Numerical analysis of reactor internals under hydrodynamic loads

Energy Technology Data Exchange (ETDEWEB)

Kim, Da Hye; Chang, Yoon Suk [Kyung Hee Univ., Yongin (Korea, Republic of); Jhung, Myung Jo [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

2013-10-15

In the present study, six kinds of major equipment of a typical reactor internals were identified by incorporating recent research trend. Based on this, detailed numerical models were developed and used for establishment of optimum analysis methodology subjected to hydrodynamic loads. As a result, stress values of the major equipment were calculated through the acoustic-structure analysis under periodic hydrodynamic load and the turbulence-structure analysis under random hydrodynamic load. The numerical analysis scheme can be used for development of preventive action plan and management procedures of the reactor internals. Reactor internals installed in a pressure vessel have been exposed to harsh environment such as high neutron irradiation and temperature with complex fluid flow. As the increase of operational years of NPPs(Nuclear Power Plants), possibility of functional loss of the reactor internals is increased due to degradation caused by radiation embrittlement, thermal aging, fatigue, corrosion and FIV(Flow-Induced Vibration) etc. In practice, defects were detected at core support structure as well as upper and lower parts of structural assembly in European and United States NPPs. Recently, in a GALL(Generic Aging Lessons Learned) report, US NRC(Nuclear Regulatory Commission) identified reactor internals as a high priority component and addressed relevant management programs. In Korea, similar activities have been conducted for long-term operation beyond design lifetime but most of them were limited to qualitative evaluation based on examination and maintenance programs. Therefore, not only to reduce repair and replacement efforts but also to secure the stability of NPPs, necessity for development of quantitative evaluation technique as well as establishment of preventive action plan and management procedures is on the rise. The FIV represents the structural vibration phenomenon induced by liquid flow and generally occurs at contact surfaces. In the present

Reliability tests for reactor internals rejuvenation technology

International Nuclear Information System (INIS)

Fujimaki, Katsumi; Hitoki, Yoichi; Otsubo, Toru; Uchiyama, Junichi

1998-01-01

Structural damage due to aging degradation of LWR reactor internals has been reported in several nuclear plants. NUPEC has started a project to test the reliability of the technology for rejuvenating reactor internals which has been funded by the Ministry of International Trade and Industry (MITI) of Japan since 1995. The project follows the policy of a report that the MITI has formally issued in April 1996 summarizing the countermeasures to be considered for aging nuclear plants and equipment. This paper gives an outline of the test plans and results which are directed at preventive maintenance before damage and repair after damage for reactor internals aging degradation. The test results for the replacement methods of ICM housing and BWR core shroud have shown that the methods were reliable and the structural integrity was appropriate based on the evaluation. (author)

Reliability tests for reactor internals replacement technology

International Nuclear Information System (INIS)

Fujimaki, K.; Uchiyama, J.; Ohtsubo, T.

2000-01-01

Structural damage due to aging degradation of LWR reactor internals has been reported in several nuclear plants. NUPEC has started a project to test the reliability of the technology for replacing reactor internals, which was directed at preventive maintenance before damage and repair after damage for the aging degradation. The project has been funded by the Ministry of International Trade and Industry (MITI) of Japan since 1995, and it follows the policy of a report that the MITI has formally issued in April 1996 summarizing the countermeasures to be considered for aging nuclear plants and equipment. This paper gives an outline of the whole test plans and the test results for the BWR reactor internals replacement methods; core shroud, ICM housing, and CRD Housing and stub tube. The test results have shown that the methods were reliable and the structural integrity was appropriate based on the evaluation. (author)

Introduction to reactor internal materials for pressurized water reactor

Energy Technology Data Exchange (ETDEWEB)

Ryu, Woo Suk; Hong, Joon Hwa; Jee, Se Hwan; Lee, Bong Sang; Kuk, Il Hyun [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1994-06-01

This report reviewed the R and D states of reactor internal materials in order to be a reference for researches and engineers who are concerning on localization of the materials in the field or laboratory. General structure of PWR internals and material specification for YGN 3 and 4 were reviewed. States-of-arts on R and D of stainless steel and Alloy X-750 were reviewed, and degradation mechanisms of the components were analyzed. In order to develop the good domestic materials for reactor internal, following studies would be carried out: microstructure, sensitization behavior, fatigue property, irradiation-induced stress corrosion cracking/radiation-induced segregation, radiation embrittlement. (Author) 7 refs., 14 figs., 5 tabs.,.

Introduction to reactor internal materials for pressurized water reactor

International Nuclear Information System (INIS)

Ryu, Woo Suk; Hong, Joon Hwa; Jee, Se Hwan; Lee, Bong Sang; Kuk, Il Hyun

1994-06-01

This report reviewed the R and D states of reactor internal materials in order to be a reference for researches and engineers who are concerning on localization of the materials in the field or laboratory. General structure of PWR internals and material specification for YGN 3 and 4 were reviewed. States-of-arts on R and D of stainless steel and Alloy X-750 were reviewed, and degradation mechanisms of the components were analyzed. In order to develop the good domestic materials for reactor internal, following studies would be carried out: microstructure, sensitization behavior, fatigue property, irradiation-induced stress corrosion cracking/radiation-induced segregation, radiation embrittlement. (Author) 7 refs., 14 figs., 5 tabs.,

Results of assembly test of HTTR reactor internals

International Nuclear Information System (INIS)

Maruyama, S.; Saikusa, A.; Shiozawa, S.; Tsuji, N.; Miki, T.

1996-01-01

The assembly test of the HTTR actual reactor internals had been carried out at the works, prior to their installation in the actual reactor pressure vessel(RPV) at the construction site. The assembly test consists of several items such as examining fabricating precision of each component and alignment of piled-up structures, measuring circumferential coolant velocity profile in the passage between the simulated RPV and the reactor internals as well as under the support plates, measuring by-pass flow rate through gaps between the reactor internals, and measuring the binding force of the core restraint mechanism. Results of the test showed good performance of the HTTR reactor internals. Installation of the reactor internals in the actual RPV was started at the construction site of HTTR in April, 1995. In the installation process, main items of the assembly test at the works were repeated to investigate the reproducibility of installation. (author). 5 refs, 11 figs

INTRANS. A computer code for the non-linear structural response analysis of reactor internals under transient loads

International Nuclear Information System (INIS)

Ramani, D.T.

1977-01-01

The 'INTRANS' system is a general purpose computer code, designed to perform linear and non-linear structural stress and deflection analysis of impacting or non-impacting nuclear reactor internals components coupled with reactor vessel, shield building and external as well as internal gapped spring support system. This paper describes in general a unique computational procedure for evaluating the dynamic response of reactor internals, descretised as beam and lumped mass structural system and subjected to external transient loads such as seismic and LOCA time-history forces. The computational procedure is outlined in the INTRANS code, which computes component flexibilities of a discrete lumped mass planar model of reactor internals by idealising an assemblage of finite elements consisting of linear elastic beams with bending, torsional and shear stiffnesses interacted with external or internal linear as well as non-linear multi-gapped spring support system. The method of analysis is based on the displacement method and the code uses the fourth-order Runge-Kutta numerical integration technique as a basis for solution of dynamic equilibrium equations of motion for the system. During the computing process, the dynamic response of each lumped mass is calculated at specific instant of time using well-known step-by-step procedure. At any instant of time then, the transient dynamic motions of the system are held stationary and based on the predicted motions and internal forces of the previous instant. From which complete response at any time-step of interest may then be computed. Using this iterative process, the relationship between motions and internal forces is satisfied step by step throughout the time interval

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

Reactor building with internal structure of which the movements are independent of those of the general raft and process for building these internal structures

International Nuclear Information System (INIS)

Hista, J.C.

1982-01-01

This reactor building includes a containment enclosure for the internal structures composed of a slab wedged on its periphery against the containment enclosure gusset and resting on the general raft by means of a peripheral bearing ring, a compressible layer being provided between the general raft and the slab [fr

Remote Inspection Techniques for Reactor Internals of Liquid Metal Reactor by using Ultrasonic Waveguide Sensor

International Nuclear Information System (INIS)

Joo, Young Sang; Kim, Seok Hun; Lee, Jae Han

2006-02-01

The primary components such as a reactor core, heat exchangers, pumps and internal structures 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 and continuous monitoring as major in-service inspection (ISI) methods of reactor internal structures. Reactor core and internal structures of LMR can not be visually examined due to an opaque liquid sodium. The under-sodium viewing and remote inspection techniques by using an ultrasonic wave should be applied for the in-service inspection of reactor internals. The remote inspection techniques using ultrasonic wave have been developed and applied for the visualization and ISI of reactor internals. The under sodium viewing technique has a limitation for the application of LMR due to the high temperature and irradiation environment. In this study, an ultrasonic waveguide sensor with a strip plate has been developed for an application to the under-sodium viewing and remote inspection. The Lamb wave propagation of a waveguide sensor has been analyzed and the zero-order antisymmetric A 0 plate wave was selected as the application mode of the sensor. The A 0 plate wave can be propagated in the dispersive low frequency range by using a liquid wedge clamped to the waveguide. A new technique is presented which is capable of steering the radiation beam angle of a waveguide sensor without a mechanical movement of the sensor assembly. The steering function of the ultrasonic radiation beam can be achieved by a frequency tuning method of the excitation pulse in the dispersive range of the A 0 mode. The technique provides an opportunity to overcome the scanning limitation of a waveguide sensor. The beam steering function has been evaluated by an experimental verification. The ultrasonic C-scanning experiments are performed in water and the feasibility of the ultrasonic waveguide sensor has been verified. The various remote inspection

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)

ITER [International Thermonuclear Experimental Reactor] reactor building design study

International Nuclear Information System (INIS)

Thomson, S.L.; Blevins, J.D.; Delisle, M.W.

1989-01-01

The International Thermonuclear Experimental Reactor (ITER) is at the midpoint of a two-year conceptual design. The ITER reactor building is a reinforced concrete structure that houses the tokamak and associated equipment and systems and forms a barrier between the tokamak and the external environment. It provides radiation shielding and controls the release of radioactive materials to the environment during both routine operations and accidents. The building protects the tokamak from external events, such as earthquakes or aircraft strikes. The reactor building requirements have been developed from the component designs and the preliminary safety analysis. The equipment requirements, tritium confinement, and biological shielding have been studied. The building design in progress requires continuous iteraction with the component and system designs and with the safety analysis. 8 figs

Degradation of fastener in reactor internal of PWR

Energy Technology Data Exchange (ETDEWEB)

Kim, D. W.; Ryu, W. S.; Jang, J. S.; Kim, S. H.; Kim, W. G.; Chung, M. K.; Han, C. H

2000-03-01

Main component degraded in reactor internal structure of PWR is fastener such as bolts, stud, cap screw, and pins. The failure of these components may damage nuclear fuel and limits the operation of nuclear reactor. In foreign reactors operated more than 10 years, an increasing number of incidents of degraded thread fasteners have been reported. The degradation of these components impair the integrity of reactor internal structure and limit the life extension of nuclear power plant. To solve the problem of fastener failure, the incidents of failure and main mechanisms should be investigated. the purpose of this state-of-the -art report is to investigate the failure incidents and mechanisms of fastener in foreign and domestic PWR and make a guide to select a proper materials. There is no intent to describe each event in detail in this report. This report covers the failures of fastener and damage mechanisms reported by the licensees of operating nuclear power plants and the applications of plants constructed after 1964. This information is derived from pertinent licensee event report, reportable occurrence reports, operating reactor event memoranda, failure analysis reports, and other relevant documents. (author)

Vibration analysis of reactor assembly internals for Prototype Fast Breeder Reactor

International Nuclear Information System (INIS)

Chellapandi, P.; Jalaldeen, S.; Srinivasan, R.; Chetal, S.C.; Bhoje, S.B.

2003-01-01

Vibration analysis of the reactor assembly components of 500 MWe Prototype Fast Breeder Reactor (PFBR) is presented. The vibration response of primary pump as well as dynamic forces developed at its supports are predicted numerically. The stiffness properties of hydrostatic bearing are determined by formulating and solving governing fluid and structural mechanics equations. The dynamic forces exerted by pump are used as input data for the dynamic response of reactor assembly components, mainly inner vessel, thermal baffle and control plug. Dynamic response of reactor assembly components is also predicted for the pressure fluctuations caused by sodium free level oscillations. Thermal baffle (weir shell) which is subjected to fluid forces developed at the associated sodium free levels is analysed by formulating and solving a set of non-linear equations for fluids, structures and fluid structure interaction (FSI). The control rod drive mechanism is analysed for response under flow induced forces on the parts subjected to cross flow in the zone just above the core top, taking into account FSI between sheaths of control and safety rod and absorber pin bundle. Based on the analysis results, it is concluded that the reactor assembly internals are free from any risk of mechanical as well as flow induced vibrations. (author)

Transactions of the 8th International Conference on Structure Mechanics in Reactor Technology

International Nuclear Information System (INIS)

Browzin, B.S.

1985-06-01

These Transactions of the JK-panel session include preprints of papers or abstracts which are listed in Volume A, ''Introduction, General Contents, Authors Index,'' Proceedings of the 8th International Conference on Structural Mechanics in Reactor Technology. These papers represent the body of the JK-panel session, ''Status of Research in Structural and Mechanical Engineering for Nuclear Power Plants,'' sponsored by the US Nuclear Regulatory Commission. Additional papers are expected at this session, which will be available at the session. The purpose of publishing these Transactions is to inform the participants of the JK-panel session in advance on the papers to be presented and discussed at the session

Structural characteristics of proposed ITER [International Thermonuclear Experimental Reactor] TF [toroidal field] coil conductor

International Nuclear Information System (INIS)

Gibson, C.R.; Miller, J.R.

1988-01-01

This paper analyzes the effect of transverse loading on a cable-in-conduit conductor which has been proposed for the toroidal field coils of the International Thermonuclear Experimental Reactor. The primary components of this conductor are a loose cable of superconducting wires, a thin-wall tube for helium containment, and a U-shaped structural channel. A method is given where the geometry of this conductor can be optimized for a given set of operating conditions. It is shown, using finite-element modeling, that the structural channel is effective in supporting loads due to transverse forces and internal pressure. In addition, it is shown that the superconducting cable is effectively shielded from external transverse loads that might otherwise degrade its current carrying capacity. 10 refs., 10 figs., 3 tabs

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)

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.

HTGR Metallic Reactor Internals Core Shell Cutting & Machining Antideformation Technique Study

International Nuclear Information System (INIS)

Xing Huiping; Xue Song

2014-01-01

The reactor shell assembly of HTGR nuclear power station demonstration project metallic reactor internals is key components of reactor, remains with high-precision large component with large-sized thin-walled straight cylinder-shaped structure, and is the first manufacture in China. As compared with other reactor shell, it has a larger ID (Φ5360mm), a longer length (19000mm), a smaller wall thickness (40mm) and a higher precision requirement. During the process of manufacture, the deformation due to cutting & machining will directly affect the final result of manufacture, the control of structural deformation and cutting deformation shall be throughout total manufacture process of such assembly. To realize the control of entire core shell assembly geometry, the key is to innovate and make breakthroughs on anti-deformation technique and then provide reliable technological foundations for the manufacture of HTGR metallic reactor internals. (author)

Analysis of the fluid-structure dynamic interaction of reactor pressure vessel internals during blowdown

International Nuclear Information System (INIS)

Schlechtendahl, E.G.; Krieg, R.; Schumann, U.

1977-01-01

The loadings on reactor internal structures (in particular the core barrel) induced during a PWR-blowdown must not result in excessive stresses and strains. The deformations are strongly influenced by the coupling of fluid and structure dynamics and it is necessary, therefore, to develop and apply new coupled analysis tools. In this paper a survey is given over work currently in progress in the Nuclear Research Center Karlsruhe and the Los Alamos Scientific Laboratory which aim towards 'best estimate codes'. The new methods will be verified by means of the HDR-blowdown tests and other experiments. The results of several scoping calculations are presented and illustrated by movie films. (orig.) [de

Nuclear fast neutron reactor cooled by a liquid metal and of which internal structures are equipped with a thermal protection device

International Nuclear Information System (INIS)

Lemercier, G.; Lions, N.

1986-01-01

The internal structures of a nuclear fast neutron reactor are covered at least partially, on the most hot side, by a thermal protection device. This device comprises modular plates arranged end to end with a certain play between themselves and taking approximately the shape of the internal structures. Each plate is fixed in its center on the internal structures by a stud. A small plate fixed at one of the corners of each plate and covering partially the adjacent plates ensures the safety fixing of these ones [fr

Development of cutting technique of reactor core internals by CO laser

International Nuclear Information System (INIS)

Takano, G.; Beppu, S.; Matsumoto, O.; Sakamoto, N.; Onozawa, T.; Sugihara, M.; Miya, K.

1995-01-01

The CO laser is superior in the absorption characteristic to materials to the CO 2 laser due to its shorter wavelength. In consideration of this characteristic Nuclear Power Engineering Corporation is studying this applicability sponsored by the Ministry of International Trade Industry of Japan to cutting of reactor core internals of commercial nuclear power plant. In decommissioning of reactor core internals it is necessary to cut stainless steel plates of 305 mm thick. The authors cut stainless steel plates of up to 310mm thick in air and those of up to 150 mm thick underwater with a 20kW class laser. Further, models simulating key structural elements of PWR core internals were cut and secondary products to clarify the applicability of the CO laser cutting to reactor core internals were evaluated. (author)

International breeder reactor development

International Nuclear Information System (INIS)

Traube, K.

1976-01-01

For more than a decade, sodium cooled breeder reactors have now been in the focus of advanced nuclear power development in the major industrialized countries. In the sixties, a total of seven small experimental nuclear power stations were commissioned. Two of these have been shut down in the meantime, the others continue to work satisfactorily, their main purpose being the development of fuel elements. The years 1972-1974 saw the commissioning of the prototype power stations in the 300 MWe power category in France, the United Kingdom and the Soviet Union. Presently, other experimental reactors are under construction in the Federal Republic of Germany, Italy, Japan, the United States, plus another Soviet 600 MWe prototype reactor and the SNR 300 DeBeNeLux prototype at Kalkar. A comparison of the technological features either implemented or planned in the prototype and experimental power plants and of their fuel elements reveals a remarkable similarity in the basic concepts pursued in different countries. The two types of breeder reactors, viz. the loop and the pool types, show a closer resemblance to each other than do pressurized and boilling water reactors. The growing awareness of administrative problems emerging in the approaching phase of the introduction of large breeder power stations in a number of European countries has recently led to a streamlining effort in the structure of industries and to tentative steps towards international cooperation on a broad basis. (orig.) [de

Elevated temperature design of KALIMER reactor internals accounting for creep and stress-rupture effects

International Nuclear Information System (INIS)

Koo, Gyeong Hoi; Yoo, Bong

2000-01-01

In most LMFBR (Liquid Metal Fast Breed Reactor) design, the operating temperature is very high and the time-dependent creep and stress-rupture effects become so important in reactor structural design. Therefore, unlike with conventional PWR, the normal operating conditions can be basically dominant design loading because the hold time at elevated temperature condition is so long and enough to result in severe total creep ratcheting strains during total service lifetime. In this paper, elevated temperature design of the conceptually designed baffle annulus regions of KALIMER (Korea Advanced Liquid Metal Reactor) reactor internal structures is carried out for normal operating conditions which have the operating temperature 530 deg. C and the total service lifetime of 30 years. For the elevated temperature design of reactor internal structures, the ASME Code Case N-201-4 is used. Using this code, the time-dependent stress limits, the accumulated total inelastic strain during service lifetime, and the creep-fatigue damages are evaluated with the calculation results by the elastic analysis under conservative assumptions. The application procedures of elevated temperature design of the reactor internal structures using ASME code case N-201-4 with the elastic analysis method are described step by step in detail. This paper will be useful guide for actual application of elevated temperature design of various reactor types accounting for creep and stress-rupture effects. (author)

Design study for KALIMER upper internal structure and reactor refueling system

International Nuclear Information System (INIS)

Park, Jin Ho

1996-09-01

The design study for the KALIMER upper internal structure (UIS) and reactor refueling system has been described. Two distinct features are plug-in UIS and extended refueling outage. For the UIS system, the functional, structural and material requirements have been determined and the accommodation approaches to meet these functional requirements described. For the refueling system, the functional, structural, process and I and C (Instrument and Control) requirements have been established and the accommodation approaches for the functional and process requirements described. The impact on plant availability due to extension of the refueling outage has also been investigated. The accommodation approaches for UIS system show that the design concept of the system will satisfy the functional requirements with a few design issues to be resolved, such as UIS plug in/out handling system and cask design. It is also shown that the functional and process requirements of the refueling system are achievable with the design of the IVTM cask and related transfer system and the extended refueling outage has little effect (within 1%) on the plant availability if extra refueling time do not exceed 1 week. 1 refs. (Author)

Design study for KALIMER upper internal structure and reactor refueling system

Energy Technology Data Exchange (ETDEWEB)

Park, Jin Ho [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1996-09-01

The design study for the KALIMER upper internal structure (UIS) and reactor refueling system has been described. Two distinct features are plug-in UIS and extended refueling outage. For the UIS system, the functional, structural and material requirements have been determined and the accommodation approaches to meet these functional requirements described. For the refueling system, the functional, structural, process and I and C (Instrument and Control) requirements have been established and the accommodation approaches for the functional and process requirements described. The impact on plant availability due to extension of the refueling outage has also been investigated. The accommodation approaches for UIS system show that the design concept of the system will satisfy the functional requirements with a few design issues to be resolved, such as UIS plug in/out handling system and cask design. It is also shown that the functional and process requirements of the refueling system are achievable with the design of the IVTM cask and related transfer system and the extended refueling outage has little effect (within 1%) on the plant availability if extra refueling time do not exceed 1 week. 1 refs. (Author).

Generation IV reactors: international projects

International Nuclear Information System (INIS)

Carre, F.; Fiorini, G.L.; Kupitz, J.; Depisch, F.; Hittner, D.

2003-01-01

Generation IV international forum (GIF) was initiated in 2000 by DOE (American department of energy) in order to promote nuclear energy in a long term view (2030). GIF has selected 6 concepts of reactors: 1) VHTR (very high temperature reactor system, 2) GHR (gas-cooled fast reactor system), 3) SFR (sodium-cooled fast reactor system, 4) SCWR (super-critical water-cooled reactor system), 5) LFR (lead-cooled fast reactor system), and 6) MFR (molten-salt reactor system). All these 6 reactor systems have been selected on criteria based on: - a better contribution to sustainable development (through their aptitude to produce hydrogen or other clean fuels, or to have a high energy conversion ratio...) - economic profitability, - safety and reliability, and - proliferation resistance. The 6 concepts of reactors are examined in the first article, the second article presents an overview of the results of the international project on innovative nuclear reactors and fuel cycles (INPRO) within IAEA. The project finished its first phase, called phase-IA. It has produced an outlook into the future role of nuclear energy and defined the need for innovation. The third article is dedicated to 2 international cooperations: MICANET and HTR-TN. The purpose of MICANET is to propose to the European Commission a research and development strategy in order to develop the assets of nuclear energy for the future. Future reactors are expected to be more multiple-purposes, more adaptable, safer than today, all these developments require funded and coordinated research programs. The aim of HTR-TN cooperation is to promote high temperature reactor systems, to develop them in a long term perspective and to define their limits in terms of burn-up and operating temperature. (A.C.)

DESIGN OF A VIBRATION AND STRESS MEASUREMENT SYSTEM FOR AN ADVANCED POWER REACTOR 1400 REACTOR VESSEL INTERNALS COMPREHENSIVE VIBRATION ASSESSMENT PROGRAM

OpenAIRE

KO, DO-YOUNG; KIM, KYU-HYUNG

2013-01-01

In accordance with the US Nuclear Regulatory Commission (US NRC), Regulatory Guide 1.20, the reactor vessel internals comprehensive vibration assessment program (RVI CVAP) has been developed for an Advanced Power Reactor 1400 (APR1400). The purpose of the RVI CVAP is to verify the structural integrity of the reactor internals to flow-induced loads prior to commercial operation. The APR1400 RVI CVAP consists of four programs (analysis, measurement, inspection, and assessment). Thoughtful prepa...

International Thermonuclear Experimental Reactor

International Nuclear Information System (INIS)

Blevins, J.D.; Stasko, R.R.

1989-09-01

An international design team comprised of members from Canada, Europe, Japan, the Soviet Union, and the United States of America, are designing an experimental fusion test reactor. The engineering and testing objectives of this International Thermonuclear Experimental Reactor (ITER) are to validate the design and to demonstrate controlled ignition, extended burn of a deuterium and tritium plasma, and achieve steady state using technology expected to be available by 1990. The concept maximizes flexibility while allowing for a variety of plasma configurations and operating scenarios. During physics phase operation, the machine produces a 22 MA plasma current. In the technology phase, the machine can be reconfigured with a thicker shield and a breeding blanket to operate with an 18 MA plasma current at a major radius of 5.5 meters. Canada's involvement in the areas of safety, facility design, reactor configuration and maintenance builds on our internationally recognized design and operational expertise in developing tritium processes and CANDU related technologies

An Evaluation Report on the High Temperature Design of the KALIMER-600 Reactor Structures

Energy Technology Data Exchange (ETDEWEB)

Park, Chang Gyu; Lee, Jae Han

2007-03-15

This report is on the validity evaluation of high temperature structural design for the reactor structures and piping of the pool-type Liquid Metal Reactor, KALIMER-600 subjected to the high temperature thermal load condition. The structural concept of the Upper Internal Structure located above the core is analyzed and the adequate UIS conceptual design for KALIMER-600 is proposed. Also, the high temperature structural integrity of the thermal liner which is to protect the UIS bottom plate from the high frequency thermal fatigue damage was evaluated by the thermal stripping analysis. The high temperature structural design of the reactor internal structure by considering the reactor startup-shutdown cycle was carried out and the structural integrity of it for a normal operating condition as well as the transient condition of the primary pump trip accident was confirmed. Additionally the structure design of the reactor internal structural was changed to prevent the non-uniform deformation of the primary pump which is induced by the thermal expansion difference between the reactor head and the baffle plate. The arrangement of the IHTS piping system which is a part of the reactor system is carried out and the structural integrity and the accumulated deformation by considering the reactor startup-shutdown cycle of a normal operating condition were evaluated. The structural integrity and the accumulated deformation of the PDRC hot leg piping by considering the PDRC operating condition were evaluated. The validity of KALIMER-600 high temperature structural design is confirmed through this study, and it is clearly found that the methodology research to evaluate the structural integrity considering the reactor life time of 60 years ensured is necessary.

Load bearing capacities and elastic-plastic behavior of reactor vessel internals

International Nuclear Information System (INIS)

Watanabe, Keita; Nagase, Ryuichi

2017-01-01

Radial Support Keys (RSKs) are installed at the bottom of Reactor Vessel Internal (RVI) of Pressurized Water Reactor (PWR) and fit into Core Support Lugs of Reactor Pressure Vessel (RPV). This structure provides reactor core horizontal support and transmits the loads between RVI and RPV. RSK is one of the critical parts of RVI from the view point of earthquake-proof safety. In order to assure the structural integrity of Nuclear Reactor in case of massive earthquake, load bearing capacities of RSK are confirmed by static loading tests with reduced-scale mockups. In addition, collapse loads of actual components calculated by Limit Analyses are conservative enough compared to the load bearing capacities confirmed by the test. Thus, the methodology to calculate collapse load by Limit Analysis is applicable to evaluation of structural integrity for RSK. (author)

Nuclear reactor internals arrangement

International Nuclear Information System (INIS)

Frisch, E.; Andrews, H.N.

1976-01-01

A nuclear reactor internals arrangement is disclosed which facilitates reactor refueling. A reactor vessel and a nuclear core is utilized in conjunction with an upper core support arrangement having means for storing withdrawn control rods therein. The upper core support is mounted to the underside of the reactor vessel closure head so that upon withdrawal of the control rods into the upper core support, the closure head, the upper core support and the control rods are removed as a single unit thereby directly exposing the core for purposes of refueling

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.

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)

International thermal reactor development

International Nuclear Information System (INIS)

Zebroski, E.L.

1977-01-01

The worldwide development of nuclear power plants is reviewed. Charts are presented which show the commitment to light-water reactor capacity construction with breakdown by region and country. Additional charts show the major nuclear research centers which have substantial scope in light water reactor development and extensive international activities

Dismantling of JPDR reactor internals by underwater plasma arc cutting technique using robotic manipulator

International Nuclear Information System (INIS)

Yokota, M.

1988-01-01

The actual dismantling of JPDR started on December 4, 1986. As of now, equipment that surrounds the reactor has mostly been removed to provide working space in reactor containment prior to the dismantling of reactor internals. Some reactor internals have been successfully dismantled using the underwater arc cutting system with a robotic manipulator during the period of January to March 1988. The cutting system is composed of an underwater plasma arc cutting device and a robotic manipulator. The cut off reactor internals were core spray block, feedwater sparger and stabilizers for fuel upper grid tube. The plasma arc cutting device was developed to dismantle the reactor internals underwater. It mainly consists of a plasma torch, power and gas supply systems for the torch, and by-product treatment systems. It has the cutting ability of 130 mm thickness stainless steel underwater. The robotic manipulator has seven degrees of freedom of movement, enabling it to move in almost the same way as the arm of a human being. The arm of the robot is mounted on a supporting device which is suspended by three chains from the support structure set on a service floor. A plasma torch is griped by the robotic hand; its position to the structure to be cut is controlled from a remote control room, about 100 meters outside the reactor containment

Preliminary Design Concept for a Reactor-internal CRDM

International Nuclear Information System (INIS)

Lee, Jae Seon; Kim, Jong Wook; Kim, Tae Wan; Choi, Suhn; Kim, Keung Koo

2013-01-01

A rod ejection accident may cause severer result in SMRs because SMRs have relatively high control rod reactivity worth compared with commercial nuclear reactors. Because this accident would be perfectly excluded by adopting a reactor-internal CRDM (Control Rod Drive Mechanism), many SMRs accept this concept. The first concept was provided by JAERI with the MRX reactor which uses an electric motor with a ball screw driveline. Babcock and Wilcox introduced the concept in an mPower reactor that adopts an electric motor with a roller screw driveline and hydraulic system, and Westinghouse Electric Co. proposes an internal Control Rod Drive in its SMR with an electric motor with a latch mechanism. In addition, several other applications have been reported thus far. The reactor-internal CRDM concept is now widely adopted in many SMR designs, and this concept may also be applied in an evolutionary reactor development. So the preliminary study is conducted based on the SMART CRDM design. A preliminary design concept for a reactor-internal CRDM was proposed and evaluated through an electromagnetic analysis. It was found that there is an optimum design for the motor housing, and the results may contribute to the realization a reactor-internal CRDM for an evolutionary reactor development. More detailed analysis results will be reported later

[International Thermonuclear Experimental Reactor support

International Nuclear Information System (INIS)

Dean, S.O.

1990-01-01

This report summarizes the activities under LLNL Purchase Order B089367, the purpose of which is to ''support the University/Lawrence Livermore National Laboratory Magnetic Fusion Program by evaluating the status of research relative to other national and international programs and assist in long-range plans and development strategies for magnetic fusion in general and for ITER in particular.'' Two specific subtasks are included: ''to review the LLNL Magnet Technology Development Program in the context of the International Thermonuclear Experimental Reactor Design Study'' and to ''assist LLNL to organize and prepare materials for an International Thermonuclear Experimental Reactor Design Study information meeting.''

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)

Remote repairs and refurbishment of reactor internal structures of magnox plant

International Nuclear Information System (INIS)

Barnes, S.A.; Kelly, D.E.

1992-01-01

The original designers of the UK Magnox reactor plant made provision for the then perceived time dependent processes that could have influenced the operational life of the plant. Changes in graphite properties with irradiation, particularly dimensional change, were well understood and in-core samples were provided for subsequent laboratory examination to monitor the processes throughout plant life. The tendency towards embrittlement with irradiation of the steel of the reactor pressure vessels was also acknowledged and again in-core samples were provided for monitoring changes in materials properties in-service and thus provide data in support of structural analyses to sustain the reactor safety cases. (author)

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

Under-Sodium Inspection Techniques for Reactor Internals of KALIMER-600 using Ultrasonic Waveguide Sensor

International Nuclear Information System (INIS)

Joo, Young Sang; Kim, Seok Hoon; Lee, Jae Han

2005-01-01

KALIMER-600 is a pool type liquid metal reactor (LMR) which is operated with a sodium coolant. The reactor internals of KALIMER-600 are submerged in a liquid sodium pool. As the liquid sodium is opaque to the light, a conventional visual inspection can not be used for observing the internal structures under a sodium condition. An under-sodium viewing (USV) technique using an ultrasonic wave should be applied for the observation of the refueling maneuver and the in-service inspection of the reactor internals. Under-sodium inspection technology utilizing ultrasonic waves has been widely developed for a visualization of the reactor core and internal components of LMR. Immersion sensors and waveguide sensors have been applied to the USV inspection. The immersion sensor has a precise imaging capability, but may have high temperature restrictions and an uncertain life. The waveguide sensor has the advantages of simplicity and reliability, but limited in its movement. The new plate-type waveguide sensor has been developed as a useful alternative to immersion sensors for USV applications. In the viewing and monitoring applications, a beam steering function of a waveguide sensor might be required. A new waveguide sensor and technique are being developed to overcome the limitations of a waveguide ultrasonic sensor. In this study, the under-sodium inspection techniques using the newly developed waveguide sensor for the reactor internal structures of KALIMER-600 is proposed

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

Reliability test for reactor internals rejuvenation technology

International Nuclear Information System (INIS)

Uchiyama, Junichi

1998-01-01

41 transparencies were presented on the subject of 'Reliability test for reactor internals rejuvenation technology'. The items presented give an introduction on the management of plant life in Japan and introduce the Nuclear Power Engineering Corporation (NUPEC). The question of what reliability tests for rejuvenation of reactor internals are is discussed in some detail and an outline of each test is given. Altogether six methods to rejuvenate reactor internals are presented, two of which have already been applied to actual plants. The presentation was supported by many detailed drawings and images

Seismic analysis of a PWR 900 reactor: study of reactor building with soil-structure interaction and evaluation of floor spectra

International Nuclear Information System (INIS)

Gantenbein, F.; Aguilar, J.

1983-08-01

The purpose of this paper is the evaluation of seismic response and floor spectra for a typical PWR 900 reactor building with respect to soil-structure interaction for soil stiffness). The typical PWR 900 reactor building consists of a concrete cylindrical external building and roof dome, a concrete internal structure (internals) on a common foundation mat as illustrated. The seismic response is obtained by SRSS method and floor spectra directly from ground spectrum and modal properties of the structure. Seismic responses and floor spectra computation is performed in the case of two different ground spectra: EDF spectrum (mean of oscillator spectra obtained from 8 californian records) normalized to 0.2 g, and DSN spectrum (typical of shallow seism) normalized to 0.3 g. The first section is devoted to internals' modelisation, the second one to the axisymmetric model of the reactor, the third one to the seismic response, the fourth one to floor spectra

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.)

Creep-Fatigue Damage Evaluation of a Model Reactor Vessel and Reactor Internals of Sodium Test Facility according to ASME-NH and RCC-MRx Codes

International Nuclear Information System (INIS)

Lim, Dong-Won; Lee, Hyeong-Yeon; Eoh, Jae-Hyuk; Son, Seok-Kwon; Kim, Jong-Bum; Jeong, Ji-Young

2016-01-01

The objective of the STELLA-2 is to support the specific design approval for PGSFR by synthetic reviews of key safety issues and code validations through the integral effect tests. Due to its high temperature operation in SFRs (and in a testing facility) up to 550 °C, thermally induced creep-fatigue damage is very likely in components including a reactor vessel, reactor internals (interior structures), heat exchangers, pipelines, etc. In this study, structural integrity of the components such as reactor vessel and internals in STELLA-2 has been evaluated against creep-fatigue failures at a concept-design step. As 2D analysis yields far conservative results, a realistic 3D simulation is performed by a commercial software. A design integrity guarding against a creep-fatigue damage failure operating at high temperature was evaluated for the reactor vessel with its internal structure of the STELLA-2. Both the high temperature design codes were used for the evaluation, and results were compared. All the results showed the vessel as a whole is safely designed at the given operating conditions, while the ASME-NH gives a conservative evaluation

Creep-Fatigue Damage Evaluation of a Model Reactor Vessel and Reactor Internals of Sodium Test Facility according to ASME-NH and RCC-MRx Codes

Energy Technology Data Exchange (ETDEWEB)

Lim, Dong-Won; Lee, Hyeong-Yeon; Eoh, Jae-Hyuk; Son, Seok-Kwon; Kim, Jong-Bum; Jeong, Ji-Young [KAERI, Daejeon (Korea, Republic of)

2016-05-15

The objective of the STELLA-2 is to support the specific design approval for PGSFR by synthetic reviews of key safety issues and code validations through the integral effect tests. Due to its high temperature operation in SFRs (and in a testing facility) up to 550 °C, thermally induced creep-fatigue damage is very likely in components including a reactor vessel, reactor internals (interior structures), heat exchangers, pipelines, etc. In this study, structural integrity of the components such as reactor vessel and internals in STELLA-2 has been evaluated against creep-fatigue failures at a concept-design step. As 2D analysis yields far conservative results, a realistic 3D simulation is performed by a commercial software. A design integrity guarding against a creep-fatigue damage failure operating at high temperature was evaluated for the reactor vessel with its internal structure of the STELLA-2. Both the high temperature design codes were used for the evaluation, and results were compared. All the results showed the vessel as a whole is safely designed at the given operating conditions, while the ASME-NH gives a conservative evaluation.

A study on nonlinear behavior of reactor containment structures during ultimate accident condition(I)

Energy Technology Data Exchange (ETDEWEB)

Kim, Sun Hoon; Kim, Young Jin; Park, Joo Yeon [Youngdong Univ., Yeongdong (Korea, Republic of)] (and others)

2003-03-15

In this study, the following scope and contents are established for first year's study of determining ultimate pressure capacity of CANDU-type reactor containment. State-of-arts on the prediction of the ultimate pressure capacity of prestressed concrete reactor containment. Comparative study on structural characteristics and analysis model of CANDU-type reactor containment. State-of-arts on evaluation method of the ultimate pressure capacity of prestressed concrete reactor containment. Enhancement of evaluation method of the ultimate pressure capacity for PWR containment structure. In order to determine a realistic lower bound of a typical reactor containment structural capacity for internal pressure, modelling techniques and analytical investigation to predict its non-linear behavior up to ultimate capacity are required. Especially, the in-depth evaluation of modeling technique and analysis procedure for determining ultimate pressure capacity of CANDU-type reactor containment is required. Therefore, modelling techniques and analytical investigation to predict its non-linear behavior up to ultimate pressure capacity of CANDU-type reactor containment for internal pressure will be suggested in this study.

A study on nonlinear behavior of reactor containment structures during ultimate accident condition(I)

International Nuclear Information System (INIS)

Kim, Sun Hoon; Kim, Young Jin; Park, Joo Yeon

2003-03-01

In this study, the following scope and contents are established for first year's study of determining ultimate pressure capacity of CANDU-type reactor containment. State-of-arts on the prediction of the ultimate pressure capacity of prestressed concrete reactor containment. Comparative study on structural characteristics and analysis model of CANDU-type reactor containment. State-of-arts on evaluation method of the ultimate pressure capacity of prestressed concrete reactor containment. Enhancement of evaluation method of the ultimate pressure capacity for PWR containment structure. In order to determine a realistic lower bound of a typical reactor containment structural capacity for internal pressure, modelling techniques and analytical investigation to predict its non-linear behavior up to ultimate capacity are required. Especially, the in-depth evaluation of modeling technique and analysis procedure for determining ultimate pressure capacity of CANDU-type reactor containment is required. Therefore, modelling techniques and analytical investigation to predict its non-linear behavior up to ultimate pressure capacity of CANDU-type reactor containment for internal pressure will be suggested in this study

An internally illuminated monolith reactor: Pros and cons relative to a slurry reactor

NARCIS (Netherlands)

Carneiro, Joana T.; Carneiro, J.T.; Berger, Rob; Moulijn, Jacob A.; Mul, Guido

2009-01-01

In the present study, kinetic models for the photo-oxidation of cyclohexane in two different photoreactor systems are discussed: a top illumination reactor (TIR) representative of a slurry reactor, and the so-called internally illuminated monolith reactor (IIMR) representing a reactor containing

International seminar on the safety research needs for Russian-designed reactors: material presented at the international seminar 1997

International Nuclear Information System (INIS)

1997-01-01

This seminar on international, national and bilateral cooperation programmes on the safety research needs for Russian-designed reactors was held in Tokyo, Japan (1997) and hosted by the OECD Nuclear Energy Agency and the Science and Technology Agency (STA) of Japan. More than 70 participants attended the seminar. Represented were experts from OECD/NEA member countries and Russia, the International Atomic Energy Agency (IAEA), the ISTC, the INSC and the Russian INSC. Eighteen papers were presented in five sessions. The seminar was structured around four main areas of cooperation: cooperative programmes of the OECD/NEA, programmes of international organisations, bi-lateral programmes, and national programmes of OECD/NEA member countries having reactors of the VVER type. General conclusions, followed by specific technical conclusions are included

Fluid-Induced Vibration Analysis for Reactor Internals Using Computational FSI Method

Energy Technology Data Exchange (ETDEWEB)

Moon, Jong Sung; Yi, Kun Woo; Sung, Ki Kwang; Im, In Young; Choi, Taek Sang [KEPCO E and C, Daejeon (Korea, Republic of)

2013-10-15

This paper introduces a fluid-induced vibration analysis method which calculates the response of the RVI to both deterministic and random loads at once and utilizes more realistic pressure distribution using the computational Fluid Structure Interaction (FSI) method. As addressed above, the FIV analysis for the RVI was carried out using the computational FSI method. This method calculates the response to deterministic and random turbulence loads at once. This method is also a simple and integrative method to get structural dynamic responses of reactor internals to various flow-induced loads. Because the analysis of this paper omitted the bypass flow region and Inner Barrel Assembly (IBA) due to the limitation of computer resources, it is necessary to find an effective way to consider all regions in the RV for the FIV analysis in the future. Reactor coolant flow makes Reactor Vessel Internals (RVI) vibrate and may affect the structural integrity of them. U. S. NRC Regulatory Guide 1.20 requires the Comprehensive Vibration Assessment Program (CVAP) to verify the structural integrity of the RVI for Fluid-Induced Vibration (FIV). The hydraulic forces on the RVI of OPR1000 and APR1400 were computed from the hydraulic formulas and the CVAP measurements in Palo Verde Unit 1 and Yonggwang Unit 4 for the structural vibration analyses. In this method, the hydraulic forces were divided into deterministic and random turbulence loads and were used for the excitation forces of the separate structural analyses. These forces are applied to the finite element model and the responses to them were combined into the resultant stresses.

An easily regenerable enzyme reactor prepared from polymerized high internal phase emulsions

International Nuclear Information System (INIS)

Ruan, Guihua; Wu, Zhenwei; Huang, Yipeng; Wei, Meiping; Su, Rihui; Du, Fuyou

2016-01-01

A large-scale high-efficient enzyme reactor based on polymerized high internal phase emulsion monolith (polyHIPE) was prepared. First, a porous cross-linked polyHIPE monolith was prepared by in-situ thermal polymerization of a high internal phase emulsion containing styrene, divinylbenzene and polyglutaraldehyde. The enzyme of TPCK-Trypsin was then immobilized on the monolithic polyHIPE. The performance of the resultant enzyme reactor was assessed according to the conversion ability of N_α-benzoyl-L-arginine ethyl ester to N_α-benzoyl-L-arginine, and the protein digestibility of bovine serum albumin (BSA) and cytochrome (Cyt-C). The results showed that the prepared enzyme reactor exhibited high enzyme immobilization efficiency and fast and easy-control protein digestibility. BSA and Cyt-C could be digested in 10 min with sequence coverage of 59% and 78%, respectively. The peptides and residual protein could be easily rinsed out from reactor and the reactor could be regenerated easily with 4 M HCl without any structure destruction. Properties of multiple interconnected chambers with good permeability, fast digestion facility and easily reproducibility indicated that the polyHIPE enzyme reactor was a good selector potentially applied in proteomics and catalysis areas. - Graphical abstract: Schematic illustration of preparation of hypercrosslinking polyHIPE immobilized enzyme reactor for on-column protein digestion. - Highlights: • A reactor was prepared and used for enzyme immobilization and continuous on-column protein digestion. • The new polyHIPE IMER was quite suit for protein digestion with good properties. • On-column digestion revealed that the IMER was easy regenerated by HCl without any structure destruction.

An easily regenerable enzyme reactor prepared from polymerized high internal phase emulsions

Energy Technology Data Exchange (ETDEWEB)

Ruan, Guihua, E-mail: guihuaruan@hotmail.com [Guangxi Key Laboratory of Electrochemical and Magnetochemical Functional Materials, College of Chemistry and Bioengineering, Guilin University of Technology, Guangxi 541004 (China); Guangxi Collaborative Innovation Center for Water Pollution Control and Water Safety in Karst Area, Guilin University of Technology, Guilin 541004 (China); Wu, Zhenwei; Huang, Yipeng; Wei, Meiping; Su, Rihui [Guangxi Key Laboratory of Electrochemical and Magnetochemical Functional Materials, College of Chemistry and Bioengineering, Guilin University of Technology, Guangxi 541004 (China); Du, Fuyou, E-mail: dufu2005@126.com [Guangxi Key Laboratory of Electrochemical and Magnetochemical Functional Materials, College of Chemistry and Bioengineering, Guilin University of Technology, Guangxi 541004 (China); Guangxi Collaborative Innovation Center for Water Pollution Control and Water Safety in Karst Area, Guilin University of Technology, Guilin 541004 (China)

2016-04-22

A large-scale high-efficient enzyme reactor based on polymerized high internal phase emulsion monolith (polyHIPE) was prepared. First, a porous cross-linked polyHIPE monolith was prepared by in-situ thermal polymerization of a high internal phase emulsion containing styrene, divinylbenzene and polyglutaraldehyde. The enzyme of TPCK-Trypsin was then immobilized on the monolithic polyHIPE. The performance of the resultant enzyme reactor was assessed according to the conversion ability of N{sub α}-benzoyl-L-arginine ethyl ester to N{sub α}-benzoyl-L-arginine, and the protein digestibility of bovine serum albumin (BSA) and cytochrome (Cyt-C). The results showed that the prepared enzyme reactor exhibited high enzyme immobilization efficiency and fast and easy-control protein digestibility. BSA and Cyt-C could be digested in 10 min with sequence coverage of 59% and 78%, respectively. The peptides and residual protein could be easily rinsed out from reactor and the reactor could be regenerated easily with 4 M HCl without any structure destruction. Properties of multiple interconnected chambers with good permeability, fast digestion facility and easily reproducibility indicated that the polyHIPE enzyme reactor was a good selector potentially applied in proteomics and catalysis areas. - Graphical abstract: Schematic illustration of preparation of hypercrosslinking polyHIPE immobilized enzyme reactor for on-column protein digestion. - Highlights: • A reactor was prepared and used for enzyme immobilization and continuous on-column protein digestion. • The new polyHIPE IMER was quite suit for protein digestion with good properties. • On-column digestion revealed that the IMER was easy regenerated by HCl without any structure destruction.

Seismic Response Analysis of Assembled Reactor Vessel Internals

International Nuclear Information System (INIS)

Je, Sang-Yun; Chang, Yoon-Suk; Kang, Sung-Sik

2015-01-01

RVIs (Reactor Vessel Internals) perform important safe-related functions such as upholding the nuclear fuel assembly as well as providing the coolant passage of the reactor core and supporting the control rod drive mechanism. Therefore, the components including RVIs have to be designed and constructed taking into account the structural integrity under various accident scenarios. The reliable seismic analysis is essentially demanded to maintain the safe-related functions of RVIs. In this study, a modal analysis was performed based on the previous researches to examine values of frequencies, mode shapes and participation factors. Subsequently, the structural integrity respecting to the earthquake was evaluated through a response spectrum analysis by using the output variables of modal analysis. In this study, the structural integrity of the assembled RVIs was carried out against the seismic event via the modal and response spectrum analyses. Even though 287MPa of the maximum stress value occurred at the connected region between UGS and CSA, which was lower than its allowable value. At present, fluid-structure interaction effects are being examined for further realistic simulation, which will be reported in the near future

Fast neutron nuclear reactor with lightened internal structure

International Nuclear Information System (INIS)

Artaud, R.; Aubert, M.; Renaux, C.

1984-01-01

The invention concerns an integrated type fast reactor. The inner vessel comprises two truncated shells, of which the large bases are connected either directly, or by a cylindrical shell of large diameter. The small base of the upper truncated shell is prolongated by a shell of small diameter and the small base of the lower truncated shell supports the reactor core. The invention allows the construction of simpler and less expansive fast reactors [fr

Development of inspection and evaluation guidelines for light water reactor internals

International Nuclear Information System (INIS)

Aoki, T.; Yamashita, H.; Sakai, K.

2002-01-01

Full text: In Japan, before a nuclear power plant reaches its 30 years of operation, the Japanese utilities carry out a 'study on plant life management'. Reflecting the results of that study into the maintenance, the utilities make efforts to maintain and improve the safety and reliability of their nuclear power plants. In this study, all safety related components are evaluated from the viewpoint of aging degradation, assuming a long-term operation. If a crack should be found at components such as reactor internals, which is deemed important for safety and are difficult to repair or replace may provide a serious impact on the plant operation and management. Reactor internals, for instance, made of austenitic stainless steel and nickel base alloy, are not completely free from aging degradation including stress corrosion cracking (SCC). Therefore, it is concluded in the study on plant life management that they are required continuous planned inspections to confirm their integrity while continuing plant operation. If an aging degradation such as SCC should be found at the reactor internals, a great amount of labor and time may be required for root cause investigation and analysis and subsequent repairs because it is very difficult to reach to the degraded portion due to structural, dimensional and environmental restrictions. Therefore, such situations may provide serious impact on plant operation and management. As to the reactor internals and other components with similar characteristics, it is strongly recommended that contingency plan should be prepared in advance. Thus, considering the significance of aging degradation, it is necessary to develop some standard rules for the inspection and evaluation of reactor internals. Such rules should specify when, where and how to inspect. They should also specify the evaluation method in case such degradation as a crack is found, and the repair method and extent if repairs are required. These standard rules must be

Process for dissolving the radioactive corrosion products from internal surfaces in nuclear reactors

International Nuclear Information System (INIS)

Brown, W.W.

1976-01-01

This invention concerns a process for dissolving in the coolant flowing in a reactor the radioactive substances from the corrosion of the internal surfaces of the reactor to which they cling. When a reactor is operating, the fission occurring in the fuel generates gases and fission substances, such as iodine 131 and 133, cesium 134 and 137, molybdenum 99, xenon 133 and activates the structural materials of the reactor such as nickel by giving off cobalt 58 and similar substances. Under this invention an oxygen rich solution is injected in the reactor coolant after the temperature and pressure reduction stage, during the preparation prior to refuelling and repairs. The oxygen in the solution speeds up the release of cobalt 58 and other radioactive substances from the internal surfaces of the reactor and their dissolving in the oxygenated cold coolant at the start of the cooling procedures of the installation. This allows them to be removed by an ion exchanger before the reactor is emptied. By utilising this process, about half a day may be gained in refuelling time when this has to be done once a week [fr

Creating an international forum - The success story of the International Forum for Reactor Aging Management (IFRAM)

International Nuclear Information System (INIS)

Brenchley, D.L.; Bond, L.J.; Carpenter, C.E.; Hwang, I.S.; Martin, O.; Reister, R.; Shoji, T.; Tilley, R.

2012-01-01

As the benefits of extending the safe operating life of nuclear power plants has become more evident, so has the need to increase international cooperation in reactor aging management research. This paper describes how individuals and organizations from Asia, Europe, and North America teamed to create the International Forum for Reactor Aging Management (IFRAM). The mission of IFRAM is to facilitate the appropriate exchange of information among those parties and organizations around the world that are presently, or are planning to, address issues of nuclear power plant (NPP) systems, structures and components aging management. The main purpose of this paper is to describe the steps to success in creating IFRAM. The desired attributes, charter, and operational methods of IFRAM are described and the key contributors for creating IFRAM are identified. (author)

Gamma dose from activation of internal shields in IRIS reactor.

Science.gov (United States)

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

2005-01-01

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

Gamma dose from activation of internal shields in IRIS reactor

International Nuclear Information System (INIS)

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

2005-01-01

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

Lifetime assessment on PWR reactor vessel internals in Korea

International Nuclear Information System (INIS)

Jung, Sung-Gyu; Jin, Tae-Eun; Jeong, Ill-Seok

2002-01-01

In order to extend the operating time of the Kori Unit 1 reactor internals, a comprehensive review of the potential ageing problems and a safety assessment have been performed. As the plant ages, reactor internal components which are subject to various ageing mechanism should be identified and evaluated based on the systematic technical procedure. In this respect, technical procedure for lifetime evaluation had been developed and applied to reactor internals. This paper describes a overall assessment and ageing management procedure and evaluation results for reactor internals. Also this paper suggests the optimal ageing management programs to maintain the integrity of reactor internals beyond design life based on the evaluation results. A review of all known potential ageing mechanisms was performed for each of the reactor internal subcomponents. From these results, 8 ageing mechanisms such as void swelling, irradiation and thermal embrittlement, fatigue, stress corrosion cracking, IASCC, stress relaxation, and wear for the reactor internal components were expected to be of major concerns during the current or extended plant life. In this study, 8 ageing mechanisms were identified for lifetime evaluation. For these ageing mechanisms, lifetime assessment was performed. As a result of this evaluation, it is expected that core barrel will exceed the IASCC threshold value during 40 operating years, and baffle/former and baffle former bolts will exceed the threshold value for void swelling, irradiation embrittlement, IASCC, stress relaxation during 40 operating years. However, for all other reactor internals subcomponents, thermal embrittlement, fatigue, SCC, and wear were identified as nonsignificant. As a result of lifetime evaluations, 4 ageing mechanisms were established to be plausible for 3 subcomponents. These results are shown. The existing ageing management programs (AMPs) for Kori Unit 1, such as ISI, water chemistry control, rod drop time testing etc., were

OECD - HRP Summer School on Light Water Reactor Structural Materials. August 26th - 30th, 2002

Energy Technology Data Exchange (ETDEWEB)

NONE

2002-07-01

In cooperation with the OECD Nuclear Energy Agency (NEA), the Halden Reactor Project organised a Summer School on Light Water Reactor Structural Materials in the period August 26 - 30, 2002. The summer school was primarily intended for people who wanted to become acquainted with materials-related subjects and issues without being experts. It is especially hoped that the summer school served to transfer knowledge to the ''young generation'' in the field of nuclear. Experts from Halden Project member organisations were solicited for the following programme: (1) Overview of The Nuclear Community and Current Issues, (2) Regulatory Framework for Ensuring Structural Integrity, (3) Non-Destructive Testing for Detection of Cracks, (4) Part I - Basics of Radiation and Radiation Damage, (5) Part II - Radiation Effects on Reactor Internal Materials, (6) Water Chemistry and Radiolysis Effects in LWRs, (7) PWR and Fast Breeder Reactor Internals, (8) PWR and Fast Breeder Reactor Internals, (9) Secondary Side Corrosion Cracking of PWR Steam Generator Tubes, (10) BWR Materials and Their Interaction with the Environment, (11) Radiation Damage in Reactor Pressure Vessels.

OECD - HRP Summer School on Light Water Reactor Structural Materials. August 26th - 30th, 2002

International Nuclear Information System (INIS)

2002-01-01

In cooperation with the OECD Nuclear Energy Agency (NEA), the Halden Reactor Project organised a Summer School on Light Water Reactor Structural Materials in the period August 26 - 30, 2002. The summer school was primarily intended for people who wanted to become acquainted with materials-related subjects and issues without being experts. It is especially hoped that the summer school served to transfer knowledge to the ''young generation'' in the field of nuclear. Experts from Halden Project member organisations were solicited for the following programme: (1) Overview of The Nuclear Community and Current Issues, (2) Regulatory Framework for Ensuring Structural Integrity, (3) Non-Destructive Testing for Detection of Cracks, (4) Part I - Basics of Radiation and Radiation Damage, (5) Part II - Radiation Effects on Reactor Internal Materials, (6) Water Chemistry and Radiolysis Effects in LWRs, (7) PWR and Fast Breeder Reactor Internals, (8) PWR and Fast Breeder Reactor Internals, (9) Secondary Side Corrosion Cracking of PWR Steam Generator Tubes, (10) BWR Materials and Their Interaction with the Environment, (11) Radiation Damage in Reactor Pressure Vessels

Aging considerations for PWR [pressurized water reactor] control rod drive mechanisms and reactor internals

International Nuclear Information System (INIS)

Ware, A.G.

1988-01-01

This paper describes age-related degradation mechanisms affecting life extension of pressurized water reactor control rod drive mechanisms and reactor internals. The major sources of age-related degradation for control rod drive mechanisms are thermal transients such as plant heatups and cooldowns, latchings and unlatchings, long-term aging effects on electrical insulation, and the high temperature corrosive environment. Flow induced loads, the high-temperature corrosive environment, radiation exposure, and high tensile stresses in bolts all contribute to aging related degradation of reactor internals. Another problem has been wear and fretting of instrument guide tubes. The paper also discusses age-related failures that have occurred to date in pressurized water reactors

Design of A Vibration and Stress Measurement System for an Advanced Power Reactor 1400 Reactor Vessel Internals Comprehensive Vibration Assessment Program

International Nuclear Information System (INIS)

Ko, Doyoung; Kim, Kyuhyung

2013-01-01

In accordance with the US Nuclear Regulatory Commission (US NRC), Regulatory Guide 1.20, the reactor vessel internals comprehensive vibration assessment program (RVI CVAP) has been developed for an Advanced Power Reactor 1400 (APR1400). The purpose of the RVI CVAP is to verify the structural integrity of the reactor internals to flow-induced loads prior to commercial operation. The APR1400 RVI CVAP consists of four programs (analysis, measurement, inspection, and assessment). Thoughtful preparation is essential to the measurement program, because data acquisition must be performed only once. The optimized design of a vibration and stress measurement system for the RVI CVAP is essential to verify the integrity of the APR1400 RVI. We successfully designed a vibration and stress measurement system for the APR1400 RVI CVAP based on the design materials, the hydraulic and structural analysis results, and performance tests of transducers in an extreme environment. The measurement system designed in this paper will be utilized for the APR1400 RVI CVAP as part of the first construction project in Korea

DESIGN OF A VIBRATION AND STRESS MEASUREMENT SYSTEM FOR AN ADVANCED POWER REACTOR 1400 REACTOR VESSEL INTERNALS COMPREHENSIVE VIBRATION ASSESSMENT PROGRAM

Directory of Open Access Journals (Sweden)

DO-YOUNG KO

2013-04-01

Full Text Available In accordance with the US Nuclear Regulatory Commission (US NRC, Regulatory Guide 1.20, the reactor vessel internals comprehensive vibration assessment program (RVI CVAP has been developed for an Advanced Power Reactor 1400 (APR1400. The purpose of the RVI CVAP is to verify the structural integrity of the reactor internals to flow-induced loads prior to commercial operation. The APR1400 RVI CVAP consists of four programs (analysis, measurement, inspection, and assessment. Thoughtful preparation is essential to the measurement program, because data acquisition must be performed only once. The optimized design of a vibration and stress measurement system for the RVI CVAP is essential to verify the integrity of the APR1400 RVI. We successfully designed a vibration and stress measurement system for the APR1400 RVI CVAP based on the design materials, the hydraulic and structural analysis results, and performance tests of transducers in an extreme environment. The measurement system designed in this paper will be utilized for the APR1400 RVI CVAP as part of the first construction project in Korea.

The international thermonuclear reactor project

International Nuclear Information System (INIS)

James, T.R.

1993-01-01

The International Thermonuclear Experimental Reactor Project is a 6-year collaborative effort involving the U.S., Europe, Japan, and the Russian Federation to produce a detailed engineering design for the next-step fusion device

State of the art of toshiba maintenance techniques for reactor internals

International Nuclear Information System (INIS)

Maekawa, Osamu; Hattori, Yasuhiro; Sudo, Akira

2002-01-01

As the number of aged plants increases, maintaining the integrity of the reactor pressure vessel and reactor internals in aged plants has become an essential issue to ensure continued stable operation and achieve higher plant operability. A major issue with regard to reactor internals is stress corrosion cracks (SCCs). Laser-applying techniques have many features suitable for preventive maintenance work on reactor internals. Toshiba has developed various laser-applying preventive maintenance techniques and accumulated considerable field experience utilizing these techniques in various aged plants. Moreover, in view of the importance of confirming the soundness of reactor internals in aged plants, Toshiba has developed and applied sophisticated nondestructive testing techniques for this purpose. (author)

International topical meeting. Research Reactor Fuel Management (RRFM) and meeting of the International Group on Reactor Research (IGORR)

Energy Technology Data Exchange (ETDEWEB)

NONE

2007-07-01

Nuclear research and test reactors have been in operation for over 60 years, over 270 research reactors are currently operating in more than 50 countries. This meeting is dedicated to different aspects of research reactor fuels: new fuels for new reactors, the conversion to low enriched uranium fuels, spent fuel management and computational tools for core simulation. About 80 contributions are reported in this document, they are organized into 7 sessions: 1) international topics and overview on new projects and fuel, 2) new projects and upgrades, 3) fuel development, 4) optimisation and research reactor utilisation, 5) innovative methods in research reactors physics, 6) safety, operation and research reactor conversion, 7) fuel back-end management, and a poster session. Experience from Australian, Romanian, Libyan, Syrian, Vietnamese, South-African and Ghana research reactors are reported among other things. The Russian program for research reactor spent fuel management is described and the status of the American-driven program for the conversion to low enriched uranium fuels is presented. (A.C.)

Technical Meeting on Liquid Metal Reactor Concepts: Core Design and Structural Materials. Working Material

International Nuclear Information System (INIS)

2013-01-01

The objective of the TM on “Liquid metal reactor concept: core design and structural materials” was 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. Main results arising from national and international R&D programmes and projects in the field were reviewed, and new activities to be carried out under the IAEA aegis were identified on the basis of the analysis of current research and technology gaps

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.)

Reactor Vessel and Reactor Vessel Internals Segmentation at Zion Nuclear Power Station - 13230

Energy Technology Data Exchange (ETDEWEB)

Cooke, Conrad; Spann, Holger [Siempelkamp Nuclear Services: 5229 Sunset Blvd., (Suite M), West Columbia, SC, 29169 (United States)

2013-07-01

Zion Nuclear Power Station (ZNPS) is a dual-unit Pressurized Water Reactor (PWR) nuclear power plant located on the Lake Michigan shoreline, in the city of Zion, Illinois approximately 64 km (40 miles) north of Chicago, Illinois and 67 km (42 miles) south of Milwaukee, Wisconsin. Each PWR is of the Westinghouse design and had a generation capacity of 1040 MW. Exelon Corporation operated both reactors with the first unit starting production of power in 1973 and the second unit coming on line in 1974. The operation of both reactors ceased in 1996/1997. In 2010 the Nuclear Regulatory Commission approved the transfer of Exelon Corporation's license to ZionSolutions, the Long Term Stewardship subsidiary of EnergySolutions responsible for the decommissioning of ZNPS. In October 2010, ZionSolutions awarded Siempelkamp Nuclear Services, Inc. (SNS) the contract to plan, segment, remove, and package both reactor vessels and their respective internals. This presentation discusses the tools employed by SNS to remove and segment the Reactor Vessel Internals (RVI) and Reactor Vessels (RV) and conveys the recent progress. SNS's mechanical segmentation tooling includes the C-HORCE (Circumferential Hydraulically Operated Cutting Equipment), BMT (Bolt Milling Tool), FaST (Former Attachment Severing Tool) and the VRS (Volume Reduction Station). Thermal segmentation of the reactor vessels will be accomplished using an Oxygen- Propane cutting system. The tools for internals segmentation were designed by SNS using their experience from other successful reactor and large component decommissioning and demolition (D and D) projects in the US. All of the designs allow for the mechanical segmentation of the internals remotely in the water-filled reactor cavities. The C-HORCE is designed to saw seven circumferential cuts through the Core Barrel and Thermal Shield walls with individual thicknesses up to 100 mm (4 inches). The BMT is designed to remove the bolts that fasten the Baffle

The International Reactor Dosimetry File (IRDF-85)

International Nuclear Information System (INIS)

Cullen, D.E.; McLaughlin, P.K.

1985-04-01

This document describes the contents of the second version of the International Reactor Dosimetry File (IRDF-85), distributed by the Nuclear Data Section of the International Atomic Energy Agency. This library superseded IRDF-82. (author)

Reactor internals design/analysis for normal, upset, and faulted conditions

International Nuclear Information System (INIS)

Burke, F.R.

1977-06-01

The analytical procedures used by Babcock and Wilcox to demonstrate the structural integrity of the 205-FA reactor internals are described. Analytical results are presented and compared to ASME Code allowable limits for Normal, Upset, and Faulted conditions. The particular faulted condition considered is a simultaneous loss-of-coolant accident and safe shutdown earthquake. The operating basis earthquake is addressed as an Upset condition

The IAEA's international project on innovative nuclear reactors and fuel cycles (INPRO)

International Nuclear Information System (INIS)

Kuptiz, Juergen; )

2002-01-01

This paper presents the IAEA International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO). It defines its rationale, key objectives and specifies the organizational structure. The IAEA General Conference (2000) has invited all interested Member states to combine their efforts under the aegis of the Agency in considering the issues of the nuclear fuel cycle, in particular by examining innovative and proliferation-resistant nuclear technology and invited Member states to consider to contribute to a task force on innovative nuclear reactors and fuel cycle

Outlines of guidelines for the inspection and evaluation of reactor vessel internals

International Nuclear Information System (INIS)

Seki, Hiroaki; Kobayashi, Hiroyuki; Nakano, Morihito; Murai, Soutarou; Nomoto, Toshiharu

2014-01-01

'The guideline committee for the inspection and evaluation of Reactor Vessel Internals' of JANSI (Japan Nuclear Safety Institute) has been developing many guidelines based on principle which the conservative methodology, and covered both individual inspection method of reactor internals and application of repair methods for reactor internals. In this paper, some aspects of the JANSI-VIP-03 (Guidelines for the inspection and evaluation of Reactor Vessel Internals, revised Dec.2013) which is summary document of the committee activity, are introduced. (author)

International Working Group on Fast Reactors Sixth Annual Meeting. Summary Report

International Nuclear Information System (INIS)

1973-01-01

The Agenda of the Meeting was as follows: 1. Review of IWGFR Activities - 1a. Approval of the minutes of the Fifth IWGFR Meeting. 1b. Report by Scientific Secretary regarding the activities of the Group. 2. Comments on National Programmes on Fast Breeder Reactors. 3. International Coordination of the Schedule for Major Fast Reactor Meetings and other major international meetings having a predominant fast reactor interest. 4. Consideration of Conferences on Fast Reactors. 4a. IAEA Symposium on Fuel and Fuel Elements for Fast Reactors, Brussels, Belgium 2-6 July 1973. 4b. International Symposium on Physics of Fast Reactors, Tokyo, Japan, 16 to 23 October 1973. 4c. International Conference on Fast Reactor Power Stations, London, UK, 11 to 14 March 1974 . 4d. Suggestions of the IWGFR members on other conferences. 5. Consideration of a Schedule for Specialists' Meetings in 1973-74. 6. Other Business - 6a. First-aid in Sodium Burns. 6b. Principles of Good Practice for Safe Operation of Sodium Circuits. 6c. Bibliography on Fast Reactors. 7. The Date and Place of the Seventh Annual Meeting of the IWGFR

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.)

Completion of the fabrication and assembly of the internal parts and pressure vessel of the LABGENE reactor

International Nuclear Information System (INIS)

Guimaraes, Leonam dos Santos

2005-01-01

The Navy's Technological Center in Sao Paulo (CTMSP) has successfully concluded in 2005 the final assembly of the internals of the Laboratory of Energy Generation's Reactor (LABGENE). This structure together with the fuel elements and the control rods drives mechanisms are part of a PWR type Nuclear Reactor. (author)

Nuclear reactor internals construction and failed fuel rod detection system

International Nuclear Information System (INIS)

Frisch, E.; Andrews, H.N.

1976-01-01

A system is provided for determining during operation of a nuclear reactor having fluid pressure operated control rod mechanisms the exact location of a fuel assembly with a defective fuel rod. The construction of the reactor internals is simplified in a manner to facilitate the testing for defective fuel rods and the reduce the cost of producing the upper internals of the reactor. 13 claims, 10 drawing figures

Dimensional control and check of field machining parts for reactor internals installation

International Nuclear Information System (INIS)

Zhang Caifang

2010-01-01

Some key issues of dimensional control for reactor internals installation are analyzed, and important technical requirements of crucial quality control elements on the measurement, machining, and checking of reactor internals filed machining parts are discussed. Moreover, provisions on quality control and risk prevention of reactor internals filed machining parts are presented in this paper. (author)

International seminar on structural investigations on pulsed neutron sources. Proceedings

International Nuclear Information System (INIS)

Aksenov, V.L.; Balagurov, A.M.; Taran, Yu.V.

1993-01-01

The proceedings of the International seminar on structural investigations using pulsed neutron sources are presented. The seminar is dedicated to the memory of Dr. Yu.M. Ostanevich, a world acknowledged physicist. The problems of structural analysis using pulsed neutron source at the IBR-2 reactor are discussed

Influence of high dose irradiation on core structural and fuel materials in advanced reactors

International Nuclear Information System (INIS)

1998-08-01

The IAEA International Working Group on Fast Reactors (IWGFR) periodically organizes meeting to discuss and review important aspects of fast reactor technology. The fifth meeting held in Obninsk, Russian Federation, 16-19 June 1997, was devoted to the influence of high dose irradiation on the mechanical properties of reactor core structural and fuel materials. The proceedings includes the papers submitted at this meeting each with a separate abstract

Computer-generated vibratory signatures for EDF PWR reactor vessel internals

International Nuclear Information System (INIS)

Trenty, A.; Lefevre, F.; Garreau, D.

1992-07-01

This paper presents a device for generation of characteristic signatures for normal or faulty vibrations on EDF PWR internal structures. The objective is to test the efficiency of methods for diagnosing faults in these structures. With this device, it is possible to build an entire PSD in several phases: choice of a general basic shape, localized addition of several kinds of background noise, generation of peaks of variable shapes, adjustment of local or global amplifications... It also offers the possibility of distorting real PSDs acquired from the reactor: shifting frequency or modifying peak shape, eliminating or adding existing shapes or shapes to be created, smoothing curves... One example is given of simulated loss of function in a hold-down spring on a computer-generated PSD of ex-core neutron noise. The device is now being used to test the potential of neural networks in recognizing faults on internal structures

IRPhEP-handbook, International Handbook of Evaluated Reactor Physics Benchmark Experiments

International Nuclear Information System (INIS)

Sartori, Enrico; Blair Briggs, J.

2008-01-01

1 - Description: The purpose of the International Reactor Physics Experiment Evaluation Project (IRPhEP) is to provide an extensively peer-reviewed set of reactor physics-related integral data that can be used by reactor designers and safety analysts to validate the analytical tools used to design next-generation reactors and establish the safety basis for operation of these reactors. This work of the IRPhEP is formally documented in the 'International Handbook of Evaluated Reactor Physics Benchmark Experiments,' a single source of verified and extensively peer-reviewed reactor physics benchmark measurements data. The IRPhE Handbook is available on DVD. You may request a DVD by completing the DVD Request Form available at: http://irphep.inl.gov/handbook/hbrequest.shtml The evaluation process entails the following steps: 1. Identify a comprehensive set of reactor physics experimental measurements data, 2. Evaluate the data and quantify overall uncertainties through various types of sensitivity analysis to the extent possible, verify the data by reviewing original and subsequently revised documentation, and by talking with the experimenters or individuals who are familiar with the experimental facility, 3. Compile the data into a standardized format, 4. Perform calculations of each experiment with standard reactor physics codes where it would add information, 5. Formally document the work into a single source of verified and peer reviewed reactor physics benchmark measurements data. The International Handbook of Evaluated Reactor Physics Benchmark Experiments contains reactor physics benchmark specifications that have been derived from experiments that were performed at various nuclear experimental facilities around the world. The benchmark specifications are intended for use by reactor physics personal to validate calculational techniques. The 2008 Edition of the International Handbook of Evaluated Reactor Physics Experiments contains data from 25 different

Age-related degradation of boiling water reactor vessel internals

International Nuclear Information System (INIS)

Ware, A.G.; Shah, V.N.

1992-01-01

Researchers at the Idaho National Engineering Laboratory performed an assessment of the aging of the reactor internals in boiling water reactors (BWRs), and identified the unresolved technical issues related to the degradation of these components. The overall life-limiting mechanism is intergranular stress corrosion cracking (IGSCC). Irradiation-assisted stress corrosion cracking, fatigue, and thermal aging embrittlement are other potential degradation mechanisms. Several failures in BWR internals have been caused by a combination of factors such as environment, high residual or preload stresses, and flow-induced vibration. The ASME Code Section XI in-service inspection requirements are insufficient for detecting aging-related degradation at many locations in reactor internals. Many of the potential locations for IGSCC or fatigue are not accessible for inspection. (orig.)

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.)

Segmentation and packaging reactor vessels internals

International Nuclear Information System (INIS)

Boucau, Joseph

2014-01-01

Document available in abstract form only, full text follows: With more than 25 years of experience in the development of reactor vessel internals and reactor vessel segmentation and packaging technology, Westinghouse has accumulated significant know-how in the reactor dismantling market. The primary challenges of a segmentation and packaging project are to separate the highly activated materials from the less-activated materials and package them into appropriate containers for disposal. Since disposal cost is a key factor, it is important to plan and optimize waste segmentation and packaging. The choice of the optimum cutting technology is also important for a successful project implementation and depends on some specific constraints. Detailed 3-D modeling is the basis for tooling design and provides invaluable support in determining the optimum strategy for component cutting and disposal in waste containers, taking account of the radiological and packaging constraints. The usual method is to start at the end of the process, by evaluating handling of the containers, the waste disposal requirements, what type and size of containers are available for the different disposal options, and working backwards to select a cutting method and finally the cut geometry required. The 3-D models can include intelligent data such as weight, center of gravity, curie content, etc, for each segmented piece, which is very useful when comparing various cutting, handling and packaging options. The detailed 3-D analyses and thorough characterization assessment can draw the attention to material potentially subject to clearance, either directly or after certain period of decay, to allow recycling and further disposal cost reduction. Westinghouse has developed a variety of special cutting and handling tools, support fixtures, service bridges, water filtration systems, video-monitoring systems and customized rigging, all of which are required for a successful reactor vessel internals

IGORR-IV: Proceedings of the fourth meeting of the International Group On Research Reactors

International Nuclear Information System (INIS)

Rosenbalm, K.F.

1995-01-01

The fourth meeting of the International Group on Research Reactors (IGORR-IV) was attended by was good 55 registered participants from 28 organizations in 13 countries, which compares well with the previous meetings. Twenty-nine papers were presented in five sessions over the two-day meeting. Session subjects were: Operating Research Reactors and Facility Upgrades; Research Reactors in Desin and Construction; Research, Development, and Analysis Results of Thermal Hydraulic Calculations, U 3 Si 2 Fuel Performance and Faibrication; Structural Materials Performance; Neutronics; Severe Accident analysis. Written versions of the papers or hard copies of the viewgraphs used are published in these Proceedings

Design of reactor internals in larger high-temperature reactors with spherical fuel elements

International Nuclear Information System (INIS)

Elter, C.

1981-01-01

In his paper, the author analyzes and summarizes the present state of the art with emphasis on the prototype reactor THTR 300 MWe, because in addition to spherical fuel elements, this type includes other features of future HTR design such as the same flow direction of cooland gas through the core. The paper on hand also elaborates design guidelines for reactor internals applicable with large HTR's of up to 1200 MWe. Proved designs will be altered so as to meet the special requirements of larger cores with spherical elements to be reloaded according to the OTTO principle. This paper is furthermore designed as a starting point for selective and swift development of reactor internals for large HTR's to be refuelled according to the OTTO principle. (orig./GL) [de

Development of advanced design features for KNGR reactor vessel and internals

Energy Technology Data Exchange (ETDEWEB)

Park, Jong Kyun; Ru, Bong; Lee, Jae Han; Lee, Hyung Yeon; Kim, Jong Bum; Ku, Kyung Heoy; Lee, Ki Young; Lee, Jun; Kim, Young In [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1995-12-01

Developments of KNGR design require to enhance the design to implement the design requirements, such as plant life time from 40 years to 60 years, safety, performance and structure and components design. The designs used for existing nuclear power plants should be modified or improved to meet the requirements in KNGR design. The purpose of the task is to develop the Advanced Design Features (ADF) related to mechanical and structural design for KNGR reactor vessel and reactor internals. The structural integrity for the System 80+ reactor vessel, of which design life is 60 years, was reviewed. EPRI-URD, CESSAR-DC, and the present design status and characteristics of System 80+ reactor vessel were comparatively studied and the improvement of reactor vessel surveillance program was investigated. The performance and aseismic characteristics of the CE-type CEDM, which will be used in System 80+, are investigated. The driving cycles of CEDM are evaluated for the load follow operation(LFO), of which Mode K is being developed by KAERI. The position of the USNRC, EPRI, ABB-CE, and industries on the elimination of OBE are reviewed, and especially ABB-CE System 80+ FSER is reviewed in detail. For the pre-stage of the verification of the OBE elimination from the design, the review of the seismic responses, i.e.. shear forces and moments, of YGN 3/4 RI was performed and the ratio of OBE response to SSE response was analysed. The screening criteria were reviewed to evaluate the integrity against pressurized thermal shock (PTS) for RV belt-line of System 80+. The evaluation methods for fracture integrity when screening criteria are not met were reviewed. The structural characteristics of IRWST spargers of System 80+ were investigated and the effect of hydrodynamic loads on NSSS was reviewed. 18 figs., 9 tabs., 40 refs. (Author) .new.

Development of advanced design features for KNGR reactor vessel and internals

International Nuclear Information System (INIS)

Park, Jong Kyun; Ru, Bong; Lee, Jae Han; Lee, Hyung Yeon; Kim, Jong Bum; Ku, Kyung Heoy; Lee, Ki Young; Lee, Jun; Kim, Young In

1995-12-01

Developments of KNGR design require to enhance the design to implement the design requirements, such as plant life time from 40 years to 60 years, safety, performance and structure and components design. The designs used for existing nuclear power plants should be modified or improved to meet the requirements in KNGR design. The purpose of the task is to develop the Advanced Design Features (ADF) related to mechanical and structural design for KNGR reactor vessel and reactor internals. The structural integrity for the System 80+ reactor vessel, of which design life is 60 years, was reviewed. EPRI-URD, CESSAR-DC, and the present design status and characteristics of System 80+ reactor vessel were comparatively studied and the improvement of reactor vessel surveillance program was investigated. The performance and aseismic characteristics of the CE-type CEDM, which will be used in System 80+, are investigated. The driving cycles of CEDM are evaluated for the load follow operation(LFO), of which Mode K is being developed by KAERI. The position of the USNRC, EPRI, ABB-CE, and industries on the elimination of OBE are reviewed, and especially ABB-CE System 80+ FSER is reviewed in detail. For the pre-stage of the verification of the OBE elimination from the design, the review of the seismic responses, i.e.. shear forces and moments, of YGN 3/4 RI was performed and the ratio of OBE response to SSE response was analysed. The screening criteria were reviewed to evaluate the integrity against pressurized thermal shock (PTS) for RV belt-line of System 80+. The evaluation methods for fracture integrity when screening criteria are not met were reviewed. The structural characteristics of IRWST spargers of System 80+ were investigated and the effect of hydrodynamic loads on NSSS was reviewed. 18 figs., 9 tabs., 40 refs. (Author) .new

Feedback from Westinghouse experience on segmentation of reactor vessel internals - 59013

International Nuclear Information System (INIS)

Kreitman, Paul J.; Boucau, Joseph; Segerud, Per; Fallstroem, Stefan

2012-01-01

With more than 25 years of experience in the development of reactor vessel internals segmentation and packaging technology, Westinghouse has accumulated significant know-how in the reactor dismantling market. Building on tooling concepts and cutting methodologies developed decades ago for the successful removal of nuclear fuel from the damaged Three Mile Island Unit 2 reactor (TMI-2), Westinghouse has continuously improved its approach to internals segmentation and packaging by incorporating lessons learned and best practices into each successive project. Westinghouse has developed several concepts to dismantle reactor internals based on safe and reliable techniques, including plasma arc cutting (PAC), abrasive water-jet cutting (AWJC), metal disintegration machining (MDM), or mechanical cutting. Westinghouse has applied its technology to all types of reactors covering Pressurized Water Reactors (PWR's), Boiling Water Reactors (BWR's), Gas Cooled Reactors (GCR's) and sodium reactors. The primary challenges of a segmentation and packaging project are to separate the highly activated materials from the less-activated materials and package them into appropriate containers for disposal. Since space is almost always a limiting factor it is therefore important to plan and optimize the available room in the segmentation areas. The choice of the optimum cutting technology is important for a successful project implementation and depends on some specific constraints like disposal costs, project schedule, available areas or safety. Detailed 3-D modeling is the basis for tooling design and provides invaluable support in determining the optimum strategy for component cutting and disposal in waste containers, taking account of the radiological and packaging constraints. Westinghouse has also developed a variety of special handling tools, support fixtures, service bridges, water filtration systems, video-monitoring systems and customized rigging, all of which are required for a

Magnet systems for the International Thermonuclear Experimental Reactor

International Nuclear Information System (INIS)

Henning, C.D.; Miller, J.R.

1988-01-01

The definition phase for the International Thermonuclear Experimental Reactor (ITER) has been nearly completed, thus beginning a three-year design effort by teams from the European Community (EC), Japan, US, and USSR. Preliminary parameters for the superconducting magnet system have been established to guide more detailed design work. Radiation tolerance of the superconductors and insulators has been important because it sets requirements for the neutron-shield dimension and sensitively influences reactor size. Major levels of mechanical stress appear in the structural cases of the inboard legs of the toroidal-field (TF) coils. The winding packs of the TF coils include significant fractions of steel that provide support against in-plane separating loads, but they offer little support against out-of-plane loads unless shear-bonding of the conductors can be maintained. Heat removal from nuclear and ac loads has not limited the fundamental design, but it has nonnegligible economic consequences. 3 refs., 3 figs., 5 tabs

PWR reactor pressure vessel internals license renewal industry report; revision 1. Final report

International Nuclear Information System (INIS)

Schwirian, R.; Robison, G.

1994-07-01

The U.S. nuclear power industry, through coordination by the Nuclear Management and Resources Council (NUMARC), and sponsorship by the U.S. Department of Energy (DOE) and the Electric Power Research Institute (EPRI), has evaluated age-related degradation effects for a number of major plant systems, structures and components, in the license renewal technical Industry Reports (IRs). License renewal applicants may choose to reference these IRs in support of their plant-specific license renewal applications, as an equivalent to the integrated plant assessment provisions of the license renewal rule (10 CFR Part 54). Pressurized water reactor (PWR) reactor pressure vessel (RPV) internals designed by all three U.S. PWR nuclear steam supply system vendors have been evaluated relative to the effects of age-related degradation mechanisms; the capability of current design limits; inservice examination, testing, repair, refurbishment, and other programs to manage these effects; and the assurance that these internals can continue to perform their intended safety functions in the license renewal term. This industry report (IR), one of a series of ten, provides a generic technical basis for evaluation of PWR reactor pressure vessel internals for license renewal

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

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)

Fabrication of internally instrumented reactor fuel rods

International Nuclear Information System (INIS)

Schmutz, J.D.; Meservey, R.H.

1975-01-01

Procedures are outlined for fabricating internally instrumented reactor fuel rods while maintaining the original quality assurance level of the rods. Instrumented fuel rods described contain fuel centerline thermocouples, ultrasonic thermometers, and pressure tubes for internal rod gas pressure measurements. Descriptions of the thermocouples and ultrasonic thermometers are also contained

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)

The International Thermonuclear Experimental Reactor configuration evolution

International Nuclear Information System (INIS)

Lousteau, D.C.; Nelson, B.E.; Lee, V.D.; Thomson, S.L.; Miller, J.M.; Lindquist, W.B.

1989-01-01

The International Thermonuclear Experimental Reactor (ITER) conceptual design activities consist of two phases: a definition phase, completed in September 1988, and a design phase, now in progress. The definition phase was successful in identifying a consistent set of technical characteristics and the broad definition of the required reactor configuration and hardware. Scheduled for completion in November 1990, the design phase is producing a more detailed definition of the required components, a first cost estimate, and a description of site requirements. A major activity in the ITER design phase is the period of joint work conducted at the Max Planck Institute for Plasma Physics, Garching, Federal Republic of Germany, from June through October 1989. An official report of the findings and conclusions of this activity will be submitted to and published by the International Atomic Energy Agency (IAEA). This paper highlights the evolution of the reactor mechanical configuration since the conclusion of the definition phase. 8 figs., 2 tabs

Induction apparatus monitoring structural strains in liquid-metal-cooled nuclear reactor

International Nuclear Information System (INIS)

Dean, S.A.; Evans, R.A.

1981-01-01

An improved method of monitoring induced torsional and linear strains in the internal structures of liquid metal cooled nuclear reactors is described. An electrical induction apparatus indicates the variation of magnetic coupling caused by a ferromagnetic member of the apparatus being subjected to such strains. (U.K.)

International Reactor Physics Experiment Evaluation (IRPhE) Project

International Nuclear Information System (INIS)

2013-01-01

The International Reactor Physics Experiment Evaluation (IRPhE) Project aims to provide the nuclear community with qualified benchmark data sets by collecting reactor physics experimental data from nuclear facilities, worldwide. More specifically the objectives of the expert group are as follows: - maintaining an inventory of the experiments that have been carried out and documented; - archiving the primary documents and data released in computer-readable form; - promoting the use of the format and methods developed and seek to have them adopted as a standard. For those experiments where interest and priority is expressed by member countries or working parties and executive groups within the NEA provide guidance or co-ordination in: - compiling experiments into a standard international agreed format; - verifying the data, to the extent possible, by reviewing original and subsequently revised documentation, and by consulting with the experimenters or individuals who are familiar with the experimenters or the experimental facility; - analysing and interpreting the experiments with current state-of-the-art methods; - publishing electronically the benchmark evaluations. The expert group will: - identify gaps in data and provide guidance on priorities for future experiments; - involve the young generation (Masters and PhD students and young researchers) to find an effective way of transferring know-how in experimental techniques and analysis methods; - provide a tool for improved exploitation of completed experiments for Generation IV reactors; - coordinate closely its work with other NSC experimental work groups in particular the International Criticality Safety Benchmark Evaluation Project (ICSBEP), the Shielding Integral Benchmark Experiment Data Base (SINBAD) and others, e.g. knowledge preservation in fast reactors of the IAEA, the ANS Joint Benchmark Activities; - keep a close link with the working parties on scientific issues of reactor systems (WPRS), the expert

Comparison of advanced reactors program of different international vendors

International Nuclear Information System (INIS)

Agnihotri, N.K.

2001-01-01

The full text follows. Proposal for presenting a paper on Advanced Reactor Program Given below is the abstract for Track 6 session on Advanced Reactor at the ninth International Conference on Nuclear Engineering being held in Nice, France from April 8. through 12. 2001. This paper will provide an update on Advanced Reactor Program of different vendors in the United States, Japan, and Europe. Specifically the paper will look at the history of different Advanced Reactor Programs, international experience, aspect of economy due to standardization, and the highlights of technical specifications. The paper will also review aspects of Economy due to standardization, public acceptance, required construction time, and the experience of different vendors. The objective of the presentation is to underscore the highlights of the Reactor Program of different vendors in order to keep the attendees of the conference up-to-date. The presentation will be an impartial overview from an outsider's (not part of the Nuclear Steam Supply System's staff). (author)

IGORR-IV: Proceedings of the fourth meeting of the International Group On Research Reactors

Energy Technology Data Exchange (ETDEWEB)

Rosenbalm, K F [comp.

1995-07-01

The fourth meeting of the International Group on Research Reactors (IGORR-IV) was attended by was good 55 registered participants from 28 organizations in 13 countries, which compares well with the previous meetings. Twenty-nine papers were presented in five sessions over the two-day meeting. Session subjects were: Operating Research Reactors and Facility Upgrades; Research Reactors in Desin and Construction; Research, Development, and Analysis Results of Thermal Hydraulic Calculations, U{sub 3}Si{sub 2} Fuel Performance and Faibrication; Structural Materials Performance; Neutronics; Severe Accident analysis. Written versions of the papers or hard copies of the viewgraphs used are published in these Proceedings.

Iser: an international inherently safe reactor concept

International Nuclear Information System (INIS)

Wakabayashi, Hiroaki

1988-01-01

Iser is a modular standardised 200-300 MWe power reactor based on the PIUS principle. It differs from PIUS in being simpler, and making full use of existing steel-vessel-based LWR technology. Iser is an inherently safe reactor concept under development in Japan. It is a generic concept, not a patented commodity, and it is expected that an international association to develop the concept will be formed. (U.K.)

Effects of fluid communications between fluid volumes on the seismic behaviour of nuclear breeder reactor internals

International Nuclear Information System (INIS)

Durandet, E.; Gibert, R.J.

1987-01-01

The internal structures of a breeder reactor as SUPERPHENIX are mainly axisymmetrial shells separated by fluid volumes which are connected by small communications holes. These communications can destroy the axisymmetry of the problem and their effects on the inertial terms due to the fluid are important. An equivalent axisymmetrical element based on a local tridimensional solution in the vicinity of the fluid communication is defined. An axisymmetrical modelization using this type of element is built in order to calculate the horizontal seismic behaviour of the reactor internals. The effect due to three typical fluid communications are studied and compared. (orig.)

Design of the upper internals structure for the Clinch River Breeder Reactor Plant

International Nuclear Information System (INIS)

Thompson, D.C.; Novendstern, E.H.

1977-01-01

The Upper Internals Structure (UIS) is located above the core and is supported from the head at four locations. It is designed to perform the following primary functions: provide secondary core holddown in the event of a malfunction of the core hydraulic holddown system; provide support for routing all in-vessel instrumentation to core assemblies; maintain alignment between the core assemblies, the UIS and the closure head; provide guidance and crossflow protection for the control rod drivelines; and mix/duct flow to the upper region of the vessel outlet plenum to minimize rapid temperature changes to components during a reactor trip transient. In accomplishing these functions, the UIS will experience a sodium environment with temperatures up to 1200 0 F (649 0 C), and as many as 7 x 10 8 cycles of fluid temperature fluctuations up to 250 0 F (121 0 C) at full power operation. It must be designed to survive these conditions in combination with seismic and flow-induced vibration loadings for its 30 year design life. The design program of designing to controlled functional requirements and design conditions is discussed. Included is a description of the significant parts of the design and the approach used to balance the requirement of tight joints. The thermal and hydraulic environment including the results of a comprehensive test program are discussed. The test program results establish the basis of the thermal boundary used in the structural evaluation, and the UIS vibration characteristics. A summary of the areas which have required design changes is included with a summary of the structural evaluation of these changes

Nuclear data for reactors. Proceedings of the second international conference. Vol. II

International Nuclear Information System (INIS)

1970-01-01

The Second International Conference on Nuclear Data for Reactors, held in Helsinki at the invitation of the Finnish Government, was convened by the International Atomic Energy Agency from 15 to 19 June 1970. The Conference, held as a result of recommendations made by the International Nuclear Data Committee, was attended by 163 participants from 28 countries and four international organizations, and 21 invited and 98 contributed papers were presented. This Conference was the second held by the IAEA on Nuclear Data for Reactors. Almost four years have elapsed since the first was held in Paris in 1966. During these years gratifying progress has been made by reactor, nuclear and evaluation physicists, whose collaboration has been greatly enhanced. As a result, many laboratories have concentrated their efforts on items of particular importance for reactor research and development, and many measurements are now available. The main purpose of this Conference was to provide an opportunity to review results of recent basic neutron-physics investigations against a background need for basic information, especially concerning reactors. The Conference itself, together with the preparatory meetings of IAEA experts in Studsvik on the status of α( 239 Pu) and the ν-bar-values for fissionable nuclei, showed an emphasis on the nuclear data aspects most important for nuclear technology. Most contributors dealt with the measurement and analysis of neutron cross-sections. This extensive new cross-section information can be attributed to several factors, the most important being the development and systematic exploitation of high-intensity neutron sources, such as modern linear accelerators, modern cyclotrons and underground nuclear explosions, improvements in instrumentation and in sample preparation techniques, and other technical improvements. Compared with the first IAEA Conference on Nuclear Data for Reactors this one has many more contributions on neutron data evaluation. Many

Nuclear data for reactors. Proceedings of the second international conference. Vol. I

International Nuclear Information System (INIS)

1970-01-01

The Second International Conference on Nuclear Data for Reactors, held in Helsinki at the invitation of the Finnish Government, was convened by the International Atomic Energy Agency from 15 to 19 June 1970. The Conference, held as a result of recommendations made by the International Nuclear Data Committee, was attended by 163 participants from 28 countries and four international organizations, and 21 invited and 98 contributed papers were presented. This Conference was the second held by the IAEA on Nuclear Data for Reactors. Almost four years have elapsed since the first was held in Paris in 1966. During these years gratifying progress has been made by reactor, nuclear and evaluation physicists, whose collaboration has been greatly enhanced. As a result, many laboratories have concentrated their efforts on items of particular importance for reactor research and development, and many measurements are now available. The main purpose of this Conference was to provide an opportunity to review results of recent basic neutron-physics investigations against a background need for basic information, especially concerning reactors. The Conference itself, together with the preparatory meetings of IAEA experts in Studsvik on the status of α( 239 Pu) and the ν-bar-values for fissionable nuclei, showed an emphasis on the nuclear data aspects most important for nuclear technology. Most contributors dealt with the measurement and analysis of neutron cross-sections. This extensive new cross-section information can be attributed to several factors, the most important being the development and systematic exploitation of high-intensity neutron sources, such as modern linear accelerators, modern cyclotrons and underground nuclear explosions, improvements in instrumentation and in sample preparation techniques, and other technical improvements. Compared with the first IAEA Conference on Nuclear Data for Reactors this one has many more contributions on neutron data evaluation. Many

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)

Operational feedback on internal structure vibration in 54 French PWRs during 300 fuel cycles

International Nuclear Information System (INIS)

Trenty, A.

1995-01-01

EDF has acquired extensive feedback on vibration of reactor vessel internals by analysing ex-core neutron noise on its 54 pressurized water reactors during the course of over 300 fuel cycles. This feedback has been built up by processing more than 3,000 vibratory signatures acquired since the startup of its reactors. These signatures are now centralized for the whole of France in the ''SINBAD'' data base. Signature processing has enabled: distinguishing between mechanical phenomena and signature variation linked to unit operation: in particular, the impact on signature level of unit operating parameters such as initial fuel enrichment and burn-up rate was assessed; among the purely mechanical phenomena, pointing up slight changes in position of vessel internals and the first signs of structural wear; relaxation (in the hold-down spring and fuel rod assemblies) and wear on surfaces of contact between internals and reactor vessel were detected; lastly and most importantly, automatic recognition of the various types of vibratory behavior of internals. It was consequently possible to draw up user requirement specifications for automated monitoring of internals, which should soon be integrated in PSAD, a system which groups several reactor monitoring functions. (author)

Method of starting internal pumps of a nuclear reactor

International Nuclear Information System (INIS)

Kumagami, Shoji.

1985-01-01

Purpose: To reduce the noise effects by decreasing the invading current into the main line upon starting an internal pump type nuclear reactor adapted to forcively recycle the reactor water by a plurality of internal pumps. Method: A plurality of internal pumps are divided into several groups and, upon starting pumps belonging to the individual unit group, the starting instances for the respective pumps are deviated to reduce the surges applied to the main line and suppress the invading current lower to reduce the earth noises. As a result, effects caused to other devices or equipments can be moderated to improve the reliability. Furthermore, by actuating the respective pumps on every group units in a starting pattern along the orthogonal line, flow rate distribution in the reactor can be balanced. Then, the instability region during low rotation of pumps, that is, instability of the flow rate near the resonance frequency can be decreased. (Kawakami, Y.)

Determination of thermal-hydraulic loads on reactor internals in a DBA-situation

International Nuclear Information System (INIS)

Ville Lestinen; Timo Toppila

2005-01-01

Full text of publication follows: According to Finnish regulatory requirements, reactor internals have to stay intact in a design basis accident (DBA) situation, so that control rods can still penetrate into the core. To fulfill this demand some criteria must be followed in periodical in-service inspections. This is the motivation for studying and developing more detailed methods for analysis of thermal-hydraulic loads on reactor internals during the DBA-situation for the Loviisa NPP in Finland. The objective of this research program is to connect thermal-hydraulic and mechanical analysis methods with the goal to produce a reliable method for determination of thermal-hydraulic and mechanical loads on reactor internals in the accident situation. The tools studied are thermal-hydraulic system codes, computational fluid dynamics (CFD) codes and finite element analysis (FEA) codes. This paper concentrates mainly on thermal-hydraulic part of the research, but also the mechanical aspects are discussed. Firstly, the paper includes a short literary review of the available methods to analyse the described problem including both thermal-hydraulic and structural analysis parts. Secondly, different possibilities to carry out thermal-hydraulic analyses have been studied. The DBA-case includes complex physical phenomena and therefore modelling is difficult. The accident situation can be for example LLOCA. When the pipe has broken, the pressure decreases and water starts to evaporate, which consumes energy and that way limits the pressure decrease. After some period of time, the system reaches a new equilibrium state. To perform exact thermal-hydraulic analysis also two phase phenomena must be included. Therefore CFD codes are not capable of modelling the DBA situation very well, but the use of CFD codes requires that the effect of two phase flow must be added somehow. One method to calculate two phase phenomena with CFD codes is to use thermal-hydraulic system codes to calculate

Development of inertia-increased reactor internal pump

International Nuclear Information System (INIS)

Tanaka, Masaaki; Matsumura, Seiichi; Kikushima, Jun; Kawamura, Shinichi; Yamashita, Norimichi; Kurosaki, Toshikazu; Kondo, Takahisa

2000-01-01

The Reactor Internal Pump (RIP) was adopted for the Reactor Recirculation System (RRS) of Advanced Boiling Water Reactor (ABWR) plants, and ten RIPs are located at the bottom of the reactor pressure vessel. In order to simplify the power supply system for the RIPs, a new inertia-increased RIP was developed, which allows to eliminate the Motor-Generator (M-G) sets. The rotating inertia was increased approximately 2.5 times of current RIP inertia by addition of flywheel on its main shaft. A full scale proving test of the inertia-increased RIP under actual plant operating conditions using full scale test loop was performed to evaluate vibration characteristics and coast down characteristics. From the results of this proving test, the validity of the new inertia-increased RIP and its power supply system (without M-G sets) was confirmed. (author)

Cutting technique for reactor internals by laser beam

International Nuclear Information System (INIS)

Matsumoto, O.; Sugihara, M.; Matsuda, K.; Miya, K.

1990-01-01

At present in Japan the verification tests on the commercial nuclear power reactor decommissioning technology are being conducted as the project of The Ministry of International Trade and Industry by Nuclear Power Engineering Test Center. This paper summarizes the interim results of the verification test for the reactor core internals decommissioning technology, which is being conducted from 1986 as a theme of the above project. All core internals to be studied here are made of stainless steel, and the maximum wall thickness is about 500mm (the maximum one to be cut is about 300mm) for the PWR's, and about 100mm for the BWR's. Though the plasma cutting, arc saw cutting method, etc. have been studied u p to now as the cutting technology for decommissioning these core internals, the authors are carrying out the development and verification test of the cutting technology with the laser beam, which is expected to increase its power in future and can be applied to various materials

International Nuclear Safety Center database on thermophysical properties of reactor materials

International Nuclear Information System (INIS)

Fink, J.K.; Sofu, T.; Ley, H.

1997-01-01

The International Nuclear Safety Center (INSC) database has been established at Argonne National Laboratory to provide easily accessible data and information necessary to perform nuclear safety analyses and to promote international collaboration through the exchange of nuclear safety information. The INSC database, located on the World Wide Web at http://www.insc.anl.gov, contains critically assessed recommendations for reactor material properties for normal operating conditions, transients, and severe accidents. The initial focus of the database is on thermodynamic and transport properties of materials for water reactors. Materials that are being included in the database are fuel, absorbers, cladding, structural materials, coolant, and liquid mixtures of combinations of UO 2 , ZrO 2 , Zr, stainless steel, absorber materials, and concrete. For each property, the database includes: (1) a summary of recommended equations with uncertainties; (2) a detailed data assessment giving the basis for the recommendations, comparisons with experimental data and previous recommendations, and uncertainties; (3) graphs showing recommendations, uncertainties, and comparisons with data and other equations; and (4) property values tabulated as a function of temperature

Physical model study of neutron noise induced by vibration of reactor internals

International Nuclear Information System (INIS)

Liu Jinhui; Gu Fangyu

1999-01-01

The author presents a physical model of neutron noise induced by reactor internals vibration in frequency domain. Based on system control theory, the reactor dynamic equations are coupled with random vibration equation, and non-linear terms are also taken into accounted while treating the random vibration. Experiments carried out on a zero-power reactor show that the model can be used to describe dynamic character of neutron noise induced by internals' vibration. The model establishes a method to help to determine internals'vibration features, and to diagnosis anomalies through neutron noise

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.

Review of the International Atomic Energy Agency International database on reactor pressure vessel materials and US Nuclear Regulatory Commission/Oak Ridge National Laboratory embrittlement data base

International Nuclear Information System (INIS)

Wang, J.A.; Kam, F.B.K.

1998-02-01

The International Atomic Energy Agency (IAEA) has supported neutron radiation effects information exchange through meetings and conferences since the mid-1960s. Through an International Working Group on Reliability of Reactor Pressure Components, information exchange and research activities were fostered through the Coordinated Research Program (CRP) sponsored by the IAEA. The final CRP meeting was held in November 1993, where it was recommended that the IAEA coordinate the development of an International Database on Reactor Pressure Vessel Material (IDRPVM) as the first step in generating an International Database on Aging Management. The purpose of this study was to provide special technical assistance to the NRC in monitoring and evaluating the IAEA activities in developing the IAEA IDRPVM, and to compare the IDRPVM with the Nuclear Regulatory Commission (NRC) - Oak Ridge National Laboratory (ORNL) Power Reactor Embrittlement Data Base (PR-EDB) and provide recommendations for improving the PR-EDB. A first test version of the IDRPVM was distributed at the First Meeting of Liaison Officers to the IAEA IDRPVM, in November 1996. No power reactor surveillance data were included in this version; the testing data were mainly from CRP Phase III data. Therefore, because of insufficient data and a lack of power reactor surveillance data received from the IAEA IDRPVM, the comparison is made based only on the structure of the IDRPVM. In general, the IDRPVM and the EDB have very similar data structure and data format. One anticipates that because the IDRPVM data will be collected from so many different sources, quality assurance of the data will be a difficult task. The consistency of experimental test results will be an important issue. A very wide spectrum of material characteristics of RPV steels and irradiation environments exists among the various countries. Hence the development of embrittlement prediction models will be a formidable task. 4 refs., 2 figs., 4 tabs

Modelling of internal structure in seismic analysis of a PHWR building

International Nuclear Information System (INIS)

Reddy, G.R.; Vaze, K.K.; Kushawaha, H.S.; Ingle, R.K.; Subramanian, K.V.

1991-01-01

Seismic analysis of complex and large structures, consisting of thick shear walls, such as Reactor Building is very involved and time consuming. It is a standard practice to model the structure as a stick model to predict reasonably the dynamic behaviour of the structure. It is required to determine approximate equivalent sectional properties of Internal Structure for representation in the stick model. The restraint to warping can change the stress distribution thus affecting the centre of rigidity and torsional inertia, Hence, standard formulae does not hold good for determination of sectional properties of the Internal Structure. In this case the equivalent sectional properties for the Internal Structure are calculated using a Finite Element Model (FEM) of the Internal Structure and applying unit horizontal forces in each direction. A 3-D stick model is developed using the guidelines. Using the properties calculated by FEM and also by standard formulae, the responses of the 3-D stick model are compared. (J.P.N.)

An overview of reactor vessel internals segmentation for nuclear plant decommissioning

International Nuclear Information System (INIS)

Litka, T.J.

1994-01-01

Several nuclear plants have undergone reactor vessel (RV) internals segmentation as part of or in preparation for decommissioning the plant. In addition, several other nuclear facilities are planning for similar work efforts. The primary technology used for segmentation of RV internals, whether in-air or underwater is Plasma Arc Cutting (PAC). Metal Disintegration Machining (MDM) is also used for difficult to make cuts. PAC and MDM are deployed by various means including Long Handled Tools (LHTs), fixtures, tracks, and multi-axis manipulators. These enable remote cutting due to the radiation and/or underwater environment. A Boiling Water Reactor (BWR), a Pressurized Water Reactor (PWR), and a High Temperature Gas Reactor (HTGR) have had their internals removed and segmented using PAC and MDM. The cutting technology used for each component, location of cut, cut geometry and environment had to be determined well before the actual cutting operations. This allowed for the design, fabrication, and testing of the delivery systems. The technologies, selection process, and methodology for RV internals segmentation will be discussed in this paper

An Overview of the International Reactor Physics Experiment Evaluation Project

International Nuclear Information System (INIS)

Briggs, J. Blair; Gulliford, Jim

2014-01-01

Interest in high-quality integral benchmark data is increasing as efforts to quantify and reduce calculational uncertainties associated with advanced modeling and simulation accelerate to meet the demands of next generation reactor and advanced fuel cycle concepts. Two Organization for Economic Cooperation and Development (OECD) Nuclear Energy Agency (NEA) activities, the International Criticality Safety Benchmark Evaluation Project (ICSBEP), initiated in 1992, and the International Reactor Physics Experiment Evaluation Project (IRPhEP), initiated in 2003, have been identifying existing integral experiment data, evaluating those data, and providing integral benchmark specifications for methods and data validation for nearly two decades. Data provided by those two projects will be of use to the international reactor physics, criticality safety, and nuclear data communities for future decades. An overview of the IRPhEP and a brief update of the ICSBEP are provided in this paper.

Crosscutting Requirements in the International Project on Innovative Reactors and Fuel Cycles (INPRO)

International Nuclear Information System (INIS)

Steur, Ronald; Lyubenov Yaven, Yanko; Gueorguiev, Boris; Mahadeva, Rao; Shen, Wenquan

2002-01-01

There are two categories of requirements: (i) user requirements that need to be met by the designers and manufacturers of innovative reactors and fuel cycles, and (ii) a wide spectrum of requirements that need to be met by countries, willing to successfully deploy innovative nuclear reactors for energy production. This part of the International Project on Innovative Reactors and Fuel Cycles will mainly deal with the second category of requirements. Both categories of requirements will vary depending on the institutional development, infrastructure availability and social attitude in any given country. Out of the need for sustainable development requirements will also more specific in the future. Over a 50-year time frame both categories of requirements will evolve with social and economic development as nuclear technology develops further. For example, the deployment of innovative reactors in countries with marginal or non-existing nuclear infrastructures would be possible only if the reactors are built, owned and operated by an international nuclear utility or if they are inherently safe and can be delivered as a 'black box - nuclear battery'. A number of issues will need to be addressed and conditions and requirements developed if this is going to become a reality. One general requirement for wider utilization of innovative nuclear power will be the public and environmental considerations, which will play a role in the decision making processes. Five main clusters of topics will be handled: - Infra-structural aspects, typology and consequences for nuclear development. - Industrial requirements for the different innovative concepts. - Institutional developments and requirements for future deployment of nuclear energy. (National as well as international) - Socio-political aspects, a.o. public acceptance and role of governments. - Sustainability: requirements following the need for sustainability Analysis will be made of the evolution of national and international

International Working Group on Fast Reactors Second Annual Meeting. Summary Report

International Nuclear Information System (INIS)

1969-01-01

The Agenda of the Meeting was as follows: Opening of the meeting. 2. Appraisal of the IWGFB's activity for the period from the first annual meeting of the Group. 3. Comments on national programmes on fast breeder reactors. 4. Presentation of general findings and conclusions of national and regional meetings on fast reactor problems held in represented countries and international organisations last year. 5. Comments on the programmes of international meetings on fast reactors to be held in 1969. 6. Consideration of a schedule for meetings on fast reactors in 1970. 7. Suggestions for the topics and location of specialists' meetings in 1969-1970. 8. Suggestions for reviews and studies in the field of fast reactors. 9. The time and place of the third annual meeting of the IWGFR. 10. Closing of the meeting

Vibration behavior of PWR reactor internals Model experiments and analysis

International Nuclear Information System (INIS)

Assedo, R.; Dubourg, M.; Livolant, M.; Epstein, A.

1975-01-01

In the late 1971, the CEA and FRAMATOME decided to undertake a comprehensive joint program of studying the vibration behavior of PWR internals of the 900 MWe, 50 cycle, 3 loop reactor series being built by FRAMATOME in France. The PWR reactor internals are submitted to several sources of excitation during normal operation. Two main sources of excitation may effect the internals behavior: the large flow turbulences which could generate various instabilities such as: vortex shedding: the pump pressure fluctuations which could generate acoustic noise in the circuit at frequencies corresponding to shaft speed frequencies or blade passing frequencies, and their respective harmonics. The flow induced vibrations are of complex nature and the approach selected, for this comprehensive program, is semi-empirical and based on both theoretical analysis and experiments on a reduced scale model and full scale internals. The experimental support of this program consists of: the SAFRAN test loop which consists of an hydroelastic similitude of a 1/8 scale model of a PWR; harmonic vibration tests in air performed on full scale reactor internals in the manufacturing shop; the GENNEVILLIERS facilities which is a full flow test facility of primary pump; the measurements carried out during start up on the Tihange reactor. This program will be completed in April 1975. The results of this program, the originality of which consists of studying separately the effects of random excitations and acoustic noises, on the internals behavior, and by establishing a comparison between experiments and analysis, will bring a major contribution for explaining the complex vibration phenomena occurring in a PWR

Boiling-Water Reactor internals aging degradation study

International Nuclear Information System (INIS)

Luk, K.H.

1993-09-01

This report documents the results of an aging assessment study for boiling water reactor (BWR) internals. Major stressors for BWR internals are related to unsteady hydrodynamic forces generated by the primary coolant flow in the reactor vessel. Welding and cold-working, dissolved oxygen and impurities in the coolant, applied loads and exposures to fast neutron fluxes are other important stressors. Based on results of a component failure information survey, stress corrosion cracking (SCC) and fatigue are identified as the two major aging-related degradation mechanisms for BWR internals. Significant reported failures include SCC in jet-pump holddown beams, in-core neutron flux monitor dry tubes and core spray spargers. Fatigue failures were detected in feedwater spargers. The implementation of a plant Hydrogen Water Chemistry (HWC) program is considered as a promising method for controlling SCC problems in BWR. More operating data are needed to evaluate its effectiveness for internal components. Long-term fast neutron irradiation effects and high-cycle fatigue in a corrosive environment are uncertainty factors in the aging assessment process. BWR internals are examined by visual inspections and the method is access limited. The presence of a large water gap and an absence of ex-core neutron flux monitors may handicap the use of advanced inspection methods, such as neutron noise vibration measurements, for BWR

Advances in fast reactor technology. Proceedings of the 30. meeting of the International Working Group on Fast Reactors

International Nuclear Information System (INIS)

1998-04-01

Individual States were largely responsible for early developments in experimental and prototype liquid metal fast reactors (LMFRs). However, for development of advanced LMFRs, international co-operation plays an important role. The IAEA seeks to promote such co-operation. For R and D incorporating innovative features, international co-operation allows pooling of resources and expertise in areas of common interest. Information on experience gained from R and D, and from the operation and construction of fast reactors, has been reviewed periodically by the International Working Group on Fast Reactors (IWGFR). These proceedings contain updated a new information on the status of LMFR development, as reported at the 30th meeting of the IWGFR, held in Beijing, China, from 13 to 16 May 1997

Advances in fast reactor technology. Proceedings of the 30. meeting of the International Working Group on Fast Reactors

Energy Technology Data Exchange (ETDEWEB)

NONE

1998-04-01

Individual States were largely responsible for early developments in experimental and prototype liquid metal fast reactors (LMFRs). However, for development of advanced LMFRs, international co-operation plays an important role. The IAEA seeks to promote such co-operation. For R and D incorporating innovative features, international co-operation allows pooling of resources and expertise in areas of common interest. Information on experience gained from R and D, and from the operation and construction of fast reactors, has been reviewed periodically by the International Working Group on Fast Reactors (IWGFR). These proceedings contain updated a new information on the status of LMFR development, as reported at the 30th meeting of the IWGFR, held in Beijing, China, from 13 to 16 May 1997. Refs,figs,tabs.

Proceedings of the international symposium on research reactor safety operations and modifications

International Nuclear Information System (INIS)

1990-03-01

The International Symposium on Research Reactor Safety, Operations and Modifications was organized by the International Atomic Energy Agency in cooperation with Atomic Energy of Canada Limited-Research Company. The main objectives of this Symposium were: (1) to exchange information and to discuss current perspectives and concerns relating to all aspects to research reactor safety, operations, and modifications; and, (2) to present views and to discuss future initiatives and directions for research reactor design, operations, utilization, and safety. The symposium topics included: research reactor programmes and experience; research reactor design safety and analysis; research reactor modifications and decommissioning; research reactor licensing; and new research reactors. These topics were covered during eight oral sessions and three poster sessions. These Proceedings include the full text of the 93 papers presented. The subject of Symposium was quite wide-ranging in that it covered essentially all aspects of research reactor safety, operations, and modifications. This was considered to be appropriate and timely given the 326 research reactors currently in operation in some 56 countries; given the degree of their utilization which ranges from pure and applied research to radioisotopes production to basic training and manpower development; and given that many of these reactors are undergoing extensive modifications, core conversions, power upratings, and are becoming the subject of safety reassessment and regulatory reviews. Although the Symposium covered many topics, the majority of papers and discussions tended to focus mainly on research reactor safety. This was seen as a clear sign of the continuing recognition of the fundamental importance of identifying and addressing, particularly through international cooperation, issues and concerns associated with research reactor safety

Fluid-structure-interaction analyses of reactor vessel using improved hybrid Lagrangian Eulerian code ALICE-II

Energy Technology Data Exchange (ETDEWEB)

Wang, C.Y.

1993-06-01

This paper describes fluid-structure-interaction and structure response analyses of a reactor vessel subjected to loadings associated with postulated accidents, using the hybrid Lagrangian-Eulerian code ALICE-II. This code has been improved recently to accommodate many features associated with innovative designs of reactor vessels. Calculational capabilities have been developed to treat water in the reactor cavity outside the vessel, internal shield structures and internal thin shells. The objective of the present analyses is to study the cover response and potential for missile generation in response to a fuel-coolant interaction in the core region. Three calculations were performed using the cover weight as a parameter. To study the effect of the cavity water, vessel response calculations for both wet- and dry-cavity designs are compared. Results indicate that for all cases studied and for the design parameters assumed, the calculated cover displacements are all smaller than the bolts` ultimate displacement and no missile generation of the closure head is predicted. Also, solutions reveal that the cavity water of the wet-cavity design plays an important role of restraining the downward displacement of the bottom head. Based on these studies, the analyses predict that the structure integrity is maintained throughout the postulated accident for the wet-cavity design.

Fluid-structure-interaction analyses of reactor vessel using improved hybrid Lagrangian Eulerian code ALICE-II

Energy Technology Data Exchange (ETDEWEB)

Wang, C.Y.

1993-01-01

This paper describes fluid-structure-interaction and structure response analyses of a reactor vessel subjected to loadings associated with postulated accidents, using the hybrid Lagrangian-Eulerian code ALICE-II. This code has been improved recently to accommodate many features associated with innovative designs of reactor vessels. Calculational capabilities have been developed to treat water in the reactor cavity outside the vessel, internal shield structures and internal thin shells. The objective of the present analyses is to study the cover response and potential for missile generation in response to a fuel-coolant interaction in the core region. Three calculations were performed using the cover weight as a parameter. To study the effect of the cavity water, vessel response calculations for both wet- and dry-cavity designs are compared. Results indicate that for all cases studied and for the design parameters assumed, the calculated cover displacements are all smaller than the bolts' ultimate displacement and no missile generation of the closure head is predicted. Also, solutions reveal that the cavity water of the wet-cavity design plays an important role of restraining the downward displacement of the bottom head. Based on these studies, the analyses predict that the structure integrity is maintained throughout the postulated accident for the wet-cavity design.

Summary of SMIRT20 Preconference Topical Workshop - Identifying Structural Issues in Advanced Reactors

International Nuclear Information System (INIS)

Richins, William; Novascone, Stephen; O'Brien, Cheryl

2009-01-01

The Idaho National Laboratory (INL, USA) and IASMiRT sponsored an international forum Nov 5-6, 2008 in Porvoo, Finland for nuclear industry, academic, and regulatory representatives to identify structural issues in current and future advanced reactor design, especially for extreme conditions and external threats. The purpose of this Topical Workshop was to articulate research, engineering, and regulatory Code development needs. The topics addressed by the Workshop were selected to address critical industry needs specific to advanced reactor structures that have long lead times and can be the subject of future SMiRT technical sessions. The topics were; (1) structural/materials needs for extreme conditions and external threats in contemporary (Gen. III) and future (Gen. IV and NGNP) advanced reactors and (2) calibrating simulation software and methods that address topic 1. The workshop discussions and research needs identified are presented. The Workshop successfully produced interactive discussion on the two topics resulting in a list of research and technology needs. It is recommended that IASMiRT communicate the results of the discussion to industry and researchers to encourage new ideas and projects. In addition, opportunities exist to retrieve research reports and information that currently exists, and encourage more international cooperation and collaboration. It is recommended that IASMiRT continue with an off-year workshop series on select topics.

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)

Repair and preventive maintenance technology for BWR reactor internals and piping

International Nuclear Information System (INIS)

Ootsubo, Tooru; Itou, Takashi; Sakashita, Akihiro

2009-01-01

Stress corrosion cracking of welding portion has found in many domestic and foreign BWR reactor internals and Primary Loop Recirculation piping. Also, repair and preventive maintenance technologies for SCC has been developed and/or adopted to BWRs in recent years. This paper introduces the sample of these technologies, such as seal-welding for SCC on BWR reactor internals, preventive maintenance technology for PLR piping such as Corrosion Resistant Cladding, Internal Polishing and Induction Heating Stress Improvement. These technologies are introduced on 'E-Journal of Advanced Maintenance', which is an international journal on a exclusive website of Japan Society of Maintenology. (author)

The international thermonuclear reactor (ITER)

International Nuclear Information System (INIS)

Fowler, T.K.; Henning, C.D.

1987-01-01

Four governmental groups, representing Europe, Japan, USSR and U.S. met in March 1987 to consider a new international design of a magnetic fusion device for the 1990's. An interim group was appointed. The author gives a brief synopsis of what might be thought of as a draft charter. The starting point is the objective of the ITER device, which is summarized as demonstrating both scientific and technical feasibility of fusion. The paper presents an update on the current thinking and technical aspects for the International Thermonuclear Experimental Reactor (ITER). This covers not only what is happening in the U.S. but also some reports of preliminary thinking of the last technical work that occurred in Vienna

Maintenance technologies for reactor internals

Energy Technology Data Exchange (ETDEWEB)

Sato, Kenji [Nuclear Energy Systems and Services Div., Toshiba Corp., Tokyo (Japan); Kobayashi, Masahiro [Toshiba Corp., Yokohama (Japan). Keihin Product Operations; Sano, Yuji; Kimura, Seiichiro [Power and Industrial Systems Research and Development Center, Toshiba Corp., Tokyo(Japan)

2000-10-01

Toshiba places the highest priority on maintenance technologies for the reactor pressure vessel (RPV) and its internals in operating nuclear power plants. This paper summarizes the status of applied laser maintenance technologies, both preventive and repair. For laser peeing and laser desensitization treatment (LDT) technologies in particular, field applications are also described in detail. In the future, the area of field applications for preventive maintenance, repair, and inspection technologies will be further expanded. (author)

Reactors licensing: proposal of an integrated quality and environment regulatory structure for nuclear research reactors in Brazil

International Nuclear Information System (INIS)

Serra, Reynaldo Cavalcanti

2014-01-01

A new integrated regulatory structure based on quality and integrated issues has been proposed to be implemented on the licensing process of nuclear research reactors in Brazil. The study starts with a literature review about the licensing process in several countries, all of them members of the International Atomic Energy Agency. After this phase it is performed a comparative study with the Brazilian licensing process to identify good practices (positive aspects), the gaps on it and to propose an approach of an integrated quality and environmental management system, in order to contribute with a new licensing process scheme in Brazil. The literature review considered the following research nuclear reactors: Jules-Horowitz and OSIRIS (France), Hanaro (Korea), Maples 1 and 2 (Canada), OPAL (Australia), Pallas (Holand), ETRR-2 (Egypt) and IEA-R1 (Brazil). The current nuclear research reactors licensing process in Brazil is conducted by two regulatory bodies: the Brazilian National Nuclear Energy Commission (CNEN) and the Brazilian Institute of Environment and Renewable Natural Resources (IBAMA). CNEN is responsible by nuclear issues, while IBAMA by environmental one. To support the study it was applied a questionnaire and interviews based on the current regulatory structure to four nuclear research reactors in Brazil. Nowadays, the nuclear research reactor’s licensing process, in Brazil, has six phases and the environmental licensing process has three phases. A correlation study among these phases leads to a proposal of a new quality and environmental integrated licensing structure with four harmonized phases, hence reducing potential delays in this process. (author)

Summary remarks and recommended reactions for an international data file for dosimetry applications for LWR, FBR, and MFR reactor research, development and testing programs

International Nuclear Information System (INIS)

McElroy, W.N.; Lippincott, E.P.; Grundl, J.A.; Fabry, A.; Dierckx, R.; Farinelli, U.

1979-01-01

The need for the use of an internationally accepted data file for dosimetry applications for light water reactor (LWR), fast breeder reactor (FBR), and magnetic fusion reactor (MFR) research, development, and testing programs continues to exist for the Nuclear Industry. The work of this IAEA meeting, therefore, will be another important step in achieving consensus agreement on an internationally recommended file and its purpose, content, structure, selected reactions, and associated uncertainy files. Summary remarks and a listing of recommended reactions for consideration in the formulation of an ''International Data File for Dosimetry Applications'' are presented in subsequent sections of this report

Fluid structure interaction studies on acoustic load response of light water nuclear reactor core internals under blowdown condition

International Nuclear Information System (INIS)

Moses Lemuel Raj, G.; Singh, R.K.; Kushwaha, H.S.; Venkat Raj, V.

1998-12-01

Acoustic load evaluation within two phase medium and the related fluid-structure interaction analysis in case of Loss of Coolant Accidents (LOCA) for light water reactor systems is an important inter-disciplinary area. The present work highlights the development of a three-dimensional finite element code FLUSHEL to analyse LOCA induced depressurization problems for Pressurised Water Reactor (PWR) core barrel and Boiling Water Reactor (BWR) core shroud. With good comparison obtained between prediction made by the present code and the experimental results of HDR-PWR test problem, coupled fluid-structure interaction analysis of core shroud of Tarapur Atomic Power Station (TAPS) is presented for recirculation line break. It is shown that the acoustic load induced stresses in the core shroud are small and downcomer acoustic cavity modes are decoupled with the shell multi-lobe modes. Thus the structural integrity of TAPS core shroud for recirculation line break induced acoustic load is demonstrated. (author)

Baffle-former arrangement for nuclear reactor vessel internals

International Nuclear Information System (INIS)

Machado, O.J.; Berringer, R.T.

1978-01-01

A baffle-former arrangement for the reactor vessel internals of a nuclear reactor is described. The arrangement includes positioning of formers at the same elevations as the fuel assembly grids, and positioning flow holes in the baffle plates directly beneath selected former grid elevations. The arrangement reduces detrimental cross flows, maintains proper core barrel and baffle temperatures, and alleviates the potential of overpressurization within the baffle-former assembly under assumed major accident conditions

Sloshing response of a reactor tank with internals

International Nuclear Information System (INIS)

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

1984-01-01

The sloshing response of a large reactor tank with in-tank components is presented. The study indicates that the presence of the internal components can significantly change the dynamic characteristics of the sloshing motion. The sloshing frequency of a tank with internals is considerably higher than that of a tank without internal. The higher sloshing frequency reduces the sloshing wave height on the free-surface but increases the dynamic pressure in the fluid

International Reactor Physics Experiment Evaluation (IRPhE) Project. IRPhE Handbook - 2015 edition

International Nuclear Information System (INIS)

Bess, John D.; Gullifor, Jim

2015-03-01

The purpose of the International Reactor Physics Experiment Evaluation (IRPhE) Project is to provide an extensively peer-reviewed set of reactor physics-related integral data that can be used by reactor designers and safety analysts to validate the analytical tools used to design next-generation reactors and establish the safety basis for operation of these reactors. This work of the IRPhE Project is formally documented in the 'International Handbook of Evaluated Reactor Physics Benchmark Experiments', a single source of verified and extensively peer-reviewed reactor physics benchmark measurements data. The evaluation process entails the following steps: Identify a comprehensive set of reactor physics experimental measurements data, Evaluate the data and quantify overall uncertainties through various types of sensitivity analysis to the extent possible, verify the data by reviewing original and subsequently revised documentation, and by talking with the experimenters or individuals who are familiar with the experimental facility, Compile the data into a standardized format, Perform calculations of each experiment with standard reactor physics codes where it would add information, Formally document the work into a single source of verified and peer reviewed reactor physics benchmark measurements data. The International Handbook of Evaluated Reactor Physics Benchmark Experiments contains reactor physics benchmark specifications that have been derived from experiments that were performed at nuclear facilities around the world. The benchmark specifications are intended for use by reactor designers, safety analysts and nuclear data evaluators to validate calculation techniques and data. Example calculations are presented; these do not constitute a validation or endorsement of the codes or cross-section data. The 2015 edition of the International Handbook of Evaluated Reactor Physics Benchmark Experiments contains data from 143 experimental series that were

International cooperation on breeder reactors

International Nuclear Information System (INIS)

Gray, J.E.; Kratzer, M.B.; Leslie, K.E.; Paige, H.W.; Shantzis, S.B.

1978-01-01

In March 1977, as the result of discussions which began in the fall of 1976, the Rockefeller Foundation requested International Energy Associates Limited (IEAL) to undertake a study of the role of international cooperation in the development and application of the breeder reactor. While there had been considerable international exchange in the development of breeder technology, the existence of at least seven major national breeder development programs raised a prima facie issue of the adequacy of international cooperation. The final product of the study was to be the identification of options for international cooperation which merited further consideration and which might become the subject of subsequent, more detailed analysis. During the course of the study, modifications in U.S. breeder policy led to an expansion of the analysis to embrace the pros and cons of the major breeder-related policy issues, as well as the respective views of national governments on those issues. The resulting examination of views and patterns of international collaboration emphasizes what was implicit from the outset: Options for international cooperation cannot be fashioned independently of national objectives, policies and programs. Moreover, while similarity of views can stimulate cooperation, this cannot of itself provide compelling justification for cooperative undertakings. Such undertakings are influenced by an array of other national factors, including technological development, industrial infrastructure, economic strength, existing international ties, and historic experience

International standardization of safety requirements for fast reactors

International Nuclear Information System (INIS)

2011-06-01

Japan Atomic Energy Agency (JAEA) is conducting the FaCT (Fast Reactor Cycle Technology Development) project in cooperation with Japan Atomic Power Company (JAPC) and Mitsubishi FBR systems inc. (MFBR), where an advanced loop-type fast reactor named JSFR (Japan Sodium-cooled Fast Reactor) is being developed. It is important to develop software technologies (a safety guideline, safety design criteria, safety design standards etc.) of FBRs as well as hardware ones (a reactor plant itself) in order to address prospective worldwide utilization of FBR technology. Therefore, it is expected to establish a rational safety guideline applicable to the JSFR and harmonized with national nuclear-safety regulations as well, including Japan, the United States and the European Union. This report presents domestic and international status of safety guideline development for sodium-cooled fast reactors (SFRs), results of comparative study for safety requirements provided in existing documents and a proposal for safety requirements of future SFRs with a roadmap for their refinement and worldwide utilization. (author)

Radiation Dosimetry of the Pressure Vessel Internals of the High Flux Beam Reactor

Science.gov (United States)

Holden, Norman E.; Reciniello, Richard N.; Hu, Jih-Perng; Rorer, David C.

2003-06-01

In preparation for the eventual decommissioning of the High Flux Beam Reactor after the permanent removal of its fuel elements from the Brookhaven National Laboratory, both measurements and calculations of the decay gamma-ray dose rate have been performed for the reactor pressure vessel and vessel internal structures which included the upper and lower thermal shields, the Transition Plate, and the Control Rod blades. The measurements were made using Red Perspex™ polymethyl methacrylate high-level film dosimeters, a Radcal "peanut" ion chamber, and Eberline's high-range ion chamber. To compare with measured gamma-ray dose rates, the Monte Carlo MCNP code and geometric progressive MicroShield code were used to model the gamma-ray transport and dose buildup.

RADIATION DOSIMETRY OF THE PRESSURE VESSEL INTERNALS OF THE HIGH FLUX BEAM REACTOR

International Nuclear Information System (INIS)

HOLDEN, N.E.; RECINIELLO, R.N.; HU, J.P.; RORER, D.C.

2002-01-01

In preparation for the eventual decommissioning of the High Flux Beam Reactor after the permanent removal of its fuel elements from the Brookhaven National Laboratory, both measurements and calculations of the decay gamma-ray dose rate have been performed for the reactor pressure vessel and vessel internal structures which included the upper and lower thermal shields, the transition plate, and the control rod blades. The measurements were made using Red Perspex(trademark) polymethyl methacrylate high-level film dosimeters, a Radcal ''peanut'' ion chamber, and Eberline's high-range ion chamber. To compare with measured gamma-ray dose rate, the Monte Carlo MCNP code and geometric progressive Microshield code were used to model the gamma transport and dose buildup

Experience with mechanical segmentation of reactor internals

International Nuclear Information System (INIS)

Carlson, R.; Hedin, G.

2003-01-01

Operating experience from BWE:s world-wide has shown that many plants experience initial cracking of the reactor internals after approximately 20 to 25 years of service life. This ''mid-life crisis'', considering a plant design life of 40 years, is now being addressed by many utilities. Successful resolution of these issues should give many more years of trouble-free operation. Replacement of reactor internals could be, in many cases, the most favourable option to achieve this. The proactive strategy of many utilities to replace internals in a planned way is a market-driven effort to minimize the overall costs for power generation, including time spent for handling contingencies and unplanned outages. Based on technical analyses, knowledge about component market prices and in-house costs, a cost-effective, optimized strategy for inspection, mitigation and replacements can be implemented. Also decommissioning of nuclear plants has become a reality for many utilities as numerous plants worldwide are closed due to age and/or other reasons. These facts address a need for safe, fast and cost-effective methods for segmentation of internals. Westinghouse has over the last years developed methods for segmentation of internals and has also carried out successful segmentation projects. Our experience from the segmentation business for Nordic BWR:s is that the most important parameters to consider when choosing a method and equipment for a segmentation project are: - Safety, - Cost-effectiveness, - Cleanliness, - Reliability. (orig.)

Hydrodynamic study of an internal airlift reactor for microalgae culture.

Science.gov (United States)

Rengel, Ana; Zoughaib, Assaad; Dron, Dominique; Clodic, Denis

2012-01-01

Internal airlift reactors are closed systems considered today for microalgae cultivation. Several works have studied their hydrodynamics but based on important solid concentrations, not with biomass concentrations usually found in microalgae cultures. In this study, an internal airlift reactor has been built and tested in order to clarify the hydrodynamics of this system, based on microalgae typical concentrations. A model is proposed taking into account the variation of air bubble velocity according to volumetric air flow rate injected into the system. A relationship between riser and downcomer gas holdups is established, which varied slightly with solids concentrations. The repartition of solids along the reactor resulted to be homogenous for the range of concentrations and volumetric air flow rate studied here. Liquid velocities increase with volumetric air flow rate, and they vary slightly when solids are added to the system. Finally, liquid circulation time found in each section of the reactor is in concordance with those employed in microalgae culture.

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

International Nuclear Information System (INIS)

1988-06-01

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

Nuclear reactor

International Nuclear Information System (INIS)

Gibbons, J.F.; McLaughlin, D.J.

1978-01-01

In the pressure vessel of the water-cooled nuclear reactor there is provided an internal flange on which the one- or two-part core barrel is hanging by means of an external flange. A cylinder is extending from the reactor vessel closure downwards to a seat on the core cupport structure and serves as compression element for the transmission of the clamping load from the closure head to the core barrel (upper guide structure). With the core barrel, subject to tensile stress, between the vessel internal flange and its seat on one hand and the compression of the cylinder resp. hold-down element between the closure head and the seat on the other a very strong, elastic sprung structure is obtained. (DG) [de

Proceedings 21. International Conference on Applied Physics of Condensed Matter and of the Scientific Conference Advanced Fast Reactors

International Nuclear Information System (INIS)

Vajda, J.; Jamnicky, I.

2015-01-01

The 21. International Conference on Applied Physics of Condensed Matter was held on 24-26 June, 2015 on Strbske Pleso, Strba, Slovakia. The Scientific Conference the Advanced Fast Reactors was part of the 21 st International Conference on APCOM 2015. The specialists discussed various aspects of modern problems in: Physical properties and structural aspects of solid materials and their influencing; Advanced fast reactors; Physical properties and structural aspects of solid materials and their influencing; Nuclear science and technology, influence of irradiation on physical properties of materials, radiation detection; Computational physics and theory of physical properties of matter; interdisciplinary physics of condensed matter; Nuclear science and technology, influence of irradiation on physical properties of materials, radiation detection; Optical phenomena in materials, photovoltaics and photonics, new principles in sensors and detection methods. Fifty seven contributions relevant of INIS interest has been inputted to INIS.

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)

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

International collaboration between nuclear research centres and the role of research reactors

International Nuclear Information System (INIS)

Dodd, B.

2001-01-01

A research reactor is a core facility in many nuclear research centres (NRCs) of Member States and it is logical that it should be the focus of any international collaboration between such centres. There are several large and sophisticated research reactors in operation in both developed and developing Member States, such as Belgium, China, Egypt, France, Hungary, Indonesia, India, Japan, ROK, Netherlands, South Africa and the USA. There are also several new, large reactors under construction or being planned such as those in Australia, Canada, China, France, Germany, and Thailand. It is felt that the utilization of these reactors can be enhanced by international co-operation to achieve common goals in research and applications. (author)

International standardization of nuclear reactor designs - the way forward

International Nuclear Information System (INIS)

Raetzke, Christian

2010-01-01

The concept of 'International Standardization of Nuclear Reactor Designs' means that vendors could build their designs in every country without having to adapt it specifically to national safety requirements. Such standardization would have two main effects. It would greatly facilitate nuclear new build worldwide by giving greater efficiency and certainty to the national licensing procedures; by taking into account the fact that vendors, and nowadays also utilities, are active across borders; by helping developing countries to establish their nuclear new build programmes; and by reducing the strain on human resources on both the regulators' and the industry's side. The second valuable effect of standardization would be to further enhance safety by improving the exchange of construction and operating experience among a number of reactors belonging to fleets of the same design. The World Nuclear Association's CORDEL (Cooperation in Reactor Design Evaluation and Licensing) Group has developed a concept for implementation of international standardization of reactor designs. It has defined a number of steps to be taken by industry. At the same time, possibilities offered by national and international regulatory mechanisms would have to be fully made use of, and some changes in regulatory frameworks might be necessary. Some steps especially towards greater cooperation of regulators have already been taken; however, much still remains to be done. The concept of deploying standardized reactor designs across a number of countries supposes an alignment and, if possible, harmonization of national safety standards; a streamlining of national licensing procedures, making them more efficient and predictable; and the willingness of national regulators to take into account licensing done in other countries. In the end, this should lead to a mutual acceptance of design approvals or, in a more distant future, even to a multinational design approval process. All in all, the concept

Moderator behaviour and reactor internals integrity at Atucha I NPP

International Nuclear Information System (INIS)

Berra, S.; Guala, M.; Herzovich, P.; Chocron, M.; Lorenzo, A.; Raffo Calderon, Ma. C. del; Urrutia, G.

1996-01-01

Atucha I is a Pressure Vessel Heavy Water Cooled Heavy Water Moderator Reactor. In this kind of reactor the moderator tank is physically connected to the primary coolant. Since neutron economy requires the moderator to be as cold as possible, it is necessary that even when physically connected, it should have a separated cooling system, which in this case is also used as a feed-water preheater, and also heat mass transfer with primary coolant should be minimized. This condition requires that some reactor internals are designed in principle to last the whole life of the plant. However, in 1988 the failure of one internal produced a 16 month shut down. This incident could have been prevented but the idea that reactor internals would not have failures due to aging was dominant at that time avoiding the early detection of the failure. However, the analysis of the records after the incident showed that some process variables had changed previously to the incident, i.e., power exchanged at the moderator heat exchanger had increased. Since the station restart up some changes in the moderator process variables and a flow rate reduction of about 10% through the primary side of one moderator cooler were observed. In order to understand the flow reduction and the overall behaviour of moderators parameters, two models were developed that predict moderator and moderator cooler behavior under the new conditions. The present paper refers to these models, which together with the improvement of process variables measurements mentioned in another paper presented at this meeting permits to understand current moderator behaviour and helps to early diagnostic of an eventual reactor internal failure. (author). 2 refs, 4 figs, 1 tab

Moderator behaviour and reactor internals integrity at Atucha I NPP

Energy Technology Data Exchange (ETDEWEB)

Berra, S; Guala, M; Herzovich, P [Central Nuclear Atucha I, Nucleoelectrica Argentina, Lima, Buenos Aires (Argentina); Chocron, M; Lorenzo, A; Raffo Calderon, Ma. C. del; Urrutia, G [Comision Nacional de Energia Atomica, Buenos Aires (Argentina). Centro Atomico Constituyentes

1997-12-31

Atucha I is a Pressure Vessel Heavy Water Cooled Heavy Water Moderator Reactor. In this kind of reactor the moderator tank is physically connected to the primary coolant. Since neutron economy requires the moderator to be as cold as possible, it is necessary that even when physically connected, it should have a separated cooling system, which in this case is also used as a feed-water preheater, and also heat mass transfer with primary coolant should be minimized. This condition requires that some reactor internals are designed in principle to last the whole life of the plant. However, in 1988 the failure of one internal produced a 16 month shut down. This incident could have been prevented but the idea that reactor internals would not have failures due to aging was dominant at that time avoiding the early detection of the failure. However, the analysis of the records after the incident showed that some process variables had changed previously to the incident, i.e., power exchanged at the moderator heat exchanger had increased. Since the station restart up some changes in the moderator process variables and a flow rate reduction of about 10% through the primary side of one moderator cooler were observed. In order to understand the flow reduction and the overall behaviour of moderators parameters, two models were developed that predict moderator and moderator cooler behavior under the new conditions. The present paper refers to these models, which together with the improvement of process variables measurements mentioned in another paper presented at this meeting permits to understand current moderator behaviour and helps to early diagnostic of an eventual reactor internal failure. (author). 2 refs, 4 figs, 1 tab.

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

Development of Alloy 718 tubular product for nuclear-reactor internals

International Nuclear Information System (INIS)

1981-01-01

The Upper Internals Structure (UIS) of the Clinch River Breeder Reactor Plant (CRBRP) provides mixing and flow direction of the core outlet flow. Alloy 718 tubes are the major components used in the UIS to provide this flow direction. The UIS is located directly above the reactor core and is exposed to a severe environment. This environment consists of high temperature sodium, alternating temperatures induced by mixing high temperature core assembly outlet flow with cooler core assembly outlet flow and rapid changes in temperature of the core outlet flow. The paper presents the UIS configuration, functions and environmental conditions that led to the selection of Alloy 718 as the material used to protect the basic UIS structure and to provide flow direction. The use of Alloy 718 is derived from the technology produced from the Department of Energy sponsored development programs. Alloy 718 tubes are made by a roll-extrusion process. The paper describes the tube fabrication process, the development of a finish sanding procedure and the results of high temperature thermal cycle testing. The high temperature thermal cycle testing generates peak strains on the metal surfaces, where the surface effects have a maximum influence on the fatigue strength. 9 figs., 2 tabs

3-dimensional finite element modelling of reactor building internal structure for static analysis

International Nuclear Information System (INIS)

Joshi, M.H.; Reddy, V.J.; Kushwaha, H.S.; Reddy, G.R.; Karandikar, G.V.

1991-01-01

a) Thin shell element gives fairly accurate results when compared to 3-D Brick element for the type of structure and loading in Reactor Building. b) The maximum element size is fixed from model 3(c) i.e. 2.0 m. c) Openings with size smaller than 0.5 m can be neglected without affecting the results very much. d) For any such problem, the methodology described in this paper can be used to take rational decisions which will ensure reasonable accuracy. (author)

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

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.

International tokamak reactor conceptual design overview

International Nuclear Information System (INIS)

Stacey, W.M. Jr.

1983-01-01

The International Tokamak Reactor (INTOR) Workshop is an unique collaborative effort among Euratom, Japan, the USA and the USSR, under the auspices of the IAEA, to assess, define, design, construct and operate the next major experiment in the World Tokamak Program beyond the TFTR, JET, JT-60, T-15 generation. During the Zero-Phase (1979), a technical data base assessment was performed, leading to a positive assessment of feasibility. During Phase-I (1/80-6/81), a conceptual design was developed to define the concept. The programmatic objectives are that INTOR should: (1) be the maximum reasonable step beyond the TFTR, JET, JT-60, T-15 generation of tokamaks, (2) demonstrate the plasma performance required for tokamak DEMOs, (3) test the development and integration into a reactor system of those technologies required for a DEMO, (4) serve as a test facility for blanket, tritium production, materials, and plasma engineering technology, (5) test fusion reactor component reliability, (6) test the maintainability of a fusion reactor, and (7) test the factors affecting the reliability, safety and environmental acceptability of a fusion reactor. A conceptual design has been developed to define a device which is consistent with these objectives. The design concept could, with a reasonable degree of confidence, be developed into a workable engineering design of a tokamak that met the performance objectives of INTOR. There is some margin in the design to allow for uncertainty. While design solutions have been found for all of the critical issues, the overall design may not yet be optimal. (author)

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.)

FFTF scale-model characterization of flow-induced vibrational response of reactor internals

International Nuclear Information System (INIS)

Ryan, J.A.; Julyk, L.J.

1977-01-01

As an integral part of the Fast Test Reactor Vibration Program for Reactor Internals, the flow-induced vibrational characteristics of scaled Fast Test Reactor core internal and peripheral components were assessed under scaled and simulated prototype flow conditions in the Hydraulic Core Mockup. The Hydraulic Core Mockup, a 0.285 geometric scale model, was designed to model the vibrational and hydraulic characteristics of the Fast Test Reactor. Model component vibrational characteristics were measured and determined over a range of 36 percent to 111 percent of the scaled prototype design flow. Selected model and prototype components were shaker tested to establish modal characteristics. The dynamic response of the Hydraulic Core Mockup components exhibited no anomalous flow-rate dependent or modal characteristics, and prototype response predictions were adjudged acceptable

FFTF scale-model characterization of flow induced vibrational response of reactor internals

Energy Technology Data Exchange (ETDEWEB)

Ryan, J A; Julyk, L J [Hanford Engineering Development Laboratory, Richland, WA (United States)

1977-12-01

As an integral part of the Fast Test Reactor Vibration Program for Reactor Internals, the flow-induced vibrational characteristics of scaled Fast Test Reactor core internal and peripheral components were assessed under scaled and simulated prototype flow conditions in the Hydraulic Core Mockup. The Hydraulic Core Mockup, a 0.285 geometric scale model, was designed to model the vibrational and hydraulic characteristics of the Fast Test Reactor. Model component vibrational characteristics were measured and determined over a range of 36% to 111% of the scaled prototype design flow. Selected model and prototype components were shaker tested to establish modal characteristics. The dynamic response of the Hydraulic Core Mockup components exhibited no anomalous flow-rate dependent or modal characteristics, and prototype response predictions were adjudged acceptable. (author)

FFTF scale-model characterization of flow induced vibrational response of reactor internals

International Nuclear Information System (INIS)

Ryan, J.A.; Julyk, L.J.

1977-01-01

As an integral part of the Fast Test Reactor Vibration Program for Reactor Internals, the flow-induced vibrational characteristics of scaled Fast Test Reactor core internal and peripheral components were assessed under scaled and simulated prototype flow conditions in the Hydraulic Core Mockup. The Hydraulic Core Mockup, a 0.285 geometric scale model, was designed to model the vibrational and hydraulic characteristics of the Fast Test Reactor. Model component vibrational characteristics were measured and determined over a range of 36% to 111% of the scaled prototype design flow. Selected model and prototype components were shaker tested to establish modal characteristics. The dynamic response of the Hydraulic Core Mockup components exhibited no anomalous flow-rate dependent or modal characteristics, and prototype response predictions were adjudged acceptable. (author)

Technical Requirements For Reactors To Be Deployed Internationally For the Global Nuclear Energy Partnership

International Nuclear Information System (INIS)

Ingersoll, Daniel T.

2007-01-01

The Global Nuclear Energy Partnership (GNEP) seeks to create an international regime to support large-scale growth in the worldwide use of nuclear energy. Fully meeting the GNEP vision may require the deployment of thousands of reactors in scores of countries, many of which do not use nuclear energy currently. Some of these needs will be met by large-scale Generation III and III+ reactors (>1000 MWe) and Generation IV reactors when they are available. However, because many developing countries have small and immature electricity grids, the currently available Generation III(+) reactors may be unsuitable since they are too large, too expensive, and too complex. Therefore, GNEP envisions new types of reactors that must be developed for international deployment that are 'right sized' for the developing countries and that are based on technologies, designs, and policies focused on reducing proliferation risk. The first step in developing such systems is the generation of technical requirements that will ensure that the systems meet both the GNEP policy goals and the power needs of the recipient countries. Reactor systems deployed internationally within the GNEP context must meet a number of requirements similar to the safety, reliability, economics, and proliferation goals established for the DOE Generation IV program. Because of the emphasis on deployment to nonnuclear developing countries, the requirements will be weighted differently than with Generation IV, especially regarding safety and non-proliferation goals. Also, the reactors should be sized for market conditions in developing countries where energy demand per capita, institutional maturity and industrial infrastructure vary considerably, and must utilize fuel that is compatible with the fuel recycle technologies being developed by GNEP. Arrangements are already underway to establish Working Groups jointly with Japan and Russia to develop requirements for reactor systems. Additional bilateral and multilateral

Performance Assessment of Turbulence Models for the Prediction of the Reactor Internal Flow in the Scale-down APR+

International Nuclear Information System (INIS)

Lee, Gonghee; Bang, Youngseok; Woo, Swengwoong; Kim, Dohyeong; Kang, Minku

2013-01-01

The types of errors in CFD simulation can be divided into the two main categories: numerical errors and model errors. Turbulence model is one of the important sources for model errors. In this study, in order to assess the prediction performance of Reynolds-averaged Navier-Stokes (RANS)-based two equations turbulence models for the analysis of flow distribution inside a 1/5 scale-down APR+, the simulation was conducted with the commercial CFD software, ANSYS CFX V. 14. In this study, in order to assess the prediction performance of turbulence models for the analysis of flow distribution inside a 1/5 scale-down APR+, the simulation was conducted with the commercial CFD software, ANSYS CFX V. 14. Both standard k-ε model and SST model predicted the similar flow pattern inside reactor. Therefore it was concluded that the prediction performance of both turbulence models was nearly same. Complex thermal-hydraulic characteristics exist inside reactor because the reactor internals consist of fuel assembly, control rod assembly, and the internal structures. Either flow distribution test for the scale-down reactor model or computational fluid dynamics (CFD) simulation have been conducted to understand these complex thermal-hydraulic features inside reactor

International project on innovative nuclear reactors and fuel cycles

International Nuclear Information System (INIS)

Cherepnin, Yu.S.; Bezzubtsev, V.S.; Gabaraev, B.A.

2002-01-01

international and national measures required for reaching the desired level of innovations in nuclear reactors and fuel cycles using a reliable and economically competitive technology based (as much as possible) on systems with inherent safety features and for the minimizing the proliferation risk and environmental impact. The practical work on the INPRO project was started since January 2001. Currently, ten European, Asian and American countries participate in the the project. The first project stage - preparation of requirements and criteria for evaluation of innovative nuclear power and fuel cycle projects - is nearing completion. The report presents the purposes and objectives of the INPRO project, management principles, current status and organization structure of the project work in Russia

TIBER (Tokamak Ignition/Burn Experimental Reactor) II as a precursor to an international thermonuclear experimental reactor

International Nuclear Information System (INIS)

Henning, C.D.; Gilleland, J.R.

1988-01-01

The Tokamak Ignition/Burn Experimental Reactor (TIBER) was pursued in the US as one option for an International Thermonuclear Experimental Reactor (ITER). This concept evolved from earlier work on the Tokamak Fusion Core Experiment (TFCX) to develop a small, ignited tokamak. While the copper-coil versions of TFCX became the short-pulsed, 1.23-m radius, Compact Ignition Tokamak (CIT), the superconducting TIBER with long pulse or steady state and a 2.6-m radius was considered for international collaboration. Recently the design was updated to TIBER II, to accommodate more conservative confinement scaling, double-poloidal divertors for impurity control, steady-state current drive, and nuclear testing. 18 refs., 1 fig

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)

Ratcheting problems for ITER [International Thermonuclear Experimental Reactor

International Nuclear Information System (INIS)

Majumdar, S.

1991-01-01

Because of the presence of high cyclic thermal stress, pressure-induced primary stress, and disruption-induced high cyclic primary stress, ratcheting of the first wall poses a serious challenge to the designers of ITER (International Thermonuclear Experimental Reactor). Existing design tools such as the Bree diagram in the ASME Boiler and Pressure Vessels Code, are not directly applicable to ITER, because of important differences in geometry and loading modes. Available alternative models for ratcheting are discussed and new Bree diagrams, that are more relevant for fusion reactor applications, are proposed. 9 refs., 17 figs

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

Seismic response analysis of a nuclear reactor structure considering nonlinear soil-structure interaction

International Nuclear Information System (INIS)

Bhaumik, Lopamudra; Raychowdhury, Prishati

2013-01-01

Highlights: • Seismic response analysis of an internal shearwall of a reactor is done. • Incremental dynamic analysis is performed with 30 recorded ground motions. • Equivalent viscous damping increases up to twice when nonlinear SSI is considered. • Roof drift demand increases up to 25% upon consideration of foundation nonlinearity. • Base shear, base moment and ductility reduce up to 62%, 40%, and 35%, respectively. - Abstract: This study focuses on the seismic response analysis of an internal shearwall of a typical Indian reactor resting on a medium dense sandy silty soil, incorporating the nonlinear behavior of the soil-foundation interface. The modeling is done in an open-source finite element framework, OpenSees, where the soil-structure interaction (SSI) is modeled using a Beam-on-Nonlinear-Winkler-Foundation (BNWF) approach. Static pushover analysis and cyclic analysis are performed followed by an incremental dynamic analysis (IDA) with 30 recorded ground motions. For performing IDA, the spectral acceleration of each motion corresponding to the fundamental period, S a (T 1 )is incremented from 0.1 g to 1.0 g with an increment step of 0.1 g. It is observed from the cyclic analysis that the equivalent viscous damping of the system increases upto twice upon incorporation of inelastic SSI. The IDA results demonstrate that the average peak base shear, base moment and displacement ductility demand reduces as much as 62%, 40%, and 35%, respectively, whereas the roof drift demand increases up to 25% upon consideration of foundation nonlinearity for the highest intensity motion. These observations indicate the need of critical consideration of nonlinear soil-structure interaction as any deficient modeling of the same may lead to an inaccurate estimation of the seismic demands of the structure

Seismic response analysis of a nuclear reactor structure considering nonlinear soil-structure interaction

Energy Technology Data Exchange (ETDEWEB)

Bhaumik, Lopamudra, E-mail: lbhaumi2@illinois.edu [University of Illinois at Urbana-Champaign (United States); Raychowdhury, Prishati, E-mail: prishati@iitk.ac.in [Indian Institute of Technology Kanpur (India)

2013-12-15

Highlights: • Seismic response analysis of an internal shearwall of a reactor is done. • Incremental dynamic analysis is performed with 30 recorded ground motions. • Equivalent viscous damping increases up to twice when nonlinear SSI is considered. • Roof drift demand increases up to 25% upon consideration of foundation nonlinearity. • Base shear, base moment and ductility reduce up to 62%, 40%, and 35%, respectively. - Abstract: This study focuses on the seismic response analysis of an internal shearwall of a typical Indian reactor resting on a medium dense sandy silty soil, incorporating the nonlinear behavior of the soil-foundation interface. The modeling is done in an open-source finite element framework, OpenSees, where the soil-structure interaction (SSI) is modeled using a Beam-on-Nonlinear-Winkler-Foundation (BNWF) approach. Static pushover analysis and cyclic analysis are performed followed by an incremental dynamic analysis (IDA) with 30 recorded ground motions. For performing IDA, the spectral acceleration of each motion corresponding to the fundamental period, S{sub a}(T{sub 1})is incremented from 0.1 g to 1.0 g with an increment step of 0.1 g. It is observed from the cyclic analysis that the equivalent viscous damping of the system increases upto twice upon incorporation of inelastic SSI. The IDA results demonstrate that the average peak base shear, base moment and displacement ductility demand reduces as much as 62%, 40%, and 35%, respectively, whereas the roof drift demand increases up to 25% upon consideration of foundation nonlinearity for the highest intensity motion. These observations indicate the need of critical consideration of nonlinear soil-structure interaction as any deficient modeling of the same may lead to an inaccurate estimation of the seismic demands of the structure.

International Conference on Physics and Technology of Reactors and Applications

International Nuclear Information System (INIS)

2011-01-01

Full text : The international conference on physics and technology of reactors is organized by the Moroccan Association for Nuclear enggineering and Reactor Technology (GMTR) with the collaboration of the Centre for Energy and Nuclear Sciences and Techniques (CNESTEN) and under the auspices of the ministry of Energy, Mining, Water and Environment. The programme of the PHYTRA2 conference covers a wide variety of topics. The conference is organised in one plenary session, eight oral technical sessions and one poster session. The oral and poster technical sessions covers the usual topics of nuclear engineering including one session on research reactors utilisation and computational methods for research reactors

Process and apparatus for adjusting a new upper reactor internals in a reactor vessel of a PWR

International Nuclear Information System (INIS)

Frizot, A.; Cadaureille, G.; Lalere, C.; Machuron, J.Y.

1987-01-01

On the new upper reactor internals is mounted devices for alignment and clearances, before introducing in the reactor vessel. After introducing alignment and clearances are measured. Adjustment pieces are provided for optimum clearances and alignment and fixed after removal from vessel. Decontamination is made by using water jets prior to fitting recess parts [fr

Fast nuclear reactors. Associated international projects. State of the art and assessment of the concepts

International Nuclear Information System (INIS)

Azpitarte, O.; Ramilo, L.

2013-01-01

The recognition of the strategic importance of nuclear energy as a source of sustainable energy may be perceived in the continuous development, in many countries, of the technology of fast nuclear reactors with an associated closed fuel cycle, assuming that these Generation IV innovative systems will be required in the future. These reactors fulfill international requirements for safety and reliability, economic competitiveness, sustainability and proliferation resistance. They have the potential of using more efficiently the natural resources of Uranium and of reducing the volume and radiotoxicity of the nuclear waste by partitioning and transmutation of Minor Actinides. The national and international programs being carried out today are concentrated in the following concepts: Sodium Fast Reactor (SFR), Lead Fast Reactor (LFR), Gas Fast Reactor (GFR), Super Critical Water Reactor (SCWR) and Molten Salt Reactor (MSR). This article presents a short review of the technology of the mentioned concepts and details the current state of the main national and international related projects. (author)

Nuclear reactor melt-retention structure to mitigate direct containment heating

International Nuclear Information System (INIS)

Tutu, N.K.; Ginsberg, T.; Klages, J.R.

1991-01-01

This patent describes a nuclear reactor melt-retention structure that functions to retain molten core material within a melt retention chamber to mitigate the extent of direct containment heating. The structure being adapted to be positioned within or adjacent to a pressurized or boiling water nuclear reactor containment building at a location such that at least a portion of the melt retention structure is lower than and to one side of the nuclear reactor pressure vessel, and such that the structure is adjacent to a gas escape channel means that communicates between the reactor cavity and the containment building of the reactor. It comprises a melt-retention chamber, wall means defining a passageway extending between the reactor cavity underneath the reactor pressure vessel and one side of the chamber, the passageway including vent means extending through an upper wall portion thereof. The vent means being in communication with the upper region of the reactor containment building, whereby gas and steam discharged from the reactor pressure vessel are vented through the passageway and vent means into the gas-escape channel means and the reactor containment building

Draft I.E.C. standard for monitoring PWR internal structures; Projet de norme C.E.I. pour la surveillance des structures internes des REP

Energy Technology Data Exchange (ETDEWEB)

Trenty, A.

1994-06-01

EDF has proposed to the International Electrotechnical Commission a draft standard for monitoring the vessel internal structures of PWRs. The standard applies to systems used for monitoring the vibratory behavior of the internal structures of PWRs (core barrel, thermal shield, fuel assemblies) on the basis of neutron fluctuations observed outside the vessel as well as of vessel vibrations. It covers the systems characteristics and the monitoring procedures. It should facilitate standardization of monitoring and comparisons on an international level. This paper presents the main features of the draft standard: -principles of measurement: correlation between movements of internals and ex core neutron noise on the one hand, forced vibrations of the vessel on the other hand; -sampling and conditioning of the signals; -monitoring equipment and in particular spectral analysis device; -functions of the monitoring software used for spectral analysis, peak detection and calculation of structure displacement; -studies preliminary to setting up the monitoring (calculation of internal vibratory modes, defect simulation on mockup, qualification on reactor during hot test...); -monitoring procedures (periodicity of analysis and what to do in case of anomaly); -documentation necessary to the monitoring. A diagnostic procedure is given as an example. The draft standard, written in 1994, will be presented in Frankfurt (Germany) in February 1995. (author). 1 annexe.

Structural design of shield-integrated thin-wall vacuum vessel and manufacturing qualification tests for International Thermonuclear Experimental Reactor (ITER)

International Nuclear Information System (INIS)

Shimizu, Katsusuke; Shibui, Masanao; Koizumi, Koichi; Kanamori, Naokazu; Nishio, Satoshi; Sasaki, Takashi; Tada, Eisuke

1992-09-01

Conceptual design of shield-integrated thin-wall vacuum vessel has been done for ITER (International Thermonuclear Experimental Reactor). The vacuum vessel concept is based on a thin-double-wall structure, which consists of inner and outer plates and rib stiffeners. Internal shielding structures, which provide neutron irradiation shielding to protect TF coils, are set up between the inner plate and the outer plate of the vessel to avoid complexity of machine systems such as supporting systems of blanket modules. The vacuum vessel is assembled/disassembled by remote handling, so that welding joints are chosen as on-site joint method from reliability of mechanical strength. From a view point of assembling TF coils, the vacuum vessel is separated at the side of port, and is divided into 32 segments similar to the ITER-CDA reference design. Separatrix sweeping coils are located in the vacuum vessel to reduce heat fluxes onto divertor plates. Here, the coil structure and attachment to the vacuum vessel have been investigated. A sectorized saddle-loop coil is available for assembling and disassembling the coil. To support electromagnetic loads on the coils, they are attached to the groove in the vacuum vessel by welding. Flexible multi-plate supporting structure (compression-type gravity support), which was designed during CDA, is optimized by investigating buckling and frequency response properties, and concept on manufacturing and fabrication of the gravity support are proposed. Partial model of the vacuum vessel is manufactured for trial, so that fundamental data on welding and fabrication are obtained. From mechanical property tests of weldment and partial models, mechanical intensity and behaviors of the weldment are obtained. Informations on FEM-modeling are obtained by comparing analysis results with experimental results. (author)

Choosing a standard reactor: International competition and domestic politics in Chinese nuclear policy

International Nuclear Information System (INIS)

Ramana, M.V.; Saikawa, Eri

2011-01-01

China has ambitious plans to expand its nuclear power capacity. One of the policy goals that high-level policymakers have desired is to base the nuclear program on a standardized reactor design. However, this has not materialized so far. By examining its nuclear reactor choices for individual projects, we argue that China’s policymaking process has been greatly influenced by international competition and domestic politics. Multiple international nuclear vendors are intent upon maintaining their respective niches in the expanding Chinese reactor market, and they have used various forms of economic and political pressure to achieve their objectives. On the other hand, China’s policymaking process is fragmented and the shifting power balances among powerful domestic actors do not allow a fixed path to be followed. Further, because of the high costs and potential profits involved, nuclear reactor choices in China have been driven not just by technical considerations but also by foreign and trade policy objectives. All of these make it unlikely that China will standardize the reactor type it constructs in the near future. -- Highlights: ► China’s nuclear power policymaking has been fragmented and without central control. ► Multiple domestic actors have pursued independent agendas. ► International nuclear vendors have intensely competed for Chinese reactor contracts. ► Economic, political and foreign policy goals have driven reactor contract decisions. ► China is unlikely to construct only a standardized reactor design.

Status of national programmes on fast reactors 1997/98. 31. annual meeting of the International Working Group on Fast Reactors

International Nuclear Information System (INIS)

1998-01-01

The objective of the meeting was to co-ordinate the exchange of information on the status of fast reactor development and operational experience, including experience with experimental types of reactor; to consider meeting arrangements for 1998 and 1999; and to review the IAEA co-ordinated research activities in the field of fast reactor, as well as co-ordination of the International Working Group on Fast Reactors activities with other organizations

Developments in the regulation of research reactors

International Nuclear Information System (INIS)

Loy, J.

2003-01-01

The International Atomic Energy Agency (IAEA) has data on over 670 research reactors in the world. Fewer than half of them are operational and a significant number are in a shutdown but not decommissioned state. The International Nuclear Safety Advisory Group (INSAG) has expressed concerns about the safety of many research reactors and this has resulted in a process to draw up an international Code of Conduct on the Safety of Research Reactors. The IAEA is also reviewing its safety standards applying to research reactors. On the home front, regulation of the construction of the Replacement Research Reactor continues. During the construction phase, regulation has centred around the consideration of Requests for Approval (RFA) for the manufacture and installation of systems, structures and components important for safety. Quality control of construction of systems, structures and components is the central issue. The process for regulation of commissioning is under consideration

Overview of International Thermonuclear Experimental Reactor (ITER) engineering design activities*

Science.gov (United States)

Shimomura, Y.

1994-05-01

The International Thermonuclear Experimental Reactor (ITER) [International Thermonuclear Experimental Reactor (ITER) (International Atomic Energy Agency, Vienna, 1988), ITER Documentation Series, No. 1] project is a multiphased project, presently proceeding under the auspices of the International Atomic Energy Agency according to the terms of a four-party agreement among the European Atomic Energy Community (EC), the Government of Japan (JA), the Government of the Russian Federation (RF), and the Government of the United States (US), ``the Parties.'' The ITER project is based on the tokamak, a Russian invention, and has since been brought to a high level of development in all major fusion programs in the world. The objective of ITER is to demonstrate the scientific and technological feasibility of fusion energy for peaceful purposes. The ITER design is being developed, with support from the Parties' four Home Teams and is in progress by the Joint Central Team. An overview of ITER Design activities is presented.

The International conference on fast reactors and related fuel cycles: next generation nuclear systems for sustainable development. Book of abstracts

International Nuclear Information System (INIS)

2017-01-01

The materials of the International Conference on Fast Reactors and Related Fuel Cycles (June 26-29, 2017, Yekaterinburg) are presented. The forum was organized by the IAEA with the assistance of Rosatom State Corporation. The theme of the conference: “The New Generation of Nuclear Systems for Sustainable Development”. About 700 specialists from more than 30 countries took part in the conference. The state and prospects for the development of the direction of fast reactors in countries dealing with this topic were discussed. A wide range of scientific issues covered the concepts of prospective reactors, reactor cores, fuel and fuel cycles, operation and decommissioning, safety, licensing, structural materials, industrial implementation [ru

Kazakhstan participation in International Experimental Reactor ITER Construction project. Work status and prospects

International Nuclear Information System (INIS)

Tazhibayeva, I.L.; Tukhvatullin, Sh.T.; Shestakov, V.P.; Kuznetsov, B.A.

2002-01-01

Kazakhstan takes part in ITER project in partnership with Russian Federation since the year of 1994. At present the technical stage of the project is completed and ITER Council should take a decision on the site for international reactor. Four countries such as Canada, Japan, Spain and France have offered their territories for being used as site for launching ITER construction. ITER partners started preparing new international agreement that will cover activities on construction, operation and decommissioning of ITER. It will also include the list of research and experimental work that is conducted in support of ITER project. Kazakhstan has already made an important contribution into technical stage realization of ITER project due to scientific and technical researches conducted by National Nuclear Center, by Institute of Experimental and Theoretical Physics and by JSC 'Ulba Metallurgical plant' ('UMP'). Research activity carried out for the support of ITER project is performed in accordance with the following main trends: Tritium safety (permeability and retentin of hydrogen isotopes during in-pile irradiation in various structural materials, co-deposed layers and protective coatings); Verification of computer codes (LOCA type) loss of coolant accidents modeling in ITER reactor; Investigation of liquid metal blanket of thermonuclear reactor (tritium production in lithium containing eutectics Li17Pb83 and ceramics Li 2 TiO 3 , study of tritium permeability). At present the working group of ITER project participants started introducing proposals for cost distribution and for placing the orders on reactor construction. Further Kazakhstan participation in ITER project may be in manufacturing high-tech parts and assemblies from commercial grades of beryllium. They will be used for armouring the reactor first wall, for its thermal protection and for protection of superconductor's components for magnetic systems that are at JSC U MP'. Scientific and technical support of

Inspection, evaluation and maintenance guidelines for reactor vessel internals in JAPAN

International Nuclear Information System (INIS)

Sakashita, Akihiro; Goto, Tomoya; Hirano, Shinro; Dozaki, Koji

2010-01-01

Inspection and Evaluation Guidelines for reactor internals has been taken into the Rules on Fitness-for–Service for Nuclear Power Plants of The Japan Society of Mechanical Engineers. It is a base of the maintenance plan of each Nuclear Power Plant. A plant maintenance methodology will have more importance to maintain the plant safety and stable plant operation. This paper introduces the systematization of the maintenance such as repair, replacement, preventive maintenance in these guidelines. Maintenance methodologies are classified follows. Repair: methodology to reinforce degraded parts by some methods or prevent progress of degradation of without replacement of the existing structure when the degradation of structure is actualized. Replacement: methodology to replace the existing structure with new one when the degradation of structure is actualized. Preventive maintenance : methodology to mitigate the damaged condition. When the maintenance methodologies are implemented in the actual plant, we have to consider the feedback of the inspection program and plant life management. (author)

Study of the strength of the internal can for internally and externally cooled fuel elements intended for gas graphite reactors

International Nuclear Information System (INIS)

Boudouresque, B.; Courcon, P.; Lestiboubois, G.

1964-01-01

The cartridge of an internally and externally cooled annular fuel element used in gas-graphite reactors is made up of an uranium fuel tube, an external can and an internal can made of magnesium alloy. For the thermal exchange between the internal can and the fuel to be satisfactory, it is necessary for the can to stay in contact with the uranium under all temperature conditions. This report, based on a theoretical study, shows how the internal can fuel gap varies during the processes of canning, charging into the reactor and thermal cycling. The following parameters are considered: tube diameter, pressure of the heat carrying gas, gas entry temperature, plasticity of the can alloy. It is shown that for all operating conditions the internal can of a 77 x 95 element, planned for a gas-graphite reactor with a 40 kg/cm 2 gas pressure, should remain in contact with the fuel. (authors) [fr

The development of reactor vessel internal heavy forging for 1000 MW pressurized-water reactor nuclear power plant

International Nuclear Information System (INIS)

Zhang Zhifeng; Chen Yongbo; Ding Xiuping; Zhang Lingfang

2012-01-01

This Paper introduced the development of Reactor Vessel Internal (RVI) heavy forgings for 1000 MW Pressurized Water Reactor (PWR) nuclear power plant, analyzed the manufacture difficulties and technical countermeasures. The testing result of the product indicated that the performance of RVI heavy forgings manufactured by Shanghai Heavy Machinery Plant Ld. (SHMP) is outstanding and entirely satisfy the technical requirements for RVI product. (authors)

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)

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.

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

Proceedings of the international symposium on materials testing reactors

International Nuclear Information System (INIS)

Ishihara, Masahiro; Kawamura, Hiroshi

2009-01-01

This report is the Proceedings of the International Symposium on Materials Testing Reactors hosted by Japan Atomic Energy Agency (JAEA). The symposium was held on July 16 to 17, 2008, at the Oarai Research and Development Center of JAEA. This symposium was also held for the 40th anniversary ceremony of Japan Materials Testing Reactor (JMTR) from achieving its first criticality. The objective of the symposium is to exchange the information on current status, future plan and so on among each testing reactors for the purpose of mutual understanding. There were 138 participants from Argentina, Belgium, France, Indonesia, Kazakhstan, Korea, the Russian Federation, Sweden, the United State, Vietnam and Japan. The symposium was divided into four technical sessions and three topical sessions. Technical sessions addressed the general topics of 'status and future plan of materials testing reactors', 'material development for research and testing reactors', irradiation technology (including PIE technology)' and 'utilization with materials testing reactors', and 21 presentations were made. Also the topical sessions addressed 'establishment of strategic partnership', 'management on re-operation work at reactor trouble' and 'basic technology for neutron irradiation tests in MTRs', and panel discussion was made. The 21 of the presented papers are indexed individually. (J.P.N.)

Improving nuclear safety at international research reactors: The Integrated Research Reactor Safety Enhancement Program (IRRSEP)

International Nuclear Information System (INIS)

Huizenga, David; Newton, Douglas; Connery, Joyce

2002-01-01

Nuclear energy continues to play a major role in the world's energy economy. Research and test reactors are an important component of a nation's nuclear power infrastructure as they provide training, experiments and operating experience vital to developing and sustaining the industry. Indeed, nations with aspirations for nuclear power development usually begin their programs with a research reactor program. Research reactors also are vital to international science and technology development. It is important to keep them safe from both accident and sabotage, not only because of our obligation to prevent human and environmental consequence but also to prevent corresponding damage to science and industry. For example, an incident at a research reactor could cause a political and public backlash that would do irreparable harm to national nuclear programs. Following the accidents at Three Mile Island and Chernobyl, considerable efforts and resources were committed to improving the safety posture of the world's nuclear power plants. Unsafe operation of research reactors will have an amplifying effect throughout a country or region's entire nuclear programs due to political, economic and nuclear infrastructure consequences. (author)

International centres of excellence based on research reactors

International Nuclear Information System (INIS)

Alldred, K.; Tozser, S.M.; Adelfang, P.

2013-01-01

A number of high flux research reactors were, or will be constructed. Each of these high flux facilities has the potential to be an important regional or International Centre of Excellence based on Research Reactors (ICERR) and scientific hub for research and materials investigations. Some are so organized currently, but for many there is a strongly national focus and scope for a significant expansion of their international role. There are manifold benefits of an expanded international role both for the ICERR's themselves and for the institutes that affiliate with them. These benefits include increased utilization and financial stability, increased international prestige, and enhanced scientific resources and capabilities. There are significant hurdles to obtaining the benefits from an expanded international role. For example, to achieve its full potential an ICERR must accommodate scientists from other nations, and include the plans and aspirations of the international community in the ICERR governance. The ICERR must also fully meet the national responsibilities for safety and security. Balancing these potentially conflicting requirements and finding a path through the organisational and legal issues is a significant challenge for any institute. The existing ICERR's therefore provide important case studies and examples of best practice that could inform the actions of other potential ICERR's. This paper describes an IAEA initiative to encourage and support the formation of new ICERR's, strengthen existing ones, and increase training resources available to Member States. The initiative will seek to share best practice and facilitate meetings and technical exchanges between the existing and potential ICERRs, and between the potential ICERR's and potential subscribing or affiliating institutes. (orig.)

International Centers of Excellence based on Research Reactors

International Nuclear Information System (INIS)

Alldred, K.; Tozser, S. M.; Adelfang, P.

2012-01-01

A number of high flux research reactors were, or will be constructed. Each of these high flux facilities has the potential to be an important regional or International Centre of Excellence based on Research Reactors (ICERR) and scientific hub for research and materials investigations. Some are so organized currently, but for many there is a strongly national focus and scope for a significant expansion of their international role. There are manifold benefits of an expanded international role both for the ICERR's themselves and for the institutes that affiliate with them. These benefits include increased utilization and financial stability, increased international prestige, and enhanced scientific resources and capabilities. There are significant hurdles to obtaining the benefits from an expanded international role. For example, to achieve its full potential an ICERR must accommodate scientists from other nations, and include the plans and aspirations of the international community in the ICERR governance. The ICERR must also fully meet the national responsibilities for safety and security. Balancing these potentially conflicting requirements and finding a path through the organisational and legal issues is a significant challenge for any institute. The existing ICERR's therefore provide important case studies and examples of best practice that could inform the actions of other potential ICERR's. This paper describes an IAEA initiative to encourage and support the formation of new ICERR's, strengthen existing ones, and increase training resources available to Member States. The initiative will seek to share best practice and facilitate meetings and technical exchanges between the existing and potential ICERRs, and between the potential ICERR's and potential subscribing or affiliating institutes. (authors)

ENS RRFM 2005: 9th international topical meeting on research reactor fuel management. Transactions

International Nuclear Information System (INIS)

2005-01-01

The ENS topical meeting on research reactor fuel management is an annual conference launched successfully in 1997. It has since then grown into well established international forum for the exchange and expertise on all significant aspects of the nuclear fuel cycle of research reactors. Oral presentations at this meeting were divided in the following four sessions: International Topics; Fuel Development, Qualification, Fabrication and Licensing; Reactor Operation, Fuel Safety and Core Conversion; Spent Fuel Management, Back-end Options, Transportation. The three poster sessions were devoted to fuel development, qualification, fabrication and licensing; reactor operation, fuel safety, core conversion, spent fuel; spent fuel management, fuel cycle back-end options, transportation

Summary of the fifth international conference on reactor shielding

International Nuclear Information System (INIS)

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

1977-01-01

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

Aging management of PWR reactor internals in U.S. plants

International Nuclear Information System (INIS)

Amberge, K.J.; Demma, A.

2015-01-01

This paper describes the development, aging management strategies and inspection results of the Pressurized Water Reactor (PWR) vessel internals inspection and evaluation guidelines. The goal of these guidelines is to provide PWR owners with robust aging management strategies to monitor degradation of internals components to support life extension as well as the current period of operation and power up-rate activities. The implementation of these guidelines began in 2010 within the U.S. PWR fleet and several examinations have been performed since. Examples of inspection results are presented for selected vessel internals components and are compared with simulation results. In summary, to date there have been no observations of austenitic stainless steel stress corrosion cracking (SCC), which is consistent with expectations based on the current understanding of the mechanism. Observations of irradiation assisted stress corrosion cracking (IASCC) have been limited and only found in baffle former bolting. Additionally, no macroscopic effects or global observations of void swelling impacts on general conditions of reactor internal hardware have been observed. (authors)

International cooperation in accident analysis of RBMK reactors

International Nuclear Information System (INIS)

Kaliatka, A.; Isag

2005-01-01

Chouha Michel (Institute for Radiological Protection and Nuclear Safety), D'Auria Francesco (Institute of Pisa), Kaliatka Algirdas (Lithuanian Energy Institute), Uspuras Eugenijus (Lithuanian Energy Institute). The safety of nuclear power plants is a primary concern of the European Union (EU) and its Member States. In the early 1990s, the European Union decided to take a prominent role in international efforts to help the New Independent States (NIS) and countries of central Europe to ensure the safety of their nuclear reactors. The Commission's approach to nuclear safety in central and Eastern Europe and the NIS is based on two main objectives, which are fully in line with the policy of the international community as decided by the G7 in 1992: 1) In the short term, to improve operational safety; to make near term technical improvements to plants based on safety assessments and to enhance regulatory regimes; 2) In the longer term, to examine the scope for replacing less safe plants by the development of alternative energy sources and more efficient use of energy and to examine the potential for upgrading plants of more recent design. In this paper the safety concerns, related to RBMK type reactors (Russian acronym for 'Channelized Large Power Reactor) are discussed. These reactors were not exported and were built exclusively in the territory of the former Soviet Union. There are presently plants at Saint Petersburg (Sosnovy Bor), Kursk, Chernobyl and Smolensk. A total of 17 such reactors have been built and 12 are currently in operation. Two international projects: TACIS project 'Development of a code system for severe accident analysis in RBMK reactors' and PHARE projects 'Support to VATESI for Important Tasks Relevant to the Licensing Activities of Ignalina Nuclear Power Plant' are presented. The aim of the TACIS project is to help the Russian Authorities to build such capabilities, for their RBMK nuclear power plants (NPPs). The drawing of the Tacis nuclear

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.)

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)

On Brazil's participation in the International Project on Innovative Nuclear Reactors and Fuels Cycles (INPRO)

International Nuclear Information System (INIS)

Goncalves Filho, Orlando Joao Agostinho

2007-01-01

In response to a resolution of its 44th General Conference (GC(44)/RES/21) held in September 2000, the International Atomic Energy Agency launched in May 2001 the International Project on Innovative Nuclear Reactors and Fuels Cycles (INPRO) with the objective of supporting the safe, sustainable, economic and proliferation-resistant use of nuclear technology to meet the global energy needs of the 21st century. Brazil joined the project from its beginnings and in 2005 submitted a proposal for the screening assessment using INPRO methodology of two small-size light-water reactors as potential components of an innovative nuclear reactor system (INS) completed with a conventional open nuclear fuel cycle. The INS reactor components currently being assessed are the International Reactor Innovative and Secure (IRIS) that is being developed by an international consortium made of 21 organizations from 10 countries (Brazil included) led by the Westinghouse Company, and the Fixed Bed Nuclear Reactor (FBNR) that is being developed at the Federal University of Rio Grande do Sul. This paper gives an overview of Brazil's participation in INPRO, highlighting the objective, scope and intermediate results of the assessment study being performed, and the possibilities for participation in one or two collaborative research projects under INPRO Phase 2 Action Plan for 2008-2009. (author)

An experimental study of hypervapotron structure in external reactor vessel cooling

Energy Technology Data Exchange (ETDEWEB)

Zhao, Yufeng; Zhang, Ming [State Nuclear Power Technology R& D Center (Beijing), Beijing (China); Hou, Fangxin [Institute of Nuclear and New Energy Technology, Tsinghua University, Beijing (China); Gao, Tianfang [State Nuclear Power Technology R& D Center (Beijing), Beijing (China); Chen, Peipei, E-mail: chenpeipei@snptc.com.cn [State Power Investment Group Corporation, Beijing (China)

2016-07-15

Highlights: • Experiments are performed to study the application of hypervapotron in ERVC design. • CHF experiments on two surfaces are conducted under different flow conditions. • Hypervapotron improves CHF performance by 40–60% compared with smooth surface. • Visualization shows fin structure removes vapor mushroom for better liquid supply. - Abstract: In vessel retention (IVR) is one of the key strategies for many advanced LWR designs to mitigate postulated severe accidents. The success of IVR substantially relies on external reactor vessel cooling (ERVC) by which the decay heat is removed from the melt core in the reactor vessel lower head. The main challenge of IVR is to provide an adequate safety margin of ERVC against critical heat flux (CHF) of subcooled flow boiling in the reactor lower head flow channel. Due to uncertainties in corium melt pool configuration, large CHF margin of ERVC is usually required by regulatory authorities to demonstrate reliability of severe accident mitigation methods. Various CHF enhancement designs have been proposed and studied in literature. In this paper, an experimental study of hypervapotron structure as a novel design to improve CHF performance of ERVC is conducted. Hypervapotron is chosen as one of the potential engineering options for International Thermonuclear Experimental Reactor (ITER) program as a divertor structure to remove highly intense heat from fusion chamber. This study is to conduct CHF experiments at typical PWR ERVC working conditions. The CHF experiments are performed in a 30 mm by 61 mm rectangular flow channel with a 200 mm long heated surface along the flow direction. Both smooth and hypervapotron surface are tested at various inclination angles of the test section to simulate various positions of the reactor lower head. The hypervapotron is found to have a 40–60% CHF improvement compared with the smooth surface. The high speed visualization indicates that hypervapotron is able to

Internally Heated Screw Pyrolysis Reactor (IHSPR) heat transfer performance study

Science.gov (United States)

Teo, S. H.; Gan, H. L.; Alias, A.; Gan, L. M.

2018-04-01

1.5 billion end-of-life tyres (ELT) were discarded globally each year and pyrolysis is considered the best solution to convert the ELT into valuable high energy-density products. Among all pyrolysis technologies, screw reactor is favourable. However, conventional screw reactor risks plugging issue due to its lacklustre heat transfer performance. An internally heated screw pyrolysis reactor (IHSPR) was developed by local renewable energy industry, which serves as the research subject for heat transfer performance study of this particular paper. Zero-load heating test (ZLHT) was first carried out to obtain the operational parameters of the reactor, followed by the one dimensional steady-state heat transfer analysis carried out using SolidWorks Flow Simulation 2016. Experiments with feed rate manipulations and pyrolysis products analyses were conducted last to conclude the study.

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.

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.

Vibration test on KMRR reactor structure and primary cooling system piping

Energy Technology Data Exchange (ETDEWEB)

Chung, Seung Hoh; Kim, Tae Ryong; Park, Jin Hoh; Park, Jin Suk; Ryoo, Jung Soo [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1994-10-01

Most equipments, piping systems and reactor structures in nuclear power plants are subjected to flow induced vibration due to high temperature and high pressure coolant flowing inside or outside of the equipments, systems and structures. Because the flow induced vibration sometimes causes significant damage to reactor structures and piping systems, it is important and necessary to evaluate the vibration effect on them and to prove their structural integrity. Korea Multipurpose Research Reactor (KMRR) being constructed by KAERI is 30 MWt pool type research reactor. Since its main structures and piping systems were designed and manufactured in accordance with the standards and guidelines for commercial nuclear power plant, it was decided to evaluate their vibratory response in accordance with the standards and guidelines for commercial NPP. The objective of this vibration test is the assessment of vibration levels of KMRR reactor structure and primary cooling piping system for their structural integrity under the steady-state or transient operating condition. 38 figs, 14 tabs, 2 refs. (Author).

Vibration test on KMRR reactor structure and primary cooling system piping

International Nuclear Information System (INIS)

Chung, Seung Hoh; Kim, Tae Ryong; Park, Jin Hoh; Park, Jin Suk; Ryoo, Jung Soo

1994-10-01

Most equipments, piping systems and reactor structures in nuclear power plants are subjected to flow induced vibration due to high temperature and high pressure coolant flowing inside or outside of the equipments, systems and structures. Because the flow induced vibration sometimes causes significant damage to reactor structures and piping systems, it is important and necessary to evaluate the vibration effect on them and to prove their structural integrity. Korea Multipurpose Research Reactor (KMRR) being constructed by KAERI is 30 MWt pool type research reactor. Since its main structures and piping systems were designed and manufactured in accordance with the standards and guidelines for commercial nuclear power plant, it was decided to evaluate their vibratory response in accordance with the standards and guidelines for commercial NPP. The objective of this vibration test is the assessment of vibration levels of KMRR reactor structure and primary cooling piping system for their structural integrity under the steady-state or transient operating condition. 38 figs, 14 tabs, 2 refs. (Author)

INTEGRAL BENCHMARKS AVAILABLE THROUGH THE INTERNATIONAL REACTOR PHYSICS EXPERIMENT EVALUATION PROJECT AND THE INTERNATIONAL CRITICALITY SAFETY BENCHMARK EVALUATION PROJECT

Energy Technology Data Exchange (ETDEWEB)

J. Blair Briggs; Lori Scott; Enrico Sartori; Yolanda Rugama

2008-09-01

Interest in high-quality integral benchmark data is increasing as efforts to quantify and reduce calculational uncertainties accelerate to meet the demands of next generation reactor and advanced fuel cycle concepts. The International Reactor Physics Experiment Evaluation Project (IRPhEP) and the International Criticality Safety Benchmark Evaluation Project (ICSBEP) continue to expand their efforts and broaden their scope to identify, evaluate, and provide integral benchmark data for method and data validation. Benchmark model specifications provided by these two projects are used heavily by the international reactor physics, nuclear data, and criticality safety communities. Thus far, 14 countries have contributed to the IRPhEP, and 20 have contributed to the ICSBEP. The status of the IRPhEP and ICSBEP is discussed in this paper, and the future of the two projects is outlined and discussed. Selected benchmarks that have been added to the IRPhEP and ICSBEP handbooks since PHYSOR’06 are highlighted, and the future of the two projects is discussed.

INTEGRAL BENCHMARKS AVAILABLE THROUGH THE INTERNATIONAL REACTOR PHYSICS EXPERIMENT EVALUATION PROJECT AND THE INTERNATIONAL CRITICALITY SAFETY BENCHMARK EVALUATION PROJECT

International Nuclear Information System (INIS)

J. Blair Briggs; Lori Scott; Enrico Sartori; Yolanda Rugama

2008-01-01

Interest in high-quality integral benchmark data is increasing as efforts to quantify and reduce calculational uncertainties accelerate to meet the demands of next generation reactor and advanced fuel cycle concepts. The International Reactor Physics Experiment Evaluation Project (IRPhEP) and the International Criticality Safety Benchmark Evaluation Project (ICSBEP) continue to expand their efforts and broaden their scope to identify, evaluate, and provide integral benchmark data for method and data validation. Benchmark model specifications provided by these two projects are used heavily by the international reactor physics, nuclear data, and criticality safety communities. Thus far, 14 countries have contributed to the IRPhEP, and 20 have contributed to the ICSBEP. The status of the IRPhEP and ICSBEP is discussed in this paper, and the future of the two projects is outlined and discussed. Selected benchmarks that have been added to the IRPhEP and ICSBEP handbooks since PHYSOR-06 are highlighted, and the future of the two projects is discussed

Use of MCNP for characterization of reactor vessel internals waste from decommissioned nuclear reactors

International Nuclear Information System (INIS)

Love, E.F.; Pauley, K.A.; Reid, B.D.

1995-09-01

This study describes the use of the Monte Carlo Neutron-Photon (MCNP) code for determining activation levels of irradiated reactor vessel internals hardware. The purpose of the analysis is to produce data for the Department of Energy's Greater-Than-Class C Low-Level Radioactive Waste Program. An MCNP model was developed to analyze the Yankee Rowe reactor facility. The model incorporates reactor geometry, material compositions, and operating history data acquired from Yankee Atomic Electric Company. In addition to the base activation analysis, parametric studies were performed to determine the sensitivity of activation to specific parameters. A component sampling plan was also developed to validate the model results, although the plan was not implemented. The calculations for the Yankee Rowe reactor predict that only the core baffle and the core support plates will be activated to levels above the Class C limits. The parametric calculations show, however, that the large uncertainties in the material compositions could cause errors in the estimates that could also increase the estimated activation level of the core barrel to above the Class C limits. Extrapolation of the results to other reactor facilities indicates that in addition to the baffle and support plates, core barrels may also be activated to above Class C limits; however the classification will depend on the specific operating conditions of the reactor and the specific material compositions of the metal, as well as the use of allowable concentration averaging practices in packaging and classifying the waste

Basic data for surveillance test on core support graphite structures for the high temperature engineering test reactor (HTTR)

International Nuclear Information System (INIS)

Sumita, Junya; Shibata, Taiju; Kikuchi, Takayuki; Iyoku, Tatsuo; Fujimoto, Nozomu; Ishihara, Masahiro; Sawa, Kazuhiro

2007-02-01

Both of the visual inspection by a TV camera and the measurement of material properties by surveillance test on core support graphite structures are planned for the High Temperature Engineering Test Reactor (HTTR) to confirm their structural integrity and characteristics. The surveillance test is aimed to investigate the change of material properties by aging effects such as fast neutron irradiation and oxidation. The obtained data will be used not only for evaluating the structural integrity of the core support graphite structures of the HTTR but also for design of advanced Very High Temperature Reactor (VHTR) discussed at generation IV international forum. This report describes the initial material properties of surveillance specimens before installation and installed position of surveillance specimens in the HTTR. (author)

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)

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.)

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.

Proceedings of the international topical meeting on advanced reactors safety: Volume 2

International Nuclear Information System (INIS)

1997-01-01

In this volume, 89 papers are grouped under the following headings: advances in research/test reactor safety; advanced reactor accident management and emergency actions; advanced reactors instrumentation/controls/human factors; probabilistic risk/safety and reliability assessments; steam explosion research and issues; advanced reactor severe accident issues and research (analysis and assessments); advanced reactor thermal hydraulics; accelerator-driven source safety; liquid-metal reactor safety; structural assessments and issues; late papers

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.)

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.

Local drying underwater cutting of reactor core internals by CO laser

International Nuclear Information System (INIS)

Beppu, Seiji; Takano, Genta; Matsumoto, Osa; Sugihara, Masaaki; Miya, Kenzo.

1991-01-01

Since the CO laser operates at shorter wavelength than the CO 2 laser, the former has superior ability to cut materials. Its applicability to the cutting of reactor core internals in a nuclear power plant has been studied. In order to use such a laser for cutting actual structures, it would be useful to develop underwater cutting technology for the purpose of minimizing radiation exposure during cutting. However, since the laser beam is absorbed by water, a path needs to be cleared along the laser's line of sight. The authors have developed a nozzle that forms a local dry zone and have demonstrated that CO laser cutting can be carried out under water as it is in air. (author)

Fast breeder reactor research

International Nuclear Information System (INIS)

1975-01-01

Full text: The meeting was attended by 15 participants from seven countries and two international organizations. The Eighth Annual Meeting of the International Working Group on Fast Reactors (IWGFR) was attended by representatives from France, Fed. Rep. Germany, Italy, Japan, United Kingdom, Union of Soviet Socialist Republics and the United States of America - countries that have made significant progress in developing the technology and physics of sodium cooled fast reactors and have extensive national programmes in this field - as well as by representatives of the Commission of the European Communities and the IAEA. The design of fast-reactor power plants is a more difficult task than developing facilities with thermal reactors. Different reactor kinetics and dynamics, a hard neutron spectrum, larger integral doses of fuel and structural material irradiation, higher core temperatures, the use of an essentially novel coolant, and, as a result of all these factors, the additional reliability and safety requirements that are imposed on the planning and operation of sodium cooled fast reactors - all these factors pose problems that can be solved comprehensively only by countries with a high level of scientific and technical development. The exchange of experience between these countries and their combined efforts in solving the fundamental problems that arise in planning, constructing and operating fast reactors are promoting technical progress and reducing the relative expenditure required for various studies on developing and introducing commercial fast reactors. For this reason, the meeting concentrated on reviewing and discussing national fast reactor programmes. The situation with regard to planning, constructing and operating fast experimental and demonstration reactors in the countries concerned, the experience accumulated in operating them, the difficulties arising during operation and ways of over-coming them, the search for optimal designs for the power

The effects of impurity composition and concentration in reactor structure material on neutron activation inventory in pressurized water reactor

Energy Technology Data Exchange (ETDEWEB)

Cha, Gil Yong; Kim, Soon Young [RADCORE, Daejeon (Korea, Republic of); Lee, Jae Min [TUV Rheinland Korea, Seoul (Korea, Republic of); Kim, Yong Soo [Hanyang University, Seoul (Korea, Republic of)

2016-06-15

The neutron activation inventories in reactor vessel and its internals, and bio-shield of a PWR nuclear power plant were calculated to evaluate the effect of impurity elements contained in the structural materials on the activation inventory. Carbon steel is, in this work, used as the reactor vessel material, stainless steel as the reactor vessel internals, and ordinary concrete as the bio-shield. For stainless steel and carbon steel, one kind of impurity concentration was employed, and for ordinary concrete five kinds were employed in this study using MCNP5 and FISPACT for the calculation of neutron flux and activation inventory, respectively. As the results, specific activities for the cases with impurity elements were calculated to be more than twice than those for the cases without impurity elements in stainless and carbon steel. Especially, the specific activity for the concrete material with impurity elements was calculated to be 30 times higher than that without impurity. Neutron induced reactions and activation inventories in each material were also investigated, and it is noted that major radioactive nuclide in steel material is Co-60 from cobalt impurity element, and, in concrete material, Co-60 and Eu-152 from cobalt and europium impurity elements, respectively. The results of this study can be used for nuclear decommissioning plan during activation inventory assessment and regulation, and it is expected to be used as a reference in the design phase of nuclear power plant, considering the decommissioning of nuclear power plants or nuclear facilities.

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. In conclusion: FSI phenomena depend highly on constraints around FSI zone, so that the constraints must be dealt with realistically in analytical models. Although a two-dimensional model is superior to a quasi-two-dimensional model. The former needs long calculation time, so it is very expensive using in parametric study. Therefore, it is desirable that the two-dimensional model is used in the final study of reactor design and the quasi-two-dimensional model is used in parametric study. The blanket affects on the acoustic pressure and the deformations of radial structures, but affects scarcely on the upper vessel deformation. The blanket also affects on the mechanical work largely. The core barrel gives scarcely the effects on pressure in single phase but gives highly the effects on pressure in two-phase and deformation of reactor structures in this study. For studying the more realistic phenomena of FSI in the reactor design, the following works should be needed. (i) Spatial Distribution of FSI Region Spatial and time-dependent distribution of fuel temperature and molten fuel fraction must be taken in realistic simulation of accident condition. To this purpose, the code will

International Conference on Physics and Technology of Reactors and Applications

International Nuclear Information System (INIS)

2007-01-01

The first international conference on physics and technology of reactors and applications (PHYTRA 1) which took place in Marrakech (Morocco) from 14 to 16 March 2007, was designed to bring together scientists, teachers and students from universities, research centres and industry and other institutions to exchange knowledge and to discuss ideas and future issues. The programmes of the PHYTRA 1 conference covers a wide variety topics, the conference was organised in three plenary sessions, ten oral technical sessions and two poster sessions. The plenary sessions covers the following topics : The prospects of nuclear energy, The situation of nuclear sciences and energy in Morocco and Africa, and the new development in reactor physics and reactor design [fr

ITER at the international conference on fusion reactor materials

International Nuclear Information System (INIS)

Kalinin, G.; Barabash, V.; Matera, R.

1998-01-01

The reports summarizes the topics of the eighth International Conference on Fusion Reactor Materials (ICFRM-8) which was held in Sendai, Japan, on 26-31 October 1997. The ICFRM is focused on the whole spectrum of materials and technologies to be applied in fusion reactors and related facilities. The total number of conference participants was over 500, representing 24 countries and about 600 oral and poster papers were presented at the conference. Three sessions were devoted to ITER materials: (i) Design-Materials Interface and ITER (oral session); (ii) ITER, Irradiation Facility and Technology, (poster session); (iii) ITER and Beyond (discussion session)

Progress in liquid metal fast reactor technology. Proceedings of the 28th meeting of the International Working Group on Fast Reactors

International Nuclear Information System (INIS)

1996-04-01

The key objectives and activities of Member State liquid metal fast reactor (LMFR) programmes are: Demonstration of effective designs; demonstration of system safety; demonstration of economic competitiveness with other power generation systems. The International Working Group on Fast Reactors (IWGFR) at its 1995 meeting observed that while some countries (as a result of static or falling power demand) are reducing the research and development programmes or delaying the commercial deployment of fast reactors, other countries are planning to introduce these reactors and are embarking on their own development programmes. In these circumstances the international exchange of information and experience is of increasing importance. These proceedings contain updated information from long standing members of the IWGFR and new information on the status of LMFR research and development from new members of the Group: Brazil, China, Republic of Kazakhstan and the Republic of Korea. Refs, figs, tabs

Progress in liquid metal fast reactor technology. Proceedings of the 28th meeting of the International Working Group on Fast Reactors

Energy Technology Data Exchange (ETDEWEB)

NONE

1996-04-01

The key objectives and activities of Member State liquid metal fast reactor (LMFR) programmes are: Demonstration of effective designs; demonstration of system safety; demonstration of economic competitiveness with other power generation systems. The International Working Group on Fast Reactors (IWGFR) at its 1995 meeting observed that while some countries (as a result of static or falling power demand) are reducing the research and development programmes or delaying the commercial deployment of fast reactors, other countries are planning to introduce these reactors and are embarking on their own development programmes. In these circumstances the international exchange of information and experience is of increasing importance. These proceedings contain updated information from long standing members of the IWGFR and new information on the status of LMFR research and development from new members of the Group: Brazil, China, Republic of Kazakhstan and the Republic of Korea. Refs, figs, tabs.

Renewable side reflector structure for a pebble bed high temperature reactor

International Nuclear Information System (INIS)

Martin, Roger.

1977-01-01

The description is given of a renewable side reflector structure for a pebble bed high temperature reactor of the kind comprising a cylindrical graphite vessel constituting the neutron reflector, this vessel being filled with graphite pebbles containing the nuclear fuel and enclosed in a concrete protective containment. The internal peripheral area of the vessel is constituted by a line of adjacent graphite rods mounted so that they can rotate about their longitudinal axis and manoeuvrable from outside the concrete containment by means of a shaft passing into it [fr

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 ...

International project GT-MHR - New generation of nuclear reactors

International Nuclear Information System (INIS)

Vasyaev, A.; Kodochigov, N.; Kuzavkov, N.; Kuznetsov, L.

2001-01-01

Gas turbine-modular helium reactor (GT-MHR) is the reactor of new generation, which satisfies the requirements of the progressing large-scale nuclear power engineering. The activities in GT-MHR Project started in 1995. In 1997 the Conceptual Design was developed under four-side Agreement (MINATOM, General Atomics, FRAMATOME, Fuji Electric); it has passed through the internal and international reviews, has been approved and recommended for further development as one of new trends in creation of new generation plants. Starting from 1999, the activities in the development of the Preliminary Design of the plant were deployed under the Agreement between the Government of the United States of America and the Government of the Russian Federation on Scientific and Technical Cooperation in the Management of Plutonium That Has Been Withdrawn From Nuclear Military Programs dated July 24, 1998. The activities are established under the Contract between MINATOM and OKBM Russia, and under the General Agreement between Department of Energy (DOE), USA and OKBM. The GT-MHR Project is included into 'Development Strategy of Russian Nuclear Power in the first Half of the XXI-st Century' providing for 'the participation in an international project on the development and construction of GT-MHR nuclear power plant till year 2010 and 'operation of GT-MHR prototype unit and creation of fuel fabrication facility (within framework of International Project) till year 2030'. (author)

Fluid-structure-interaction of the pressurized water reactor core internals during blowdown - numerical simulation with a homogenization model

International Nuclear Information System (INIS)

Benner, J.

1984-03-01

A method for the numerical simulation of the Pressurized Water Reactor (PWR) core internal's behaviour during a blowdown accident is described, by which the motion of the reactor core and the interaction of the fuel elements with the core barrel and the coolant medium is calculated. Furthermore, some simple models for the support columns, lower and upper core support and the grid plate are provided. All these models have been implemented into the code Flux-4. For the solution of the very complex, coupled equations of motions for fluid and fuel rods an efficient numerical solution technique has been developed. With the new code-version Flux-5 the PWR-blowdown is parametically investigated. The calculated core barrel loadings are compared with Flux-4 results, simulating the core's inertia by a mass ring of HDR type. (orig.) [de

Structural Integrity Evaluation of the KALIMER-600 Reactor Core Support Structure

International Nuclear Information System (INIS)

Park, Chang Gyu; Kim, Jong Bum; Lee, Jae Han

2005-01-01

KALIMER-600(Korea Advanced LIquid MEtal Reactor, 600MWe) is a pool type sodium-cooled liquid metal reactor. Since the normal operating temperature of KALIMER-600 is 545 .deg. C, the reactor structures in the hot pool region are designed and evaluated according to the elevated temperature design rules such as the ASME Boiler and Pressure Vessel Code Section III, Subsection NH. Since the core support structure of KALIMER-600 is in the cold pool region under 400 .deg. C, a high temperature inelastic behavior is not expected. Thus the stress and fatigue limits are the main concerns to assure the structural design integrity following the ASME Subsection NG. In this paper, the evaluations of the stress and fatigue damage for the core support structure of KALIMER-600 are carrried out in the case of a normal operation condition using the rules of ASME Subsection NG. To obtain the stress values, a heat transfer analysis and a stress analysis under a combined loading condition are performed. From the stress distribution results, the critical sections are selected and the stress and fatigue limits are evaluated for the selected regions

Scale-model characterization of flow-induced vibrational response of FFTF reactor internals

International Nuclear Information System (INIS)

Ryan, J.A.; Mahoney, J.J.

1980-10-01

Fast Test Reactor core internal and peripheral components were assessed for flow-induced vibrational characteristics under scaled and simulated prototype flow conditions in the Hydraulic Core Mockup as an integral part of the Fast Test Reactor Vibration Program. The Hydraulic Core Mockup was an 0.285 geometric scale model of the Fast Test Reactor internals designed to simulate prototype vibrational and hydraulic characteristics. Using water to simulate sodium coolant, vibrational characteristics were measured and determined for selected model components over the scaled flow range of 36 to 110%. Additionally, in-situ shaker tests were conducted on selected Hydraulic Core Mockup outlet plenum components to establish modal characteristics. Most components exhibited resonant response at all test flow rates; however, the measured dynamic response was neither abnormal nor anomalously flow-rate dependent, and the predicted prototype components' response were deemed acceptable

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.)

International Reactor Innovative and Secure (IRIS) summary

International Nuclear Information System (INIS)

Carelli, Mario D.

2001-01-01

The IRIS (International Reactor Innovative and Secure) reactor is described in the first part of the presentation. IRIS is a light water cooled reactor with an integral configuration, where steam generators, pumps and pressurizer are inside the reactor vessel. Partially funded by the DOE NERI program, IRIS is being developed by an international consortium of 16 organizations from seven countries. A key IRIS characteristic is its 'safety by design' approach which strives to eliminate, by design, as many accidents as possible rather than coping with their consequences. Initial returns are very positive; out of the eight Class IV accidents considered in the AP600 only one remains as a Class IV in IRIS, and at much reduced probability. Small-to-medium LOCAs have minimal consequences as the core remains safely under water for days, without the need for safety injection or water makeup. In spite of its novelty IRIS is firmly grounded on proven LWR technology and therefore a prototype is not needed to assure design certification. Rather, very extensive scaled tests will be performed to investigate the performance of in-vessel components such as steam generators and pumps, both individually and as interactive systems. Accident sequences will also be simulated and tested to prove IRIS safety by design claims. The first core fuel is less than 5% enriched and the fuel assembly is very similar to existing PWR assemblies, so there is no licensing challenge regarding the fuel. Because of the safety by design approach, yielding simplifications In design and accident management (e.g., IRIS does not have an emergency core cooling system), some accident scenarios are eliminated and others have lesser consequences. Thus, simplification and streamlining of the regulatory process might be possible. Risk informed regulation will be coupled with safety by design to show lower accident and damage probabilities. This could lead to a relaxation of siting regulatory requirements. It is

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.)

GROWTH OF THE INTERNATIONAL CRITICALITY SAFETY AND REACTOR PHYSICS EXPERIMENT EVALUATION PROJECTS

Energy Technology Data Exchange (ETDEWEB)

J. Blair Briggs; John D. Bess; Jim Gulliford

2011-09-01

Since the International Conference on Nuclear Criticality Safety (ICNC) 2007, the International Criticality Safety Benchmark Evaluation Project (ICSBEP) and the International Reactor Physics Experiment Evaluation Project (IRPhEP) have continued to expand their efforts and broaden their scope. Eighteen countries participated on the ICSBEP in 2007. Now, there are 20, with recent contributions from Sweden and Argentina. The IRPhEP has also expanded from eight contributing countries in 2007 to 16 in 2011. Since ICNC 2007, the contents of the 'International Handbook of Evaluated Criticality Safety Benchmark Experiments1' have increased from 442 evaluations (38000 pages), containing benchmark specifications for 3955 critical or subcritical configurations to 516 evaluations (nearly 55000 pages), containing benchmark specifications for 4405 critical or subcritical configurations in the 2010 Edition of the ICSBEP Handbook. The contents of the Handbook have also increased from 21 to 24 criticality-alarm-placement/shielding configurations with multiple dose points for each, and from 20 to 200 configurations categorized as fundamental physics measurements relevant to criticality safety applications. Approximately 25 new evaluations and 150 additional configurations are expected to be added to the 2011 edition of the Handbook. Since ICNC 2007, the contents of the 'International Handbook of Evaluated Reactor Physics Benchmark Experiments2' have increased from 16 different experimental series that were performed at 12 different reactor facilities to 53 experimental series that were performed at 30 different reactor facilities in the 2011 edition of the Handbook. Considerable effort has also been made to improve the functionality of the searchable database, DICE (Database for the International Criticality Benchmark Evaluation Project) and verify the accuracy of the data contained therein. DICE will be discussed in separate papers at ICNC 2011. The status of the

Overview of Experiments for Physics of Fast Reactors from the International Handbooks of Evaluated Criticality Safety Benchmark Experiments and Evaluated Reactor Physics Benchmark Experiments

Energy Technology Data Exchange (ETDEWEB)

Bess, J. D.; Briggs, J. B.; Gulliford, J.; Ivanova, T.; Rozhikhin, E. V.; Semenov, M. Yu.; Tsibulya, A. M.; Koscheev, V. N.

2017-07-01

Overview of Experiments to Study the Physics of Fast Reactors Represented in the International Directories of Critical and Reactor Experiments John D. Bess Idaho National Laboratory Jim Gulliford, Tatiana Ivanova Nuclear Energy Agency of the Organisation for Economic Cooperation and Development E.V.Rozhikhin, M.Yu.Sem?nov, A.M.Tsibulya Institute of Physics and Power Engineering The study the physics of fast reactors traditionally used the experiments presented in the manual labor of the Working Group on Evaluation of sections CSEWG (ENDF-202) issued by the Brookhaven National Laboratory in 1974. This handbook presents simplified homogeneous model experiments with relevant experimental data, as amended. The Nuclear Energy Agency of the Organization for Economic Cooperation and Development coordinates the activities of two international projects on the collection, evaluation and documentation of experimental data - the International Project on the assessment of critical experiments (1994) and the International Project on the assessment of reactor experiments (since 2005). The result of the activities of these projects are replenished every year, an international directory of critical (ICSBEP Handbook) and reactor (IRPhEP Handbook) experiments. The handbooks present detailed models of experiments with minimal amendments. Such models are of particular interest in terms of the settlements modern programs. The directories contain a large number of experiments which are suitable for the study of physics of fast reactors. Many of these experiments were performed at specialized critical stands, such as BFS (Russia), ZPR and ZPPR (USA), the ZEBRA (UK) and the experimental reactor JOYO (Japan), FFTF (USA). Other experiments, such as compact metal assembly, is also of interest in terms of the physics of fast reactors, they have been carried out on the universal critical stands in Russian institutes (VNIITF and VNIIEF) and the US (LANL, LLNL, and others.). Also worth mentioning

Naval Reactors Prime Contractor Team (NRPCT) Experiences and Considerations With Irradiation Test Performance in an International Environment

International Nuclear Information System (INIS)

MH Lane

2006-01-01

This letter forwards a compilation of knowledge gained regarding international interactions and issues associated with Project Prometheus. The following topics are discussed herein: (1) Assessment of international fast reactor capability and availability; (2) Japanese fast reactor (JOYO) contracting strategy; (3) NRPCT/Program Office international contract follow; (4) Completion of the Japan Atomic Energy Agency (JAEA)/Pacific Northwest National Laboratory (PNNL) contract for manufacture of reactor test components; (5) US/Japanese Departmental interactions and required Treaties and Agreements; and (6) Non-technical details--interactions and considerations

Naval Reactors Prime Contractor Team (NRPCT) Experiences and Considerations With Irradiation Test Performance in an International Environment

Energy Technology Data Exchange (ETDEWEB)

MH Lane

2006-02-15

This letter forwards a compilation of knowledge gained regarding international interactions and issues associated with Project Prometheus. The following topics are discussed herein: (1) Assessment of international fast reactor capability and availability; (2) Japanese fast reactor (JOYO) contracting strategy; (3) NRPCT/Program Office international contract follow; (4) Completion of the Japan Atomic Energy Agency (JAEA)/Pacific Northwest National Laboratory (PNNL) contract for manufacture of reactor test components; (5) US/Japanese Departmental interactions and required Treaties and Agreements; and (6) Non-technical details--interactions and considerations.

Technical committee meeting on Liquid Metal Fast Reactor (LMFR) developments. 33rd annual meeting of the International Working Group on Fast Reactors (IWG-FR). Working material

International Nuclear Information System (INIS)

2000-01-01

Over the past 33 years, the IAEA has actively encouraged and advocated international cooperation in fast reactor technology. The present publication contains information on the status of fast reactor development and on worldwide activities in this advanced nuclear power technology during 1999/2000, as reported at the 33. annual meeting of the International Working Group on Fast Reactors. It is intended to provide information regarding the current status of LMFR development in IAEA Member States

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

Structural Integrity Assessment of Reactor Containment Subjected to Aircraft Crash

International Nuclear Information System (INIS)

Kim, Junyong; Chang, Yoonsuk

2013-01-01

When an accident occurs at the NPP, containment building which acts as the last barrier should be assessed and analyzed structural integrity by internal loading or external loading. On many occasions that can occur in the containment internal such as LOCA(Loss Of Coolant Accident) are already reflected to design. Likewise, there are several kinds of accidents that may occur from the outside of containment such as earthquakes, hurricanes and strong wind. However, aircraft crash that at outside of containment is not reflected yet in domestic because NPP sites have been selected based on the probabilistic method. After intentional aircraft crash such as World Trade Center and Pentagon accident in US, social awareness for safety of infrastructure like NPP was raised world widely and it is time for assessment of aircraft crash in domestic. The object of this paper is assessment of reactor containment subjected to aircraft crash by FEM(Finite Element Method). In this paper, assessment of structural integrity of containment building subjected to certain aircraft crash was carried out. Verification of structure integrity of containment by intentional severe accident. Maximum stress 61.21MPa of horizontal shell crash does not penetrate containment. Research for more realistic results needed by steel reinforced concrete model

Internal corium catcher of a nuclear reactor

International Nuclear Information System (INIS)

Anatolii S Vlasov; Vladimir N Mineev; Aleksandr S Sidorov; Yuri A Zeigarnik

2005-01-01

Full text of publication follows: A corium catcher is one of the main devices of a nuclear reactor that provides corium melt and fission products retention within a containment during severe accidents. Several studies and design developments have shown that corium retention within a reactor vessel can be attained with a moderate capacity of the latter (up to 600 - 650 MW el.). With a higher reactor capacity external corium catchers are applied both at Russian (VVER-1000) and European (EPR) reactors. In the external catcher of a VVER-1000 reactor, most technological problems are solved due to using sacrificial material. They are as follows: (a) endo-thermal interaction of corium and sacrificial material reduces a level of the temperatures in the final melt pool; (b) solution in the melt of a great amount of the sacrificial material reduces the specific heat release density and the heat flux density at the boundaries of a melt; (c) due to changing of the oxide-component density an inverse stratification of the metallic and oxide components of the corium takes place, thus excluding heat-flux focusing in the zone of the metallic layer and making it possible to supply water on the free surface of the corium without a danger of incipience of the vapor explosion; (d) final oxidation of zirconium occurs without hydrogen generation. The above principles have been realized in the external catcher of the VVER- 1000 reactor at Tyanvan NPS that is presently under construction in China. Successfully solving of the problems concerning to the external catcher makes it possible to return on the new conceptual and technological basis to the idea of retention of the corium melt inside the vessel of a nuclear reactor of large capacity, that is, to provide the reactor vessel to play a role of an internal catcher. For this purpose, a reactor vessel is elongated by approximately two meters. In the lower part of the vessel, on elliptical bottom, pieces of sacrificial material are arranged

Aging management strategy for reactor internals of Korean nuclear power plants

International Nuclear Information System (INIS)

Hwang, Seong Sik; Kim, Soung Woo; Lee, Sam Lai; Hong, Seung Mo; Kim, Hong Pyo; Kim, Dong Jin; Lim, Yun Soo; Kim, Joung Soo; Jung, Man Kyo; Park, Jang Yul

2010-01-01

This report describes various factors on the IASCC of reactor internals in terms of fluence, stress, water chemistries and materials. Materials of each components of Korean nuclear power plants have been surveyed. A technical report for a management of reactor internals issued by EPRI was reviewed for a selection of most susceptible area among many components. Baffle former bolts are considered to be the most susceptible parts due to high irradiation level(fluence) and high tensile stress. Neutron fluence of Kori-1 and Kori-2 was calculated based on fuel exchange history, fuel performance and plant operation history. This report will be used for more advanced inspection and maintenance guidelines development by supplying inspection intervals and components (most susceptible regions) for the long term operation plants

Fusion Reactor Materials

International Nuclear Information System (INIS)

Decreton, M.

2002-01-01

The objective of SCK-CEN's programme on fusion reactor materials is to contribute to the knowledge on the radiation-induced behaviour of fusion reactor materials and components as well as to help the international community in building the scientific and technical basis needed for the construction of the future reactor. Ongoing projects include: the study of the mechanical and chemical (corrosion) behaviour of structural materials under neutron irradiation and water coolant environment; the investigation of the characteristics of irradiated first wall material such as beryllium; investigations on the management of materials resulting from the dismantling of fusion reactors including waste disposal. Progress and achievements in these areas in 2001 are discussed

Management of research reactor ageing

International Nuclear Information System (INIS)

1995-03-01

As of December 1993, about one quarter of the operating research reactors were over 30 years old. The long life of research reactors has raised some concern amongst research reactor operators, regulators and, to some extent, the general public. The International Atomic Energy Agency commenced activities on the topic of research reactor ageing by appointing an internal working group in 1988 and convening a Consultants Meeting in 1989. The subject was also discussed at an international symposium and a regional seminar held in 1989 and 1992 respectively. A draft document incorporating information and experience exchanged at the above meetings was reviewed by a Technical Committee Meeting held in Vienna in 1992. The present TECDOC is the outcome of this meeting and contains recommendations, guidelines and information on the management of research reactor ageing, which should be used in conjunction with related publications of the IAEA Research Reactor Safety Programme, which are referenced throughout the text. This TECDOC will be of interest to operators and regulators involved with the safe operation of any type of research reactor to (a) understand the behaviour and influence of ageing mechanisms on the reactor structures, systems and components; (b) detect and assess the effect of ageing; (c) establish preventive and corrective measures to mitigate these effects; and (d) make decisions aimed at the safe and continued operation of a research reactor. 32 refs, tabs

Management of research reactor ageing

Energy Technology Data Exchange (ETDEWEB)

NONE

1995-03-01

As of December 1993, about one quarter of the operating research reactors were over 30 years old. The long life of research reactors has raised some concern amongst research reactor operators, regulators and, to some extent, the general public. The International Atomic Energy Agency commenced activities on the topic of research reactor ageing by appointing an internal working group in 1988 and convening a Consultants Meeting in 1989. The subject was also discussed at an international symposium and a regional seminar held in 1989 and 1992 respectively. A draft document incorporating information and experience exchanged at the above meetings was reviewed by a Technical Committee Meeting held in Vienna in 1992. The present TECDOC is the outcome of this meeting and contains recommendations, guidelines and information on the management of research reactor ageing, which should be used in conjunction with related publications of the IAEA Research Reactor Safety Programme, which are referenced throughout the text. This TECDOC will be of interest to operators and regulators involved with the safe operation of any type of research reactor to (a) understand the behaviour and influence of ageing mechanisms on the reactor structures, systems and components; (b) detect and assess the effect of ageing; (c) establish preventive and corrective measures to mitigate these effects; and (d) make decisions aimed at the safe and continued operation of a research reactor. 32 refs, tabs.

Optimization of reactor pressure vessel internals segmentation in Korea

Energy Technology Data Exchange (ETDEWEB)

Lee, Byung-Sik [Dankook Univ., Chungnam (Korea, Republic of). Dept. of Nuclear Engineering

2017-11-15

One of the most challenging tasks during plant decommissioning is the removal of highly radioactive internal components from the reactor pressure vessel (RPV). For RPV internals dismantling, it is essential that all activities are thoroughly planned and discussed in the early stage of the decommissioning project. One of the key activities in the detailed planning is to prepare the segmentation and packaging plan that describes the sequential steps required to segment, separate, and package each individual component of RPV, based on an activation analysis and component characterization study.

International benchmark on the natural convection test in Phenix reactor

International Nuclear Information System (INIS)

Tenchine, D.; Pialla, D.; Fanning, T.H.; Thomas, J.W.; Chellapandi, P.; Shvetsov, Y.; Maas, L.; Jeong, H.-Y.; Mikityuk, K.; Chenu, A.; Mochizuki, H.; Monti, S.

2013-01-01

Highlights: ► Phenix main characteristics, instrumentation and natural convection test are described. ► “Blind” calculations and post-test calculations from all the participants to the benchmark are compared to reactor data. ► Lessons learned from the natural convection test and the associated calculations are discussed. -- Abstract: The French Phenix sodium cooled fast reactor (SFR) started operation in 1973 and was stopped in 2009. Before the reactor was definitively shutdown, several final tests were planned and performed, including a natural convection test in the primary circuit. During this natural convection test, the heat rejection provided by the steam generators was disabled, followed several minutes later by reactor scram and coast-down of the primary pumps. The International Atomic Energy Agency (IAEA) launched a Coordinated Research Project (CRP) named “control rod withdrawal and sodium natural circulation tests performed during the Phenix end-of-life experiments”. The overall purpose of the CRP was to improve the Member States’ analytical capabilities in the field of SFR safety. An international benchmark on the natural convection test was organized with “blind” calculations in a first step, then “post-test” calculations and sensitivity studies compared with reactor measurements. Eight organizations from seven Member States took part in the benchmark: ANL (USA), CEA (France), IGCAR (India), IPPE (Russian Federation), IRSN (France), KAERI (Korea), PSI (Switzerland) and University of Fukui (Japan). Each organization performed computations and contributed to the analysis and global recommendations. This paper summarizes the findings of the CRP benchmark exercise associated with the Phenix natural convection test, including blind calculations, post-test calculations and comparisons with measured data. General comments and recommendations are pointed out to improve future simulations of natural convection in SFRs

Installation technology of reactor internals on shroud replacement work

International Nuclear Information System (INIS)

Miyano, Hiroshi

1999-01-01

Since the replacement of large welded reactor internals much as a core shroud did not have a precedent in the world, quite a few technologies had to be developed. Especially for the installation of new core shroud, jet pumps, core plate and top guide, the accurate weld and fit-up techniques for large structures was required to secure their integrity. The vessel shielding system was utilized to reduce general area dose rate such that all replacement work. For jet pump installation, automatic remote welding machines were used for high radiation area. As for the core shroud, shroud support weld prep machining tool with high accuracy, jacking system to support fit-up, new weld machine for small work space and low heat input weld joint were developed. Shroud replacement work in Fukushima Dai-ichi NPS Unit 3 (1F-3) with application of these development techniques, was successfully accomplished. The technology is applied for 1F-2 replacement work also. (author)

Westinghouse experience in using mechanical cutting for reactor vessel internals segmentation

International Nuclear Information System (INIS)

Boucau, Joseph; Fallstroem, Stefan; Segerud, Per; Kreitman, Paul J.

2010-01-01

Some commercial nuclear power plants have been permanently shut down to date and decommissioned using dismantling methods. Other operating plants have decided to undergo an upgrade process that includes replacement of reactor internals. In both cases, there is a need to perform a segmentation of the reactor vessel internals with proven methods for long term waste disposal. Westinghouse has developed several concepts to dismantle reactor internals based on safe and reliable techniques. Mechanical cutting has been used by Westinghouse since 1999 for both PWRs and BWRs and its process has been continuously improved over the years. Detailed planning is essential to a successful project, and typically a 'Segmentation and Packaging Plan' is prepared to document the effort. The usual method is to start at the end of the process, by evaluating the waste disposal requirements imposed by the waste disposal agency, what type and size of containers are available for the different disposal options, and working backwards to select the best cutting tools and finally the cut geometry required. These plans are made utilizing advanced 3-D CAD software to model the process. Another area where the modelling has proven invaluable is in determining the logistics of component placement and movement in the reactor cavity, which is typically very congested when all the internals are out of the reactor vessel in various stages of segmentation. The main objective of the segmentation and packaging plan is to determine the strategy for separating the highly activated components from the less activated material, so that they can be disposed of in the most cost effective manner. Usually, highly activated components cannot be shipped off-site, so they must be packaged such that they can be dry stored with the spent fuel in an Independent Spent Fuel Storage Installation (ISFSI). Less activated components can be shipped to an off-site disposal site depending on space availability. Several of the

The Effect of the Holes Size Change of Lower-Support-Structure-Bottom Plate on the Reactor Core-Inlet Flow-Distribution

International Nuclear Information System (INIS)

Lee, Gong Hee; Bang, Young Seok; Cheong, Ae Ju

2015-01-01

Complex thermal-hydraulic phenomena exist inside PWR because reactor interiors include a fuel assembly, control rod assembly, ICI (In-Core Instrumentation), and other internal structures. Because changes to reactor design may influence interior, thermal-hydraulic characteristics, licensing applicants commonly conduct a flow-distribution test and use test results (e.g., core-inlet flow-rate distribution) as the input data for a core thermal-margin analysis program. Because the APR+ (Advanced Power Reactor Plus) had more fuel assemblies (241EA → 257EA) and the design of some internal structures was changed (from those of APR1400), the core-inlet flow-rate distribution for a 1/5 scaled-down reactor model was measured and high flow-rates were found especially near the outer region of the reactor core. In this study, to examine the effect of the holes size change (i.e. smaller diameter) in the outer region of the LSSBP, not a 50% blockage of the flow holes, on the reactor core-inlet flow-distribution, simulations were conducted with the commercial CFD (Computational Fluid Dynamics) software, ANSYS CFX R.14. The predicted results were compared with those of the original LSSBP. In this study, to examine the effect of the holes size change (smaller diameter) in the outer region of the LSSBP on the reactor core-inlet flow-distribution, simulations were conducted with the commercial CFD software, ANSYS CFX R.14. The predicted results were compared with those of the original LSSBP. Through these comparisons it was concluded that a more uniform distribution of the mass-flow rate at the core-inlet plane could be obtained by reducing the holes size in the outer region of the LSSBP

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.)

IGORR 2: Proceedings of the 2. meeting of the International Group On Research Reactors

Energy Technology Data Exchange (ETDEWEB)

NONE

1992-07-01

The International group on Research Reactors was formed to facilitate the sharing of knowledge and experience among those institutions and individuals who are actively working to design, build, and promote new research reactors or to make significant upgrades to existing facilities. Sessions during this second meeting were devoted to research reactor reports (GRENOBLE reactor, FRM-II, HIFAR, PIK, reactors at JAERI, MAPLE, ANS, NIST, MURR, TRIGA, BR-2, SIRIUS 2); other neutron sources; and two workshops were dealing with research and development results and needs and reports on progress in needed of R and D areas identified at IGORR 1.

IGORR 2: Proceedings of the 2. meeting of the International Group On Research Reactors

International Nuclear Information System (INIS)

1992-01-01

The International group on Research Reactors was formed to facilitate the sharing of knowledge and experience among those institutions and individuals who are actively working to design, build, and promote new research reactors or to make significant upgrades to existing facilities. Sessions during this second meeting were devoted to research reactor reports (GRENOBLE reactor, FRM-II, HIFAR, PIK, reactors at JAERI, MAPLE, ANS, NIST, MURR, TRIGA, BR-2, SIRIUS 2); other neutron sources; and two workshops were dealing with research and development results and needs and reports on progress in needed of R and D areas identified at IGORR 1

Inspection and repair of reactor pressure vessel (RPV) internals

International Nuclear Information System (INIS)

Bohmann, W.; Poetz, F.; Nicolai, M.

1996-01-01

The past 10 years of operation of light water reactors were characterized by intensive inspection- and repair work on vital components. For boiling water reactors (BWR) it was typical to totally replace the piping system and for pressurized water reactors (PWR) it was the step to complete steam generator (SG) replacement - besides the development of increasingly diligent inspection and repair methods for SG tubes. It can be expected that in the 10 years to come the development of inspection- and repair methods will be aimed mainly at the core internals of BWR's as well as PWR's. Our prediction is that before the end of this decade a first complete replacement of these components will be performed. Already to date a broad range of techniques are available which enable the utilities to carry out inspections and repair of components of core internals in a relatively short time and acceptable expenses. Using examples such as Fuel Alignment Pin Inspection and Replacement, Baffle Former Bolt Inspection and Replacement, Core Barrel Former Bolt Inspection which are typical for PWR's we will in the following describe the existing methods, their development and - last but not least - their successful utilization. What is going to happen in the future? Ageing of the operating plants will continue, thus requesting the plant operators as well as the service companies to work on advanced technologies to fulfill the needs of the industry. (author)

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.)

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.)

Reactor Core Internals Replacement of Ikata Units 1 and 2

International Nuclear Information System (INIS)

Ikeda, K.; Ishikawa, T.; Miyoshi, T.; Takagi, T.

2012-01-01

This paper presents an overview of the reactor core internals replacement project carried out at the Ikata Nuclear Power Station in Japan, which was the first of its kind among PWRs in the world. Failure of baffle former bolts was first reported in 1989 at Bugey 2 in France. Since then, similar incidents have been reported in Belgium and in the U.S., but not in Japan. However, the possibility of these bolts failing in Japanese plants cannot be denied in the future as operating hours increase. Ageing degradation mechanisms for the reactor core internals include irradiation-assisted stress corrosion cracking of baffle former bolts and mechanical wear of control rod guide cards. Two different approaches can be taken to address these ageing issues: to inspect and repair whenever a problem is found; and to replace the entire core internals with those of a new design having advanced features to prevent ageing degradation problems. The choice of our company was the latter. This paper explains the reasons for the choice and summarizes the replacement project activities at Ikata Units 1 and 2 as well as the improvements incorporated in the new design. (author)

The Jules Horowitz Reactor - A new High Performance European Material Testing Reactor open to International Users Present Status and Objectives

International Nuclear Information System (INIS)

Iracane, Daniel; Bignan, Gilles; Lindbaeck, Jan-Erik; Blomgren, Jan

2010-01-01

The development of sustainable nuclear energy requires R and D on fuel and material behaviour under irradiation with a high level of performance in order to meet the needs and challenges for the benefit of industry, research and public bodies. These stakes require a sustainable and secured access to an up-to-date high performance Material Testing Reactor. Following a broad survey within the European Research Area, the international community agreed that the need for Material Test Reactors in support of nuclear power plant safety and operation will continue in the context of sustainable nuclear energy. The Jules Horowitz Reactor project (JHR) copes with this context. JHR is designed as a user facility addressing the needs of the international community. This means: - flexibility with irradiation loops able to reproduce a large variation in operation conditions of different power reactor technologies, - high flux capacity to address Generations II, III, and IV needs. JHR is designed, built and operated as an international user facility because: - Given the maturity and globalization of the industry, domestic tools have no more the required level of economic and technical efficiency. Meanwhile, countries with nuclear energy need an access to high performance irradiation experimental capabilities to support technical skill and guarantee the competitiveness and safety of nuclear energy. - Many research items related to safety or public policy (waste management, etc.) require international cooperation to share costs and benefits of resulting consensus. JHR design is optimised for offering high performance material and fuel irradiation capability for the coming decades. This project is driven and funded by an international consortium gathering vendors, utilities and public stakeholders. This consortium has been set up in March 2007 when the construction began. The construction is in progress and the start of operation is scheduled for 2014. The JHR is a research

HEU core conversion of Russian production reactors: a major threat to the international RERTR regime

International Nuclear Information System (INIS)

Kuperman, Alan J.; Leventhal, Paul L.

1998-01-01

This paper calls the attention for the major threat to the International Reduced Enrichment for Research and Test Reactors (RERTR) program, represented by the HEU core conversion of russian production reactors. This program aims to reduce and eventually eliminate international civilian commerce in nuclear weapons-usable, highly enriched uranium , and thereby significantly lower risks of the material being stolen or diverted by terrorist or states for producing nuclear weapons

International Reactor Physics Experiment Evaluation (IRPhE) Project. IRPhE Handbook - 2017 edition

International Nuclear Information System (INIS)

2017-01-01

The International Reactor Physics Evaluation (IRPhE) Project was initiated as a pilot in 1999 by the Nuclear Energy Agency (NEA) Nuclear Science Committee (NSC). The project was endorsed as an official activity of the NSC in June 2003. While the NEA co-ordinates and administers the IRPhE Project at the international level, each participating country is responsible for the administration, technical direction and priorities of the project within their respective countries. The information and data included in this handbook are available to NEA member countries, to all contributing countries and to others on a case-by-case basis. The IRPhE Project is patterned after the International Criticality Safety Benchmark Evaluation Project (ICSBEP). It closely co-ordinates with the ICSBEP to avoid duplication of efforts and publication of conflicting information. Some benchmark data are applicable to both nuclear criticality safety and reactor physics technology. Some have already been evaluated and published by the ICSBEP, but have been extended to include other types of measurements in addition to the critical configuration. Through this effort, the IRPhE Project will be able to 1) consolidate and preserve the existing worldwide information base; 2) retrieve lost data; 3) identify areas where more data are needed; 4) draw upon the resources of the international reactor physics community to help fill knowledge gaps; 5) identify discrepancies between calculations and experiments due to deficiencies in reported experimental data, cross-section data, cross-section processing codes and neutronics codes; 6) eliminate a large amount of redundant research and processing of reactor physics experiment data, and 7) improve future experimental planning, execution and reporting. This handbook contains reactor physics benchmark specifications that have been derived from experiments performed at nuclear facilities around the world. The benchmark specifications are intended for use by

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)

Investigation of structural materials of reactors using high-energy heavy-ion irradiations

International Nuclear Information System (INIS)

Wang Zhiguang

2007-01-01

Radiation damage in structural materials of fission/fusion reactors is mainly attributed to the evolution of intensive atom displacement damage induced by energetic particles (n, α and/or fission fragments) and high-rate helium doping by direct α particle bombardments and/or (n, α) reactions. It can cause severe degradation of reactor structural materials such as surface blistering, bulk void swelling, deformation, fatigue, embrittlement, stress erosion corrosion and so on that will significantly affect the operation safety of reactors. However, up to now, behavior of structural materials at the end of their service can hardly be fully tested in a real reactor. In the present work, damage process in reactor structural materials is briefly introduced, then the advantages of energetic ion implantation/irradiation especially high-energy heavy ion irradiation are discussed, and several typical examples on simulation of radiation effects in reactor candidate structural materials using high-energy heavy ion irradiations are pronounced. Experimental results and theoretical analysis suggested that irradiation with energetic particles especially high-energy heavy ions is very useful technique for simulating the evolution of microstructures and macro-properties of reactor structural materials. Furthermore, an on-going plan of material irradiation experiments using high energy H- and He-ions based on the Heavy Ion Research Facilities in Lanzhou (HIRFL) is also briefly interpreted. (authors)

Common floor system vertical earthquake-proof structure for reactor equipment

International Nuclear Information System (INIS)

Morishita, Masaki.

1996-01-01

In an LMFBR type reactor, a reactor container, a recycling pump and a heat exchanger are disposed on a common floor. Vertical earthquake-proof devices which can be stretched only in vertical direction formed by laminating large-sized bellevilles are disposed on a concrete wall at the circumference of each of reactor equipments. A common floor is placed on all of the vertical earthquake-proof devices to support the entire earthquake-proof structure simultaneously. If each of reactor equipments is loaded on the common floor and the common floor is entirely supported against earthquakes altogether, since the movement of each of the reactor equipments loaded on the common floor is identical, relative dislocation is not exerted on the main pipelines which connect the equipments. In addition, since the entire earthquake structure has a flat common floor and each of the reactor equipments is suspended to minimize the distance between a gravitational center and a support point, locking vibration is less caused to the horizontal earthquake. (N.H.)

Fusion Reactor Materials

Energy Technology Data Exchange (ETDEWEB)

Decreton, M

2002-04-01

The objective of SCK-CEN's programme on fusion reactor materials is to contribute to the knowledge on the radiation-induced behaviour of fusion reactor materials and components as well as to help the international community in building the scientific and technical basis needed for the construction of the future reactor. Ongoing projects include: the study of the mechanical and chemical (corrosion) behaviour of structural materials under neutron irradiation and water coolant environment; the investigation of the characteristics of irradiated first wall material such as beryllium; investigations on the management of materials resulting from the dismantling of fusion reactors including waste disposal. Progress and achievements in these areas in 2001 are discussed.

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

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.)

The obscure factor analysis on the vibration reliability of the internals of nuclear power plant reactor and anti-vibration measures

International Nuclear Information System (INIS)

Fu Geyan; Zhu Qirong

1998-11-01

It is pointed out that the main reason making nuclear power plants reactors leak is the vibration of internals of reactors. The factors which lead the vibration all have randomness and obscureness. The obscure reliability theory is introduced to the vibration system of internals of nuclear power reactor. Based on a quantity of designing and moving data, the obscure factors effecting the vibration reliability of the internals of nuclear power plant reactor are analyzed and the anti-vibration reliability criteria and the evaluating model are given. And the anti-vibration reliability measures are advanced from different quarters of the machine design and building, the thermohydraulics design, the control of reactivity, etc.. They may benefit the theory and practice for building and perfecting the vibration obscure reliability model of the reactor internals

Structural and piping issues in the design certification of advanced reactors

International Nuclear Information System (INIS)

Ali, S.A.; Terao, D.; Bagchi, G.

1996-01-01

The purpose of this paper is to discuss the design certification of structures and piping for evolutionary and passive advanced light water reactors. Advanced reactor designs are based on a set of assumed site-related parameters that are selected to envelop a majority of potential nuclear power plant sites. Multiple time histories are used as the seismic design basis in order to cover the majority of potential sites in the US. Additionally, design are established to ensure that surface motions at a particular site will not exceed the enveloped standard design surface motions. State-of-the-art soil-structure interaction (SSI) analyses have been performed for the advanced reactors, which include structure-to-structure interaction for all seismic Category 1 structures. Advanced technology has been utilized to exclude the dynamic effects of pipe rupture from structural design by demonstrating that the probability of pipe rupture is extremely low. For piping design, the advanced reactor vendors have developed design acceptance criteria (DAC) which provides the piping design analysis methods, design procedures, and acceptance criteria. In SECY-93-087 the NRC staff recommended that the Commission approve the approach to eliminate the OBE from the design of structures and piping in advanced reactors and provided guidance which identifies the necessary changes to existing seismic design criteria. The supplemental criteria address fatigue, seismic anchor motion, and piping stress limits when the OBE is eliminated

China: EDF's feedback experience of reactor operating is essential to win international markets

International Nuclear Information System (INIS)

Maillart, H.

2016-01-01

The main assets of EDF on the Chinese nuclear power market is first, its very important feedback experience of reactor operations (EDF cumulates one year of reactor operations every week due to its fleet of 58 reactors), secondly the cooperation with China allowed China to enter nuclear energy in 1983 with the construction of the Daya Bay plant and now to develop its own technology: the CPR-1000 reactor. China is the world leader in terms of nuclear market dynamism with 30 reactors in operation, 24 reactors being built and 40 others planned. A new stage in the Franco-China cooperation would be to share relevant good practices in the managing of both French and Chinese fleets of reactors. EDF has upgraded its commercial international offer, it now proposes to cover all the stages of the nuclear power plant from site selection to plant deconstruction via construction, operation, maintenance and waste management which constitutes a commitment over a 100 year period. (A.C.)

Effects of neutron radiation and residual stresses on the corrosion of welds in light water reactor internals

International Nuclear Information System (INIS)

Schaaf, Bob van der; Gavillet, Didier; Lapena, Jesus; Ohms, Carsten; Roth, Armin; Dyck, Steven van

2006-01-01

After many years of operation in Light Water Reactors (LWR) Irradiation Assisted Stress Corrosion Cracking (IASCC) of internals has been observed. In particular the heat-affected zone (HAZ) has been associated with IASCC attack. The welding process induces residual stresses and micro-structural modifications. Neutron irradiation affects the materials response to mechanical loading. IASCC susceptibility of base materials is widely studied, but the specific conditions of irradiated welds are rarely assessed. Core component relevant welds of Type 304 and 347 steels have been fabricated and were irradiated in the High Flux Reactor (HFR) in Petten to 0.3 and 1 dpa (displacement per atom). In-service welds were cut from the thermal shield of the decommissioned BR-3 reactor. Residual stresses, measured using neutron diffraction, ring core tests and X-ray showed residual stress levels up to 400 MPa. Micro-structural characterization showed higher dislocation densities in the weld and HAZ. Neutron radiation increased the dislocation density, resulting in hardening and reduced fracture toughness. The sensitization degree of the welds, measured with the electrochemical potentio-dynamic reactivation method, was negligible. The Slow Strain Rate Tensile (SSRT) tests, performed at 290 deg. C in water with 200 ppb dissolved oxygen, (DO), did not reveal inter-granular cracking. Inter-granular attack of in-service steel is observed in water with 8 ppm (DO), attributed not only to IASCC, but also to IGSCC from thermal sensitization during fabrication. Stress-relieve annealing has caused Cr-grain boundary precipitation, indicating the sensitization. The simulated internal welds, irradiated up to 1.0 dpa, did not show inter-granular cracking with 8 ppm DO. (authors)

Combined development of international nuclear fusion test reactors

International Nuclear Information System (INIS)

Anon.

1987-01-01

Ambassadors of the four most important partners (Common Market, Japan, USA and USSR) in the IAEA sponsored INTOR project, met on the 15 and 16 March 1987 in Vienna under the auspices of the IAEA. A press release was issued acknowledging the considerable technical progress made in magnetic nuclear fusion research. Future design concepts, assistance in research and development work and other activities towards the provision of an international experimental thermonuclear reactor were discussed. (G.T.H.)

Ukrainian Nuclear Society International Conference 'Modernization of the NPP with VVER reactor' (abstracts)

International Nuclear Information System (INIS)

Bar'yakhtar, V.G.

1999-01-01

Abstracts of the papers presented at International conference of the Ukrainian Nuclear Society 'Modernization of the NPP with VVER reactor'. The following problems are considered: improving the NPP's safety and reliability; reactor modernization, the lifetime prolongation; increasing of the reactor operating characteristics; methods of capacity factor increasing: refueling control, maintenance control; technical and economical aspects of NPP modernization; modernization of the automated control system of the fuel process at the NPP's; technical features and methods for the continued radiation and technology control at the NPP's; training, increasing the staff qualification and NPP modernization

Fuel, structural material and coolant for an advanced fast micro-reactor

International Nuclear Information System (INIS)

Nascimento, Jamil A. do; Guimaraes, Lamartine N.F.; Ono, Shizuca

2011-01-01

The use of nuclear reactors in space, seabed or other Earth hostile environment in the future is a vision that some Brazilian nuclear researchers share. Currently, the USA, a leader in space exploration, has as long-term objectives the establishment of a permanent Moon base and to launch a manned mission to Mars. A nuclear micro-reactor is the power source chosen to provide energy for life support, electricity for systems, in these missions. A strategy to develop an advanced micro-reactor technologies may consider the current fast reactor technologies as back-up and the development of advanced fuel, structural and coolant materials. The next generation reactors (GEN-IV) for terrestrial applications will operate with high output temperature to allow advanced conversion cycle, such as Brayton, and hydrogen production, among others. The development of an advanced fast micro-reactor may create a synergy between the GEN-IV and space reactor technologies. Considering a set of basic requirements and materials properties this paper discusses the choice of advanced fuel, structural and coolant materials for a fast micro-reactor. The chosen candidate materials are: nitride, oxide as back-up, for fuel, lead, tin and gallium for coolant, ferritic MA-ODS and Mo alloys for core structures. The next step will be the neutronic and burnup evaluation of core concepts with this set of materials. (author)

International Forum for Reactor Aging Management (IFRAM): Progress in cooperation to promote longer term operations (LTO)

International Nuclear Information System (INIS)

Shoji, T.; Bond, L.J.; Brenchley, D.L.; Carpenter, C.E.; Hwang, I.S.; Martin, O.; Reister, R.; Tilley, R.

2012-01-01

The vision and goals for the International Forum for Reactor Aging Management (IFRAM) are an organization that will complement existing networks and provide the framework to facilitate the appropriate exchange of information among those parties and organizations around the world that are presently addressing, or are planning to address, issues of nuclear power plant systems, structures and components aging management. This paper will provide a status report on IFRAM and a summary of the inaugural meeting, and discuss activities and plans for the path forward. (author)

Review of inspection of Torness reactor internals using remote techniques

International Nuclear Information System (INIS)

Dynan, J.

1993-01-01

Inspection of reactor internals is increasingly being achieved by deployment of cameras into areas of sometimes high radiation fields using manipulators. The manipulators have been developed over a number of years and from a variety of sources and those provided at Torness can maybe be seen to be the state of the art development of such machines for AGRS. Torness and Heysham 2 reactors were specifically designed for inspection using remote techniques and special facilities have been provided for this purpose. This paper is written by an operator and not a manipulator specialist and is intended to show what the operator needs versus what the designer gave him. (author)

The IAEA's International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO)

International Nuclear Information System (INIS)

Juergen Kupitz

2002-01-01

This paper presents the IAEA International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO). It defines its rationale, key objectives and specifies the organizational structure. The IAEA General Conference (2000) has invited 'all interested Member States to combine their efforts under the aegis of the Agency in considering the issues of the nuclear fuel cycle, in particular by examining innovative and proliferation-resistant nuclear technology' (GC(44)/RES/21) and invited Member States to consider to contribute to a task force on innovative nuclear reactors and fuel cycle (GC(44)/RES/22). In response to this invitation, the IAEA initiated an 'International Project on Innovative Nuclear Reactors and Fuel Cycles', INPRO. The Terms of Reference for INPRO were adopted at a preparatory meeting in November 2000, and the project was finally launched by the INPRO Steering Committee in May 2001. At the General Conference in 2001, first progress was reported, and the General Conference adopted a resolution on 'Agency Activities in the Development of Innovative Nuclear Technology' [GC(45)/RES/12, Tab F], giving INPRO a broad basis of support. The resolution recognized the 'unique role that the Agency can play in international collaboration in the nuclear field'. It invited both 'interested Member States to contribute to innovative nuclear technology activities' at the Agency as well as the Agency itself 'to continue it's efforts in these areas'. Additional endorsement came in a UN General Assembly resolution in December 2001 (UN GA 2001, A/RES/56/94), that again emphasized 'the unique role that the Agency can play in developing user requirements and in addressing safeguards, safety and environmental questions for innovative reactors and their fuel cycles' and stressed 'the need for international collaboration in the development of innovative nuclear technology'. As of February 2002, the following countries or entities have become members of INPRO: Argentina

Program of assessment of mechanical and corrosion mechanical properties of reactor internals materials due to operation conditions in WWERs

International Nuclear Information System (INIS)

Ruscak, M.; Zamboch, M.

1998-01-01

Reactor internals are subject to three principle operation influences: neutron and gamma irradiation, mechanical stresses, both static and dynamic, and coolant chemistry. Several cases of damage have been reported in previous years in both boiling and pressure water reactors. They are linked with the term of irradiation assisted stress corrosion cracking as a possible damage mechanism. In WWERs, the principal material used for reactor internals is austenitic titanium stabilized stainless steel 08Kh18N10T, however high strength steels are used as well. To assess the changes of mechanical properties and to determine whether sensitivity to intergranular cracking can be increased by high neutron fluences, the experimental program has been started. The goal is to assure safe operation of the internals as well as life management for all planned operation period. The program consists of tests of material properties, both mechanical and corrosion-mechanical. Detailed neutron fluxes calculation as well as stress and deformation calculations are part of the assessment. Model of change will be proposed in order to plan inspections of the facility. In situ measurements of internals will be used to monitor exact status of structure during operation. Tensile specimens manufactured from both base metal and model weld joint have been irradiated to the total fluences of 3-20 dpa. Changes of mechanical properties are tested by the tensile test, stress corrosion cracking tests are performed in the autoclave with water loop and active loading. Operation temperature, pressure and water chemistry are chosen for the tests. (author)

Preliminary structural evaluations of the STAR-LM reactor vessel and the support design

International Nuclear Information System (INIS)

Koo, Gyeong-Hoi; Sienicki, James J.; Moisseytsev, Anton

2007-01-01

In this paper, preliminary structural evaluations of the reactor vessel and support design of the STAR-LM (The Secure, Transportable, Autonomous Reactor - Liquid Metal variant), which is a lead-cooled reactor, are carried out with respect to an elevated temperature design and seismic design. For an elevated temperature design, the structural integrity of a direct coolant contact to the reactor vessel is investigated by using a detail structural analysis and the ASME-NH code rules. From the results of the structural analyses and the integrity evaluations, it was found that the design concept of a direct coolant contact to the reactor vessel cannot satisfy the ASME-NH rules for a given design condition. Therefore, a design modification with regards to the thermal barrier is introduced in the STAR-LM design. For a seismic design, detailed seismic time history response analyses for a reactor vessel with a consideration of a fluid-structure interaction are carried out for both a top support type and a bottom support type. And from the results of the hydrodynamic pressure responses, an investigation of the minimum thickness design of the reactor vessel is tentatively carried out by using the ASME design rules

Dual-cell reduction and group effect in an internal microelectrolysis reactor

International Nuclear Information System (INIS)

Ying, Diwen; Peng, Juan; Li, Kan; Wang, Yalin; Pan, Siwen; Jia, JinPing

2013-01-01

Highlights: ► 2,4-DCP was removed simultaneously in cathode and anode cells in IME reactor. ► Mechanism of dual-cell reduction gave an insight of cathode and anode cells. ► Significant V oc increase with one/two electrodes couples being installed in series. ► Group effect reveals possible high redox potential in IME reactor. -- Abstract: To address some of the fundamental questions regarding the functions of cathodes and anodes (e.g., iron and granular active carbon) and what happens when numerous particles (microscopic galvanic cells) are combined in the widely used and efficient wastewater treatment of an internal microelectrolysis (IME) reactor, we employed a specifically designed dual-cell IME reactor with single-couple electrodes and a narrow cell IME reactor with multi-couple electrodes. The simultaneous removal of 6.36 mg L −1 and 2.93 mg L −1 of 2,4-dichlorophenol within 30 min from the graphite cathode and the iron anode cells, respectively, was observed in the dual-cell reactor. The innovative concept behind this phenomenon is that the anode, which is generally believed to be oxidative, could probably be reductive in an IME reactor. Thus, it is important to understand the unique performance of IME reactors. The group effect, which provides a 45% increase of open-circuit potential with just two additional electrode couples in aqueous solution, was tested and verified in a multi-couple electrode reactor. It suggests a complex potential distribution in the IME reactor and that compounds even with high redox potential could possibly be reduced, which was generally believed to be difficult to accomplish

Design considerations for ITER [International Thermonuclear Experimental Reactor] toroidal field coils

International Nuclear Information System (INIS)

Kalsi, S.S.; Lousteau, D.C.; Miller, J.R.

1987-01-01

The International Thermonuclear Experimental Reactor (ITER) is a new tokamak design project with joint participation from Europe, Japan, the Union of Soviet Socialist Republics (USSR), and the United States. This paper describes a magnetic and mechanical design methodology for toroidal field (TF) coils that employs Nb/sub 3/Sn superconductor technology. Coil winding is sized by using conductor concepts developed for the US TIBER concept. The nuclear heating generated during operation is removed from the windings by helium flowing through the conductor. The heat in the coil case is removed through a separate cooling circuit operating at approximately 20 K. Manifold concepts are presented for the complete coil cooling system. Also included are concepts for the coil structural arrangement. The effects of in-plane and out-of-plane loads are included in the design considerations for the windings and case. Concepts are presented for reacting these loads with a minimum amount of additional structural material. Concepts discussed in this paper could be considered for the ITER TF coils. 6 refs., 5 figs., 1 tab

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)

Nuclear reactor melt-retention structure to mitigate direct containment heating

Science.gov (United States)

Tutu, Narinder K.; Ginsberg, Theodore; Klages, John R.

1991-01-01

A light water nuclear reactor melt-retention structure to mitigate the extent of direct containment heating of the reactor containment building. The structure includes a retention chamber for retaining molten core material away from the upper regions of the reactor containment building when a severe accident causes the bottom of the pressure vessel of the reactor to fail and discharge such molten material under high pressure through the reactor cavity into the retention chamber. In combination with the melt-retention chamber there is provided a passageway that includes molten core droplet deflector vanes and has gas vent means in its upper surface, which means are operable to deflect molten core droplets into the retention chamber while allowing high pressure steam and gases to be vented into the upper regions of the containment building. A plurality of platforms are mounted within the passageway and the melt-retention structure to direct the flow of molten core material and help retain it within the melt-retention chamber. In addition, ribs are mounted at spaced positions on the floor of the melt-retention chamber, and grid means are positioned at the entrance side of the retention chamber. The grid means develop gas back pressure that helps separate the molten core droplets from discharged high pressure steam and gases, thereby forcing the steam and gases to vent into the upper regions of the reactor containment building.

IGORR-IV -- Proceedings of the fourth meeting of the International Group on Research Reactors

Energy Technology Data Exchange (ETDEWEB)

Rosenbalm, K.F. [comp.

1995-12-31

The International Group on Research Reactors was formed to facilitate the sharing of knowledge and experience among those institutions and individuals who are actively working to design, build, and promote new research reactors or to make significant upgrades to existing facilities. Twenty-nine papers were presented in five sessions and written versions of the papers or hard copies of the vugraphs used are published in these proceedings. The five sessions were: (1) Operating Research Reactors and Facility Upgrades; (2) Research Reactors in Design and Construction; (3) ANS Closeout Activities; (4) and (5) Research, Development, and Analysis Results.

IGORR-IV - Proceedings of the fourth meeting of the International Group on Research Reactors

International Nuclear Information System (INIS)

Rosenbalm, K.F.

1995-01-01

The International Group on Research Reactors was formed to facilitate the sharing of knowledge and experience among those institutions and individuals who are actively working to design, build, and promote new research reactors or to make significant upgrades to existing facilities. Twenty-nine papers were presented in five sessions and written versions of the papers or hard copies of the vugraphs used are published in these proceedings. The five sessions were: (1) Operating Research Reactors and Facility Upgrades; (2) Research Reactors in Design and Construction; (3) ANS Closeout Activities; (4) and (5) Research, Development, and Analysis Results

Structural analysis of cellular blocks for a prestressed cast iron reactor pressure vessel

International Nuclear Information System (INIS)

Thomas, R.G.; Head, J.L.

1979-01-01

The cast segments from which the prestressed cast iron nuclear reactor pressure vessel may be constructed are not readily amenable to detailed three-dimensional finite element analysis because their complex internal web structure requires a very large number of elements if reasonable aspect ratios are to be retained. A technique has been developed of modelling these blocks using plate bending elements from the ASKA code. By this means it has been possible to study in detail several designs of casting and to identify favourable features. The results of these studies, and others in which assessments are made of the sensitivity of the structure to prestressing load changes and machining errors, are reported. (orig.)

Upgraded prototype-reactor internal pump for ABWR

International Nuclear Information System (INIS)

Kumagai, Mikio; Amemori, Shiro; Saito, Takehiko

1988-01-01

In 1983, Toshiba, using their own technology, manufactured a commercial grade reactor internal pump (RIP). Recently, however, a licensing agreement with KSB of West Germany covering the RIP technology, has combined the know-how of KSB with Toshiba's technology to produce a truly high-quality prototype RIP. The pump produces the required coreflow for ABWR at low speed and with high efficiency, and simply by increasing the pump speed to the prior level, the coreflow can be further increased for such advantages as improved fuel cycle economy. Here, the advanced features and test results of the RIP are summarized. (author)

3. International conference on catalysis in membrane reactors

Energy Technology Data Exchange (ETDEWEB)

NONE

1998-09-01

The 3. International Conference on Catalysis in Membrane Reactors, Copenhagen, Denmark, is a continuation of the previous conferences held in Villeurbanne 1994 and Moscow 1996 and will deal with the rapid developments taking place within membranes with emphasis on membrane catalysis. The approx. 80 contributions in form of plenary lectures and posters discuss hydrogen production, methane reforming into syngas, selectivity and specificity of various membranes etc. The conference is organised by the Danish Catalytic Society under the Danish Society for Chemical Engineering. (EG)

The assessment of structural dynamics problems in nuclear reactor safety

International Nuclear Information System (INIS)

Liebe, R.

1978-10-01

The paper discusses important physical features of structural dynamics problems in reactor safety. First a general characterization is given of the following problems: Containment deformation due to pool-dynamics during BWR-blowdown; behavior of the core internals due to PWR-blowdown loads; dynamic response of a nuclear power plant during an earthquake; fuel element deformation due to local pressure pulses in an LMFBR core. Several criterias are formulated to classify typical problems so that a better choise can be made both of appropriate mathematical/numerical as well as experimental techniques. The degree of physical coupling between structural dynamics and fluid dynamics is discussed in more detail since it requires particular attention when selecting problem-oriented methods of solution. Some examples are given to illustrate the application and to compare advantages and disadvantages of several numerical methods. Then description is given of experimental techniques in structural dynamics and typical problem areas are identified. Finally some results are presented concerning the fuel element deformation problem in LMFBRs and from the general considerations some important conclusions are summarized. (orig.) 891 RW 892 AP [de

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.

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)

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)

Control Rod Drive Mechanism Installed in the Internal of Reactor Pressure Vessel

Energy Technology Data Exchange (ETDEWEB)

Choi, M. H.; Choi, S.; Park, J. S.; Lee, J. S.; Kim, D. O.; Hur, N. S.; Hur, H.; Yu, J. Y

2008-09-15

This report describes the review results and important technologies related to the in-vessel type control rod drive mechanism. Generally, most of the CRDMs used in the PWR are attached outside of the reactor pressure vessel, and the pernetration of the vessel head can not avoid. However, in-vessel type CRDMs, which are installed inside the reactor vessel, can eliminate the possibility of rod ejection accidents and the penetration of the vessel head, and provide a compact design of the reactor vessel and containment. There are two kinds of in-vessel type CRDM concerning the driving force-driven by a driving motor and by a hydraulic force. Motor driven CRDMs have been mainly investigated in Japan(MRX, IMR, DRX, next generation BWR etc.), and developed the key components such as a canned motor, an integrated rod position indicator, a separating ball-nut and a ball bearing that can operate under the water conditions of a high temperature and pressure. The concept of hydraulically driven CRDMs have been first reported by KWU and Siemens for KWU 200 reactor, and Argentina(CAREM) and China(NHR-5, NHR-200) have been developed the internal CRDM with the piston and cylinder of slightly different geometries. These systems are driven by the hydraulic force which is produced by pumps outside of the reactor vessel and transmitted through a pipe penetrating the reactor vessel, and needs complicated control and piping systems including pumps, valves and pipes etc.. IRIS has been recently decided the internal CRDMs as the reference design, and an analytical and experimental investigations of the hydraulic drive concept are performed by POLIMI in Italy. Also, a small French company, MP98 has been developed a new type of control rods, called 'liquid control rods', where reactivity is controlled by the movement of a liquid absorber in a manometer type device.

Joint ICFRM-14 (14. international conference on fusion reactor materials) and IAEA satellite meeting on cross-cutting issues of structural materials for fusion and fission applications. PowerPoint presentations

International Nuclear Information System (INIS)

2009-01-01

The Conference was devoted to the challenges in the development of new materials for advanced fission, fusion and hybrid reactors. The topics discussed include fuels and materials research under the high neutron fluence; post-irradiation examination; development of radiation resistant structural materials utilizing fission research reactors; core materials development for the advanced fuel cycle initiative; qualification of structural materials for fission and fusion reactor systems; application of charged particle accelerators for radiation resistance investigations of fission and fusion structural materials; microstructure evolution in structural materials under irradiation; ion beams and ion accelerators

Structural capacity assessment of WWER-1000 MW reactor containment. Progress report

International Nuclear Information System (INIS)

Jordanov, M.

1999-01-01

The objective of the project is to provide assessment of the structural behaviour and safety capacity of the WWER-1000 MW Reactor Building Containment at Kozloduy NPP under critical combination of loads according to the current international requirements. The analysis is focused on a realistic assessment of the Containment taking into account the non-linear shell behaviour of the pre-stressed reinforced concrete structure. Previous assessments of the status of pre stressing cables pointed out that the efficiency of the Containment as a final defence barrier for internal and external events depends on their reliability. Due to this, the experimental data obtained from embedded sensors (gauges) at pre-stressed shell structure is to be compared with the results from analytical investigations. The reliability of the WWER-1000 MW accident prevention system is under evaluation in the project. The Soviet standard design WWER-1000 MW type units installed in Kozloduy NPP were originally designed for a Safe Shutdown Earthquake (SSE) with a peak ground acceleration (PGA) of 0.1g. The new site seismicity studies revealed that the seismic hazard for the site significantly exceeds the originally estimated and a Review Level Earthquake (RLE) anchored to PGA=0.20g was proposed for re-assessment of the structures and equipment at Kozloduy NPP. The scope of the study is a re-assessment of the Containment structure under critical combination of loads according to the current safety and reliability requirements, including comparison between the Russian design requirements and the international regulations. Additionally, an investigation of the pre-stressing technology and the annual control of the cables' pre-stressing of the Containment is to be made. The crane influence on the dynamic behaviour of the Containment will be done as well as a study of the integrity of the Containment as a final defence barrier

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)

Introduction [International Reactor Dosimetry File 2002 (IRDF-2002)

International Nuclear Information System (INIS)

Paviotti-Corcuera, R.; Zolnay, E.M.

2006-01-01

The most recently tested version of the International Reactor Dosimetry File, IRDF-90 Version 2 (IRDF-90.2), was released in 1993. Most of the evaluations used in this file were prepared in the mid-1980s, and in the meantime a large amount of new experimental data has become available, along with two new national reactor dosimetry libraries (the Russian Reactor Dosimetry File (RRDF-98) and the Japanese Evaluated Nuclear Data Library (JENDL/D-99)). The cross-sections and related uncertainties for several reactions in these libraries may be of better quality than the data in the older IRDF-90 file. These developments have resulted in different cross-section values being applied to the evaluation of experimental data, creating difficulties in comparing the results of reactor dosimetry calculations from the same types of nuclear facility. Therefore, there has been a strong demand from the reactor dosimetry community for an updated and standardized version of the IRDF. The IAEA has in the past supported similar efforts to improve the quality of data for reactor dosimetry applications. A major objective of the present data development project was to prepare and distribute a standardized, updated and tested reactor dosimetry cross-section library accompanied by uncertainty information (IRDF-2002) for use in service life assessments of nuclear power reactors. In order to achieve this objective, two technical meetings were organized. Both meetings were held at the IAEA in Vienna. The first meeting took place from 27 to 29 August 2002, the second from 1 to 3 October 2003. Recommendations were made concerning the following topics and the preparation of the library: reactions to be included, requirements for new evaluations or revisions, nuclear decay data, radiation damage data, testing of the data in benchmark fields and inclusion of computer codes. The participants emphasized that good quality nuclear data for reactor dosimetry are essential to improve assessments of the

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)

International tokamak reactor conceptual design overview

International Nuclear Information System (INIS)

Stacey, W.M. Jr.

1981-01-01

The International Tokamak Reactor (INTOR) Workshop is an unique collaborative effort among Euratom, Japan, the USA and USSR. The Zero-Phase of the INTOR Workshop, which was conducted during 1979, assessed the technical data base that would support the construction of the next major device in the tokamak program to operate in the early 1990s and defined the objectives and characteristics of this device. The INTOR workshop was extended into phase-1, the Definition Phase, in early 1980. The objective of the Phase-1 Workshop was to develop a conceptual design of the INTOR experiment. The purpose of this paper is to give an overview of the work of the Phase-1 INTOR Workshop (January 1980-June 1981, with emphasis upon the conceptual design

IRPhE - International Reactor Physics Experiments database

International Nuclear Information System (INIS)

Sartori, E.

2004-01-01

The OECD/NEA Nuclear Science Committee (NSC) has identified the need to establish international databases containing all the important experiments that are available for sharing among the specialists and has set up or sponsored specific activities to achieve this. The aim is to preserve them in an agreed standard format in computer accessible form, to use them for international activities involving validation of current and new calculational schemes including computer codes and nuclear data libraries, for assessing uncertainties, confidence bounds and safety margins, and to record measurement methods and techniques. It is a significant saving results from disseminating a standard benchmark set to be used worldwide. A framework for professionals that use the standard benchmark set to validate and verify modeling codes and data for radiation transport, criticality safety and reactor physics applications guarantees a comparative set of analyses. It represents also a good basis for pinpointing important gaps and where efforts should be concentrated and ensures knowledge and competence preservation, management and transfer in nuclear science and engineering. A large number of experimentalists, physicists, evaluators, modelers have devoted large amounts of their efforts and competencies to produce the data on which the methods we are using today are based. These data are far from having been exploited fully for the different nuclear and radiation technologies. This wealth of information needs to be preserved in a form more easily exploitable by modern information technology and for use in connection with novel and refined computational models with limitations of the past removed. These data will form the basis for the studies of more advanced nuclear technology, will be instrumental in identifying areas where there is a lack of knowledge and thus provide support to justifying new experiments that would reduce design uncertainties and consequently costs. Improvement of

Proceedings of the 1994 international meeting on reduced enrichment for research and test reactors

International Nuclear Information System (INIS)

1997-08-01

This meeting brought together participants in the international effort to minimize and eventually eliminate the use of highly enriched uranium in civilian nuclear programs. Papers cover the following topics: National programs; fuel cycle; nuclear fuels; analyses; advanced reactors; and reactor conversions. Selected papers have been indexed separately for inclusion to the Energy Science and Technology Database

Proceedings of the 1994 international meeting on reduced enrichment for research and test reactors

Energy Technology Data Exchange (ETDEWEB)

NONE

1997-08-01

This meeting brought together participants in the international effort to minimize and eventually eliminate the use of highly enriched uranium in civilian nuclear programs. Papers cover the following topics: National programs; fuel cycle; nuclear fuels; analyses; advanced reactors; and reactor conversions. Selected papers have been indexed separately for inclusion to the Energy Science and Technology Database.

Summary report of the final technical meeting on 'International Reactor Dosimetry File: IRDF-2002'

International Nuclear Information System (INIS)

Griffin, Patrick J.; Paviotti-Corcuera, R.

2003-10-01

Presentations, recommendations and conclusions of the Final Technical Meeting on 'International Reactor Dosimetry File: IRDF-2002' are summarized in this report. The main aims of this meeting were to discuss scientific and technical matters related to reactor dosimetry and to assign responsibilities for the preparation of the final version of the IRDF- 2002 library and the associated TECDOC. Tasks were assigned and deadlines were agreed. Participants emphasized that accurate and complete nuclear data for reactor dosimetry are essential to improve the assessment accuracies for reactor pressure vessel service lifetimes in nuclear power plants, as well as for other neutron metrology applications such as boron neutron capture therapy, therapeutic use of medical isotopes, nuclear physics measurements, and reactor safety applications. (author)

Focus on International Cooperation in D and D Cost Estimations for Research Reactors and the Role of the IAEA in It

International Nuclear Information System (INIS)

Daniska, V.; Laraia, M.

2009-01-01

Estimating decommissioning costs and collecting funds are the prerequisites for safe, timely and cost effective decommissioning. These issues were addressed by main international organizations involved in decommissioning - IAEA, OECD/NEA and European Commission. While the activities of OECD/NEA and European Commission are focused mainly on nuclear power plants, the activities of IAEA involve the whole spectrum of nuclear facilities worldwide and the aspects of decommissioning of research reactors, due to their spread in member states, have the special position in IAEA activities. The experience in decommissioning of research reactors was accumulated in developed countries and international activities are needed to support the planning and implementation of decommissioning of research reactors in other IAEA Member States. Decommissioning costing plays the key role among these issues. One of the main results of the common effort of the above listed organisations in costing is the standardised structure of decommissioning cost items which was jointly published as the common platform for harmonisation in decommissioning costing and is recommended for general use. Main subject of the paper is the preliminary presentation of a new IAEA project planned for the period 2009-2011 with the objectives to develop the robust standardised cost estimation methodology, which implements the standardised cost structure. The aim of the proposed methodology is to support the Member States to prepare their own specific cost estimates for research reactors. The planned target is the parametric cost estimating model applicable for various types of research reactors and for various national decommissioning backgrounds. The model is supported by facility inventory database and database of unit factors and data specific for facility to be decommissioned. The paper reviews main types of research reactors; the currant practice for decommissioning costing which will be implemented in the

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

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

Regulatory Assessment Technologies for Aging of Reactor Vessel Internals

Energy Technology Data Exchange (ETDEWEB)

Jhung, Myung Jo; Park, Jeong Soon; Ko, Hanok [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

2013-10-15

In order to develop the audit calculation system, it is required to develop crack evaluation, seismic analysis and thermal-hydraulic analysis techniques for RVIs so that integrity of RVIs under the aging environment can be evaluated and be assured. In addition, regulatory requirements including safety review and inspection guides should be developed in order to assure the quality and uniformity of safety reviews and inspections regarding aging assessment and management of RVIs. Since Reactor Vessel Internals (RVIs) are installed within the reactor pressure vessel and surround the fuel assemblies, some of them are exposed to the environment such as high neutron irradiation, high temperature and reactor coolant flow. Those environmental factors can cause damage to RVIs including cracks, loss of material, fatigue, loss of fracture toughness and change of dimension as the operation time of nuclear power plants (NPPs) increases. For long-term operation more than 40 years, aging management of RVIs is important. The final objectives of this study are to establish the audit calculation system for RVIs and to develop regulatory requirements for aging assessment and management of RVIs considering their operating conditions, materials, and possible aging mechanisms.

The numerical simulation of the WWER-440/V-213 reactor pressure vessel internals response to maximum hypothetical large break loss of coolant accident

International Nuclear Information System (INIS)

Hermansky, P.; Krajcovic, M.

2012-01-01

The reactor internals are designed to ensure cooling of the fuel, to ensure the movement of emergency control assemblies under all operating conditions including accidents and facilitate removal of the fuel and of the internals following an accident This paper presents results of the numerical simulation of the WWER-440/V213 reactor vessel internals dynamic response to maximum hypothetical Large-Break Loss of Coolant Accident. The purpose of this analysis is to determine the reactor vessel internals response due to rapid depressurization and to prove no such deformations occur in the reactor vessel internals which would prevent timely and proper activation of the emergency control assemblies. (Authors)

BWR internals life assurance

International Nuclear Information System (INIS)

Herrera, M.L.; Stancavage, P.P.

1988-01-01

Boiling water reactor (BWR) internal components play an important role in power plant life extension. Many important internals were not designed for easy removal and changes in material properties and local environmental effects due to high radiation makes stress corrosion cracking more likely and more difficult to correct. Over the past several years, operating experience has shown that inspection, monitoring and refurbishment can be accomplished for internal structures with existing technology. In addition, mitigation techniques which address the causes of degradation are available to assure that life extension targets can be met. This paper describes the many considerations and aspects when evaluating life extension for reactor vessel internals

Proceedings of the 1988 International Meeting on Reduced Enrichment for Research and Test Reactors

Energy Technology Data Exchange (ETDEWEB)

1993-07-01

The international effort to develop and implement new research reactor fuels utilizing low-enriched uranium, instead of highly- enriched uranium, continues to make solid progress. This effort is the cornerstone of a widely shared policy aimed at reducing, and possibly eliminating, international traffic in highly-enriched uranium and the nuclear weapon proliferation concerns associated with this traffic. To foster direct communication and exchange of ideas among the specialists in this area, the Reduced Enrichment Research and Test Reactor (RERTR) Program, at Argonne National Laboratory, sponsored this meeting as the eleventh of a series which began 1978. Individual papers presented at the meeting have been cataloged separately.

Proceedings of the 1988 International Meeting on Reduced Enrichment for Research and Test Reactors

International Nuclear Information System (INIS)

1993-07-01

The international effort to develop and implement new research reactor fuels utilizing low-enriched uranium, instead of highly- enriched uranium, continues to make solid progress. This effort is the cornerstone of a widely shared policy aimed at reducing, and possibly eliminating, international traffic in highly-enriched uranium and the nuclear weapon proliferation concerns associated with this traffic. To foster direct communication and exchange of ideas among the specialists in this area, the Reduced Enrichment Research and Test Reactor (RERTR) Program, at Argonne National Laboratory, sponsored this meeting as the eleventh of a series which began 1978. Individual papers presented at the meeting have been cataloged separately

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

Internal fluid flow management analysis for Clinch River Breeder Reactor Plant sodium pumps

International Nuclear Information System (INIS)

Cho, S.M.; Zury, H.L.; Cook, M.E.; Fair, C.E.

1978-12-01

The Clinch River Breeder Reactor Plant (CRBRP) sodium pumps are currently being designed and the prototype unit is being fabricated. In the design of these large-scale pumps for elevated temperature Liquid Metal Fast Breeder Reactor (LMFBR) service, one major design consideration is the response of the critical parts to severe thermal transients. A detailed internal fluid flow distribution analysis has been performed using a computer code HAFMAT, which solves a network of fluid flow paths. The results of the analytical approach are then compared to the test data obtained on a half-scale pump model which was tested in water. The details are presented of pump internal hydraulic analysis, and test and evaluation of the half-scale model test results

International conference on fast reactors and related fuel cycles (FR09): Challenges and opportunities. CN-176 presentations

International Nuclear Information System (INIS)

2009-01-01

Renewed interest in nuclear energy is driven by the need to develop carbon free energy sources, by demographics and development in emerging economies, as well as by security of supply concerns. It is expected that nuclear energy will deliver huge amounts of energy to both emerging and developed economies. However, acceptance of large scale contributions would depend on satisfaction of key drivers to enhance sustainability in terms of economics, safety, adequacy of natural resources, waste reduction, non-proliferation and public acceptance. Fast spectrum reactors with recycle enhance the sustainability indices significantly. This has led to the focus on fast spectrum reactors with recycle in the Generation IV International Forum (GIF) and the International Project on Innovative Nuclear Reactors and Fuel Cycles (INPRO) initiative of the IAEA. It is expected that 2009 will register major events in the domain of fast spectrum reactors, that is, the restart of Monju in Japan, the first criticality of the China Experimental Fast Reactor in China, as well as new insights through end-of-life studies in Phenix, France. New fast reactors are expected to be commissioned in the near future: the 500 MW(e) Prototype Fast Breeder Reactor in India and the BN-800 unit in the Russian Federation. Moreover, China, France, India, Japan, Republic of Korea and the United States of America are preparing advanced prototypes/ demonstrations and/or commercial reactors for the 2020-2030 horizon. The necessary condition for successful fast reactor deployment in the near and mid-term is the understanding and assessment of innovative technological and design options, based on both past knowledge and experience, as well as on ongoing research and technology development efforts. In this respect, the need for in-depth international information exchange is underscored by the fact that the last large international fast reactor conference was held as far back as 1991. Since then, progress in research

Materials for advanced water cooled reactors

International Nuclear Information System (INIS)

1992-09-01

The current IAEA programme in advanced nuclear power technology promotes technical information exchange between Member States with major development programmes. The International Working Group on Advanced Technologies for Water Cooled Reactors recommended to organize a Technical Committee Meeting for the purpose of providing an international forum for technical specialists to review and discuss aspects regarding development trends in material application for advanced water cooled reactors. The experience gained from the operation of current water cooled reactors, and results from related research and development programmes, should be the basis for future improvements of material properties and applications. This meeting enabled specialists to exchange knowledge about structural materials application in the nuclear island for the next generation of nuclear power plants. Refs, figs, tabs

Overview of the 2014 Edition of the International Handbook of Evaluated Reactor Physics Benchmark Experiments (IRPhEP Handbook)

Energy Technology Data Exchange (ETDEWEB)

John D. Bess; J. Blair Briggs; Jim Gulliford; Ian Hill

2014-10-01

The International Reactor Physics Experiment Evaluation Project (IRPhEP) is a widely recognized world class program. The work of the IRPhEP is documented in the International Handbook of Evaluated Reactor Physics Benchmark Experiments (IRPhEP Handbook). Integral data from the IRPhEP Handbook is used by reactor safety and design, nuclear data, criticality safety, and analytical methods development specialists, worldwide, to perform necessary validations of their calculational techniques. The IRPhEP Handbook is among the most frequently quoted reference in the nuclear industry and is expected to be a valuable resource for future decades.

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

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)

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

Rotating bed reactor for CLC: Bed characteristics dependencies on internal gas mixing

International Nuclear Information System (INIS)

Håkonsen, Silje Fosse; Grande, Carlos A.; Blom, Richard

2014-01-01

Highlights: • A mathematical model for the rotating CLC reactor has been developed. • The model reflects the gas distribution in the reactor during CLC operation. • Radial dispersion in the rotating bed is the main cause for internal gas mixing. • The model can be used to optimize the reactor design and particle characteristics. - Abstract: A newly designed continuous lab-scale rotating bed reactor for chemical looping combustion using CuO/Al 2 O 3 oxygen carrier spheres and methane as fuel gives around 90% CH 4 conversion and >90% CO 2 capture efficiency based on converted methane at 800 °C. However, from a series of experiments using a broad range of operating conditions potential CO 2 purities only in the range 20–65% were yielded, mostly due to nitrogen slip from the air side of the reactor into the effluent CO 2 stream. A mathematical model was developed intending to understand the air-mixing phenomena. The model clearly reflects the gas slippage tendencies observed when varying the process conditions such as rotation frequency, gas flow and the flow if inert gas in the two sectors dividing the air and fuel side of the reactor. Based on the results, it is believed that significant improvements can be made to reduce gas mixing in future modified and scaled-up reactor versions

Transactions of the 10th international conference on structual mechanics in reactor technology

International Nuclear Information System (INIS)

Hadjian, A.H.

1989-01-01

In this book, a wide spectrum of subjects is covered, including theoretical developments in structural mechanics, loading conditions, behavior of materials, fluid mechanics, operating experience, accident sequences, and calculational procedures. As a result, 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. The papers are theoretical or applied, or they address both of these aspects to demonstrate application of developed methods to solve specific design problems and show how well actual behavior correlates with theory. These paper explore in detail the mechanical design and system integration of fusion power reactors; thermohydraulics, structural mechanics and life-time evaluations of reactor components as first wall diverter/limiter, plasma heating devices, breeding blanket and shielding, magnet coils and supports, and vacuum containment systems, and structural analysis and comparison with measured data

Modeling the dynamics of internal flooding - verification analysis

International Nuclear Information System (INIS)

Filipov, K.

2011-01-01

The results from conducted software WATERFOW's verification analysis, developed for the purposes of reactor building internal flooding analysis have been presented. For the purpose of benchmarking the integrated code MELCOR is selected. Considering the complex structure of reactor building, the sample tests were used to cover the characteristic points of the internal flooding analysis. The inapplicability of MELCOR to the internal flooding study has been proved

Vibrations measurement in fast and PWR reactor study

International Nuclear Information System (INIS)

Tigeot, Y.; Epstein, A.; Hareux, F.

1975-01-01

In the past severe damages have occured in several nuclear reactors, by structural vibrations induced by the primary cooling flow. To avoid this kind of troubles, the SEMT makes studies for two different types of reactors. For the light pressurized water reactors, some tests have been made on the SAFRAN test loop which is a three loop 1/8 scale internal model of a 900 MWe reactor. This study is actually undertaken jointly with Framatome. Elsewhere, measurements have been made on the Phenix fast breeder sodium reactor, and studies are planned for the Super Phenix reactor [fr

International Thermonuclear Experimental Reactor (ITER). Engineering Design Activities (EDA). Agreement and protocol 1

International Nuclear Information System (INIS)

1992-01-01

This document contains protocol 1 to the agreement among the European Atomic Energy Community, the government of Japan, the Government of the Russian Federation, and the Government of the United States of America on cooperation in the engineering design activities for the International Thermonuclear Experimental Reactor, which activities shall be conducted under the auspices of the International Atomic Energy Agency

Development of Safety Review Guide for the Periodic Safety Review of Reactor Vessel Internals

International Nuclear Information System (INIS)

Park, Jeongsoon; Ko, Hanok; Kim, Seonjae; Jhung, Myungjo

2013-01-01

Aging management of the reactor vessel internals (RVIs) is one of the important issues for long-term operation of nuclear power plants (NPPs). Safety review on the assessment and management of the RVI aging is conducted through the process of a periodic safety review (PSR). The regulatory body should check that reactor facilities sustain safety functions in light of degradation due to aging and that the operator of a nuclear power reactor establishes and implements management program to deal with degradation due to aging in order to guarantee the safety functions and the safety margin as a result of PSR. KINS(Korea Institute of Nuclear Safety) has utilized safety review guides (SRG) which provide guidance to KINS staffs in performing safety reviews in order to assure the quality and uniformity of staff safety reviews. The KINS SRGs for the continued operation of pressurized water reactors (PWRs) published in 2006 contain areas of review regarding aging management of RVIs in chapter 2 (III.2.15, Appendix 2.0.1). However unlike the SRGs for the continued operation, KINS has not officially published the SRGs for the PSR of PWRs, but published them as a form of the research report. In addition to that, the report provides almost same review procedures for aging assessment and management of RVIs with the ones provided in the SRGs for the continued operation, it cannot provide review guidance specific to PSRs. Therefore, a PSR safety review guide should be developed for RVIs in PWRs. In this study, a draft PSR safety review guide for reactor vessel internals in PWRs is developed and provided. In this paper, a draft PSR safety review guide for reactor vessel internals (PSR SRG-RVIs) in PWRs is introduced and main contents of the draft are provided. However, since the PSR safety review guides for areas other than RVIs in the pressurized water reactors (PWRs) are expected to be developed in the near future, the draft PSR SRG-RVIs should be revisited to be compatible with

Mathematical models in Slowpoke reactor internal irradiation site

International Nuclear Information System (INIS)

Raza, J.

2007-01-01

The main objective is to build representative mathematical models of neutron activation analysis in a Slowpoke internal irradiation site. Another significant objective is to correct various elements neutron activation analysis measured mass using these models. The neutron flux perturbation is responsible for the measured under-estimation of real masses. We supposed that neutron flux perturbation measurements taken during the Ecole Polytechnique de Montreal Slowpoke reactor first fuel loading were still valid after the second fuelling. .We also supposed that the thermal neutrons spatial and kinetic energies distributions as well as the absorption microscopic cross section dependence on the neutrons kinetic energies were important factors to satisfactorily represent neutron activation analysis results. In addition, we assumed that the neutron flux is isotropic in the laboratory system. We used experimental results from the Slowpoke reactor internal irradiation sites, in order to validate our mathematical models. Our models results are in close agreement with these experimental results..We established an accurate global mathematical correlation of the neutron flux perturbation in function of samples volumes and macroscopic neutron absorption cross sections. It is applicable to sample volumes ranging from 0,1 to 1,3 ml and macroscopic neutron absorption cross section up to 5 moles-b for seven (7) elements with atomic numbers (Z) ranging from 5 to 79. We first came up with a heuristic neutron transport mathematical semi-analytical model, in order to better understand neutrons behaviour in presence of one of several different nuclei samples volumes and mass. In order to well represent the neutron flux perturbation, we combined a neutron transport solution obtained from the spherical harmonics method of a finite cylinder and a mathematical expression combining two cylindrical harmonic functions..With the help of this model and the least squares method, we made extensive

Trajectory-tracking control of underwater inspection robot for nuclear reactor internals using Time Delay Control

International Nuclear Information System (INIS)

Park, Joon-Young; Cho, Byung-Hak; Lee, Jae-Kyung

2009-01-01

This paper addresses the trajectory control problem of an underwater inspection robot for nuclear reactor internals. From the viewpoint of control engineering, the trajectory control of the underwater robot is a difficult task due to its nonlinear dynamics, which includes various hydraulic forces such as buoyancy and hydrodynamic damping, the difference between the centres of gravity and buoyancy, and disturbances from a tether cable. To solve such problems, we applied Time Delay Control to the underwater robot. This control law has a very simple structure not requiring nonlinear plant dynamics, and was proven to be highly robust against nonlinearities, uncertainties and disturbances. We confirmed its effectiveness through experiments.

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.

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

A structure for the protection of nuclear-reactor pressurized-vessels against rupture

International Nuclear Information System (INIS)

Marcellin, J.-P.; Aubert, Gilles

1974-01-01

Description is given of a structure for the protection of nuclear-reactor pressurized-vessels against rupture. Said structure comprises a pre-stressed concrete tank adapted to surround the tank side-wall and bottom, said tank being higher than said vessel, said tank being provided with ports for passing cooling fluid ducts therethrough, and a crown adapted to rest along the periphery of the reactor-cover and made integral therewith. This can be applied to reactors of the PWR type [fr

Computing radiation dose to reactor pressure vessel and internals

International Nuclear Information System (INIS)

1996-01-01

Within the next twenty years many of the nuclear reactors currently in service will reach their design lifetime. One of the key factors affecting decisions on license extensions will be the ability to confidently predict the integrity of the reactor pressure vessel and core structural components which have been subjected to many years of cumulative radiation exposure. This report gives an overview of the most recent scientific literature and current methodologies for computational dosimetry in the OECD/NEA Member countries. Discussion is extended to consider some related issues of materials science, such as the metals, and limitations of the models in current use. Proposals are made for further work. (author)

Design considerations for ITER [International Thermonuclear Experimental Reactor] magnet systems

International Nuclear Information System (INIS)

Henning, C.D.; Miller, J.R.

1988-01-01

The International Thermonuclear Experimental Reactor (ITER) is now completing a definition phase as a beginning of a three-year design effort. Preliminary parameters for the superconducting magnet system have been established to guide further and more detailed design work. Radiation tolerance of the superconductors and insulators has been of prime importance, since it sets requirements for the neutron-shield dimension and sensitively influences reactor size. The major levels of mechanical stress in the structure appear in the cases of the inboard legs of the toroidal-field (TF) coils. The cases of the poloidal-field (PF) coils must be made thin or segmented to minimize eddy current heating during inductive plasma operation. As a result, the winding packs of both the TF and PF coils includes significant fractions of steel. The TF winding pack provides support against in-plane separating loads but offers little support against out-of-plane loads, unless shear-bonding of the conductors can be maintained. The removal of heat due to nuclear and ac loads has not been a fundamental limit to design, but certainly has non-negligible economic consequences. We present here preliminary ITER magnetic systems design parameters taken from trade studies, designs, and analyses performed by the Home Teams of the four ITER participants, by the ITER Magnet Design Unit in Garching, and by other participants at workshops organized by the Magnet Design Unit. The work presented here reflects the efforts of many, but the responsibility for the opinions expressed is the authors'. 4 refs., 3 figs., 4 tabs

Flow model study of 'Monju' reactor vessel

International Nuclear Information System (INIS)

Miyaguchi, Kimihide

1980-01-01

In the case of designing the structures in nuclear reactors, various problems to be considered regarding thermo-hydrodynamics exist, such as the distribution of flow quantity and the pressure loss in reactors and the thermal shock to inlet and outlet nozzles. In order to grasp the flow characteristics of coolant in reactors, the 1/2 scale model of the reactor structure of ''Monju'' was attached to the water flow testing facility in the Oarai Engineering Center, and the simulation experiment has been carried out. The flow characteristics in reactors clarified by experiment and analysis so far are the distribution of flow quantity between high and low pressure regions in reactors, the distribution of flow quantity among flow zones in respective regions of high and low pressure, the pressure loss in respective parts in reactors, the flow pattern and the mixing effect of coolant in upper and lower plenums, the effect of the twisting angle of inlet nozzles on the flow characteristics in lower plenums, the effect of internal cylinders on the flow characteristics in upper plenums and so on. On the basis of these test results, the improvement of the design of structures in reactors was made, and the confirmation test on the improved structures was carried out. The testing method, the calculation method, the test results and the reflection to the design of actual machines are described. (Kako, I.)

Growth and Expansion of the International Criticality Safety Benchmark Evaluation Project and the Newly Organized International Reactor Physics Experiment Evaluation Project

International Nuclear Information System (INIS)

J. Blair Briggs; Lori Scott; Yolanda Rugama; Enrico Satori

2007-01-01

Since ICNC 2003, the International Criticality Safety Benchmark Evaluation Project (ICSBEP) has continued to expand its efforts and broaden its scope. Criticality-alarm/shielding type benchmarks and fundamental physics measurements that are relevant to criticality safety applications are not only included in the scope of the project, but benchmark data are also included in the latest version of the handbook. A considerable number of improvements have been made to the searchable database, DICE and the criticality-alarm/shielding benchmarks and fundamental physics measurements have been included in the database. There were 12 countries participating on the ICSBEP in 2003. That number has increased to 18 with recent contributions of data and/or resources from Brazil, Czech Republic, Poland, India, Canada, and China. South Africa, Germany, Argentina, and Australia have been invited to participate. Since ICNC 2003, the contents of the ''International Handbook of Evaluated Criticality Safety Benchmark Experiments'' have increased from 350 evaluations (28,000 pages) containing benchmark specifications for 3070 critical or subcritical configurations to 442 evaluations (over 38,000 pages) containing benchmark specifications for 3957 critical or subcritical configurations, 23 criticality-alarm-placement/shielding configurations with multiple dose points for each, and 20 configurations that have been categorized as fundamental physics measurements that are relevant to criticality safety applications in the 2006 Edition of the ICSBEP Handbook. Approximately 30 new evaluations and 250 additional configurations are expected to be added to the 2007 Edition of the Handbook. Since ICNC 2003, a reactor physics counterpart to the ICSBEP, The International Reactor Physics Experiment Evaluation Project (IRPhEP) was initiated. Beginning in 1999, the IRPhEP was conducted as a pilot activity by the by the Organization of Economic Cooperation and Development (OECD) Nuclear Energy Agency

New Reactor Siting, Licensing and Construction Experience. Proceedings of the 2. CNRA International Workshop on 'New Reactor Siting, Licensing and Construction Experience'

International Nuclear Information System (INIS)

2013-01-01

This report documents the proceedings from the 2. Workshop on New Reactor Siting, Licensing and Construction Experience. A total of 45 specialists from 16 countries and international organisations attended. The meeting was sponsored by the OECD Nuclear Energy Agency Committee on Nuclear Regulatory Activities and hosted by the US Nuclear Regulatory Commission (U.S.NRC). The objectives of the workshop were to provide a forum to exchange information on lessons learned from siting, licensing and constructing new nuclear power plants around the world. Key focus areas included siting practices and regulatory positions that have been enhanced as a result of the Fukushima accident; lessons learned from licensing and design review approaches and challenges, construction experience and recommendations for regulatory oversight; and regulatory cooperation on generic and design specific issues through the MDEP specific working groups. The workshop was structured in 4 technical sessions, each followed by ample time for panel discussions. The first technical session was devoted to regulatory cooperation on generic and design specific issues, MDEP working groups (EPR, AP1000), vendor inspection co-operation, digital I and C, and codes and standards. The second technical session was intended to discuss and share regulatory positions on siting practices and enhancements as a result of lessons learned from Fukushima accident. The third technical session addressed the construction experience and regulatory oversight of new reactor construction activities. And the fourth technical session included presentations on the lessons learned from regulatory licensing reviews of new reactor designs

Progress toward international agreement to improve reactor safety

International Nuclear Information System (INIS)

Lieberman, J.I.; Graham, B.

1993-01-01

Representatives of nearly one-half of the 114 member states of the International Atomic Energy Agency (IAEA), including the United States, have participated in the development of an international nuclear safety conventions proposed multilateral treaty to improve civil nuclear power reactor safety. A preliminary draft of the convention has been developed (referred to as the draft convention for this report), but discussions are continuing, and when the final convention text will be completed and presented to IAEA member states for signature is uncertain. This report responds to the former and current Chairman's request that we provide information on the development of the nuclear safety convention, including a discussion of (1) the draft convention's scope and objectives, (2) how the convention will be implemented and monitored, (3) the views of selected country representatives on what provisions should be included in the draft convention, and (4) the convention's potential benefits and limitations

Status of international HTGR [high-temperature gas-cooled reactor] development

International Nuclear Information System (INIS)

Homan, F.J.; Simon, W.A.

1988-01-01

Programs for the development of high-temperature gas-cooled reactor (HTGR) technology over the past 30 years in eight countries are briefly described. These programs have included both government sector and industrial participation. The programs have produced four electricity-producing prototype/demonstration reaactors, two in the United States, and two in the Federal Republic of Germany. Key design parameters for these reactors are compared with the design parameters planned for follow-on commercial-scale HTGRs. The development of HTGR technology has been enhanced by numerous cooperative agreements over the years, involving both government-sponsored national laboratories and industrial participants. Current bilateral cooperative agreements are described. A relatively new component in the HTGR international cooperation is that of multinational industrial alliances focused on supplying commercial-scale HTGR power plants. Current industrial cooperative agreements are briefly discussed

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

Statistical analysis of the vibration loading of the reactor internals and fuel assemblies of reactor units type WWER-440 from deferent projects

International Nuclear Information System (INIS)

Ovcharov, O.; Pavelko, V.; Usanov, A.; Arkadov, G.; Dolgov, A.; Molchanov, V.; Anikeev, J.; Pljush, A.

2006-01-01

In this paper the following items have been presented: 1) Vibration noise instrument channels; 2) Vibration loading characteristics of control assemblies, internals and design peculiarities of internals of WWER-440 deferent projects; 3) Coolant flow rate through the reactor, reactor core, fuel assemblies and control assemblies for different projects WWER-440 and 4) Noise measurements of coolant speed per channel. The change of auto power spectrum density of absolute displacement detector signal for the last 12 years of SUS monitoring of the Kola NPP unit 2; the coherence functions groups between two SPND of the same level for the Kola NPP unit 1; the measured coolant flow rate at Paks NPP and the auto power spectrum density group of SPND signals from 11 neutron measuring channels of the Kola NPP unit 1 are given. The main factors of vibration loading of internals and fuel assemblies for Kola NPP units 1-4, Bohunice NPP units 1 and 2 and Novovoronezh NPP units 3 and 4 are also discussed

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.)

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.

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

Nuclear reactor construction with bottom supported reactor vessel

International Nuclear Information System (INIS)

Sharbaugh, J.E.

1987-01-01

This patent describes an improved liquid metal nuclear reactor construction comprising: (a) a nuclear reactor core having a bottom platform support structure; (b) a reactor vessel for holding a large pool of low pressure liquid metal coolant and housing the core; (c) a containment structure surrounding the reactor vessel and having a sidewall spaced outwardly from the reactor vessel side wall and having a base mat spaced below the reactor vessel bottom end wall; (d) a central small diameter post anchored to the containment structure base mat and extending upwardly to the reactor vessel to axially fix the bottom end wall of the reactor vessel and provide a center column support for the lower end of the reactor core; (e) annular support structure disposed in the reactor vessel on the bottom end wall and extending about the lower end of the core; (f) structural support means disposed between the containment structure base mat and bottom end of the reactor vessel wall and cooperating for supporting the reactor vessel at its bottom end wall on the containment structure base mat to allow the reactor vessel to expand radially but substantially prevent any lateral motions that might be imposed by the occurrence of a seismic event; (g) a bed of insulating material disposed between the containment structure base mat and the bottom end wall of the reactor vessel and uniformly supporting the reactor vessel at its bottom end wall; freely expand radially from the central post as it heats up while providing continuous support thereof; (h) a deck supported upon the wall of the containment vessel above the top open end of the reactor vessel; and (i) extendible and retractable coupling means extending between the deck and the top open end of the reactor vessel and flexibly and sealably interconnecting the reactor vessel at its top end to the deck

Technical mechanics in constructional reactor safety

International Nuclear Information System (INIS)

Matthees, W.

1979-01-01

Reactor safety is based on close cooperation between a number of technical and scientific disciplines; most problems of reactor technology can be solved with the aid of technical mechanics. At the 5th International Conference on Structural Mechanics in Reactor Technology (5th SMIRT), one of the biggest conferences in the field of applied technical mechanics, about 800 papers were read giving the latest state of knowledge in the field of constructional reactor safety. The main subject of the conference was the analysis of material behaviour under high loads; the information and methods of these analysis go far beyond what is required in the conventional field. (orig./UA) [de

Summary report of the technical meeting on 'International Reactor Dosimetry File: IRDF-2002'

International Nuclear Information System (INIS)

Greenwood, L.R.; Paviotti-Corcuera, R.

2002-09-01

This report summarizes the presentations, recommendations and conclusions of the Technical Meeting on 'International Reactor Dosimetry File: IRDF-2002.' The purpose of this meeting was to discuss scientific and technical matters related to the subject and coordinate related tasks. Discussions were held and recommendations were given for the preparation of the files on topics related to: reactions to be included, need for new evaluations or revisions, decay data, radiation damage data, integral testing in benchmark fields, and computer codes to be included. Tasks were assigned and deadlines were set. The participants emphasized that accurate and complete knowledge of nuclear data for reactor dosimetry are essential for improving the accuracy of the reactor pressure vessel service life assessment of nuclear power plants as well as in other neutron metrology applications such as boron neutron capture therapy, therapeutic use of medical isotopes, nuclear physics measurements, and reactor safety applications. (author)

The IRIS consortium: international cooperation in advanced reactor development

International Nuclear Information System (INIS)

Carelli, M.; Petrovic, B.; Miller, K.; Lombardi, C.; Ricotti, M.E.

2005-01-01

Besides its many outstanding technical innovations in the design and safety, the most innovative feature of the International Reactor Innovative and Secure (IRIS), is perhaps the international cooperation which carries on its development. IRIS is designed by an international consortium which currently numbers 21 organizations from ten countries across four continents. It includes reactor, fuel and fuel cycle vendors, component manufacturers, laboratories, academia, architect engineers and power producers. The defining organizational characteristics of IRIS is that while Westinghouse has overall lead and responsibility, this lead is of the type of 'primus inter pares' (first among equals) rather than the traditional owner versus suppliers/contractors relationship. All members of the IRIS consortium contribute and expect to have a return, should IRIS be successfully deployed, commensurate to their investment. The nature of such return will be tailored to the type of each organization, because it will of course be of a different nature for say a component manufacturer, university, or architect engineer. One fundamental tenet of the consortium is that all members, regardless of their amount of contribution, have equal access to all information developed within the project. Technical work is thus being coordinated by integrated subgroups and the whole team meets twice a year to perform an overall review of the work, discuss policy and strategy and plan future activities. Personnel from consortium members have performed internships, mostly at Westinghouse locations in Pittsburgh, Pennsylvania, and Windsor, Connecticut, but also at other members, as it has been the case for several graduate students. In fact, more than one hundred students at the various universities have been working on IRIS, most of them conducting graduate theses at the master or doctoral level. The IRIS experience has proved very helpful to the students in successfully landing their employment choice

Remote maintenance for fusion experimental reactor

International Nuclear Information System (INIS)

Koizumi, Koichi; Takeda, Nobukazu

2000-01-01

Here was introduced on maintenance of reactor core portion operated by remote control among maintenance of the International Thermonuclear Experimental Reactor (ITER) begun on its design since 1988 under international cooperation of U.S.A., Europe, Russia and Japan. Every appliances constructing the reactor core portion is necessary to carry out all of their inspection and maintenance by using remote controlled apparatus because of their radiation due to neutron generated by DT combustion of plasma. For engineering design activity (EDA) in ITER, not only design and development of the remote control appliances but also design under consideration of remote maintenance for from structural design of maintained objective appliances to access method to appliances, transportation and preservation method of radiated matters, and out-reactor maintenance in a hot cell, is now under progress. Here were also reported on basic concept on maintenance and conservation of ITER, maintenance design of diverter and blanket with high maintenance frequency and present state on development of maintenance appliances. (G.K.)

Fifteenth annual meeting of the International Working Group on Fast Reactors. Summary report

International Nuclear Information System (INIS)

1982-09-01

The Fifteenth Annual Meeting of the IWGFR was held in accordance with the recommendation of the previous Annual Group Meeting, in Obninsk, USSR, Vienna from 30 March to 2 April 1982. The meeting was attended by the Member States of the group: France, the Federal Republic of Germany, Italy, Japan, the United Kingdom, and the USA, as well as by representatives from CEC, IAEA and OECD and observer from the USSR. This document includes: review of the IWGFR Activities for the period since the Eleventh Annual Meeting of the Group; preliminary programme of international conference on breeder reactors as a world energy resource and the breeder fuel cycle; list of meetings on atomic energy which may be of interest to the IWGFR Members; IWGFR criteria for supporting some of the international conferences; list of proposed topics for the IWGFR Specialists' Meetings; list of topics for review articles on LMFBR recommended for publication by the IAEA; list of meetings sponsored by the IWGFR; a list of members of the International Working Group on Fast Reactors

Sixteenth annual meeting of the International Working Group on Fast Reactors. Summary report

Energy Technology Data Exchange (ETDEWEB)

NONE

1983-10-01

The Sixteenth Annual Meeting of the IWGFR was held in accordance with the recommendation of the previous Annual Meeting Group, in Vienna from 12-15 April 1983. The meeting was attended by the Member States of the group: France, the Federal Republic of Germany, Italy, Japan, the United Kingdom, and the USA, as well as by representatives from CEC, IAEA and OECD and observer from the USSR. This document includes: review of the IWGFR Activities for the period since the Eleventh Annual Meeting of the Group; preliminary programme of international conference on breeder reactors as a world energy resource and the breeder fuel cycle; list of meetings on atomic energy which may be of interest to the IWGFR Members; IWGFR criteria for supporting some of the international conferences; list of proposed topics for the IWGFR Specialists' Meetings; list of topics for review articles on LMFBR recommended for publication by the IAEA; list of meetings sponsored by the IWGFR; a list of members of the International Working Group on Fast Reactors.

Fifteenth annual meeting of the International Working Group on Fast Reactors. Summary report

Energy Technology Data Exchange (ETDEWEB)

NONE

1982-09-01

The Fifteenth Annual Meeting of the IWGFR was held in accordance with the recommendation of the previous Annual Group Meeting, in Obninsk, USSR, Vienna from 30 March to 2 April 1982. The meeting was attended by the Member States of the group: France, the Federal Republic of Germany, Italy, Japan, the United Kingdom, and the USA, as well as by representatives from CEC, IAEA and OECD and observer from the USSR. This document includes: review of the IWGFR Activities for the period since the Eleventh Annual Meeting of the Group; preliminary programme of international conference on breeder reactors as a world energy resource and the breeder fuel cycle; list of meetings on atomic energy which may be of interest to the IWGFR Members; IWGFR criteria for supporting some of the international conferences; list of proposed topics for the IWGFR Specialists' Meetings; list of topics for review articles on LMFBR recommended for publication by the IAEA; list of meetings sponsored by the IWGFR; a list of members of the International Working Group on Fast Reactors.

Sixteenth annual meeting of the International Working Group on Fast Reactors. Summary report

International Nuclear Information System (INIS)

1983-10-01

The Sixteenth Annual Meeting of the IWGFR was held in accordance with the recommendation of the previous Annual Meeting Group, in Vienna from 12-15 April 1983. The meeting was attended by the Member States of the group: France, the Federal Republic of Germany, Italy, Japan, the United Kingdom, and the USA, as well as by representatives from CEC, IAEA and OECD and observer from the USSR. This document includes: review of the IWGFR Activities for the period since the Eleventh Annual Meeting of the Group; preliminary programme of international conference on breeder reactors as a world energy resource and the breeder fuel cycle; list of meetings on atomic energy which may be of interest to the IWGFR Members; IWGFR criteria for supporting some of the international conferences; list of proposed topics for the IWGFR Specialists' Meetings; list of topics for review articles on LMFBR recommended for publication by the IAEA; list of meetings sponsored by the IWGFR; a list of members of the International Working Group on Fast Reactors

Wear plates control rod guide tubes top internal reactor vessel C. N. VANDELLOS II

International Nuclear Information System (INIS)

2010-01-01

The guide tubes for control rods forming part of the upper internals of the reactor vessel, its function is to guide the control rod to permit its insertion in the reactor core. These guide tubes are suspended from the upper support plate which are fixed by bolts and extending to the upper core plate which is fastened by clamping bolts (split pin) to prevent lateral displacement of the guide tubes, while allowing axial expansion.

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)

Multipurpose research reactors

International Nuclear Information System (INIS)

1988-01-01

The international symposium on the utilization of multipurpose research reactors and related international co-operation was organized by the IAEA to provide for information exchange on current uses of research reactors and international co-operative projects. The symposium was attended by about 140 participants from 36 countries and two international organizations. There were 49 oral presentations of papers and 24 poster presentations. The presentations were divided into 7 sessions devoted to the following topics: neutron beam research and applications of neutron scattering (6 papers and 1 poster), reactor engineering (6 papers and 5 posters), irradiation testing of fuel and material for fission and fusion reactors (6 papers and 10 posters), research reactor utilization programmes (13 papers and 4 posters), neutron capture therapy (4 papers), neutron activation analysis (3 papers and 4 posters), application of small reactors in research and training (11 papers). A separate abstract was prepared for each of these papers. Refs, figs and tabs

The FRJ-1 (MERLIN) research reactor: its main activity inventory has been removed by successful demolition of the reactor block

International Nuclear Information System (INIS)

Stahn, B.; Printz, R.; Matela, K.; Zehbe, C.; Poeppinghaus, J.; Cremer, J.

2004-01-01

The FRJ-1 (MERLIN) research reactor was decommissioned in 1985 after twenty-three years of operation. Demolition of the plant was begun in 1996. The article contains a survey of the demolition steps carried out so far within the framework of three partial permits. The main activity is the demolition of the reactor core structures as a precondition for subsequent measures to ensure clearance measurements of the building. The core structures are demolished which were exposed to high neutron fluxes during reactor operation and now show the highest activity and dose rate levels, except for the core internals. For demolition and disassembly of the metal structures in this part of the plant, the tools specially designed and made include a remotely operated sawing system and a pipe cutting system for internal segmentation of the beam lines. The universal demolition tool for use also above and beyond the concrete structures has been found to be a remotely controlled electrohydraulic demolition shovel. Spreading contamination in the course of the demolition work was avoided. One major reason for this success was the fact that no major airborne contamination existed at any time as a consequence of the quality of the material demolished and also of the consistent use of technical tools. While the reactor block was being demolished, an application for clearance measurement of the reactor hall and subsequent release from the scope of the Atomic Energy Act was filed as early as in mid-2003. The fourth partial permit covering these activities is expected to be issued in the spring of 2004. (orig.)

The inner containment of an EPR trademark pressurized water reactor

Energy Technology Data Exchange (ETDEWEB)

Ostermann, Dirk; Krumb, Christian; Wienand, Burkhard [AREVA GmbH, Offenbach (Germany)

2014-08-15

On February 12, 2014 the containment pressure and subsequent leak tightness tests on the containment of the Finnish Olkiluoto 3 EPR trademark reactor building were completed successfully. The containment of an EPR trademark pressurized water reactor consists of an outer containment to protect the reactor building against external hazards (such as airplane crash) and of an inner containment that is subjected to internal overpressure and high temperature in case of internal accidents. The current paper gives an overview of the containment structure, the design criteria, the validation by analyses and experiments and the containment pressure test.

Development and Applicability Demonstration of a Remote Inspection Module for Inspection of Reactor Internals in an SFR

International Nuclear Information System (INIS)

Kim, Hoewoong; Joo, Youngsang; Park, Changgyu; Kim, Jongbum; Bae, Jinho

2014-01-01

Since liquid sodium is optically opaque, the ultrasonic inspection technique has been mainly employed for inspection of reactor internals in a Sodium-cooled Fast Reactor (SFR). Until now, two types of ultrasonic sensors have been mainly developed; immersion and waveguide sensors. An immersion sensor can provide a high-resolution image, but it may have problems in terms of reliability and life time because the sensor is exposed to high temperature during inspection. On the other hand, a waveguide sensor can maintain its performance during long-term inspection in high temperature because it installs an ultrasonic transducer in a cold region even though such a high-frequency ultrasonic wave cannot be used owing to the long propagation distance [4-6]. In this work, a remote inspection module employing four 10 m long waveguide sensors was newly developed and several performance tests were carried out in water to demonstrate the applicability of the developed remote inspection module to inspection of reactor internals in an SFR. In this work, a remote inspection module for inspection of reactor internals in an SFR was newly developed. The developed remote inspection module employs four 10 m long waveguide sensors for multiple inspection applications: a horizontal beam waveguide sensor for ranging inspection, two vertical beam waveguide sensors for viewing inspection and a 45 .deg. angle beam waveguide sensor for identification inspection. Several performance tests such as ranging, viewing and identification inspections were carried out for simulated nuclear fuel assembly specimens in water, and the applicability of the developed remote inspection module to inspection of reactor internals in an SFR was demonstrated

Development and Applicability Demonstration of a Remote Inspection Module for Inspection of Reactor Internals in an SFR

Energy Technology Data Exchange (ETDEWEB)

Kim, Hoewoong; Joo, Youngsang; Park, Changgyu; Kim, Jongbum [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Bae, Jinho [National Fusion Research Institute, Daejeon (Korea, Republic of)

2014-05-15

Since liquid sodium is optically opaque, the ultrasonic inspection technique has been mainly employed for inspection of reactor internals in a Sodium-cooled Fast Reactor (SFR). Until now, two types of ultrasonic sensors have been mainly developed; immersion and waveguide sensors. An immersion sensor can provide a high-resolution image, but it may have problems in terms of reliability and life time because the sensor is exposed to high temperature during inspection. On the other hand, a waveguide sensor can maintain its performance during long-term inspection in high temperature because it installs an ultrasonic transducer in a cold region even though such a high-frequency ultrasonic wave cannot be used owing to the long propagation distance [4-6]. In this work, a remote inspection module employing four 10 m long waveguide sensors was newly developed and several performance tests were carried out in water to demonstrate the applicability of the developed remote inspection module to inspection of reactor internals in an SFR. In this work, a remote inspection module for inspection of reactor internals in an SFR was newly developed. The developed remote inspection module employs four 10 m long waveguide sensors for multiple inspection applications: a horizontal beam waveguide sensor for ranging inspection, two vertical beam waveguide sensors for viewing inspection and a 45 .deg. angle beam waveguide sensor for identification inspection. Several performance tests such as ranging, viewing and identification inspections were carried out for simulated nuclear fuel assembly specimens in water, and the applicability of the developed remote inspection module to inspection of reactor internals in an SFR was demonstrated.

Design of a new research reactor : 1st year conceptual design

International Nuclear Information System (INIS)

Park, Cheol; Lee, B. C.; Chae, H. T.

2004-01-01

A new research reactor model satisfying the strengthened regulatory environments and the changed circumstances around nuclear society should be prepared for the domestic and international demand of research reactor. This can also lead to the improvement of technologies and fostering manpower obtained during the construction and the operation of HANARO. In this aspect, this study has been launched and the 1st year conceptual design has been carried out in 2003. The major tasks performed at the first year of conceptual design stage are as follows; Establishments of general design requirements of research reactors and experimental facilities, Establishment of fuel and reactor core concepts, Preliminary analysis of reactor physics and thermal-hydraulics for conceptual core, Conceptual design of reactor structure and major systems, International cooperation to establish foundations for exporting

Effect of hydroelastic coupling on the response of a nuclear reactor to ground acceleration

International Nuclear Information System (INIS)

Au-Yang, M.K.; Skinner, D.A.

1977-01-01

The dynamical characteristics of a nuclear reactor vessel and its internal components is affected by the coolant inside the vessel. Recent studies in flow-induced vibration of reactor internal components show that the effect of the entrapped coolant can be properly accounted for by adding a 'hydrodynamic mass' matrix to the physical mass of the fluid structure system. In the past few years, analytical expressions for this hydrodynamic mass matrix have been derived, usually under greatly simplifying assumptions on the geometry of the structure. Typical examples are slender-cylinder and simply-supported-cylinder assumptions. While expressions derived based on these assumptions can still bring out the general characteristics of hydroelastic coupling of structure, their application to seismic analysis of reactor components is limited because these structutres, even though generally cylindrical, are usually neither slender nor simply supported. This paper presents an anlytical and experimental study of the effects of hydroelastic coupling on the seismic response of a reactor vessel and its internal components. The hydrodynamic mass matrix for cylindrical shell structures with arbitrary D/l ratios. Two specific examples are included to illustrate the effect of hydroelastic coupling on the response of a PWR to ground acceleration. (Auth.)

International trend on development of an innovative nuclear reactor and its meanings

Energy Technology Data Exchange (ETDEWEB)

Matsui, Kazuaki [Institute of Applied Energy, Tokyo (Japan)

2002-01-01

On outlining on flow of so-called innovative or new type nuclear reactor, at first, an improvement line of large-scale WHR, contains ABWR-2, APWR and its successive APWR+ in Japan, APR in Korea, and EPR in Europe, all of which have super large-scale output of 1.5MKW to use their scale merits in maximum. And, the second type is fast reactor only in Russia and Japan which are under reviewing its actual using plan of its already established development route. Furthermore, nuclear industry in the world is allowable to say a has-been industry, even its R and D system is decrepit, its researchers are much aged, and even utilization and foreign development of nuclear energy as a protecting measure of global warming are pronounced its self-control at the Bonn Conference in last year. However, the Generation 4 International Forum led by U.S.A. since early of 2000 and the Innovative Reactor Development Program (INPRO) through the International Atomic Energy Association (IAEA) due to initiative of Russia are planned to cooperatively promote their programs. In order to obtain any priority on small-scale production considerable technical jump is required or R and D and technical development elements with technical gap is necessary, which must be proved establishment of a target to overcome their scale demerit. (G.K.)

Specificities of micro-structured reactors for hydrogen production and purification

Energy Technology Data Exchange (ETDEWEB)

Dupont, N.; Germani, G.; Van Veen, A.C.; Schuurman, Y.; Mirodatos, C. [Institut de Recherches sur la Catalyse - CNRS, 2, Avenue Albert Einstein, 69626 Villeurbanne Cedex (France); Schaefer, G. [Atotech Deutschland GmbH, PO Box 210780, 10507 Berlin (Germany)

2007-07-15

This paper presents the specificities of micro-structured reactors as compared to conventional fixed-bed reactors through two case studies devoted to (i) hydrogen production by methanol steam reforming, (ii) hydrogen purification by water-gas shift (WGS). Key features like catalyst coating stability, temperature and pressure management, effects of operating conditions (residence time, pressure drops, etc.) are well identified as controlling the micro-reactor performances for methanol reforming. These devices are also shown to be excellent tools for fast access to reaction kinetics as exemplified for the WGS reaction, subject to operating conditions carefully chosen to ensure proper hydrodynamics, in order to use conventional plug flow reactor models for extracting rate constants. (author)

Growth and Expansion of the International Criticality Safety Benchmark Evaluation Project and the Newly Organized International Reactor Physics Experiment Evaluation Project

Energy Technology Data Exchange (ETDEWEB)

J. Blair Briggs; Lori Scott; Yolanda Rugama; Enrico Satori

2007-05-01

Since ICNC 2003, the International Criticality Safety Benchmark Evaluation Project (ICSBEP) has continued to expand its efforts and broaden its scope. Criticality-alarm / shielding type benchmarks and fundamental physics measurements that are relevant to criticality safety applications are not only included in the scope of the project, but benchmark data are also included in the latest version of the handbook. A considerable number of improvements have been made to the searchable database, DICE and the criticality-alarm / shielding benchmarks and fundamental physics measurements have been included in the database. There were 12 countries participating on the ICSBEP in 2003. That number has increased to 18 with recent contributions of data and/or resources from Brazil, Czech Republic, Poland, India, Canada, and China. South Africa, Germany, Argentina, and Australia have been invited to participate. Since ICNC 2003, the contents of the “International Handbook of Evaluated Criticality Safety Benchmark Experiments” have increased from 350 evaluations (28,000 pages) containing benchmark specifications for 3070 critical or subcritical configurations to 442 evaluations (over 38,000 pages) containing benchmark specifications for 3957 critical or subcritical configurations, 23 criticality-alarm-placement / shielding configurations with multiple dose points for each, and 20 configurations that have been categorized as fundamental physics measurements that are relevant to criticality safety applications in the 2006 Edition of the ICSBEP Handbook. Approximately 30 new evaluations and 250 additional configurations are expected to be added to the 2007 Edition of the Handbook. Since ICNC 2003, a reactor physics counterpart to the ICSBEP, The International Reactor Physics Experiment Evaluation Project (IRPhEP) was initiated. Beginning in 1999, the IRPhEP was conducted as a pilot activity by the by the Organization of Economic Cooperation and Development (OECD) Nuclear Energy