon, G.

    1977-01-01

    This paper describes the philosophy and the provisions of the Order of 26 February 1974 concerning application of the regulations on pressurized components for light water reactor steam supply systems. The aim is to show how these regulations which differ from other regulations on pressurized components and is more detailed on many points, is applied in practice in France in the various stages of the design, construction and operation of PWRs. (NEA) [fr

  10. COMPARISON OF COOLING SCHEMES FOR HIGH HEAT FLUX COMPONENTS COOLING IN FUSION REACTORS

    Directory of Open Access Journals (Sweden)

    Phani Kumar Domalapally

    2015-04-01

    Full Text Available Some components of the fusion reactor receives high heat fluxes either during the startup and shutdown or during the operation of the machine. This paper analyzes different ways of enhancing heat transfer using helium and water for cooling of these high heat flux components and then conclusions are drawn to decide the best choice of coolant, for usage in near and long term applications.

  11. Earthquake-proof supporting structure in reactor vessel

    International Nuclear Information System (INIS)

    Sakurai, Akio; Sekine, Katsuhisa; Madokoro, Manabu; Katoono, Shin-ichi; Konno, Mutsuo; Suzuki, Takuro.

    1990-01-01

    Conventional earthquake-proof structure comprises a vessel vibration stopper integrated to a reactor vessel, powder for restricting the horizontal displacements, a safety vessel surrounds the outer periphery of the reactor vessel and a safety vessel vibration stopper integrated therewith, which are fixed to buildings. However, there was a problem that a great amount of stresses are generated in the base of the reactor vessel vibration stopper due to reaction of the powders which restrict thermal expansion. In order to remarkably reduce the reaction of the powers, powders are charged into a spaces formed between each of the reactor vessel vibration stopper, the safety vessel vibration stopper and the flexible member disposed between them. According to this constitution, the reactor vessel vibration stopper does not undergo a great reaction of the powers upon thermal expansion of the reactor vessel to moderate the generated stresses, maintain the strength and provide earthquake-proof supporting function. (N.H.)

  12. The efficiency of two anaerobic reactor components; Eficiencias de dos componentes de un reactor anaerobio

    Energy Technology Data Exchange (ETDEWEB)

    Vazquez Borges, E.; Mendez Novelo, R.; Magana Pietra, A. [Facultad de Ingenieria. Universidad de Yucatan (Mexico); Martinez Pereda, P.; Fernandez Villagomez, G. [Universidad Nacional Autonoma de Mexico. Division de Estudios de posgrado de la Facultad de Ingenieria. Mexico (Mexico)

    1997-09-01

    This study examined the behaviour of an anaerobic digester in treating pig farm sewage. The experimental model consisted of a UASB reactor at the bottom and a high-rate sedimentator at the top with a total capacity of 534 litres. The digester was installed on a pig farm and its performance under different operating conditions was determined, with hydraulic retention time (HRT) as the critical parameter for evaluating the anaerobic system`s efficiency. The results obtained during the experiment to establish the critical operating parameters are reported. The organic loads applied for a HRT of 1 day were 7.3 kg/m``3/day of total DQO and 3 kg/m``3/day of soluble DQO, following organic matter removal rates (as total DQO) of 36% and 49% respectively and removal rates (as soluble DQO) of 74% in the UASB and 8% in the sedimentator. The efficiency of the reactor as a whole at this HRT time was a removal rate of 74% of total DQO and 75% of soluble DQO. (Author) 25 refs.

  13. Analysis of Removal Alternatives for the Heavy Water Components Test Reactor at the Savannah River Site

    International Nuclear Information System (INIS)

    Owen, M.B.

    1996-08-01

    This engineering study was developed to evaluate different options for decommissioning of the Heavy Water Components Test Reactor (HWCTR) at the Savannah River Site. This document will be placed in the DOE-SRS Area reading rooms for a period of 30 days in order to obtain public input to plans for the demolition of HWCTR

  14. Development and application of computer codes for multidimensional thermalhydraulic analyses of nuclear reactor components

    International Nuclear Information System (INIS)

    Carver, M.B.

    1983-01-01

    Components of reactor systems and related equipment are identified in which multidimensional computational thermal hydraulics can be used to advantage to assess and improve design. Models of single- and two-phase flow are reviewed, and the governing equations for multidimensional analysis are discussed. Suitable computational algorithms are introduced, and sample results from the application of particular multidimensional computer codes are given

  15. 10 CFR 110.26 - General license for the export of nuclear reactor components.

    Science.gov (United States)

    2010-01-01

    ... 10 Energy 2 2010-01-01 2010-01-01 false General license for the export of nuclear reactor components. 110.26 Section 110.26 Energy NUCLEAR REGULATORY COMMISSION (CONTINUED) EXPORT AND IMPORT OF... Denmark Finland France Germany Greece Indonesia Ireland Italy Japan Latvia Lithuania Luxembourg...

  16. Thermal aging of some decommissioned reactor components and methodology for life prediction

    International Nuclear Information System (INIS)

    Chung, H.M.

    1989-03-01

    Since a realistic aging of cast stainless steel components for end-of-life or life-extension conditions cannot be produced, it is customary to simulate the thermal aging embrittlement by accelerated aging at ∼400 degree C. In this investigation, field components obtained from decommissioned reactors have been examined after service up to 22 yr to provide a benchmark of the laboratory simulation. The primary and secondary aging processes were found to be identical to those of the laboratory-aged specimens, and the kinetic characteristics were also similar. The extent of the aging embrittlement processes and other key factors that are known to influence the embrittlement kinetics have been compared for the decommissioned reactor components and materials aged under accelerated conditions. On the basis of the study, a mechanistic understanding of the causes of the complex behavior in kinetics and activation energy of aging (i.e., the temperature dependence of aging embrittlement between the accelerated and reactor-operating conditions) is presented. A mechanistic correlation developed thereon is compared with a number of available empirical correlations to provide an insight for development of a better methodology of life prediction of the reactor components. 18 refs., 18 figs., 5 tabs

  17. Dynamic interaction of components, structure, and foundation of nuclear power facilities

    International Nuclear Information System (INIS)

    Pajuhesh, J.; Hadjian, A.H.

    1977-01-01

    A solution is formulated for the dynamic analysis of structures and components with different stiffness and damping characteristics, including the consideration of soil-structure interaction effects. Composite structures are often analysed approximately, in particular with regards to damping. For example, the reactor and other equipment in nuclear power plant structures are often analysed by assuming them uncoupled from the supporting structures. To achieve a better accuracy, the coupled system is hereby analysed as a composite component-structure-soil system. To demonstrate the assembly technique, two examples are considered: (a) a steel structure sitting on a concrete stem and linked by a steel bridge to another concrete structure, and (b) an actual model of a nuclear power plant containment structure. (Auth.)

  18. Aging Management Strategy and Requirements of Pressurized Water Reactor Internal Components

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Jun Seog; Oh, Sung Jin; Won, Se Yol; Jeong, Sun Mi [KHNP, Daejeon (Korea, Republic of)

    2016-05-15

    The demonstration that the effects of degradation in the components of PWR internals are adequately managed is essential for maintaining a healthy fleet and ensuring the continued functionality of the reactor internals. It is also very important to determine when and where irradiation susceptibility may occur for the continued operation. This paper introduces the aging management strategies and requirements for PWR internals components and discusses effects of irradiation aging results from the functionality assessments based on the categorization of internal components. This paper introduces aging management strategies and requirements for PWR internals components. The aging management requirements for PWR internals are specified in four final component groups, which are Primary, Expansion, Existing Program and No Additional Measures. Among these groups, Primary groups include any restriction on general applicability, degradation mechanism, forward link to any Expansion components, examination method, initial examination and frequency, and examination coverage and accessibility. Expansion groups are backward link to the Primary component.

  19. Reliability Prediction Of System And Component Of Process System Of RSG-GAS Reactor

    International Nuclear Information System (INIS)

    Sitorus Pane, Jupiter

    2001-01-01

    The older the reactor the higher the probability of the system and components suffer from loss of function or degradation. This phenomenon occurred because of wear, corrosion, and fatigue. Study on component reliability was generally performed deterministically and statistically. This paper would describe an analysis of using statistical method, i.e. regression Cox, in order to predict the reliability of the components and their environmental influence's factors. The result showed that the dynamics, non safety related, and mechanic components have higher risk of failure, whereas static, safety related, and electric have lower risk of failures. The relative risk value for variable of components dynamics, quality, dummy 1 and dummy 2 are of 1.54, 1.59, 1.50, and 0.83 compare to other components type with each variable. Component with the higher risk have lower reliability than lower one

  20. Internal structure of reactor building for Madras Atomic Power Project

    International Nuclear Information System (INIS)

    Pandit, D.P.

    1975-01-01

    The structural configuration and analysis of structural elements of the internal structure of reactor building for the Madras Atomic Power Project has been presented. Two methods of analysis of the internal structure, viz. Equivalent Plane Frame and Finite Element Method, are explained and compared with the use of bending moments obtained. (author)

  1. Centralized Reliability Data Organization (CREDO) assessment of critical component unavailability in liquid metal reactors

    International Nuclear Information System (INIS)

    Koger, K.H.; Haire, M.J.; Humphrys, B.L.; Manneschmidt, J.F.; Setoguchi, K.; Nakai, R.

    1988-01-01

    The Centralized Reliability Data Organization (CREDO) is the largest repository of liquid metal reactor (LMR) component reliability data in the world. It is jointly sponsored by the US Dept. of Energy (DOE) and the Power Reactor and Nuclear Fuel Development Corporation (PNC) of Japan. The CREDO data base contains information on a population of more than 20,000 components and approximately 1500 event records. A conservative estimation is that the total component operating hours is approaching 2.2 billion hours. The work reported here focuses on the availability information contained in CREDO and the development of availability critical items lists. That is, individual components are ranked in prioritized lists from worst to best performers from an availability standpoint. Availability as used here is an inherent characteristics of the component and is not necessarily related to plant operability. A major observation is that a few components have a much higher unavailability factor than the average. The top fifteen components contribute 93%, 77%, and 87% of the total system unavailability for EBR-II, FFTF, and JOYO respectively. Critical components common to all three sites are mechanical pumps and electromagnetic pumps. Application of resources to these components with the highest unavailability will have the greatest effect on overall availability. All three sites demonstrate that low maintainability (i.e., long repair times), rather than unreliability (i.e., high failure rates), are the main contributors, by about a two-to-one margin, to liquid metal system unavailability

  2. Nuclear data for structural materials of fission and fusion reactors

    International Nuclear Information System (INIS)

    Goulo, V.

    1989-06-01

    The document presents the status of nuclear reaction theory concerning optical model development, level density models and pre-equilibrium and direct processes used in calculation of neutron nuclear data for structural materials of fission and fusion reactors. 6 refs

  3. A model for structural analysis of nuclear reactor pressure vessel flanges

    International Nuclear Information System (INIS)

    Oliveira, C.A. de.

    1987-01-01

    Due to the recent Brazilian advances in the nuclear technology area, it has been necessary the development of design and analysis methods for pressurized water reactor components, also as other components of a nuclear plant. This work proposes a methodology for the structural analysis of large diameter nuclear reactor pressure vessel flanges. In the analysis the vessel is divided into shell-of-revolution elements, the flanges are represented by rigid rings, and the bolts are treated as beams. The flexibility method is used for solving the problem. A computer program is shown, and the given results (displacements and stresses) are compared with results obtained by the finite element method. Although developed for nuclear reactor pressure vessel calculations, the program is more general, being possible its use for the analysis of any structure composed by shells of revolution. (author)

  4. Reactor pressure vessel and reactor coolant circuit cast duplex stainless steel components contribution of the expertise for life management studies

    International Nuclear Information System (INIS)

    Bezdikian, Georges

    2006-09-01

    The life management of French Nuclear Power Plants is a major stake from an economic and a technical point of view considering the aging management assessment of the key components of the plant. The actual life evaluation is the result of prediction of life assessment from important program of expertise for the 3-loop PWR and 4-loop PWR plants in operation. To optimize the strategic policy in order to achieve the best possible performance and to prepare the technical and economical choice and decision, the paper presents the association of life management strategy and the program of expertise considering: - the identification of degradation for different components and prediction criteria proposed; - the large database from cast reactor coolant and component removed from nuclear power plants and expertise studies to confirm the prediction; - the life evaluation of RPV with radiation surveillance program based on the expertise of irradiation capsules, it is particularly shown how the expertise is in the center of the strategic choice. The French utility has organized the life management of nuclear plant as a function of several programs of expertise of knowledge on the long term experience feedback and the maintenance program for life. This paper shows updated on RPV and reactor coolant equipment activities engaged by utility on: - periodic maintenance and volume of expertise; - Alternative maintenance actions; - Large volume of expertise and how are managed these results to predict the aging management. (author)

  5. Mirror fusion reactor design

    International Nuclear Information System (INIS)

    Neef, W.S. Jr.; Carlson, G.A.

    1979-01-01

    Recent conceptual reactor designs based on mirror confinement are described. Four components of mirror reactors for which materials considerations and structural mechanics analysis must play an important role in successful design are discussed. The reactor components are: (a) first-wall and thermal conversion blanket, (b) superconducting magnets and their force restraining structure, (c) neutral beam injectors, and (d) plasma direct energy converters

  6. Experience of partial dismantling and large component removal of light water reactors

    International Nuclear Information System (INIS)

    Dubourg, M.

    1987-01-01

    Not any of the French PWR reactors need to be decommissioned before the next decade or early 2000. However, feasibility studies of decommissioning have been undertaken and several dismantling scenarios have been considered including the dismantling of four PWR units and the on-site entombment of the active components into a reactor building for interim disposal. In addition to theoretical evaluation of radwaste volume and activity, several operations of partial dismantling of active components and decontamination activities have been conducted in view of dismantling for both PWR and BWR units. By analyzing the concept of both 900 and 1300 MWe PWR's, it appears that the design improvements taken into account for reducing occupational dose exposure of maintenance personnel and the development of automated tools for performing maintenance and repairs of major components, contribute to facilitate future dismantling and decommissioning operations

  7. Manufacturing requirements of reactor assembly components for PFBR (Paper No. 041)

    International Nuclear Information System (INIS)

    Murty, C.G.K.; Bhoje, S.B.

    1987-02-01

    This paper enumerates the requirements of 500 MWe Prototype Fast Breeder Reactor (PFBR) components and considering the present state of art of Indian industry an analysis is made on the challenges to be faced in manufacture highlighting the areas needing development. The large sizes and weights of the components coupled with the limitations on shop facilities and ODC transport, demand part of the fabrication to be done at shop and balance assembly work as well as certain assembly machining operations to be done at site work shop. The stringent geometrical tolerances coupled with extensive destructive and non-destructive examinations call for balanced and low heat input welding techniques and special inspection equipment like electronic co-ordinate determination system. The present paper deals with the specific manufacturing problems of the main reactor components. (author)

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

    International Nuclear Information System (INIS)

    Khoury, G.A.

    1995-01-01

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

  9. The role of materials in the analysis of fast breeder reactor components

    International Nuclear Information System (INIS)

    Aubert, Michel; Petrequin, Pierre.

    1982-09-01

    The analysis of fast breeder reactor components involves the knowledge of certain properties of the materials used. The latter consist of the following: - a body of data required for calculations, including allowable stresses and fatigue strength, as well as the rules applicable to these data, - a number of qualitative requirements serving to guarantee that the quality of the material fully justifies the use of the previously established elements. This duality of concerns is illustrated by some recent examples which occured during the construction of the Super Phenix reactor [fr

  10. An approach to development of structural design criteria for highly irradiated core components

    International Nuclear Information System (INIS)

    Nelson, D.V.

    1980-01-01

    The advent of the fast breeder reactor presents novel challenges in structural design and materials engineering. For instance, the core components of these reactors experience high energy neutron irradiation at elevated temperature, which causes significant time-dependent changes in material behaviour, such as a progressive loss of ductility. New structural design criteria are needed to extend elevated temperature design-by-analysis to account for these changes. Alloys best able to cope with the demands of the core operating environment are being explored and their structural behaviour characterized. The purpose of this paper is to illustrate an approach used in the development of core component structural design criteria. To do this, several design rules, plus brief rationale, from draft RDT Standards F9-7, -8 and -9 will be presented. These recently completed standards ('Structural Design Guidelines for Breeder Reactor Core Components') were prepared for the U.S. Department of Energy and represent a consensus among most organizations participating in the U.S. breeder program. (author)

  11. Structural integrity assessment of piping components

    International Nuclear Information System (INIS)

    Kushwaha, H.S.; Chattopadhyay, J.

    2008-01-01

    Integrity assessment of piping components is very essential for safe and reliable operation of power plants. Over the last several decades, considerable work has been done throughout the world to develop a methodology for integrity assessment of pipes and elbows, appropriate for the material involved. However, there is scope of further development/improvement of issues, particularly for pipe bends, that are important for accurate integrity assessment of piping. Considering this aspect, a comprehensive Component Integrity Test Program was initiated in 1998 at Bhabha Atomic Research Centre (BARC), India. In this program, both theoretical and experimental investigations were undertaken to address various issues related to the integrity assessment of pipes and elbows. Under the experimental investigations, fracture mechanics tests have been conducted on pipes and elbows of 200-400 mm nominal bore (NB) diameter with various crack configurations and sizes under different loading conditions. Tests on small tensile and three point bend specimens, machined from the tested pipes, have also been done to evaluate the actual stress-strain and fracture resistance properties of pipe/elbow material. The load-deflection curve and crack initiation loads predicted by non-linear finite element analysis matched well with the experimental results. The theoretical collapse moments of throughwall circumferentially cracked elbows, predicted by the recently developed equations, are found to be closer to the test data compared to the other existing equations. The role of stress triaxialities ahead of crack tip is also shown in the transferability of J-Resistance curve from specimen to component. The cyclic loading and system compliance effect on the load carrying capacity of piping components are investigated and new recommendations are made. (author)

  12. U.S. fast reactor materials and structures program

    International Nuclear Information System (INIS)

    Harms, W.O.; Purdy, C.M.

    1984-01-01

    The U.S. DOE has sponsored a vigorous breeder reactor materials and structures program for 15 years. Important contributions have resulted from this effort in the areas of design (inelastic rules, verified methods, seismic criteria, mechanical properties data); resolution of licensing issues (technical witnessing, confirmatory testing); construction (fabrication/welding procedures, nondestructive testing techniques); and operation (sodium purification, instrumentation and chemical analysis, radioactivity control, and in-service inspection. The national LMFBR program currently is being restructured. The Materials and Structures Program will focus its efforts in the following areas: (1) removal of anticipated licensing impediments through confirmation of the adequacy of structural design methods and criteria for components containing welds and geometric discontinuities, the generation of mechanical properties for stainless steel castings and weldments, and the evaluation of irradiation effects; (2) qualification of modified 9 Cr-1 Mo steel and tribological coatings for design flexibility; (3) development of improved inelastic design guidelines and procedures; (4) reform of design codes and standards and engineering practices, leading to simpler, less conservative rules and to simplified design analysis methods; and (5) incorporation of information from foreign program

  13. Support structure for reactor core constituent element

    International Nuclear Information System (INIS)

    Aida, Yasuhiko.

    1993-01-01

    A connection pipe having an entrance nozzle inserted therein as a reactor core constituent element is protruded above the upper surface of a reactor core support plate. A through hole is disposed to the protruding portion of the connection pipe. The through hole and a through hole disposed to the reactor core support plate are connected by a communication pipe. A shear rod is disposed in a horizontal portion at the inside of the communication pipe and is supported by a spring horizontally movably. Further, a groove is disposed at a position of the entrance nozzle opposing to the shear rod. The shear rod is urged out of the communication pipe by the pressure of the high pressure plenum and the top end portion of the shear rod is inserted to the groove of the entrance nozzle during operation. Accordingly, the shear rod is positioned in a state where it is extended from the through hole of the communication pipe to the groove of the entrance nozzle. This can mechanically constrain the rising of the reactor core constituent elements by the shear rod upon occurrence of earthquakes. (I.N.)

  14. Nuclear reactor structural material forming less radioactive corrosion product

    International Nuclear Information System (INIS)

    Nakazawa, Hiroshi.

    1988-01-01

    Purpose: To provide nuclear reactor structural materials forming less radioactive corrosion products. Constitution: Ni-based alloys such as inconel alloy 718, 600 or inconel alloy 750 and 690 having excellent corrosion resistance and mechanical property even in coolants at high temperature and high pressure have generally been used as nuclear reactor structural materials. However, even such materials yield corrosion products being attacked by coolants circulating in the nuclear reactor, which produce by neutron irradiation radioactive corrosion products, that are deposited in primary circuit pipeways to constitute exposure sources. The present invention dissolves dissolves this problems by providing less activating nuclear reactor structural materials. That is, taking notice on the fact that Ni-58 contained generally by 68 % in Ni changes into Co-58 under irradiation of neutron thereby causing activation, the surface of nuclear reactor structural materials is applied with Ni plating by using Ni with a reduced content of Ni-58 isotopes. Accordingly, increase in the radiation level of the nuclear reactor structural materials can be inhibited. (K.M.)

  15. Factor structure underlying components of allostatic load.

    Directory of Open Access Journals (Sweden)

    Jeanne M McCaffery

    Full Text Available Allostatic load is a commonly used metric of health risk based on the hypothesis that recurrent exposure to environmental demands (e.g., stress engenders a progressive dysregulation of multiple physiological systems. Prominent indicators of response to environmental challenges, such as stress-related hormones, sympatho-vagal balance, or inflammatory cytokines, comprise primary allostatic mediators. Secondary mediators reflect ensuing biological alterations that accumulate over time and confer risk for clinical disease but overlap substantially with a second metric of health risk, the metabolic syndrome. Whether allostatic load mediators covary and thus warrant treatment as a unitary construct remains to be established and, in particular, the relation of allostatic load parameters to the metabolic syndrome requires elucidation. Here, we employ confirmatory factor analysis to test: 1 whether a single common factor underlies variation in physiological systems associated with allostatic load; and 2 whether allostatic load parameters continue to load on a single common factor if a second factor representing the metabolic syndrome is also modeled. Participants were 645 adults from Allegheny County, PA (30-54 years old, 82% non-Hispanic white, 52% female who were free of confounding medications. Model fitting supported a single, second-order factor underlying variance in the allostatic load components available in this study (metabolic, inflammatory and vagal measures. Further, this common factor reflecting covariation among allostatic load components persisted when a latent factor representing metabolic syndrome facets was conjointly modeled. Overall, this study provides novel evidence that the modeled allostatic load components do share common variance as hypothesized. Moreover, the common variance suggests the existence of statistical coherence above and beyond that attributable to the metabolic syndrome.

  16. Irradiation of electronic components and circuits at the Portuguese Research Reactor: Lessons learned

    Energy Technology Data Exchange (ETDEWEB)

    Marques, J.G.; Ramos, A.R.; Fernandes, A.C.; Santos, J.P. [Centro de Ciencias e Tecnologias Nucleares, Instituto Superior Tecnico, Universidade de Lisboa, Estrada Nacional 10, 2695-066 Bobadela LRS (Portugal)

    2015-07-01

    The behavior of electronic components and circuits under radiation is a concern shared by the nuclear industry, the space community and the high-energy physics community. Standard commercial components are used as much as possible instead of radiation hard components, since they are easier to obtain and allow a significant reduction of costs. However, these standard components need to be tested in order to determine their radiation tolerance. The Portuguese Research Reactor (RPI) is a 1 MW pool-type reactor, operating since 1961. The irradiation of electronic components and circuits is one area where a 1 MW reactor can be competitive, since the fast neutron fluences required for testing are in most cases well below 10{sup 16} n/cm{sup 2}. A program was started in 1999 to test electronics components and circuits for the LHC facility at CERN, initially using a dedicated in-pool irradiation device and later a beam line with tailored neutron and gamma filters. Neutron filters are essential to reduce the intensity of the thermal neutron flux, which does not produce significant defects in electronic components but produces unwanted radiation from activation of contacts and packages of integrated circuits and also of the printed circuit boards. In irradiations performed within the line-of-sight of the core of a fission reactor there is simultaneous gamma radiation which complicates testing in some cases. Filters can be used to reduce its importance and separate testing with a pure gamma radiation source can contribute to clarify some irradiation results. Practice has shown the need to introduce several improvements to the procedures and facilities over the years. We will review improvements done in the following areas: - Optimization of neutron and gamma filters; - Dosimetry procedures in mixed neutron / gamma fields; - Determination of hardness parameter and 1 MeV-equivalent neutron fluence; - Temperature measurement and control during irradiation; - Follow-up of reactor

  17. Simulating the behaviour of zirconium-alloy components in nuclear reactors

    International Nuclear Information System (INIS)

    Coleman, C.E.

    2001-12-01

    To prevent failure in nuclear components one needs to understand the interactions between adjacent materials and the changes in their physical properties during all phases of reactor operation. Three examples from CANDU reactors are described to illustrate the use of simulations that imitate complicated reactor situations. These are: swelling tests that led to a method for increasing the tolerance or Zircaloy fuel cladding to power ramps; observations of the behaviour of leaking cracks in Zr-2.5Nb pressure tubes that provide confidence in the use of leak-before-break as part of the defence against flaw development; and contact boiling tests on modifications to the surfaces of Zircaloy calandria tubes that enhance the ability of the heavy water moderator to act as a heat sink after a postulated loss-of-coolant accident. (author)

  18. Reactor Safety Gap Evaluation of Accident Tolerant Components and Severe Accident Analysis

    International Nuclear Information System (INIS)

    Farmer, Mitchell T.; Bunt, R.; Corradini, M.; Ellison, Paul B.; Francis, M.; Gabor, John D.; Gauntt, R.; Henry, C.; Linthicum, R.; Luangdilok, W.; Lutz, R.; Paik, C.; Plys, M.; Rabiti, Cristian; Rempe, J.; Robb, K.; Wachowiak, R.

    2015-01-01

    The overall objective of this study was to conduct a technology gap evaluation on accident tolerant components and severe accident analysis methodologies with the goal of identifying any data and/or knowledge gaps that may exist, given the current state of light water reactor (LWR) severe accident research, and additionally augmented by insights obtained from the Fukushima accident. The ultimate benefit of this activity is that the results can be used to refine the Department of Energy's (DOE) Reactor Safety Technology (RST) research and development (R&D) program plan to address key knowledge gaps in severe accident phenomena and analyses that affect reactor safety and that are not currently being addressed by the industry or the Nuclear Regulatory Commission (NRC).

  19. ARCHER Project: Progress on Material and component activities for the Advanced High Temperature Reactor

    International Nuclear Information System (INIS)

    Buckthorpe, D.E.

    2014-01-01

    The ARCHER (Advanced High-Temperature Reactors for Cogeneration of Heat and Electricity R&D) integrated project is a four year project which was started in 2011 as part of the European Commission 7th Framework Programme (FP7) to perform High Temperature Reactor technology R&D in support of reactor demonstration. The project consortium encompasses conventional and Nuclear Industry, Utilities, Technical Support Organizations, Research & Development Organizations and Academia. The activities involved contribute to the Generation IV (GIF) International Forum and collaborate with related projects in the US, China, Japan, and the Republic of Korea in cooperation with IAEA and ISTC. This paper addresses the progress of the work on ARCHER materials and component activities since the start of the project and underlines some of the main conclusions reached. (author)

  20. Material and component progress within ARCHER for advanced high temperature reactor

    International Nuclear Information System (INIS)

    Buckthorpe, D.E.; Davies, M.; Pra, F.; Bonnamy, P.; Fokkens, J.; Heijna, M.; Bout, N. de; Vreeling, A.; Bourlier, F.; Lhachemi, D.; Woayehune, A.; Dubiez-le-Goff, S.; Hahner, P.; Futterer, M.; Berka, J.; Kalivodora, J.; Pouchon, M.A.; Schmitt, R.; Homerin, P.; Marsden, B.; Mummery, P.; Mutch, G.; Ponca, D.; Buhl, P.; Hoffmann, M.; Rondet, F.; Pecherty, A.; Baurand, F.; Alenda, F.; Esch, M.; Kohlz, N.; Reed, J.; Fachinger, J.; Klower, Dr.

    2014-01-01

    The ARCHER (Advanced High-Temperature Reactors for Cogeneration of Heat and Electricity R and D) integrated project started in 2011 as part of the European Commission 7. Framework Programme (FP7) for a period of four years to perform High Temperature Reactor technology R and D in support of reactor demonstration. The project consortium encompasses conventional and Nuclear Industry, Utilities, Technical Support Organizations, Research and Development Organizations and Academia. The activities involved contribute to the Generation IV (GIF) International Forum and collaborate with related projects in the US, China, Japan, and the Republic of Korea in cooperation with IAEA and ISTC. This paper addresses the progress of the work on materials and component technologies within ARCHER over the first two years of the project. (authors)

  1. Reactor Safety Gap Evaluation of Accident Tolerant Components and Severe Accident Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Farmer, Mitchell T. [Argonne National Lab. (ANL), Argonne, IL (United States); Bunt, R. [Southern Nuclear, Atlanta, GA (United States); Corradini, M. [Univ. of Wisconsin, Madison, WI (United States); Ellison, Paul B. [GE Power and Water, Duluth, GA (United States); Francis, M. [Argonne National Lab. (ANL), Argonne, IL (United States); Gabor, John D. [Erin Engineering, Walnut Creek, CA (United States); Gauntt, R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Henry, C. [Fauske and Associates, Burr Ridge, IL (United States); Linthicum, R. [Exelon Corp., Chicago, IL (United States); Luangdilok, W. [Fauske and Associates, Burr Ridge, IL (United States); Lutz, R. [PWR Owners Group (PWROG); Paik, C. [Fauske and Associates, Burr Ridge, IL (United States); Plys, M. [Fauske and Associates, Burr Ridge, IL (United States); Rabiti, Cristian [Idaho National Lab. (INL), Idaho Falls, ID (United States); Rempe, J. [Rempe and Associates LLC, Idaho Falls, ID (United States); Robb, K. [Argonne National Lab. (ANL), Argonne, IL (United States); Wachowiak, R. [Electric Power Research Inst. (EPRI), Knovville, TN (United States)

    2015-01-31

    The overall objective of this study was to conduct a technology gap evaluation on accident tolerant components and severe accident analysis methodologies with the goal of identifying any data and/or knowledge gaps that may exist, given the current state of light water reactor (LWR) severe accident research, and additionally augmented by insights obtained from the Fukushima accident. The ultimate benefit of this activity is that the results can be used to refine the Department of Energy’s (DOE) Reactor Safety Technology (RST) research and development (R&D) program plan to address key knowledge gaps in severe accident phenomena and analyses that affect reactor safety and that are not currently being addressed by the industry or the Nuclear Regulatory Commission (NRC).

  2. Polymer matrix nanocomposites for automotive structural components

    Science.gov (United States)

    Naskar, Amit K.; Keum, Jong K.; Boeman, Raymond G.

    2016-12-01

    Over the past several decades, the automotive industry has expended significant effort to develop lightweight parts from new easy-to-process polymeric nanocomposites. These materials have been particularly attractive because they can increase fuel efficiency and reduce greenhouse gas emissions. However, attempts to reinforce soft matrices by nanoscale reinforcing agents at commercially deployable scales have been only sporadically successful to date. This situation is due primarily to the lack of fundamental understanding of how multiscale interfacial interactions and the resultant structures affect the properties of polymer nanocomposites. In this Perspective, we critically evaluate the state of the art in the field and propose a possible path that may help to overcome these barriers. Only once we achieve a deeper understanding of the structure-properties relationship of polymer matrix nanocomposites will we be able to develop novel structural nanocomposites with enhanced mechanical properties for automotive applications.

  3. Investigation on structural integrity of graphite component during high temperature 950degC continuous operation of HTTR

    International Nuclear Information System (INIS)

    Sumita, Junya; Shimazaki, Yosuke; Shibata, Taiju

    2014-01-01

    Graphite material is used for internal structures in high temperature gas-cooled reactor. The core components and graphite core support structures are so designed as to maintain the structural integrity to keep core cooling capability. To confirm that the core components and graphite core support structures satisfy the design requirements, the temperatures of the reactor internals are measured during the reactor operation. Surveillance test of graphite specimens and in-service inspection using TV camera are planned in conjunction with the refueling. This paper describes the evaluation results of the integrity of the core components and graphite core support structures during the high temperature 950degC continuous operation, a high temperature continuous operation with reactor outlet temperature of 950degC for 50 days, in high temperature engineering test reactor. The design requirements of the core components and graphite core support structures were satisfied during the high temperature 950degC continuous operation. The dimensional change of graphite which directly influences the temperature of coolant was estimated considering the temperature profiles of fuel block. The magnitude of irradiation-induced dimensional change considering temperature profiles was about 1.2 times larger than that under constant irradiation temperature of 1000degC. In addition, the programs of surveillance test and ISI using TV camera were introduced. (author)

  4. Selection of hardfacing material for components of the Indian Prototype Fast Breeder Reactor

    International Nuclear Information System (INIS)

    Bhaduri, A.K.; Indira, R.; Albert, S.K.; Rao, B.P.S.; Jain, S.C.; Asokkumar, S.

    2004-01-01

    Nickel-base hardfacing alloys have been chosen to replace cobalt-base alloys as hardfacing material for components of the Indian Prototype Fast Breeder Reactor, for minimising the dose rate to personnel during maintenance and decommissioning, and to reduce the shielding thickness required for component handling. Induced activity, dose rate and shielding computations showed that replacing cobalt-base alloys with nickel-base alloys for hardfacing of components would result in a marked reduction in both the dose rate from the components and the thickness of lead handling flasks. Long-term ageing studies on the nickel-base hardface deposits on austenitic stainless steel showed that the hardface deposit would retain adequate hardness at the end of the components' design service-life of 40 years of exposure at 823 K

  5. Structural analysis of reactor fuel elements

    International Nuclear Information System (INIS)

    Weeks, R.W.

    1977-01-01

    An overview of fuel-element modeling is presented that traces the development of codes for the prediction of light-water-reactor and fast-breeder-reactor fuel-element performance. It is concluded that although the mathematical analysis is now far advanced, the development and incorporation of mechanistic constitutive equations has not kept pace. The resultant reliance on empirical correlations severely limits the physical insight that can be gained from code extrapolations. Current efforts include modeling of alternate fuel systems, analysis of local fuel-cladding interactions, and development of a predictive capability for off-normal behavior. Future work should help remedy the current constitutive deficiencies and should include the development of deterministic failure criteria for use in design

  6. SMART - Structure mechanical analysis in reactor technology

    International Nuclear Information System (INIS)

    Argyris, J.H.; Faust, G.; Szimmat, J.; Warnke, E.P.; Willam, K.J.

    1975-01-01

    The programme system SMART was developed in the years 1970-75 to calculate prestressed-concrete reactor pressure vessels with finite elements. The present report outlines the course and present state of research and development work. Following the specification of SMART, a brief presentation of the analytical possibilities and of the expansions for investigating creep, ultimate load behaviour and thermodiffusion is given. In conclusion, the fields of application of SMART are illustrated by means of examples. (orig./LH) [de

  7. Evolution of microstructure in zirconium alloy core components of nuclear reactors during service

    International Nuclear Information System (INIS)

    Griffiths, M.; Coleman, C.E.; Holt, R.A.; Sagat, S.; Urbanic, V.F.; Chow, C.K.

    1993-03-01

    X-ray diffraction and analytical electron microscopy have been used to characterise microstructural and microchemical changes produced by neutron irradiation of Zr-2.5Nb, Zircaloy-2 and Zircaloy-4 nuclear reactor core components. In many cases there is a clear relationship between the radiation damage microstructure and the physical properties of in-service core components. For example, the difference in delayed hydride cracking velocity between the inlet and outlet ends of Zr-2.5Nb pressure tubes in pressurised heavy water reactors can be directly correlated with variations in a-dislocation density and β-Zr phase decomposition. For the same tubes, the variation of fracture toughness has the same fluence dependence as dislocation loop density and improvements in corrosion behaviour can be linked with decreases in the Nb concentration in the α-Zr matrix due to Nb precipitation during irradiation. For pressurised water reactors and boiling water reactors the onset of 'breakaway' growth in Zircaloy-4 guide tubes can be directly correlated with the appearance of basal plane dislocation loops in the microstructure. (author). 37 refs., 28 figs., 4 tabs

  8. Evolution of microstructure in zirconium alloy core components of nuclear reactors during service

    Energy Technology Data Exchange (ETDEWEB)

    Griffiths, M; Coleman, C E; Holt, R A; Sagat, S; Urbanic, V F [Atomic Energy of Canada Ltd., Chalk River, ON (Canada). Chalk River Nuclear Labs.; Chow, C K [Atomic Energy of Canada Ltd., Pinawa, MB (Canada). Whiteshell Nuclear Research Establishment

    1993-03-01

    X-ray diffraction and analytical electron microscopy have been used to characterise microstructural and microchemical changes produced by neutron irradiation of Zr-2.5Nb, Zircaloy-2 and Zircaloy-4 nuclear reactor core components. In many cases there is a clear relationship between the radiation damage microstructure and the physical properties of in-service core components. For example, the difference in delayed hydride cracking velocity between the inlet and outlet ends of Zr-2.5Nb pressure tubes in pressurised heavy water reactors can be directly correlated with variations in a-dislocation density and {beta}-Zr phase decomposition. For the same tubes, the variation of fracture toughness has the same fluence dependence as dislocation loop density and improvements in corrosion behaviour can be linked with decreases in the Nb concentration in the {alpha}-Zr matrix due to Nb precipitation during irradiation. For pressurised water reactors and boiling water reactors the onset of `breakaway` growth in Zircaloy-4 guide tubes can be directly correlated with the appearance of basal plane dislocation loops in the microstructure. (author). 37 refs., 28 figs., 4 tabs.

  9. Base isolation strategies for structures and components

    International Nuclear Information System (INIS)

    Varma, Veto; Reddy, G.R.; Vaze, K.K.; Kushwaha, H.S.

    2003-08-01

    In the present report the effect of laminated rubber bearing (LRB) system on the dynamic response of the structure was studied. A LRB system was designed and tested in the laboratory for its dynamic characteristics. Finite element analysis was also performed and based on this analysis, isolator for PHWR nuclear power plant was designed. Analysis of the building was performed with and without isolator. Comparison of responses was made in terms of frequencies, accelerations and displacements and floor response spectra. (author)

  10. Generic aging management programs for license renewal of BWR reactor coolant systems components

    International Nuclear Information System (INIS)

    Shah, V.N.; Liu, Y.Y.

    2002-01-01

    The paper reviews the existing generic aging management programs (AMPs) for the reactor coolant system (RCS) components in boiling water reactors (BWRs), including the reactor pressure vessel and internals, the reactor recirculation system, and the connected piping. These programs have been evaluated in the U.S. Nuclear Regulatory Commission (NRC) report, Generic Aging Lessons Learned (GALL), NUREG-1801, for their use in the license renewal process to manage several aging effects, including loss of material, crack initiation and growth, loss of fracture toughness, loss of preload, wall thinning, and cumulative fatigue damage. The program evaluation includes a review of ten attributes (scope of program, preventive actions, parameters monitored/inspected, detection of aging effects, monitoring and trending, acceptance criteria, corrective actions, confirmative process, administrative control, and operating experience) for their effectiveness in managing a specific aging effect in a given component(s). The generic programs are based on the ASME Section XI inservice inspection requirements; industry guidelines for inspection and evaluation of aging effects in BWR reactor vessel, internals, and recirculation piping; monitoring and control of BWR water chemistry; and operating experience as reported in the USNRC generic communications and industry reports. The review concludes that all generic AMPs are acceptable for managing aging effects in BWR RCS components during an extended period of operation and do not need further evaluation. However, the plant-specific programs for managing aging in certain RCS components during an extended period of operation do require further evaluation. For some plant-specific AMPs, the GALL report recommends an aging management activity to verify their effectiveness. An example of such an activity is a one-time inspection of Class 1 small-bore piping to ensure that service-induced weld cracking is not occurring in the piping. Several of

  11. Generic Aging Management Programs for License Renewal of BWR Reactor Coolant System Components

    International Nuclear Information System (INIS)

    Shah, V.N.; Liu, Y.Y.

    2002-01-01

    The paper reviews the existing generic aging management programs (AMPs) for the reactor coolant system (RCS) components in boiling water reactors (BWRs), including the reactor pressure vessel and internals, the reactor recirculation system, and the connected piping. These programs have been evaluated in the U.S. Nuclear Regulatory Commission (NRC) report, Generic Aging Lessons Learned (GALL), NUREG-1801, for their use in the license renewal process to manage several aging effects, including loss of material, crack initiation and growth, loss of fracture toughness, loss of preload, wall thinning, and cumulative fatigue damage. The program evaluation includes a review of ten attributes (scope of program, preventive actions, parameters monitored/inspected, detection of aging effects, monitoring and trending, acceptance criteria, corrective actions, confirmative process, administrative control, and operating experience) for their effectiveness in managing a specific aging effect in a given component(s). The generic programs are based on the ASME Section XI inservice inspection requirements; industry guidelines for inspection and evaluation of aging effects in BWR reactor vessel, internals, and recirculation piping; monitoring and control of BWR water chemistry; and operating experience as reported in the USNRC generic communications and industry reports. The review concludes that all generic AMPs are acceptable for managing aging effects in BWR RCS components during an extended period of operation and do not need further evaluation. However, the plant-specific programs for managing aging in certain RCS components during an extended period of operation do require further evaluation. For some plant-specific AMPs, the GALL report recommends an aging management activity to verify their effectiveness. An example of such an activity is a one-time inspection of Class 1 small-bore piping to ensure that service-induced weld cracking is not occurring in the piping. Several of

  12. Study of the WWR-S IFIN-HH reactor main components stare, after 40 years working, using nondestructive methods

    International Nuclear Information System (INIS)

    Dragolici, A. C.; Zorliu, A.; Ripeanu, R.; Petran, C.; Mincu, I.

    2000-01-01

    The main goal of these investigations was to establish the security level after 40 years of working of the WWR-S research reactor of Horia Hulubei National Institute of Research and Development for Physics and Nuclear Engineering, Bucharest-Magurele. The purpose of these investigations was: checking the functionality and the physical integrity of the main components of the reactor. The physical integrity of the components is usually affected by slow processes, such as: corrosion, erosion, aging, deformations and initially hidden flaws with very slow evolutions. The methods used to determine the effects of these processes and to infer conclusions about the physical integrity of the facility are: visualizations by optical means (endoscopy and video camera), examination using ultrasounds and gammagraphy. The objective of the endoscopic checking was the view of the state of interior surfaces of the tubes and pipes, specially the inaccessible areas of the non-dismantling parts of the reactor. Big size components, such as reactor vessel, the biologic protection vessel and the main large diameter pipes of the primary cooling system, were investigated using a special device that contains a video camera connected to a PC. To obtain more information regarding the evolution of the corrosion spots, scratches and harmed areas on the investigated surfaces, their depth was checked by ultrasounds, and the welding seams structure was determined by gammagraphy. A table is given with some significant results obtained from ultrasound measurements in different points of reactor vessel, thermal column, horizontal tubes, etc. After these tests, the conclusions are: the maximum corrosion depth is 0.2 mm; - scratches are superficially, not exceeding 0.2-0.5 mm; - the traces of harmed areas are produced by the electromagnetic device utilization used for manipulation of aluminium capsules which contain irradiated substances. They are superficial, with maximum area of about 1 cm 2 ; the

  13. Study on dynamic characteristics of reduced analytical model for PWR reactor internal structures

    International Nuclear Information System (INIS)

    Yoo, Bong; Lee, Jae Han; Kim, Jong Bum; Koo, Kyeong Hoe

    1993-01-01

    The objective of this study is to establish the procedure of the reduced analytical modeling technique for the PWR reactor internal(RI) structures and to carry out the sensitivity study of the dynamic characteristics of the structures by varying the structural parameters such as the stiffness, the mass and the damping. Modeling techniques for the PWR reactor internal structures and computer programs used for the dynamic analysis of the reactor internal structures are briefly investigated. Among the many components of RI structures, the dynamic characteristics for CSB was performed. The sensitivity analysis of the dynamic characteristics for the reduced analytical model considering the variations of the stiffnesses for the lower and upper flanges of the CSB and for the RV Snubber were performed to improve the dynamic characteristics of the RI structures against the external loadings given. In order to enhance the structural design margin of the RI components, the nonlinear time history analyses were attempted for the RI reduced models to compare the structural responses between the reference model and the modified one. (Author)

  14. Conceptual design of a fission-based integrated test facility for fusion reactor components

    International Nuclear Information System (INIS)

    Watts, K.D.; Deis, G.A.; Hsu, P.Y.S.; Longhurst, G.R.; Masson, L.S.; Miller, L.G.

    1982-01-01

    The testing of fusion materials and components in fission reactors will become increasingly important because of lack of fusion engineering test devices in the immediate future and the increasing long-term demand for fusion testing when a fusion reactor test station becomes available. This paper presents the conceptual design of a fission-based Integrated Test Facility (ITF) developed by EG and G Idaho. This facility can accommodate entire first wall/blanket (FW/B) test modules such as those proposed for INTOR and can also accommodate smaller cylindrical modules similar to those designed by Oak Ridge National laboratory (ORNL) and Westinghouse. In addition, the facility can be used to test bulk breeder blanket materials, materials for tritium permeation, and components for performance in a nuclear environment. The ITF provides a cyclic neutron/gamma flux as well as the numerous module and experiment support functions required for truly integrated tests

  15. Preliminary Development of Online Monitoring Acoustic Emission System for the Integrity of Research Reactor Components

    Science.gov (United States)

    Bakhri, S.; Sumarno, E.; Himawan, R.; Akbar, T. Y.; Subekti, M.; Sunaryo, G. R.

    2018-02-01

    Three research reactors owned by BATAN have been more than 25 years. Aging of (Structure, System and Component) SSC which is mainly related to mechanical causes become the most important issue for the sustainability and safety operation. Acoustic Emission (AE) is one of the appropriate and recommended methods by the IAEA for inspection as well as at the same time for the monitoring of mechanical SSC related. However, the advantages of AE method in detecting the acoustic emission both for the inspection and the online monitoring require a relatively complex measurement system including hardware software system for the signal detection and analysis purposes. Therefore, aim of this work was to develop an AE system based on an embedded system which capable for doing both the online monitoring and inspection of the research reactor’s integrity structure. An embedded system was selected due to the possibility to install the equipment on the field in extreme environmental condition with capability to store, analyses, and send the required information for further maintenance and operation. The research was done by designing the embedded system based on the Field Programmable Gate Array (FPGA) platform, because of their execution speed and system reconfigurable opportunities. The AE embedded system is then tested to identify the AE source location and AE characteristic under tensile material testing. The developed system successfully acquire the AE elastic waveform and determine the parameter-based analysis such as the amplitude, peak, duration, rise time, counts and the average frequency both for the source location test and the tensile test.

  16. Development of underwater laser cladding and underwater laser seal welding techniques for reactor components

    International Nuclear Information System (INIS)

    Hino, Takehisa; Tamura, Masataka; Tanaka, Yoshimi; Kouno, Wataru; Makino, Yoshinobu; Kawano, Shohei; Matsunaga, Keiji

    2009-01-01

    Stress corrosion cracking (SCC) has been reported at the aged components in many nuclear power plants. Toshiba has been developing the underwater laser welding. This welding technique can be conducted without draining the water in the reactor vessel. It is beneficial for workers not to exposure the radiation. The welding speed can be attaining twice as fast as that of Gas Tungsten Arc Welding (GTAW). The susceptibility of SCC can also be lower than the Alloy 600 base metal. (author)

  17. Thermal-hydraulic limitations on water-cooled fusion reactor components

    International Nuclear Information System (INIS)

    Cha, Y.S.; Misra, B.

    1986-01-01

    An assessment of the cooling requirements for fusion reactor components, such as the first wall and limiter/divertor, was carried out using pressurized water as the coolant. In order to establish the coolant operating conditions, a survey of the literature on departure from nucleate boiling, critical heat flux, asymmetrical heating and heat transfer augmentation techniques was carried out. The experimental data and the empirical correlations indicate that thermal protection for the fusion reactor components based on conventional design concepts can be provided with an adequate margin of safety without resorting to either high coolant velocities, excessive coolant pressures, or heat transfer augmentation techniques. If, however, the future designs require unconventional shapes or heat transfer enhancement techniques, experimental verification would be necessary since no data on heat transfer augmentation techniques exist for complex geometries, especially under asymmetrically heated conditions. Since the data presented herein are concerned primarily with thermal protection of the reactor components, the final design should consider other factors such as thermal stresses, temperature limits, and fatigue

  18. Research of application of new material to light water reactor components

    International Nuclear Information System (INIS)

    Mihara, Tanetoyo

    1992-01-01

    Advanced Nuclear Equipment Research Institute (ANERI) has been doing the research to apply the new material including metal, fine ceramics and high polymer which were developed and applied in other industries to components and parts of light water reactor for the purpose of Improvement of reliability of components, improvement of efficiency of periodic inspection, improvement of repair and reduction of radiation exposure of worker. This project started upon the sponsorship of Ministry of International Trade and Industry (MITI) by the schedule of FY1985-FY1993 (9 years) and effective results has been obtained. (author)

  19. Computer-controlled ultrasonic equipment for automatic inspection of nuclear reactor components after manufacturing

    International Nuclear Information System (INIS)

    Moeller, P.; Roehrich, H.

    1983-01-01

    After foundation of the working team ''Automated US-Manufacture Testing'' in 1976 the realization of an ultrasonic test facility for nuclear reactor components after manufacturing has been started. During a period of about 5 years, an automated prototype facility has been developed, fabricated and successfully tested. The function of this facility is to replace the manual ultrasonic tests, which are carried out autonomically at different stages of the manufacturing process and to fulfil the test specification under improved economic conditions. This prototype facility has been designed as to be transported to the components to be tested at low expenditure. Hereby the reproduceability of a test is entirely guaranteed. (orig.) [de

  20. Materials for heat flux components of the first wall in fusion reactors

    International Nuclear Information System (INIS)

    Hoven, H.; Koizlik, K.; Linke, J.; Nickel, H.; Wallura, E.

    1985-08-01

    Materials of the First Wall in near-fusion plasma machines are subjected to a complex load system resulting from the plasma-wall interaction. The materials for their part also influence the plasma. Suitable materials must be available in order to ensure that the wall components achieve a sufficiently long dwell time and that their effects on the plasma remain small and controllable. The present report discusses relations between the plasma-wall interaction, the reactions of the materials and testing and examination methods for specific problems in developing and selecting suitable materials for highly stressed components on the First Wall of fusion reactors. (orig.)

  1. Structure optimization of CFB reactor for moderate temperature FGD

    Energy Technology Data Exchange (ETDEWEB)

    Li, Yuan; Zhang, Jie; Zheng, Kai; You, Changfu [Tsinghua Univ., Beijing (China). Dept. of Thermal Engineering; Ministry of Education, Beijing (China). Key Lab. for Thermal Science and Power Engineering

    2013-07-01

    The gas velocity distribution, sorbent particle concentration distribution and particle residence time in circulating fluidized bed (CFB) reactors for moderate temperature flue gas desulfurization (FGD) have significant influence on the desulfurization efficiency and the sorbent calcium conversion ratio for sulfur reaction. Experimental and numerical methods were used to investigate the influence of the key reactor structures, including the reactor outlet structure, internal structure, feed port and circulating port, on the gas velocity distribution, sorbent particle concentration distribution and particle residence time. Experimental results showed that the desulfurization efficiency increased 5-10% when the internal structure was added in the CFB reactor. Numerical analysis results showed that the particle residence time of the feed particles with the average diameter of 89 and 9 {mu}m increased 40% and 17% respectively, and the particle residence time of the circulating particles with the average diameter of 116 {mu}m increased 28% after reactor structure optimization. The particle concentration distribution also improved significantly, which was good for improving the contact efficiency between the sorbent particles and SO{sub 2}. In addition, the optimization guidelines were proposed to further increase the desulfurization efficiency and the sorbent calcium conversion ratio.

  2. The Performance of Structured Packings in Trickle-Bed Reactors

    NARCIS (Netherlands)

    Frank, M.J.W.; Kuipers, J.A.M.; Versteeg, G.F.; Swaaij, W.P.M. van

    1999-01-01

    An experimental study was carried out to investigate whether the use of structured packings might improve the mass transfer characteristics and the catalyst effectiveness of a trickle-bed reactor. Therefore, the performances of a structured packing, consisting of KATAPAK elements, and a dumped

  3. Thermal performance of an insulating structure for a reactor vessel

    International Nuclear Information System (INIS)

    Aranovitch, E.; Crutzen, S.; LeDet, M.; Denis, R.

    This report describes the installations used to test the HTGR reactor vessel insulating structure called ''Casali'' and details the experimental results in 3 groups: general experiments, systematic study, and technological experiments. The results obtained make it possible to satisfactorily predict the behavior of the structure in a practical application

  4. Analysis of removal alternatives for the Heavy Water Components Test Reactor at the Savannah River Site. Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    Owen, M.B.

    1997-04-01

    This engineering study evaluates different alternatives for decontamination and decommissioning of the Heavy Water Components Test Reactor (HWCTR). Cooled and moderated with pressurized heavy water, this uranium-fueled nuclear reactor was designed to test fuel assemblies for heavy water power reactors. It was operated for this purpose from march of 1962 until December of 1964. Four alternatives studied in detail include: (1) dismantlement, in which all radioactive and hazardous contaminants would be removed, the containment dome dismantled and the property restored to a condition similar to its original preconstruction state; (2) partial dismantlement and interim safe storage, where radioactive equipment except for the reactor vessel and steam generators would be removed, along with hazardous materials, and the building sealed with remote monitoring equipment in place to permit limited inspections at five-year intervals; (3) conversion for beneficial reuse, in which most radioactive equipment and hazardous materials would be removed and the containment building converted to another use such as a storage facility for radioactive materials, and (4) entombment, which involves removing hazardous materials, filling the below-ground structure with concrete, removing the containment dome and pouring a concrete cap on the tomb. Also considered was safe storage, but this approach, which has, in effect, been followed for the past 30 years, did not warrant detailed evaluation. The four other alternatives were evaluate, taking into account factors such as potential effects on the environment, risks, effectiveness, ease of implementation and cost. The preferred alternative was determined to be dismantlement. This approach is recommended because it ranks highest in the comparative analysis, would serve as the best prototype for the site reactor decommissioning program and would be most compatible with site property reuse plans for the future.

  5. Analysis of removal alternatives for the Heavy Water Components Test Reactor at the Savannah River Site. Revision 1

    International Nuclear Information System (INIS)

    Owen, M.B.

    1997-04-01

    This engineering study evaluates different alternatives for decontamination and decommissioning of the Heavy Water Components Test Reactor (HWCTR). Cooled and moderated with pressurized heavy water, this uranium-fueled nuclear reactor was designed to test fuel assemblies for heavy water power reactors. It was operated for this purpose from march of 1962 until December of 1964. Four alternatives studied in detail include: (1) dismantlement, in which all radioactive and hazardous contaminants would be removed, the containment dome dismantled and the property restored to a condition similar to its original preconstruction state; (2) partial dismantlement and interim safe storage, where radioactive equipment except for the reactor vessel and steam generators would be removed, along with hazardous materials, and the building sealed with remote monitoring equipment in place to permit limited inspections at five-year intervals; (3) conversion for beneficial reuse, in which most radioactive equipment and hazardous materials would be removed and the containment building converted to another use such as a storage facility for radioactive materials, and (4) entombment, which involves removing hazardous materials, filling the below-ground structure with concrete, removing the containment dome and pouring a concrete cap on the tomb. Also considered was safe storage, but this approach, which has, in effect, been followed for the past 30 years, did not warrant detailed evaluation. The four other alternatives were evaluate, taking into account factors such as potential effects on the environment, risks, effectiveness, ease of implementation and cost. The preferred alternative was determined to be dismantlement. This approach is recommended because it ranks highest in the comparative analysis, would serve as the best prototype for the site reactor decommissioning program and would be most compatible with site property reuse plans for the future

  6. A kinetic model for impact/sliding wear of pressurized water reactor internal components: Application to rod cluster control assemblies

    International Nuclear Information System (INIS)

    Zbinden, M.

    1996-01-01

    Certain internal components of Pressurized Water Reactors are damaged by wear when subjected to vibration induced by flow. In order to enable predictive calculation of such wear, one must have a model which takes account reliably of real damages. The modelling of wear represents a final link in a succession of numerical calculations which begins by the determination of hydraulic excitations induced by the flow. One proceeds, then, in the dynamic response calculation of the structure to finish up with an estimation of volumetric wear and of the depth of wear scars. A new concept of industrial wear model adapted to components of nuclear plants is proposed. Its originality is to be supported, on one hand, by experimental results obtained via wear machines of relatively short operational times, and, on the other hand, by the information obtained from the operating feedback over real wear kinetics of the reactors components. The proposed model is illustrated by an example which correspond to a specific real situation. The determination of the coefficients permitting to cover all assembly of configurations and the validation of the model in these configurations have been the object of the most recent work

  7. Structure of pool in reactor building

    International Nuclear Information System (INIS)

    Yokoyama, Shigeki.

    1997-01-01

    Shielding walls made of iron-reinforced concrete having a metal liner including two body walls rigidly combined to the upper surface of a reactor container are disposed at least to one of an equipment pool or spent fuel storage pool in a reactor building. A rack for temporarily placing an upper lattice plate is detachably attached at least above one of a steam dryer or a gas/liquid separator temporarily placed in the temporary pool, and the height from the bottom portion to the upper end of the shielding wall is determined based on the height of an upper lattice plate temporary placed on the rack and the water depth required for shielding radiation from the upper lattice plate. An operator's exposure on the operation floor can be reduced by the shielding wall, and radiation dose from the spent fuels is reduced. The increase of the height of a pool guarder enhances bending resistance as a ceiling. In addition, the total height of them is made identical with the depth of the spent fuel storage pool thereby enabling to increase storage area for spent fuels. (N.H.)

  8. Performance tests of the reactor containment structures of HTTR

    International Nuclear Information System (INIS)

    Sakaba, Nariaki; Iigaki, Kazuhiko; Kawaji, Satoshi; Iyoku, Tatsuo

    1998-03-01

    The containment structures of the HTTR consist of the reactor containment vessel (CV), service area (SA) and emergency air purification system, which minimize the release of FPs in the postulated accidents with FP release from the reactor facilities. The CV is designed to withstand the temperature and pressure transients and to be leak-tight within the specified leakage limit even in the case of a rupture of the primary concentric hot gas duct. The pressure of inside of the SA should be maintained slightly lower than that of atmosphere by the emergency air purification system. The radioactive materials are released from the stack to environment via the emergency air purification system under the accident condition. Then the emergency air purification system should remove airborne radio-activities and should maintain proper pressure in the SA. We established the method to measure leak rate of the CV with closed reactor coolant pressure boundary although it is normally measured under opened reactor coolant pressure boundary as employed in LWRs. The CV leak rate test was carried out by the newly developed method and the expected performance was obtained. The SA and emergency air purification system were also confirmed by the performance test. We concluded that the reactor containment structures were fabricated to minimize the release of FPs in the postulated accidents with FP release from the reactor facilities. (author)

  9. Studies on the behavior of graphite structures irradiated in the Dragon Reactor. Dragon Project report

    Energy Technology Data Exchange (ETDEWEB)

    Everett, M. R.; Graham, L. W.; Ridealgh, F.

    1971-11-15

    Design data for the physical and mechanical property changes which occur in graphite structural and fuel body components irradiated in an HTR are largely obtained from small specimens tested in the laboratory and in materials test reactors. A brief data summary is given. This graphite physics data can be used to predict dimensional changes, internal stress generation and strength changes in the graphite materials of HTR fuel elements irradiated in the Dragon Reactor. In this paper, the results which have been obtained from post-irradiation examination of a number of fuel pins, are compared with prediction.

  10. Economic evaluation of small modular nuclear reactors and the complications of regulatory fee structures

    International Nuclear Information System (INIS)

    Vegel, Benjamin; Quinn, Jason C.

    2017-01-01

    Carbon emission concerns and volatility in fossil fuel resources have renewed world-wide interest in nuclear energy as a solution to growing energy demands. Several large nuclear reactors are currently under construction in the United States, representing the first new construction in over 30 years. Small Modular Reactors (SMRs) have been in design for many years and offer potential technical and economic advantages compared with traditionally larger reactors. Current SMR capital and operational expenses have a wide range of uncertainty. This work evaluates the potential for SMRs in the US, develops a robust techno-economic assessment of SMRs, and leverages the model to evaluate US regulatory fees structures. Modeling includes capital expenses of a factory facility and capital and operational expenses with multiple scenarios explored through a component-level capital cost model. Policy regarding the licensing and regulation of SMRs is under development with proposed annual US regulatory fees evaluated through the developed techno-economic model. Results show regulatory fees are a potential barrier to the economic viability of SMRs with an alternate fee structure proposed and evaluated. The proposed fee structure is based on the re-distribution of fees for all nuclear reactors under a single structure based on reactor thermal power rating. - Highlights: • Potential demand for new small modular nuclear power in the US is established. • Capital costs are broken down on component level and include factory production. • US regulatory fees structures are evaluated, results show potential barrier. • An additional fee structure is proposed and compared with current US fee structures.

  11. Nuclear heat source component design considerations for HTGR process heat reactor plant concept

    International Nuclear Information System (INIS)

    McDonald, C.F.; Kapich, D.; King, J.H.; Venkatesh, M.C.

    1982-01-01

    Using alternate energy sources abundant in the U.S.A. to help curb foreign oil imports is vitally important from both national security and economic standpoints. Perhaps the most forwardlooking opportunity to realize national energy goals involves the integrated use of two energy sources that have an established technology base in the U.S.A., namely nuclear energy and coal. The coupling of a high-temperature gas-cooled reactor (HTGR) and a chemical process facility has the potential for long-term synthetic fuel production (i.e., oil, gasoline, aviation fuel, hydrogen, etc.) using coal as the carbon source. Studies are in progress to exploit the high-temperature capability of an advanced HTGR variant for nuclear process heat. The process heat plant discussed in this paper has a 1170-MW(t) reactor as the heat source and the concept is based on indirect reforming, i.e., the high-temperature nuclear thermal energy is transported (via an intermediate heat exchanger (IHX)) to the externally located process plant by a secondary helium transport loop. Emphasis is placed on design considerations for the major nuclear heat source (NHS) components, and discussions are presented for the reactor core, prestressed concrete reactor vessel (PCRV), rotating machinery, and heat exchangers

  12. Control of activation levels to simplify waste management of fusion reactor ferritic steel components

    International Nuclear Information System (INIS)

    Wiffen, F.W.; Santoro, R.T.

    1983-01-01

    Activation characteristics of a material for service in the neutron flux of a fusion reactor first wall fall into three areas: waste management, reactor maintenance and repair, and safety. Of these, the waste management area is the most likely to impact the public acceptance of fusion reactors for power generation. The decay of the activity in steels within tens of years could lead to simplified waste disposal or possibly even to materials recycle. Whether or not these can be achieved will be controlled by (1) selection of alloying elements, (2) control of critical impurity elements, and (3) control of cross contamination from other reactor components. Several criteria can be used to judge the acceptability of potential alloying elements in iron, and to define the limits on content of critical impurity elements. One approach is to select and limit alloying additions on the basis of the activity. If material recycle is a goal, N, Al, Ni, Cu, Nb, and Mo must be excluded. If simplified waste storage by shallow land burial is the goal, regulations limit the concentration of only a few isotopes. For first-wall material that will be exposed to 9 MW-y/m 2 service, allowable initial concentration limits include (in at. ppM) Ni < 20,000; Mo < 3650; N < 3650, Cu < 2400; and Nb < 1.0. The other constituent elements of ferritic steels will not be limited. Possible substitutes for the molybdenum normally used to strengthen the steels include W, Ta, Ti, and V

  13. Identifying the Component Structure of Satisfaction Scales by Nonlinear Principal Components Analysis

    NARCIS (Netherlands)

    Manisera, M.; Kooij, A.J. van der; Dusseldorp, E.

    2010-01-01

    The component structure of 14 Likert-type items measuring different aspects of job satisfaction was investigated using nonlinear Principal Components Analysis (NLPCA). NLPCA allows for analyzing these items at an ordinal or interval level. The participants were 2066 workers from five types of social

  14. Component Degradation Susceptibilities As The Bases For Modeling Reactor Aging Risk

    International Nuclear Information System (INIS)

    Unwin, Stephen D.; Lowry, Peter P.; Toyooka, Michael Y.

    2010-01-01

    The extension of nuclear power plant operating licenses beyond 60 years in the United States will be necessary if we are to meet national energy needs while addressing the issues of carbon and climate. Characterizing the operating risks associated with aging reactors is problematic because the principal tool for risk-informed decision-making, Probabilistic Risk Assessment (PRA), is not ideally-suited to addressing aging systems. The components most likely to drive risk in an aging reactor - the passives - receive limited treatment in PRA, and furthermore, standard PRA methods are based on the assumption of stationary failure rates: a condition unlikely to be met in an aging system. A critical barrier to modeling passives aging on the wide scale required for a PRA is that there is seldom sufficient field data to populate parametric failure models, and nor is there the availability of practical physics models to predict out-year component reliability. The methodology described here circumvents some of these data and modeling needs by using materials degradation metrics, integrated with conventional PRA models, to produce risk importance measures for specific aging mechanisms and component types. We suggest that these measures have multiple applications, from the risk-screening of components to the prioritization of materials research.

  15. Data base formation for important components of reactor TRIGA MARK II

    Energy Technology Data Exchange (ETDEWEB)

    Jordan, R; Mavko, B; Kozuh, M [Inst. Jozef Stefan, Ljubljana (Slovenia)

    1992-07-01

    The paper represents specific data base formation for reactor TRIGA MARK II in Podgorica. Reactor operation data from year 1985 to 1990 were collected. Two groups of collected data were formed. The first group includes components data and the second group covers data of reactor scrams. Time related and demand related models were used for data evaluation. Parameters were estimated by classical method. Similar data bases are useful everywhere where components unavailabilities may have severe drawback. (author) [Slovenian] V referatu smo prikazali raziskavo, v okviru katere smo za raziskovalni reaktor TRIGA MARK II v Podgorici izoblikovali specificno bazo podatkov. Zbrali smo podatke obratovanja reaktorja od leta 1985 do 1990. Rezultate raziskave dogodkov smo razdelili v dve glavni skupini. V prvo spadajo obratovalni podatki o komponentah, v drugo skupino pa spadajo zagoni oz. zaustavitve reaktorja. Podatke smo ovrednotili z modelom v casovnem prostoru in z modelom na zahtevo. Parametre modelov smo dolocili s klasicno metodo. Opisane baze podatkov so uporabne povsod, kjer so lahko posledice nezanesljivega delovanja sistemov velike. [author].

  16. Comparison of control systems applied to the handling of radioactive reactor components

    International Nuclear Information System (INIS)

    Robinson, C.; Harris, E.G.; Dyer, P.C.; Williams, J.G.B.

    1985-01-01

    The first generation of nuclear power stations have individual reactors each incorporating complete facilities for servicing components and refuelling. In the later designs, each power station has two reactors which are connected by a central block. This central block contains one set of facilities to service both reactors, but to improve the station capability, some of these are to be replicated. The central block incorporates a hoist well which was used during construction for the accessing of complete components. On completion of this work, the physical size of the hoist well is such as to permit the incorporation of additional facilities if these are shown to be operationally and economically desirable. Since a number of years of power operation has elapsed, the advantages of back-fitting to existing fuel-handling facilities has been illustrated. Since the mechanical arrangements and operating procedures are substantially similar for both the original and new handling facilities, the paper will illustrate the control systems provided for each. The configuration of the system is arranged to have two channels of control which complies with the current standard requirements in the United Kingdom. These requirements are more stringent than when the existing facility was designed and constructed, as described in the relevant sections of the paper. The new system has been designed and is being manufactured to comply with the Central Electricity Generating Board standard for nuclear fuel route interlock and control systems. (author)

  17. Ageing evaluation model of nuclear reactors structural elements

    International Nuclear Information System (INIS)

    Ziliukas, A.; Jutas, A.; Leisis, V.

    2002-01-01

    In this article the estimation of non-failure probability by random faults on the structural elements of nuclear reactors is presented. Ageing is certainly a significant factor in determining the limits of nuclear plant lifetime or life extensions. Usually the non failure probability rates failure intensity, which is characteristic for structural elements ageing in nuclear reactors. In practice the reliability is increased incorrectly because not all failures are fixed and cumulated. Therefore, the methodology with using the fine parameter of the failures flow is described. The comparison of non failure probability and failures flow is carried out. The calculation of these parameters in the practical example is shown too. (author)

  18. Safety classification of nuclear power plant systems, structures and components

    International Nuclear Information System (INIS)

    1992-01-01

    The Safety Classification principles used for the systems, structures and components of a nuclear power plant are detailed in the guide. For classification, the nuclear power plant is divided into structural and operational units called systems. Every structure and component under control is included into some system. The Safety Classes are 1, 2 and 3 and the Class EYT (non-nuclear). Instructions how to assign each system, structure and component to an appropriate safety class are given in the guide. The guide applies to new nuclear power plants and to the safety classification of systems, structures and components designed for the refitting of old nuclear power plants. The classification principles and procedures applying to the classification document are also given

  19. The potential for reducing the radiological consequences of reactor decommissioning through selection of construction materials for activated components

    International Nuclear Information System (INIS)

    Woollam, P.B.

    1984-08-01

    This report considers whether it may be possible to reduce the radiological consequences of reactor decommissioning by careful attention to the specification of the elemental concentration of materials used in the reactor's construction. In particular, consideration is given to the potential for reduction of the concentration of elements known to activate to long lived daughter isotopes. Two particular areas are addressed, both applied to Sizewell 'B' PWR. The first is the choice of raw materials for the construction of the concrete bioshield to minimise future waste arisings. The second is the specification of some trace element concentrations in the steel pressure vessel and reactor internal structures to minimise personnel exposure at decommissioning time. The report presents extensive analyses of many of the candidate raw materials for Sizewell 'B' concrete, including PFA, and derives the radiological consequences for the eventual disposal of these materials to a hypothetical municipal land fill waste site. Data are also presented on the concentrations of important elements activating to gamma emitting daughters in type 304 stainless steels, leading to an assessment of likely dose equivalent rates at decommissioning time from the pressure vessel and from the internal components. (author)

  20. Design study of blanket structure for tokamak experimental fusion reactor

    International Nuclear Information System (INIS)

    1979-11-01

    Design study of the blanket structure for JAERI Experimental Fusion Reactor (JXFR) has been carried out. Studied here were fabrication and testing of the blanket structure (blanket cells, blanket rings, piping and blanket modules), assembly and disassembly of the blanket module, and monitering and testing technique. Problems in design and fabrication of the blanket structure could be revealed. Research and development problems for the future were also disclosed. (author)

  1. Reliability of redundant structures of nuclear reactor protection systems

    International Nuclear Information System (INIS)

    Vojnovic, B.

    1983-01-01

    In this paper, reliability of various redundant structures of PWR protection systems has been analysed. Structures of reactor tip systems as well as the systems for activation of safety devices have been presented. In all those systems redundancy is achieved by means of so called majority voting logic ('r out of n' structures). Different redundant devices have been compared, concerning probability of occurrence of safe as well as unsafe failures. (author)

  2. Apparatus for securing structural tubes in nuclear reactor fuel assemblies

    International Nuclear Information System (INIS)

    Kerry, J.S.

    1987-01-01

    This patent describes a nuclear reactor fuel assembly having a structural tube with a predetermined inside diameter, a generally cylindrical insert of an axial length substantially smaller than the axial length of the structural tube and having a generally cylindrical passageway of a predetermined diameter smaller than the predetermined inside diameter for providing an effectively reduced inside diameter for the structural tube. The insert comprises: means, having an outside diameter approximately equal to the predetermined inside diameter, for coaxially centering the insert within the structural tube; forming lobes, operable when expanded to locally deform against the structural tube thereby locking the insert within the structural tube

  3. Study on advanced structural design for commercialized fast breeder reactors

    International Nuclear Information System (INIS)

    Morishita, Masaki; Aoto, Kazumi; Kasahara, Naoto; Asayama, Tai

    2003-05-01

    Japan Nuclear Cycle Development Institute (JNC) and Japan Atomic Power Company (JAPC) launched joint research programs on structural design and three-dimensional seismic isolation technologies, as part of the supporting R and D activities for the feasibility studies on commercialized fast breeder reactor cycle systems. A research project by JAPC under the auspices of the Ministry of Economy, Trade, and Industry (METI) with technical support by JNC is included in this joint study. This report contains the results of the research on the structural design technology. The research scope was identified as (1) FDS (FBR Design Standard), (2) Standardization of new material, and (3) System Based Code for Integrity, and the results of this year's studies are summarized as follows. (1) FDS (FBR Design Standard). R and D policy of Phase II study considering to the newest needs demanded for structural design were clarified, and R and D items were settled concretely. As for failure criteria, preliminary ratcheting-fatigue tests were conducted with rational methods, and Negligible Creep curves that settle creep designing region rationally were expanded their applying condition. R and D policy and items of guideline for inelastic analysis were clarified, and analyzing methodologies were studied. Summering up exemplification of thermal load prediction methodologies were progressed. To predict thermal striping loads, advanced frequency response function of thermal stress, and fatigue evaluation methodology were studied. (2) Standardization of new material. As for candidate 12-chromium stainless steel (added tungsten, non-added tungsten), that is expected to improve strength of components of commercialized FR, short and medium-term material tests, and creep-fatigue tests at small strain range were conducted. As for above candidate steel, TIG (Tungsten Inert Gas) welding method was recommended. Requirements to standardize new materials and index to select new materials were studied

  4. Performance of materials in the component cooling water systems of pressurized water reactors

    International Nuclear Information System (INIS)

    Lee, B.S.

    1993-01-01

    The component cooling water (CCW) system provides cooling water to several important loads throughout the plant under all operating conditions. An aging assessment CCW systems in pressurized water reactors (PWRs) was conducted as part of Nuclear Plant Aging Research Program (NPAR) instituted by the US Nuclear Regulatory Commission. This paper presents some of the results on the performances of materials in respect of their application in CCW Systems. All the CCW system failures reported to the Nuclear Plant Reliability Data System (NPRDS) from January 1988 to June 1990 were reviewed; it is concluded that three of the main contributors to CCW system failures are valves, pumps, and heat exchangers. This study identified the modes and causes of failure for these components; most of the causes for the aging-related failures could be related to the performance of materials. Also, in this paper the materials used for these components are reviewed, and there aging mechanisms under CCW system conditions are discussed

  5. Development of fiber-delivered laser peening system to prevent stress corrosion cracking of reactor components

    International Nuclear Information System (INIS)

    Sano, Y.; Kimura, M.; Yoda, M.; Mukai, N.; Sato, K.; Uehara, T.; Ito, T.; Shimamura, M.; Sudo, A.; Suezono, N.

    2001-01-01

    The authors have developed a system to deliver water-penetrable intense laser pulses of frequency-doubled Nd:YAG laser through optical fiber. The system is capable of improving a residual stress on water immersed metal material remotely, which is effective to prevent the initiation of stress corrosion cracking (SCC) of reactor components. Experimental results showed that a compressive residual stress with enough amplitude and depth was built in the surface layer of type 304 stainless steel (SUS304) by irradiating laser pulses through optical fiber with diameter of 1 mm. A prototype peening head with miniaturized dimensions of 88 mm x 46 mm x 25 mm was assembled to con-firm the accessibility to the heat affected zone (HAZ) along weld lines of a reactor core shroud. The accessibility was significantly improved owing to the flexible optical fiber and the miniaturized peening head. The fiber delivered system opens up the possibility of new applications of laser peening. (author)

  6. Transference of know-how for the fabrication of heavy components for nuclear power reactors

    International Nuclear Information System (INIS)

    Meier, F.

    1977-01-01

    1) Heavy components for nuclear power reactors. Reactor pressure vessels with total weight of 540 tons; steam generators: heat exchangers with U-type tube bundles, total weight 420 tons. 2) Choice of know-how recipient. Technical criteria, i.e. manufacturing facilities, existing quality assurance system, location of the workshops, possibilities for training, infrastructures. 3. Measures for transferring know-how to a newly established company. Planning and erection of the factory: organisational set up of the company; personnel selection and training; transfer of documentation; transfer of know-how that cannot be transferred in a written form. 4) Contracts for assuring the transfer of know-how. Stipulation of mutual rights and obligations of the know-how owner and receiver in individual contracts: engineering services contract, technical information contract, personnel training contract, license contract. (orig.) [de

  7. The application of mechanical desktop in the design of the reactor core structure of China advanced research reactor

    International Nuclear Information System (INIS)

    Lang Ruifeng

    2002-01-01

    The three-dimensional parameterization design method is introduced to the design of reactor core structure for China advanced research reactor. Based on the modeling and dimension variable driving of the main parts as well as the modification of dimension variable, the preliminary design and modification of reactor core is carried out with high design efficiency and quality as well as short periods

  8. Reactor pressure vessel structural integrity research

    International Nuclear Information System (INIS)

    Pennell, W.E.; Corwin, W.R.

    1994-01-01

    Development continues on the technology used to assess the safety of irradiation-embrittled nuclear reactor pressure vessels (RPVs) containing flaws. Fracture mechanics tests on RPV steel, coupled with detailed elastic-plastic finite-element analyses of the crack-tip stress fields, have shown that (1) constraint relaxation at the crack tip of shallow surface flaws results in increased data scatter but no increase in the lower-bound fracture toughness, (2) the nil ductility temperature (NDT) performs better than the reference temperature for nil ductility transition (RT NDT ) as a normalizing parameter for shallow-flaw fracture toughness data, (3) biaxial loading can reduce the shallow-flaw fracture toughness, (4) stress-based dual-parameter fracture toughness correlations cannot predict the effect of biaxial loading on shallow-flaw fracture toughness because in-plane stresses at the crack tip are not influenced by biaxial loading, and (5) an implicit strain-based dual-parameter fracture toughness correlation can predict the effect of biaxial loading on shallow-flaw fracture toughness. Experimental irradiation investigations have shown that (1) the irradiation-induced shift in Charpy V-notch vs temperature behavior may not be adequate to conservatively assess fracture toughness shifts due to embrittlement, and (2) the wide global variations of initial chemistry and fracture properties of a nominally uniform material within a pressure vessel may confound accurate integrity assessments that require baseline properties

  9. Structural materials for fusion reactor blanket systems

    International Nuclear Information System (INIS)

    Bloom, E.E.; Smith, D.L.

    1984-01-01

    Consideration of the required functions of the blanket and the general chemical, mechanical, and physical properties of candidate tritium breeding materials, coolants, structural materials, etc., leads to acceptable or compatible combinations of materials. The presently favored candidate structural materials are the austenitic stainless steels, martensitic steels, and vanadium alloys. The characteristics of these alloy systems which limit their application and potential performance as well as approaches to alloy development aimed at improving performance (temperature capability and lifetime) will be described. Progress towards understanding and improving the performance of structural materials has been substantial. It is possible to develop materials with acceptable properties for fusion applications

  10. Structural design and dynamic analysis of underground nuclear reactor containments

    International Nuclear Information System (INIS)

    Kierans, T.W.; Reddy, D.V.; Heale, D.G.

    1975-01-01

    Present actual experience in the structural design of undeground containments is limited to only four rather small reactors all located in Europe. Thus proposals for future underground reactors depend on the transposition of applicable design specifications, constraints and criteria from existing surface nuclear power plants to underground, and the use of many years of experience in the structural design of large underground cavities and cavity complexes for other purposes such as mining, hydropower stations etc. An application of such considerations in a recent input for the Underground Containment sub-section of the Seismic Task Group Report to the ASCE Committee for Nuclear Structures and Materials is presented as follows: underground concept considerations, siting criteria and structural selection, structural types, analytical and semi-analytical approaches, design and other miscellaneous considerations

  11. Research and development issues for fast reactor structural design standard (FDS)

    International Nuclear Information System (INIS)

    Kasahara, Naoto; Ando, Masanori; Morishita, Masaki

    2003-01-01

    For realization of safe and economical fast reactor (FR) plants, Japan Nuclear Cycle Development Institute (JNC) and Japan Atomic Power Company (JAPC) are cooperating on 'Feasibility Study on Commercialized FR Cycle Systems'. To certify the design concepts and validate their structural integrity, the research and development of 'Fast Reactor Structural Design Standard (FDS)' is recognized as an essential theme. FDS considers general characteristics of FRs and design needs for their rationalization. Three main subjects were settled in research and development issues of FDS. One is rationalization of failure criteria' taking characteristic design conditions into account. Next is development of 'a guideline on inelastic analysis for design' in order to predict elastic plastic and creep behaviours of high temperature components. Furthermore, efforts are being made toward preparing a guideline on thermal loads modeling' for FR component design where thermal loads are dominant. (author)

  12. Packaging, Transportation, and Disposal Logistics for Large Radioactively Contaminated Reactor Decommissioning Components

    International Nuclear Information System (INIS)

    Lewis, Mark S.

    2008-01-01

    The packaging, transportation and disposal of large, retired reactor components from operating or decommissioning nuclear plants pose unique challenges from a technical as well as regulatory compliance standpoint. In addition to the routine considerations associated with any radioactive waste disposition activity, such as characterization, ALARA, and manifesting, the technical challenges for large radioactively contaminated components, such as access, segmentation, removal, packaging, rigging, lifting, mode of transportation, conveyance compatibility, and load securing require significant planning and execution. In addition, the current regulatory framework, domestically in Titles 49 and 10 and internationally in TS-R-1, does not lend itself to the transport of these large radioactively contaminated components, such as reactor vessels, steam generators, reactor pressure vessel heads, and pressurizers, without application for a special permit or arrangement. This paper addresses the methods of overcoming the technical and regulatory challenges. The challenges and disposition decisions do differ during decommissioning versus component replacement during an outage at an operating plant. During decommissioning, there is less concern about critical path for restart and more concern about volume reduction and waste minimization. Segmentation on-site is an available option during decommissioning, since labor and equipment will be readily available and decontamination activities are routine. The reactor building removal path is also of less concern and there are more rigging/lifting options available. Radionuclide assessment is necessary for transportation and disposal characterization. Characterization will dictate the packaging methodology, transportation mode, need for intermediate processing, and the disposal location or availability. Characterization will also assist in determining if the large component can be transported in full compliance with the transportation

  13. Assuring the reliability of structural components - experimental data and non-destructive examination requirements

    International Nuclear Information System (INIS)

    Lucia, A.C.

    1984-01-01

    The probability of failure of a structural component can be estimated by either statistical methods or a probabilistic structural reliability approach (where the failure is seen as a level crossing of a damage stochastic process which develops in space and in time). The probabilistic approach has the advantage that it makes available not only an absolute value of the failure probability but also a lot of additional information. The disadvantage of the probabilistic approach is its complexity. It is discussed for the following situations: reliability of a structural component, material properties, data for fatigue crack growth evaluation, a bench mark exercise on reactor pressure vessel failure probability computation, and non-destructive examination for assuring a given level of structural reliability. (U.K.)

  14. Structural design of SBWR reactor building complex using microcomputers

    International Nuclear Information System (INIS)

    Mandagi, K.; Rajagopal, R.S.; Sawhney, P.S.; Gou, P.F.

    1993-01-01

    The design concept of Simplified Boiling Water Reactor (SBWR) plant is based on simplicity and passive features to enhance safety and reliability, improve performance, and increase economic viability. The SBWR utilizes passive systems such as Gravity Driven Core-Cooling System (GDCS) and Passive Containment Cooling System (PCCS). To suit these design features the Reactor Building (RB) complex of the SBWR is configured as an integrated structure consisting of a cylindrical Reinforced Concrete Containment Vessel (RCCV) surrounded by square reinforced concrete safety envelope and outer box structures, all sharing a common reinforced concrete basemat. This paper describes the structural analysis and design aspects of the RB complex. A 3D STARDYNE finite element model has been developed for the structural analysis of the complex using a PC Compaq 486/33L microcomputer. The structural analysis is performed for service and factored load conditions for the applicable loading combinations. The dynamic responses of containment structures due to pool hydrodynamic loads have been calculated by an axisymmetric shell model using COSMOS/M program. The RCCV is designed in accordance with ASME Section 3, Division 2 Code. The rest of the RB which is classified as Seismic Category 1 structure is designed in accordance with the ACI 349 Code. This paper shows that microcomputers can be efficiently used for the analysis and design of large and complex structures such as RCCV and Reactor Building complex. The use of microcomputers can result in significant savings in the computational cost compared with that of mainframe computers

  15. General Description of the Mechanic Design of the Pressure Vessel and the Internal Mechanical Component of the CAREM Reactor

    International Nuclear Information System (INIS)

    Diez, F.; Horro, R.

    2000-01-01

    This paper presents a brief description of the CAREM reactor pressure vessel and its main internal mechanical components and summarizes the functional requirements and approaches applied for their design, together with a review of the normative applicable in each case

  16. Tube sheet structural analysis of intermediate heat exchanger for fast breeder reactor 'Monju'

    International Nuclear Information System (INIS)

    Nakagawa, Y.; Ueno, T.; Fukuda, Y.; Ichimiya, M.

    1983-01-01

    The Prototype Fast Breeder Reactor 'Monju' is the first power generating fast breeder reactor in Japan. We have been designing the components of the plant for manufacturing. Among these is the intermediate heat exchanger (IHX) which exchanges heat between primary and secondary sodium loop. The tube sheet of IHX (shell to ligament junction) is a difficult area from the view point of structural strength design under elevated temperature. To validate the structural integrity of tube sheet we performed the series of inelastic analysis and tube sheet thermal shock test using test pieces and half scale model of actual design. The results of inelastic analyses showed there is little progressive deformation around shell to ligament structural discontinuous junction. Furthermore, thermal shock tests showed no increase of an accumulative deformation. By these analyses and experiments, structural reliability of tube sheet could be shown. (author)

  17. Use of the local-global concept in detecting component vibration in reactors

    International Nuclear Information System (INIS)

    Al-Ammar, M.A.

    1981-01-01

    The local-global concept, based on the detector adjoint function, was used to develop the response of a detector to an absorber vibrating in one dimension. A one-dimensional two-group diffusion code was developed to calculate the frequency dependent detector response as a function of detector and absorber positions for the coupled-core UTR-10 reactor. Results from this code indicated the best possible detector and absorber locations, where more detailed calculations were made using a two-group, three-dimensional diffusion code with finite detector and absorber volumes. An experiment was then designed, for the chosen positions, using a vibrating cadmium absorber with a detector on each side. The assembly was placed in the vertical central stringer of the reactor. Investigations were carried out for vibrations in two flux gradients and experimental data were analyzed in the frequency domain using a microcomputer-based data acquisition system. The experimental investigation showed the validity of the local-global concept. A normalized outputs cross power spectral density was developed that correctly predicted the different flux tilts in the two flux gradients. It was also shown that the frequency response of the local component had a wide plateau region. Monitoring the behavior of the normalized cross power spectral density was thought to be a promising indicator for the detection and localization of malfunctioning vibrating components. It might also be used to detect flux irregularities in the vicinity of a vibrating component

  18. Mechanical design philosophy for the graphite components of the core structure of an HTGR

    International Nuclear Information System (INIS)

    Bodmann, E.

    1987-01-01

    Parallel to the layout and design of the graphite components for THTRs and the succeeding high temperature reactor projects, the design methods for graphite components have been improved over the years. The aim of this works is to develop the design methods which take into account both the particular properties of graphite and the particular functions of the components. Because of the close relation ship between materials and design codes, this development work has progressed with the development, testing and qualification of German reactor graphite. In this paper, the experience in this field of Hochtemperatur Reaktorbau GmbH and the results of the work and approach to the design problems are reported. The example of a HTR 500 design for a 550 MWe power station is taken up, and the core structure is explained. The graphite components are divided into three classes according to the stress limits. The loading of these components is reviewed. The aim of the design is not the complete avoidance of failure, but to avoid the failure of a single component from leading to a disadvantageous consequence which is not allowable. The classification of loading events, Weibull statistics and maximum allowable stress, the formation of the permissible stress, the assessment of stress due to multiaxial loading and so on are described. (Kako, I.)

  19. Structural strength analysis of lug component for RTP fuel transfer cask

    International Nuclear Information System (INIS)

    Ahmad Nabil Abdul Rahim; Tonny Lanyau; Mazleha Maskin

    2012-01-01

    A three decade aging PUSPATI TRIGA Reactor will soon require spent fuels to be replaced with fresh fuels. Initial procedure requires these spent fuels be cooled down due to high temperature and high radioactivity resulted from fission products activities and then transferred out of reactor tank into a spent fuel pool. Performing this task will need a fuel transfer cask (FTC) to provide sufficient radiation shielding for the workers during fuels transferring task. Currently the only economically viable route FTC to get into reactor tank is through the top of reactor tank using crane. The huge and high density lead filled FTC will be hanging throughout the task of transferring each fuel element thus FTC is subjected to tensile stress during its operation. Lug component is anticipated to be one of the main weaknesses of FTC design subjected under tensile load due to it diminutive in size but very large amount of stress to be introduced on it. Therefore lugs component requires an adequate stress analysis to prove the design reliability. This paper will discuss the structural strength using conventional engineering calculation based on best estimate approach. (author)

  20. Residual life estimation of cracked aircraft structural components

    OpenAIRE

    Maksimović, Mirko S.; Vasović, Ivana V.; Maksimović, Katarina S.; Trišović, Nataša; Maksimović, Stevan M.

    2018-01-01

    The subject of this investigation is focused on developing computation procedure for strength analysis of damaged aircraft structural components with respect to fatigue and fracture mechanics. For that purpose, here will be defined computation procedures for residual life estimation of aircraft structural components such as wing skin and attachment lugs under cyclic loads of constant amplitude and load spectrum. A special aspect of this investigation is based on using of the Strain Energy Den...

  1. Use of stainless steel as structural materials in reactor cores

    International Nuclear Information System (INIS)

    Teodoro, C.A.

    1990-01-01

    Austenitic stainless steels are used as structural materials in reactor cores, due to their good mechanical properties at working temperatures and high generalized corrosion resistance in aqueous medium. The objective of this paper is to compare several 300 series austenitic stainless steels related to mechanical properties, localized corrosion resistance (SCC and intergranular) and content of delta ferrite. (author)

  2. Periodic reviews of structural integrity of gas-cooled reactors

    International Nuclear Information System (INIS)

    Banks, P.J.; Stokoe, T.Y.; Thomas, D.L.

    1995-01-01

    Nuclear Electric operates 12 gas-cooled reactor power stations which have been in service for between 5 and 30 years. Periodically, comprehensive reviews of the safety cases are carried out for each station. The approach followed in these reviews in respect of structural integrity is outlined with the use of illustrative examples. (author)

  3. A review of the US joining technologies for plasma facing components in the ITER fusion reactor

    International Nuclear Information System (INIS)

    Odegard, B.C. Jr.; Cadden, C.H.; Watson, R.D.; Slattery, K.T.

    1998-02-01

    This paper is a review of the current joining technologies for plasma facing components in the US for the International Thermonuclear Experimental Reactor (ITER) project. Many facilities are involved in this project. Many unique and innovative joining techniques are being considered in the quest to join two candidate armor plate materials (beryllium and tungsten) to a copper base alloy heat sink (CuNiBe, OD copper, CuCrZr). These techniques include brazing and diffusion bonding, compliant layers at the bond interface, and the use of diffusion barrier coatings and diffusion enhancing coatings at the bond interfaces. The development and status of these joining techniques will be detailed in this report

  4. Method and alloys for fabricating wrought components for high-temperature gas-cooled reactors

    International Nuclear Information System (INIS)

    Thompson, L.D.; Johnson, W.R.

    1983-01-01

    Wrought, nickel-based alloys, suitable for components of a high-temperature gas-cooled reactor exhibit strength and excellent resistance to carburization at elevated temperatures and include aluminum and titanium in amounts and ratios to promote the growth of carburization resistant films while preserving the wrought character of the alloys. These alloys also include substantial amounts of molybdenum and/or tungsten as solid-solution strengtheners. Chromium may be included in concentrations less than 10% to assist in fabrication. Minor amounts of carbon and one or more carbide-forming metals also contribute to high-temperature strength. The range of compositions of these alloys is given. (author)

  5. Proceedings of a specialist meeting on the ultrasonic inspection of reactor components

    International Nuclear Information System (INIS)

    1976-01-01

    Beside synthesis of two conferences on nondestructive testing and on inspection, the contributions of this conference are reporting experimental observations and research works on ultrasonic techniques, methods, procedures (pre-service or in-service) and equipment for the inspection of nuclear reactor components (pressure vessels, tubing and piping), generally in stainless steel (often austenitic or ferritic) material or in zirconium alloy. Some contributions are also dealing with the relationship between material microstructure and ultrasonic inspection method and equipment, or with the detection and sizing precision of flaws (cracks)

  6. Justification and manufacturing quality assurance for the use of hot Isostatically pressed, reactor coolant system components in PWR plant

    International Nuclear Information System (INIS)

    Sulley, J. L.; Hookham, I. D.

    2008-01-01

    This paper presents an overview of the work undertaken by Rolls-Royce to introduce Hot Isostatically Pressed (HIP) components into Pressurised Water Reactor plant. It presents the work from a design justification and manufacturing quality assurance perspective, rather than from a pure metallurgical perspective, although some metallurgical and mechanical property comparisons with the traditional forged material are presented. Although the HIP process is not new, it was new in its application to Rolls-Royce designed nuclear reactor plant. In order to satisfy the regulatory requirement of 'Proven Engineering Practices' with regard to the introduction of new material processes, and to provide a robust manufacturing substantiation leg of a multi-legged safety case, Rolls-Royce has implemented an evolving, staged approach, starting with HIP bonding of solid valve seats into small bore valve pressure boundaries. This was followed by powder HIP consolidation of leak-limited, thin-walled toroids, and has culminated in the powder HIP consolidation of components, such as steam generator headers, large bore valves and pipe sections. The paper provides an overview of each of these stages and the approach taken with respect to justification. The paper describes the benefits that Rolls-Royce has realised so far through the introduction of HIPed components, and improvements planned for the future. Structural integrity benefits are described, such as improved grain structure, mechanical properties, and ultrasonic inspection. Project-based benefits are also described, such as provision of an alternative strategic sourcing route, cost and lead-time reduction. A full description is provided of key quality assurance steps applied to the process to ensure a high quality product is delivered commensurate with a high integrity nuclear application. 2008 Rolls-Royce plc. (authors)

  7. Proceedings of the 17th international conference on structural mechanics in reactor technology

    International Nuclear Information System (INIS)

    2003-01-01

    The conference was divided into the following divisions and subdivisions: DIVISION A: Plenary lectures and panel; DIVISION B: Computational mechanics (Structural and thermal analysis; High-non linear analysis, material behaviour; Vibration and fluid dynamics analysis); DIVISION C: Fuel and core structures (Fuel vibration and fretting; Fuel design and constitutive modelling; Fuel failure under operation and accident conditions; Fuel failure under operation and accident conditions; Components and material behaviour under irradiation; Integrity of fuel systems under transient conditions); DIVISION D: Aging, Life Extension and Licence Renewal (International Regulatory and Economic Perspectives; Utility perspectives, WWER technology; Fatigue, corrosion and crack issues; Component integrity; Aging assessment and monitoring; Containment and other structures); DIVISION F: Design methods and rules for components (International codes and standards; Tube, piping codes and standards; Analyses; Fatigue and life assessment; Creep; Bolted connections and gaskets); DIVISION G: Fracture mechanics (Reactor pressure vessel integrity; Dynamic loading; Fracture considerations for various applications; Failure assessment of Zr alloy; Pipe integrity; Integrity of welds; Failure of non-metallic materials; Leak before break (LBB); Corrosion aspects); DIVISION H: Concrete Containment and Other Structures (Concrete materials and performance; Tests of scale prestressed concrete containment vessel; Shear wall test and analysis; Structural analysis and containment design; Structural integrity and analysis); DIVISION J: Analysis and design for dynamic and extreme load (Vibration of shells and plates; Impact analysis; Piping vibration; Structural dynamics; Experimental and other topics); DIVISION K: Seismic analysis, design and qualification (General seismic issues; Ground motion and sitting; Soil-structure interaction; Seismic response of structures; Seismic re-evaluation; Seismic response and

  8. Structural aging program to assess the adequacy of critical concrete components in nuclear power plants

    International Nuclear Information System (INIS)

    Naus, D.J.; Marchbanks, M.F.; Oland, C.B.; Arndt, E.G.

    1989-01-01

    The Structural Aging (SAG) Program is carried out by the Oak Ridge National Laboratory (ORNL) under sponsorship of the United States Nuclear Regulatory Commission (USNRC). The Program has evolved from preliminary studies conducted to evaluate the long-term environmental challenges to light-water reactor safety-related concrete civil structures. An important conclusion of these studies was that a damage methodology, which can provide a quantitative measure of a concrete structure's durability with respect to potential future requirements, needs to be developed. Under the SAG Program, this issue is being addressed through: establishment of a structural materials information center, evaluation of structural component assessment and repair technologies, and development of a quantitative methodology for structural aging determinations. Progress to date of each of these activities is presented as well as future plans. 7 refs., 5 figs

  9. Present status of SiCf/SiC composites as low-activation structural materials of fusion reactor in Japan

    International Nuclear Information System (INIS)

    Kohyama, A.; Katoh, Y.; Hasegawa, A.; Noda, T.

    2001-01-01

    The outline of research subjects on SiCf/SiC composites to apply to the structural components of fusion reactors are described and present status on material development of SiCf/SiC composites in Japan is reviewed. Irradiation experiments of the composites using fission reactors conducted by international collaborations to clarify their radiation response and to optimize the fabrication processes are introduced. (author)

  10. Mechanical structure and problem of thorium molten salt reactor

    International Nuclear Information System (INIS)

    Kamei, Takashi

    2011-01-01

    After Fukushima Daiichi accident, there became great interest in Thorium Molten Salt Reactor (MSR) for the safety as station blackout leading to auto drainage of molten salts with freeze valve. This article described mechanical structure of MSR and problems of materials and pipes. Material corrosion problem by molten salts would be solved using modified Hastelloy N with Ti and Nb added, which should be confirmed by operation of an experimental reactor. Trends in international activities of MSR were also referred including China declaring MSR development in January 2011 to solve thorium contamination issues at rare earth production and India rich in thorium resources. (T. Tanaka)

  11. Mechanical properties along interfaces of bonded structures in fusion reactors

    International Nuclear Information System (INIS)

    Hassan, M.H.; Kulcinski, G.L.

    1993-01-01

    Proper assessment of the mechanical properties along interfaces of bonded structures currently used in many fusion reactor designs is essential to compare the different fabrication techniques. A Mechanical Properties Microprobe (MPM) was used to measure hardness and Young's modules along the interfaces of Be/Cu bonded structure. The MPM was able to distinguish different fabrication techniques by a direct measurement of the hardness, Young's modules, and H/E 2 which reflects the ability of deformation of the interfacial region

  12. Results of 1989/90 research and development activities at KfK Institute for Reactor Components

    International Nuclear Information System (INIS)

    1991-03-01

    R and D activities at IRB (Institut fuer Reaktorbauelemente - Institute for Reactor Components) are dedicated to thermodynamics and fluid dynamics. Emphasis is on the design of nuclear reactor and fusion reactor components. Environmental engineering was added recently. Most activities are applications-oriented. Fundamental investigations focus on energy research and energy technology. The activities are carried out in the framework of different projects (PKF/nuclear fusion, PSF/nuclear safety, PSU/pollution control). Points of main effort are the development of basic liquid-metal-cooled blanket solutions, investigations on natural convection in reactor ranks, and the cooling properties of future containments for pressurized water reactors in the case of nuclear fusion accidents. (orig./GL) [de

  13. Application of probabilistic fracture mechanics to the reliability analysis of pressure-bearing reactor components

    International Nuclear Information System (INIS)

    Schmitt, W.; Roehrich, E.; Wellein, R.

    1977-01-01

    Since no failures in the primary reactor components have been reported so far, it is impossible to estimate the failure probability of those components just by means of statistics. Therefore the way of probabilistic fracture mechanics has been proposed. Here the material properties, the loads and the crack distributions are treated as statistical variables with certain distributions. From the distributions of these data probability density functions can be established for the loading of a component (e.g. the stress intensity factor) as well as for the resistance of this component (e.g. the fracture toughness). From these functions the failure probability for a given failure mode (e.g. brittle fracture) is easily obtained either by the application of direct integration procedures which are shortly reviewed here, or by the use of Monte Carlo techniques. The most important part of the concept is the collection of a sufficiently large amount of raw data from different sources (departments within the company or external). These data need to be processed so that they can be transformed into probability density functions. The method of data collection and processing in terms of histograms, plots of probability density functions etc, is described. The choice of the various types of distribution functions is discussed. As an example the derivation of the probability density function for cracks of a given size in a component is presented. (Auth.)

  14. Investigation of the local component of power-reactor noise via diffusion theory

    International Nuclear Information System (INIS)

    Kosaly, G.

    1975-03-01

    The aim of the paper is to provide a theoretical background for the phenomenological model, which postulates the existence of a local component in the neutron noise of a light water cooled boiling water reactor. After the introductory review of the phenomenological model, noise calculation are performed by help of the one-group and two-group diffusion theory. Only in the two-group diffusion model it is succeeded to find a term in the response to a propagating disturbance of density which results in a small volume of neutrons physical sensivity around the point of observation. The problem, whether this local component can be a dominating term in the solution or not, is investigated in the Appenix. (Sz.Z.)

  15. Protein structure similarity from principle component correlation analysis

    Directory of Open Access Journals (Sweden)

    Chou James

    2006-01-01

    Full Text Available Abstract Background Owing to rapid expansion of protein structure databases in recent years, methods of structure comparison are becoming increasingly effective and important in revealing novel information on functional properties of proteins and their roles in the grand scheme of evolutionary biology. Currently, the structural similarity between two proteins is measured by the root-mean-square-deviation (RMSD in their best-superimposed atomic coordinates. RMSD is the golden rule of measuring structural similarity when the structures are nearly identical; it, however, fails to detect the higher order topological similarities in proteins evolved into different shapes. We propose new algorithms for extracting geometrical invariants of proteins that can be effectively used to identify homologous protein structures or topologies in order to quantify both close and remote structural similarities. Results We measure structural similarity between proteins by correlating the principle components of their secondary structure interaction matrix. In our approach, the Principle Component Correlation (PCC analysis, a symmetric interaction matrix for a protein structure is constructed with relationship parameters between secondary elements that can take the form of distance, orientation, or other relevant structural invariants. When using a distance-based construction in the presence or absence of encoded N to C terminal sense, there are strong correlations between the principle components of interaction matrices of structurally or topologically similar proteins. Conclusion The PCC method is extensively tested for protein structures that belong to the same topological class but are significantly different by RMSD measure. The PCC analysis can also differentiate proteins having similar shapes but different topological arrangements. Additionally, we demonstrate that when using two independently defined interaction matrices, comparison of their maximum

  16. Development of radiation resistant structural materials utilizing fission research reactors in Japan (Role of research reactors)

    International Nuclear Information System (INIS)

    Shikama, T.; Tanigawa, H.; Nozawa, T.; Muroga, T.; Aoyama, T.; Kawamura, H.; Ishihara, M.; Ito, C.; Kaneda, S.; Mimura, S.

    2009-01-01

    Structural materials for next-generation nuclear power systems should have a good radiation resistance, where the expected accumulation dose will largely exceed 10 dpa. Among several candidate materials, materials of five categories, 1. Austenitic steels, including high nickel alloys, 2. Low activation ferritic martensitic steels, 3. ODS steels (austenitic and ferritic), 4. Vanadium based alloys, 5. Silicon carbide composites (SiC/SiCf). All have been most extensively studied in Japan, in collaboration among industries, national institutes such as Japan Atomic Energy Agency (JAEA), National Institute for Fusion Science (NIFS) and National Institute for Materials Science (NIMS), and universities. The high nickel base alloys were studied for their low swelling behaviors mainly by the NIMS and the austenitic steels are studied for their reliable engineering data base and their reliable performance in irradiation environments mainly by the JAEA, mainly for their application in the near-term projects such as the ITER and the Sodium Cooled Fast Reactors. The most extensive studies are now concentrated on the Low Activation Ferritic Marsensitic steels and ODS steels, for their application in a demonstration fusion reactor and prototype sodium cooled fast reactors. Fundamental studies on radiation effects are carried out, mainly utilizing Japan Materials Testing Rector (JMTR) with its flexible irradiation ability, up to a few dpa. For higher dpa irradiation, a fast test reactor, JOYO is utilized up to several 10s dpa. Some international collaborations such as Japan/USA and Japan/France are effective to utilize reactors abroad, such as High Flux Isotope Reactor (HFIR) of Oak Ridge National Laboratory, and sodium cooled high flux fast reactors in France. Silicon carbide based composites are extensively studied by university groups led by Kyoto University and the JAEA. For their performance in heavy irradiation environments, the Japan/USA collaboration plays an important role

  17. Structural design of nuclear reactor machinery and equipment

    International Nuclear Information System (INIS)

    Hara, Hideki

    1992-01-01

    Since the machinery, equipment and piping which compose nuclear power station facilities are diverse, when those are designed, consideration is given sufficiently to the objective of use and the importance of the object machinery and equipment so that those can maintain the soundness over the design life. In this report, on the contents and the design standard in the design techniques for nuclear reactor machinery and equipment, the way of thinking is shown, taking an example of reactor pressure vessel which is stipulated as the vessel kind 1 in the 'Technical standard of structures and others regarding nuclear facilities for electric power generation', Notice No. 501 of the Ministry of International Trade and Industry. The reactor pressure vessel of 1350 MWe improved type BWR (ABWR) is used under the condition of 87.9 kg/cm 2 and 302 degC, and the inside diameter is about 7.2 m, the inside height is about 21 m, and the wall thickness is about 170 mm. The design standard for reactor pressure vessels and its way of thinking, breakdown prevention design and the design techniques for reactor pressure vessels are described. (K.I.)

  18. Damage by radiation in structural materials of BWR reactor vessels

    International Nuclear Information System (INIS)

    Robles, E.; Balcazar, M.; Alpizar, A.M.; Calderon, B.E.

    2002-01-01

    The structural materials which are manufactured the pressure vessels of the BWR reactors undergo degradation in their mechanical properties mainly due to the damage produced by the fast neutrons (E> 1 MeV) coming from the reactor core. The mechanisms of neutron damage in this type of materials are experimentally studied, through the irradiation of vessel steel in experimental reactors for a quickly ageing. Alternately the neutron damage through steel irradiation with heavy ions is simulated. In this work the first results of the damage induced by irradiation of a similar steel to the vessel of a BWR reactor are shown. The irradiation was performed with fast neutrons (E> 1 MeV, fluence of 1.45 x 10 18 n/cm 2 ) in the TRIGA Mark III Salazar reactor and separately with Ni +3 ions in a Tandetrom accelerator (E= 4.8 MeV and an ion flux rank of 0.1 to 53 ions/A 2 ). (Author)

  19. Russian Federation: Passive Safety Components for Lead-Cooled Reactor Facilities

    International Nuclear Information System (INIS)

    Sarkulov, M.K.

    2015-01-01

    There is a specific range of engineered features used traditionally in nuclear technology. As a rule, main reactivity control systems use conventional active actuators with solid-body control members and/or liquid systems with active injection of liquid absorber. Other operation principles are normally chosen for additional systems. Currently, the traditional approach to improving the reliability of a reactor facility suggests an increase in the number of safety components and systems which provide for mutual assurance or assist each other. There is a great variety of additional reactivity control members designed for the reactor facility control and shutdown, including hydrodynamic members in the form of rods (acting from the coolant flow); floating-type members (absorbers and displacers); storage-type and liquid members (used in separate channels); bulk members (pebble absorber); gas-based members (with a gas absorber); shape-memory members and others. Hydrodynamic systems were introduced at Beloyarsk NPP Units 1 and 2 and proposed for use in other facility designs, Gases and bulk materials have not been commonly accepted: the former because of the high cost of high-efficiency gaseous absorbers, and the latter because of the complecated monitoring of the bulk material position. It is rather difficult and not always necessary to use the same engineering approaches in new lead-cooled reactor facilities as in traditional ones. Similarly to the development of traditional safety systems, passive safety components (devices) shall be designed according to the essential requirements of the nuclear regulations of the Russian Federation

  20. Earthquake analysis with nonlinear soil-structure interaction and nonlinear supports of components

    International Nuclear Information System (INIS)

    Hansson, V.

    1990-01-01

    For the determination of the seismic response of a structure the soil-structure interaction in most cases is modelled by a mass-spring-damper-system. Normally design concepts for components and piping are based on linear calculations and stress limitations. A concept for a reactor building for the HTR 100 consisted of a relatively high structure compared with the dimensions of the foundation. The structure was comparatively deep embedded in the soil, so here the embedment influences significantly the soil-structure interaction. The assembly of reactor vessel, heat exchanger and circulators has a height of about 37 m. Supports are arranged at different levels. Due to temperature deformations of the vessel and of the support constructions small gaps at the supports may only be avoided by complicated constructions of the supports. Nonlinear analyses were performed for soil, building and component with all supports. The finite element analyses used time histories. In order to describe the radiation damping the hysteresis of the soil with 1 percent material damping was considered. Nonlinearities in the interface of soil and foundation and due to gaps and friction at the supports were taken into account. The stiffness of the support constructions influences reactions and accelerations to a high extent. Properly chosen stiffnesses of the support constructions lead to a behaviour similar to linear elastic behaviour. 13 figs

  1. Estimate of the induced activity in the fixed structure of the Pluto Reactor

    International Nuclear Information System (INIS)

    Goodill, D.R.; Moore, D.C.; Tymons, B.J.

    1984-11-01

    This report presents an inventory of neutron-induced activity in the main components of a Materials Testing Reactor at the end of reactor life. The calculations were carried out for the PLUTO reactor at Harwell which is taken to be typical of all MTRs. The results were derived by using the FISPIN computer code, taking into account the geometry and construction of the reactor components. (author)

  2. Management of research reactor; dynamic characteristics analysis for reactor structures related with vibration of HANARO fuel assembly

    Energy Technology Data Exchange (ETDEWEB)

    Ahn, Chang Kee; Shim, Joo Sup [Shinwa Technology Information, Seoul (Korea)

    2001-04-01

    The objective of this study is to deduce the dynamic correlation between the fuel assembly and the reactor structure. Dynamic characteristics analyses for reactor structure related with vibration of HANARO fuel assembly have been performed For the dynamic characteristic analysis, the in-air models of the round and hexagonal flow tubes, 18-element and 36-element fuel assemblies, and reactor structure were developed. By calculating the hydrodynamic mass and distributing it on the in-air models, the in-water models of the flow tubes, the fuel assemblies, and the reactor structure were developed. Then, modal analyses for developed in-air and in-water models have been performed. Especially, two 18-element fuel assemblies and three 36-element fuel assemblies were included in the in-water reactor models. For the verification of the modal analysis results, the natural frequencies and the mode shapes of the fuel assembly were compared with those obtained from the experiment. Finally the analysis results of the reactor structure were compared with them performed by AECL Based on the reactor model without PCS piping, the in-water reactor model including the fuel assemblies was developed, and its modal analysis was performed. The analysis results demonstrate that there are no resonance between the fuel assembly and the reactor structures. 26 refs., 419 figs., 85 tabs. (Author)

  3. Method for producing components with internal architectures, such as micro-channel reactors, via diffusion bonding sheets

    Science.gov (United States)

    Alman, David E [Corvallis, OR; Wilson, Rick D [Corvallis, OR; Davis, Daniel L [Albany, OR

    2011-03-08

    This invention relates to a method for producing components with internal architectures, and more particularly, this invention relates to a method for producing structures with microchannels via the use of diffusion bonding of stacked laminates. Specifically, the method involves weakly bonding a stack of laminates forming internal voids and channels with a first generally low uniaxial pressure and first temperature such that bonding at least between the asperites of opposing laminates occurs and pores are isolated in interfacial contact areas, followed by a second generally higher isostatic pressure and second temperature for final bonding. The method thereby allows fabrication of micro-channel devices such as heat exchangers, recuperators, heat-pumps, chemical separators, chemical reactors, fuel processing units, and combustors without limitation on the fin aspect ratio.

  4. Helium effect on mechanical property of fusion reactor structural materials

    International Nuclear Information System (INIS)

    Yamamoto, Norikazu; Chuto, Toshinori; Murase, Yoshiharu; Nakagawa, Johsei

    2004-01-01

    High-energy neutrons produced in fusion reactor core caused helium in the structural materials of fusion reactors, such as blankets. We injected alpha particles accelerated by the cyclotron to the samples of martensite steel (9Cr3WVTaB). Equivalent helium doses injected to the sample is estimated to be up to 300 ppm, which were estimated to be equivalent to helium accumulation after the 1-year reactor operation. Creep tests of the samples were made to investigate helium embrittlement. There were no appreciable changes in the relation between the stresses and the rupture time, the minimum creep rate and the applied stress. Grain boundary effect by helium was not observed in ruptured surfaces. Fatigue tests were made for SUS304 samples, which contain helium up to 150 ppm. After 0.05 Hz cyclic stress tests, it was shown that the fatigue lifetime (cycles to rupture and extension to failure) are 1/5 in 150 ppm helium samples compared with no helium samples. The experimental results suggest martensite steel is promising for structural materials of fusion reactors. (Y. Tanaka)

  5. Design of reactor components (non replaceable) of 500 MWe PHWR for enhanced life

    International Nuclear Information System (INIS)

    Dwivedi, K.P.; Seth, V.K.

    1994-01-01

    A nuclear power station is characterised by large initial cost and low operating cost. So a plant which is capable of operating for a longer period of time will be economically more attractive. In the past approach had been to design a nuclear power plant for 30 to 40 years of life time. However, with the improvement in technology and incorporation of redundant and diverse safety features it is now possible to design a nuclear power plant for longer life. Now internationally it is being realised that without sacrificing safety features, plant life should be extended till the cost of maintenance or refurbishment is larger than the cost of the replacement capacity. In order to meet the objective of long life, for the components which cannot be easily replaced the life time of about 100 years is being considered as the design objective. For other items replacement, layout space, shielding, access route and lifting capacity and component design are receiving additional emphasis so as to provide a long total station life time. With the above background, design improvements to enhance the life of reactor components for 500 MWe PHWR namely calandria, end shields and calandria vault liners which cannot be replaced and on which any repair is extremely difficult, have been made. This paper deals with design life of these components and the modifications incorporated in the design. (author). 3 refs., 2 tabs., 3 figs

  6. Structural materials for Gen-IV nuclear reactors: Challenges and opportunities

    Science.gov (United States)

    Murty, K. L.; Charit, I.

    2008-12-01

    Generation-IV reactor design concepts envisioned thus far cater toward a common goal of providing safer, longer lasting, proliferation-resistant and economically viable nuclear power plants. The foremost consideration in the successful development and deployment of Gen-IV reactor systems is the performance and reliability issues involving structural materials for both in-core and out-of-core applications. The structural materials need to endure much higher temperatures, higher neutron doses and extremely corrosive environment, which are beyond the experience of the current nuclear power plants. Materials under active consideration for use in different reactor components include various ferritic/martensitic steels, austenitic stainless steels, nickel-base superalloys, ceramics, composites, etc. This paper presents a summary of various Gen-IV reactor concepts, with emphasis on the structural materials issues depending on the specific application areas. This paper also discusses the challenges involved in using the existing materials under both service and off-normal conditions. Tasks become increasingly complex due to the operation of various fundamental phenomena like radiation-induced segregation, radiation-enhanced diffusion, precipitation, interactions between impurity elements and radiation-produced defects, swelling, helium generation and so forth. Further, high temperature capability (e.g. creep properties) of these materials is a critical, performance-limiting factor. It is demonstrated that novel alloy and microstructural design approaches coupled with new materials processing and fabrication techniques may mitigate the challenges, and the optimum system performance may be achieved under much demanding conditions.

  7. Japanese contributions to containment structure, assembly and maintenance and reactor building for ITER

    International Nuclear Information System (INIS)

    Shibanuma, Kiyoshi; Honda, Tsutomu; Kanamori, Naokazu

    1991-06-01

    Joint design work on Conceptual Design Activity of International Thermonuclear Experimental Reactor (ITER) with four parties, Japan, the United States, the Soviet Union and the European Community began in April 1988 and was successfully completed in December 1990. In Japan, the home team was established in wide range of collaboration between JAERI and national institute, universities and heavy industries. The Fusion Experimental Reactor (FER) Team at JAERI is assigned as a core of the Japanese home team to support the joint Team activity and mainly conducted the design and R and D in the area of containment structure, remote handling and plant system. This report mainly describes the Japanese contribution on the ITER containment structure, remote handling and reactor building design. Main areas of contributions are vacuum vessel, attaching locks, electromagnetic analysis, cryostat, port and service line layout for containment structure, in-vessel handling equipment design and analysis, blanket handling equipment design and related short term R and D for assembly and maintenance, and finally reactor building design and analysis based on the equipment and service line layout and components flow during assembly and maintenance. (author)

  8. Transmutation and activation of fusion reactor wall and structural materials

    International Nuclear Information System (INIS)

    Jarvis, O.N.

    1979-01-01

    This report details the extent of the nuclear data needed for inclusion in a data library to be used for general assessments of fusion reactor structure activation and transmutation, describes the sources of data available, reviews the literature and explores the reliability of current calculations by providing an independent assessment of the activity inventory to be expected from five structural materials in a simple blanket design for comparison with the results of other workers. An indication of the nuclear reactions which make important contributions to the activity, transmutation and gas production rates for these structural materials is also presented. (author)

  9. Full scale seismic simulation of a nuclear reactor with parallel finite element analysis code for assembled structure

    International Nuclear Information System (INIS)

    Yamada, Tomonori

    2010-01-01

    The safety requirement of nuclear power plant attracts much attention nowadays. With the growing computing power, numerical simulation is one of key technologies to meet this safety requirement. Center for Computational Science and e-Systems of Japan Atomic Energy Agency has been developing a finite element analysis code for assembled structure to accurately evaluate the structural integrity of nuclear power plant in its entirety under seismic events. Because nuclear power plant is very huge assembled structure with tens of millions of mechanical components, the finite element model of each component is assembled into one structure and non-conforming meshes of mechanical components are bonded together inside the code. The main technique to bond these mechanical components is triple sparse matrix multiplication with multiple point constrains and global stiffness matrix. In our code, this procedure is conducted in a component by component manner, so that the working memory size and computing time for this multiplication are available on the current computing environment. As an illustrative example, seismic simulation of a real nuclear reactor of High Temperature engineering Test Reactor, which is located at the O-arai research and development center of JAEA, with 80 major mechanical components was conducted. Consequently, our code successfully simulated detailed elasto-plastic deformation of nuclear reactor and its computational performance was investigated. (author)

  10. Dynamic interactions of components, structure, and foundation of nuclear power facilities

    International Nuclear Information System (INIS)

    Pajuhesh, J.; Hadjian, A.H.

    1977-01-01

    A solution is formulated for the dynamic analysis of structures and components with different stiffness and damping characteristics, including the consideration of soil-structure interaction effects. Composite structures are often analysed approximately, in particular with regards to damping. For example, the reactor and other equipment in nuclear power plant structures are often analysed by assuming them uncoupled from the supporting structures. To achieve a better accuracy, the coupled system is hereby analysed as a composite component-structure-soil system. Although derivation of mass and stiffness matrices for the component-structure-soil system is a simple problem, the determination of the damping characteristics of such a system is more complex. This emphasis on the proper evaluation of system damping is warranted on the grounds that, when resonance conditions occur, the response amplitude is governed to a significant degree by the system damping. The damping information is usually available for each sub-structure separately with its based fixed or devoid of rigid-body modes of motion. The rigid-body motions are often free of damping resistance but sometimes, such as in the case of soil-structure interaction, or in the case of aerodynamic resistance, are uniquely defined. The composite damping matrix for the complete structure is hereby derived from the above-mentioned information. Thus, the damping matrix is first obtained for the free-free model of each sub-structure (the model containing the structural degrees of freedom together with rigid-body modes of motion), and then the submatrices for the free-free models are assembled to form the composite damping matrix in acccordance with an assembly technique relating the sub-structure coordinates to the global coordinates of the composite structure

  11. Engineering solutions for components facing the plasma in experimental power reactors

    International Nuclear Information System (INIS)

    Casini, G.; Farfaletti-Casali, F.

    1985-01-01

    A review of the engineering problems related to the structures in front of the plasma of experimental Tokamak-type reactors is made. Attention is focused on the so-named ''first wall'', i.e. the wall side of the blanket segments facing the plasma, and on the collector plates of the impurity control system, in particular for the case of the single-null poloidal divertor. Even if the uncertainties related to the plasma-wall interaction are stil relevant, some engineering solutions which look manageable are identified and described. (orig.)

  12. Non-structural Components influencing Hospital Disaster Preparedness in Malaysia

    Science.gov (United States)

    Samsuddin, N. M.; Takim, R.; Nawawi, A. H.; Rosman, M. R.; SyedAlwee, S. N. A.

    2018-04-01

    Hospital disaster preparedness refers to measures taken by the hospital’s stakeholders to prepare, reduce the effects of disaster and ensure effective coordination during incident response. Among the measures, non-structural components (i.e., medical laboratory equipment & supplies; architectural; critical lifeline; external; updated building document; and equipment & furnishing) are critical towards hospital disaster preparedness. Nevertheless, over the past few years these components are badly affected due to various types of disasters. Hence, the objective of this paper is to investigate the non-structural components influencing hospital’s disaster preparedness. Cross-sectional survey was conducted among thirty-one (31) Malaysian hospital’s employees. A total of 6 main constructs with 107 non-structural components were analysed and ranked by using SPSS and Relative Importance Index (RII). The results revealed that 6 main constructs (i.e. medical laboratory equipment & supplies; architectural; critical lifeline; external; updated building document; and equipment & furnishing) are rated as ‘very critical’ by the respondents. Among others, availability of medical laboratory equipment and supplies for diagnostic and equipment was ranked first. The results could serve as indicators for the public hospitals to improve its disaster preparedness in terms of planning, organising, knowledge training, equipment, exercising, evaluating and corrective actions through non-structural components.

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

    International Nuclear Information System (INIS)

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

    2016-01-01

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

  14. Study on reactor building structure using ultrahigh strength materials, 1

    International Nuclear Information System (INIS)

    Ishimura, Kikuo; Odajima, Masahiro; Irino, Kazuo; Hashiba, Toshio.

    1991-01-01

    This study was promoted to be aimed at realization of the optimal nuclear reactor building structure of the future. As the first step, the study regarding ultrahigh strength reinforced concrete (abbr. RC) shear wall was selected. As the result of various tests, the application of ultrahigh strength RC shear walls was verified. The tests conducted were relevant to; ultrahigh strength concrete material tests; pure shear tests of RC flat panels; and bending shear tests and its simulation analysis of RC shear walls. (author)

  15. Shock absorber in combination with a nuclear reactor core structure

    International Nuclear Information System (INIS)

    Housman, J.J.

    1976-01-01

    This invention relates to the provision of shock absorbers for use in blind control rod passages of a nuclear reactor core structure which are not subject to degradation. The shock absorber elements are made of a porous brittle carbonaceous material, a porous brittle ceramic material, or a porous brittle refractory oxide and have a void volume of between 30% and 70% of the total volume of the element for energy absorption by fracturing due to impact loading by a control rod. (UK)

  16. The problem of helium in structural materials for fusion reactor

    International Nuclear Information System (INIS)

    Nikiforov, A.S.; Zakharov, A.P.; Chuev, V.I.

    1982-01-01

    The processes of helium buildup in some metals and alloys at different energy neutron flux irradiation under thermonuclear reactor conditions are considered. The data on high temperature helium embrittlement of a number of stainless steels, titanium and aluminium alloys etc. are given A review of experiments concerning the implanted helium behaviour is presented. Possible reactions between helium atoms and point defects or their clusters are discussed. Analysed are material structure variations upon buildup in them up to 1 at % of helium

  17. Nuclear reactors. Introduction

    International Nuclear Information System (INIS)

    Boiron, P.

    1997-01-01

    This paper is an introduction to the 'nuclear reactors' volume of the Engineers Techniques collection. It gives a general presentation of the different articles of the volume which deal with: the physical basis (neutron physics and ionizing radiations-matter interactions, neutron moderation and diffusion), the basic concepts and functioning of nuclear reactors (possible fuel-moderator-coolant-structure combinations, research and materials testing reactors, reactors theory and neutron characteristics, neutron calculations for reactor cores, thermo-hydraulics, fluid-structure interactions and thermomechanical behaviour of fuels in PWRs and fast breeder reactors, thermal and mechanical effects on reactors structure), the industrial reactors (light water, pressurized water, boiling water, graphite moderated, fast breeder, high temperature and heavy water reactors), and the technology of PWRs (conceiving and building rules, nuclear parks and safety, reactor components and site selection). (J.S.)

  18. Residual radioactivity guidelines for the heavy water components test reactor at the Savannah River Site

    International Nuclear Information System (INIS)

    Owen, M.B. Smith, R.; McNeil, J.

    1997-04-01

    Guidelines were developed for acceptable levels of residual radioactivity in the Heavy Water Components Test Reactor (HWCTR) facility at the conclusion of its decommissioning. Using source terms developed from data generated in a detailed characterization study, the RESRAD and RASRAD-BUILD computer codes were used to calculate derived concentration guideline levels (DCGLs) for the radionuclides that will remain in the facility. The calculated DCGLs, when compared to existing concentrations of radionuclides measured during a 1996 characterization program, indicate that no decontamination of concrete surfaces will be necessary. Also, based on the results of the calculations, activated concrete in the reactor biological shield does not have to be removed, and imbedded radioactive piping in the facility can remain in place. Viewed in another way, the results of the calculations showed that the present inventory of residual radioactivity in the facility (not including that associated with the reactor vessel and steam generators) would produce less than one millirem per year above background to a hypothetical individual on the property. The residual radioactivity is estimated to be approximately 0.04 percent of the total inventory in the facility as of March, 1997. According to the results, the only radionuclides that would produce greater than 0.0.1-millirem per year are Am-241 (0.013 mrem/yr at 300 years), C-14 (0.022 mrem/yr at 1000 years) and U-238 (0.034 mrem/yr at 6000 years). Human exposure would occur only through the groundwater pathways, that is, from water drawn from, a well on the property. The maximum exposure would be approximately one percent of the 4 millirem per year ground water exposure limit established by the U.S. Environmental Protection Agency. 11 refs., 13 figs., 15 tabs

  19. Application of probabilistic fracture mechanics to the reliability analysis of pressure-bearing reactor components

    International Nuclear Information System (INIS)

    Schmitt, W.; Roehrich, E.; Wellein, R.

    1977-01-01

    Since no failures in the primary reactor components have been reported so far, it is impossible to estimate the failure probability of those components just by means of statistics. Therefore the way of probabilistic fracture mechanics has been proposed. Here the material properties, the loads and the crack distributions are treated as statistical variables with certain distributions. From the distributions of these data probability density functions can be established for the loading of a component as well as for the resistance of this component. From these functions the failure probability for a given failure mode is easily obtained either by the application of direct integration procedures which are shortly reviewed here, or by the use of Monte Carlo techniques. The most important part of the concept is the collection of a sufficiently large amount of raw data from different sources. These data need to be processed so that they can be transformed into probability density functions. The method of data collection and processing in terms of histograms, plots of probability density functions etc. is described. The choice of the various types of distribution functions is discussed. As an example, the derivation of the probability density function for cracks of a given size in a component is presented. Here the raw data, i.e. the ultrasonic results, are transformed into real crack sizes by means of a conservative conversion rule. The true distribution of the indications is obtained by taking into account a detection probability function. The final probability density function is influenced by the fact that indications exceeding certain values need to be re

  20. The effect of vertical earthquake component on the uplift of the nuclear reactor building

    International Nuclear Information System (INIS)

    Kobayashi, Toshio

    1986-01-01

    During a strong earthquake, the base mat of a nuclear reactor building may be lifted partially by the response overturning moment. And it causes geometrical nonlinear interaction between the base mat and rock foundation beneath it. In order to avoid this uplift phenomena, the base mat and/or plan of the building is enlarged in some cases. These special design need more cost and/or time in construction. In the evaluation of the uplift phenomena, a parameter ''η'' named ''contact ratio'' is used defined as the ratio of compression stress zone area of base mat for total area of base mat. Usually this contact ratio is calculated under the combination of the maximum overturning moment obtained by the linear earthquake response analysis and the normal force by the gravity considering the effect of the vertical earthquake component. In this report, the effect of vertical earthquake component for the uplift phenomena is studied and it concludes that the vertical earthquake component gives little influence on the contact ratio. In order to obtain more reasonable contact retio, the nonlinear rocking analysis subjected to horizontal and vertical earthquake motions simultaneously is proposed in this report. As the second best method, the combination of the maximum overturning moment obtained by linear analysis and the normal force by only the gravity without the vertical earthquake effect is proposed. (author)

  1. Regulatory instrument review: Management of aging of LWR [light water reactor] major safety-related components

    International Nuclear Information System (INIS)

    Werry, E.V.

    1990-10-01

    This report comprises Volume 1 of a review of US nuclear plant regulatory instruments to determine the amount and kind of information they contain on managing the aging of safety-related components in US nuclear power plants. The review was conducted for the US Nuclear Regulatory Commission (NRC) by the Pacific Northwest Laboratory (PNL) under the NRC Nuclear Plant Aging Research (NPAR) Program. Eight selected regulatory instruments, e.g., NRC Regulatory Guides and the Code of Federal Regulations, were reviewed for safety-related information on five selected components: reactor pressure vessels, steam generators, primary piping, pressurizers, and emergency diesel generators. Volume 2 will be concluded in FY 1991 and will also cover selected major safety-related components, e.g., pumps, valves and cables. The focus of the review was on 26 NPAR-defined safety-related aging issues, including examination, inspection, and maintenance and repair; excessive/harsh testing; and irradiation embrittlement. The major conclusion of the review is that safety-related regulatory instruments do provide implicit guidance for aging management, but include little explicit guidance. The major recommendation is that the instruments be revised or augmented to explicitly address the management of aging

  2. Manufacturing and testing in reactor relevant conditions of brazed plasma facing components of the ITER divertor

    International Nuclear Information System (INIS)

    Bisio, M.; Branca, V.; Marco, M. Di; Federici, A.; Grattarola, M.; Gualco, G.; Guarnone, P.; Luconi, U.; Merola, M.; Ozzano, C.; Pasquale, G.; Poggi, P.; Rizzo, S.; Varone, F.

    2005-01-01

    A fabrication route based on brazing technology has been developed for the realization of the high heat flux components for the ITER vertical target and Dome-Liner. The divertor vertical target is armoured with carbon fiber reinforced carbon and tungsten in the lower straight part and in the upper curved part, respectively. The armour material is joined to heat sinks made of precipitation hardened copper-chromium-zirconium alloy. The plasma facing units of the dome component are based on a tungsten flat tile design with hypervapotron cooling. An innovative brazing technique based on the addition of carbon fibers to the active brazing alloy, developed by Ansaldo Ricerche for applications in the field of the energy production, has been used for the carbon fiber composite to copper joint to reduce residual stresses. The tungsten-copper joint has been realized by direct casting. A proper brazing thermal cycle has been studied to guarantee the required mechanical properties of the precipitation hardened alloy after brazing. The fabrication route of plasma facing components for the ITER vertical target and dome based on the brazing technology has been proved by means of thermal fatigue tests performed on mock-ups in reactor relevant conditions

  3. Component evaluation for intersystem loss-of-coolant accidents in advanced light water reactors

    International Nuclear Information System (INIS)

    Ware, A.G.

    1994-07-01

    Using the methodology outlined in NUREG/CR-5603 this report evaluates (on a probabilistic basis) design rules for components in ALWRs that could be subjected to intersystem loss-of-coolant accidents (ISLOCAs). The methodology is intended for piping elements, flange connections, on-line pumps and valves, and heat exchangers. The NRC has directed that the design rules be evaluated for BWR pressures of 7.04 MPa (1025 psig), PWR pressures of 15.4 MPa (2235 psig), and 177 degrees C (350 degrees F), and has established a goal of 90% probability that system rupture will not occur during an ISLOCA event. The results of the calculations in this report show that components designed for a pressure of 0.4 of the reactor coolant system operating pressure will satisfy the NRC survival goal in most cases. Specific recommendations for component strengths for BWR and PWR applications are made in the report. A peer review panel of nationally recognized experts was selected to review and critique the initial results of this program

  4. Argonne Liquid-Metal Advanced Burner Reactor : components and in-vessel system thermal-hydraulic research and testing experience - pathway forward.

    Energy Technology Data Exchange (ETDEWEB)

    Kasza, K.; Grandy, C.; Chang, Y.; Khalil, H.; Nuclear Engineering Division

    2007-06-30

    from normal high-flow operation to natural circulation. Low-flow coolant events are the most difficult to design for because they involve the most complex thermal-hydraulic behavior induced by the dominance of thermal-buoyancy forces acting on the coolants. Such behavior can cause multiple-component flow interaction phenomena, which are not adequately understood or appreciated by reactor designers as to their impact on reactor performance and safety. Since the early 1990s, when DOE canceled the U.S. Liquid Metal Fast Breeder Reactor (LMFBR) program, little has been done experimentally to further understand the importance of the complex thermal-buoyancy phenomena and their impact on reactor design or to improve the ability of three-dimensional (3-D) transient computational fluid dynamics (CFD) and structures codes to model the phenomena. An improved experimental data base and the associated improved validated codes would provide needed design tools to the reactor community. The improved codes would also facilitate scale-up from small-scale testing to prototype size and would facilitate comparing performance of one reactor/component design with another. The codes would also have relevance to the design and safety of water-cooled reactors. To accomplish the preceding, it is proposed to establish a national GNEP-LMR research and development center at Argonne having as its foundation state-of-art science-based infrastructure consisting of: (a) thermal-hydraulic experimental capabilities for conducting both water and sodium testing of individual reactor components and complete reactor in-vessel models and (b) a computational modeling development and validation capability that is strongly interfaced with the experimental facilities. The proposed center would greatly advance capabilities for reactor development by establishing the validity of high-fidelity (i.e., close to first principles) models and tools. Such tools could be used directly for reactor design or for qualifying

  5. A procedure for evaluating residual life of major components in light water reactors

    International Nuclear Information System (INIS)

    Uchida, S.; Fujimori, H.; Ibe, E.; Kuniya, J.; Hayashi, M.; Fuse, M.; Yamauchi, K.

    1995-01-01

    A computer program for evaluating residual life of major components in boiling water reactors is proposed. It divides the stress corrosion cracking process into two stages; a probabilistic crack generation stage and a deterministic crack propagation one. The minimum period of the crack generation stage is evaluated assuming an exponential distribution of the stage. The crack propagation rate is calculated by the slip-dissolution/film-rupture model. The neutron flux and fluence dependence of the neutron radiation effects on material properties was evaluated by using theoretical models of radiation damage. The computer program works on an engineering work station. Evaluated results are displayed as a map of the residual life, or as graphs of crack length evolution

  6. Integrated network for structural integrity monitoring of critical components in nuclear facilities, RIMIS

    International Nuclear Information System (INIS)

    Roth, Maria; Constantinescu, Dan Mihai; Brad, Sebastian; Ducu, Catalin; Malinovschi, Viorel

    2008-01-01

    The round table aims to join specialists working in the research area of the Romanian R and D Institutes and Universities involved in structural integrity assessment of materials, especially those working in the nuclear field, together with the representatives of the end user, the Cernavoda NPP. This scientific event will offer the opportunity to disseminate the theoretical, experimental and modelling activities, carried out to date, in the framework of the National Program 'Research of Excellence', Module I 2006-2008, managed by the National Authority for Scientific Research. Entitled 'Integrated Network for Structural Integrity Monitoring of Critical Components in Nuclear Facilities, RIMIS, the project has two main objectives: 1. - to elaborate a procedure applicable to the structural integrity assessment of critical components used in Romanian nuclear facilities (CANDU type Reactor, Hydrogen Isotopes Separation installations); 2. - to integrate the national networking into a similar one of European level, and to enhance the scientific significance of Romanian R and D organisations as well as to increase the contribution in solving major issues of the nuclear field. The topics of the round table will be focused on: 1. Development of a Structural Integrity Assessment Methodology applicable to the nuclear facilities components; 2. Experimental investigation methods and procedures; 3. Numeric simulation of nuclear components behaviour; 4. Further activities to finalize the assessment procedure. Also participations and contributions to sustain the activity in the European Network NULIFE, FP6 will be discussed. (authors)

  7. Experimental and theoretical investigations of soil-structure interaction effect at HDR-reactor-building

    International Nuclear Information System (INIS)

    Wassermann, K.

    1983-01-01

    Full-scale dynamic testing on intermediate and high levels was performed at the Heissdampfreaktor (HDR) in 1979. Various types of dynamic forces were applied and response of the reactor containment structure and internal components was measured. Precalculations of dynamic behaviour and response of the structure were made through different mathematical models for the structure and the underlying soil. Soil-Structure Interaction effects are investigated and different theoretical models are compared with experimental results. In each model, the soil is represented by springs attached to the structural model. Stiffnesses of springs are calculated by different finite-element idealizations and half-space approximations. Eigenfrequencies and damping values related to interaction effects (translation, rocking, torsion) are identified from test results. The comparisons of dynamic characteristic of the soil-structure system between precalculations and test results show good agreement in general. (orig.)

  8. Estimated long lived isotope activities in ET-RR-1 reactor structural materials for decommissioning study

    International Nuclear Information System (INIS)

    Ashoub, N.; Saleh, H.

    1995-01-01

    The first Egyptian research reactor, ET-RR-1 is tank type with light water as a moderator, coolant and reflector. Its nominal power is 2MWt and the average thermal neutron flux is 10 13 n/cm 2 sec -1 . Its criticality was on the fall of 1961. The reactor went through several modifications and updating and is still utilized for experimental research. A plan for decommissioning of ET-RR-1 reactor should include estimation of radioactivity in structural materials. The inventory will help in assessing the radiological consequences of decommissioning. This paper presents a conservative calculation to estimate the activity of the long lived isotopes which can be produced by neutron activation. The materials which are presented in significant quantities in the reactor structural materials are aluminum, cast iron, graphite, ordinary and iron shot concrete. The radioactivity of each component is dependent not only upon the major elements, but also on the concentration of the trace elements. The main radioactive inventory are expected to be from 60 Co and 55 Fe which are presented in aluminium as trace elements and in large quantities in other construction materials. (author)

  9. A fracture mechanics method of evaluating structural integrity of a reactor vessel due to thermal shock effects following LOCA condition

    International Nuclear Information System (INIS)

    Ramani, D.T.

    1977-01-01

    The importance of knowledge of structural integrity of a reactor vessel due to thermal shock effects, is related to safety and operational requirements in assessing the adequacy and flawless functioing of the nuclear power systems. Followig a loss-of-coolant accident (LOCA) condition the integrity of the reactor vessel due to a sudden thermal shock induced by actuation of emergency core cooling system (ECCS), must be maintained to ensure safe and orderly shutdown of the reactor and its components. The paper encompasses criteria underlaying a fracture mechanics method of analysis to evaluate structural integrity of a typical 950 MWe PWR vessel as a result of very drastic changes in thermal and mechanical stress levels in the reactor vessel wall. The main object of this investigation therefore consists in assessing the capability of a PWR vessel to withstand the most critical thermal shock without inpairing its ability to conserve vital coolant owing to probable crack propagation. (Auth.)

  10. Online stress corrosion crack and fatigue usages factor monitoring and prognostics in light water reactor components: Probabilistic modeling, system identification and data fusion based big data analytics approach

    Energy Technology Data Exchange (ETDEWEB)

    Mohanty, Subhasish M. [Argonne National Lab. (ANL), Argonne, IL (United States); Jagielo, Bryan J. [Argonne National Lab. (ANL), Argonne, IL (United States); Oakland Univ., Rochester, MI (United States); Iverson, William I. [Argonne National Lab. (ANL), Argonne, IL (United States); Univ. of Illinois at Urbana-Champaign, Champaign, IL (United States); Bhan, Chi Bum [Argonne National Lab. (ANL), Argonne, IL (United States); Pusan National Univ., Busan (Korea, Republic of); Soppet, William S. [Argonne National Lab. (ANL), Argonne, IL (United States); Majumdar, Saurin M. [Argonne National Lab. (ANL), Argonne, IL (United States); Natesan, Ken N. [Argonne National Lab. (ANL), Argonne, IL (United States)

    2014-12-10

    Nuclear reactors in the United States account for roughly 20% of the nation's total electric energy generation, and maintaining their safety in regards to key component structural integrity is critical not only for long term use of such plants but also for the safety of personnel and the public living around the plant. Early detection of damage signature such as of stress corrosion cracking, thermal-mechanical loading related material degradation in safety-critical components is a necessary requirement for long-term and safe operation of nuclear power plant systems.

  11. Survey of creep data on structural materials of fast breeder reactor

    International Nuclear Information System (INIS)

    Yoshida, S.

    1977-11-01

    The reactor vessels and other components of fast breeder reactor is affected by high neutron irradiation at elevated temperatures. However, in this regard, related test data on creep property of component materials and welds at elevated temperatures are a few in Japan, and especially, there are no data available on the irradiation effect. It will take 3 to 7 years before the results of currently planned research and development on prototype fast breeder become available. On the other hand, establishment of design base for prototype fast breeder and other needs call for early solution to such problems. The Committee should, therefore, collect from documents the latest data on experiments on structural materials overseas and in our country, and survey and analyze the problems in order to proceed with the future research and development in the most effective way. It was for this purpose that the Fourth Subcommittee at Technical Research Association for Integrity of Structures at Elevated Service Temperatures was commissioned by Power Reactor and Nuclear Fuel Development Corporation to conduct the examination and study of related data by establishing Group 41G. This collection of data is the compilation of the above results. (author)

  12. Design and Fabrication Technique of the Key Components for Very High Temperature Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Ho Jin; Song, Ki Nam; Kim, Yong Wan

    2006-12-15

    The gas outlet temperature of Very High Temperature Reactor (VHTR) may be beyond the capability of conventional metallic materials. The requirement of the gas outlet temperature of 950 .deg. C will result in operating temperatures for metallic core components that will approach very high temperature on some cases. The materials that are capable of withstanding this temperature should be prepared, or nonmetallic materials will be required for limited components. The Ni-base alloys such as Alloy 617, Hastelloy X, XR, Incoloy 800H, and Haynes 230 are being investigated to apply them on components operated in high temperature. Currently available national and international codes and procedures are needed reviewed to design the components for HTGR/VHTR. Seven codes and procedures, including five ASME Codes and Code cases, one French code (RCC-MR), and on British Procedure (R5) were reviewed. The scope of the code and code cases needs to be expanded to include the materials with allowable temperatures of 950 .deg. C and higher. The selection of compact heat exchangers technology depends on the operating conditions such as pressure, flow rates, temperature, but also on other parameters such as fouling, corrosion, compactness, weight, maintenance and reliability. Welding, brazing, and diffusion bonding are considered proper joining processes for the heat exchanger operating in the high temperature and high pressure conditions without leakage. Because VHTRs require high temperature operations, various controlled materials, thick vessels, dissimilar metal joints, and precise controls of microstructure in weldment, the more advanced joining processes are needed than PWRs. The improved solid joining techniques are considered for the IHX fabrication. The weldability for Alloy 617 and Haynes 230 using GTAW and SMAW processes was investigated by CEA.

  13. Technical Needs for Prototypic Prognostic Technique Demonstration for Advanced Small Modular Reactor Passive Components

    Energy Technology Data Exchange (ETDEWEB)

    Meyer, Ryan M.; Coble, Jamie B.; Hirt, Evelyn H.; Ramuhalli, Pradeep; Mitchell, Mark R.; Wootan, David W.; Berglin, Eric J.; Bond, Leonard J.; Henager, Charles H.

    2013-05-17

    This report identifies a number of requirements for prognostics health management of passive systems in AdvSMRs, documents technical gaps in establishing a prototypical prognostic methodology for this purpose, and describes a preliminary research plan for addressing these technical gaps. AdvSMRs span multiple concepts; therefore a technology- and design-neutral approach is taken, with the focus being on characteristics that are likely to be common to all or several AdvSMR concepts. An evaluation of available literature is used to identify proposed concepts for AdvSMRs along with likely operational characteristics. Available operating experience of advanced reactors is used in identifying passive components that may be subject to degradation, materials likely to be used for these components, and potential modes of degradation of these components. This information helps in assessing measurement needs for PHM systems, as well as defining functional requirements of PHM systems. An assessment of current state-of-the-art approaches to measurements, sensors and instrumentation, diagnostics and prognostics is also documented. This state-of-the-art evaluation, combined with the requirements, may be used to identify technical gaps and research needs in the development, evaluation, and deployment of PHM systems for AdvSMRs. A preliminary research plan to address high-priority research needs for the deployment of PHM systems to AdvSMRs is described, with the objective being the demonstration of prototypic prognostics technology for passive components in AdvSMRs. Greater efficiency in achieving this objective can be gained through judicious selection of materials and degradation modes that are relevant to proposed AdvSMR concepts, and for which significant knowledge already exists. These selections were made based on multiple constraints including the analysis performed in this document, ready access to laboratory-scale facilities for materials testing and measurement, and

  14. Design and Fabrication Technique of the Key Components for Very High Temperature Reactor

    International Nuclear Information System (INIS)

    Lee, Ho Jin; Song, Ki Nam; Kim, Yong Wan

    2006-12-01

    The gas outlet temperature of Very High Temperature Reactor (VHTR) may be beyond the capability of conventional metallic materials. The requirement of the gas outlet temperature of 950 .deg. C will result in operating temperatures for metallic core components that will approach very high temperature on some cases. The materials that are capable of withstanding this temperature should be prepared, or nonmetallic materials will be required for limited components. The Ni-base alloys such as Alloy 617, Hastelloy X, XR, Incoloy 800H, and Haynes 230 are being investigated to apply them on components operated in high temperature. Currently available national and international codes and procedures are needed reviewed to design the components for HTGR/VHTR. Seven codes and procedures, including five ASME Codes and Code cases, one French code (RCC-MR), and on British Procedure (R5) were reviewed. The scope of the code and code cases needs to be expanded to include the materials with allowable temperatures of 950 .deg. C and higher. The selection of compact heat exchangers technology depends on the operating conditions such as pressure, flow rates, temperature, but also on other parameters such as fouling, corrosion, compactness, weight, maintenance and reliability. Welding, brazing, and diffusion bonding are considered proper joining processes for the heat exchanger operating in the high temperature and high pressure conditions without leakage. Because VHTRs require high temperature operations, various controlled materials, thick vessels, dissimilar metal joints, and precise controls of microstructure in weldment, the more advanced joining processes are needed than PWRs. The improved solid joining techniques are considered for the IHX fabrication. The weldability for Alloy 617 and Haynes 230 using GTAW and SMAW processes was investigated by CEA

  15. Automatic capability to store and retrieve component data and to calculate structural integrity of these components

    International Nuclear Information System (INIS)

    McKinnis, C.J.; Toor, P.M.

    1985-01-01

    In structural analysis, assimilation of material, geometry, and service history input parameters is very cumbersome. Quite often with changing service history and revised material properties and geometry, an analysis has to be repeated. To overcome the above mentioned difficulties, a computer program was developed to provide the capability to establish a computerized library of all material, geometry, and service history parameters for components. The program also has the capability to calculate the structural integrity based on the Arrhenius type equations, including the probability calculations. This unique combination of computerized input information storage and automated analysis procedure assures consistency, efficiency, and accuracy when the hardware integrity has to be reassessed

  16. Vessel supporting structure for liquid metal cooled nuclear reactors

    International Nuclear Information System (INIS)

    Mahe, Armel; Jullien, Georges

    1974-01-01

    The supporting structure described is for a liquid metal cooled nuclear reactor, the vessel being of the type suspended to the end slab of the reactor. It includes a ring connected at one of its two ends to a single shell and at the other end to two shells. One of these three shells connected to the lower end of the ring forms the upper part of the vessel to be supported. The two other shells are embedded in two sperate parts of the slab. The ring and shell assembly is housed in an annular space provided in the end slab and separating it into two parts, namely a central part and a peripheral part [fr

  17. A susceptor heating structure in MOVPE reactor by induction heating

    International Nuclear Information System (INIS)

    Li, Zhiming; Li, Hailing; Zhang, Jincheng; Li, Jinping; Jiang, Haiying; Fu, Xiaoqian; Han, Yanbin; Xia, Yingjie; Huang, Yimei; Yin, Jianqin; Zhang, Lejuan; Hu, Shigang

    2014-01-01

    A novel susceptor with a revolutionary V-shaped slot of solid of revolution form is proposed in the metalorganic vapor phase epitaxy (MOVPE) reactor by induction heating. This slot changes the heat transfer rate as the generated heat is transferred from the high temperature region of the susceptor to the substrate, which improves the uniformity of the substrate temperature distribution. By using finite element method (FEM), the susceptor with this structure for heating the substrate of six inches in diameter is optimized. It is observed that this optimized susceptor with the V-shaped slot makes the uniformity of the substrate temperature distribution improve more than 80%, which can be beneficial to the film growth. - Highlights: •A novel susceptor with V-shaped slot in MOVPE reactor is proposed. •Temperature in the substrate is optimized. •Great temperature uniformity of the substrate is obtained

  18. Structural materials for ITER in-vessel component design

    Energy Technology Data Exchange (ETDEWEB)

    Kalinin, G. [Max-Planck-Inst. fur Plasmaphys., Garching (Germany). ITER Garching JWS; Gauster, W. [Max-Planck-Inst. fur Plasmaphys., Garching (Germany). ITER Garching JWS; Matera, R. [Max-Planck-Inst. fur Plasmaphys., Garching (Germany). ITER Garching JWS; Tavassoli, A.-A.F. [CEA Centre d`Etudes Nucleaires de Saclay, 91 - Gif-sur-Yvette (France); Rowcliffe, A. [Oak Ridge National Lab., TN (United States); Fabritsiev, S. [Research Inst. of Electrophysical Apparatus, St. Petersburg (Russian Federation); Kawamura, H. [JAERI, IMTR Project, Ibaraki (Japan). Blanket Irradiation Lab.

    1996-10-01

    The materials proposed for ITER in-vessel components have to exhibit adequate performance for the operating lifetime of the reactor or for specified replacement intervals. Estimates show that maximum irradiation dose to be up to 5-7 dpa (for 1 MWa/m{sup 2} in the basic performance phase (BPP)) within a temperature range from 20 to 300 C. Austenitic SS 316LN-ITER Grade was defined as a reference option for the vacuum vessel, blanket, primary wall, pipe lines and divertor body. Conventional technologies and mill products are proposed for blanket, back plate and manifold manufacturing. HIPing is proposed as a reference manufacturing method for the primary wall and blanket and as an option for the divertor body. The existing data show that mechanical properties of HIPed SS are no worse than those of forged 316LN SS. Irradiation will result in property changes. Minimum ductility has been observed after irradiation in an approximate temperature range between 250 and 350 C, for doses of 5-10 dpa. In spite of radiation-induced changes in tensile deformation behavior, the fracture remains ductile. Irradiation assisted corrosion cracking is a concern for high doses of irradiation and at high temperatures. Re-welding is one of the critical issues because of the need to replace failed components. It is also being considered for the replacement of shielding blanket modules by breeding modules after the BPP. (orig.).

  19. Characterization and management of radioactive sodium and other reactor components as input data for the decommissioning of liquid metal-cooled fast reactors. A compilation of data produced of data produced by members of the IAEA technical working group on fast reactors (TWG-FR) at two consultancies and one technical committee meeting. Working material

    International Nuclear Information System (INIS)

    2002-01-01

    A number of liquid metal cooled fast reactors (LMFRs) are in operation and, some have already been shut down; other reactors will reach the end of their design lifetime in a few years and become candidates for decommissioning. It is unfortunate that little consideration was devoted to decommissioning of reactors at the plant design and construction stage. It is with this focus that the Technical Working Group on Fast Reactors (TWGFR) recommended that the IAEA organize the exchange of information on LMFRs decommissioning technology. It was pointed out that the decommissioning of small sodium-cooled reactors has shown that there are two basic differences between thermal and fast reactors decommissioning: on the one side, the treatment and disposal of radioactive sodium coolant, and on the other side, the management of reactor components, for which the structural materials are activated in depth by fast neutrons. To this end, a Technical Committee Meeting on Sodium Removal and Disposal from LMFRs in Normal Operation and in the framework of Decommissioning (Aix-en-Provence, France, November 1997) and two Consultancies on Decommissioning of the Kazakh BN-350 LMFR (Vienna, Austria, October 1996; Obninsk, Russian Federation, February 1998) were convened by the IAEA. These Meetings brought together a group of experts from France, Russia, Kazakhstan, the UK, and the USA to exchange information on, and to review current technical knowledge and experience in the management of radioactive coolant and reactor components following closing of LMFRs, as well as their design features and operating experience relevant for decommissioning procedures. The report provides general and detailed information on activation characteristics of the primary coolant; treatment and disposal of the spent sodium; removal of the residual sodium deposits and decontamination; the activation characteristics of the reactor components and the management of the latter. The recurring theme is finding

  20. Simplified calculation of thermal stresses - on the reduction of effort in the stress analysis of reactor components

    International Nuclear Information System (INIS)

    Karow, K.

    1984-01-01

    The fatigue behaviour of reactor components is predominantly determined from the in-service thermal stresses. The calculation of such stresses for a number of temperature transients in the adjacent fluid may be expensive, particularly with complicated structures. Under certain conditions this expense can be reduced considerably with the aid of a rule, which permits interpolation of thermal stresses from known reference values instead of calculation. This paper presents the derivation and method of application of this interpolation rule. The derivation procedure is based on well-known proportionalities between thermal stress range Δsigma in the structure and temperature change ΔT and rate of change T of the fluid in the extreme cases of an ideal thermal shock and quasi-steady-state conditions, respectively. For the real transients in between the relationship Δsigma proportional (ΔT)sup(x) Tsup(1-x)αsup(y) is proposed, where x is the shock-degree and lies between 0 and 1, and, additionally, y designates the influence of the heat transfer coefficient α. This formula yields the interpolation rule. The rule permits interpolation of stress ranges for additional thermal transients from at least 3 reference stresses via x and y. The procedure is applicable to any metallic structure, reduces fatigue analysis effort considerably and yields excellent results. The paper is split up into 2 parts. In the following the derivation of the rule is presented. The second part describes its application and will be published shortly. (orig.)

  1. Structural analysis of the Upper Internals Structure for the Clinch River Breeder Reactor Plant

    International Nuclear Information System (INIS)

    Houtman, J.L.

    1979-01-01

    The Upper Internals Structure (UIS) of the Clinch River Breeder Reactor Plant (CRBRP) provides control of core outlet flow to prevent severe thermal transients from occuring at the reactor vessel and primary heat transport outlet piping, provides instrumentation to monitor core performance, provides support for the control rod drivelines, and provides secondary holddown of the core. All of the structural analysis aspects of assuring the UIS is structurally adequate are presented including simplified and rigorous inelastic analysis methods, elevated temperature criteria, environmental effects on material properties, design techniques, and manufacturing constraints

  2. Structural evaluation of safety class components to natural phenomena loadings

    International Nuclear Information System (INIS)

    Conrads, T.J.

    1989-01-01

    This paper addresses the efforts completed at the US Department of Energy Hanford Site near Richland, Washington, to qualify structurally a number of existing safety class components in the Plutonium Finishing Plant complex. Design, fabrication, and installation of the facility occurred in the 1950s and 1960s and were based on the Uniform Building Code criteria for wind and earthquake loads. Recently the buildings were qualified to site-specific wind and seismic hazards. The methodology employed to qualify seismically the safety class components is discussed

  3. INFLUENCE OF STRUCTURE COMPONENTS ON MACHINE TOOL ACCURACY

    Directory of Open Access Journals (Sweden)

    ConstantinSANDU

    2017-11-01

    Full Text Available For machine tools, the accuracy of the parts of the machine tool structure (after roughing should be subject to relief and natural or artificial aging. The performance of the current accuracy of machine tools as linearity or flatness was higher than 5 μm/m. Under this value there are great difficulties. The performance of the structure of the machine tools in the manufacture of structural parts of machine tools, with a flatness accuracy that the linearity of about 2 μm/m, are significant deviations form of their half-finished. This article deals with the influence of errors of form of semifinished and machined parts on them, on their shape and especially what happens to structure machine tools when the components of the structure were assembling this.

  4. Generic component reliability data for research reactors PSA. Final report of the CRP on data acquisition for research reactor PSA. Working material

    International Nuclear Information System (INIS)

    1993-01-01

    The scope of this document is to provide the final reference generic component reliability database information for a variety of research reactor types. As noted in Section 2.1 and Table 3a, many years of component data are represented in the database so that it is expected that the report should provide representative data valid for a number of years. The database provides component failure rates on a time and/or a demand related basis according to the operational modes of the components. At the current time an update of the database is not planned. As a result of the implementation of data collection systems in the research reactors represented in these studies, updating of data from individual facilities could be made available from the contributing research reactor facilities themselves. As noted in Section 1.2, the report does not include detailed discussion of information regarding component classification and reliability parameter definitions. The report does provide some insights and discussion regarding the practicalities of the data collection process and some guidelines for database usage. 9 refs, tabs

  5. Generic component reliability data for research reactors PSA. Final report of the CRP on data acquisition for research reactor PSA. Working material

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1994-12-31

    The scope of this document is to provide the final reference generic component reliability database information for a variety of research reactor types. As noted in Section 2.1 and Table 3a, many years of component data are represented in the database so that it is expected that the report should provide representative data valid for a number of years. The database provides component failure rates on a time and/or a demand related basis according to the operational modes of the components. At the current time an update of the database is not planned. As a result of the implementation of data collection systems in the research reactors represented in these studies, updating of data from individual facilities could be made available from the contributing research reactor facilities themselves. As noted in Section 1.2, the report does not include detailed discussion of information regarding component classification and reliability parameter definitions. The report does provide some insights and discussion regarding the practicalities of the data collection process and some guidelines for database usage. 9 refs, tabs.

  6. Maximum power gains of radio-frequency-driven two-energy-component tokamak reactors

    International Nuclear Information System (INIS)

    Jassby, D.L.

    1974-11-01

    Two-energy-component fusion reactors in which the suprathermal component (D) is produced by harmonic cyclotron ''runaway'' of resonant ions are considered. In one ideal case, the fast hydromagnetic wave at ω = 2ω/sub cD/ produces an energy distribution f(W) approximately constant (up to W/sub max/) that includes all deuterons, which then thermalize and react with the cold tritons. In another ideal case, f(W) approximately constant is maintained by the fast wave at ω = ω/sub cD/. If one neglects (1) direct rf input to the bulk-plasma electrons and tritons, and (2) the fact that many deuterons are not resonantly accelerated, then the maximum ideal power gain is about 0.85 Q/sub m/ in the first case and 1.05 Q/sub m/ in the second case, where Q/sub m/ is the maximum fusion gain in the beam-injection scheme (e.g., Q/sub m/ = 1.9 at T/sub e/ = 10 keV). Because of nonideal effects, the cyclotron runaway phenomenon may find its most practical use in the heating of 50:50 D--T plasmas to ignition. (auth)

  7. Cladding Effects on Structural Integrity of Nuclear Components

    International Nuclear Information System (INIS)

    Sattari-Far, Iradi; Andersson, Magnus

    2006-06-01

    Based on this study, the following conclusions and recommendations can be made: Due to significant differences in the thermal and mechanical properties between the austenitic cladding and the ferritic base metal, residual stresses are induced in the cladding and the underlying base metal. These stresses are left in clad components even after Post-Weld Heat Treatment (PWHT). The different restraint conditions of the clad component have a minor influence on the magnitude of the cladding residual stresses in the cladding layer. The thickness of the clad object is the main impacting geometrical dimension in developing cladding residual stresses. A clad object having a base material thickness exceeding 10 times the cladding thickness would be practically sufficient to introduce cladding residual stresses of a thick reactor pressure vessel. For a clad component that received PWHT, the peak tensile stress is in the cladding layer, and the residual stresses in the underlying base material are negligible. However, for clad components not receiving PWHT, for instance the repair welding of the cladding, the cladding residual stresses of tensile type exist even in the base material. This implies a higher risk for underclad cracking for clad repairs that received no PWHT. For certain clad geometries, like nozzles, the profile of the cladding residual stresses depends on the clad thickness and position, and significant tensile stresses can also exist in the base material. Based on different measurements reported in the literature, a value of 150 GPa can be used as Young's Modulus of the austenitic cladding material at room temperature. The control measurements of small samples from the irradiated reactor pressure vessel head did not reveal a significant difference of Young's Modulus between the irradiated and the unirradiated cladding material condition. No significant differences between the axial and tangential cladding residual stresses are reported in the measurement of

  8. Cladding Effects on Structural Integrity of Nuclear Components

    Energy Technology Data Exchange (ETDEWEB)

    Sattari-Far, Iradi; Andersson, Magnus [lnspecta Technology AB, Stockholm (Sweden)

    2006-06-15

    Based on this study, the following conclusions and recommendations can be made: Due to significant differences in the thermal and mechanical properties between the austenitic cladding and the ferritic base metal, residual stresses are induced in the cladding and the underlying base metal. These stresses are left in clad components even after Post-Weld Heat Treatment (PWHT). The different restraint conditions of the clad component have a minor influence on the magnitude of the cladding residual stresses in the cladding layer. The thickness of the clad object is the main impacting geometrical dimension in developing cladding residual stresses. A clad object having a base material thickness exceeding 10 times the cladding thickness would be practically sufficient to introduce cladding residual stresses of a thick reactor pressure vessel. For a clad component that received PWHT, the peak tensile stress is in the cladding layer, and the residual stresses in the underlying base material are negligible. However, for clad components not receiving PWHT, for instance the repair welding of the cladding, the cladding residual stresses of tensile type exist even in the base material. This implies a higher risk for underclad cracking for clad repairs that received no PWHT. For certain clad geometries, like nozzles, the profile of the cladding residual stresses depends on the clad thickness and position, and significant tensile stresses can also exist in the base material. Based on different measurements reported in the literature, a value of 150 GPa can be used as Young's Modulus of the austenitic cladding material at room temperature. The control measurements of small samples from the irradiated reactor pressure vessel head did not reveal a significant difference of Young's Modulus between the irradiated and the unirradiated cladding material condition. No significant differences between the axial and tangential cladding residual stresses are reported in the

  9. Prestressed concrete nuclear reactor containment structures. Revision 3

    International Nuclear Information System (INIS)

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

    1975-02-01

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

  10. Transactions of the 9th international conference on structural mechanics in reactor technology. Vol. M

    International Nuclear Information System (INIS)

    Wittmann, F.H.

    1987-01-01

    For obvious reasons reliability plays a dominant role in reactor technology. The area to be covered by Division M which deals with this subject, can be briefly summarized as follows: Probabilistic safety assessment (PSA) of structures and uncertainty modelling in structural design. Pre-service and in-service inspection with respect to evaluation of the probability of failure in time of structure. Stochastic loads modelling. External events (earthquakes, aircraft-impacts, etc.). Stochastic damage models of materials and structures. Probabilistic fracture mechanics. Model for ageing of components and structures. Reliability analysis of large and complex systems. Benchmark exercises. Analysis of operational experience. Precursor-studies. Man-machine interactions. Relationship between availability and PSA. Using probabilistic methods in setting up codes, standards and safety goals. Risk assessment of nuclear power plants and of nuclear fuel cycle installations. All 65 papers are separately indexed in the database. (orig./HP)

  11. Statistical properties of material strength for reliability evaluation of components of fast reactors. Austenitic stainless steels

    International Nuclear Information System (INIS)

    Takaya, Shigeru; Sasaki, Naoto; Tomobe, Masato

    2015-03-01

    Many efforts have been made to implement the System Based Code concept of which objective is to optimize margins dispersed in several codes and standards. Failure probability is expected to be a promising quantitative index for optimization of margins, and statistical information for random variables is needed to evaluate failure probability. Material strength like tensile strength is an important random variable, but the statistical information has not been provided enough yet. In this report, statistical properties of material strength such as creep rupture time, steady creep strain rate, yield stress, tensile stress, flow stress, fatigue life and cyclic stress-strain curve, were estimated for SUS304 and 316FR steel, which are typical structural materials for fast reactors. Other austenitic stainless steels like SUS316 were also used for statistical estimation of some material properties such as fatigue life. These materials are registered in the JSME code of design and construction of fast reactors, so test data used for developing the code were used as much as possible in this report. (author)

  12. Optimization of tungsten-steel joints for plasma facing components in fusion reactors

    Energy Technology Data Exchange (ETDEWEB)

    Heuer, Simon; Linsmeier, Christian [Forschungszentrum Juelich GmbH, Institut fuer Energie- und Klimaforschung - Plasmaphysik, Juelich (Germany); Weber, Thomas; Linke, Jochen [Forschungszentrum Juelich GmbH, Institut fuer Energie- und Klimaforschung - Werkstoffstruktur und -eigenschaften, Juelich (Germany); Matejicek, Jiri [Institute of Plasma Physics, Academy of Sciences of the Czech Republic, Prague (Czech Republic)

    2015-07-01

    Tungsten, joint to a martensitic-ferritic EUROFER97 structure, is a promising plasma facing material composite for fusion reactors. Due to the effect of mismatch in thermo-mechanical properties direct bonding is not feasible. Current research is therefore ongoing on interlayer systems. While the adhesion was already improved by the utilization of a discrete Cu, Ti or V interlayer, that is able to relax stresses by plastic deformation, joints still do not resist the expected load cycles in a fusion reactor. Therefore, alternatives for the interface are needed. This contribution presents research on functionally graded materials (FGM). The particular microstructure of a graded interlayer allows re-distributing macro stresses from a discrete interface to a greater volume while avoiding in particular Cu which tends to swell under neutron irradiation. A parameter study on the basis of finite element analysis will be presented as well as first results of several processing routes for FGM that shall be evaluated and benchmarked by mechanical as well as thermal testing.

  13. Relationships between chemical oxygen demand (COD) components and toxicity in a sequential anaerobic baffled reactor/aerobic completely stirred reactor system treating Kemicetine

    Energy Technology Data Exchange (ETDEWEB)

    Sponza, Delia Teresa, E-mail: delya.sponza@deu.edu.tr [Department of Environmental Engineering, Faculty of Engineering, Dokuz Eyluel University, Buca Kaynaklar Campus, Tinaztepe, 35160 Izmir (Turkey); Demirden, Pinar, E-mail: pinar.demirden@kozagold.com [Environmental Engineer, Koza Gold Company, Environmental Department, Ovacik, Bergama Izmir (Turkey)

    2010-04-15

    In this study the interactions between toxicity removals and Kemicetine, COD removals, intermediate products of Kemicetine and COD components (CODs originating from slowly degradable organics, readily degradable organics, inert microbial products and from the inert compounds) were investigated in a sequential anaerobic baffled reactor (ABR)/aerobic completely stirred tank reactor (CSTR) system with a real pharmaceutical wastewater. The total COD and Kemicetine removal efficiencies were 98% and 100%, respectively, in the sequential ABR/CSTR systems. 2-Amino-1 (p-nitrophenil)-1,3 propanediol, l-p-amino phenyl, p-amino phenol and phenol were detected in the ABR as the main readily degradable inter-metabolites. In the anaerobic ABR reactor, the Kemicetin was converted to corresponding inter-metabolites and a substantial part of the COD was removed. In the aerobic CSTR reactor the inter-metabolites produced in the anaerobic reactor were completely removed and the COD remaining from the anerobic reactor was biodegraded. It was found that the COD originating from the readily degradable organics did not limit the anaerobic degradation process, while the CODs originating from the slowly degradable organics and from the inert microbial products significantly decreased the anaerobic ABR reactor performance. The acute toxicity test results indicated that the toxicity decreased from the influent to the effluent of the aerobic CSTR reactor. The ANOVA test statistics showed that there was a strong linear correlation between acute toxicity, CODs originating from the slowly degradable organics and inert microbial products. A weak correlation between acute toxicity and CODs originating from the inert compounds was detected.

  14. Relationships between chemical oxygen demand (COD) components and toxicity in a sequential anaerobic baffled reactor/aerobic completely stirred reactor system treating Kemicetine

    International Nuclear Information System (INIS)

    Sponza, Delia Teresa; Demirden, Pinar

    2010-01-01

    In this study the interactions between toxicity removals and Kemicetine, COD removals, intermediate products of Kemicetine and COD components (CODs originating from slowly degradable organics, readily degradable organics, inert microbial products and from the inert compounds) were investigated in a sequential anaerobic baffled reactor (ABR)/aerobic completely stirred tank reactor (CSTR) system with a real pharmaceutical wastewater. The total COD and Kemicetine removal efficiencies were 98% and 100%, respectively, in the sequential ABR/CSTR systems. 2-Amino-1 (p-nitrophenil)-1,3 propanediol, l-p-amino phenyl, p-amino phenol and phenol were detected in the ABR as the main readily degradable inter-metabolites. In the anaerobic ABR reactor, the Kemicetin was converted to corresponding inter-metabolites and a substantial part of the COD was removed. In the aerobic CSTR reactor the inter-metabolites produced in the anaerobic reactor were completely removed and the COD remaining from the anerobic reactor was biodegraded. It was found that the COD originating from the readily degradable organics did not limit the anaerobic degradation process, while the CODs originating from the slowly degradable organics and from the inert microbial products significantly decreased the anaerobic ABR reactor performance. The acute toxicity test results indicated that the toxicity decreased from the influent to the effluent of the aerobic CSTR reactor. The ANOVA test statistics showed that there was a strong linear correlation between acute toxicity, CODs originating from the slowly degradable organics and inert microbial products. A weak correlation between acute toxicity and CODs originating from the inert compounds was detected.

  15. Probabilistic structural analysis methods for select space propulsion system components

    Science.gov (United States)

    Millwater, H. R.; Cruse, T. A.

    1989-01-01

    The Probabilistic Structural Analysis Methods (PSAM) project developed at the Southwest Research Institute integrates state-of-the-art structural analysis techniques with probability theory for the design and analysis of complex large-scale engineering structures. An advanced efficient software system (NESSUS) capable of performing complex probabilistic analysis has been developed. NESSUS contains a number of software components to perform probabilistic analysis of structures. These components include: an expert system, a probabilistic finite element code, a probabilistic boundary element code and a fast probability integrator. The NESSUS software system is shown. An expert system is included to capture and utilize PSAM knowledge and experience. NESSUS/EXPERT is an interactive menu-driven expert system that provides information to assist in the use of the probabilistic finite element code NESSUS/FEM and the fast probability integrator (FPI). The expert system menu structure is summarized. The NESSUS system contains a state-of-the-art nonlinear probabilistic finite element code, NESSUS/FEM, to determine the structural response and sensitivities. A broad range of analysis capabilities and an extensive element library is present.

  16. In-situ measurement of mechanical properties of structural components using cyclic ball indentation technique

    International Nuclear Information System (INIS)

    Chatterjee, S.; Madhusoodanan, K.; Panwar, Sanjay; Rupani, B.B.

    2007-01-01

    Material properties of components change during service due to environmental conditions. Measurement of mechanical properties of the components is important for assessing their fitness for service. In many instances, it is not possible to remove sizable samples from the component for doing the measurement in laboratory. In-situ technique for measurement of mechanical properties has great significance in such cases. One of the nondestructive methods that can be adopted for in-situ application is based on cyclic ball indentation technique. It involves multiple indentation cycles (at the same penetration location) on a metallic surface by a spherical indenter. Each cycle consists of indentation, partial unload and reload sequences. Presently, commercial systems are available for doing indentation test on structural component for limited applications. But, there is a genuine need of remotely operable compact in-situ property measurement system. Considering the importance of such applications Reactor Engineering Division of BARC has developed an In-situ Property Measurement System (IProMS), which can be used for in-situ measurement of mechanical properties of a flat or tubular component. This paper highlights the basic theory of measurement, qualification tests on IProMS and results from tests done on flat specimens and tubular component. (author)

  17. Fusion component design for the moving-ring field-reversed mirror reactor

    International Nuclear Information System (INIS)

    Carlson, G.A.

    1981-01-01

    This partial report on the reactor design contains sections on the following: (1) burner section magnet system design, (2) plasma ring energy recovery, (3) vacuum system, (4) cryogenic system, (5) tritium flows and inventories, and (6) reactor design and layout

  18. Safety analyses for transient behavior of plasma and in-vessel components during plasma abnormal events in fusion reactor

    International Nuclear Information System (INIS)

    Honda, Takuro; Okazaki, Takashi; Bartels, H.W.; Uckan, N.A.; Seki, Yasushi.

    1997-01-01

    Safety analyses on plasma abnormal events have been performed using a hybrid code of a plasma dynamics model and a heat transfer model of in-vessel components. Several abnormal events, e.g., increase in fueling rate, were selected for the International Thermonuclear Experimental Reactor (ITER) and transient behavior of the plasma and the invessel components during the events was analyzed. The physics model for safety analysis was conservatively prepared. In most cases, the plasma is terminated by a disruption or it returns to the original operation point. When the energy confinement improves by a factor of 2.0 in the steady state, which is a hypothetical assumption under the present plasma data, the maximum fusion power reaches about 3.3 GW at about 3.6 s and the plasma is terminated due to a disruption. However, the results obtained in this study show the confinement boundary of ITER can be kept almost intact during the abnormal plasma transients, as long as the cooling system works normally. Several parametric studies are needed to comprehend the overpower transient including structure behavior, since many uncertainties are connected to the filed of the plasma physics. And, future work will need to discuss the burn control scenario considering confinement mode transition, system specifications, experimental plans and safety regulations, etc. to confirm the safety related to the plasma anomaly. (author)

  19. Technical Meeting on Liquid Metal Reactor Concepts: Core Design and Structural Materials. Presentations

    International Nuclear Information System (INIS)

    2013-01-01

    The objective of the Technical Meeting is to present and discuss innovative liquid metal fast reactor (LMFR) core designs with special focus on the choice, development, testing and qualification of advanced reactor core structural materials

  20. Biofiltration of paint solvent mixtures in two reactor types: overloading by polar components.

    Science.gov (United States)

    Paca, Jan; Halecky, Martin; Misiaczek, Ondrej; Kozliak, Evguenii I; Jones, Kim

    2012-01-01

    Steady-state performances of a trickle bed reactor (TBR) and a biofilter (BF) in loading experiments with increasing inlet concentrations of polar solvents, acetone, methyl ethyl ketone, methyl isobutyl ketone and n-butyl acetate, were investigated, along with the system's dynamic responses. Throughout the entire experimentation time, a constant loading rate of aromatic components of 4 g(c)·m(-3)·h(-1) was maintained to observe the interactions between the polar substrates and aromatic hydrocarbons. Under low combined substrate loadings, the BF outperformed TBR not only in the removal of aromatic hydrocarbons but also in the removal of polar substrates. However, increasing the loading rate of polar components above the threshold value of 31-36 g(c)·m(-3)·h(-1) resulted in a steep and significant drop in the removal efficiencies of both polar (except for butyl acetate) and hydrophobic components, which was more pronounced in the BF; so the relative TBR/BF efficiency became reversed under such overloading conditions. A step-drop of the overall OL(POLAR) (combined loading by polar air pollutants) from overloading values to 7 g(c)·m(-3)·h(-1) resulted in an increase of all pollutant removal efficiencies, although in TBR the recovery was preceded by lag periods lasting between 5 min (methyl ethyl ketone) to 3.7 h (acetone). The occurrence of lag periods in the TBR recovery was, in part, due to the saturation of mineral medium with water-soluble polar solvents, particularly, acetone. The observed bioreactor behavior was consistent with the biological steps being rate-limiting.

  1. The prediction of reliability and residual life of reactor pressure components

    International Nuclear Information System (INIS)

    Nemec, J.; Antalovsky, S.

    1978-01-01

    The paper deals with the problem of PWR pressure components reliability and residual life evaluation and prediction. A physical model of damage cumulation which serves as a theoretical basis for all considerations presents two major aspects. The first one describes the dependence of the degree of damage in the crack leading-edge in pressure components on the reactor system load-time history, i.e. on the number of transient loads. Both stages, fatigue crack initiation and growth through the wall until the critical length is reached, are investigated. The crack is supposed to initiate at the flaws in a strength weld joint or in the bimetallic weld of the base ferritic steel and the austenitic stainless overlay cladding. The growth rates of developed cracks are analysed in respect to different load-time histories. Important cyclic properties of some steels are derived from the low-cycle fatigue theory. The second aspect is the load-time history-dependent process of precipitation, deformation and radiation aging, characterized entirely by the critical crack-length value mentioned above. The fracture point, defined by the equation ''crack-length=critical value'' and hence the residual life, can be evaluated using this model and verified by in-service inspection. The physical model described is randomized by considering all the parameters of the model as random. Monte Carlo methods are applied and fatigue crack initiation and growth is simulated. This permits evaluation of the reliability and residual life of the component. The distributions of material and load-time history parameters are needed for such simulation. Both the deterministic and computer-simulated probabilistic predictions of reliability and residual life are verified by prior-to-failure sequential testing of data coming from in-service NDT periodical inspections. (author)

  2. Biofiltration of paint solvent mixtures in two reactor types: overloading by hydrophobic components.

    Science.gov (United States)

    Paca, Jan; Halecky, Martin; Misiaczek, Ondrej; Jones, Kim; Kozliak, Evguenii; Sobotka, Miroslav

    2010-12-01

    Steady-state performance characteristics of a trickle bed reactor (TBR) and a biofilter (BF) in loading experiments with increasing toluene/xylenes inlet concentrations while maintaining a constant loading rate of hydrophilic components (methyl ethyl and methyl isobutyl ketones, acetone, and n-butyl acetate) of 4 g m⁻³ h⁻¹ were evaluated and compared, along with the systems' dynamic responses. At the same combined substrate loading of 55 g m⁻³ h⁻¹ for both reactors, the TBR achieved more than 1.5 times higher overall removal efficiency (RE(W)) than the BF. Increasing the loading rate of aromatics resulted in a gradual decrease of their REs. The degradation rates of acetone and n-butyl acetate were also inhibited at higher loads of aromatics, thus revealing a competition in cell catabolism. A step-drop in loading of aromatics resulted in an immediate increase of RE(W) with variations in the TBR, while the new steady-state value in the BF took 6-7 h to achieve. The TBR consistently showed a greater performance than BF in removing toluene and xylenes. Increasing the loading rate of aromatics resulted in a gradual decrease of their REs. The degradation rates of acetone and n-butyl acetate were also lower at higher OL(AROM), revealing a competition in the cell catabolism. The results obtained are consistent with the proposed hypothesis of greater toxic effects under low water content, i.e., in the biofilter, caused by aromatic hydrocarbons in the presence of polar ketones and esters, which may improve the hydrocarbon partitioning into the aqueous phase.

  3. Treatment of core components from nuclear power plants with PWR and BWR reactors - 16043

    International Nuclear Information System (INIS)

    Viermann, Joerg; Friske, Andreas; Radzuweit, Joerg

    2009-01-01

    During operation of a Nuclear Power Plant components inside the RPV get irradiated. Irradiation has an effect on physical properties of these components. Some components have to be replaced after certain neutron doses or respectively after a certain operating time of the plant. Such components are for instance water channels and control rods from Boiling Water Reactors (BWR) or control elements, poisoning elements and flow restrictors from Pressurized Water Reactors (PWR). Most of these components are stored in the fuel pool for a certain time after replacement. Then they have to be packaged for further treatment or for disposal. More than 25 years ago GNS developed a system for disposal of irradiated core components which was based on a waste container suitable for transport, storage and disposal of Intermediate Level Waste (ILW), the so-called MOSAIK R cask. The MOSAIK R family of casks is subject of a separate presentation at the ICEM 09 conference. Besides the MOSAIK R cask the treatment system developed by GNS comprised underwater shears to cut the components to size as well as different types of equipment to handle the components, the shears and the MOSAIK R casks in the fuel pool. Over a decade of experience it showed that this system although effective needed improvement for BWR plants where many water channels and control rods had to be replaced after a certain operating time. Because of the large numbers of components the time period needed to cut the components in the pool had a too big influence on other operational work like rearranging of fuel assemblies in the pool. The system was therefore further developed and again a suitable cask was the heart of the solution. GNS developed the type MOSAIK R 80 T, a cask that is capable to ship the unsegmented components with a length of approx. 4.5 m from the Power plants to an external treatment centre. This treatment centre consisting of a hot cell installation with a scrap shear, super-compactor and a heavy

  4. Structural ECM components in the premetastatic and metastatic niche

    DEFF Research Database (Denmark)

    Høye, Anette M; Erler, Janine T

    2016-01-01

    The aim of this review is to give an overview of the extracellular matrix (ECM) components that are important for creating structural changes in the premetastatic and metastatic niche. The successful arrival and survival of cancer cells that have left the primary tumor and colonized distant sites...... aimed at targeting cell-ECM interactions may well be one of the best viable approaches to combat metastasis and thus improve patient care....

  5. Seismic fragility analysis of structural components for HFBR facilities

    International Nuclear Information System (INIS)

    Park, Y.J.; Hofmayer, C.H.

    1992-01-01

    The paper presents a summary of recently completed seismic fragility analyses of the HFBR facilities. Based on a detailed review of past PRA studies, various refinements were made regarding the strength and ductility evaluation of structural components. Available laboratory test data were analysed to evaluate the formulations used to predict the ultimate strength and deformation capacities of steel, reinforced concrete and masonry structures. The biasness and uncertainties were evaluated within the framework of the fragility evaluation methods widely accepted in the nuclear industry. A few examples of fragility calculations are also included to illustrate the use of the presented formulations

  6. Mechanical and materials engineering of modern structure and component design

    CERN Document Server

    Altenbach, Holm

    2015-01-01

    This book presents the latest findings on mechanical and materials engineering as applied to the design of modern engineering materials and components. The contributions cover the classical fields of mechanical, civil and materials engineering, as well as bioengineering and advanced materials processing and optimization. The materials and structures discussed can be categorized into modern steels, aluminium and titanium alloys, polymers/composite materials, biological and natural materials, material hybrids and modern nano-based materials. Analytical modelling, numerical simulation, state-of-the-art design tools and advanced experimental techniques are applied to characterize the materials’ performance and to design and optimize structures in different fields of engineering applications.

  7. Development of a Versatile Ultrasonic Internal Pipe/Vessel Component Monitor for In-Service Inspection of Nuclear Reactor Components

    Energy Technology Data Exchange (ETDEWEB)

    Searfass, Clifford T. [Structural Integrity Associates, Inc., State College, PA (United States); Malinowski, Owen M. [Structural Integrity Associates, Inc., State College, PA (United States); Van Velsor, Jason K. [Structural Integrity Associates, Inc., State College, PA (United States)

    2015-03-22

    The stated goal of this work was to develop a versatile system which could accurately measure vessel and valve internal vibrations and cavitation formation under in-service conditions in nuclear power plants, ultrasonically. The developed technology will benefit the nuclear power generation industry by allowing plant operators to monitor valve and vessel internals during operation. This will help reduce planned outages and plant component failures. During the course of this work, Structural Integrity Associates, Inc. gathered information from industry experts that target vibration amplitudes to be detected should be in the range of 0.001-in to 0.005-in (0.025-mm to 0.127-mm) and target vibration frequency ranges which should be detected were found to be between 0-Hz and 300-Hz. During the performed work, an ultrasonic measuring system was developed which utilized ultrasonic pulse-echo time-of-flight measurements to measure vibration frequency and amplitude. The developed system has been shown to be able to measure vibration amplitudes as low as 0.0008-in (0.020-mm) with vibration frequencies in the range of 17-Hz to 1000-Hz. Therefore, the developed system was able to meet the industry needs for vibration measurement. The developed ultrasonic system was also to be able to measure cavitation formation by monitoring the received ultrasonic time- and frequency-domain signals. This work also demonstrated the survivability of commercially available probes at temperatures up to 300-F for several weeks.

  8. An automatic chip structure optical inspection system for electronic components

    Science.gov (United States)

    Song, Zhichao; Xue, Bindang; Liang, Jiyuan; Wang, Ke; Chen, Junzhang; Liu, Yunhe

    2018-01-01

    An automatic chip structure inspection system based on machine vision is presented to ensure the reliability of electronic components. It consists of four major modules, including a metallographic microscope, a Gigabit Ethernet high-resolution camera, a control system and a high performance computer. An auto-focusing technique is presented to solve the problem that the chip surface is not on the same focusing surface under the high magnification of the microscope. A panoramic high-resolution image stitching algorithm is adopted to deal with the contradiction between resolution and field of view, caused by different sizes of electronic components. In addition, we establish a database to storage and callback appropriate parameters to ensure the consistency of chip images of electronic components with the same model. We use image change detection technology to realize the detection of chip images of electronic components. The system can achieve high-resolution imaging for chips of electronic components with various sizes, and clearly imaging for the surface of chip with different horizontal and standardized imaging for ones with the same model, and can recognize chip defects.

  9. Extraction of Independent Structural Images for Principal Component Thermography

    Directory of Open Access Journals (Sweden)

    Dmitry Gavrilov

    2018-03-01

    Full Text Available Thermography is a powerful tool for non-destructive testing of a wide range of materials. Thermography has a number of approaches differing in both experiment setup and the way the collected data are processed. Among such approaches is the Principal Component Thermography (PCT method, which is based on the statistical processing of raw thermal images collected by thermal camera. The processed images (principal components or empirical orthogonal functions form an orthonormal basis, and often look like a superposition of all possible structural features found in the object under inspection—i.e., surface heating non-uniformity, internal defects and material structure. At the same time, from practical point of view it is desirable to have images representing independent structural features. The work presented in this paper proposes an approach for separation of independent image patterns (archetypes from a set of principal component images. The approach is demonstrated in the application of inspection of composite materials as well as the non-invasive analysis of works of art.

  10. Structural Components of Lifestyle and Beyond: The Case of Hungary

    Directory of Open Access Journals (Sweden)

    Tamás Keller

    2011-06-01

    Full Text Available This paper deals with the question of when and how lifestyle and its components are important in social stratification. There is considerable consensus among scholars about the structure of the society being a consequence of hierarchical dimensions like occupation, income, or wealth. Some thirty years ago, largely based on Bourdieu’s “Distinction”, a new paradigm emerged highlighting the lifestyle components and the value-oriented cultural and material consumption in stratifi cation. The idea refl ects the empirical fi nding that inequality between social classes has largely decreased, giving priority to horizontal lifestyle differentiation instead of vertical inequality dimensions. From a theoretical viewpoint, a challenge in the approach is finding out to what extent lifestyle typology is of a non-vertical character in reality. This social determination of lifestyle is investigated for Hungary when comparing an occupation-based typology with a consumption-based one. On the one hand, results reveal that the effects of structural components on social status are stronger than those of lifestyle. On the other hand, lifestyle turns out to be less independent of social position and the top and bottom levels of the lifestyle typology are particularly predictable by structural measures.

  11. Nuclear reactor component populations, reliability data bases, and their relationship to failure rate estimation and uncertainty analysis

    International Nuclear Information System (INIS)

    Martz, H.F.; Beckman, R.J.

    1981-12-01

    Probabilistic risk analyses are used to assess the risks inherent in the operation of existing and proposed nuclear power reactors. In performing such risk analyses the failure rates of various components which are used in a variety of reactor systems must be estimated. These failure rate estimates serve as input to fault trees and event trees used in the analyses. Component failure rate estimation is often based on relevant field failure data from different reliability data sources such as LERs, NPRDS, and the In-Plant Data Program. Various statistical data analysis and estimation methods have been proposed over the years to provide the required estimates of the component failure rates. This report discusses the basis and extent to which statistical methods can be used to obtain component failure rate estimates. The report is expository in nature and focuses on the general philosophical basis for such statistical methods. Various terms and concepts are defined and illustrated by means of numerous simple examples

  12. Trends in large-scale testing of reactor structures

    International Nuclear Information System (INIS)

    Blejwas, T.E.

    2003-01-01

    Large-scale tests of reactor structures have been conducted at Sandia National Laboratories since the late 1970s. This paper describes a number of different large-scale impact tests, pressurization tests of models of containment structures, and thermal-pressure tests of models of reactor pressure vessels. The advantages of large-scale testing are evident, but cost, in particular limits its use. As computer models have grown in size, such as number of degrees of freedom, the advent of computer graphics has made possible very realistic representation of results - results that may not accurately represent reality. A necessary condition to avoiding this pitfall is the validation of the analytical methods and underlying physical representations. Ironically, the immensely larger computer models sometimes increase the need for large-scale testing, because the modeling is applied to increasing more complex structural systems and/or more complex physical phenomena. Unfortunately, the cost of large-scale tests is a disadvantage that will likely severely limit similar testing in the future. International collaborations may provide the best mechanism for funding future programs with large-scale tests. (author)

  13. Flow induced vibrational excitation of nuclear reactor structures

    International Nuclear Information System (INIS)

    Gibert, R.J.

    1979-01-01

    The pressure fluctuations generated by disturbed flows, encountered in nuclear reactors induce vibrations in the structures. In order to make forecastings for these vibrational levels, it is necessary to know the characteristics of the random pressure fluctuations induced in the walls by the main flow peculiarities of the circuits. This knowledge is essentially provided by experimentation which shows that most of the energy from these fluctuations is in the low frequency area. It is also necessary to determine the transfer functions of the fluid-structure coupled system. Given the frequency range of the excitations, a calculation of the characteristics of the first eigenmodes is generally sufficient. This calculation is carried out by finite element codes, the modal dampings being assessed separately. In this paper, emphasis is placed mainly on the analysis of the sources of excitation due to flow peculiarities. Some examples will also be given of assessments of vibrations in real structures (pipes, reactor internals, etc.) and of comparisons with the experimental results obtained on models or on a site [fr

  14. Evaluation of materials for heat exchanging components in advanced helium-cooled reactors

    International Nuclear Information System (INIS)

    Schubert, F.

    1984-01-01

    The qualification of metallic materials for advanced HTR applications is based on creep behaviour, fatigue properties, structural stability and corrosion resistance. A brief state of the art is provided for the materials for heat exchanging components. The experimental results are treated with respect to the importance for the design, the characteristic of time-depend materials behaviour are evaluated. Of specific interest are the possible effects of helium on the mechanical properties. Helium, which serves as primary coolant, contains traces of reactive impurities such as hydrogen, methane, carbon monoxide and water vapor. The evaluation of the HTR materials program serves as basis for structural design rules of components with operation temperatures above 800 deg C. The materials mechanical topics are discussed. Alloy improvement and the progress in development of new alloys are reviewed. (author)

  15. Grain boundary engineering for structure materials of nuclear reactors

    Science.gov (United States)

    Tan, L.; Allen, T. R.; Busby, J. T.

    2013-10-01

    Grain boundary engineering (GBE), primarily implemented by thermomechanical processing, is an effective and economical method of enhancing the properties of polycrystalline materials. Among the factors affecting grain boundary character distribution, literature data showed definitive effect of grain size and texture. GBE is more effective for austenitic stainless steels and Ni-base alloys compared to other structural materials of nuclear reactors, such as refractory metals, ferritic and ferritic-martensitic steels, and Zr alloys. GBE has shown beneficial effects on improving the strength, creep strength, and resistance to stress corrosion cracking and oxidation of austenitic stainless steels and Ni-base alloys.

  16. Swelling of structural materials in fast neutron reactors

    International Nuclear Information System (INIS)

    Seran, J.L.

    1983-06-01

    The physical origin of swelling in irradiated materials and the main parameters acting on swelling of SS 316 are examined: temperature, neutron dose, dose rate, chemical composition, strain hardening. Results obtained, in Rapsodie and Phenix reactors, with fuel cans and with the hexagonal tube containing the fuel pins are analyzed and compared with results found in litterature. In conclusion hot swelling of SS 316 is too important at high doses and is will be replaced by austenitic steels stabilized by Ti and ferritic steels or high nickel steels with structural hardening [fr

  17. Grain boundary engineering for structure materials of nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Tan, L., E-mail: tanl@ornl.gov [Materials Science and Technology Division, Oak Ridge National Laboratory (United States); Allen, T.R. [Department of Engineering Physics, University of Wisconsin–Madison (United States); Busby, J.T. [Materials Science and Technology Division, Oak Ridge National Laboratory (United States)

    2013-10-15

    Grain boundary engineering (GBE), primarily implemented by thermomechanical processing, is an effective and economical method of enhancing the properties of polycrystalline materials. Among the factors affecting grain boundary character distribution, literature data showed definitive effect of grain size and texture. GBE is more effective for austenitic stainless steels and Ni-base alloys compared to other structural materials of nuclear reactors, such as refractory metals, ferritic and ferritic–martensitic steels, and Zr alloys. GBE has shown beneficial effects on improving the strength, creep strength, and resistance to stress corrosion cracking and oxidation of austenitic stainless steels and Ni-base alloys.

  18. Capabilities and limitations of fracture mechanics methods in the assessment of integrity of light water reactor components

    Energy Technology Data Exchange (ETDEWEB)

    Burdekin, F M

    1988-12-31

    This document deals with fracture mechanics methods used for the assessment of Light Water Reactor (LWR) components. The background to analysis methods using elastic plastic parameters is described. Several results obtained with these methods are presented as well as results of reliability analysis methods. (TEC). 27 refs.

  19. Assessment of core structural materials and surveillance programme of research reactors. Report of the consultants meeting. Working material

    International Nuclear Information System (INIS)

    2009-01-01

    A series of presentations on the assessment of core structural components and materials at their facilities were given by the experts. The different issues related to degradation mechanisms were discussed. The outputs include a more thorough understanding of the specific challenges related to Research Reactors (RRs) as well as proposals for activities which could assist RR organizations in their efforts to address the issues involved. The experts recommend that research reactor operators consider implementation of surveillance programs for materials of core structural components, as part of ageing management program (TECDOC-792 and DS-412). It is recognised by experts that adequate archived structural material data is not available for many RRs. Access to this data and extension of existing material databases could help many operating organisations extend the operation of their RRs. The experts agreed that an IAEA Technical Meeting (TM) on Assessment of Core Structural Materials should be organised in December 2009 (IAEA HQ Vienna). The proposed objectives of the TM are: (i) exchange of detailed technical information on the assessment and ageing management of core structural materials, (ii) identification of materials of interest for further investigation, (iii) proposal for a new IAEA CRP on Assessment of Core Structural Materials, and (iv) identification of RRs prepared to participate in proposed CRP. Based on the response to a questionnaire prepared for the 2008 meeting of the Technical Working Group for Research Reactors, the number of engineering capital projects related to core structural components is proportionally lower than those related to,for example, I and C or electrical power systems. This implies that many operating research reactors will be operating longer using their original core structural components and justifies the assessment and evaluation programmes and activities proposed in this report. (author)

  20. Development of expert system for structural design of FBR components

    International Nuclear Information System (INIS)

    Ueda, Hiroyoshi; Uno, Masayoshi; Ogawa, Hiroshi; Shimakawa, Takashi; Yoshimura, Shinobu; Yagawa, Genki.

    1995-01-01

    The characteristics of structural design processes for nuclear components can be summarized as follows : (1) Many engineers belonging to different fields are working in parallel, exchanging a huge amount of data and information. (2) A final solution is determined after a number of iterative design processes. (3) Solutions have to be examined many times based on sophisticated design codes. (4) Sophisticated calculation methods such as the finite element method are frequently utilized, and experts' knowledge on such analyses plays important roles in the design process. Taking these issues into consideration, a new expert system for structural design is developed in the present study. Here, the object-oriented data flow mechanism and the blackboard model are utilized to systematize structural design processes in a computer. An automated finite element calculation module is implemented, and experts' knowledge is stored in knowledge base. In addition, a new algorithm is employed to automatically draw the design window, which is defined as an area of permissible solutions in a design parameter space. The developed system is successfully applied to obtain the design windows of four components selected from the demonstration FBR structures. (author)

  1. Effects of air blast on power plant structures and components

    International Nuclear Information System (INIS)

    Kot, C.A.; Valentin, R.A.; McLennan, D.A.; Turula, P.

    1978-10-01

    The effects of air blast from high explosives detonation on selected power plant structures and components are investigated analytically. Relying on a synthesis of state of the art methods estimates of structural response are obtained. Similarly blast loadings are determined from compilations of experimental data reported in the literature. Plastic-yield line analysis is employed to determine the response of both concrete and steel flat walls (plates) under impulsive loading. Linear elastic theory is used to investigate the spalling of concrete walls and mode analysis methods predict the deflection of piping. The specific problems considered are: the gross deformation of reinforced concrete shield and containment structures due to blast impulse, the spalling of concrete walls, the interaction or impact of concrete debris with steel containments and liners, and the response of exposed piping to blast impulse. It is found that for sufficiently close-in detonations and/or large explosive charge weights severe damage or destruction will result. This is particularly true for structures or components directly exposed to blast impulse

  2. Structure of the vault, a ubiquitous celular component.

    Science.gov (United States)

    Kong, L B; Siva, A C; Rome, L H; Stewart, P L

    1999-04-15

    The vault is a ubiquitous and highly conserved ribonucleoprotein particle of approximately 13 MDa. This particle has been shown to be upregulated in certain multidrug-resistant cancer cell lines and to share a protein component with the telomerase complex. Determination of the structure of the vault was undertaken to provide a first step towards understanding the role of this cellular component in normal metabolism and perhaps to shed some light on its role in mediating drug resistance. Over 1300 particle images were combined to calculate an approximately 31 A resolution structure of the vault. Rotational power spectra did not yield a clear symmetry peak, either because of the thin, smooth walls or inherent flexibility of the vault. Although cyclic eightfold (C8) symmetry was imposed, the resulting reconstruction may be partially cylindrically averaged about the eightfold axis. Our results reveal the vault to be a hollow, barrel-like structure with two protruding caps and an invaginated waist. Although the normal cellular function of the vault is as yet undetermined, the structure of the vault is consistent with either a role in subcellular transport, as previously suggested, or in sequestering macromolecular assemblies.

  3. Detailed Structural Analysis of Critical Wendelstein 7-X Magnet System Components

    International Nuclear Information System (INIS)

    Egorov, K.

    2006-01-01

    The Wendelstein 7-X (W7-X) stellarator experiment is presently under construction and assembly in Greifswald, Germany. The goal of the experiment is to verify that the stellarator magnetic confinement concept is a viable option for a fusion reactor. The complex W7-X magnet system requires a multi-level approach to structural analysis for which two types of finite element models are used: Firstly, global models having reasonably coarse meshes with a number of simplifications and assumptions, and secondly, local models with detailed meshes of critical regions and elements. Widely known sub-modelling technique with boundary conditions extracted from the global models is one of the approaches for local analysis with high assessment efficiency. In particular, the winding pack (WP) of the magnet coils is simulated in the global model as a homogeneous orthotropic material with effective mechanical characteristic representing its real composite structure. This assumption allows assessing the whole magnet system in terms of general structural factors like forces and moments on the support elements, displacements of the main components, deformation and stress in the coil casings, etc. In a second step local models with a detailed description of more critical WP zones are considered in order to analyze their internal components like conductor jackets, turn insulation, etc. This paper provides an overview of local analyses of several critical W7-X magnet system components with particular attention on the coil winding packs. (author)

  4. The maintenance optimization of structural components in nuclear power plants

    International Nuclear Information System (INIS)

    Bryla, P.; Ardorino, F.; Aufort, P.; Jacquot, J.P.; Magne, L.; Pitner, P.; Verite, B.; Villain, B.; Monnier, B.

    1997-10-01

    An optimization process, called 'OMF-Structures', is developed by Electricite de France (EDF) in order to extend the current 'OMF' Reliability Centered Maintenance to piping structural components. The Auxiliary Feedwater System of a 900 MW French nuclear plant has been studied in order to lay the foundations of the method. This paper presents the currently proposed principles of the process. The principles of the OMF-Structures process include 'Risk-Based Inspection' concepts within an RCM process. Two main phases are identified: The purpose of the first phase is to select the risk-significant failure modes and associated elements. This phase consists of two major steps: potential consequences evaluation and reliability performance evaluation. The second phase consists of the definition of preventive maintenance programs for piping elements that are associated with risk-significant failure modes. (author)

  5. Methodologies to assess PWSCC susceptibility of primary component Alloy 600 locations in pressurized water reactors

    International Nuclear Information System (INIS)

    Rao, G.V.

    1993-01-01

    Methodologies to assess susceptibility to Primary Water Stress Corrosion Cracking (PWSCC) of Alloy 600 component locations in the Primary System of Pressurized Water Reactors are presented. The assessment methodologies are presented. The assessment methodologies are based on Relative Susceptibility Index (RSI) and Cumulative Susceptibility Index (CSI) models utilizing key contributing parameters such as service and residual stresses, yield strength, service temperature, material condition and microstructure, and the accumulated service time. To aid in the development of future inspection plans, a method of ranking of the assessed susceptibilities by 'bench marking' with respect to the susceptibility of a reference location of known PWSCC history of a reference location of known PWSCC history is presented. Means of utilizing the susceptibility ranking results in developing a prioritized inspection plan are discussed. A follow-up investigative plan to the initial inspection is proposed, which includes identification of critical sampling locations, sample extraction, sample investigations and testing to ensure that the potentially highest susceptibility locations are free from near term PWSCC and, further, to provide a basis for established schedules for future inspections. Finally, parametric considerations of the contributing factor are presented to help the utility choose suitable option to mitigate the PWSCC issue while minimizing the impact on continued service

  6. Fault detection in nonlinear chemical processes based on kernel entropy component analysis and angular structure

    Energy Technology Data Exchange (ETDEWEB)

    Jiang, Qingchao; Yan, Xuefeng; Lv, Zhaomin; Guo, Meijin [East China University of Science and Technology, Shanghai (China)

    2013-06-15

    Considering that kernel entropy component analysis (KECA) is a promising new method of nonlinear data transformation and dimensionality reduction, a KECA based method is proposed for nonlinear chemical process monitoring. In this method, an angle-based statistic is designed because KECA reveals structure related to the Renyi entropy of input space data set, and the transformed data sets are produced with a distinct angle-based structure. Based on the angle difference between normal status and current sample data, the current status can be monitored effectively. And, the confidence limit of the angle-based statistics is determined by kernel density estimation based on sample data of the normal status. The effectiveness of the proposed method is demonstrated by case studies on both a numerical process and a simulated continuous stirred tank reactor (CSTR) process. The KECA based method can be an effective method for nonlinear chemical process monitoring.

  7. Fault detection in nonlinear chemical processes based on kernel entropy component analysis and angular structure

    International Nuclear Information System (INIS)

    Jiang, Qingchao; Yan, Xuefeng; Lv, Zhaomin; Guo, Meijin

    2013-01-01

    Considering that kernel entropy component analysis (KECA) is a promising new method of nonlinear data transformation and dimensionality reduction, a KECA based method is proposed for nonlinear chemical process monitoring. In this method, an angle-based statistic is designed because KECA reveals structure related to the Renyi entropy of input space data set, and the transformed data sets are produced with a distinct angle-based structure. Based on the angle difference between normal status and current sample data, the current status can be monitored effectively. And, the confidence limit of the angle-based statistics is determined by kernel density estimation based on sample data of the normal status. The effectiveness of the proposed method is demonstrated by case studies on both a numerical process and a simulated continuous stirred tank reactor (CSTR) process. The KECA based method can be an effective method for nonlinear chemical process monitoring

  8. Activation and Radiation Damage Behaviour of Russian Structural Materials for Fusion Reactors in the Fission and Fusion Reactors

    International Nuclear Information System (INIS)

    Blokhin, A.; Demin, N.; Chernov, V.; Leonteva-Smirnova, M.; Potapenko, M.

    2006-01-01

    Various structural low (reduced) activated materials have been proposed as a candidate for the first walls-blankets of fusion reactors. One of the main problems connected with using these materials - to minimise the production of long-lived radionuclides from nuclear transmutations and to provide with good technological and functional properties. The selection of materials and their metallurgical and fabrication technologies for fusion reactor components is influenced by this factor. Accurate prediction of induced radioactivity is necessary for the development of the fusion reactor materials. Low activated V-Ti-Cr alloys and reduced activated ferritic-martensitic steels are a leading candidate material for fusion first wall and blanket applications. At the present time a range of compositions and an impurity level are still being investigated to better understand the sensitive of various functional and activation properties to small compositional variations and impurity level. For the two types of materials mentioned above (V-Ti-Cr alloys and 9-12 % Cr f/m steels) and manufactured in Russia (Russia technologies) the analysis of induced activity, hydrogen and helium-production as well as the accumulation of such elements as C, N, O, P, S, Zn and Sn as a function of irradiation time was performed. Materials '' were irradiated '' by fission (BN-600, BOR-60) and fusion (Russian DEMO-C Reactor Project) typical neutron spectra with neutron fluency up to 10 22 n/cm 2 and the cooling time up to 1000 years. The calculations of the transmutation of elements and the induced radioactivity were carried out using the FISPACT inventory code, and the different activation cross-section libraries like the ACDAM, FENDL-2/A and the decay data library FENDL-2/D. It was shown that the level of impurities controls a long-term behaviour of induced activity and contact dose rate for materials. From this analysis the concentration limits of impurities were obtained. The generation of gas

  9. Technical and management challenges associated with structural materials degradation in nuclear reactors in the future

    International Nuclear Information System (INIS)

    Ford, F.P.

    2007-01-01

    There are active plans worldwide to increase nuclear power production by significant amounts. In the near term (i.e. by 2020) this will be accomplished by, (a) increasing the power output of the existing reactors and extending their life, and by, (b) constructing new reactors that are very similar to the current water-cooled designs. Beyond 2025-2030, it is possible that new reactors (i.e. the 'GEN IV' designs) will be very different from those currently in service. A full discussion of the technical and management concerns associated with materials degradation that might arise over the next 40 years would need to address a wide range of topics. Quite apart from discussing the structural integrity issues for the materials of construction and the fuel cladding, the debate would also need to cover, for example, fuel resources and the associated issues of fuel cycle management and waste disposal, manufacturing capacity, inspection capabilities, human reliability, etc., since these all impact to one degree or another on the choice of material and the reactor operating conditions. For brevity, the scope of this article is confined to the integrity of the materials of construction for passive components in the current water-cooled reactors and the evolutionary designs (which will dominate the near term new constructions), and the very different GEN IV reactor designs. In all cases the operating environments will be more aggressive than currently encountered. For instance, the concerns for flow accelerated corrosion and flow-induced vibration will be increased under extended power uprate conditions for the current water-cooled reactors. Of greater concern, the design life will be at least 60 years for all of the new reactors and for those current reactors operating with extended licenses. This automatically presents challenges with regard to managing both irradiation damage in metallic and non-metallic materials of construction, and environmentally assisted cracking. This

  10. Structural stability analysis considerations in fusion reactor plasma chamber design

    International Nuclear Information System (INIS)

    Delaney, M.J.; Cramer, B.A.

    1978-01-01

    This paper presents an approach to analyzing a toroidal plasma chamber for the prevention of both static and dynamic buckling. Results of stability analyses performed for the doublet shaped plasma chamber of the General Atomic 3.8 meter radius TNS ignition test reactor are presented. Load conditions are the static external atmospheric pressure load and the dynamic plasma disruption pulse load. Methods for analysis of plasma chamber structures are presented for both types of load. Analysis for static buckling is based on idealizing the plasma chamber into standard structural shapes and applying classical cylinder and circular torus buckling equations. Results are verified using the Buckling of Shells of Revolution (BOSOR4) finite difference computer code. Analysis for the dynamic loading is based on a pulse buckling analysis method for circular cylinders

  11. Improved computation method in residual life estimation of structural components

    Directory of Open Access Journals (Sweden)

    Maksimović Stevan M.

    2013-01-01

    Full Text Available This work considers the numerical computation methods and procedures for the fatigue crack growth predicting of cracked notched structural components. Computation method is based on fatigue life prediction using the strain energy density approach. Based on the strain energy density (SED theory, a fatigue crack growth model is developed to predict the lifetime of fatigue crack growth for single or mixed mode cracks. The model is based on an equation expressed in terms of low cycle fatigue parameters. Attention is focused on crack growth analysis of structural components under variable amplitude loads. Crack growth is largely influenced by the effect of the plastic zone at the front of the crack. To obtain efficient computation model plasticity-induced crack closure phenomenon is considered during fatigue crack growth. The use of the strain energy density method is efficient for fatigue crack growth prediction under cyclic loading in damaged structural components. Strain energy density method is easy for engineering applications since it does not require any additional determination of fatigue parameters (those would need to be separately determined for fatigue crack propagation phase, and low cyclic fatigue parameters are used instead. Accurate determination of fatigue crack closure has been a complex task for years. The influence of this phenomenon can be considered by means of experimental and numerical methods. Both of these models are considered. Finite element analysis (FEA has been shown to be a powerful and useful tool1,6 to analyze crack growth and crack closure effects. Computation results are compared with available experimental results. [Projekat Ministarstva nauke Republike Srbije, br. OI 174001

  12. Principal Component Analysis Based Measure of Structural Holes

    Science.gov (United States)

    Deng, Shiguo; Zhang, Wenqing; Yang, Huijie

    2013-02-01

    Based upon principal component analysis, a new measure called compressibility coefficient is proposed to evaluate structural holes in networks. This measure incorporates a new effect from identical patterns in networks. It is found that compressibility coefficient for Watts-Strogatz small-world networks increases monotonically with the rewiring probability and saturates to that for the corresponding shuffled networks. While compressibility coefficient for extended Barabasi-Albert scale-free networks decreases monotonically with the preferential effect and is significantly large compared with that for corresponding shuffled networks. This measure is helpful in diverse research fields to evaluate global efficiency of networks.

  13. Design and structural calculation of nuclear power plant mechanical components

    International Nuclear Information System (INIS)

    Amaral, J.A.R. do

    1986-01-01

    The mechanical components of a nuclear power plant must show high quality and safety due to the presence of radioactivity. Besides the perfect functioning during the rigid operating conditions, some postulated loadings are foreseen, like earthquake and loss of coolant accidents, which must be also considered in the design. In this paper, it is intended to describe the design and structural calculations concept and development, the interactions with the piping and civil designs, as well as their influences in the licensing process with the authorities. (Author) [pt

  14. Assessment of design limits and criteria requirements for Eurofer structures in TBM components

    Energy Technology Data Exchange (ETDEWEB)

    Aiello, G., E-mail: giacomo.aiello@cea.fr [CEA, DEN/DANS/DM2S, F-91191 Gif-sur-Yvette (France); Aktaa, J. [Forschungszentrum Karlsruhe (FZK), Institute for Materials Research II, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Cismondi, F. [Forschungszentrum Karlsruhe (FZK), Institut fuer Neutronenphysik und Reaktortechnik, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Rampal, G.; Salavy, J.-F. [CEA, DEN/DANS/DM2S, F-91191 Gif-sur-Yvette (France); Tavassoli, F. [CEA, DEN/DANS/DMN/DIR, F-91191 Gif-sur-Yvette (France)

    2011-07-01

    Eurofer97 is a Reduced Activation Ferritic-Martensitic (RAFM) steel developed for use as structural material in fusion power reactors blankets and in particular the future DEMOnstration power plant that should follow ITER. In order to evaluate the performances of the different blanket concepts in a fusion-relevant environment, the ITER experimental programme foresees the installation of dedicated Test Blanket Modules (TBMs), representative of the corresponding DEMO blankets, in selected equatorial ports. To be fully relevant, TBMs will have to be designed and fabricated using DEMO relevant technologies and will, in particular, use Eurofer97 as structural material. While the use of ferritic/martensitic steels is not new in the nuclear industry, the fusion environment in ITER poses new challenges for the structural materials. Besides, contrary to DEMO, ITER is characterised by a strongly pulsed mode of operation that could have severe consequences on the lifetime of the components. This paper gives an overview of the issues related to the design of Eurofer97 structures in TBM components, discussing the choice of reference Codes and Standards and the consistency of the design rules with Eurofer97 mechanical properties.

  15. Magnons in one-dimensional k-component Fibonacci structures

    Energy Technology Data Exchange (ETDEWEB)

    Costa, C. H., E-mail: carloshocosta@hotmail.com [Departamento de Física Teórica e Experimental, Universidade Federal do Rio Grande do Norte, 59072-970 Natal-RN (Brazil); Escola de Ciências e Tecnologia, Universidade Federal do Rio Grande do Norte, 59072-970 Natal-RN (Brazil); Vasconcelos, M. S. [Escola de Ciências e Tecnologia, Universidade Federal do Rio Grande do Norte, 59072-970 Natal-RN (Brazil)

    2014-05-07

    We have studied the magnon transmission through of one-dimensional magnonic k-component Fibonacci structures, where k different materials are arranged in accordance with the following substitution rule: S{sub n}{sup (k)}=S{sub n−1}{sup (k)}S{sub n−k}{sup (k)} (n≥k=0,1,2,…), where S{sub n}{sup (k)} is the nth stage of the sequence. The calculations were carried out in exchange dominated regime within the framework of the Heisenberg model and taking into account the RPA approximation. We have considered multilayers composed of simple cubic spin-S Heisenberg ferromagnets, and, by using the powerful transfer-matrix method, the spin wave transmission is obtained. It is demonstrated that the transmission coefficient has a rich and interesting magnonic pass- and stop-bands structures, which depends on the frequency of magnons and the k values.

  16. Simulation approaches to probabilistic structural design at the component level

    International Nuclear Information System (INIS)

    Stancampiano, P.A.

    1978-01-01

    In this paper, structural failure of large nuclear components is viewed as a random process with a low probability of occurrence. Therefore, a statistical interpretation of probability does not apply and statistical inferences cannot be made due to the sparcity of actual structural failure data. In such cases, analytical estimates of the failure probabilities may be obtained from stress-strength interference theory. Since the majority of real design applications are complex, numerical methods are required to obtain solutions. Monte Carlo simulation appears to be the best general numerical approach. However, meaningful applications of simulation methods suggest research activities in three categories: methods development, failure mode models development, and statistical data models development. (Auth.)

  17. Multigroup Moderation Test in Generalized Structured Component Analysis

    Directory of Open Access Journals (Sweden)

    Angga Dwi Mulyanto

    2016-05-01

    Full Text Available Generalized Structured Component Analysis (GSCA is an alternative method in structural modeling using alternating least squares. GSCA can be used for the complex analysis including multigroup. GSCA can be run with a free software called GeSCA, but in GeSCA there is no multigroup moderation test to compare the effect between groups. In this research we propose to use the T test in PLS for testing moderation Multigroup on GSCA. T test only requires sample size, estimate path coefficient, and standard error of each group that are already available on the output of GeSCA and the formula is simple so the user does not need a long time for analysis.

  18. Development on methods for evaluating structure reliability of piping components

    International Nuclear Information System (INIS)

    Schimpfke, T.; Grebner, H.; Peschke, J.; Sievers, J.

    2003-01-01

    In the frame of the German reactor safety research program of the Federal Ministry of Economics and Labour, GRS has started to develop an analysis code named PROST (PRObabilistic STructure analysis) for estimating the leak and break probabilities of piping systems in nuclear power plants. The development is based on the experience achieved with applications of the public available US code PRAISE 3.10 (Piping Reliability Analysis Including Seismic Events), which was supplemented by additional features regarding the statistical evaluation and the crack orientation. PROST is designed to be more flexible to changes and supplementations. Up to now it can be used for calculating fatigue problems. The paper mentions the main capabilities and theoretical background of the present PROST development and presents a parametric study on the influence by changing the method of stress intensity factor and limit load calculation and the statistical evaluation options on the leak probability of an exemplary pipe with postulated axial crack distribution. Furthermore the resulting leak probability of an exemplary pipe with postulated circumferential crack distribution is compared with the results of the modified PRAISE computer program. The intention of this investigation is to show trends. Therefore the resulting absolute values for probabilities should not be considered as realistic evaluations. (author)

  19. To the application of TV and optical equipment for in-service inspection of reactor vessel and primary circuit component materials

    International Nuclear Information System (INIS)

    Afonin, Eh.M.; Bachelis, I.M.; Tokarev, E.A.; Yastrebov, V.E.

    1985-01-01

    Some problems of application of TV and optical equipment for inspection of reactor vessel and primary circuit component materials are considered taking the most widespread WWER-440 type reactor as an example. The most advanrageous objects of the inspection and typical zones of equipment arrangement are shown. Methods and peculiarities of the inspection with the use of TV and optical equipment are presented. Recommendations on rational application of the equipment for the inspection of WWER-440 reactor vessel components are given

  20. Method of providing extended life expectancy for components of boiling water reactors

    International Nuclear Information System (INIS)

    Niedrach, L.W.

    1992-01-01

    This patent describes a containment for a boiling water nuclear reactor, a stainless steel containment, the containment having a deposit of a metal of the platinum group of metals on the surfaces thereof exposed to high temperature, high pressure water of the boiling water nuclear reactor

  1. Research and development of selected components of the high-temperature reactor for process heat generation - results and their application

    International Nuclear Information System (INIS)

    Theymann, W.; Lange, G.

    1989-01-01

    For the process heat supplying high-temperature reactor (PNP) a comprehensive research and development program was performed. Investigations in three fields of the program are reported: heat transfer, gas flow guidance components, and seismic properties of the core structure. Results are presented for the statistics of heat transfer in the core and for heat transfer under operational conditions of a PNP-plant. Further topics are cooling of the side reflector, hot gas mixing in the core bottom region, optimization of inlet flow into the steam generator, and flow tests on a large diameter shut-off valve. Performance tests on hot gas insulations in a special test facility are described as well as tests on connecting elements for coaxial ducts. The measured data on dynamic excitation of the pebble bed with the SAMSON test facility allow an analytical description of the pebble bed core with respect to seismic behaviour. The results of experiments and calculations, using the computer codes CRUNCH-1D and -2D, for seismic excitation of the suspended top reflector are discussed. The seismic tests will be completed in 1989 with the side reflector investigations. A comprehensive seismic verification will then be available. (orig.)

  2. Prospect of Ti-Ni shape memory alloy applied in reactor structures

    International Nuclear Information System (INIS)

    Duan Yuangang

    1995-01-01

    Shape memory effect mechanism, physical property, composition, manufacturing process and application in mechanical structure of Ti-Ni shape memory alloy are introduced. Applications of Ti-Ni shape memory alloy in reactor structure are prospected and some necessary technical conditions of shape memory alloy applied in the reactor structure are put forward initially

  3. On the structural integrity evaluation about aged components

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2013-08-15

    About one third of the nuclear power plants in Japan have been operated more than 30 years and flaws due to age-related degradation mechanisms have been detected in some components such as piping systems or core shrouds these years. Moreover, several severe earthquakes such as the Tohoku District - off the Pacific Ocean Earthquake or the Niigata-ken Chuetsu-oki Earthquake have struck some nuclear power plants in Japan recent years. Therefore, the structural integrity evaluation about nuclear installations and components considering seismic loads and aging mechanisms has become more and more important. In this study, several evaluation methods were proposed to assess the crack growth rate under the seismic loading conditions, to assess the failure conditions or the realistic failure capacities of the aged piping systems considering seismic or general loading conditions. Furthermore, analysis codes were developed considering aging mechanisms to carry out the integrity evaluation, or the failure probability evaluation which is useful in the seismic PSA evaluation. All of these assessment methods and analysis codes are being used and will be used more and more in the cross-check analyses or the safety reviews about nuclear installations and components. (author)

  4. Structure of steel reactor building and construction method therefor

    International Nuclear Information System (INIS)

    Yamakawa, Toshikimi.

    1997-01-01

    The building of the present invention contains a reactor pressure vessel, and has double steel plate walls endurable to elevation of inner pressure and keeping airtightness, and shielding concretes are filled between the double steel plate walls. It also has empty double steel plate walls not filled with concretes and has pipelines, vent ducts, wirings and a support structures for attaching them between the double steel plate walls. It is endurable to a great inner pressure satisfactory and keeps airtightness by the two spaced steel plates. It can be greatly reduced in the weight, and can be manufactured efficiently with high quality in a plant by so called module construction, and the dimension of the entire of the reactor building can be reduced. It is constructed in a dock, transported on the sea while having the space between the two steel plate walls as a ballast tanks, placed in the site, and shielding concretes are filled between the double steel plate walls. The term for the construction can be reduced, and the cost for the construction can be saved. (N.H.)

  5. An analysis of reactor structural response to fuel sodium interaction in a hypothetical core disruptive accident

    International Nuclear Information System (INIS)

    Suzuki, K.; Tashiro, M.; Sasanuma, K.; Nagashima, K.

    1976-01-01

    This study shows the effect of constraints around FSI zone on FSI phenomena and deformations of reactor structures. SUGAR-PISCES code system has been developed to evaluate the phenomena of FSI and the response of reactor structure. SUGAR calculates the phenomena of FSI. PISCES, developed by Physics International Company in U.S.A., calculates the dynamic response of reactor structure in two-dimensional, time-dependent finite-difference Lagrangian model. The results show that the peak pressure and energy by FSI and the deformation of reactor structures are about twice in case of FSI zone surrounding by blanket than by coolant. The FSI phenomena highly depend on the reactor structure and the realistic configuration around core must be considered for analyzing hypothetical core disruptive accident. This work was supported by a grant from Power Reactor and Nuclear Fuel Development Corporation. (auth.)

  6. Reactor

    International Nuclear Information System (INIS)

    Fujibayashi, Toru.

    1976-01-01

    Object: To provide a boiling water reactor which can enhance a quake resisting strength and flatten power distribution. Structure: At least more than four fuel bundles, in which a plurality of fuel rods are arranged in lattice fashion which upper and lower portions are supported by tie-plates, are bundled and then covered by a square channel box. The control rod is movably arranged within a space formed by adjoining channel boxes. A spacer of trapezoidal section is disposed in the central portion on the side of the channel box over substantially full length in height direction, and a neutron instrumented tube is disposed in the central portion inside the channel box. Thus, where a horizontal load is exerted due to earthquake or the like, the spacers come into contact with each other to support the channel box and prevent it from abnormal vibrations. (Furukawa, Y.)

  7. Structural Components of Synaptic Plasticity and Memory Consolidation

    Science.gov (United States)

    Bailey, Craig H.; Kandel, Eric R.; Harris, Kristen M.

    2015-01-01

    Consolidation of implicit memory in the invertebrate Aplysia and explicit memory in the mammalian hippocampus are associated with remodeling and growth of preexisting synapses and the formation of new synapses. Here, we compare and contrast structural components of the synaptic plasticity that underlies these two distinct forms of memory. In both cases, the structural changes involve time-dependent processes. Thus, some modifications are transient and may contribute to early formative stages of long-term memory, whereas others are more stable, longer lasting, and likely to confer persistence to memory storage. In addition, we explore the possibility that trans-synaptic signaling mechanisms governing de novo synapse formation during development can be reused in the adult for the purposes of structural synaptic plasticity and memory storage. Finally, we discuss how these mechanisms set in motion structural rearrangements that prepare a synapse to strengthen the same memory and, perhaps, to allow it to take part in other memories as a basis for understanding how their anatomical representation results in the enhanced expression and storage of memories in the brain. PMID:26134321

  8. Structure and creep of Russian reactor steels with a BCC structure

    Science.gov (United States)

    Sagaradze, V. V.; Kochetkova, T. N.; Kataeva, N. V.; Kozlov, K. A.; Zavalishin, V. A.; Vil'danova, N. F.; Ageev, V. S.; Leont'eva-Smirnova, M. V.; Nikitina, A. A.

    2017-05-01

    The structural phase transformations have been revealed and the characteristics of the creep and long-term strength at 650, 670, and 700°C and 60-140 MPa have been determined in six Russian reactor steels with a bcc structure after quenching and high-temperature tempering. Creep tests were carried out using specially designed longitudinal and transverse microsamples, which were fabricated from the shells of the fuel elements used in the BN-600 fast neutron reactor. It has been found that the creep rate of the reactor bcc steels is determined by the stability of the lath martensitic and ferritic structures in relation to the diffusion processes of recovery and recrystallization. The highest-temperature oxide-free steel contains the maximum amount of the refractory elements and carbides. The steel strengthened by the thermally stable Y-Ti nanooxides has a record high-temperature strength. The creep rate at 700°C and 100 MPa in the samples of this steel is lower by an order of magnitude and the time to fracture is 100 times greater than that in the oxide-free reactor steels.

  9. Use of principal components analysis and three-dimensional atmospheric-transport models for reactor-consequence evaluation

    International Nuclear Information System (INIS)

    Gudiksen, P.H.; Walton, J.J.; Alpert, D.J.; Johnson, J.D.

    1982-01-01

    This work explores the use of principal components analysis coupled to three-dimensional atmospheric transport and dispersion models for evaluating the environmental consequences of reactor accidents. This permits the inclusion of meteorological data from multiple sites and the effects of topography in the consequence evaluation; features not normally included in such analyses. The technique identifies prevailing regional wind patterns and their frequencies for use in the transport and dispersion calculations. Analysis of a hypothetical accident scenario involving a release of radioactivity from a reactor situated in a river valley indicated the technique is quite useful whenever recurring wind patterns exist, as is often the case in complex terrain situations. Considerable differences were revealed in a comparison with results obtained from a more conventional Gaussian plume model using only the reactor site meteorology and no topographic effects

  10. Flaw evaluation methodology for class 2, 3 components in light water reactors

    International Nuclear Information System (INIS)

    Miura, Naoki; Kashima, Koichi; Miyazaki, Katsumasa; Hasegawa, Kunio; Oritani, Naohiko

    2006-01-01

    It is quite important to validate the structural integrity of operating plant components as aged LWR plants are gradually increasing in Japan. The rules on fitness-for-service for nuclear power plants constituted by the JSME provides flaw evaluation methodology. They are mainly focused on Class 1 components, while flaw evaluation criteria for Class 2, 3 components are not consolidated. As such, they also required from the viewpoints of in-service inspection request, reduction of operating cost and systematization of consistent code/standard. In this study, basic concept of flaw evaluation for Class 2, 3 piping was considered, and it is concluded that the same evaluation procedure as Class 1 piping in the current rules is applicable. Some technical issues on practical flaw evaluation for Class 2, 3 piping were listed up, and a countermeasure for each issue was devised. Especially, both allowable flaw sizes in acceptance standards and critical flaw sizes in acceptance criteria have to be determined in consideration of degraded fracture toughness. (author)

  11. Final Report: Safety of Plasma Components and Aerosol Transport During Hard Disruptions and Accidental Energy Release in Fusion Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Bourham, Mohamed A.; Gilligan, John G.

    1999-08-14

    Safety considerations in large future fusion reactors like ITER are important before licensing the reactor. Several scenarios are considered hazardous, which include safety of plasma-facing components during hard disruptions, high heat fluxes and thermal stresses during normal operation, accidental energy release, and aerosol formation and transport. Disruption events, in large tokamaks like ITER, are expected to produce local heat fluxes on plasma-facing components, which may exceed 100 GW/m{sup 2} over a period of about 0.1 ms. As a result, the surface temperature dramatically increases, which results in surface melting and vaporization, and produces thermal stresses and surface erosion. Plasma-facing components safety issues extends to cover a wide range of possible scenarios, including disruption severity and the impact of plasma-facing components on disruption parameters, accidental energy release and short/long term LOCA's, and formation of airborne particles by convective current transport during a LOVA (water/air ingress disruption) accident scenario. Study, and evaluation of, disruption-induced aerosol generation and mobilization is essential to characterize database on particulate formation and distribution for large future fusion tokamak reactor like ITER. In order to provide database relevant to ITER, the SIRENS electrothermal plasma facility at NCSU has been modified to closely simulate heat fluxes expected in ITER.

  12. Comprehending the structure of a vacuum vessel and in-vessel components of fusion machines. 2. Comprehending the divertor structure

    International Nuclear Information System (INIS)

    Suzuki, Satoshi; Akiba, Masato; Saito, Masakatsu

    2006-01-01

    Divertor is given the largest heat load in the in-vessel components of fusion machine. The functions and conditions of divertor are stated from the point of view of thermal and structural dynamics. The way of thinking of structure design of divertor of JT-60 and the ITER (International Thermonuclear Experimental Reactor) is explained. As the conditions of divertor, the materials for large heat load, heat removal, pressure boundary, control of damage, and thermal stress/strain are considered. The divertor has to be changed periodically. The materials are required the heat removal function for high heat load. CuCrZr will be used to cooling tube and heat sink, and CFC materials for the surface. The cross section of ITER, a part of divertor, heat load of divertor and other components, the thermal conductivity of CFC and metal materials, conditions of cooling water for divertor of BWR, PWR and ITER, the thermal stress produced on rod, vertical target of ITER, structure of cooling tube, distribution of temperature and critical heart flux of inner wall of cooling tube, and fatigue clack of cooling tube are shown. (S.Y.)

  13. Analysis of soil-structure interaction and floor response spectrum of reactor building for China advanced research reactor

    International Nuclear Information System (INIS)

    Rong Feng; Wang Jiachun; He Shuyan

    2006-01-01

    Analysis of Soil-Structure Interaction (SSI) and calculation of Floor Response Spectrum (FRS) is substantial for anti-seismic design for China Advanced Research Reactor (CARR) project. The article uses direct method to analyze the seismic reaction of the reactor building in considering soil-structure interaction by establishing two-dimensional soil-structure co-acting model for analyzing and inputting of seismic waves from three directions respectively. The seismic response and floor response spectrum of foundation and floors of the building under different cases have been calculated. (authors)

  14. Scheme of database structure on decommissioning of the research reactor

    International Nuclear Information System (INIS)

    Park, H. S.; Park, S. K.; Kim, H. R.; Lee, D. K.; Jung, K. J.

    2001-01-01

    ISP (Information Strategy Planning), which is the first step of the whole database development, has been studied to manage effectively information and data related to the decommissioning activities of the Korea Research Reactor 1 and 2 (KRR-1 and 2). Since Korea has not acquired the technology of the decommissioning database management system, some record management system (RMS) of large nuclear facilities of national experience such as in the U.S.A, Japan, Belgium, and Russian were reviewed. In order to construct the database structure of the whole decommissioning activities such as the working information, radioactive waste treatment, and radiological surveying and analysis has been extracted from the whole dismantling process. These information and data will be used as the basic data to analyzed the matrix to find the entity relationship diagram and will contribute to the establishment of a business system design and the development of a decommissioning database system as well

  15. Mechanics of structures and maintenance of pressurized water reactors

    International Nuclear Information System (INIS)

    Hutin, J.P.

    1992-01-01

    Electricite de France nowadays has in operation 34 units of 900 MW and 17 units of 1300 MW of PWR. Since the first unit was run, this means that more than 350 reactor-years have been performed, to which should be added the experience already gained on fossil fuel or natural uranium plants. This enabled EDF to build its own philosophy and a strategy for maintenance that are best suited for the specific requirements of the hardware with which the actual nuclear boilers are made-up. This philosophy and strategy rest upon an analysis which calls widely for the mechanics of structures, to such an extent that major decisions concerning maintenance depend on the ability that one has for resolving problems within the scope of that discipline

  16. Structural Component Fabrication and Characterization of Advanced Radiation Resistant ODS Steel for Next Generation Nuclear Systems

    International Nuclear Information System (INIS)

    Noh, Sang Hoon; Kim, Young Chun; Jin, Hyun Ju; Choi, Byoung Kwon; Kang, Suk Hoon; Kim, Tae Kyu

    2016-01-01

    In a sodium-cooled fast reactor (SFR), the coolant outlet temperature and peak temperature of the fuel cladding tube will be about 545 .deg. C and 700 .deg. C with 250 dpa of a very high neutron dose rate. To realize this system, it is necessary to develop an advanced structural material having high creep and irradiation resistance at high temperatures. Austenitic stainless steel may be one of the candidates because of good strength and corrosion resistance at the high temperatures, however irradiation swelling severely occurred to 120dpa at high temperatures and this eventually leads to a decrease of the mechanical properties and dimensional stability. Advanced radiation resistant ODS steel (ARROS) has been newly developed for the in-core structural components in SFR, which has very attractive microstructures to achieve both superior creep and radiation resistances at high temperatures [4]. Nevertheless, the use of ARROS as a structural material essentially requires the fabrication technology development for component parts such as sheet, plate and tube. In this study, plates and tubes were tentatively fabricated with a newly developed alloy, ARROS. Microstructures as well as mechanical properties were also investigated to determine the optimized condition of the fabrication processes.

  17. Present activities for the preparation of a Japanese draft of structural design guidelines for the experimental fusion reactor

    International Nuclear Information System (INIS)

    Miya, K.; Muto, Y.; Takatsu, H.; Hada, K.; Koizumi, K.; Jitsukawa, S.; Arai, T.; Ohkawa, Y.; Shimakawa, T.; Aoto, K.; Shiraishi, H.; Takagi, T.; Miki, N.; Takahashi, S.; Sato, K.; Takemasa, F.; Kasaba, M.; Kudough, F.; Fujita, J.; Kajiura, S.; Kinoshita, S.

    1996-01-01

    Since November 1990, systematic research has been carried out in preparation for a Japanese draft of structural design guidelines for the experimental fusion reactor. This report summarizes the major results of the work and the status of these efforts. A classification of components and definition of operating conditions are proposed on the basis of the ITER-CDA design, in the light of the safety characteristics of the fusion reactor and relevant conventions for the existing fission reactor design code. Specific issues regarding the structural design of the experimental fusion reactor are discussed based on the experimental and analytical work. The validity of the existing structural design method is confirmed for the use of irradiated 316 SS, irrespective of the significant reduction in uniform elongation capability caused by heavy neutron irradiation. Further important phenomena are treated such as magnetic damping, magnetic stiffness and fracture due to electromagnetic forces. Finally, the issues concerned with welding and non-destructive examinations are discussed with relevance to component classification. (orig.)

  18. Conceptual design and technology development of containment structure in Fusion Experimental Reactor (FER)

    International Nuclear Information System (INIS)

    Nishio, Satoshi; Sato, Keisuke; Matsuoka, Fushiki; Kanamori, Naokazu; Koizumi, Koichi; Abe, Tetsuya; Hosobuchi, Hideo; Tada, Eisuke; Yamada, Masao.

    1991-05-01

    A conceptual design of FER (Fusion Experimental Reactor) containment structure and its associated R and D activities, conducted from '89 to '90, are described. The FER containment structure system which mainly consists of a vacuum vessel, shielding structures, in-vessel replaceable components, ports, a cooling pipe system, has been developed to fullfil the required function. As an initial stage of R and D activities, the elemental technologies common to a tokamak reactor have been developed. Among them, a locking mechanism for supporting in-vessel replaceable components and a technique for insulation/conduction are described. For the locking mechanism, a caulking cotter driven by hydraulic pressure has been employed. Three kinds of hydraulic driving mechanism have been manufactured by trial: a 'piston jack' type, a 'bellows' type and a 'flexible tube' type. In the latter type, the stroke is obtained by changing the cross section of the flexible tube from a flat racetrack shape to a fat shape by hydraulic pressure. As the result of preliminary performance test, the shape of 'flexible tube' has been found to be improved. For the insulation coating, Al 2 O 3 has been selected as the material and a plasma spray method has been applied as the coating procedure. For the conduction coating, Cr 3 C 2 has been selected as the material and JET-KOTE method has been applied as the coating procedure. Both methods have been successfully developed and have been confirmed to be applicable the actual machine. A one fifth scale model has been fabricated in order to verify the design feasibility, mainly geometrical consistency. Then some design modifications were found to be needed for some of the components based on the manufacturing experience. (author)

  19. Modular fabrication and characterization of complex silicon carbide composite structures Advanced Reactor Technologies (ART) Research Final Report (Feb 2015 – May 2017)

    Energy Technology Data Exchange (ETDEWEB)

    Khalifa, Hesham [General Atomics, San Diego, CA (United States)

    2017-08-03

    Advanced ceramic materials exhibit properties that enable safety and fuel cycle efficiency improvements in advanced nuclear reactors. In order to fully exploit these desirable properties, new processing techniques are required to produce the complex geometries inherent to nuclear fuel assemblies and support structures. Through this project, the state of complex SiC-SiC composite fabrication for nuclear components has advanced significantly. New methods to produce complex SiC-SiC composite structures have been demonstrated in the form factors needed for in-core structural components in advanced high temperature nuclear reactors. Advanced characterization techniques have been employed to demonstrate that these complex SiC-SiC composite structures provide the strength, toughness and hermeticity required for service in harsh reactor conditions. The complex structures produced in this project represent a significant step forward in leveraging the excellent high temperature strength, resistance to neutron induced damage, and low neutron cross section of silicon carbide in nuclear applications.

  20. Oxidation damage evaluation by non-destructive method for graphite components in high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Shibata, Taiju; Tada, Tatsuya; Sumita, Junya; Sawa, Kazuhiro

    2008-01-01

    To develop non-destructive evaluation methods for oxidation damage on graphite components in High Temperature Gas-cooled Reactors (HTGRs), the applicability of ultrasonic wave and micro-indentation methods were investigated. Candidate graphites, IG-110 and IG-430, for core components of Very High Temperature Reactor (VHTR) were used in this study. These graphites were oxidized uniformly by air at 500degC. The following results were obtained from this study. (1) Ultrasonic wave velocities with 1 MHz can be expressed empirically by exponential formulas to burn-off, oxidation weight loss. (2) The porous condition of the oxidized graphite could be evaluated with wave propagation analysis with a wave-pore interaction model. It is important to consider the non-uniformity of oxidized porous condition. (3) Micro-indentation method is expected to determine the local oxidation damage. It is necessary to assess the variation of the test data. (author)