Conceptual design study of small long-life PWR based on thorium cycle fuel

International Nuclear Information System (INIS)

Subkhi, M. Nurul; Su'ud, Zaki; Waris, Abdul; Permana, Sidik

2014-01-01

A neutronic performance of small long-life Pressurized Water Reactor (PWR) using thorium cycle based fuel has been investigated. Thorium cycle which has higher conversion ratio in thermal region compared to uranium cycle produce some significant of 233 U during burn up time. The cell-burn up calculations were performed by PIJ SRAC code using nuclear data library based on JENDL 3.3, while the multi-energy-group diffusion calculations were optimized in whole core cylindrical two-dimension R-Z geometry by SRAC-CITATION. this study would be introduced thorium nitride fuel system which ZIRLO is the cladding material. The optimization of 350 MWt small long life PWR result small excess reactivity and reduced power peaking during its operation

Development of small and medium reactors for power and heat production

International Nuclear Information System (INIS)

Becka, J.

1978-01-01

Data are given on the current state of development of small and medium-power reactors designed mainly for electric power production in small power grids, for heat production for small- and medium-power desalination plants with possible electric power generation, for process steam production and heat development for district heating systems, again combined with electric power generation, and for propelling big and fast passenger ships. A diagram is shown of the primary system of an integrated PWR derived from the Otto Hahn reactor. The family is listed of the standard sizes of the integral INTERATOM company pressurized water reactors. Also listed are the specifications and design of CAS 2CG and AS 3G type reactors used mainly for long-distance heating systems. (J.B.)

Directives and general design requirements for a small PWR

International Nuclear Information System (INIS)

Arrieta, L.A.

1992-08-01

A proposal of directives and general requirements for the development of a small PWR conceptual design is presented. These directives address the main safety, performance and economic design aspects. The purpose is to use this work as a base for a wide discussion, involving all project participants, culminating with the definition of the final directives and general requirements. (author)

Utilization of thorium in PWR type reactors

International Nuclear Information System (INIS)

Correa, F.

1977-01-01

Uranium 235 consumption is comparatively evaluated with thorium cycle for a PWR type reactor. Modifications are only made in fuels components. U-235 consumption is pratically unchanged in both cycles. Some good results are promised to the mixed U-238/Th-232 fuel cycle in 1/1 proportion [pt

Deboration in nuclear stations of the PWR type

International Nuclear Information System (INIS)

1978-01-01

Reactivity control in nuclear power stations of the PWR type is realised with boric acid. A method to concentrate boric acid without an evaporator has been studied. A flow-sheet with reverse osmosis is proposed. (author)

Serious accidents of PWR type reactors for power generation

International Nuclear Information System (INIS)

2008-12-01

This document presents the great lines of current knowledge on serious accidents relative to PWR type reactors. First, is exposed the physics of PWR type reactor core meltdown and the possible failure modes of the containment building in such a case. Then, are presented the dispositions implemented with regards to such accidents in France, particularly the pragmatic approach that prevails for the already built reactors. Then, the document tackles the case of the European pressurized reactor (E.P.R.), for which the dimensioning takes into account explicitly serious accidents: it is a question of objectives conception and their respect must be the object of a strict demonstration, by taking into account uncertainties. (N.C.)

Ductile crack growth resistance of PWR components. Application for structural integrity assessment

International Nuclear Information System (INIS)

Bethmont, M.; Eripret, C.; Le Delliou, P.; Frund, J.M.

1995-01-01

Structural integrity assessment of PWR components, as pressure vessel and piping, needs to evaluate the ductile crack growth resistance which is generally characterized by J resistance curves (or J-R curves) based on the path-independent J Integral. These curves are more often obtained from laboratory tests with small specimens as CT-specimens and their application to large component safety analysis could be questionable Indeed, it is well known that J-R curves could depend on the specimen size and on the loading mode (i.e. bending stress versus tensile stress) but this dependency could be different from one material to another. This means that it would depend not only on the stress-strain state but also on the actual local fracture mechanisms (i. e. the damage) occurring before the crack initiation or during the crack propagation. The purpose of this paper is to gather some results of crack growth resistance measurement studied at EDF with different materials in order to show how the effect of the parameters, as specimen geometry and mode of loading, is directly related to the local fracture mechanisms or the microstructure of the materials. For that a number of results are analysed by means of the local approach of fracture which is a very useful tool to predict quantitatively the J-R curve dependency, related to fracture mechanisms (authors). 12 refs., 9 figs

Development of technologies for nuclear reactors of small and medium sized

International Nuclear Information System (INIS)

2011-08-01

This meeting include: countries presentations, themes and objectives of the training course, reactor types, design, EPR, APR1400, A P 1000, A PWR, ATMEA 1, VVER-1000, A PWR, ATMEA 1, VVER 1200, Boiling Water Reactor, A BWR, A BWR -II, ESBUR, Ke ren, AREVA, Heavy Water Reactor, Candu 6, Acr-1000, HWR, Bw, Iris, CAREM NuCcale, Smart, KLT-HOS, Westinghouse small modular Reactor, Gas Cooled Reactors, PBMR, React ores enfriados con metales liquidos, Hs, Prism,Terra Power, Hyper ion, appliance's no electric as de energia, Generation IV Reactors,VHTR, Gas Fast Reactor, Sodium Fast Reactor, Molten salt Reactor, Lfr, Water Cooled Reactor, Technology Assessment Process, Fukushima accident.

Development and application of integrated digital I and C system in Japanese PWR plants

International Nuclear Information System (INIS)

Tominaga, M.

1995-01-01

The Integrated Digital Instrumentation and Control (I and C) System has been developed and applied to non-safety grade I and C systems in the latest 5 Japanese PWR plants in 1990's. Based on the experience in these plants, the Integrated Digital I and C System will be planned to apply also to safety grade I and C systems in Advanced PWR (APWR) as the overall application of digital technology. The basic design task has been just started for APWR which is to be in commercial operation in early 2000's and under the development about various issues of safety grade digital I and C systems. On the other hand, in conventional Japanese PWR plants, digital I and C systems have been applied step by step since 1980's. For example, digital I and C systems for radio-active waste processing system have been adopted to 13 units, and dedicated digital I and C systems for Local loop control system to 8 units. The trend and status of development and application of the digital I and C systems, especially the Integrated Digital I and C System in Japanese PWR plants are presented. (5 refs., 4 figs.)

Studies of a small PWR for onsite industrial power

International Nuclear Information System (INIS)

Klepper, O.H.; Smith, W.R.

1977-01-01

Information on the use of a 300 to 400 MW(t) PWR type reactor for industrial applications is presented concerning the potential market, reliability considerations, reactor plant description, construction techniques, comparison between nuclear and fossil-fired process steam costs, alternative fossil-fired steam supplies, and industrial application

Experiment and analyses on intentional secondary-side depressurization during PWR small break LOCA. Effects of depressurization rate and break area on core liquid level behavior

International Nuclear Information System (INIS)

Asaka, Hideaki; Ohtsu, Iwao; Anoda, Yoshinari; Kukita, Yutaka

1997-01-01

The effects of the secondary-side depressurization rate and break area on the core liquid level behavior during a PWR small-break LOCA were studied using experimental data from the Large Scale Test Facility (LSTF) and by using analysis results obtained with a JAERI modified version of RELAP5/MOD3 code. The LSTF is a 1/ 48 volumetrically scaled full-height integral model of a Westinghouse-type PWR. The code reproduced the thermal-hydraulic responses, observed in the experiment, for important parameters such as the primary and secondary side pressures and core liquid level behavior. The sensitivity of the core minimum liquid level to the depressurization rate and break area was studied by using the code assessed above. It was found that the core liquid level took a local minimum value for a given break area as a function of secondary side depressurization rate. Further efforts are, however, needed to quantitatively define the maximum core temperature as a function of break area and depressurization rate. (author)

A numerical integration approach suitable for simulating PWR dynamics using a microcomputer system

International Nuclear Information System (INIS)

Zhiwei, L.; Kerlin, T.W.

1983-01-01

It is attractive to use microcomputer systems to simulate nuclear power plant dynamics for the purpose of teaching and/or control system design. An analysis and a comparison of feasibility of existing numerical integration methods have been made. The criteria for choosing the integration step using various numerical integration methods including the matrix exponential method are derived. In order to speed up the simulation, an approach is presented using the Newton recursion calculus which can avoid convergence limitations in choosing the integration step size. The accuracy consideration will dominate the integration step limited. The advantages of this method have been demonstrated through a case study using CBM model 8032 microcomputer to simulate a reduced order linear PWR model under various perturbations. It has been proven theoretically and practically that the Runge-Kutta method and Adams-Moulton method are not feasible. The matrix exponential method is good at accuracy and fairly good at speed. The Newton recursion method can save 3/4 to 4/5 time compared to the matrix exponential method with reasonable accuracy. Vertical Barhis method can be expanded to deal with nonlinear nuclear power plant models and higher order models as well

Economic targets for small PWR reactor designs

International Nuclear Information System (INIS)

Board, J.

1991-01-01

Small reactors are likely to be less economic than large reactors, but the lower financial exposure with small reactors may be attractive to utilities contemplating a restart to a nuclear programme. New nuclear plant can be economic, but success will depend more on how the plant are built, rather than what type or size is built. A target for new plant for operation early in the next century should be a generation cost of 3p to 3.5 p/kWh. This corresponds to an overnight capital cost of Pound 1000/kWh to Pound 1100/kWh. (author)

Optimal unit sizing for small-scale integrated energy systems using multi-objective interval optimization and evidential reasoning approach

International Nuclear Information System (INIS)

Wei, F.; Wu, Q.H.; Jing, Z.X.; Chen, J.J.; Zhou, X.X.

2016-01-01

This paper proposes a comprehensive framework including a multi-objective interval optimization model and evidential reasoning (ER) approach to solve the unit sizing problem of small-scale integrated energy systems, with uncertain wind and solar energies integrated. In the multi-objective interval optimization model, interval variables are introduced to tackle the uncertainties of the optimization problem. Aiming at simultaneously considering the cost and risk of a business investment, the average and deviation of life cycle cost (LCC) of the integrated energy system are formulated. In order to solve the problem, a novel multi-objective optimization algorithm, MGSOACC (multi-objective group search optimizer with adaptive covariance matrix and chaotic search), is developed, employing adaptive covariance matrix to make the search strategy adaptive and applying chaotic search to maintain the diversity of group. Furthermore, ER approach is applied to deal with multiple interests of an investor at the business decision making stage and to determine the final unit sizing solution from the Pareto-optimal solutions. This paper reports on the simulation results obtained using a small-scale direct district heating system (DH) and a small-scale district heating and cooling system (DHC) optimized by the proposed framework. The results demonstrate the superiority of the multi-objective interval optimization model and ER approach in tackling the unit sizing problem of integrated energy systems considering the integration of uncertian wind and solar energies. - Highlights: • Cost and risk of investment in small-scale integrated energy systems are considered. • A multi-objective interval optimization model is presented. • A novel multi-objective optimization algorithm (MGSOACC) is proposed. • The evidential reasoning (ER) approach is used to obtain the final optimal solution. • The MGSOACC and ER can tackle the unit sizing problem efficiently.

Design concepts and safety concerns of the small and medium size reactors (SMR)

International Nuclear Information System (INIS)

Seul, Kwang Won; Lee, Jae Hun; Kim, Hho Jung

1998-01-01

The small and medium size reactors (SMR) and interface facilities such as desalination plant are expected to be located near the population area because of restrictions in transporting the plant products such as fresh water to long distance area. To protect the public around the plant facility from the possible release of radioactive materials, the design development of the SMR is focusing on an enhancement of the safety and reliability as well as the economics. In this study, the major safety concepts of the SMR designs significantly different from the current PWR designs are investigated and the safety concerns applicable to the integrated SMR design of Korea (called SMART), were identified. Those safety issues include the use of proven technology, application of strengthening defense in depth, event categorization and selection, simplification of emergency planning, determination of accident source terms and so on. The efforts to resolve the safety concerns in the design stage will provide an improvement of the safety of the SMART design

CAREM: an innovative-integrated PWR

International Nuclear Information System (INIS)

Mazzi, R.

2005-01-01

A promising future in view of the increasing worldwide acknowledgment of the Nuclear Power as a bulk-environmentally friendly energy source is envisaged; nevertheless the widespread concerns about nuclear safety means the uppermost challenge to the nuclear designers to achieve massive public acceptance of NPP. CAREM is an Argentine project aimed to achieve the development, design and construction of an innovative, small and integrated Nuclear Power Plant (NPP). The reactor has an indirect cycle with some distinctive features that greatly simplify the design, and also contributes to a higher safety level. Some of the design highlight are: integrated primary cooling system, primary cooling by natural circulation, self-pressurised primary system and safety systems relying on passive features. The innovative solutions are embraced in the 'CAREM Concept' aimed to enhance safety by using simpler and more reliable solutions to tackle major safety design challenges of the nuclear generation industry. The goal is achieved by drastically reducing the conceivable list of initiating events jointly with a large primary water/power ratio that results in spontaneous slow and mild transients even after most severe system or component failure. In addition all Safety Systems are also based on simple and reliable solutions that increase sharply the overall plant reliability at reduced costs. The concept has been engineering developed for the CAREM 25 (prototype, 100 MWth, 27 MWe) considered an appropriate size to display the performance related with the reactor core cooling and safety systems. This module while not cost effective if compared with major sized NPP's installation and operating cost, results appropriate for applications such as supplying domestic or industrial electricity and/or steam (i.e. for a water demineralising plant) at isolated or difficult to access, mid size, populations. A promising market is envisaged for the evolution of the CAREM Concept towards higher

Does company size matter? Validation of an integrative model of safety behavior across small and large construction companies.

Science.gov (United States)

Guo, Brian H W; Yiu, Tak Wing; González, Vicente A

2018-02-01

Previous safety climate studies primarily focused on either large construction companies or the construction industry as a whole, while little is known about whether company size has significant effects on workers' understanding of safety climate measures and relationships between safety climate factors and safety behavior. Thus, this study aims to: (a) test the measurement equivalence (ME) of a safety climate measure across workers from small and large companies; (b) investigate if company size alters the causal structure of the integrative model developed by Guo, Yiu, and González (2016). Data were collected from 253 construction workers in New Zealand using a safety climate measure. This study used multi-group confirmatory factor analyses (MCFA) to test the measurement equivalence of the safety climate measure and structure invariance of the integrative model. Results indicate that workers from small and large companies understood the safety climate measure in a similar manner. In addition, it was suggested that company size does not change the causal structure and mediational processes of the integrative model. Both measurement equivalence of the safety climate measure and structural invariance of the integrative model were supported by this study. Practical applications: Findings of this study provided strong support for a meaningful use of the safety climate measure across construction companies in different sizes. Safety behavior promotion strategies designed based on the integrative model may be well suited for both large and small companies. Copyright © 2017 National Safety Council and Elsevier Ltd. All rights reserved.

Control in fabrication of PWR and BWR type reactor fuel elements

International Nuclear Information System (INIS)

Gorskij, V.V.

1981-01-01

Both destructive and non-destructive testing methods now in use in fabrication of BWR and PWR type reactor fuel elements at foreign plants are reviewed. Technological procedures applied in fabrication of fuel elements and fuel assemblies are described. Major attention is paid to radiographic, ultrasonic, metallographic, visual and autoclavic testings. A correspondence of the methods applied to the ASTM standards is discussed. The most part of the countries are concluded the apply similar testing methods enabling one to reliably evaluate the quality of primary materials and fabricated fuel elements and thus meeting the demands to contemporary PWR and BWR type reactor fuel elements. Practically all fuel element and pipe fabrication plants in Western Europe, Asia and America use the ASTM standards as the basis for the quality contr [ru

PWR cold-leg small break loca with faulty HPI

International Nuclear Information System (INIS)

Kumamaru, H.; Kukita, Y.

1991-01-01

The ROSA-IV Large Scale Test Facility (LSTF) is a 1/48 volumetrically-scaled model of a pressurized water reactor (PWR). At the LSTF are performed cold-leg small-break loss-of-coolant accident (LOCA) tests with faulty high pressure injection (HPI) system for break areas from 0.5% to 10% and an intentional primary system depressurization test following a small-break LOCA test. A simple prediction model is proposed for prediction of times of major events. Test data and calculations show that intentional primary system depressurization with use of the pressurizer power-operated relief valves (PORVs) is effective for break areas of approximately 0.5% or less, is unnecessary for breaks of 5% or more, and is insufficient for intermediate break areas to maintain adequate core cooling. (author)

Effects of Burnable Absorbers on PWR Spent Nuclear Fuel

International Nuclear Information System (INIS)

O'Leary, P.M.; Pitts, M.L.

2000-01-01

Burnup credit is an ongoing issue in designing and licensing transportation and storage casks for spent nuclear fuel (SNF). To address this issue, in July 1999, the U.S. Nuclear Regulatory Commission (NRC), Spent Fuel Project Office, issued Interim Staff Guidance-8 (ISG-8), Revision 1 allowing limited burnup credit for pressurized water reactor (PWR) spent nuclear fuel (SNF) to be used in transport and storage casks. However, one of the key limitations for a licensing basis analysis as stipulated in ISG-8, Revision 1 is that ''burnup credit is restricted to intact fuel assemblies that have not used burnable absorbers''. Because many PWR fuel designs have incorporated burnable-absorber rods for more than twenty years, this restriction places an unnecessary burden on the commercial nuclear power industry. This paper summarizes the effects of in-reactor irradiation on the isotopic inventory of PWR fuels containing different types of integral burnable absorbers (BAs). The work presented is illustrative and intended to represent typical magnitudes of the reactivity effects from depleting PWR fuel with different types of burnable absorbers

Validating Westinghouse atom 16 x 16 and 18 x 18 PWR fuel performance

International Nuclear Information System (INIS)

Andersson, S.; Gustafson, J.; Jourdain, P.; Lindstroem, L.; Hallstadius, L.; Hofling, C.G.

2001-01-01

Westinghouse Atom designs and fabricates PWR fuel for all major European fuel types: 17 x 17 standard (12 ft) and 17 x 17 XL (14 ft) for Westinghouse type PWRs, and 16 x 16 and 18 x 18 fuel for Siemens type PWRs. The W Atom PWR fuel designs are based on the extensive Westinghouse CE PWR fuel experience from combustion engineering type PWRs. The W atom designs utilise basic design features from the W CE fuel tradition, such as all-Zircaloy mid grids and the proven ( 6 rod years) Guardian TM debris catcher, which is integrated in the bottom Inconel grid. Several new features have been developed to meet with stringent European requirements originating from requirements on very high burnup, in combination with low-leakage core operating strategies and high coolant temperatures. The overall reliability of the Westinghouse Atom PWR fuel is very high; no fuel failure has been detected since 1997. (orig.)

Problems of control of WWER-type pressurized water reactors (PWR's)

International Nuclear Information System (INIS)

Drab, F.; Grof, V.

1978-01-01

The problems are dealt with of nuclear power reactor control. Special attention is paid to the reactor of the WWER type, which will play the most important part in the Czechoslovak power system in the near future. The subsystems are described which comprise the systems of reactor control and protection. The possibilities are outlined of using Czechoslovak instrumentation for the control and safety system of the WWER-type PWR. (author)

PACTEL and PWR PACTEL Test Facilities for Versatile LWR Applications

Directory of Open Access Journals (Sweden)

Virpi Kouhia

2012-01-01

Full Text Available This paper describes construction and experimental research activities with two test facilities, PACTEL and PWR PACTEL. The PACTEL facility, comprising of reactor pressure vessel parts, three loops with horizontal steam generators, a pressurizer, and emergency core cooling systems, was designed to model the thermal-hydraulic behaviour of VVER-440-type reactors. The facility has been utilized in miscellaneous applications and experiments, for example, in the OECD International Standard Problem ISP-33. PACTEL has been upgraded and modified on a case-by-case basis. The latest facility configuration, the PWR PACTEL facility, was constructed for research activities associated with the EPR-type reactor. A significant design basis is to utilize certain parts of PACTEL, and at the same time, to focus on a proper construction of two new loops and vertical steam generators with an extensive instrumentation. The PWR PACTEL benchmark exercise was launched in 2010 with a small break loss-of-coolant accident test as the chosen transient. Both facilities, PACTEL and PWR PACTEL, are maintained fully operational side by side.

PACTEL and PWR PACTEL Test Facilities for Versatile LWR Applications

International Nuclear Information System (INIS)

Virpi Kouhia, V.; Purhonen, H.; Riikonen, V.; Puustinen, M.; Kyrki-Rajamaki, R.; Vihavainen, J.

2012-01-01

This paper describes construction and experimental research activities with two test facilities, PACTEL and PWR PACTEL. The PACTEL facility, comprising of reactor pressure vessel parts, three loops with horizontal steam generators, a pressurizer, and emergency core cooling systems, was designed to model the thermal-hydraulic behaviour of VVER-440-type reactors. The facility has been utilized in miscellaneous applications and experiments, for example, in the OECD International Standard Problem ISP-33. PACTEL has been upgraded and modified on a case-by-case basis. The latest facility configuration, the PWR PACTEL facility, was constructed for research activities associated with the EPR-type reactor. A significant design basis is to utilize certain parts of PACTEL, and at the same time, to focus on a proper construction of two new loops and vertical steam generators with an extensive instrumentation. The PWR PACTEL benchmark exercise was launched in 2010 with a small break loss-of-coolant accident test as the chosen transient. Both facilities, PACTEL and PWR PACTEL, are maintained fully operational side by side.

An integrated PWR for marine propulsion

International Nuclear Information System (INIS)

Letouze, A.; Marecaux, A.; Rollason, J.; Heap, S.; Foster, A.; Jewer, S.; Thompson, A. C.; Williams, A. M.; Beeley, P. A.

2008-01-01

Results from a design study for a nuclear propulsion plant utilising a small integrated PWR using many of the inherent safety features of the IRIS design. The design consists of a single pass, low enrichment core housed, together with all associated primary circuit components, within a reactor pressure vessel 10.3 m high and 4.1 m in diameter. Reactor physics calculations were conducted with the codes WIMS9a and MONK8b. The core design contains 21 fuel assemblies each containing 264 UO 2 fuel pins. Each fuel module has a cluster of 24 boron carbide control rods and a central instrumentation channel. The fuel enrichment was 9% in order to achieve the core lifetime requirement of 3000 EFPD at a reactor power of 120 MWth. This gives a discharge burnup of 51,000 MWd/t. To control excess reactivity, two forms of burnable poison are employed: a zirconium dibromide (ZrB 2 ) coating on the fuel compacts, and gadolinium oxide homogeneously mixed in the fuel. Thermal hydraulic calculations were performed using TRAC-P(ND) for steady-state operation and for a number of fault transients. The helical once through steam generators were modelled using heat structure and pipe components and their performance compared to independent calculations including heat transfer correlations for the helical coiled geometry. Intact circuit calculations for steady state were followed by a small break LOCA calculation including the effect of a containment volume which reproduced the gain of coolant effect reported for IRIS. It was demonstrated that the thermal limits were not exceeded for the identified key transients. The dynamic response of the reactor plant to typical power demands was modelled using AcslXtreme software. Several schemes for limiting the power overshoot that was found on rapid increase to full power were examined. It was concluded that the SG must be operated with variable secondary pressure and the best means of reducing power overshoot is to step back the throttle opening

Significance assessment of small-medium sized reactors

Energy Technology Data Exchange (ETDEWEB)

Kanno, Minoru [Japan Atomic Power Co., Research and Development Dept., Tokyo (Japan)

2002-12-01

Preliminary assessment for deployment of small-medium sized reactor (S and M reactor) as a future option has been conducted at the JAPCO (Japan Atomic Power Company) under the cooperation with the CRIERI (Central Research Institute of Electric Power Industry). Significance of the S and M reactor introduction is listed as follows; lower investment cost, possible siting near demand side, enlarged freedom of siting, shorter transmission line, good compatibility with slow increase of demand and plain explanation of safety using simpler system such as integral type vessel without piping, natural convection core cooling and passive safety system. The deployment of simpler plant system, modular shop fabrication, ship-shell structured building and longer operation period can assure economics comparable with that of a large sized reactor, coping with scale-demerit. Also the S and M reactor is preferable in size for the nuclear heat utilization such as hydrogen production. (T. Tanaka)

Study on external reactor vessel cooling capacity for advanced large size PWR

International Nuclear Information System (INIS)

Jin Di; Liu Xiaojing; Cheng Xu; Li Fei

2014-01-01

External reactor vessel cooling (ERVC) is widely adopted as a part of in- vessel retention (IVR) in severe accident management strategies. In this paper, some flow parameters and boundary conditions, eg., inlet and outlet area, water inlet temperature, heating power of the lower head, the annular gap size at the position of the lower head and flooding water level, were considered to qualitatively study the effect of them on natural circulation capacity of the external reactor vessel cooling for an advanced large size PWR by using RELAP5 code. And the calculation results provide some basis of analysis for the structure design and the following transient response behavior of the system. (authors)

Gadolinia experience and design for PWR fuel cycles

International Nuclear Information System (INIS)

Stephenson, L. C.

2000-01-01

The purpose of this paper is to describe Siemens Power Corporation's (SPC) current experience with the burnable absorber gadolinia in PWR fuel assemblies, including optimized features of SPC's PWR gadolinia designs, and comparisons with other burnable absorbers. Siemens is the world leader in PWR gadolinia experience. More than 5,900 Siemens PWR gadolinia-bearing fuel assemblies have been irradiated. The use of gadolinia-bearing fuel provides significant flexibility in fuel cycle designs, allows for low radial leakage fuel management and extended operating cycles, and reduces BOC (beginning-of-cycle) soluble boron concentrations. The optimized use of an integral burnable neutron absorber is a design feature which provides improved economic performance for PWR fuel assemblies. This paper includes a comparison between three different types of integral burnable absorbers: gadolinia, Zirconium diboride and erbia. Fuel cycle design studies performed by Siemens have shown that the enrichment requirements for 18-24 month fuel cycles utilizing gadolinia or zirconium diboride integral fuel burnable absorbers can be approximately the same. Although a typical gadolinia residual penalty for a cycle design of this length is as low as 0.02-0.03 wt% U-235, the design flexibility of gadolinia allows for very aggressive low-leakage core loading plans which reduces the enrichment requirements for gadolinia-bearing fuel. SPC has optimized its use of gadolinia in PWR fuel cycles. Typically, low (2-4) weight percent Gd 2 O 3 is used for beginning to middle of cycle reactivity hold down as well as soluble boron concentration holddown at BOC. Higher concentrations of Gd 2 O 3 , such as 6 and 8 wt%, are used to control power peaking in assemblies later in the cycle. SPC has developed core strategies that maximize the use of lower gadolinia concentrations which significantly reduces the gadolinia residual reactivity penalty. This optimization includes minimizing the number of rods with

The management of mechanical integrity inspections at small-sized 'Seveso' facilities

International Nuclear Information System (INIS)

Bragatto, Paolo A.; Pittiglio, Paolo; Ansaldi, Silvia

2009-01-01

The mechanical integrity (MI) of equipment has been controlled at all industrial facilities for many decades. Control methods and intervals are regulated by laws or codes and best practices. In European countries, the legislations implementing the Seveso Directives on the control of major accident hazards require the owner of establishments where hazardous chemicals are handled, to implement a safety management system (SMS). MI controls should be an integral part of the SMS. At large establishments this goal is achieved by adopting the RBI method, but in small-sized establishments with a limited budget and scanty personnel, a heuristic approach is more suitable. This paper demonstrates the feasibility and advantages of integrating SMS and MI by means of a simple method that includes a few basic concepts of RBI without additional costs for operator. This method, supported by a software tool, is resilient as it functions effectively in spite of eventual budget reductions and personnel turnover. The results of MI controls can also be exploited to monitor equipment condition and demonstrate the adequacy of technical systems to the Competent Authorities (CA). Furthermore, the SMS can 'capture' knowledge resulting from MI experience and exploit it for a better understanding of risk

Natural circulation in a scaled PWR integral test facility

International Nuclear Information System (INIS)

Kiang, R.L.; Jeuck, P.R. III

1987-01-01

Natural circulation is an important mechanism for cooling a nuclear power plant under abnormal operating conditions. To study natural circulation, we modeled a type of pressurized water reactor (PWR) that incorporates once-through steam generators. We conducted tests of single-phase natural circulations, two-phase natural circulations, and a boiler condenser mode. Because of complex geometry, the natural circulations observed in this facility exhibit some phenomena not commonly seen in a simple thermosyphon loop

Coolant degassing device for PWR type reactors

International Nuclear Information System (INIS)

Kita, Kaoru; Takezawa, Kazuaki; Minemoto, Masaki.

1982-01-01

Purpose: To efficiently decrease the rare gas concentration in primary coolants, as well as shorten the degassing time required for the periodical inspection in the waste gas processing system of a PWR type reactor. Constitution: Usual degassing method by supplying hydrogen or nitrogen to a volume control tank is replaced with a method of utilizing a degassing tower (method of flowing down processing liquid into the filled tower from above while uprising streams from the bottom of the tower thereby degassing the gases dissolved in the liquid into the steams). The degassing tower is combined with a hydrogen separator or hydrogen recombiner to constitute a waste gas processing system. (Ikeda, J.)

French nuclear plants PWR vessel integrity assessment and life management

Energy Technology Data Exchange (ETDEWEB)

Bezdikian, G. [Electricite de France (EDF), Div. Production Nucleaire, 93 - Saint-Denis (France); Quinot, P. [FRAMATOME, Dept. Bloc Reacteur et Boucles Primaires, 92 - Paris-La-Defence (France); Faidy, C.; Churier-Bossennec, H. [Electricite de France (EDF), Div. Ingenierie et Service, 69 - Villeurbanne (France)

2001-07-01

The Reactor Pressure Vessel life management of 56 PWR 3 loop and 4 loop reactors units was engaged by the French Utility EDF (Electricite de France) a few years ago and is yet on going on. This paper will present the work carried out within the framework of justifying why the 34 three loop reactor vessels will remain acceptable for operation for a lifetime of at least 40-years. A summary of the measures will be given. An overall review of actions will be presented describing the French approach, using important existing databases, including studies related to irradiation surveillance monitoring program and end of life fluence assessment. The last results obtained are based on generic integrity analyses for all categories of situations (normal upset emergency and faulted conditions) until the end of lifetime, postulating circumferential an radial kinds of flaw located in the stainless steel cladding or shallow sub-cladding area. The results of structural integrity analyses beginning with elastic computations and completed with three-dimensional finite element elastic plastic computations for envelope cases, are compared with code criteria for operating plants. The objective is to evaluate the margins on different parameters as RTNDT (Reference Nil Ductility Transition Temperature), toughness or crack size, to justify the global fitness for service of all these Reactor Pressure Vessels. The paper introduces EDF's maintenance strategy, related to integrity assessment, for those nuclear power plants under operation, based on NDE in-service inspection of the first thirty millimeters in the thickness of the wall and major surveillance programs of the vessels. (author)

French nuclear plants PWR vessel integrity assessment and life management

International Nuclear Information System (INIS)

Bezdikian, G.; Quinot, P.; Faidy, C.; Churier-Bossennec, H.

2001-01-01

The Reactor Pressure Vessel life management of 56 PWR 3 loop and 4 loop reactors units was engaged by the French Utility EDF (Electricite de France) a few years ago and is yet on going on. This paper will present the work carried out within the framework of justifying why the 34 three loop reactor vessels will remain acceptable for operation for a lifetime of at least 40-years. A summary of the measures will be given. An overall review of actions will be presented describing the French approach, using important existing databases, including studies related to irradiation surveillance monitoring program and end of life fluence assessment. The last results obtained are based on generic integrity analyses for all categories of situations (normal upset emergency and faulted conditions) until the end of lifetime, postulating circumferential an radial kinds of flaw located in the stainless steel cladding or shallow sub-cladding area. The results of structural integrity analyses beginning with elastic computations and completed with three-dimensional finite element elastic plastic computations for envelope cases, are compared with code criteria for operating plants. The objective is to evaluate the margins on different parameters as RTNDT (Reference Nil Ductility Transition Temperature), toughness or crack size, to justify the global fitness for service of all these Reactor Pressure Vessels. The paper introduces EDF's maintenance strategy, related to integrity assessment, for those nuclear power plants under operation, based on NDE in-service inspection of the first thirty millimeters in the thickness of the wall and major surveillance programs of the vessels. (author)

Status of development - An integral type small reactor MRX in JAERI

International Nuclear Information System (INIS)

Hoschi, T.; Ochiai, M.; Shimazaki, J.

1998-01-01

JAERI is conducting a design study on an integral type small reactor MRX for the use of nuclear ships. The basic concept of the reactor system is the integral type reactor with in-vessel steam generators and control rod drive systems, however, such new technologies as the water-filled containment, the passive decay heat removal system, the advanced automatic system, etc., are adopted to satisfy the essential requirements for the next generation ship reactors, i.e. compact, light, highly safe and easy operation. Research and development (R and D) works have being progressed on the peculiar components, the advanced automatic operation systems and the safety systems. Feasibility study and the economical evaluation of nuclear merchant ships have also being performed. The experiments and analysis of the safety carried out so far are proving that the passive safety features applied into the MRX are sufficient functions in the safety point of view. The MRX is a typical small type reactor realizing the easy operation by simplifying the reactor systems adopting the passive safety systems, therefore, it has wide variety of use as energy supply systems. This paper summarizes the present status on the design study of the MRX and the research and development activities as well as the some results of feasibility study. (author)

Pre design processing of waste of ex-resin without materials matrix from nuclear power plant type PWR 1000 MW

International Nuclear Information System (INIS)

Cerdas Tarigan

2010-01-01

Have been done pre design processing of waste ex-resin without capacities matrix materials from nuclear power plant type PWR 1000 MW During the time radioactive waste of ex-resin processed to use process of immobilization use matrix materials like mixture cement and epoxy resin and then conditioning. This process is not effective and efficient because end result volume of end product bigger than volume early operation system and maintenance of its installation more difficult. To overcome this created a design of technology processing of waste of ex- resin without matrix materials through process of strainer, drying and conditioning represent technological innovation newly processing of radioactive waste of ex-resin. Besides this process more effective and efficient, volume of end product waste much more small from volume early and operation system and maintenance of its easier installation. Pre design is expected to be used as a basis to make conceptual of pre design installation of strainer, drying and conditioning for the processing of waste of ex-resin from nuclear power plant type PWR 1000 MW. (author)

New generation nuclear power units of PWR type integral reactors

International Nuclear Information System (INIS)

Mitenkov, F.M.; Kurachen Kov, A.V.; Malamud, V.A.; Panov, Yu.K.; Runov, B.I.; Flerov, L.N.

1997-01-01

Design bases of new generation nuclear power units (nuclear power plants - NPP, nuclear co-generation plants - NCP, nuclear distract heating plants - NDHP), using integral type PWPS, developed in OKBM, Nizhny Novgorod and trends of design decisions optimization are considered in this report. The problems of diagnostics, servicing and repair of the integral reactor components in course of operation are discussed. The results of safety analysis, including the problems of several accident localization with postulated core melting and keeping corium in the reactor vessel and guard vessel are presented. Information on experimental substantiation of the suggested plant design decisions is presented. (author)

Safety Analysis for Medium/Small Size Integral Reactor: Evaluation of Safety Characteristics for Small and Medium Integral Reactor

Energy Technology Data Exchange (ETDEWEB)

Kim, Hho jung; Seul, K W; Ahn, S K; Bang, Y S; Park, D G; Kim, B K; Kim, W S; Lee, J H; Kim, W K; Shim, T M; Choi, H S; Ahn, H J; Jung, D W; Kim, G I; Park, Y M; Lee, Y J [Korea Inst. of Nuclear Safety, Taejon (Korea, Republic of)

1997-07-01

The Small and medium integral reactor is developed to be utilized for non-electric areas such as district heating and steam production for desalination and other industrial purposes, and then these applications may typically imply a closeness between the reactor and the user. It requires the reactor to be designed with the adoption of special functional and inherent safety features to ensure and promote a high level of safety and reliability, in comparison with the existing nuclear power plants. The objective of the present study is to establish the bases for the development of regulatory requirements and technical guides to address the special safety characteristics of the small and medium integral reactor. In addition, the study aims to identify and to propose resolutions to the possible safety concerns in the design of the small and medium integral reactor. 34 refs., 20 tabs. (author)

The development of emergency core cooling systems in the PWR, BWR, and HWR Candu type of nuclear power plants

International Nuclear Information System (INIS)

Mursid Djokolelono.

1976-01-01

Emergency core cooling systems in the PWR, BWR, and HWR-Candu type of nuclear power plant are reviewed. In PWR and BWR the emergency cooling can be catagorized as active high pressure, active low pressure, and a passive one. The PWR uses components of the shutdown cooling system: whereas the BWR uses components of pressure suppression contaiment. HWR Candu also uses the shutdown cooling system similar to the PWR except some details coming out from moderator coolant separation and expensive cost of heavy water. (author)

Analysis of a small PWR core with the PARCS/Helios and PARCS/Serpent code systems

International Nuclear Information System (INIS)

Baiocco, G.; Petruzzi, A.; Bznuni, S.; Kozlowski, T.

2017-01-01

Highlights: • The consistency between Helios and Serpent few-group cross sections is shown. • The PARCS model is validated against a Monte Carlo 3D model. • The fission and capture rates are compared. • The influence of the spacer grids on the axial power distribution is shown. - Abstract: Lattice physics codes are primarily used to generate cross-section data for nodal codes. In this work the methodology of homogenized constant generation was applied to a small Pressurized Water Reactor (PWR) core, using the deterministic code Helios and the Monte Carlo code Serpent. Subsequently, a 3D analysis of the PWR core was performed with the nodal diffusion code PARCS using the two-group cross section data sets generated by Helios and Serpent. Moreover, a full 3D model of the PWR core was developed using Serpent in order to obtain a reference solution. Several parameters, such as k eff , axial and radial power, fission and capture rates were compared and found to be in good agreement.

Analysis of radiation safety for Small Modular Reactor (SMR) on PWR-100 MWe type

Science.gov (United States)

Udiyani, P. M.; Husnayani, I.; Deswandri; Sunaryo, G. R.

2018-02-01

Indonesia as an archipelago country, including big, medium and small islands is suitable to construction of Small Medium/Modular reactors. Preliminary technology assessment on various SMR has been started, indeed the SMR is grouped into Light Water Reactor, Gas Cooled Reactor, and Solid Cooled Reactor and from its site it is group into Land Based reactor and Water Based Reactor. Fukushima accident made people doubt about the safety of Nuclear Power Plant (NPP), which impact on the public perception of the safety of nuclear power plants. The paper will describe the assessment of safety and radiation consequences on site for normal operation and Design Basis Accident postulation of SMR based on PWR-100 MWe in Bangka Island. Consequences of radiation for normal operation simulated for 3 units SMR. The source term was generated from an inventory by using ORIGEN-2 software and the consequence of routine calculated by PC-Cream and accident by PC Cosyma. The adopted methodology used was based on site-specific meteorological and spatial data. According to calculation by PC-CREAM 08 computer code, the highest individual dose in site area for adults is 5.34E-02 mSv/y in ESE direction within 1 km distance from stack. The result of calculation is that doses on public for normal operation below 1mSv/y. The calculation result from PC Cosyma, the highest individual dose is 1.92.E+00 mSv in ESE direction within 1km distance from stack. The total collective dose (all pathway) is 3.39E-01 manSv, with dominant supporting from cloud pathway. Results show that there are no evacuation countermeasure will be taken based on the regulation of emergency.

Small Lesion Size Is Associated with Sleep-Related Epilepsy in Focal Cortical Dysplasia Type II

Directory of Open Access Journals (Sweden)

Bo Jin

2018-02-01

Full Text Available ObjectiveTo investigate the neuroimaging and clinical features associated with sleep-related epilepsy (SRE in patients with focal cortical dysplasia (FCD type II.MethodsPatients with histopathologically proven FCD type II were included from three epilepsy centers. SRE was defined according to the video EEG findings and seizure history. Cortical surface reconstruction and volume calculation were performed using FreeSurfer. The lesions were manually delineated on T1 volumetric MRI using the ITK-SNAP software. The lesion volumes were normalized by the intracranial volume of each patient. The lesions were classified as small or large by placing a threshold based on quantitative (whether the lesion was detected on MRI report and qualitative (volume criteria.ResultsA total of 77 consecutive patients were included. Of them, 36 had SRE and 41 had non-SRE. An earlier age of epilepsy onset, high seizure frequency, regional interictal EEG findings, and favorable surgical outcome were characteristic in both groups. Small lesions were defined as those having a volume <3,217 mm3. In total, 60.9% of the patients with SRE (25/41 had small FCD lesion, which was significantly higher than the non-SRE group (9/34, 26.5%, p = 0.005. Small lesion size was the only predictor significantly associated with SRE in the overall type II group by multivariate analyses (p = 0.016. Although the proportion of patients who had frontal FCD and SRE was higher than non-frontal FCD (54.5 vs. 27.3%, p = 0.043, the relationship between SRE and lesion location was not confirmed by multivariate analysis. Thalamic volume and seizure semiology were not statistically different between the SRE and non-SRE group. The significant association between lesion size and SRE was reproducible in type IIb and IIa subgroups.SignificanceSRE is common in patients with FCD type II. Small FCD type II lesions are significantly associated with SRE. Although our findings cannot be applied to

Simulation of small break loss of coolant accident in pressurized water reactor (PWR)

International Nuclear Information System (INIS)

Abass, N. M. N.

2012-02-01

A major safety concern in pressurized-water-reactor (PWR) design is the loss-of-coolant accident (LOCA),in which a break in the primary coolant circuit leads to depressurization, boiling of the coolant, consequent reduced cooling of the reactor core, and , unless remedial measures are taken, overheating of the fuel rods. This concern has led to the development of several simulators for safety analysis. This study demonstrates how the passive and active safety systems in conventional and advanced PWR behave during the small break loss of Coolant Accident (SBLOCA). The consequences of SBOLOCA have been simulated using IAEA Generic pressurized Water Reactor Simulator (GPWRS) and personal Computer Transient analyzer (PCTRAN) . The results were presented and discussed. The study has confirmed the major safety advantage of passive plants versus conventional PWRs is that the passive safety systems provide long-term core cooling and decay heat removal without the need for operator actions and without reliance on active safety-related system. (Author)

Comparative study T-type and I-type layout of PWR nuclear power plants

International Nuclear Information System (INIS)

Eko Rudi Iswanto and Siti Alimah

2010-01-01

Determining plant layout is one of the five major stages during the life time of a nuclear power plant. Some important factors that affect in the selecting of plant layout are availability of infrastructure, economic aspects, social aspects, public and environment safety, and also easy to do. Another factor to be considered is requirements as seismic design, which refers to the principles of good security workers, communities and the environment of radiological risks. There are many layout types of nuclear power plant, two of them are T-type layout and I-type layout. Each type of the plant layout has advantage and disadvantage, therefore this study is to understand them. Good layout is able to provide a high level of security against earthquakes. In term of earthquake design, I-type layout has a higher security level than T-type layout. Therefore, I-type layout can be a good choice for PWR nuclear power plants 1000 MWe that will be built in Indonesia. (author)

Design of the control room of the N4-type PWR: main features and feedback operating experience

International Nuclear Information System (INIS)

Peyrouton, J.M.; Guillas, J.; Nougaret, Ch.

2004-01-01

This article presents the design, specificities and innovating features of the control room of the N4-type PWR. A brief description of control rooms of previous 900 MW and 1300 MW -type PWR allows us to assess the change. The design of the first control room dates back to 1972, at that time 2 considerations were taken into account: first the design has to be similar to that of control rooms for thermal plants because plant operators were satisfied with it and secondly the normal operating situation has to be privileged to the prejudice of accidental situations just as it was in a thermal plant. The turning point was the TMI accident that showed the weight of human factor in accidental situations in terms of pilot team, training, procedures and the ergonomics of the work station. The impact of TMI can be seen in the design of 1300 MW-type PWR. In the beginning of the eighties EDF decided to launch a study for a complete overhaul of the control room concept, the aim was to continue reducing the human factor risk and to provide a better quality of piloting the plant in any situation. The result is the control room of the N4-type PWR. Today the cumulated feedback experience of N4 control rooms represents more than 20 years over a wide range of situations from normal to incidental, a survey shows that the N4 design has fulfilled its aims. (A.C.)

Estimating probable flaw distributions in PWR steam generator tubes

International Nuclear Information System (INIS)

Gorman, J.A.; Turner, A.P.L.

1997-01-01

This paper describes methods for estimating the number and size distributions of flaws of various types in PWR steam generator tubes. These estimates are needed when calculating the probable primary to secondary leakage through steam generator tubes under postulated accidents such as severe core accidents and steam line breaks. The paper describes methods for two types of predictions: (1) the numbers of tubes with detectable flaws of various types as a function of time, and (2) the distributions in size of these flaws. Results are provided for hypothetical severely affected, moderately affected and lightly affected units. Discussion is provided regarding uncertainties and assumptions in the data and analyses

Assessment of void swelling in austenitic stainless steel PWR core internals

International Nuclear Information System (INIS)

Chung, H.M.

2006-01-01

As many pressurized water reactors (PWRs) age and life extension of the aged plants is considered, void swelling behavior of austenitic stainless steel (SS) core internals has become the subject of increasing attention. In this report, the available database on void swelling and density change of austenitic SSs was critically reviewed. Irradiation conditions, test procedures, and microstructural characteristics were carefully examined, and key factors that are important to determine the relevance of the database to PWR conditions were evaluated. Most swelling data were obtained from steels irradiated in fast breeder reactors at temperatures >385 C and at dose rates that are orders of magnitude higher than PWR dose rates. Even for a given irradiation temperature and given steel, the integral effects of dose and dose rate on void swelling should not be separated. It is incorrect to extrapolate swelling data on the basis of 'progressive compounded multiplication' of separate effects of factors such as dose, dose rate, temperature, steel composition, and fabrication procedure. Therefore, the fast reactor data should not be extrapolated to determine credible void swelling behavior for PWR end-of-life (EOL) or life-extension conditions. Although the void swelling data extracted from fast reactor studies is extensive and conclusive, only limited amounts of swelling data and information have been obtained on microstructural characteristics from discharged PWR internals or steels irradiated at temperatures and at dose rates comparable to those of a PWR. Based on this relatively small amount of information, swelling in thin-walled tubes and baffle bolts in a PWR is not considered a concern. As additional data and relevant research becomes available, the newer results should be integrated with existing data, and the worthiness of this conclusion should continue to be scrutinized. PWR baffle reentrant corners are the most likely location to experience high swelling rates, and

PWR station blackout transient simulation in the INER integral system test facility

International Nuclear Information System (INIS)

Liu, T.J.; Lee, C.H.; Hong, W.T.; Chang, Y.H.

2004-01-01

Station blackout transient (or TMLB' scenario) in a pressurized water reactor (PWR) was simulated using the INER Integral System Test Facility (IIST) which is a 1/400 volumetrically-scaled reduce-height and reduce-pressure (RHRP) simulator of a Westinghouse three-loop PWR. Long-term thermal-hydraulic responses including the secondary boil-off and the subsequent primary saturation, pressurization and core uncovery were simulated based on the assumptions of no offsite and onsite power, feedwater and operator actions. The results indicate that two-phase discharge is the major depletion mode since it covers 81.3% of the total amount of the coolant inventory loss. The primary coolant inventory has experienced significant re-distribution during a station blackout transient. The decided parameter to avoid the core overheating is not the total amount of the coolant inventory remained in the primary core cooling system but only the part of coolant left in the pressure vessel. The sequence of significant events during transient for the IIST were also compared with those of the ROSA-IV large-scale test facility (LSTF), which is a 1/48 volumetrically-scaled full-height and full-pressure (FHFP) simulator of a PWR. The comparison indicates that the sequence and timing of these events during TMLB' transient studied in the RHRP IIST facility are generally consistent with those of the FHFP LSTF. (author)

Analysis of size effect applicable to evaluation of fracture toughness of base metal for PWR vessel

International Nuclear Information System (INIS)

Benhamou, C.; Joly, P.; Andrieu, A.; Parrot, A.; Vidard, S.

2015-01-01

The objective of the present paper is to review the specimen size effect (also called crack front length effect) on Fracture Toughness of PWR Reactor Pressure Vessel Steel base metal. The analysis of the reality and amplitude of this effect is conducted in a first step on a database (the so-called GKSS database) including fracture toughness test results on a single representative material using specimens of different thicknesses, tested in the same temperature range. A realistic analytical form for describing the size effect observed in this data set is thus derived from statistical analyses and proposed for engineering application. In a second step, this size effect formulation is then applied to a large number of fracture toughness data, obtained in Irradiation Surveillance Programs, and also to the numerous data used for the definition of the ASME (and RCC-M) fracture toughness reference curves. This analysis allows normalizing all the available fracture toughness data with a single specimen width of 100 mm and defining the fracture toughness reference curve as the lower bound of this normalized set of data points. It is thus demonstrated that the fracture toughness reference curve is associated with a reference crack length of 100 mm, and can be used in RPV integrity analyses for other crack front length in association with the crack front length correction formula defined in the first step. (authors)

Technology selection for offshore underwater small modular reactors

International Nuclear Information System (INIS)

Shivan, Koroush; Ballinger, Ronald; Buongiorno, Jacopo; Forsberg, Charles; Kazimi, Mujid; Todreas, Neil

2016-01-01

This work examines the most viable nuclear technology options for future underwater designs that would meet high safety standards as well as good economic potential, for construction in the 2030-2040 time frame. The top five concepts selected from a survey of 13 nuclear technologies were compared to a small modular pressurized water reactor (PWR) designed with a conventional layout. In order of smallest to largest primary system size where the reactor and all safety systems are contained, the top five designs were: (1) a lead-bismuth fast reactor based on the Russian SVBR-100; (2) a novel organic cooled reactor; (3) an innovative superheated water reactor; (4) a boiling water reactor based on Toshiba's LSBWR; and (5) an integral PWR featuring compact steam generators. A similar study on potential attractive power cycles was also performed. A condensing and recompression supercritical CO 2 cycle and a compact steam Rankine cycle were designed. It was found that the hull size required by the reactor, safety systems and power cycle can be significantly reduced (50-80%) with the top five designs compared to the conventional PWR. Based on the qualitative economic consideration, the organic cooled reactor and boiling water reactor designs are expected to be the most cost effective options

Technology selection for offshore underwater small modular reactors

Energy Technology Data Exchange (ETDEWEB)

Shivan, Koroush; Ballinger, Ronald; Buongiorno, Jacopo; Forsberg, Charles; Kazimi, Mujid; Todreas, Neil [Dept. of Nuclear Science and Engineering, Massachusetts Institute of Technology, Cambridge (United States)

2016-12-15

This work examines the most viable nuclear technology options for future underwater designs that would meet high safety standards as well as good economic potential, for construction in the 2030-2040 time frame. The top five concepts selected from a survey of 13 nuclear technologies were compared to a small modular pressurized water reactor (PWR) designed with a conventional layout. In order of smallest to largest primary system size where the reactor and all safety systems are contained, the top five designs were: (1) a lead-bismuth fast reactor based on the Russian SVBR-100; (2) a novel organic cooled reactor; (3) an innovative superheated water reactor; (4) a boiling water reactor based on Toshiba's LSBWR; and (5) an integral PWR featuring compact steam generators. A similar study on potential attractive power cycles was also performed. A condensing and recompression supercritical CO{sub 2} cycle and a compact steam Rankine cycle were designed. It was found that the hull size required by the reactor, safety systems and power cycle can be significantly reduced (50-80%) with the top five designs compared to the conventional PWR. Based on the qualitative economic consideration, the organic cooled reactor and boiling water reactor designs are expected to be the most cost effective options.

Technology Selection for Offshore Underwater Small Modular Reactors

Directory of Open Access Journals (Sweden)

Koroush Shirvan

2016-12-01

Full Text Available This work examines the most viable nuclear technology options for future underwater designs that would meet high safety standards as well as good economic potential, for construction in the 2030–2040 timeframe. The top five concepts selected from a survey of 13 nuclear technologies were compared to a small modular pressurized water reactor (PWR designed with a conventional layout. In order of smallest to largest primary system size where the reactor and all safety systems are contained, the top five designs were: (1 a lead–bismuth fast reactor based on the Russian SVBR-100; (2 a novel organic cooled reactor; (3 an innovative superheated water reactor; (4 a boiling water reactor based on Toshiba's LSBWR; and (5 an integral PWR featuring compact steam generators. A similar study on potential attractive power cycles was also performed. A condensing and recompression supercritical CO2 cycle and a compact steam Rankine cycle were designed. It was found that the hull size required by the reactor, safety systems and power cycle can be significantly reduced (50–80% with the top five designs compared to the conventional PWR. Based on the qualitative economic consideration, the organic cooled reactor and boiling water reactor designs are expected to be the most cost effective options.

Semi-automatic ultrasonic inspection of PWR upper internal immersed components

International Nuclear Information System (INIS)

Dombret, P.; Coquette, A.; Cermak, J.; Verspeelt, D.

1985-01-01

The present paper describes the characteristics of a semi-automatic ultrasonic inspection system. Components inspected are the so-called flexures, small pins located at the upper part of control rod tube-guide, some of which happened to broke in a few Westinghouse type PWR's. Inspection results and other system capabilities are also mentioned

Containment for small pressurized water reactors

International Nuclear Information System (INIS)

Siler, W.C.; Marda, R.S.; Smith, W.R.

1977-01-01

Babcock and Wilcox Company has prepared studies under ERDA contract of small and intermediate size (313, 365 and 1200 MWt) PWR reactor plants, for industrial cogeneration or electric power generation. Studies and experience with nuclear plants in this size range indicate unfavorable economics. To offset this disadvantage, modular characteristics of an integral reactor and close-coupled vapor suppression containment have been exploited to shorten construction schedules and reduce construction costs. The resulting compact reactor/containment complex is illustrated. Economic studies to date indicate that the containment design and the innovative construction techniques developed to shorten erection schedules have been important factors in reducing estimated project costs, thus potentially making such smaller plants competetive with competing energy sources

Simulation of natural circulation on an integral type experimental facility, MASLWR

Energy Technology Data Exchange (ETDEWEB)

Chung, Youngjong; Lim, Sungwon; Ha, Jaejoo [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2013-05-15

The OSU MASLWR test facility was reconfigured to eliminate a recurring grounding problem and improve facility reliability in anticipation of conducting an IAEA International Collaborative Standard Problem (ICSP). The purpose of ICSP is to provide experimental data on flow instability phenomena under natural circulation conditions and coupled containment/reactor vessel behavior in integral-type reactors, and to evaluate system code capabilities to predict natural circulation phenomena for integral type PWR, by simulating an integrated experiment. A natural circulation in the primary side during various core powers is analyzed using TASS/SMR code for the integral type experimental facility. The calculation results show higher steady state primary flow than experiment. If it matches the initial flow with experiment, it shows lower primary flow than experiment according to the increase of power. The code predictions may be improved by applying a Reynolds number dependent form loss coefficient to accurately account for unrecoverable pressure losses.

Behavior of small-sized BWR fuel under reactivity initiated accident conditions

International Nuclear Information System (INIS)

Yanagisawa, Kazuaki; Fujishiro, Toshio; Horiki, Oichiro; Chen Dianshan; Takeuchi, Kiyoshi.

1992-01-01

The present work was performed on this small-sized BWR fuel, where Zr liner and rod prepressurization were taken as experimental parameters. Experiment was done under simulated reactivity initiated accident (RIA) conditions at Nuclear Safety Research Reactor (NSRR) belonged to Japan Atomic Energy Research Institute (JAERI). Major remarks obtained are as follows: (1) Three different types of the fuel rods consisted of (a) Zr lined/pressurized (0.65MPa), (b) Zr lined/non-pressurized and (c) non-Zr lined/pressurized (o.65MPa) were used, respectively. Failure thresholds of these were not less than that (260 cal/g·fuel) described in Japanese RIA Licensing Guideline. Small-sized BWR and conventional 8 x 8 BWR fuels were considered to be in almost the same level in failure threshold. Failure modes of the three were (a) cladding melt/brittle, (b) cladding melt/brittle and (c) rupture by large ballooning, respectively. (2) The magnitude of pressure pulse at fuel fragmentation was also studied by lined/pressurized and non-lined/pressurized fuels. Above the energy deposition of 370 cal/g·fuel, mechanical energy (or pressure) was found to be released from these fragmented fuels. No measurable difference was, however, observed between the tested fuels and NSRR standard (and conventional 8 x 8 BWR) fuels. (3) It is worthy of mentioning that Zr liner tended to prevent the cladding from large ballooning. Non-lined/pressurized fuel tended to cause wrinkle deformation at cladding. Hence, cladding external was notched much by the wrinkles. (4) Time to fuel failure measured from the tested BWR fuels (pressurization < 0.6MPA) was longer than that measured from PWR fuels (pressurization < 3.2MPa). The magnitude of the former was of the order of 3 ∼ 6s, while that of the latter was < 1s. (J.P.N.)

Programme of hot points eradication (Co-60) led on French PWR type reactors

International Nuclear Information System (INIS)

Rocher, A.; Ridoux, P.; Anthoni, S.; Brun, C.

1998-01-01

The question of hot points (pellets rich in cobalt 59 or in cobalt 60 in a PWR type reactor), is studied from the radiation protection point of view. The purpose is to see how to optimize the radiation protection, the elimination of these hot points can bring an improvement. (N.C.)

Sub-critical crack growth and clad integrity in a PWR reactor pressure vessel

International Nuclear Information System (INIS)

Tice, D.R.; Foreman, A.J.E.; Sharples, J.K.

1987-10-01

The possibility of in-service growth of sub-critical defects in a PWR reactor pressure vessel to a critical size which could result in vessel failure was addressed in both the 1976 and 1982 reports of the Light Water Reactor Study Group (LWRSG), under the Chairmanship of Dr W Marshall (now Lord Marshall). An addendum to this report was published by UKAEA in April 1987. The section of the addendum dealing with subcritical crack growth and the related issue of integrity of the stainless steel cladding on the inner vessel surface is reproduced in this report. This section of the LWRSG addendum provides a review of the current status of fatigue crack growth and environmentally assisted cracking research for pressure vessel steels in light water reactor environments, as well as a review of developments in crack growth assessment methods. The review concludes that the alternative assessment procedures now being developed give a more realistic prediction of in service crack growth than the ASME Section XI Appendix A fatigue crack growth curves. (author)

3. Workshop for IAEA ICSP on Integral PWR Design Natural Circulation Flow Stability and Thermo-hydraulic Coupling of Containment and Primary System during Accidents. Presentations

International Nuclear Information System (INIS)

2012-04-01

Most advanced nuclear power plant designs adopted several kinds of passive systems. Natural circulation is used as a key driving force for many passive systems and even for core heat removal during normal operation such as NuScale, CAREM, ESBWR and Indian AHWR designs. Simulation of natural circulation phenomena is very challenging since the driving force of it is weak compared to forced circulation and involves a coupling between primary system and containment for integral type reactor. The IAEA ICSP (International Collaborative Standard Problem) on 'Integral PWR Design Natural Circulation Flow Stability and Thermo-hydraulic Coupling of Containment and Primary System during Accidents' was proposed within the CRP on 'Natural Circulation Phenomena, Modelling, and Reliability of Passive Systems that utilize Natural Circulation'. Oregon State University (OSU) of USA offered to host this ICSP. This ICSP plans to conduct the following experiments and blind/open simulations with system codes: 1. Quasi-steady state operation with different core power levels: Conduct quasi-steady state operation with step-wise increase of core power level in order to observe single phase natural circulation flow according to power level. The experimental facility and operating conditions for an integral PWR will be used. 2. Thermo-hydraulic Coupling between Primary system and Containment: Conduct a loss of feedwater transient with subsequent ADS blowdown and long term cooling to determine the progression of a loss of feedwater transient by natural circulation through primary and containment systems. These tests would examine the blowdown phase as well as the long term cooling using sump natural circulation by coupling the primary to containment systems. This data could be used for the evaluation of system codes to determine if they model specific phenomena in an accurate manner. OSU completed planned two ICSP tests in July 2011 and real initial and boundary conditions measured from the

R and D status of an integral type small reactor MRX in JAERI

International Nuclear Information System (INIS)

Hoshi, Tsutao; Ochiai, Masaaki; Iida, Hiromasa; Yamaji, Akio; Shimazaki, Junya

1995-01-01

JAERI is conducting a design study on an integral type small reactor MRX for the use of nuclear ships. The basic concept of the reactor system is the integral type reactor with in-vessel steam generators and control rod drive systems, however, such new technologies as the water-filled containment, the passive decay heat removal system, the advanced automatic system, etc., are adopted to satisfy the essential requirements for the next generation ship reactors, i.e. compact, light, highly safe and easy operation. Research and development (R and D) works have being progressed on the peculiar components, the advanced automatic operation systems and the safety study of the thermal hydraulic phenomena as well as the feasibility study of the applicability to merchant ships. The experiments and analysis of the safety carried out so far are proving that the passive safety features applied into the MRX are sufficient functions in the safety point of view. The MRX is a typical small type reactor realizing the easy operation by simplifying the reactor systems adopting the passive safety systems, therefore, it has wide variety of use as energy supply systems. This paper summarizes the present status on the design study of the MRX and the research and development activities as well as the results of feasibility study. (author)

A study for small-medium LWR development of JAPC

International Nuclear Information System (INIS)

Okazaki, Toshihiko; Hida, Takahiko; Hoshi, Takashi; Kawahara, Hiroto; Tominaga, Kenji; Asano, Hiromitsu

2011-01-01

LWR (Light Water Reactor) power stations have accumulated many experiences of design, construction and operation. In addition, large-sized reactors have an advantage of economy of scale and 1,000 MWe LWR has therefore become the mainstream reactor in Japan. Meanwhile, introduction of the medium and small-sized LWRs (SMRs) has also been under review in Japan in order to respond to stagnant growth in electricity demand and electricity market liberalization or for investment risk mitigation; however, it has not been realized due to the economic disadvantage of scale. Therefore, JAPC has been developing the concept of SMR (300 MWe - 600 MWe) which is competitive to the large-sized LWR cooperating with Japanese plant makers (Hitachi, Toshiba Corporation and Mitsubishi Heavy Industries), assessing the possibility of realization of SMRs as one of the electric power sources in the future. As the result of the JAPC's study, we developed SMR concepts whose cost and safety are almost equal to large-sized LWR and confirmed technical feasibility of the concept in order to start developing basic design. In this paper, the outline of the SMR concepts and the current development status are presented. Concepts have been developed for two types of SMRs (i.e. BWR and PWR). As for the BWR type, reactor system is simplified by adopting natural circulation core method and CRD falling under gravity in order to downsize the reactor containments. As for the PWR type, the risk of LOCA occurrence is eliminated by unifying the primary system (e.g. incorporating steam generator into reactor). Furthermore, the primary system is simplified by adopting natural circulation core method in operation and containment vessel also become compact for the PWR. As for JAPC's further development of SMRs, key elements of SMR concepts are studied. In addition, the environment surrounding the SMRs has changed in recent years and the one with capacity exceeding 300-600 MWe class or small-sized reactor with

Reactor control system. PWR

International Nuclear Information System (INIS)

2009-01-01

At present, 23 units of PWR type reactors have been operated in Japan since the start of Mihama Unit 1 operation in 1970 and various improvements have been made to upgrade operability of power stations as well as reliability and safety of power plants. As the share of nuclear power increases, further improvements of operating performance such as load following capability will be requested for power stations with more reliable and safer operation. This article outlined the reactor control system of PWR type reactors and described the control performance of power plants realized with those systems. The PWR control system is characterized that the turbine power is automatic or manually controlled with request of the electric power system and then the nuclear power is followingly controlled with the change of core reactivity. The system mainly consists of reactor automatic control system (control rod control system), pressurizer pressure control system, pressurizer water level control system, steam generator water level control system and turbine bypass control system. (T. Tanaka)

Design study of a PWR of 1.300 MWe of Angra-2 type operating in the thorium cycle

International Nuclear Information System (INIS)

Andrade, E.P.; Carneiro, F.A.N.; Schlosser, G.J.

1984-01-01

The utilization of the thorium-highly enriched uranium and thorium-plutonium mixed oxide fuels in an unmodified PWR is analysed. The PWR of 1300 MWe from KWU (Angra-2 type) is taken as the reference reactor for the study. Reactor core design calculations for both types of fuels considering once-through and recycle fuels. The calculations were performed with the KWU design codes FASER-3 and MEDIUM 2.2 after introduction of the thorium chain and some addition of nuclide data in FASER-3. A two-energy group scheme and a two-dimensional (XY) representation of the reactor core were utilized. (Author) [pt

Development of a computer code for transients simulation in PWR type reactors

International Nuclear Information System (INIS)

Alvim, A.C.M.; Botelho, D.A.; Oliveira Barroso, A.C. de

1981-01-01

A computer code for the simulation of operacional-transients and accidents in PWR type reactors is being developed at IEN (Instituto de Engenharia Nuclear). Accidents will be considered in which variations in thermohydraulics parameters of fuel and coolant don't cause nucleate boiling in the reactor core, but, otherwise are sufficiently strong to justify a more detailed simulation than that used in linearized models. (E.G.) [pt

Fuel assemblies for PWR type reactors: fuel rods, fuel plates. CEA work presentation

International Nuclear Information System (INIS)

Delafosse, Jacques.

1976-01-01

French work on PWR type reactors is reported: basic knowledge on Zr and its alloys and on uranium oxide; experience gained on other programs (fast neutron and heavy water reactors); zircaloy-2 or zircaloy-4 clad UO 2 fuel rods; fuel plates consisting of zircaloy-2 clad UO 2 squares of thickness varying between 2 and 4mm [fr

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)

Assessment of erbium as candidate burnable absorber for future PWR operaning cycles: A neutronic and fabrication study

International Nuclear Information System (INIS)

Asou, M.; Dehaudt, P.; Porta, J.

1995-01-01

Erbium begins to play a role in the control of PWR core reactivity. Generally speaking, burnable absorbers were only used to establish fresh core equilibrium. In France, since the possibility of extending irradiation cycles by 12 to 18 months, then up to 24 and 30 months, has been envisaged, there is renewed interest in burnable absorbers. The fabrication of PWR pellets has been investigated, providing high density and a good erbium homogeneity. The pellets characteristics were consistent with the specifications of PWR fuel. However, with the present process, the grain size remains small. Studies in progress now shows that erbium is not only a valuable alternative to gadolinium, for long fuel cycles (≥18 months) but also a new fuel concept. (orig.)

Manufacturing and properties of closure head forging integrated with flange for PWR reactor pressure vessel

International Nuclear Information System (INIS)

Tomoharu Sasaki; Iku Kurihara; Etsuo Murai; Yasuhiko Tanaka; Koumei Suzuki

2003-01-01

Closure head forging (SA508, Gr.3 Cl.1) integrated with flange for PWR reactor pressure vessel has been developed. This is intended to enhance structural integrity of closure head resulted in elimination of ISI, by eliminating weld joint between closure head and flange in the conventional design. Manufacturing procedures have been established so that homogeneity and isotropy of the material properties can be assured in the closure head forging integrated with flange. Acceptance tensile and impact test specimens are taken and tested regarding the closure head forging integrated with flange as very thick and complex forgings. This paper describes the manufacturing technologies and material properties of the closure head forging integrated with flange. (orig.)

An integrated approach for developing a technology strategy framework for small- to medium-sized furniture manufacturers to improve competitiveness

Directory of Open Access Journals (Sweden)

Lourens, A.S.

2013-05-01

Full Text Available Low-technology firms, such as those found within the furniture manufacturing industry, have no framework or methodology to guide them successfully to acquire and integrate technology that enables them to operate more competitively. The aim of this article is to illustrate the development of a technology strategy framework for small- to medium-sized furniture manufacturers to assist them to improve their competitiveness. More specifically, this article presents an integrated technology strategy framework that enables management to integrate their business strategy with their technology requirements successfully, thus improving competitiveness.

The Conceptual Design of Innovative Safe PWR

Energy Technology Data Exchange (ETDEWEB)

Kim, Han-Gon [Centural Research Institute, Daejeon (Korea, Republic of); Heo, Sun [Korea Hydro and Nuclear Power Co., Daejeon (Korea, Republic of)

2016-10-15

Most of countries operating NPPs have been performed post-Fukushima improvements as short-term countermeasure to enhance the safety of operating NPPs. Separately, vendors have made efforts on developing passive safety systems as long-term and ultimate countermeasures. AP1000 designed by Westinghouse Electric Company has passive safety systems including the passive emergency core cooling system (PECCS), the passive residual heat removal system (PRHRS), and the passive containment cooling system (PCCS). ESBWR designed by GE-Hitachi also has passive safety systems consisting of the isolation condenser system, the gravity driven cooling system and the PCCS. Other countries including China and Russia have made efforts on developing passive safety systems for enhancing the safety of their plants. In this paper, we summarize the design goals and main design feature of innovative safe PWR, iPOWER which is standing for Innovative Passive Optimized World-wide Economical Reactor, and show the developing status and results of research projects. To mitigate an accident without electric power and enhance the safety level of PWR, the conceptual designs of passive safety system and innovative safe PWR have been performed. It includes the PECCS for core cooling and the PCCS for containment cooling. Now we are performing the small scale and separate effect tests for the PECCS and the PCCS and preparing the integral effect test for the PECCS and real scale test for the PCCS.

Development of technologies for nuclear reactors of small and medium sized; Desarrollo de Tecnologias para Reactores Nucleares de pequeno y medio tamano

Energy Technology Data Exchange (ETDEWEB)

NONE

2011-08-15

This meeting include: countries presentations, themes and objectives of the training course, reactor types, design, EPR, APR1400, A P 1000, A PWR, ATMEA 1, VVER-1000, A PWR, ATMEA 1, VVER 1200, Boiling Water Reactor, A BWR, A BWR -II, ESBUR, Ke ren, AREVA, Heavy Water Reactor, Candu 6, Acr-1000, HWR, Bw, Iris, CAREM NuCcale, Smart, KLT-HOS, Westinghouse small modular Reactor, Gas Cooled Reactors, PBMR, React ores enfriados con metales liquidos, Hs, Prism,Terra Power, Hyper ion, appliance's no electric as de energia, Generation IV Reactors,VHTR, Gas Fast Reactor, Sodium Fast Reactor, Molten salt Reactor, Lfr, Water Cooled Reactor, Technology Assessment Process, Fukushima accident.

Thermal-hydraulic analysis of PWR small assembly for irradiation test of CARR

International Nuclear Information System (INIS)

Yin Hao; Zou Yao; Liu Xingmin

2015-01-01

The thermal-hydraulic behaviors of the PWR 4 × 4 small assembly tested in the high temperature and high pressure loop of China Advanced Research Reactor were analyzed. The CFD method was used to carry out 3D simulation of the model, thus detailed thermal-hydraulic parameters were obtained. Firstly, the simplified model was simulated to give the 3D temperature and velocity distributions and analyze the heat transfer process. Then the whole scale small assembly model was simulated and the simulation results were compared with those of simplified rod bundle model. Its flow behavior was studied and flow mixing characteristics of the grids were analyzed, and the mixing factor of the grid was calculated and can be used for further thermal-hydraulic study. It is shown that the highest temperature of the fuel rod meets the design limit and the mixing effect of the grid is obvious. (authors)

Development of an advanced 16x165 Westinghouse type PWR fuel assembly for Slovenia

International Nuclear Information System (INIS)

Boone, M. L.; King, S. J.; Pulver, E. F.; Jeon, K.-L.; Esteves, R.; Kurincic, B.

2004-01-01

Industrias Nucleares do Brasil (INB), KEPCO Nuclear Fuel Company, Ltd. (KNFC), and Westinghouse Electric Company (Westinghouse) have jointly designed an advanced 16x16 Westinghouse type PWR fuel assembly. This advanced 16x16 Westinghouse type PWR fuel assembly, which will be implemented in both Kori Unit 2 (in Korea) and Angra Unit 1 (in Brazil) in January and March 2005, respectively, is an integral part of the utilities fuel management strategy. This same fuel design has also been developed for future use in Krsko Unit 1 (in Slovenia). In this paper we will describe the front-end nuclear fuel management activities utilized by the joint development team and describe how these activities played an integral part in defining the direction of the advanced 16x16 Westinghouse type PWR fuel assembly design. Additionally, this paper will describe how this design demonstrates improved margins under high duty plant operating conditions. The major reason for initiating this joint development program was to update the current 16x16 fuel assembly, which is also called 16STD. The current 16STD fuel assembly contains a non-optimized fuel rod diameter for the fuel rod pitch (i.e. 9.5 mm OD fuel rods at a 0.485 inch pitch), non-neutronic efficient components (i.e. Inconel Mid grids), no Intermediate Flow Mixer (IFM) grids, and other mechanical features. The advanced 16x16 fuel assembly is being designed for peak rod average burnups of up to 75 MWd/kgU and will use an optimized fuel rod diameter (i.e. 9.14 mm OD ZIRLO TM fuel rods), neutronic efficient components (i.e. ZIRLO TM Mid grids), ZIRLO TM Intermediate Flow Mixer (IFM) grids to improve Departure from Nucleate Boiling (DNB) margin, and many other mechanical features that improve design margins. Nuclear design activities in the areas of fuel cycle cost and fuel management were performed in parallel to the fuel assembly design efforts. As the change in reactivity due to the change in the fuel rod diameter influences directly

Model for calculating the boron concentration in PWR type reactors

International Nuclear Information System (INIS)

Reis Martins Junior, L.L. dos; Vanni, E.A.

1986-01-01

A PWR boron concentration model has been developed for use with RETRAN code. The concentration model calculates the boron mass balance in the primary circuit as the injected boron mixes and is transported through the same circuit. RETRAN control blocks are used to calculate the boron concentration in fluid volumes during steady-state and transient conditions. The boron reactivity worth is obtained from the core concentration and used in RETRAN point kinetics model. A FSAR type analysis of a Steam Line Break Accident in Angra I plant was selected to test the model and the results obtained indicate a sucessfull performance. (Author) [pt

A small long-cycle PWR core design concept using fully ceramic micro-encapsulated (FCM) and UO2–ThO2 fuels for burning of TRU

International Nuclear Information System (INIS)

Bae, Gonghoon; Hong, Ser Gi

2015-01-01

In this paper, a new small pressurized water reactor (PWR) core design concept using fully ceramic micro-encapsulated (FCM) particle fuels and UO 2 –ThO 2 fuels was studied for effective burning of transuranics from a view point of core neutronics. The core of this concept rate is 100 MWe. The core designs use the current PWR-proven technologies except for a mixed use of the FCM and UO 2 –ThO 2 fuel pins of low-enriched uranium. The significant burning of TRU is achieved with tri-isotropic particle fuels of FCM fuel pins, and the ThO 2 –UO 2 fuel pins are employed to achieve long-cycle length of ∼4 EFPYs (effective full-power year). Also, the effects of several candidate materials for reflector are analyzed in terms of core neutronics because the small core size leads to high sensitivity of reflector material on the cycle length. The final cores having 10 w/o SS303 and 90 w/o graphite reflector are shown to have high TRU burning rates of 33%–35% in FCM pins and significant net burning rates of 24%–25% in the total core with negative reactivity coefficients, low power peaking factors, and sufficient shutdown margins of control rods. (author)

Testing of elastomer seals using small-size rigs

International Nuclear Information System (INIS)

Leeks, C.W.E.; Dunford, B.; Barnfield, J.H.; Gray, I.L.S.

1997-01-01

This paper looks at the use of small size seal leakage test rigs to demonstrate the compliance of full size container seals against the IAEA Transport Regulation's limits for activity release for normal transport and accident conditions. The detailed requirements of the regulations are discussed and it is concluded that an appropriate test programme to meet these requirements using only small size test rigs, can normally be set up and carried out on a relatively short time scale. It is important that any small test rigs should be designed to represent the relevant features of the seal arrangement and the overall test programme should cover all of the conditions, specified by the regulations, for the type, classification and contents of the container under consideration. The parameters of elastomer O-rings, which affect their sealing ability, are considered and those which are amenable to small scale testing or have to be modelled at full size are identified. Generally, the seals used in leakage tests have to be modelled with a full size cross-section but can have a reduced peripheral length. (Author)

Verification test for radiation reduction effect and material integrity on PWR primary system by zinc injection

Energy Technology Data Exchange (ETDEWEB)

Kawakami, H.; Nagata, T.; Yamada, M. [Nuclear Power Engineering Corp. (Japan); Kasahara, K.; Tsuruta, T.; Nishimura, T. [Mitsubishi Heavy Industries, Ltd. (Japan); Ishigure, K. [Saitama Inst. of Tech. (Japan)

2002-07-01

Zinc injection is known to be an effective method for the reduction of radiation source in the primary water system of a PWR. There is a need to verify the effect of Zn injection operation on radiation source reduction and materials integrity of PWR primary circuit. In order to confirm the effectiveness of Zn injection, verification test as a national program sponsored by Ministry of Economy, Trade and Industry (METI) was started in 1995 for 7-year program, and will be finished by the end of March in 2002. This program consists of irradiation test and material integrity test. Irradiation test as an In-Pile-Test managed by AEAT Plc(UK) was performed using the LVR-15 reactor of NRI Rez in Check Republic. Furthermore, Out-of-Pile-Test using film adding unit was also performed to obtain supplemental data for In-Pile-Test at Takasago Engineering Laboratory of NUPEC. Material Integrity test was planned to perform constant load test, constant strain test and corrosion test at the same time using large scale Loop and slow strain extension rate testing (SSRT) at Takasago Engineering Laboratory of NUPEC. In this paper, the results of the verification test for Zinc program at present are discussed. (authors)

Study of crack propagation velocity in steel tanks of PWR type reactor

International Nuclear Information System (INIS)

Amzallac, C.; Bernard, J.L.; Slama, G.

1983-05-01

Description and results of a serie of tests carried out on crack propagation velocity of steels in PWR environment (pressurized high temperature water), in order to examine the effects of metallurgical parameters such as chemical composition of steel, especially sulfur and carbon content, and steel type (laminate or forged steels), effects of mechanical parameters such as loading ratio, cycle form, frequency and application mode of loads and of chemical parameters (anodal dissolution or fatigue with hydrogen) [fr

Sizewell: proposed site for Britain's first PWR power station

International Nuclear Information System (INIS)

1980-10-01

The pamphlet covers the following points, very briefly: nuclear power - a success story; the Government's nuclear programme; why Sizewell; the PWR (with diagram); the PWR at Sizewell (with aerial view) (location; size; cooling water; road access; fuel transport; construction; employment; environment; screening; the next steps (licensing procedures, etc.); safety; further information). (U.K.)

Reactor building seismic analysis of a PWR type - NPP

International Nuclear Information System (INIS)

Kakubo, Masao

1983-01-01

Earthquake engineering studies raised up in Brazil during design licensing and construction phases of Almirante Alvaro Alberto NPP, units 1 and 2. State of art of soil - structure interaction analysis with particular reference to the impedance function calculation analysis with particular reference to the impedance function calculation of a group of pile is presented in this M.Sc. Dissertation, as an example the reactor building dynamic response of a 1325 MWe NPP PWR type is calculated. The reactor building is supported by a pile foundation with 2002 end bearing piles. Upper and lower bound soil parameters are considered in order to observe their influence on dynamic response of structure. Dynamic response distribution on pile heads show pile-soil-pile interaction effects. (author)

Feasibility study on thermal-hydraulic design of reduced-moderation PWR-type core

International Nuclear Information System (INIS)

Yoshida, Hiroyuki; Ohnuki, Akira; Akimoto, Hajime

2000-03-01

At JAERI, a conceptual study on reduced-moderation water reactor (RMWR) has been performed as one of the advanced reactor system which is designed so as to realize the conversion ratio more than unity. In this reactor concept, the gap spacing between the fuel rods is remarkably narrower than in a reactor currently operated. Therefore, an evaluation of the core thermal margin becomes very important in the design of the RMWR. In this study, we have performed a feasibility evaluation on thermal-hydraulic design of RM-PWR type core (core thermal output: 2900 MWt, Rod gaps: 1 mm). In RM-PWR core, seed and blanket regions are exist. In the blanket region, power density is lower than that of the seed region. Then, evaluation was performed under setting a channel box to each fuel assembly in order to adjust the flow rate in each assembly, because it is possible that the coolant boils in the seed region. In the feasibility evaluations, subchannel code COBRA-IV-I was used in combination with KfK DNB (departure nucleate boiling) correlation. When coolant mass flow rate to the blanket fuel assembly is reduced by 40%, and that to the seed fuel assembly is increased, coolant boiling is not occurred in the assembly region calculation. Provided that the channel boxes to the blanket fuel assembly are set up and coolant mass flow rate to the blanket fuel assembly is reduced by 40%, it is confirmed by the whole core calculation that the boiling of the coolant is not occurred and the RM-PWR core is feasible. (author)

Measurement of mist cooling of PWR during LOCA by LDA

International Nuclear Information System (INIS)

Lee, S.L.; Sheen, H.J.; Issapour, I.

1985-01-01

The prediction of temperature distribution and heat transfer within rod bundles during the refill and reflood phase of a LOCA (loss of coolant accident) is of critical importance for determining the location and size of blockages due to clad deformation in a pressurized water reactor (PWR). Mist cooling by small droplets generated from large droplets on hitting grid spacers has been suggested as one of the most important heat transfer mechanisms which are responsible for the development of this temperature transient. The questions to be asked are whether such small droplets indeed exist and, if so, how are they related to the cooling of the fuel rods. Hereby reported is the result of a direct experimental investigation on these questions by a special laser-Doppler anemometry (LDA) particle sizing technique together with temperature measurements of the rod claddings and flow in the subchannel

Laboratory results gained from cold worked type 316Ti under simulated PWR primary environment

International Nuclear Information System (INIS)

Devrient, B.; Kilian, R.; Koenig, G.; Widera, M.; Wermelinger, T.

2015-01-01

Beginning in 2005, intergranular stress corrosion cracking (IGSCC) of barrel bolts made from cold worked type 316Ti (German Material No. 1.4571 K) was observed in several S/KWU type PWRs. This mechanism was so far less understood for PWR primary conditions. Therefore an extended joint research program was launched by AREVA GmbH and VGB e.V. to clarify the specific conditions which contributed to the observed findings on barrel bolts. In the frame of this research program beneath the evaluation of the operational experience also laboratory tests on the general cracking behavior of cold worked type 316Ti material, which followed the same production line as for barrel bolt manufacturing in the eighties, with different cold work levels covering up to 30 % were performed to determine whether there is a specific susceptibility of cold worked austenitic stainless steel specimens to suffer IGSCC under simulated PWR primary conditions. All these slow strain rate tests on tapered specimens and component specimens came to the results that first, much higher cold work levels than used for the existing barrel bolts are needed for IGSCC initiation. Secondly, additional high active plastic deformation is needed to generate and propagate intergranular cracking. And thirdly, all specimens finally showed ductile fracture at the applied strain rates. (authors)

An experimental study on effective depressurization actions for PWR vessel bottom small break LOCA with HPI failure and gas inflow (ROSA-V test SB-PV-04)

International Nuclear Information System (INIS)

Suzuki, Mitsuhiro; Takeda, Takeshi; Asaka, Hideaki; Nakamura, Hideo

2006-03-01

A small break loss-of-coolant accident (SBLOCA) experiment was conducted at the Large Scale Test Facility (LSTF) of ROSA-V program to study effects of accident management (AM) measures on core cooling, which are important in case of total failure of high pressure injection (HPI) system during an SBLOCA at a pressurized water reactor (PWR). The LSTF is a full-height and 1/48 volume-scaled facility simulating a 4-loop Westing-house-type PWR (3423 MWt). The experiment, SB-PV-04, simulated a PWR vessel bottom SBLOCA with a rupture of ten instrument-tubes which is equivalent to 0.2% cold leg break. It is clarified that AM actions with steam generator (SG) rapid depressurization by fully opening relief valves and auxiliary feedwater supply are effective to avoid core uncovery by actuating the low pressure injection (LPI) system though the primary depressurization is degraded by non-condensable gas inflow to the primary loops from the accumulator injection system. The effective core cooling was established by the rapid depressurization which contributed to preserve larger primary coolant mass than in the previous experiment (SB-PV-03) which was conducted with smaller primary cooling rate of -55 K/h as AM actions. (author)

Study on new-type fuel-related assembly handling tools for PWR NPP

International Nuclear Information System (INIS)

Fan Xiumei

2013-01-01

This article describes the design and study on a set of new-type fuel-related assembly snatching tools used for PWR NPP. The purpose is mainly to enhance the tool safety, reliability and convenientness by improvement of the mechanism and structure of the tool for snatching preciseness and avoiding from falling and abrasion of fuel-related assemblies for any condition. The new-type fuel-related assembly handling tools are compared with similar equipment in worldwide in terms of function, main technical characteristic, and safety and protection, some of them are better than the similar equipment in that they have reliable loading and unloading and conveying capabilities. (author)

Dissolution process for advanced-PWR-type fuels

International Nuclear Information System (INIS)

Black, D.E.; Decker, L.A.; Pearson, L.G.

1979-01-01

The new Fluorinel Dissolution Process and Fuel Storage (FAST) Facility at ICPP will provide underwater storage of spent PWR fuel and a new head-end process for fuel dissolution. The dissolution will be two-stage, using HF and HNO 3 , with an intermittent H 2 SO 4 dissolution for removing stainless steel components. Equipment operation is described

Program of monitoring PWR fuel in Spain; Programa de Vigilancia de Combustible pwr en Espana

Energy Technology Data Exchange (ETDEWEB)

Martinez Murillo, J. C.; Quecedo, M.; Munoz-Roja, C.

2015-07-01

In the year 2000 the PWR utilities: Centrales Nucleares Almaraz-Trillo (CNAT) and Asociacion Nuclear Asco-Vandellos (ANAV), and ENUSA Industrias Avanzadas developed and executed a coordinated strategy named PIC (standing for Coordinated Research Program), for achieving the highest level of fuel reliability. The paper will present the scope and results of this program along the years and will summarize the way the changes are managed to ensure fuel integrity. The excellent performance of the ENUSA manufactured fuel in the PWR Spanish NPPs is the best indicator that the expectations on this program are being met. (Author)

Transient performance of flow in circuits of PWR type reactors

International Nuclear Information System (INIS)

Hirdes, V.R.; Carajilescov, P.

1988-09-01

Generally, PWR's are designed with several primary loops, each one provided with a pump to circulate the coolant through the core. If one or more of these pumps fail, there would be a decrease in reactor flow rate which could cause coolant phase change in the core and components overheating. The present work establishes a simulation model for pump failure in PWR's and the SARDAN-FLOW computes code was developed, considering any combination of such failures. Based on the data of Angra I, several accident and operational transient conditions were simulated. (author) [pt

Radiation embrittlement of PWR vessel supports

International Nuclear Information System (INIS)

Cheverton, R.D.; Robinson, G.C.; Pennell, W.E.; Nanstad, R.K.

1989-01-01

Several studies pertaining to radiation damage of PWR vessel supports were conducted between 1978 and 1987. During this period, apparently there was no reason to believe that low-temperature (<100 degree C) MTR embrittlement data were not appropriate for evaluating embrittlement of PWR vessel supports. However, late in 1986, data from the High Flux Isotope Reactor (HFIR) vessel surveillance program indicated that the embrittlement rates of the several HFIR vessel materials (A212-B, A350-LF3, A105-II) were substantially greater than anticipated on the basis of MTR data. Further evaluation of the HFIR data suggested that a fluence-rate effect was responsible for the apparent discrepancy, and shortly thereafter it became apparent that this rate effect was applicable to the evaluation of LWR vessel supports. As a result, the Nuclear Regulatory Commission (NRC) requested that the Oak Ridge National Laboratory (ORNL) evaluate the impact of the apparent embrittlement rate effect on the integrity of light-water-reactor (LWR) vessel supports. The purpose of the study was to provide an indication of whether the integrity of reactor vessel supports is likely to be challenged by radiation-induced embrittlement. The scope of the evaluation included correlation of the HFIR data for application to the evaluation of LWR vessel supports; a survey and cursory evaluation of all US LWR vessel support designs, selection of two plants for specific-plant evaluation, and a specific-plant evaluation of both plants to determine critical flaw sizes for their vessel supports. 19 refs., 8 figs., 2 tabs

Status of developing advanced PWR in Japan

International Nuclear Information System (INIS)

Iida, Yotaro

1982-01-01

During past eleven years since the first PWR power plant, Mihama Unit 1 of Kansai Electric Power Co., started the commercial operation in 1970, Mitsubishi Heavy Industries has endeavored to improve PWR technologies on the basis of the advice from electric power companies and the technical information to overcome difficulties in PWR power plants. Now, the main objective is to improve the overall plant performance, and the rate of operation of Japanese PWR power plants has significantly risen. The improvement of the reliability, the shortening of regular inspection period and the reduction of radioactive waste handling were attempted. In view of the satisfactory operational experience of Westinghouse type PWRs, the basic reactor concept has not been changed so far. Mitsubishi and Westinghouse reached basic agreement in August, 1981, to develop a spectral shift type large capacity reactor as the advanced PWRs for Japan. This type of PWRs hab higher degree of freedom for extended fuel cycle operation and enhances the advantage of entire fuel cycle economy, particularly the significant reduction of uranium use. The improved neutron economy is attainable by reducing neutron loss, and the core design with low power density and the economical use of plutonium are advantageous for the fuel cycle economy. (Kako, I.)

Activity transport models for PWR primary circuits; PWR-ydinvoimalaitoksen primaeaeripiirin aktiivisuuskulkeutumismallit

Energy Technology Data Exchange (ETDEWEB)

Tanner, V; Rosenberg, R [VTT Chemical Technology, Otaniemi (Finland)

1995-03-01

The corrosion products activated in the primary circuit form a major source of occupational radiation dose in the PWR reactors. Transport of corrosion activity is a complex process including chemistry, reactor physics, thermodynamics and hydrodynamics. All the mechanisms involved are not known and there is no comprehensive theory for the process, so experimental test loops and plant data are very important in research efforts. Several activity transport modelling attempts have been made to improve the water chemistry control and to minimise corrosion in PWR`s. In this research report some of these models are reviewed with special emphasis on models designed for Soviet VVER type reactors. (51 refs., 16 figs., 4 tabs.).

Results of 15 years experiments in the PMK-2 integral-type facility for VVERs

Energy Technology Data Exchange (ETDEWEB)

Szabados, L.; Ezsoel, G.; Perneczky, L. [KFKI Atomic Energy Research Institute, Budapest (Hungary)

2001-07-01

Due to the specific features of the VVER-440/213-type reactors the transient behaviour of such a reactor system is different from the usual PWR system behaviour. To provide an experimental database for the transient behaviour of VVER systems the PMK integral-type facility, the scaled down model of the Paks NPP was designed and constructed in the early 1980's. Since the start-up of the facility 48 experiments have been performed. It was confirmed through the experiments that the facility is a suitable tool for the computer code validation experiments and to the identification of basic thermal-hydraulic phenomena occurring during plant accidents. High international interest was shown by the four Standard Problem Exercises of the IAEA and by the projects financed by the EU-PHARE. A wide range of small- and medium-size LOCA sequences have been studied to know the performance and effectiveness of ECC systems and to evaluate the thermal-hydraulic safety of the core. Extensive studies have been performed to investigate the one- and two-phase natural circulation, the effect of disturbances coming from the secondary circuit and to validate the effectiveness of accident management measures like bleed and feed. The VVER-specific case, the opening of the SG collector cover was also extensively investigated. Examples given in the report show a few results of experiments and the results of calculation analyses performed for validation purposes of codes like RELAP5, ATHLET and CATHARE. There are some other white spots in Cross Reference Matrices for VVER reactors and, therefore, further experiments are planned to perform tests primarily in further support of accident management measures at low power states of plants to facilitate the improved safety management of VVER-440-type reactors. (authors)

Results of 15 years experiments in the PMK-2 integral-type facility for VVERs

International Nuclear Information System (INIS)

Szabados, L.; Ezsoel, G.; Perneczky, L.

2001-01-01

Due to the specific features of the VVER-440/213-type reactors the transient behaviour of such a reactor system is different from the usual PWR system behaviour. To provide an experimental database for the transient behaviour of VVER systems the PMK integral-type facility, the scaled down model of the Paks NPP was designed and constructed in the early 1980's. Since the start-up of the facility 48 experiments have been performed. It was confirmed through the experiments that the facility is a suitable tool for the computer code validation experiments and to the identification of basic thermal-hydraulic phenomena occurring during plant accidents. High international interest was shown by the four Standard Problem Exercises of the IAEA and by the projects financed by the EU-PHARE. A wide range of small- and medium-size LOCA sequences have been studied to know the performance and effectiveness of ECC systems and to evaluate the thermal-hydraulic safety of the core. Extensive studies have been performed to investigate the one- and two-phase natural circulation, the effect of disturbances coming from the secondary circuit and to validate the effectiveness of accident management measures like bleed and feed. The VVER-specific case, the opening of the SG collector cover was also extensively investigated. Examples given in the report show a few results of experiments and the results of calculation analyses performed for validation purposes of codes like RELAP5, ATHLET and CATHARE. There are some other white spots in Cross Reference Matrices for VVER reactors and, therefore, further experiments are planned to perform tests primarily in further support of accident management measures at low power states of plants to facilitate the improved safety management of VVER-440-type reactors. (authors)

Method of stopping operation of PWR type reactor

International Nuclear Information System (INIS)

Ueno, Takashi; Tsuge, Ayao; Kawanishi, Yasuhira; Onimura, Kichiro; Kadokami, Akira.

1989-01-01

In PWR type reactors after long period of l00 % power operation, since boiling is caused in heat conduction pipes and water is depleted within the intergranular corrosion fracture face in the crevis portion to result in a dry-out state, impregnation and concentration of corrosion inhibitors into the intergranular corrosion fracture face are insufficient. In view of the above, the corrosion inhibitor at a high concentration is impregnated into the intergranular corrosion fracture face by keeping to inject the corrosion inhibitor from l00 % thermal power load by way of the thermal power reduction to the zero power state upon operatioin shutdown. That is, if the thermal power is reduced to or near the 0 power upon reactor shutdown, feedwater in the crevis portion is put to subcooled state, by which the steam present in the intergranular corrosion fracture face are condensated and the corrosion inhibitor at high concentration impregnated into the crevis portion are penetrated into the intergranular corrosion fracture face. (K.M.)

Development of small-size baking oven

Energy Technology Data Exchange (ETDEWEB)

Tabata, Akihisa; Kuwabara, Shigeru; Yamazawa, Yoshitaka; Shigeta, Eiji

1987-03-01

In the bakery business, oven fresh bakeries selling fresh bread by installing their own baking ovens at their shops have become popular recently. This article reports the development of a small-size gas baking oven for oven fresh bakaries. The gas convection oven developed recently is based on the structure of the conventional electric convection oven, and uses low pressure gas. The gas oven has an advantage that the combustion gas contains moisture. The convection oven bakes bread normally at the baking density approximately 2.5 times as much as that of the radiation oven, thereby the size of the oven may become smaller. This oven can bake many kinds of bread ranging from croissants to bean-jam buns by gas combnstion heat as well as radiation heat from the radiation plates installed at the top of each compartment in the oven. An ultra small air heat type burner was developed to provide stable short flames in order to make the size of the combustion chamber smaller. (20 figs, 2 tabs)

Evaluation of the fuel rod integrity in PWR reactors from the spectrometric analysis of the primary coolant

International Nuclear Information System (INIS)

Monteiro, Iara Arraes

1999-02-01

The main objective of this thesis is to provide a better comprehension of the phenomena involved in the transport of fission products, from the fuel rod to the coolant of a PWR reactor. To achieve this purpose, several steps were followed. Firstly, it was presented a description of the fuel elements and the main mechanisms of fuel rod failure, indicating the most important nuclides and their transport mechanisms. Secondly, taking both the kinetic and diffusion models for the transport of fission products as a basis, a simple analytical and semi-empirical model was developed. This model was also based on theoretical considerations and measurements of coolant's activity, according to internationally adopted methodologies. Several factors are considered in the modelling procedures: intrinsic factors to the reactor itself, factors which depend on the reactor's operational mode, isotope characteristic factors, and factors which depend on the type of rod failure. The model was applied for different reactor's operational parameters in the presence of failed rods. The main conclusions drawn from the analysis of the model's output are relative to the variation on the coolant's water activity with the fuel burnup, the linear operation power and the primary purification rate and to the different behaviour of iodine and noble gases. The model was saturated from a certain failure size and showed to be unable to distinguish between a single big fail and many small ones. (author)

In-pile data analysis of the comparative WWER/PWR test IFA-503.1. Final report.

Energy Technology Data Exchange (ETDEWEB)

Volkov, B.; Devold, H.; Ryazantzev, E.; Yakovlev, V.

1999-04-15

The comparative WWER/PWR test in IFA-503.1 was commenced in July 1995 and successfully finished at the end of November 1998. The main objective of the test was generation of representative and comparative data of standard WWER-440 fuel fabricated at the 'MSZ' Electrostal (Russia) and PWR type fuel manufactured at IFE Kjeller (Norway). The test assembly comprised two clusters, each with 3 WWER rods and 3 PWR type rods. Eight rods with two types of fuel were instrumented with expansion thermometers, four rods were equipped with both fuel stack elongation detectors and pressure transducers. All sensors worked satisfactorily during the test. The average burnups achieved in the lower and upper clusters were around 25 and 20 MWd/kgUO{sub 2}, respectively. Some difference in densification of the two types of fuel was revealed during the first irradiation period. However, the fuel temperatures and commencement of fuel stack swelling were similar despite this fact. At the end of the test the rig was moved to a higher flux position in the HBWR core with the aim of promoting FGR and to compare the behaviour of the two types of fuel under higher power. Pressure measurements indicated a comparable low FGR (around 1 percent) in both types of rods. The centreline temperatures measured in the PWR rods were very close to the Halden FGR threshold whilst the WWER fuel temperatures were slightly lower. Despite the differences found in the behaviour of the two types of fuel during the test, the analysis of the in-pile data showed that these differences would not affect the fuel efficiency, at least, up to the burnup achieved in the test. It is supposed that these differences can be related to the fuel microstructure, in particular to the fuel grain and pore sizes (author) (ml)

In-pile data analysis of the comparative WWER/PWR test IFA-503.1. Final report

Energy Technology Data Exchange (ETDEWEB)

Volkov, B.; Devold, H.; Ryazantzev, E.; Yakovlev, V

1999-04-15

The comparative WWER/PWR test in IFA-503.1 was commenced in July 1995 and successfully finished at the end of November 1998. The main objective of the test was generation of representative and comparative data of standard WWER-440 fuel fabricated at the 'MSZ' Electrostal (Russia) and PWR type fuel manufactured at IFE Kjeller (Norway). The test assembly comprised two clusters, each with 3 WWER rods and 3 PWR type rods. Eight rods with two types of fuel were instrumented with expansion thermometers, four rods were equipped with both fuel stack elongation detectors and pressure transducers. All sensors worked satisfactorily during the test. The average burnups achieved in the lower and upper clusters were around 25 and 20 MWd/kgUO{sub 2}, respectively. Some difference in densification of the two types of fuel was revealed during the first irradiation period. However, the fuel temperatures and commencement of fuel stack swelling were similar despite this fact. At the end of the test the rig was moved to a higher flux position in the HBWR core with the aim of promoting FGR and to compare the behaviour of the two types of fuel under higher power. Pressure measurements indicated a comparable low FGR (around 1 percent) in both types of rods. The centreline temperatures measured in the PWR rods were very close to the Halden FGR threshold whilst the WWER fuel temperatures were slightly lower. Despite the differences found in the behaviour of the two types of fuel during the test, the analysis of the in-pile data showed that these differences would not affect the fuel efficiency, at least, up to the burnup achieved in the test. It is supposed that these differences can be related to the fuel microstructure, in particular to the fuel grain and pore sizes (author) (ml)

An intelligent pedagogic tool for teaching the operators of PWR type reactors

International Nuclear Information System (INIS)

Cordier, B.; Guillermard, M.

1990-01-01

A tool was developed for assisting the instruction of the operators of a PWR type nuclear power plant. For achieving the objectives, an expert system and a simulator were combined. The main objective of the system is to improve the work of the operators in performing remedial actions in case of accident. The simulator applies two IBM PC AT3 and a MC 680 20 microprocessor. The use and the validation of the expert system are presented. The perspectives for the system, implanted on the Tricastin nuclear power plant, are analyzed [fr

Layout of the primary circuit with its components for PWR and BWR

International Nuclear Information System (INIS)

Meyer, P.J.

1981-01-01

The light water-moderated and cooled pressurized water reactors and boiling water reactors constitute the basis of economic utilization of nuclear energy all over the world. Pressurized water reactors up to capacities of 3,800 MWth are those most used for power generation. However, their potential capacities exceed 3,800 MWth, so that already in the near future PWR are conseivable which readily generate 1,500 to 2,000 MWe. The main problem for starting the next generation of PWRs are of safety measure and licensing questions. Interesting applications of the PWRs are nuclear district heating, generation of process steam of desalination plants, steam injection into the ground for oil production or chemical factories. A new generation of natural circulation boiling water reactors with a capacity of 200 to 400 MW will be used for development of small industrial areas or for countries without an integral grid system. The natural circulation boiling water reactor will be subject of a separate lecture. Due to the fact of the majority of the PWR all over the world this lecture will discuss mainly PWR design aspects. (orig./RW)

Proposal for a advanced PWR core with adequate characteristics for passive safety concept

International Nuclear Information System (INIS)

Perrotta, Jose Augusto

1999-01-01

This work presents a discussion upon the suitable from an advanced PWR core, classified by the EPRI as 'Passive PWR' (advanced reactor with passive safety concept to power plants with less than 600 MW electrical power). The discussion upon the type of core is based on nuclear fuel engineering concepts. Discussion is made on type of fuel materials, structural materials, geometric shapes and manufacturing process that are suitable to produce fuel assemblies which give good performance for this type of reactors. The analysis is guided by the EPRI requirements for Advanced Light Water Reactor (ALWR). By means of comparison, the analysis were done to Angra 1 (old type of 600 MWe PWR class), and the design of the Westinghouse Advanced PWR-AP600. It was verified as a conclusion of this work that the modern PWR fuels are suitable for advanced PWR's Nevertheless, this work presents a technical alternative to this kind of fuel, still using UO 2 as fuel, but changing its cylindrical form of pellets and pin type fuel element to plane shape pallets and plate type fuel element. This is not a novelty fuel, since it was used in the 50's at Shippingport Reactor and as an advanced version by CEA of France in the 70's. In this work it is proposed a new mechanical assembly design for this fuel, which can give adequate safety and operational performance to the core of a 'Passive PWR'. (author)

The Impact of Business Size and Business Type on Small Business Investment in Electronic Commerce: a study of Swedish small businesses

Directory of Open Access Journals (Sweden)

Robert MacGregor

2002-05-01

Full Text Available In the past, organisations relied on traditional quantitative metrics, such as Return on Investment (ROI to make decisions when investing in technology. With the advent of electronic commerce (EC, organisations have had to rethink their investment and acquisition decisions due to the strategic nature of electronic commerce. Where ROI measures have failed, they have been replaced with a plethora of organisational driving forces. This paper focuses on the driving forces behind EC adoption by small and medium enterprises (SME's and aims to determine the impact of organisational factors such as size and type of business on EC acquisition criteria. The results of a research study carried out in Sweden are presented and suggest that there exist high levels of significance between the size of the business and customer demand, reduced costs, developing new markets and improvement to marketing as driving forces, and the type of business and customer demand, pressure from competition, increased sales and improvement of relationship with business partners as driving forces for EC adoption.

RNL NDT studies related to PWR pressure vessel inlet nozzle inspection

International Nuclear Information System (INIS)

Rogerson, A.; Poulter, L.N.J.; Clough, P.; Cooper, A.

1984-01-01

Non-destructive examinations of the Reactor Pressure Vessel (RPV) of a Pressurized Water Reactor (PWR) play an important role in assuring vessel integrity throughout its operational life. Automated ultrasonic techniques for the detection and sizing of flaws in thick-section seam welds and near-surface regions in a PWR RPV have been under development at RNL for some time. Techniques for the inspection of complex geometry welds and other regions of the vessel are now being assessed and further developed as part of the UK NDT development programme in support of the Sizewell PWR. One objective of this programme is to demonstrate that the range of ultrasonic techniques already shown to be effective for the inspection of seam welds and inlet nozzle corner regions, through exercises such as the Defect Detection Trials, can also be effective for inspection of these other vessel regions. The nozzle-to-vessel welds and nozzle crotch corners associated with the RPV water inlet and outlet nozzles are two such regions being examined in this programme. In this paper, a review is given of the work performed at RNL in the development of a laboratory-based inspection system for inlet nozzle inspection. The main features of the system in its current stage of development are explained. (author)

The AMEBA PWR, a new concept in the technology of nuclear reactor safety

Energy Technology Data Exchange (ETDEWEB)

Novelli, A

2000-05-01

AMEBA is an Italian acronym which stands for 'alta moderazione e basso arricchimento' (high moderation and low enrichment). The AMEBA reactor is nothing more than a PWR which possesses very unusual values of both volumetric ratio moderator/fuel and U-235 enrichment of UO{sub 2}. The possibility is shown of the technical realisation of a nuclear power plant equipped with an AMEBA PWR reactor. Among the most enticing properties of AMEBA are the following: self-shut-down in any abnormal condition, elimination of all need for control rods and boric acid dissolution in the water, absolute impossibility of reaching values of reactivity greater than a fraction of a dollar, intrinsic subcriticality, attaining to several dollars, in non-operative condition when the water is at ambient temperature, normal operation with a very small-sized pressurizer, self-start-up.

The AMEBA PWR, a new concept in the technology of nuclear reactor safety

International Nuclear Information System (INIS)

Novelli, A.

2000-01-01

AMEBA is an Italian acronym which stands for 'alta moderazione e basso arricchimento' (high moderation and low enrichment). The AMEBA reactor is nothing more than a PWR which possesses very unusual values of both volumetric ratio moderator/fuel and U-235 enrichment of UO 2 . The possibility is shown of the technical realisation of a nuclear power plant equipped with an AMEBA PWR reactor. Among the most enticing properties of AMEBA are the following: self-shut-down in any abnormal condition, elimination of all need for control rods and boric acid dissolution in the water, absolute impossibility of reaching values of reactivity greater than a fraction of a dollar, intrinsic subcriticality, attaining to several dollars, in non-operative condition when the water is at ambient temperature, normal operation with a very small-sized pressurizer, self-start-up

Method of starting up PWR type reactor

International Nuclear Information System (INIS)

Kadokami, Akira; Ueno, Ryuji; Tsuge, Ayao; Onimura, Kichiro; Ochi, Tatsuya.

1988-01-01

Purpose: To start-up a PWR type reactor so as to effectively impregnate and concentrate corrosion inhibitors in intergranular corrosive faces. Method: Upon reactor start-up, after transferring from the warm zero output state to thermal power loaded state and injecting corrosion inhibitors, thermal power is returned to zero and, subsequently, increased up to a rated power. By selecting the thermal power upon injecting the corrosion inhibitors to a steam generator body, that is, by selecting a thermal power load that starts to boil in heat conduction tubes, feedwater in the clavis portion can be formed into an appropriate boiling convection and, accordingly, the corrosion inhibitors can be penetrated to the clevis portion at a higher rate and in a greater amount as compared with those under zero power condition. Subsequently, when the thermal power is reduced, a sub-cooled state is attained in the clevis portion, in which steams present in the intergranular corrosion faces in the heat conduction tubes are condensated. As a result, the corrosion inhibitors at high concentration are impregnated into the intergranular corrosive faces to provide excellent effects. (Kamimura, M.)

Development of an engineering simulator for integral type PWR for nuclear ship

International Nuclear Information System (INIS)

Takahashi, Teruo; Shimazaki, Junya; Nakazawa, Toshio

2000-01-01

JAERI has developed a real-time engineering simulator for the integral type reactor MRX (Marine Reactor X) of power 100 MWt to evaluate the design and operational performance and to study highly automatic operations of a reactor plant. Marine reactor is operated under the conditions of pitching and rolling and load change, in comparison with a reactor for a land-based generating plant. And the MRX has systems with structural features, such as water-filled containment vessel, once-through type steam generator and emergency decay heat removal system. Considerations are paid to take these operational conditions and structural features into the simulation model. It is shown that the simulated results are consistent with the planned design and operational performance, and on the other hand present us some technical issues to be investigated in the design specifications. (author)

Competitiveness basis of integrated PWR technology

International Nuclear Information System (INIS)

Florido, Pablo C.; Bergallo, Juan E.

1999-01-01

The present nuclear power market is dominated by LWR, with a minor fraction of CANDU contribution due to its advantages in small and medium electric grid. Nowadays the demand for nuclear reactors is concentrated in Asia and this situation will probably not change in the short and medium term, even under the financial crisis. The Asian market needs large plants, and feedback regarding demand and products has been done (AP 600 and Sbwr) in studies of passive systems to higher power range. Present active nuclear power programs take place in countries with high GNP growth and related electricity energy demand far from expectancies in the '80. Four of these countries were in the list of the top 10 most dynamical nations. Traditionally, nuclear countries have large economies and electric grids or strong national nuclear commitment. However, in the near future the demand of nuclear reactors seems to move towards faster growing economies and countries, with new electrical grids. From the point of view of the competitiveness, higher power nuclear plants are seriously questioned in this new market. The International Atomic Energy Agency has promoted studies in the field of small nuclear power plants (around 300 M We) for a long time. These designs have never reached commercial level in industrialized countries. One reason has been that the reactor technology naturally moves towards the best economy, i.e. the greater power. The second reason has been that the nuclear designers were always in countries with large electric systems. Consequently the utilities never have been attracted by prototypes of small reactors. The only option to develop a competitive medium nuclear reactors is to develop a design with different scale economy, like the scale economy of CCGT. If LWR technology is used, the design could be the best alternative for the utilities, because the availability of present LWR is one of the main advantages as a proved power source. The most promising type of LWR with

Scaling studies - PWR

International Nuclear Information System (INIS)

Sonneck, G.

1983-05-01

A RELAP 4/MOD 6 study was made based on the blowdown phase of the intermediate break experiment LOFT L5-1. The method was to set up a base model and to vary parametrically some areas where it is known or suspected that LOFT differs from a commercial PWR. The aim was not to simulate LOFT or a PWR exactly but to understand the influence of the following parameters on the thermohydraulic behaviour of the system and the clad temperature: stored heat in the downcomer (LOFT has rather large filler blocks in this part of the pressure vessel); bypass between downcomer and upper plenum; and core length. The results show that LOFT is prototypical for all calculated blowdowns. As the clad temperatures decrease with decreasing stored energy in the downcomer, increased bypass and increased core length, LOFT results seem to be realistic as long as realistic bypass sizes are considered; they are conservative in the two other areas. (author)

Integrated training support system for PWR operator training simulator

International Nuclear Information System (INIS)

Sakaguchi, Junichi; Komatsu, Yasuki

1999-01-01

The importance of operator training using operator training simulator has been recognized intensively. Since 1986, we have been developing and providing many PWR simulators in Japan. We also have developed some training support systems connected with the simulator and the integrated training support system to improve training effect and to reduce instructor's workload. This paper describes the concept and the effect of the integrated training support system and of the following sub-systems. We have PES (Performance Enhancement System) that evaluates training performance automatically by analyzing many plant parameters and operation data. It can reduce the deviation of training performance evaluation between instructors. PEL (Parameter and Event data Logging system), that is the subset of PES, has some data-logging functions. And we also have TPES (Team Performance Enhancement System) that is used aiming to improve trainees' ability for communication between operators. Trainee can have conversation with virtual trainees that TPES plays automatically. After that, TPES automatically display some advice to be improved. RVD (Reactor coolant system Visual Display) displays the distributed hydraulic-thermal condition of the reactor coolant system in real-time graphically. It can make trainees understand the inside plant condition in more detail. These sub-systems have been used in a training center and have contributed the improvement of operator training and have gained in popularity. (author)

Size Exclusion HPLC Detection of Small-Size Impurities as a Complementary Means for Quality Analysis of Extracellular Vesicles

Directory of Open Access Journals (Sweden)

Tao Huang

2015-07-01

Full Text Available For extracellular vesicle research, whether for biomarker discoveries or therapeutic applications, it is critical to have high-quality samples. Both microscopy and NanoSight Tracking Analysis (NTA for size distribution have been used to detect large vesicles. However, there is currently no well-established method that is convenient for routine quality analysis of small-size impurities in vesicle samples. In this paper we report a convenient method, called ‘size-exclusion high-performance liquid chromatography’ (SE-HPLC, alongside NTA and Microscopy analysis to guide and qualify the isolation and processing of vesicles. First, the SE-HPLC analysis was used to detect impurities of small-size proteins during the ultra-centrifugation process of vesicle isolation; it was then employed to test the changes of vesicles under different pH conditions or integrity after storage. As SE-HPLC is generally accessible in most institutions, it could be used as a routine means to assist researchers in examining the integrity and quality of extracellular vesicles along with other techniques either during isolation/preparation or for further engineering and storage.

PWR-to-PWR fuel cycle model using dry process

International Nuclear Information System (INIS)

Iqbal, M.; Jeong, Chang Joon; Rho, Gyu Hong

2002-03-01

PWR-to-PWR fuel cycle model has been developed to recycle the spent fuel using the dry fabrication process. Two types of fuels were considered; first fuel was based on low initial enrichment with low discharge burnup and second one was based on more initial enrichment with high discharge burnup in PWR. For recycling calculations, the HELIOS code was used, in which all of the available fission products were considered. The decay of 10 years was applied for reuse of the spent fuel. Sensitivity analysis for the fresh feed material enrichment has also been carried out. If enrichment of the mixing material is increased the saving of uranium reserves would be decreased. The uranium saving of low burned fuel increased from 4.2% to 7.4% in fifth recycling step for 5 wt% to 19.00wt% mixing material enrichment. While for high burned fuel, there was no uranium saving, which implies that higher uranium enrichment required than 5 wt%. For mixing of 15 wt% enriched fuel, the required mixing is about 21.0% and 37.0% of total fuel volume for low and high burned fuel, respectively. With multiple recycling, reductions in waste for low and high burned fuel became 80% and 60%, for first recycling, respectively. In this way, waste can be reduced more and the cost of the waste disposal reduction can provide the economic balance

In-pile post-DNB behavior of a nine-rod PWR-type fuel bundle

International Nuclear Information System (INIS)

Gunnerson, F.S.; MacDonald, P.E.

1980-01-01

The results of an in-pile power-cooling-mismatch (PCM) test designed to investigate the behavior of a nine-rod, PWR-type fuel bundle under intermittent and sustained periods of high temperature film boiling operation are presented. Primary emphasis is placed on the DNB and post-DNB events including rod-to-rod interactions, return to nucleate boiling (RNB), and fuel rod failure. A comparison of the DNB behavior of the individual bundle rods with single-rod data obtained from previous PCM tests is also made

Fuel assembly for pressure loss variable PWR type reactor

International Nuclear Information System (INIS)

Yoshikuni, Masaaki.

1993-01-01

In a PWR type reactor, a pressure loss control plate is attached detachably to a securing screw holes on the lower surface of a lower nozzle to reduce a water channel cross section and increase a pressure loss. If a fuel assembly attached with the pressure loss control plate is disposed at a periphery of the reactor core where the power is low and heat removal causes no significant problem, a flowrate at the periphery of the reactor core is reduced. Since this flowrate is utilized for removal of heat from fuel assemblies of high powder at the center of the reactor core where a pressure loss control plate is not attached, a thermal limit margin of the whole reactor core is increased. Thus, a limit of power peaking can be moderated, to obtain a fuel loading pattern improved with neutron economy. (N.H.)

Preventive testing and leakage detection in pipe-lines of steam condensers and generators of a PWR type reactor

International Nuclear Information System (INIS)

Canalini, A.; Carvalho, N.C. de

1985-01-01

The non-destructive methods: Spum, Helium and Hydrostatic used in leakage detection in condenser pipelines for PWR type reactors are presented. The time, costs, sensitivity, resources necessary and personnel development factors are considered to choose adequated method, in function of nuclear power plant conditions. The leakage tests are applied in pressurized systems or vacuum. Eddy Current testing is used in condensers and steam generators aiming to avoid leakage in these equipments. The spume testing for leakage detection in condenser pipelines - which operation - and hydrostatic testing for leakage detection through reaming with shutdown - were most efficients. The Helium testing applied in pressurized systems or submitted to vacuum systems presented satisfactory results. The Eddy Current testing in condenser and steam generator pipelines reached desired objective, reducing leakage in the first and preserving the integrity in the second. (M.C.K.) [pt

Small-sized test of gravity separation and preliminary assessment of technology and economics in Guangshigou granite pegmatite type uranium deposit

Energy Technology Data Exchange (ETDEWEB)

Zhifu, Sun; Mingyue, Feng; Jiashu, Rong; Ziyang, Xu [Beijing Research Inst. of Uranium Geology (China)

1994-11-01

The small-sized test of gravity separation in Guangshigou granite pegmatite type uranium deposit has found a new avenue for the industrial utilization of ores from such uranium deposit, especially those low grade ones. The test has proved that the gravity separation is superior to hydrometallurgy in the aspect of uranium recovery from ores of the granite pegmatite type uranium deposit, by-product recovery and protection against environmental pollution.

Small-sized test of gravity separation and preliminary assessment of technology and economics in Guangshigou granite pegmatite type uranium deposit

International Nuclear Information System (INIS)

Sun Zhifu; Feng Mingyue; Rong Jiashu; Xu Ziyang

1994-01-01

The small-sized test of gravity separation in Guangshigou granite pegmatite type uranium deposit has found a new avenue for the industrial utilization of ores from such uranium deposit, especially those low grade ones. The test has proved that the gravity separation is superior to hydrometallurgy in the aspect of uranium recovery from ores of the granite pegmatite type uranium deposit, by-product recovery and protection against environmental pollution

Development of Cost Estimation Methodology of Decommissioning for PWR

International Nuclear Information System (INIS)

Lee, Sang Il; Yoo, Yeon Jae; Lim, Yong Kyu; Chang, Hyeon Sik; Song, Geun Ho

2013-01-01

The permanent closure of nuclear power plant should be conducted with the strict laws and the profound planning including the cost and schedule estimation because the plant is very contaminated with the radioactivity. In Korea, there are two types of the nuclear power plant. One is the pressurized light water reactor (PWR) and the other is the pressurized heavy water reactor (PHWR) called as CANDU reactor. Also, the 50% of the operating nuclear power plant in Korea is the PWRs which were originally designed by CE (Combustion Engineering). There have been experiences about the decommissioning of Westinghouse type PWR, but are few experiences on that of CE type PWR. Therefore, the purpose of this paper is to develop the cost estimation methodology and evaluate technical level of decommissioning for the application to CE type PWR based on the system engineering technology. The aim of present study is to develop the cost estimation methodology of decommissioning for application to PWR. Through the study, the following conclusions are obtained: · Based on the system engineering, the decommissioning work can be classified as Set, Subset, Task, Subtask and Work cost units. · The Set and Task structure are grouped as 29 Sets and 15 Task s, respectively. · The final result shows the cost and project schedule for the project control and risk management. · The present results are preliminary and should be refined and improved based on the modeling and cost data reflecting available technology and current costs like labor and waste data

The Role of Branding in Small and Medium-Sized Enterprises

Directory of Open Access Journals (Sweden)

Franc Vidic

2013-12-01

Full Text Available The purpose of this article is to show the relationship between branding and brand management in small and medium-sized enterprises (SMEs. Traditionally, branding was associated with large and global corporations. However, we often forget that small and medium-sized enterprises also deal with their own names (brands in their own way. The study identified four types of businesses, regardless of their association with brands. We named these four types, as follows: Ignorant; User; Low-Cost Producer; and Differentiation Producer. If the first two types (i.e. Ignorant and User differ primarily in the extent to which they use simple branding activities, and are used mainly in the local market where the enterprises tend to operate, we found that the last two types (i.e. low-cost producers and differentiation producers design their branding strategies in accordance with their generic strategies and mode of growth.

A neural networks based ``trip`` analysis system for PWR-type reactors; Um sistema de analise de ``trip`` em reatores PWR usando redes neuronais

Energy Technology Data Exchange (ETDEWEB)

Alves, Antonio Carlos Pinto Dias

1993-12-31

The analysis short after automatic shutdown (trip) of a PWR-type nuclear reactor takes a considerable amount of time, not only because of the great number of variables involved in transients, but also the various equipment that compose a reactor of this kind. On the other hand, the transients`inter-relationship, intended to the detection of the type of the accident is an arduous task, since some of these accidents (like loss of FEEDWATER and station BLACKOUT, for example), generate transients similar in behavior (as cold leg temperature and steam generators mixture levels, for example). Also, the sequence-of-events analysis is not always sufficient for correctly pin point the causes of the trip. (author) 11 refs., 39 figs.

A neural networks based ``trip`` analysis system for PWR-type reactors; Um sistema de analise de ``trip`` em reatores PWR usando redes neuronais

Energy Technology Data Exchange (ETDEWEB)

Alves, Antonio Carlos Pinto Dias

1994-12-31

The analysis short after automatic shutdown (trip) of a PWR-type nuclear reactor takes a considerable amount of time, not only because of the great number of variables involved in transients, but also the various equipment that compose a reactor of this kind. On the other hand, the transients`inter-relationship, intended to the detection of the type of the accident is an arduous task, since some of these accidents (like loss of FEEDWATER and station BLACKOUT, for example), generate transients similar in behavior (as cold leg temperature and steam generators mixture levels, for example). Also, the sequence-of-events analysis is not always sufficient for correctly pin point the causes of the trip. (author) 11 refs., 39 figs.

Integrity of PWR pressure vessels during overcooling accidents

International Nuclear Information System (INIS)

Cheverton, R.D.; Iskander, S.K.; Whitman, G.D.

1982-01-01

The reactor pressure vessel in a pressurized water reactor is normally subjected to temperatures and pressures that preclude propagation of sharp, crack-like defects that might exist in the wall of the vessel. However, there is a class of postulated accidents, referred to as overcooling accidents, that can subject the pressure vessel to severe thermal shock while the pressure is substantial. As a result of such accidents vessels containing high concentrations of copper and nickel, which enhance radiation embrittlement, may possess a potential for extensive propagation of preexistent inner surface flaws prior to the vessel's normal end of life. For the purpose of evaluating this problem a state-of-the-art fracture mechanics model was developed and has been used for conducting parametric analyses and for calculating several recorded PWR transients. Results of the latter analysis indicate that there may be some vessels that have a potential for failure today if subjected to a Rancho Seco (1978) or TMI-2 (1979) type transient. However, the calculational model may be excessively conservative, and this possibility is under investigation

Integrity of PWR pressure vessels during overcooling accidents

International Nuclear Information System (INIS)

Cheverton, R.D.; Iskander, S.K.; Whitman, G.D.

1982-01-01

The reactor pressure vessel in a pressurized water reactor is normally subjected to temperatures and pressures that preclude propagation of sharp, crack-like defects that might exist in the wall of the vessel. However, there is a class of postulated accidents, referred to as overcooling accidents, that can subject the pressure vessel to severe thermal shock while the pressure is substantial. As a result of such accidents, vessels containing high concentrations of copper and nickel, which enhance radiation embrittlement, may possess a potential for extensive propagation of preexistent inner surface flaws prior to the vessel's normal end of life. A state-of-the-art fracture-mechanics model was developed and has been used for conducting parametric analyses and for calculating several recorded PWR transients. Results of the latter analysis indicate that there may be some vessels that have a potential for failure in a few years if subjected to a Rancho Seco-type transient. However, the calculational model may be excessively conservative, and this possibility is under investigation

The PWR integrated leak rate test, a review of experiences and results

International Nuclear Information System (INIS)

Keogh, P.

1985-01-01

The paper reviews the Integrated Leak Rate Test (ILRT) as it has been carried out in the USA and as reported in papers in European countries. The test procedures are critically appraised and recommendations are given for modifications to them. The values used in a PWR are identified as a main source of leaks and possibilities for improvement are discussed. The use of a part pressure test and its limitations are considered. A part pressure test cannot give the same assurance as a full pressure test but may be useful for the identification of gross leaks. Secondary effects such as weather and the use of Van de Waals equations are considered and are found to be not important for concrete containments. (orig.)

Use of sales and operations planning in small and medium-sized enterprises

Directory of Open Access Journals (Sweden)

Michał Adamczak

2013-03-01

Full Text Available Background: Increasing competitiveness in the market, customer expectations related to the shortening of the deadlines and the reduction of prices of products and services force companies to improve the efficiency of internal processes. The integration of planning process is one of possible ways to achieve this aim. The integration of planning processes by the use of SOP model (Sales and Operations Planning is a method to implement this idea. The study allowed to identify ways to implement the process of sales and operations planning in small and medium-sized enterprises. Material and methods: The study was conducted in companies from different industries. The research method was in-depth interviews conducted with managers of companies or persons occupying management positions in the organizational process of implementing sales and operations planning. Results: During the survey, 10 companies were asked about the use of sales and operations planning, its elements and organizational aspects of its development, by the company. Conclusions: The use of sales and operations plan is closely dependent on the size of the company and its localization in the supply chain. Small enterprises are not interested in the integration of the planning process due to the small scale of operations and the centralization of decision-making process. Medium-sized enterprises, due to the increased complexity of the processes of planning, see the benefits of their integration in the SOP model.

Quenching effect in an optical fiber type small size dosimeter irradiated with 290 MeV·u{sup -1} carbon ions

Energy Technology Data Exchange (ETDEWEB)

Hirata, Yuho; Watanabe, Kenichi; Uritani, Akira; Yamazaki, Atsushi [Graduate School of Engineering, Nagoya University, Nagoya (Japan); Koba, Yusuke; Matsufuji, Naruhiro [National Institute of Radiological Sciences, Chiba (Japan)

2016-09-15

We are developing a small size dosimeter for dose estimation in particle therapies. The developed dosimeter is an optical fiber based dosimeter mounting an radiation induced luminescence material, such as an OSL or a scintillator, at a tip. These materials generally suffer from the quenching effect under high LET particle irradiation. We fabricated two types of the small size dosimeters. They used an OSL material Eu:BaFBr and a BGO scintillator. Carbon ions were irradiated into the fabricated dosimeters at Heavy Ion Medical Accelerator in Chiba (HIMAC). The small size dosimeters were set behind the water equivalent acrylic phantom. Bragg peak was observed by changing the phantom thickness. An ion chamber was also placed near the small size dosimeters as a reference. Eu:BaFBr and BGO dosimeters showed a Bragg peak at the same thickness as the ion chamber. Under high LET particle irradiation, the response of the luminescence-based small size dosimeters deteriorated compared with that of the ion chamber due to the quenching effect. We confirmed the luminescence efficiency of Eu:BaFBr and BGO decrease with the LET. The reduction coefficient of luminescence efficiency was different between the BGO and the Eu:BaFBr. The LET can be determined from the luminescence ratio between Eu:BaFBr and BGO, and the dosimeter response can be corrected. We evaluated the LET dependence of the luminescence efficiency of the BGO and Eu:BaFBr as the quenching effect. We propose and discuss the correction of the quenching effect using the signal intensity ratio of the both materials. Although the correction precision is not sufficient, feasibility of the proposed correction method is proved through basic experiments.

PWR and WWER fuel performance. A comparison of major characteristics

International Nuclear Information System (INIS)

Weidinger, H.

2006-01-01

PWR and WWER fuel technologies have the same basic performance targets: most effective use of the energy stored in the fuel and highest possible reliability. Both fuel technologies use basically the same strategies to reach these targets: 1) Optimized reload strategies; 2) Maximal use of structural material with low neutron cross sections; 3) Decrease the fuel failure frequency towards a 'zero failure' performance by understanding and eliminating the root causes of those defects. The key driving force of the technology of both, PWR and WWER fuel is high burn-up. Presently a range of 45 - 50 MWD/kgU have been reached commercially for PWR and WWER fuel. The main technical limitations to reach high burn-up are typically different for PWR and WWER fuel: for PWR fuel it is the corrosion and hydrogen uptake of the Zr-based materials; for WWER fuel it is the mechanical and dimensional stability of the FA (and the whole core). Corrosion and hydrogen uptake of Zr-materials is a 'non-problem' for WWER fuel. Other performance criteria that are important for high burn-up are the creep and growth behaviour of the Zr materials and the fission gas release in the fuel rod. There exists a good and broad data base to model and design both fuel types. FA and fuel rod vibration appears to be a generic problem for both fuel types but with more evidence for PWR fuel performance reliability. Grid-to-rod fretting is still a major issue in the fuel failure statistics of PWR fuel. Fuel rod cladding defects by debris fretting is no longer a key problem for PWR fuel, while it still appears to be a significant root cause for WWER fuel failures. 'Zero defect' fuel performance is achievable with a high probability, as statistics for US PWR and WWER-1000 fuel has shown

Experimental study on secondary depressurization action for PWR vessel bottom small break LOCA with HPI failure and gas inflow (ROSA-V/LSTF test SB-PV-03)

International Nuclear Information System (INIS)

Suzuki, Mitsuhiro; Takeda, Takeshi; Asaka, Hideaki; Nakamura, Hideo

2005-06-01

A small break loss-of-coolant accident (SBLOCA) experiment was conducted at the Large Scale Test Facility (LSTF) of ROSA-V program to study effects of accident management (AM) measures on core cooling, which is important in case of high pressure injection (HPI) system failure during an SBLOCA at a pressurized water reactor (PWR). The LSTF is a full-height and 1/48 volume-scaled facility simulating 4-loop Westinghouse-type PWR (3423 MWt). The experiment, SB-PV-03, simulated a PWR vessel bottom SBLOCA with a rupture of ten instrument-tubes which is equivalent to 0.2% cold leg break. Total HPI failure, non-condensable gas inflow from accumulator injection system (AIS) and operator AM actions on steam generator (SG) secondary depressurization at a rate of -55 K/h and auxiliary feedwater (AFW) supply for 30 minutes were assumed as experiment conditions. It is clarified that the AM actions are effective on primary system depressurization until the end of AIS injection at 1.6 MPa, but thereafter become less effective due to inflow of the non-condensable gas, resulting in delay of low pressure injection (LPI) actuation and whole core heatup under continuous water discharge through the bottom break. The report describes these thermohydraulic phenomena related with transient primary coolant mass and AM actions in addition to estimation of non-condensable gas behavior which affected primary-to-secondary heat transfer. (author)

Small head size after atomic irradiation

International Nuclear Information System (INIS)

Miller, R.W.; Mulvihill, J.J.

1975-01-01

A study of children exposed to nuclear explosions in Hiroshima and Nagasaki showed small head size and mental retardation when exposure occurred less than 18 weeks of gestational age. Increased frequency of small head size occurred when maternal exposure was 10 to 19 rad. Tables and graphs are presented to show relationships between dose, gestational age, and frequency of small head size

Small PWR 'PFPWR50' using cermet fuel of Th-Pu particles

International Nuclear Information System (INIS)

Hirayama, Takashi; Shimazu, Yoichiro

2009-01-01

An innovative concept of PFPWR50 has been studied. The main feature of PFPWR50 has been to adopt TRISO coated fuel particles in a conventional PWR cladding. Coated fuel particle provides good confining ability of fission products. But it is pointed out that swelling of SiC layer at low temperature by irradiation has possibilities of degrading the integrity of coated fuel particle in the LWR environment. Thus, we examined the use of Cermet fuel replacing SiC layer to Zr metal or Zr compound. And the nuclear fuel has been used as fuel compact, which is configured to fix coated fuel particles in the matrix material to the shape of fuel pellet. In the previous study, graphite matrix is adopted as the matrix material. According to the burnup calculations of the several fuel concepts with those covering layers, we decide to use Zr layer embedded in Zr metal base or ZrC layer with graphite matrix. But carbon has the problem at low temperature by irradiation as well as SiC. Therefore, Zr covering layer and Zr metal base are finally selected. The other feature of PFPWR50 concept has been that the excess reactivity is suppressed during a cycle by initially loading burnable poison (gadolinia) in the fuels. In this study, a new loading pattern is determined by combining 7 types of assemblies in which the gadolinia concentration and the number of the fuel rods with gadolinia are different. This new core gives 6.7 equivalent full power years (EFPY) as the core life of a cycle. And the excess reactivity is suppressed to less than 2.0%Δk/k during the cycle. (author)

Automatic welding processes for reactor coolant pipes used in PWR type nuclear power plant

International Nuclear Information System (INIS)

Hamada, T.; Nakamura, A.; Nagura, Y.; Sakamoto, N.

1979-01-01

The authors developed automatic welding processes (submerged arc welding process and TIG welding process) for application to the welding of reactor coolant pipes which constitute the most important part of the PWR type nuclear power plant. Submerged arc welding process is suitable for flat position welding in which pipes can be rotated, while TIG welding process is suitable for all position welding. This paper gives an outline of the two processes and the results of tests performed using these processes. (author)

Experiments for simulating a great leak in the primary coolant circuit of a PWR type reactor

International Nuclear Information System (INIS)

Liebig, E.

1977-01-01

A loss of coolant accident is to be simulated on a high pressure test rig. The accident is initiated by an externally induced rupture of a pair of rupture-disks installed in a coolant ejection device. Several problems of simulating leaks in the primary coolant circuit of PWR type reactors are dealt with. The selection of appropriate rupture-disks for such experiments is described

Application of the integrated analysis of safety (ISA) to sequences of Total loss of feed water in a PWR Reactor

International Nuclear Information System (INIS)

Moreno Chamorro, P.; Gallego Diaz, C.

2011-01-01

The main objective of this work is to show the current status of the implementation of integrated analysis of safety (ISA) methodology and its SCAIS associated tool (system of simulation codes for ISA) to the sequence analysis of total loss of feedwater in a PWR reactor model Westinghouse of three loops with large, dry containment.

Radionuclide compositions of spent fuel and high level waste for the uranium and plutonium fuelled PWR

International Nuclear Information System (INIS)

Fairclough, M.P.; Tymons, B.J.

1985-06-01

The activities of a selection of radionuclides are presented for three types of reactor fuel of interest in radioactive waste management. The fuel types are for a uranium 'burning' PWR, a plutonium 'burning' PWR using plutonium recycled from spent uranium fuel and a plutonium 'burning' PWR using plutonium which has undergone multiple recycle. (author)

Performance of PWR study in the technology supplier countries: south korea and japan case

International Nuclear Information System (INIS)

Sriyana

2007-01-01

Electricity is needed as an infrastructure to support the national economic growth. For economic development sustainability, energy alternatives should be provided. Nuclear Power Plant (NPP) become the alternative of electricity generation for optimum energy mix in Indonesia and planned to operate in the 2016. Several studies have already done to prepare the NPP construction. This study focused on NPP performance especially PWR type in Asia, namely Japan and South Korea. Methodology used in this is literature tracing and a small calculation. The energy availability per unit per year is used as a parameter for evaluating the NPP performance. This conclusion are 1) the amount of NPP - PWR type in Japan is 22 units with total operational experiences 526 reactor-years and the average energy availability factor about 70.7% per unit per year. Meanwhile for the same type South Korea has 16 unit with total operational experience 222 reactor-years and average availability factor per unit per year is about 86.9%. 2) the 1000 class of PWR type both South Korea and Japan have 14 units. The operational experiences for thi class is 170 reactor-year for South Korean and 307 reactor-year for Japan. Meanwhile the average availability factor per unit per year is about 87.0% for South Korea and 69.6% for Japan. 3) the average availability factor is closed to capacity factor, so is important for real figure in assuming the techno-economic parameters, because it will influence the result o economic calculation. (author)

Conceptual design of simplified PWR

International Nuclear Information System (INIS)

Tabata, Hiroaki

1996-01-01

The limited availability for location of nuclear power plant in Japan makes plants with higher power ratings more desirable. Having no intention of constructing medium-sized plants as a next generation standard plant, Japanese utilities are interested in applying passive technologies to large ones. So, Japanese utilities have studied large passive plants based on AP600 and SBWR as alternative future LWRs. In a joint effort to develop a new generation nuclear power plant which is more friendly to operator and maintenance personnel and is economically competitive with alternative sources of power generation, JAPC and Japanese Utilities started the study to modify AP600 and SBWR, in order to accommodate the Japanese requirements. During a six year program up to 1994, basic concepts for 1000 MWe class Simplified PWR (SPWR) and Simplified BWR (SBWR) were developed, though there still remain several areas to be improved. These studies have now stepped into the phase of reducing construction cost and searching for maximum power rating that can be attained by reasonably practical technology. These results also suggest that it is hopeful to develop a large 3-loop passive plant (∼1200 MWe). Since Korea mainly deals with PWR, this paper summarizes SPWR study. The SPWR is jointly studied by JAPC, Japanese PWR Utilities, EdF, WH and Mitsubishi Heavy Industry. Using the AP-600 reference design as a basis, we enlarged the plant size to 3-loops and added engineering features to conform with Japanese practice and Utilities' preference. The SPWR program definitively confirmed the feasibility of a passive plant with an NSSS rating about 1000 MWe and 3 loops. (J.P.N.)

FLUOLE-2: An Experiment for PWR Pressure Vessel Surveillance

Directory of Open Access Journals (Sweden)

Thiollay Nicolas

2016-01-01

Full Text Available FLUOLE-2 is a benchmark-type experiment dedicated to 900 and 1450 MWe PWR vessels surveillance dosimetry. This two-year program started in 2014 and will end in 2015. It will provide precise experimental data for the validation of the neutron spectrum propagation calculation from core to vessel. It is composed of a square core surrounded by a stainless steel baffe and internals: PWR barrel is simulated by steel structures leading to different steel-water slides; two steel components stand for a surveillance capsule holder and for a part of the pressure vessel. Measurement locations are available on the whole experimental structure. The experimental knowledge of core sources will be obtained by integral gamma scanning measurements directly on fuel pins. Reaction rates measured by calibrated fission chambers and a large set of dosimeters will give information on the neutron energy and spatial distributions. Due to the low level neutron flux of EOLE ZPR a special, high efficiency, calibrated gamma spectrometry device will be used for some dosimeters, allowing to measure an activity as low as 7. 10−2 Bq per sample. 103mRh activities will be measured on an absolute calibrated X spectrometry device. FLUOLE-2 experiment goal is to usefully complete the current experimental benchmarks database used for the validation of neutron calculation codes. This two-year program completes the initial FLUOLE program held in 2006–2007 in a geometry representative of 1300 MWe PWR.

PWR type overpower tests at 1620 GJ/KGU (18,800 MWD/MTU)

International Nuclear Information System (INIS)

Knudsen, P.; Bagger, C.; Carlsen, H.

1979-01-01

Three PWR type test fuel pins accumulated a burnup of 1620 GJ/kgU at heat loads decreasing from 45 to 28 kW/m (avg. test levels). One pin was ramped to 43 kW/m at 31 W/m/s; after 15 ks the power was increased to 45 kW/m and held constant for 1.9 Ms without failure indication. The other two pins were ramped to 44 kW/m at 23 W/m/s and then to 49 kW/m in a further 1.2 ks; both failed after max. 360 s. The post-irradiation examination revealed large stress-corrosion (SCC) type cladding cracks. Other cracks, down to a few μm deep, were probably early stages of large SCC cracks. Fission gas release in the intact pin was as high as 42% and estimated to be much lower for the two failed pins

Assessment of the TASS 1-D neutronics model for the westinghouse and ABB-CE type PWR reactivity induced transients

International Nuclear Information System (INIS)

Choi, J.D.; Yoon, H.Y.; Um, K.S.; Kim, H.C.; Sim, S.K.

1997-01-01

Best estimate transient analysis code, TASS, has been developed for the normal and transient simulation of the Westinghouse and ABB-CE type PWRs. TASS thermal hydraulic model is based on the non-homogeneous, non-equilibrium two-phase continuity, energy and mixture momentum equations with constitutive relations for closure. Core neutronics model employs both the point kinetics and one-dimensional neutron diffusion model. Semi-implicit numerical scheme is used to solve the discretized finite difference equations. TASS one dimensional neutronics core model has been assessed through the reactivity induced transient analyses for the KORI-3, three loop Westinghouse PWR, and Younggwang-3 (YGN-3), two-loop ABB-CE PWR, nuclear power plants currently operating in Korea. The assessment showed that the TASS one dimensional neutronics core model can be applied for the Westinghouse and ABB-CE type PWRs to gain thermal margin which is necessary for a potential use of the high fuel burnup, extended fuel cycle, power upgrading and for the plant life extension

Cost comparisons of wet and dry interim storage facilities for PWR spent nuclear fuel in Korea

International Nuclear Information System (INIS)

Cho, Chun-Hyung; Kim, Tae-Man; Seong, Ki-Yeoul; Kim, Hyung-Jin; Yoon, Jeong-Hyoun

2011-01-01

Research highlights: → We compare the costs of wet and dry interim storage facilities for PWR spent fuel. → We use the parametric method and quotations to deduce unknown cost items. → Net present values and levelized unit prices are calculated for cost comparisons. → A system price is the most decisive factor in cost comparisons. - Abstract: As a part of an effort to determine the ideal storage solution for pressurized water reactor (PWR) spent nuclear fuel, a cost assessment was performed to better quantify the competitiveness of several storage types. Several storage solutions were chosen for comparison, including three dry storage concepts and a wet storage concept. The net present value (NPV) and the levelized unit cost (LUC) of each solution were calculated, taking into consideration established scenarios and facility size. Wet storage was calculated to be the most expensive solution for a 1700 MTU facility, and metal cask storage marked the highest cost for a 5000 MTU facility. Sensitivity analyses on discount rate, metal cask price, operation and maintenance cost, and facility size revealed that the system price is the most decisive factor affecting competitiveness among the storage types.

Cost comparisons of wet and dry interim storage facilities for PWR spent nuclear fuel in Korea

Energy Technology Data Exchange (ETDEWEB)

Cho, Chun-Hyung, E-mail: skycho@krmc.or.kr [Korea Radioactive Waste Management Corporation, 1045 Daedeokdaero, Yuseong-Gu, Daejeon 305-353 (Korea, Republic of); Kim, Tae-Man; Seong, Ki-Yeoul; Kim, Hyung-Jin; Yoon, Jeong-Hyoun [Korea Radioactive Waste Management Corporation, 1045 Daedeokdaero, Yuseong-Gu, Daejeon 305-353 (Korea, Republic of)

2011-05-15

Research highlights: > We compare the costs of wet and dry interim storage facilities for PWR spent fuel. > We use the parametric method and quotations to deduce unknown cost items. > Net present values and levelized unit prices are calculated for cost comparisons. > A system price is the most decisive factor in cost comparisons. - Abstract: As a part of an effort to determine the ideal storage solution for pressurized water reactor (PWR) spent nuclear fuel, a cost assessment was performed to better quantify the competitiveness of several storage types. Several storage solutions were chosen for comparison, including three dry storage concepts and a wet storage concept. The net present value (NPV) and the levelized unit cost (LUC) of each solution were calculated, taking into consideration established scenarios and facility size. Wet storage was calculated to be the most expensive solution for a 1700 MTU facility, and metal cask storage marked the highest cost for a 5000 MTU facility. Sensitivity analyses on discount rate, metal cask price, operation and maintenance cost, and facility size revealed that the system price is the most decisive factor affecting competitiveness among the storage types.

Westinghouse technologies and integration with Toshiba

International Nuclear Information System (INIS)

Noda, Tetsuya; Tanazawa, Takeshi; Yoshida, Hiroyuki

2007-01-01

With Westinghouse Electric Company (WEC) now a member of the Toshiba Group, Toshiba is capable of supplying both boiling water reactor (BWR) and pressurized water reactor (PWR) systems. WEC is well experienced worldwide in the nuclear business and by integrating the technologies of both Toshiba and WEC. Toshiba will be able to provide a greater range of services in the global market. We will build a cooperative structure not only for the maintenance service and fuel businesses but also for the development of innovative reactors while aiming for global expansion with the AP 1000 PWR, the most advanced PWR in the nuclear power plant business. We will continue making efforts so as to be able to provide all types of products and services as one-stop solutions regardless of the type of reactor. (author)

Development of gamma-ray-suppression type of small-sized neutron detector based on a 6Li-glass scintillator

International Nuclear Information System (INIS)

Matsumoto, T.; Harano, H.; Shimoyama, T.; Kudo, K.; Uritani, A.

2005-01-01

A small-sized thermal neutron detector based on a 6 Li-glass scintillator and a plastic optical fiber was developed for measurement of a dose distribution of thermal neutrons in a thermal neutron standard field. A contribution of gamma rays can not be neglected in the neutron measurement with this detector, although the 6 Li-glass scintillator can be distinguishable for the neutrons and the gamma rays by difference of each pulse height. Moreover, to reduce an uncertainty of neutron counts caused by the gamma ray background around a discrimination level, we suggested a gamma-ray-suppression type of small-sized thermal neutron detector with a 6 Li-glass scintillator, a hollow CsI(Tl) scintillator and plastic optical fibers. The detector can reject signals due to the gamma rays with an anti-coincidence method. In the present paper, we evaluated an ability of a gamma-ray suppression of the detector using the EGS4 electron-photon transport Monte-Carlo code with the PRESTA routine. As the results, the sufficient gamma-ray suppression effect was shown. (author)

Design and static simulation of secondary loop of small PWR nuclear power plants

International Nuclear Information System (INIS)

Martin Lopez, L.A.N.

1989-01-01

A computer program that has been developed with the purpose of making easier the decisions concerning the design of the secondary loop of small PWR nuclear power plants through numerical experiments of low running costs and short time is presented. Initially, the first part of the computer program is described. It aims to preliminarily design several major components of the secondary circuit from user-defined design conditions. Next, the second part of the computer program is presented. It simulates the steady state operation at part-load conditions of the preliminary design of the plant by generating and solving systems of simultaneous nonlinear algebraic equations, their number varying from 17 to 107. The computer program has been tested for several application cases. The program results are discussed in the last part of the work, along with several aspects to be added to the program in future works. (author)

Electropolishing of replacement steam generator channel heads at Millstone-2 PWR

International Nuclear Information System (INIS)

Hudson, M.J.B.; Raney, H.; Raney, D.; Spalaris, C.N.

1992-07-01

A field application of EPRI-developed steam generator electropolishing technique was performed at Millstone-2 PWR. The process was qualified under previous programs on a laboratory scale, but it was thought appropriate to scale up application to full size components. Replacement of steam generators at Millstone-2 provided a unique opportunity to demonstrate that electropolishing can be applied safely and at a cost which was judged to be recoverable after a small number of fuel cycles. The project, preparation, electropolishing and cleanup, was completed at the reactor site in 25 working days. An alternate, less costly electrolyte solution was qualified for use in future applications

Thermohydraulic calculations of PWR primary circuits

International Nuclear Information System (INIS)

Botelho, D.A.

1984-01-01

Some mathematical and numerical models from Retran computer codes aiming to simulate reactor transients, are presented. The equations used for calculating one-dimensional flow are integrated using mathematical methods from Flash code, with steam code to correlate the variables from thermodynamic state. The algorithm obtained was used for calculating a PWR reactor. (E.G.) [pt

EPRI PWR primary water chemistry guidelines revision

International Nuclear Information System (INIS)

McElrath, Joel; Fruzzetti, Keith

2014-01-01

EPRI periodically updates the PWR Primary Water Chemistry Guidelines as new information becomes available and as required by NEI 97-06 (Steam Generator Program Guidelines) and NEI 03-08 (Guideline for the Management of Materials Issues). The last revision of the PWR water chemistry guidelines identified an optimum primary water chemistry program based on then-current understanding of research and field information. This new revision provides further details with regard to primary water stress corrosion cracking (PWSCC), fuel integrity, and shutdown dose rates. A committee of industry experts, including utility specialists, nuclear steam supply system (NSSS) and fuel vendor representatives, Institute of Nuclear Power Operations (INPO) representatives, consultants, and EPRI staff collaborated in reviewing the available data on primary water chemistry, reactor water coolant system materials issues, fuel integrity and performance issues, and radiation dose rate issues. From the data, the committee updated the water chemistry guidelines that all PWR nuclear plants should adopt. The committee revised guidance with regard to optimization to reflect industry experience gained since the publication of Revision 6. Among the changes, the technical information regarding the impact of zinc injection on PWSCC initiation and dose rate reduction has been updated to reflect the current level of knowledge within the industry. Similarly, industry experience with elevated lithium concentrations with regard to fuel performance and radiation dose rates has been updated to reflect data collected to date. Recognizing that each nuclear plant owner has a unique set of design, operating, and corporate concerns, the guidelines committee has retained a method for plant-specific optimization. Revision 7 of the Pressurized Water Reactor Primary Water Chemistry Guidelines provides guidance for PWR primary systems of all manufacture and design. The guidelines continue to emphasize plant

77 FR 30227 - Small Business Size Regulations, Small Business Innovation Research (SBIR) Program and Small...

Science.gov (United States)

2012-05-22

... Business Size Regulations, Small Business Innovation Research (SBIR) Program and Small Business Technology... public Webinar and Roundtable Meetings regarding its proposal to amend its regulations governing size and eligibility for the Small Business Innovation Research (SBIR) and Small Business Technology Transfer (STTR...

Gamma and Neutron Radiolysis in the 21-PWR Waste Package

Energy Technology Data Exchange (ETDEWEB)

J.S. Tang

2001-05-03

The objective of this calculation is to compute gamma and neutron dose rates in order to determine the maximum radiolytic production of nitric acid and other chemical species inside the 21-PWR (pressurized-water reactor) waste package (WP). The scope of this calculation is limited to the time period between 5,000 and 100,000 years after emplacement. The information provided by the sketches attached to this calculation is that of the potential design for the type of WP considered in this calculation. The results of this calculation will be used to evaluate nitric acid corrosion of fuel cladding from radiolysis in the 21-PWR WP. This calculation was performed in accordance with the Technical Work Plan for: Waste Package Design Description for LA (Civilian Radioactive Waste Management System (CRWMS) Management and Operating Contractor (M&O) 2000a). AP-3.124, Calculations, is used to perform the calculation and develop the document. This calculation is associated with the total system performance assessment (TSPA) of which the spent fuel cladding integrity is to be evaluated.

Gamma and Neutron Radiolysis in the 21-PWR Waste Package

International Nuclear Information System (INIS)

J.S. Tang

2001-01-01

The objective of this calculation is to compute gamma and neutron dose rates in order to determine the maximum radiolytic production of nitric acid and other chemical species inside the 21-PWR (pressurized-water reactor) waste package (WP). The scope of this calculation is limited to the time period between 5,000 and 100,000 years after emplacement. The information provided by the sketches attached to this calculation is that of the potential design for the type of WP considered in this calculation. The results of this calculation will be used to evaluate nitric acid corrosion of fuel cladding from radiolysis in the 21-PWR WP. This calculation was performed in accordance with the Technical Work Plan for: Waste Package Design Description for LA (Civilian Radioactive Waste Management System (CRWMS) Management and Operating Contractor (M and O) 2000a). AP-3.124, Calculations, is used to perform the calculation and develop the document. This calculation is associated with the total system performance assessment (TSPA) of which the spent fuel cladding integrity is to be evaluated

Natural vibrations of a core banel of a PWR type reactor by elements of revolution shell

International Nuclear Information System (INIS)

Barcellos, C.S. de.

1980-01-01

Aim to estimate the behavior of the cove barrel of PWR type reactors, submitted to several load conditions, their dynamic characteristic, were determined. In order to obtain the natural modes and frequencies of the core barrel, the CYLDYFE comprete code based in the finite element method, was developed. The obtained results are compared with results obtained by other programs such as SAP, ASKA and STRUDL/DYNAL and by other analytical methods. (M.C.K.) [pt

76 FR 63216 - Small Business Size Standards: Information

Science.gov (United States)

2011-10-12

... Federal small business assistance, SBA establishes small business definitions (referred to as size... business definition or size standard established by the SBA Administrator. The SBA considers as part of its... SMALL BUSINESS ADMINISTRATION 13 CFR Part 121 RIN 3245-AG26 Small Business Size Standards...

PWR: 10 years after and perspectives

International Nuclear Information System (INIS)

1990-01-01

These proceedings of the SFEN days on PWR (Ten years after and perspectives) comprise 13 conferences bearing on: - From the occurential approach to the state approach - Evolution of calculating tools - Human factors and safety - Reactor safety in the PWR 2000 - The PWR and the electrical power grid load follow - Fuel aspect of PWR management - PWR chemistry evolution - Balance of radiation protection - PWR modifications balance and influence on reactor operation - Design and maintenance of reactor components: 4 conferences [fr

Knowledge of ageing phenomenons of materials used in the PWR power plants

International Nuclear Information System (INIS)

Vancon, D.; Meyzaud, Y.; Soulat, P.

1996-01-01

The nuclear power plants with PWR type reactors are planned to work during forty years and are the subject of studies aiming to check their integrity during all their life. The materials used to the fabrication of the components can be submitted different stress. The temperature, the mechanical constraints, the irradiation are examples of stress which can make the materials getting old. This text presents three themes: the ageing by irradiation, the thermal ageing and the corrosion, and their principle industrial consequences. (N.C.)

Cylindrization of a PWR core for neutronic calculations

International Nuclear Information System (INIS)

Santos, Rubens Souza dos

2005-01-01

In this work we propose a core cylindrization, starting from a PWR core configuration, through the use of an algorithm that becomes the process automated in the program, independent of the discretization. This approach overcomes the problem stemmed from the use of the neutron transport theory on the core boundary, in addition with the singularities associated with the presence of corners on the outer fuel element core of, existents in the light water reactors (LWR). The algorithm was implemented in a computational program used to identification of the control rod drop accident in a typical PWR core. The results showed that the algorithm presented consistent results comparing with an production code, for a problem with uniform properties. In our conclusions, we suggest, for future works, for analyzing the effect on mesh sizes for the Cylindrical geometry, and to compare the transport theory calculations versus diffusion theory, for the boundary conditions with corners, for typical PWR cores. (author)

Evaluation of the fuel rod integrity in PWR reactors from the spectrometric analysis of the primary coolant; Avaliacao da integridade de varetas combustiveis em reatores PWR a partir da analise espectrometrica da agua do primario

Energy Technology Data Exchange (ETDEWEB)

Monteiro, Iara Arraes

1999-02-15

The main objective of this thesis is to provide a better comprehension of the phenomena involved in the transport of fission products, from the fuel rod to the coolant of a PWR reactor. To achieve this purpose, several steps were followed. Firstly, it was presented a description of the fuel elements and the main mechanisms of fuel rod failure, indicating the most important nuclides and their transport mechanisms. Secondly, taking both the kinetic and diffusion models for the transport of fission products as a basis, a simple analytical and semi-empirical model was developed. This model was also based on theoretical considerations and measurements of coolant's activity, according to internationally adopted methodologies. Several factors are considered in the modelling procedures: intrinsic factors to the reactor itself, factors which depend on the reactor's operational mode, isotope characteristic factors, and factors which depend on the type of rod failure. The model was applied for different reactor's operational parameters in the presence of failed rods. The main conclusions drawn from the analysis of the model's output are relative to the variation on the coolant's water activity with the fuel burnup, the linear operation power and the primary purification rate and to the different behaviour of iodine and noble gases. The model was saturated from a certain failure size and showed to be unable to distinguish between a single big fail and many small ones. (author)

Comprehensive exergetic and economic comparison of PWR and hybrid fossil fuel-PWR power plants

International Nuclear Information System (INIS)

Sayyaadi, Hoseyn; Sabzaligol, Tooraj

2010-01-01

A typical 1000 MW Pressurized Water Reactor (PWR) nuclear power plant and two similar hybrid 1000 MW PWR plants operate with natural gas and coal fired fossil fuel superheater-economizers (Hybrid PWR-Fossil fuel plants) are compared exergetically and economically. Comparison is performed based on energetic and economic features of three systems. In order to compare system at their optimum operating point, three workable base case systems including the conventional PWR, and gas and coal fired hybrid PWR-Fossil fuel power plants considered and optimized in exergetic and exergoeconomic optimization scenarios, separately. The thermodynamic modeling of three systems is performed based on energy and exergy analyses, while an economic model is developed according to the exergoeconomic analysis and Total Revenue Requirement (TRR) method. The objective functions based on exergetic and exergoeconomic analyses are developed. The exergetic and exergoeconomic optimizations are performed using the Genetic Algorithm (GA). Energetic and economic features of exergetic and exergoeconomic optimized conventional PWR and gas and coal fired Hybrid PWR-Fossil fuel power plants are compared and discussed comprehensively.

Comparison of the long-time corrosion behavior of certain Zr alloys in PWR, BWR, and laboratory tests

International Nuclear Information System (INIS)

Garzarolli, F.; Broy, Y.; Busch, R.A.

1996-01-01

Laboratory corrosion tests have always been an important tool for Zr alloy development and optimization. However, it must be known whether a test is representative for the application in-reactor. To shed more light on this question, coupons of several Zr alloys were exposed under isothermal conditions in BWR and PWR type environments. For evaluation of the in-PWR tests and for comparison of out-of-pile and in-pile tests, the different temperatures and times were normalized to a temperature-independent normalized time by assuming an activation temperature (Q/R) of 14,200 K. Comparison of in-PWR and out-of-pile corrosion behavior of Zircaloy shows that corrosion deviates to higher values in PWR if a weight gain of about 50 mg/dm 2 is exceeded. In the case of the Zr2.5Nb alloy, a slight deviation of corrosion as compared to laboratory results starts in PWR only above a weight gain of 100 mg/dm 2 . In BWR, corrosion of Zircaloy is enhanced early in time if compared with out-of-pile. Zr2.5Nb exhibits higher corrosion results in BWR than Zircaloy-4. Alloying chemistry and material condition affect corrosion of Zr alloys. However, several of the material parameters have shown a different ranking in the different environments. Nevertheless, several material parameters influencing in-reactor corrosion like the second phase particle (SPP) size of in-PWR behavior as the Sn and Fe content can be optimized by out-of-pile corrosion tests

Pressurizer and steam-generator behavior under PWR transient conditions

International Nuclear Information System (INIS)

Wahba, A.B.; Berta, V.T.; Pointner, W.

1983-01-01

Experiments have been conducted in the Loss-of-Fluid Test (LOFT) pressurized water reactor (PWR), at the Idaho National Engineering Laboratory, in which transient phenomena arising from accident events with and without reactor scram were studied. The main purpose of the LOFT facility is to provide data for the development of computer codes for PWR transient analyses. Significant thermal-hydraulic differences have been observed between the measured and calculated results for those transients in which the pressurizer and steam generator strongly influence the dominant transient phenomena. Pressurizer and steam generator phenomena that occurred during four specific PWR transients in the LOFT facility are discussed. Two transients were accompanied by pressurizer inflow and a reduction of the heat transfer in the steam generator to a very small value. The other two transients were accompanied by pressurizer outflow while the steam generator behavior was controlled

Observations of Bright Massive Stars Using Small Size Telescopes

Science.gov (United States)

Beradze, Sopia; Kochiashvili, Nino

2017-11-01

The size of a telescope determines goals and objects of observations. During the latest decades it becomes more and more difficult to get photometric data of bright stars because most of telescopes of small sizes do not operate already. But there are rather interesting questions connected to the properties and evolution ties between different types of massive stars. Multi-wavelength photometric data are needed for solution of some of them. We are presenting our observational plans of bright Massive X-ray binaries, WR and LBV stars using a small size telescope. All these stars, which are presented in the poster are observational targets of Sopia Beradze's future PhD thesis. We already have got very interesting results on the reddening and possible future eruption of the massive hypergiant star P Cygni. Therefore, we decided to choose some additional interesting massive stars of different type for future observations. All Massive stars play an important role in the chemical evolution of galaxies because of they have very high mass loss - up to 10-4M⊙/a year. Our targets are on different evolutionary stages and three of them are the members of massive binaries. We plan to do UBVRI photometric observations of these stars using the 48 cm Cassegrain telescope of the Abastumani Astrophisical Observatory.

Human resources management within the process management in small and medium-sized enterprises

Directory of Open Access Journals (Sweden)

Marie Duspivová

2013-01-01

Full Text Available Sector of small and medium-sized enterprises is regarded as the backbone of the economy and a driving force of innovation, employment and social integration. Development of the sector of small and medium-sized enterprises in the Czech Republic has a substantial impact on economic and social development of the country and its various regions. This article deals with the human resources management in small and medium-sized enterprises, because it is more than obvious recently that the prosperity of the organization is depended on human resources and management of them can determine not only whether the organizations succeed, but whether it will be able to survive in turbulent conditions in the present world. The main aim of this paper is to analytical describe the monitoring the process of human resources management in selected categories of enterprises by business activity and number of employees including statistical analysis of causal effects. Further to analytical describe the evaluation the process of human resources management and indicators of this process, which are monitored by small and medium-sized enterprises. To achieve this aim were used selected primary data collected in the project GAJU 068/2010/S titled “Process management and its possible implementation in small and medium-sized enterprises”.

Evaluation of Advanced Thermohydraulic System Codes for Design and Safety Analysis of Integral Type Reactors

International Nuclear Information System (INIS)

2014-02-01

The integral pressurized water reactor (PWR) concept, which incorporates the nuclear steam supply systems within the reactor vessel, is one of the innovative reactor types with high potential for near term deployment. An International Collaborative Standard Problem (ICSP) on Integral PWR Design, Natural Circulation Flow Stability and Thermohydraulic Coupling of Primary System and Containment during Accidents was established in 2010. Oregon State University, which made available the use of its experimental facility built to demonstrate the feasibility of the Multi-application Small Light Water Reactor (MASLWR) design, and sixteen institutes from seven Member States participated in this ICSP. The objective of the ICSP is to assess computer codes for reactor system design and safety analysis. This objective is achieved through the production of experimental data and computer code simulation of experiments. A loss of feedwater transient with subsequent automatic depressurization system blowdown and long term cooling was selected as the reference event since many different modes of natural circulation phenomena, including the coupling of primary system, high pressure containment and cooling pool are expected to occur during this transient. The power maneuvering transient is also tested to examine the stability of natural circulation during the single and two phase conditions. The ICSP was conducted in three phases: pre-test (with designed initial and boundary conditions established before the experiment was conducted), blind (with real initial and boundary conditions after the experiment was conducted) and open simulation (after the observation of real experimental data). Most advanced thermohydraulic system analysis codes such as TRACE, RELAPS and MARS have been assessed against experiments conducted at the MASLWR test facility. The ICSP has provided all participants with the opportunity to evaluate the strengths and weaknesses of their system codes in the transient

Swing-Down of 21-PWR Waste Package

International Nuclear Information System (INIS)

A.K. Scheider

2001-01-01

The objective of this calculation is to determine the structural response of the waste package (WP) swinging down from a horizontally suspended height. The WP used for that purpose is the 21-Pressurized Water Reactor (PWR) WP. The scope of this document is limited to reporting the calculation results in terms of stress intensities. This calculation is associated with the WP design and was performed by the Waste Package Design group in accordance with the ''Technical Work Plan for: Waste Package Design Description for LA'' (Ref. 13). AP-3.12Q, ''Calculations'' (Ref. 18) is used to perform the calculation and develop the document. The information provided by the sketches attached to this calculation is that of the potential design of the type of 21-PWR WP design considered in this calculation and provides the potential dimensions and materials for the 21-PWR WP design

Coconut genome size determined by flow cytometry: Tall versus Dwarf types.

Science.gov (United States)

Freitas Neto, M; Pereira, T N S; Geronimo, I G C; Azevedo, A O N; Ramos, S R R; Pereira, M G

2016-02-11

Coconuts (Cocos nucifera L.) are tropical palm trees that are classified into Tall and Dwarf types based on height, and both types are diploid (2n = 2x = 32 chromosomes). The reproduction mode is autogamous for Dwarf types and allogamous for Tall types. One hypothesis for the origin of the Dwarf coconut suggests that it is a Tall variant that resulted from either mutation or inbreeding, and differences in genome size between the two types would support this hypothesis. In this study, we estimated the genome sizes of 14 coconut accessions (eight Tall and six Dwarf types) using flow cytometry. Nuclei were extracted from leaf discs and stained with propidium iodide, and Pisum sativum (2C = 9.07 pg DNA) was used as an internal standard. Histograms with good resolution and low coefficients of variation (2.5 to 3.2%) were obtained. The 2C DNA content ranged from 5.72 to 5.48 pg for Tall accessions and from 5.58 to 5.52 pg for Dwarf accessions. The mean genome sizes for Tall and Dwarf specimens were 5.59 and 5.55 pg, respectively. Among all accessions, Rennel Island Tall had the highest mean DNA content (5.72 pg), whereas West African Tall had the lowest (5.48 pg). The mean coconut genome size (2C = 5.57 pg, corresponding to 2723.73 Mbp/haploid set) was classified as small. Only small differences in genome size existed among the coconut accessions, suggesting that the Dwarf type did not evolve from the Tall type.

PWR water chemistry controls: a perspective on industry initiatives and trends relative to operating experience and the EPRI PWR water chemistry guidelines

International Nuclear Information System (INIS)

Fruzzetti, K.; Choi, S.; Haas, C.; Pender, M.; Perkins, D.

2010-01-01

An effective PWR water chemistry control program must address the following goals: Minimize materials degradation (e.g., PWSCC, corrosion of fuel, corrosion damage of steam generator (SG) tubes); Maintain fuel integrity and good performance; Minimize corrosion product transport (e.g., transport and deposition on the fuel, transport into the SGs where it can foul tube surfaces and create crevice environments for the concentration of corrosive impurities); Minimize dose rates. Water chemistry control must be optimized to provide overall improvement considering the sometimes variant constraints of the goals listed above. New technologies are developed for continued mitigation of materials degradation, continued fuel integrity and good performance, continued reduction of corrosion product transport, and continued minimization of plant dose rates. The EPRI chemistry program, in coordination with other EPRI programs, strives to improve these areas through application of chemistry initiatives, focusing on these goals. This paper highlights the major initiatives and issues with respect to PWR primary and secondary system chemistry and outlines the recent, on-going, and proposed work to effectively address them. These initiatives are presented in light of recent operating experience, as derived from EPRI's PWR chemistry monitoring and assessment program, and EPRI's water chemistry guidelines. (author)

BWR and PWR chemistry operating experience and perspectives

International Nuclear Information System (INIS)

Fruzzetti, K.; Garcia, S.; Lynch, N.; Reid, R.

2014-01-01

It is well recognized that proper control of water chemistry plays a critical role in ensuring the safe and reliable operation of Boiling Water Reactors (BWRs) and Pressurized Water Reactors (PWRs). State-of-the-art water chemistry programs reduce general and localized corrosion of reactor coolant system, steam cycle equipment, and fuel cladding materials; ensure continued integrity of cycle components; and reduce radiation fields. Once a particular nuclear plant component has been installed or plant system constructed, proper water chemistry provides a global tool to mitigate materials degradation problems, thereby reducing the need for costly repairs or replacements. Recognizing the importance of proper chemistry control and the value in understanding the relationship between chemistry guidance and actual operating experience, EPRI continues to collect, monitor, and evaluate operating data from BWRs and PWRs around the world. More than 900 cycles of valuable BWR and PWR operating chemistry data has been collected, including online, startup and shutdown chemistry data over more than 10 years (> 20 years for BWRs). This paper will provide an overview of current trends in BWR and PWR chemistry, focusing on plants in the U.S.. Important chemistry parameters will be highlighted and discussed in the context of the EPRI Water Chemistry Guidelines requirements (i.e., those parameters considered to be of key importance as related to the major goals identified in the EPRI Guidelines: materials integrity; fuel integrity; and minimizing plant radiation fields). Perspectives will be provided in light of recent industry initiatives and changes in the EPRI BWR and PWR Water Chemistry Guidelines. (author)

A comparison of fuzzy logic-PID control strategies for PWR pressurizer control

International Nuclear Information System (INIS)

Kavaklioglu, K.; Ikonomopoulos, A.

1993-01-01

This paper describes the results obtained from a comparison performed between classical proportional-integral-derivative (PID) and fuzzy logic (FL) controlling the pressure in a pressurized water reactor (PWR). The two methodologies have been tested under various transient scenarios, and their performances are evaluated with respect to robustness and on-time response to external stimuli. One of the main concerns in the safe operation of PWR is the pressure control in the primary side of the system. In order to maintain the pressure in a PWR at the desired level, the pressurizer component equipped with sprayers, heaters, and safety relief valves is used. The control strategy in a Westinghouse PWR is implemented with a PID controller that initiates either the electric heaters or the sprayers, depending on the direction of the coolant pressure deviation from the setpoint

Study of development of non-destructive method for determining FGR from high burned PWR type fuel rod

International Nuclear Information System (INIS)

Yanagisawa, Kazuaki; Miyanishi, Hideyuki; Kitagawa, Isamu; Iida, Shozo; Ito, Tadaharu; Amano, Hidetoshi.

1991-11-01

Experimental study was made to evaluate the FGR (Fission Product Gas Release) from high burned PWR type fuel rods by means of non-destructive method through measurement of the gamma activity of 85 Kr isotope which was accumulated in the fuel top plenum. Experimental result shows that it is possible to know the amounts of FGR at fuel plenum by the equations given in the followings. FGR = 0.28C/V f or FGR = 0.07C where, FGR (%) is the amounts of Xe and Kr released from UO 2 fuel, C (counts/h) the radioactivity of 85 Kr at plenum of the tested fuel rod and V f (ml) the plenum volume of the tested fuel rod, respectively. The present study was made by using 14 x 14 PWR type fuel rods preirradiated up to the burn-up of 42.1 MWd/kgU, followed by the pulse irradiation at Nuclear Safety Research Reactor of Japan Atomic Energy Research Institute (JAERI). The FGR of the tested segmented fuel rods were measured by puncturing and found to range from 0.6% to 12% according to the magnitude of the deposited energy given by pulse. Estimated experimental error bands against the above equations were within plus minus 30%. (author)

Failure probability of PWR reactor coolant loop piping

International Nuclear Information System (INIS)

Lo, T.; Woo, H.H.; Holman, G.S.; Chou, C.K.

1984-02-01

This paper describes the results of assessments performed on the PWR coolant loop piping of Westinghouse and Combustion Engineering plants. For direct double-ended guillotine break (DEGB), consideration was given to crack existence probability, initial crack size distribution, hydrostatic proof test, preservice inspection, leak detection probability, crack growth characteristics, and failure criteria based on the net section stress failure and tearing modulus stability concept. For indirect DEGB, fragilities of major component supports were estimated. The system level fragility was then calculated based on the Boolean expression involving these fragilities. Indirect DEGB due to seismic effects was calculated by convolving the system level fragility and the seismic hazard curve. The results indicate that the probability of occurrence of both direct and indirect DEGB is extremely small, thus, postulation of DEGB in design should be eliminated and replaced by more realistic criteria

Evaluation of the presence of a burnable absorber in an assembly 3x3 type PWR

International Nuclear Information System (INIS)

Martinez F, M. A.; Del Valle G, E.; Alonso V, G.

2008-01-01

In the present work the effect is evaluated that causes the presence of a burnable absorber in an adjustment of rods of 3x3 of a fuel assembly type PWR using CASMO-4 code, when comparing the infinite multiplication factor and some average cross sections by means of codes MCNP-4A, CASMO-3 and HELIOS. For this evaluation two cases are evaluated: first consists of an adjustment of rods of 3x3 full completely of fuel and the second consists of a central rod full with a burnable absorber type wet annular burnable absorber (WABA) and the remaining full fuel rods. In both cases the enrichment of the fissile isotopes is varied, for two types of fuel, MOX degree armament and UO 2 . (Author)

77 FR 72702 - Small Business Size Standards: Information

Science.gov (United States)

2012-12-06

... SMALL BUSINESS ADMINISTRATION 13 CFR Part 121 RIN 3245-AG26 Small Business Size Standards: Information AGENCY: U.S. Small Business Administration. ACTION: Final rule. SUMMARY: The United States Small Business Administration (SBA) is increasing the receipts based small business size standards for 15...

The advanced main control console for next japanese PWR plants

International Nuclear Information System (INIS)

Tsuchiya, A.; Ito, K.; Yokoyama, M.

2001-01-01

The purpose of the improvement of main control room designing in a nuclear power plant is to reduce operators' workload and potential human errors by offering a better working environment where operators can maximize their abilities. In order to satisfy such requirements, the design of main control board applied to Japanese Pressurized Water Reactor (PWR) type nuclear power plant has been continuously modified and improved. the Japanese Pressurized Water Reactor (PWR) Utilities (Electric Power Companies) and Mitsubishi Group have developed an advanced main control board (console) reflecting on the study of human factors, as well as using a state of the art electronics technology. In this report, we would like to introduce the configuration and features of the Advanced Main Control Console for the practical application to the next generation PWR type nuclear power plants including TOMARI No.3 Unit of Hokkaido Electric Power Co., Inc. (author)

Time integrated spectroscopy of turbid media based on the microscopic beer-lambert law: application to small-size phantoms having different boundary conditions.

Science.gov (United States)

Zhang, H; Urakami, T; Tsuchiya, Y; Lu, Z; Hiruma, T

1999-01-01

Continued work on time-integrated spectroscopy (TIS) is presented to quantify absorber concentrations in turbid media. We investigated the applicability of the TIS method to small-size media that have different boundary conditions by measuring two 20×20×50 mm3 cuboid liquid tissue-like phantoms at various absorption levels (absorption coefficients of the phantom from 2.5×10-3 to 4.4×10-2 mm-1 at 782 nm and from 3.1×10-3 to 2.7×10-2 mm-1 at 831 nm). The scattering and absorbing solution was filled into ordinary and black-anodized aluminum containers to provide different boundary conditions. By means of a single equation, the absorber concentrations have been recovered within errors of a few percent in both cases. This demonstrates that the TIS method can quantify absorbers in small-size media having different boundary conditions. © 1999 Society of Photo-Optical Instrumentation Engineers.

Maturity of the PWR

International Nuclear Information System (INIS)

Bacher, P.; Rapin, M.; Aboudarham, L.; Bitsch, D.

1983-03-01

Figures illustrating the predominant position of the PWR system are presented. The question is whether on the basis of these figures the PWR can be considered to have reached maturity. The following analysis, based on the French program experience, is an attempt to pinpoint those areas in which industrial maturity of the PWR has been attained, and in which areas a certain evolution can still be expected to take place

Integrity assessment of the cold leg piping system in a PWR

International Nuclear Information System (INIS)

Mayfield, M.E.; Leis, B.N.

1981-01-01

The purpose of this paper is to examine the integrity of a nuclear piping system, designed in accordance with Section III, in the context of a damage tolerance analysis procedure. Such a procedure directly addresses the defects and cyclic loadings that are responsible for the above noted exceptions. The analysis and results reported here are for a fatigue life analysis of the Cold Leg piping in a PWR. This piping system is particularly important from a safety standpoint since a large break is a possible initiator of a core meltdown accident. The analysis employs LEFM concepts to determine the time between the initial start-up and (1) formation of a leak, (2) detection of the leak, and (3) the final fracture of the piping. Both longitudinal and circumferential defects are considered. The defects are assumed to propagate from the pipe I.D. in a self-similar manner. Inputs to the analysis were derived from information supplied by plant operators and vendors, published data, and 'expert opinions'. The life was computed using a linear damage accumulation. (orig./GL)

The verification of PWR-fuel code for PWR in-core fuel management

International Nuclear Information System (INIS)

Surian Pinem; Tagor M Sembiring; Tukiran

2015-01-01

In-core fuel management for PWR is not easy because of the number of fuel assemblies in the core as much as 192 assemblies so many possibilities for placement of the fuel in the core. Configuration of fuel assemblies in the core must be precise and accurate so that the reactor operates safely and economically. It is necessary for verification of PWR-FUEL code that will be used in-core fuel management for PWR. PWR-FUEL code based on neutron transport theory and solved with the approach of multi-dimensional nodal diffusion method many groups and diffusion finite difference method (FDM). The goal is to check whether the program works fine, especially for the design and in-core fuel management for PWR. Verification is done with equilibrium core search model at three conditions that boron free, 1000 ppm boron concentration and critical boron concentration. The result of the average burn up fuel assemblies distribution and power distribution at BOC and EOC showed a consistent trend where the fuel with high power at BOC will produce a high burn up in the EOC. On the core without boron is obtained a high multiplication factor because absence of boron in the core and the effect of fission products on the core around 3.8 %. Reactivity effect at 1000 ppm boron solution of BOC and EOC is 6.44 % and 1.703 % respectively. Distribution neutron flux and power density using NODAL and FDM methods have the same result. The results show that the verification PWR-FUEL code work properly, especially for core design and in-core fuel management for PWR. (author)

The PWR cores management

International Nuclear Information System (INIS)

Barral, J.C.; Rippert, D.; Johner, J.

2000-01-01

During the meeting of the 25 january 2000, organized by the SFEN, scientists and plant operators in the domain of the PWR debated on the PWR cores management. The five first papers propose general and economic information on the PWR and also the fast neutron reactors chains in the electric power market: statistics on the electric power industry, nuclear plant unit management, the ITER project and the future of the thermonuclear fusion, the treasurer's and chairman's reports. A second part offers more technical papers concerning the PWR cores management: performance and optimization, in service load planning, the cores management in the other countries, impacts on the research and development programs. (A.L.B.)

The role of micro size computing clusters for small physics groups

International Nuclear Information System (INIS)

Shevel, A Y

2014-01-01

A small physics group (3-15 persons) might use a number of computing facilities for the analysis/simulation, developing/testing, teaching. It is discussed different types of computing facilities: collaboration computing facilities, group local computing cluster (including colocation), cloud computing. The author discuss the growing variety of different computing options for small groups and does emphasize the role of the group owned computing cluster of micro size.

Study of the noise propagation in PWR with coupled codes

International Nuclear Information System (INIS)

Verdu, G.; Garcia-Fenoll, M.; Abarca, A.; Miro, R.; Barrachina, T.

2011-01-01

The in-core detectors provide signals of the power distribution monitoring for the Reactor Protection System (RPS). The advanced fuel management strategies (high exposure) and the power upratings for PWR reactor types have led to an increase in the noise amplitude in detectors signals. In the present work a study of the propagation along the reactor core and the effects on the core power evolution of a small perturbation on the moderator density, using the coupled code RELAP5-MOD3.3/PARCSv2.7 is presented. The purpose of these studies is to be able to reproduce and analyze the in-core detector simulated signals. (author)

Monte Carlo based radial shield design of typical PWR reactor

Energy Technology Data Exchange (ETDEWEB)

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

2016-11-15

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

PWR type process heat reactor

International Nuclear Information System (INIS)

Aubert, Gilles; Petit, Guy.

1974-01-01

The nuclear reactor described is of the pressurized water type. It includes a prestressed concrete vessel, the upper part of which is shut by a closure, and a core surrounded by a core ring. The core fuel assemblies are supported by an initial set of vertical tubes integral with the bottom of the vessel, which serve to guide the rods of the control system. Over the core there is a second set of vertical tubes, able to receive the absorbing part of a control rod when this is raised above the core. An annular pressurizer around the core ring keeps the water in a liquid state. A pump is located above the second set of tubes and is integral with the closure. It circulates the water between the core and the intake of at least one primary heat exchanger, the exchanger (s) being placed between the wall of the vessel and the core ring [fr

Pressurized Water Reactors (PWR) and Boiling Water Reactors (BWR) are compared

International Nuclear Information System (INIS)

Greneche, D.

2014-01-01

This article compares the 2 types of light water reactors that are used to produce electricity: the Pressurized Water Reactor (PWR) and the Boiling Water Reactor (BWR). Historically the BWR concept was developed after the PWR concept. Today 80% of light water reactors operating in the world are of PWR-type. This comparison is comprehensive and detailed. First the main technical features are reviewed and compared: reactor architecture, core and fuel design, reactivity control, reactor vessel, cooling systems and reactor containment. Secondly, various aspects concerning reactor operations like reactor control, fuel management, maintenance, inspections, radiation protection, waste generation and reactor reliability are presented and compared for both reactors. As for the issue of safety, it is highlighted that the accidental situations are too different for the 2 reactors to be compared. The main features of reactor safety are explained for both reactors

Simplified model for the thermo-hydraulic simulation of the hot channel of a PWR type nuclear reactor

International Nuclear Information System (INIS)

Belem, J.A.T.

1993-09-01

The present work deals with the thermal-hydraulic analysis of the hot channel of a standard PWR type reactor utilizing a simplified mathematical model that considers constant the water mass flux during single-phase flow and reduction of the flow when the steam quality is increasing in the channel (two-phase flow). The model has been applied to the Angra-1 reactor and it has proved satisfactory when compared to other ones. (author). 25 refs, 15 figs, 3 tabs

76 FR 70667 - Small Business Size Standards: Educational Services

Science.gov (United States)

2011-11-15

... business assistance, SBA establishes small business size definitions (referred to as size standards) for... SMALL BUSINESS ADMINISTRATION 13 CFR Part 121 RIN 3245-AG29 Small Business Size Standards: Educational Services AGENCY: U.S. Small Business Administration. ACTION: Proposed rule. SUMMARY: The U.S...

JAIF report on small- and medium-size LWRs

International Nuclear Information System (INIS)

Anon.

1983-01-01

The Ministry of International Trade and Industry has made a report on the results of the study on the utilization of small and medium size LWRs. The Japan Atomic Industrial Forum undertook this research work by the commission of the Ministry. It studied six cases of PWRs and BWRs with thermal output power ranging from 20 MWt to 700 MWt as the hypothetical design for their siting conditions, reactor capacity and the examples of utilization of the corresponding steam output. A large housing estate of 5,000 units, a local core city with population of 100,000, industrial utilization for petrochemical plants and paper and pulp plants and the combined utilization for a factory and a housing estate of 5,000 units were examined. The report said that a city with population of 100,000 would be the best object to supply heat from a nuclear reactor, and it is important to develop the nuclear reactors with 200 MWt capacity. This type is versatile in the siting conditions, and can be constructed near a city, accordingly, it offers the possibility of extensive utilization. The significance of small and medium size LWRs as the alternative energy for oil, the regional development accompanying the use of small and medium size LWRs, the technical development toward scale reduction, easy operation and versatile utilization, the economical evaluation of the utilization models and the tasks for the future are reported. (Kako, I.)

Preliminary design concepts of an advanced integral reactor

International Nuclear Information System (INIS)

Moon, Kap S.; Lee, Doo J.; Kim, Keung K.; Chang, Moon H.; Kim, Si H.

1997-01-01

An integral reactor on the basis of PWR technology is being conceptually developed at KAERI. Advanced technologies such as intrinsic and passive safety features are implemented in establishing the design concepts of the reactor to enhance the safety and performance. Research and development including laboratory-scale tests are concurrently underway for confirming the technical adoption of those concepts to the rector design. The power output of the reactor will be in the range of 100MWe to 600MWe which is relatively small compared to the existing loop type reactors. The detailed analysis to assure the design concepts is in progress. (author). 3 figs, 1 tab

Evaluation of steam generator U-tube integrity during PWR station blackout with secondary system depressurization

International Nuclear Information System (INIS)

Hidaka, Akihide; Asaka, Hideaki; Sugimoto, Jun; Ueno, Shingo; Yoshino, Takehito

1999-12-01

In PWR severe accidents such as station blackout, the integrity of steam generator U-tube would be threatened early at the transient among the pipes of primary system. This is due to the hot leg countercurrent natural circulation (CCNC) flow which delivers the decay heat of the core to the structures of primary system if the core temperature increases after the secondary system depressurization. From a view point of accident mitigation, this steam generator tube rupture (SGTR) is not preferable because it results in the direct release of primary coolant including fission products (FP) to the environment. Recent SCDAP/RELAP5 analyses by USNRC showed that the creep failure of pressurizer surge line which results in release of the coolant into containment would occur earlier than SGTR during the secondary system depressurization. However, the analyses did not consider the decay heat from deposited FP on the steam generator U-tube surface. In order to investigate the effect of decay heat on the steam generator U-tube integrity, the hot leg CCNC flow model used in the USNRC's calculation was, at first, validated through the analysis for JAERI's LSTF experiment. The CCNC model reproduced well the thermohydraulics observed in the LSTF experiment and thus the model is mostly reliable. An analytical study was then performed with SCDAP/RELAP5 for TMLB' sequence of Surry plant with and without secondary system depressurization. The decay heat from deposited FP was calculated by JAERI's FP aerosol behavior analysis code, ART. The ART analysis showed that relatively large amount of FPs may deposit on steam generator U-tube inlet mainly by thermophoresis. The SCDAP/RELAP5 analyses considering the FP decay heat predicted small safety margin for steam generator U-tube integrity during secondary system depressurization. Considering associated uncertainties in the analyses, the potential for SGTR cannot be ignored. Accordingly, this should be considered in the evaluation of merits

Criticality coefficient calculation for a small PWR using Monte Carlo Transport Code

Energy Technology Data Exchange (ETDEWEB)

Trombetta, Debora M.; Su, Jian, E-mail: dtrombetta@nuclear.ufrj.br, E-mail: sujian@nuclear.ufrj.br [Coordenacao dos Programas de Pos-Graduacao em Engenharia (COPPE/UFRJ), Rio de Janeiro, RJ (Brazil); Chirayath, Sunil S., E-mail: sunilsc@tamu.edu [Department of Nuclear Engineering and Nuclear Security Science and Policy Institute, Texas A and M University, TX (United States)

2015-07-01

Computational models of reactors are increasingly used to predict nuclear reactor physics parameters responsible for reactivity changes which could lead to accidents and losses. In this work, preliminary results for criticality coefficient calculation using the Monte Carlo transport code MCNPX were presented for a small PWR. The computational modeling developed consists of the core with fuel elements, radial reflectors, and control rods inside a pressure vessel. Three different geometries were simulated, a single fuel pin, a fuel assembly and the core, with the aim to compare the criticality coefficients among themselves.The criticality coefficients calculated were: Doppler Temperature Coefficient, Coolant Temperature Coefficient, Coolant Void Coefficient, Power Coefficient, and Control Rod Worth. The coefficient values calculated by the MCNP code were compared with literature results, showing good agreement with reference data, which validate the computational model developed and allow it to be used to perform more complex studies. Criticality Coefficient values for the three simulations done had little discrepancy for almost all coefficients investigated, the only exception was the Power Coefficient. Preliminary results presented show that simple modelling as a fuel assembly can describe changes at almost all the criticality coefficients, avoiding the need of a complex core simulation. (author)

Application of the integrated analysis of safety (IAS) to sequences of Total loss of feed water in a PWR Reactor; Aplicacion del Analisis Integrado de Seguridad (ISA) a Secuencias de Perdidas Total de Agua de Alimentacion en un Reactor PWR

Energy Technology Data Exchange (ETDEWEB)

Moreno Chamorro, P.; Gallego Diaz, C.

2011-07-01

The main objective of this work is to show the current status of the implementation of integrated analysis of safety (IAS) methodology and its SCAIS associated tool (system of simulation codes for IAS) to the sequence analysis of total loss of feedwater in a PWR reactor model Westinghouse of three loops with large, dry containment.

14C Behaviour in PWR coolant

International Nuclear Information System (INIS)

Sims, Howard; Dickinson Shirley; Garbett, Keith

2012-09-01

Although 14 C is produced in relatively small amounts in PWR coolant, it is important to know its fate, for example whether it is released by gaseous discharge, removed by absorption on ion exchange (IX) resins or deposited on the fuel pin surfaces. 14 C can exist in a range of possible chemical forms: inorganic carbon compounds (probably mainly CO 2 ), elemental carbon, and organic compounds such as hydrocarbons. This paper presents results from a preliminary survey of the possible reactions of 14 C in PWR coolant. The main conclusions of the study are: - A combination of thermal and radiolytic reactions controls the chemistry of 14 C in reactor coolant. A simple chemical kinetic model predicts that CH 3 OH would be the initial product from radiolytic reactions of 14 C following its formation from 17 O. CH 3 OH is predicted to arise as a result of reactions of OH . with CH 4 and CH 3 , and it persists because there is no known radiation chemical reduction mechanism. - Thermodynamic considerations show that CH 3 OH can be thermally reduced to CH 4 in PWR conditions, although formation of CO 2 from small organics is the most thermodynamically favourable outcome. Such reactions could be catalysed on active nickel surfaces in the primary circuit. - Limited plant data would suggest that CH 4 is the dominant form in PWR and CO 2 in BWR. This implies that radiation chemistry may be important in determining the speciation. - Addition of acetate does not affect the amount of 14 C formed, but the addition of large amounts of stable carbon would lead to a large range of additional products, some of which would be expected to deposit on fuel pin surfaces as high molecular weight hydrocarbons. However, the subsequent thermal decomposition reactions of these products are not known. - Acetate addition may represent a small input of 12 C compared with organic material released from CVCS resins, although the importance of this may depend on whether that is predominantly soluble

A Multi-Physics PWR Model for the Load Following

OpenAIRE

Muniglia , Mathieu; Do , Jean-Michel; Jean-Charles , Le Pallec; Grard , Hubert; Verel , Sébastien; David , S.

2016-01-01

International audience; In this paper, a new model of a Pressurized Water Reactor (PWR) is described. This model includes the description of the core as well as a simplified secondary loop: the goal is to reproduce a load-following type transient, where the output power of the plant is controlled by the electric grid. Consequently, the control systems are also modeled, as the control rods or the soluble boron. The reference power plant is a 1300MW electrical PWR, managed with the french G mode.

Building of Nuclear Ship Engineering Simulation System development of the simulator for the integral type reactor

Energy Technology Data Exchange (ETDEWEB)

Takahashi, Teruo; Shimazaki, Junya; Yabuuchi, Noriaki; Fukuhara, Yosifumi; Kusunoki, Takeshi; Ochiai, Masaaki [Department of Nuclear Energy Systems, Tokai Research Establishment, Japan Atomic Energy Research Institute, Tokai, Ibaraki (Japan); Nakazawa, Toshio [Department of HTTR Project, Oarai Research Establishment, Japan Atomic Energy Research Institute, Oarai, Ibaraki (Japan)

2000-03-01

JAERI had carried out the design study of a light-weight and compact integral type reactor of power 100 MW{sub th} with passive safety as a power source for the future nuclear ships, and completed an engineering design. To confirm the design and operation performance and to utilize the study of automation of the operations of reactor, we developed a real-time simulator for the integral type reactor. This simulator is a part of Nuclear Ship Engineering Simulation System (NESSY) and on the same hardware as 'Mutsu' simulator which was developed to simulate the first Japanese nuclear ship Mutsu'. Simulation accuracy of 'Mutsu' simulator was verified by comparing the simulation results With data got in the experimental voyage of 'Mutsu'. The simulator for the integral type reactor uses the same programs which were used in 'Mutsu' simulator for the separate type PWR, and the simulated results are approximately consistent with the calculated values using RELAP5/MOD2 (The later points are reported separately). Therefore simulation accuracy of the simulator for the integral type reactor is also expected to be reasonable, though it is necessary to verify by comparing with the real plant data or experimental data in future. We can get the perspectives to use as a real-time engineering simulator and to achieve the above-mentioned aims. This is a report on development of the simulator for the integral type reactor mainly focused on the contents of the analytical programs expressed the structural features of reactor. (author)

Alloy 690 in PWR type reactors; Aleaciones base niquel en condiciones de primario de los reactores tipo PWR

Energy Technology Data Exchange (ETDEWEB)

Gomez Briceno, D.; Serrano, M.

2005-07-01

Alloy 690, used as replacement of Alloy 600 for vessel head penetration (VHP) nozzles in PWR, coexists in the primary loop with other components of Alloy 600. Alloy 690 shows an excellent resistance to primary water stress corrosion cracking, while Alloy 600 is very susceptible to this degradation mechanisms. This article analyse comparatively the PWSCC behaviour of both Ni-based alloys and associated weld metals 52/152 and 82/182. (Author)

Innovations in PHWR design, integration of nuclear power stations into power systems and role of small size nuclear power plants in a developing country

International Nuclear Information System (INIS)

Mehta, S.K.; Kakodkar, A.; Balakrishnan, M.R.; Ray, R.N.; Murthy, L.G.K.; Chamany, B.F.; Kati, S.L.

1977-01-01

PHWR concept of thermal reactors has been considered with a view to exploiting the limited resources of natural uranium and keeping in mind the projected nuclear power programme covering fast breeder reactors. Experience in engineering of current PHWR units in India, gradual build up of necessary infrastructure and operational experience with one unit, have helped in building up design and technological capability in the country. The R and D facilities have been so planned that additional data required for the design of bigger reactor units (i.e.500/600 MWe) could be generated with minimal augmentation. Satisfactory operation of a nuclear power station demands certain prerequisites from the connected power system. The grid should have load patterns suitable for base load operation of these stations, should be stiff so far as voltage and frequency fluctuations are concerned and should have high reliability. A typical power grid in this country is characterised by heavy loads during peak hours and very light loads during night. Regional grids are of small size and the few interconnections existing between the regional grids consist of weak tie lines. Amongst all types of the power stations, it is the nuclear system which undergoes maximum strain and economic penalty while operating when connected to such a power system. Consistent with the above, phase installation of small-size power reactor units of about 200 MWe capacity may facilitate setting up of larger unit sizes at a later date. The effect of any possible reduction in the capital cost of a larger unit power station will enable the power station to partially meet the demand of the more productive types of loads. This paper deals with some of the major design changes that are being incorporated in the PHWR type power reactors currently being set up and the research and development back-up required for the purpose. Since the unit sizes of the power reactors presently contemplated are small compared to nuclear

77 FR 28520 - Small Business Size Regulations, Small Business Innovation Research (SBIR) Program and Small...

Science.gov (United States)

2012-05-15

... SMALL BUSINESS ADMINISTRATION 13 CFR Part 121 RIN 3245-AG46 Small Business Size Regulations, Small Business Innovation Research (SBIR) Program and Small Business Technology Transfer (STTR) Program AGENCY: Small Business Administration. ACTION: Proposed rule. SUMMARY: The U.S. Small Business Administration...

Analysis of bubble pressure in the rim region of high burnup PWR fuel

Energy Technology Data Exchange (ETDEWEB)

Koo, Yang Hyun; Lee, Byung Ho; Sohn, Dong Seong [Korea Atomic Energy Research Institute, Taejeon (Korea)

2000-02-01

Bubble pressure in the rim region of high burnup PWR UO{sub 2} fuel has been modeled based on measured rim width, porosity and bubble density. Using the assumption that excessive bubble pressure in the rim is inversely proportional to its radius, proportionality constant is derived as a function of average pellet burnup and bubble radius. This approach is possible because the integration of the number of Xe atoms retained in the rim bubbles, which can be calculated as a function of bubble radius, over the bubble radius gives the total number of Xe atoms in the rim bubbles. Here the total number of Xe atoms in the rim bubbles can be derived from the measured Xe depletion fraction in the matrix and the calculated rim thickness. Then the rim bubble pressure is obtained as a function of fuel burnup and bubble size from the proportionality constant. Therefore, the present model can provide some useful information that would be required to analyze the behavior of high burnup PWR UO{sub 2} fuel under both normal and transient operating conditions. 28 refs., 9 figs. (Author)

RCC-M - Design and Conception Rules for Mechanical Components of PWR Nuclear Islands

International Nuclear Information System (INIS)

2007-01-01

The design and construction rules applicable to mechanical components of PWR Nuclear Islands (RCC-M) are a part of the collection of design and construction rules for nuclear power plants. It covers the rules applicable to the design and manufacture of pressure boundaries of mechanical equipment of pressurized water reactors (PWR). The pressure components subject to the RCC-M are specified in A 4000. They include the reactor fluid systems (primary, secondary and auxiliary systems) and other components which are not subject to pressure: vessel internals, supports for pressure components subject to the RCC-M, nuclear island storage tanks. When a pressure equipment is subject to the RCC-M, all its elements subject to pressure are also, in accordance with the provisions of A 4000, and these elements are the same class as the component. In this case all the provisions of the RCC-M are applicable: design, procurement, manufacture, inspection and pressure testing. Elements which are not subject to pressure and which are subject to the RCC-M may be covered within the Code by limited specific provisions (procurement of materials for example). The other rules applicable to this equipment must be in contractual form. The assemblies comprising pressure equipment assembled by a manufacturer to constitute an integrated and functional whole, shall be subject to the rules indicated in this Code. Main objectives of Code Requirements are to ensure the integrity and mechanical stability over the equipment design life. Function ability and operability of equipment are not directly addressed in the Code. The RCC-M contributes to ensuring compliance with regulatory requirements. These requirements depend on the applicable regulatory context. The RCC-M is representative of the state of the art as concerns the design and manufacture of PWR components, ensuring an overall safety level tested through experience. The RCC-M consists of five sections, which provide rules for the design and

Material property changes of stainless steels under PWR irradiation

International Nuclear Information System (INIS)

Fukuya, Koji; Nishioka, Hiromasa; Fujii, Katsuhiko; Kamaya, Masayuki; Miura, Terumitsu; Torimaru, Tadahiko

2009-01-01

Structural integrity of core structural materials is one of the key issues for long and safe operation of pressurized water reactors. The stainless steel components are exposed to neutron irradiation and high-temperature water, which cause significant property changes and irradiation assisted stress corrosion cracking (IASCC) in some cases. Understanding of irradiation induced material property changes is essential to predict integrity of core components. In the present study, microstructure and microchemistry, mechanical properties, and IASCC behavior were examined in 316 stainless steels irradiated to 1 - 73 dpa in a PWR. Dose-dependent changes of dislocation loops and cavities, grain boundary segregation, tensile properties and fracture mode, deformation behavior, and their interrelation were discussed. Tensile properties and deformation behavior were well coincident with microstructural changes. IASCC susceptibility under slow strain rate tensile tests, IASCC initiation under constant load tests in simulated PWR primary water, and their relationship to material changes were discussed. (author)

Test requirements for the integral effect test to simulate Korean PWR plants

Energy Technology Data Exchange (ETDEWEB)

Song, Chul Hwa; Park, C. K.; Lee, S. J.; Kwon, T. S.; Yun, B. J.; Chung, M. K

2001-02-01

In this report, the test requirements are described for the design of the integral effect test facility to simulate Korean PWR plants. Since the integral effect test facility should be designed so as to simulate various thermal hydraulic phenomena, as closely as possible, to be occurred in real plants during operation or anticipated transients, the design and operational characteristics of the reference plants (Korean Standard Nuclear Plant and Korean Next Generation Reactor)were analyzed in order to draw major components, systems, and functions to be satisfied or simulated in the test facility. The test matrix is set up by considering major safety concerns of interest and the test objectives to confirm and enhance the safety of the plants. And the analysis and prioritization of the test matrix leads to the general design requirements of the test facility. Based on the general design requirements, the design criteria is set up for the basic and detailed design of the test facility. And finally it is drawn the design requirements specific to the fluid system and measurement system of the test facility. The test requirements in this report will be used as a guideline to the scaling analysis and basic design of the test facility. The test matrix specified in this report can be modified in the stage of main testing by considering the needs of experiments and circumstances at that time.

Test requirements for the integral effect test to simulate Korean PWR plants

International Nuclear Information System (INIS)

Song, Chul Hwa; Park, C. K.; Lee, S. J.; Kwon, T. S.; Yun, B. J.; Chung, M. K.

2001-02-01

In this report, the test requirements are described for the design of the integral effect test facility to simulate Korean PWR plants. Since the integral effect test facility should be designed so as to simulate various thermal hydraulic phenomena, as closely as possible, to be occurred in real plants during operation or anticipated transients, the design and operational characteristics of the reference plants (Korean Standard Nuclear Plant and Korean Next Generation Reactor)were analyzed in order to draw major components, systems, and functions to be satisfied or simulated in the test facility. The test matrix is set up by considering major safety concerns of interest and the test objectives to confirm and enhance the safety of the plants. And the analysis and prioritization of the test matrix leads to the general design requirements of the test facility. Based on the general design requirements, the design criteria is set up for the basic and detailed design of the test facility. And finally it is drawn the design requirements specific to the fluid system and measurement system of the test facility. The test requirements in this report will be used as a guideline to the scaling analysis and basic design of the test facility. The test matrix specified in this report can be modified in the stage of main testing by considering the needs of experiments and circumstances at that time

Contribution to the study of the conversion PWR type reactors to the thorium cycle

International Nuclear Information System (INIS)

Martins Filho, J.R.

1980-01-01

The use of the thorium cycle in PWR reactors is discussed. The fuel has been calculated in the equilibrium condition for a economic comparison with the uranium cycle (in the same condition). First of all, a code named EQUILIBRIO has been developed for the fuel equilibrium calculation. The results gotten by this code, were introduced in the LEOPARD code for the fuel depletion calculation (in the equilibrium cycle). Same important physics details of fuel depletion are studied, for instance: the neutron balance, power sharing, fuel burnup, etc. The calculations have been done taking as reference the Angra-1 PWR reactor. (Author) [pt

Stress corrosion cracking of steam generator tube and primary pipe in PWR type nuclear power plants

International Nuclear Information System (INIS)

Zhang Weiguo; Gao Fengqin; Zhou Hongyi

1992-03-01

The behavior of stress corrosion cracking (SCC) was studied by slow strain rate test (SSRT), constant load test (CLT) and low frequency cyclic loading test (LFCLT). The purpose of these tests is to get the test data for evaluating the integrity of pressurized boundary of pipes in Qinshan and Guangdong nuclear power plants (NPPs). Tested materials are 316 nuclear grade stainless steel (SS) for primary pipes in welded heat affected zone (WHAZ) and tubes of heat transfer, such as Incoloy-800, Inconel-600 and 321 SS which are used for steam generator in PWR NPPs. The effects of material metallurgy, shot peening treatment, tensile load, strain rate, cyclic load and water chemistry on the behavior of SCC were considered

Stress corrosion cracking of steam generator tube and primary pipe in PWR type nuclear power plants

International Nuclear Information System (INIS)

Zhang Weiguo; Gao Fengqin; Zhou Hongyi

1993-01-01

The behavior of stress corrosion cracking (SCC) is studied by slow strain rate test (SSRT), constant load test (CLT) and low frequency cyclic loading test (LFCLT). The purpose of these tests is to get the test data for evaluating the integrity of pressurized boundary of pipes in Qinshan and Guangdong nuclear power plants. Tested materials are 316 nuclear grade stainless steel (SS) for primary pipes in welded heat affected zone (WHAZ) and steam generator tubes, such as Incoloy-800, Inconel-600, Inconel-690 and 321 SS which are used for steam generator in PWR. The effects of material metallurgy, shot-peening treatment, tensile load, strain rate, cyclic load and water chemistry on the behavior of SCC are investigated

Branding and outcomes in small and medium-sized enterprises (SMEs)

DEFF Research Database (Denmark)

Odoom, Raphael; Narteh, Bedman; Rand, John

2017-01-01

The study investigates the relationships of enterprise resources and branding capabilities with branding efforts and branding benefits. It examines the differential effect of physical resources and branding capabilities on enterprises’ branding efforts and outcomes. Empirical data for the study...... were drawn from 304 small and medium-sized enterprises (SMEs) in Ghana. The hypothesized relationships were analyzed using Structural Equation Modeling. The study found that resources and capabilities possessions might not be enough to produce the optimum branding benefits for enterprises. A better...... result, however, emerges when these resources and capabilities are integrated with well-coordinated branding efforts of the enterprises. The study offers several implications for managers of small businesses based on the findings of the study....

PWR steam generators tube integrity: plugging criteria for PWSCC in roll transition zone

International Nuclear Information System (INIS)

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

1999-01-01

One of the most important causes for tube plugging in PWR (Pressurized Water Reactor) steam generators is the degradation mechanism called Primary Water Stress Corrosion Cracking (PWSCC) in roll transition zone (RTZ) near the tubesheet, mainly for Alloy 600 tubes. To avoid an excessive tube plugging, alternative criteria have been developed based on an approach that consists in withdrawing from service any tube containing a defect for which there is a high probability of a critical size under accident conditions to be reached during next operation cycle. Predictions of the number of tubes to be plugged can be done aiming at preventive maintenance and tube repair, and even a steam generator replacement, without a large and non-planned plant outage. This work presents important aspects related to tube plugging criteria for PWSCC in RTZ based on the risk of break after a leak detection. Calculations of allowable crack length and allowable leak rate for a particular situation are also shown. (author)

Effect of operating conditions and environment on properties of materials of PWR type nuclear power plant components

International Nuclear Information System (INIS)

Vacek, M.

1987-01-01

Operating reliability and service life of PWR type nuclear power plants are discussed with respect to the material properties of the plant components. The effects of the operating environment on the material properties and the methods of their determination are characterized. Discussed are core materials, such as fuel, its cladding and regulating rod materials, and the materials of pipes, steam generators and condensers. The advances in the production of pressure vessel materials and their degradation during operation are treated in great detail. (Z.M.)

Measurement of the residual stresses in a PWR Control Rod Drive Mechanism nozzle

OpenAIRE

Coules, Harry; Smith, David

2018-01-01

Residual stress in the welds that attach Control Rod Drive Mechanism nozzles into the upper head of a PWR reactor vessel can influence the vessel's structural integrity and initiate Primary Water Stress Corrosion Cracking. PWSCC at Alloy 600 CRDM nozzles has caused primary coolant leakage in operating PWRs. We have used Deep Hole Drilling to characterise residual stresses in a PWR vessel head. Measurements of the internal cladding and nozzle attachment weld showed that although modest tensile...

Preliminary study on direct recycling of spent PWR fuel in PWR system

International Nuclear Information System (INIS)

Waris, Abdul; Nuha; Novitriana; Kurniadi, Rizal; Su'ud, Zaki

2012-01-01

Preliminary study on direct recycling of PWR spent fuel to support SUPEL (Straight Utilization of sPEnt LWR fuel in LWR system) scenario has been conducted. Several spent PWR fuel compositions in loaded PWR fuel has been evaluated to obtain the criticality of reactor. The reactor can achieve it criticality for U-235 enrichment in the loaded fresh fuel is at least 4.0 a% with the minimum fraction of the spent fuel in the core is 15.0 %. The neutron spectra become harder with the escalating of U-235 enrichment in the loaded fresh fuel as well as the amount of the spent fuel in the core.

Reactor analysis support package (RASP). Volume 7. PWR set-point methodology. Final report

International Nuclear Information System (INIS)

Temple, S.M.; Robbins, T.R.

1986-09-01

This report provides an overview of the basis and methodology requirements for determining Pressurized Water Reactor (PWR) technical specifications related setpoints and focuses on development of the methodology for a reload core. Additionally, the report documents the implementation and typical methods of analysis used by PWR vendors during the 1970's to develop Protection System Trip Limits (or Limiting Safety System Settings) and Limiting Conditions for Operation. The descriptions of the typical setpoint methodologies are provided for Nuclear Steam Supply Systems as designed and supplied by Babcock and Wilcox, Combustion Engineering, and Westinghouse. The description of the methods of analysis includes the discussion of the computer codes used in the setpoint methodology. Next, the report addresses the treatment of calculational and measurement uncertainties based on the extent to which such information was available for each of the three types of PWR. Finally, the major features of the setpoint methodologies are compared, and the principal effects of each particular methodology on plant operation are summarized for each of the three types of PWR

Instrumentation and control strategies for an integral pressurized water reactor

Directory of Open Access Journals (Sweden)

Belle R. Upadhyaya

2015-03-01

Full Text Available Several vendors have recently been actively pursuing the development of integral pressurized water reactors (iPWRs that range in power levels from small to large reactors. Integral reactors have the features of minimum vessel penetrations, passive heat removal after reactor shutdown, and modular construction that allow fast plant integration and a secure fuel cycle. The features of an integral reactor limit the options for placing control and safety system instruments. The development of instrumentation and control (I&C strategies for a large 1,000 MWe iPWR is described. Reactor system modeling—which includes reactor core dynamics, primary heat exchanger, and the steam flashing drum—is an important part of I&C development and validation, and thereby consolidates the overall implementation for a large iPWR. The results of simulation models, control development, and instrumentation features illustrate the systematic approach that is applicable to integral light water reactors.

Engineering structure design and fabrication process of small sized China helium-cooled solid breeder test blanket module

International Nuclear Information System (INIS)

Wang Zeming; Chen Lu; Hu Gang

2014-01-01

Preliminary design and analysis for china helium-cooled solid breeder (CHHC-SB) test blanket module (TBM) have been carried out recently. As partial verification that the original size module was reasonable and the development process was feasible, fabrication work of a small sized module was to be carried out targetedly. In this paper, detailed design and structure analysis of small sized TBM was carried out based on preliminary design work, fabrication process and integrated assembly process was proposed, so a fabrication for the trial engineering of TBM was layed successfully. (authors)

Evaluation of passive autocatalytic recombiners (PARS) performance for a PWR-konvoi containment type with Gothic 8.1 code

International Nuclear Information System (INIS)

Lopez-Alonso Conty, E.; Papini, D.; Jimenez Varas, G.

2016-01-01

The study presented in this work analyses the evaluation of Passive Autocatalytic Recombiners (PARs) performance for a PWR-Konvoi containment type as a result of an international collaboration between the Paul Scherrer institute (PSI) and the Universidad Politecnica de Madrid (UPM). The implementation study analyzes the size, location and number of the PARs to minimize the risk arising from a hydrogen release and its distribution in the containment building during a hypothetical severe accident. A detailed 3D model of containment was used for the simulations developed for the Gothic 8.1 code. In the first place, the hydrogen preferential pathways and points of hydrogen accumulation were studies and identified starting from the base case scenario without any mitigation measure. The severe accident scenario chosen is a fast release of hydrogen-steam mixture from hot leg creep rupture during SBO (Station Black-Out) accident. Secondly a configuration of PARs was simulated under the same conditions of the unmitigated case. The PAR configuration offered an improvement in the chosen accident scenario, decreasing the hydrogen concentration values below the flammability limit /hydrogen concentration below 7%) in all the containment compartments. (Author)

77 FR 39385 - Receipts-Based, Small Business Size Standard

Science.gov (United States)

2012-07-03

.... The NRC is increasing its receipts-based, small business size standard from $6.5 million to $7 million...-based, small business size standard increasing from $6.5 million to $7.0 million. This adjustment is to... regulatory programs. The NRC is increasing its receipts-based, small business size standard from $6.5 million...

Atmea launches Atmea1 the mid-sized generation 3+ PWR you can rely on

International Nuclear Information System (INIS)

2008-01-01

ATMEA, a daughter company of AREVA NP and Mitsubishi Heavy Industries, is developing and will supply ATMEA1, the most advanced 1100 MWe PWR plant with the combination of the unique set of competence and experience of its parent companies. This folder presents the ATMEA1 reactor main features. (J.S.)

PWR plant transient analyses using TRAC-PF1

International Nuclear Information System (INIS)

Ireland, J.R.; Boyack, B.E.

1984-01-01

This paper describes some of the pressurized water reactor (PWR) transient analyses performed at Los Alamos for the US Nuclear Regulatory Commission using the Transient Reactor Analysis Code (TRAC-PF1). Many of the transient analyses performed directly address current PWR safety issues. Included in this paper are examples of two safety issues addressed by TRAC-PF1. These examples are pressurized thermal shock (PTS) and feed-and-bleed cooling for Oconee-1. The calculations performed were plant specific in that details of both the primary and secondary sides were modeled in addition to models of the plant integrated control systems. The results of these analyses show that for these two transients, the reactor cores remained covered and cooled at all times posing no real threat to the reactor system nor to the public

PWR accident management realated tests: some Bethsy results

International Nuclear Information System (INIS)

Clement, P.; Chataing, T.; Deruaz, R.

1993-01-01

The BETHSY integral test facility which is a scaled down model of a 3 loop FRAMATOME PWR and is currently operated at the Nuclear Center of Grenoble, forms an important part of the French strategy for PWR Accident Management. In this paper the features of both the facility and the experimental program are presented. Two accident transients: a total loss of feedwater and a 2'' cold leg break in case of High Pressure Safety Injection System failure, involving either Event Oriented - or State Oriented-Emergency Operating Procedures (EO-EOP or SO-EOP) are described and the system response analyzed. CATHARE calculation results are also presented which illustrate the ability of this code to adequately predict the key phenomena of these transients. (authors). 13 figs., 11 refs., 2 tabs

Use of standard spectra for the short life radionuclides and ratios for long life radionuclides in the wastes of EDF PWR type reactors

International Nuclear Information System (INIS)

Lantes, B.; Bienvenu, Ph.

2001-01-01

This paper presents the type of declaration of radioactivity in the wastes of PWR type reactors park. Particularly, it insists on the justification of use of spectra for the declaration of short live radionuclides. It tackles the important developments of methods and measures of radiochemical analysis made by the Cea in order to determine the ratios to declare the long life radioisotopes. (N.C.)

Small-size low-temperature scanning tunnel microscope

International Nuclear Information System (INIS)

Al'tfeder, I.B.; Khajkin, M.S.

1989-01-01

A small-size scanning tunnel microscope, designed for operation in transport helium-filled Dewar flasks is described. The microscope design contains a device moving the pin to the tested sample surface and a piezoelectric fine positioning device. High vibration protection of the microscope is provided by its suspension using silk threads. The small-size scanning tunnel microscope provides for atomic resolution

MTR and PWR/PHWR in-pile loop safety in integration with the operation of multipurpose reactor - GAS

International Nuclear Information System (INIS)

Suharno; Aji, Bintoro; Sugiyanto; Rohman, Budi; Zarkasi, Amin S.; Giarno

1998-01-01

MTR and PWR/PHWR In-Pile Loop safety analysis in integration with the operation of Multipurpose Reactor - Gas has been carried out and completed. The assessment is emphasized on the function of the interface systems from the dependence of the operation and the evaluation to the possibility of leakage or failure of the in-pile part inside the reactor pool and reactor core. The analysis is refers to the logic function of the interface system and the possibility of leakage or failure of the in-pile part inside reactor pool and reactor core to consider the integrity of the core qualitatively. The results show that in normal and in transient conditions , the interface system meet the function requirement in safe integrated operation of in-pile loop and reactor. And the results of the possibility analysis of the leakage shows that the possibility based on mechanically assessment is very low and the impact to core integrity is nothing or can be eliminated. The possible position for leakage is on the flen on which one meter above the top level of the core, therefore no influence of leakage to the core

Evaluation of gross tumor size using CT, 18F-FDG PET, integrated 18F-FDG PET/CT and pathological analysis in non-small cell lung cancer

International Nuclear Information System (INIS)

Yu Huiming; Liu Yunfang; Hou Ming; Liu Jie; Li Xiaonan; Yu Jinming

2009-01-01

Purpose: The correlation of gross tumor sizes between combined 18 F-FDG PET/CT images and macroscopic surgical samples has not yet been studied in detail. In the present study, we compared CT, 18 F-FDG PET and combined 18 F-FDG PET/CT for the delineation of gross tumor volume (GTV) and validated the results through examination of the macroscopic surgical specimen. Methods: Fifty-two operable non-small cell lung cancer (NSCLC) patients had integrated 18 F-FDG PET/CT scans preoperatively and pathological examination post-operation. Four separate maximal tumor sizes at X (lateral direction), Y (ventro-dorsal direction) and Z (cranio-caudal direction) axis were measured on 18 F-FDG PET, CT, combined 18 F-FDG PET/CT and surgical specimen, respectively. Linear regression was calculated for each of the three imaging measurements versus pathological measurement. Results: No significant differences were observed among the tumor sizes measured by three images and pathological method. Compared with pathological measurement, CT size at X, Y, Z axis was larger, whereas combined 18 F-FDG PET/CT and 18 F-FDG PET size were smaller. Combined 18 F-FDG PET/CT size was more similar to the pathological size than that of 18 F-FDG PET or CT. Results of linear regressions showed that integrated 18 F-FDG PET/CT was the most accurate modality in measuring the size of cancer. Conclusions: 18 F-FDG PET/CT correlates more faithfully with pathological findings than 18 F-FDG PET or CT. Integrated 18 F-FDG PET/CT is an effective tool to define the target of GTV in radiotherapy.

eLEARNING - A CHANCE FOR SMALL AND MEDIUM SIZED ENTERPRISES

OpenAIRE

Ileana Hamburg; Christiane Lindecke; Judith Terstriep

2005-01-01

eLearning through its flexibility and facility of access is seen as a major enabler of lifelong learning (LLL), as a catalyst of change and a chance for small and medium-sized enterprises (SMEs) to improve their business and to integrate into European market. But so far the eLearning in the context of vocational educational training has been mainly adopted by large enterprises, while only little activity can be observed in SMEs. The question arises what the chances and challenges for SMEs are...

Valve testing for UK PWR safety applications

International Nuclear Information System (INIS)

George, P.T.; Bryant, S.

1989-01-01

Extensive testing and development has been done by the Central Electricity Generating Board (CEGB) to support the design, construction and operation of Sizewell B, the UK's first PWR. A Blowdown Rig for the Assessment of Valve Operability - (BRAVO) has been constructed at the CEGB Marchwood Engineering Laboratory to reproduce PWR Pressurizer fluid conditions for the full scale testing of Pressurizer Relief System (PRS) valves. A full size tandem pair of Pilot Operated Safety Relief Valves (POSRVs) is being tested under the full range of pressurizer fluid conditions. Tests to date have produced important data on the performance of the valve in its Cold Overpressure protection mode of operation and on methods for the in-service testing of the valve. Also, a full size pressurizer safety valve has been tested under full PRS fluid conditions to develop a methodology for the pre-service testing of the Sizewell valves. Further work will be carried out to develop procedures for the in-service testing of the valve. In the Main Steam Safety Valve test program carried out at the Siemens-KWU Test Facilities, a single MSSV from three potential suppliers was tested under full secondary system conditions. The test results have been analyzed and are reflected in the CEGB's arrangements for the pre-service and in-service testing of the Sizewell MSSVs. Valves required to interrupt pipebreak flow must be qualified for this duty by testing or a combination of testing and analysis. To obtain guidance on the performance of such tests gate and globe valves have been subjected to simulated pipebreaks under PWR primary circuit conditions. In the light of problems encountered with gate valve closure under these conditions, further tests are currently being carried out on the BRAVO facility on a gate valve, in preparation for the full scale flow interruption qualification testing of the Sizewell main steam isolation valve

Utilizing Content Marketing in Small and Medium-Sized Organizations

OpenAIRE

Parviainen, Ville

2014-01-01

The major objective of this study is to find out how and to what extent online content is currently utilized for marketing purposes among small and medium-sized organizations in Finland. Additionally, positive and negative future prospects concerning this type of content marketing were explored. The study is mainly qualitative by nature. The empirical part of this thesis was carried out between July 2013 and March 2014 and it consists of five semi-structured interviews with five professio...

Next generation PWR

International Nuclear Information System (INIS)

Tanaka, Toshihiko; Fukuda, Toshihiko; Usui, Shuji

2001-01-01

Development of LWR for power generation in Japan has been intended to upgrade its reliability, safety, operability, maintenance and economy as well as to increase its capacity in order, since nuclear power generation for commercial use was begun on 1970, to steadily increase its generation power. And, in Japan, ABWR (advanced BWR) of the most promising LWR in the world, was already used actually and APWR (advanced PWR) with the largest output in the world is also at a step of its actual use. And, development of the APWR in Japan was begun on 1980s, and is at a step of plan on construction of its first machine at early of this century. However, by large change of social affairs, economy of nuclear power generation is extremely required, to be positioned at an APWR improved development reactor promoted by collaboration of five PWR generation companies and the Mitsubishi Electric Co., Ltd. Therefore, on its development, investigation on effect of change in social affairs on nuclear power stations was at first carried out, to establish a design requirement for the next generation PWR. Here were described on outline, reactor core design, safety concept, and safety evaluation of APWR+ and development of an innovative PWR. (G.K.)

Electrical and control aspects of the Sizewell B PWR

International Nuclear Information System (INIS)

1992-01-01

The pressurized water reactor, Sizewell-B, which is being built in Suffolk is well on in its construction schedule. This conference looked at the electrical and control aspects of the first PWR to be built in the United Kingdom. Although based on the standard Westinghouse PWR design, modifications have been made to meet the particular requirements of the site and the UK licensing regulations. There are 11 papers on all aspects of the electrical systems, 5 papers on the cables and cable installation, 5 on the main control rooms and auxiliary shutdown room, 5 on the integrated system and centralised operation, 6 on the monitoring and protection systems and 9 on the reactor protection systems. All 41 are indexed separately. (UK)

INITIAL PLANETESIMAL SIZES AND THE SIZE DISTRIBUTION OF SMALL KUIPER BELT OBJECTS

International Nuclear Information System (INIS)

Schlichting, Hilke E.; Fuentes, Cesar I.; Trilling, David E.

2013-01-01

The Kuiper Belt is a remnant from the early solar system and its size distribution contains many important constraints that can be used to test models of planet formation and collisional evolution. We show, by comparing observations with theoretical models, that the observed Kuiper Belt size distribution is well matched by coagulation models, which start with an initial planetesimal population with radii of about 1 km, and subsequent collisional evolution. We find that the observed size distribution above R ∼ 30 km is primordial, i.e., it has not been modified by collisional evolution over the age of the solar system, and that the size distribution below R ∼ 30 km has been modified by collisions and that its slope is well matched by collisional evolution models that use published strength laws. We investigate in detail the resulting size distribution of bodies ranging from 0.01 km to 30 km and find that its slope changes several times as a function of radius before approaching the expected value for an equilibrium collisional cascade of material strength dominated bodies for R ∼< 0.1 km. Compared to a single power-law size distribution that would span the whole range from 0.01 km to 30 km, we find in general a strong deficit of bodies around R ∼ 10 km and a strong excess of bodies around 2 km in radius. This deficit and excess of bodies are caused by the planetesimal size distribution left over from the runaway growth phase, which left most of the initial mass in small planetesimals while only a small fraction of the total mass is converted into large protoplanets. This excess mass in small planetesimals leaves a permanent signature in the size distribution of small bodies that is not erased after 4.5 Gyr of collisional evolution. Observations of the small Kuiper Belt Object (KBO) size distribution can therefore test if large KBOs grew as a result of runaway growth and constrained the initial planetesimal sizes. We find that results from recent KBO

Operating function tests of the PWR type RHR pump for engineering safety system under simulated strong ground excitation

International Nuclear Information System (INIS)

Uga, Takeo; Shiraki, Kazuhiro; Homma, Toshiaki; Inazuka, Hisashi; Nakajima, Norifumi.

1979-08-01

Results are described of operating function verification tests of a PWR RHR pump during an earthquake. Of the active reactor components, the PWR residual heat removal pump was chosen from view points of aseismic classification, safety function, structural complexity and past aseismic tests. Through survey of the service conditions and structure of this pump, seismic test conditions such as acceleration level, simulated seismic wave form and earthquake duration were decided for seismicity of the operating pump. Then, plans were prepared to evaluate vibration chracteristics of the pump and to estimate its aseismic design margins. Subsequently, test facility and instrumentation system were designed and constructed. Experimental results could thus be acquired on vibration characteristics of the pump and its dynamic behavior during different kinds and levels of simulated earthquake. In conclusion: (1) Stiffeners attached to the auxiliary system piping do improve aseismic performance of the pump. (2) The rotor-shaft-bearing system is secure unless it is subjected to transient disturbunces having high frequency content. (3) The motor and pump casing having resonance frequencies much higher than frequency content of the seismic wave show only small amplifications. (4) The RHR pump possesses an aseismic design margin more than 2.6 times the expected ultimate earthquake on design basis. (author)

Performance of high burned PWR fuel during transient

International Nuclear Information System (INIS)

Yanagisawa, Kazuaki; Fujishiro, Toshio

1992-01-01

In a majority of Japanese light water type commercial powder reactors (LWRs), UO 2 pellet sheathed by zircaloy cladding is used. Licensed discharged burn-up of the PWR fuel rod is going to be increased from 39 MWd/kgU to 48 MWd/kgU. This requests the increased reliability of cladding material as a strong barrier against fission product (FP). A long time usage in the neutron field and in the high temperature coolant will cause the zircaloy hardening and embrittlement. The cladding material is also degraded by waterside corrosion. These degradations are enhanced much by increased burn-up. A increased magnitude of the pellet-cladding mechanical interaction (PCMI) is of importance for increasing the stress of cladding material. In addition, aggressive FPs released from the fuel tends to attack the cladding material to cause stress corrosion cracking (SCC). At the Nuclear Safety Research Reactor (NSRR) in JAERI, 14 x 14 PWR type fuel rods preirradiation up to 42 MWd/kgU was prepared for the transient pulse irradiation under the simulated reactivity initiated accident (RIA) conditions. This will cause a prompt increase of the fuel temperature and stress on the highly burned cladding material. In the present paper, steady-state and transient behavior observed from the tested PWR fuel rod and calculational results obtained from the computer code FPRETAIN will be described. (author)

Generation of SCALE 6 Input Data File for Cross Section Library of PWR Spent Fuel

International Nuclear Information System (INIS)

Jeong, Chang Joon; Cho, Dong Keun

2010-11-01

In order to obtain the cross section libraries of the Korean Pressurized water reactor (PWR) spent fuel (SF), SCALE 6 code input files have been generated. The PWR fuel data were obtained from the nuclear design report (NDR) of the current operating PWRs. The input file were prepared for 16 fuel types such as 4 types of Westinghouse 14x14, 3 types of OPR-1000 16x16, 4 types of Westinghouse 16x16, and 6 types of Westinghouse 17x17. For each fuel type, 5 kinds of fuel enrichments have been considered such as 1.5, 2.0 ,3.0, 4.0 and 5.0 wt%. In the SCALE 6 calculation, a ENDF-V 44 group was used. The 25 burnup step until 72000 MWD/T was used. A 1/4 symmetry model was used for 16x16 and 17x17 fuel assembly, and 1/2 symmetry model was used for 14x14 fuel assembly The generated cross section libraries will be used for the source-term analysis of the PWR SF

Probes for inspections of heat exchanges installed at nuclear power plants type PWR by eddy current method

International Nuclear Information System (INIS)

Silva, Alonso F.O.

2007-01-01

From all non destructive examination methods usable to perform integrity evaluation of critical equipment installed at nuclear power plants (NPP), eddy current test (ET) may be considered the most important one, when examining heat exchangers. For its application, special probes and reference calibration standards are employed. In pressurized water reactor (PWR) NPPs, a particularly critical equipment is the steam generator (SG), a huge heat exchanger that contains thousands of U-bend thin wall tubes. Due to its severe working conditions (pressure and temperature), that component is periodically examined by means of ET. In this paper a revision of the operating fundamentals of the main ET probes, used to perform SG inspections is presented. (author)

Condensate polishing guidelines for PWR and BWR plants

International Nuclear Information System (INIS)

Robbins, P.; Crinigan, P.; Graham, B.; Kohlmann, R.; Crosby, C.; Seager, J.; Bosold, R.; Gillen, J.; Kristensen, J.; McKeen, A.; Jones, V.; Sawochka, S.; Siegwarth, D.; Keeling, D.; Polidoroff, T.; Morgan, D.; Rickertsen, D.; Dyson, A.; Mills, W.; Coleman, L.

1993-03-01

Under EPRI sponsorship, an industry committee, similar in form and operation to other guideline committees, was created to develop Condensate Polishing Guidelines for both PWR and BWR systems. The committee reviewed the available utility and water treatment industry experience on system design and performance and incorporated operational and state-of-the-art information into document. These guidelines help utilities to optimize present condensate polisher designs as well as be a resource for retrofits or new construction. These guidelines present information that has not previously been presented in any consensus industry document. The committee generated guidelines that cover both deep bed and powdered resin systems as an integral part of the chemistry of PWR and BWR plants. The guidelines are separated into sections that deal with the basis for condensate polishing, system design, resin design and application, data management and performance and management responsibilities

TRANSPORT CHARACTERISTICS OF SELECTED PWR LOCA GENERATED DEBRIS

International Nuclear Information System (INIS)

MAJI, A. K.; MARSHALL, B.

2000-01-01

In the unlikely event of a Loss of Coolant Accident (LOCA) in a pressurized water reactor (PWR), break jet impingement would dislodge thermal insulation FR-om nearby piping, as well as other materials within the containment, such as paint chips, concrete dust, and fire barrier materials. Steam/water flows induced by the break and by the containment sprays would transport debris to the containment floor. Subsequently, debris would likely transport to and accumulate on the suction sump screens of the emergency core cooling system (ECCS) pumps, thereby potentially degrading ECCS performance and possibly even failing the ECCS. In 1998, the U. S. Nuclear Regulatory Commission (NRC) initiated a generic study (Generic Safety Issue-191) to evaluate the potential for the accumulation of LOCA related debris on the PWR sump screen and the consequent loss of ECCS pump net positive suction head (NPSH). Los Alamos National Laboratory (LANL), supporting the resolution of GSI-191, was tasked with developing a method for estimating debris transport in PWR containments to estimate the quantity of debris that would accumulate on the sump screen for use in plant specific evaluations. The analytical method proposed by LANL, to predict debris transport within the water that would accumulate on the containment floor, is to use computational fluid dynamics (CFD) combined with experimental debris transport data to predict debris transport and accumulation on the screen. CFD simulations of actual plant containment designs would provide flow data for a postulated accident in that plant, e.g., three-dimensional patterns of flow velocities and flow turbulence. Small-scale experiments would determine parameters defining the debris transport characteristics for each type of debris. The containment floor transport methodology will merge debris transport characteristics with CFD results to provide a reasonable and conservative estimate of debris transport within the containment floor pool and

Simplified model for the thermo-hydraulic simulation of the hot channel of a PWR type nuclear reactor; Modelo simplificado para simulacao do comportamento termohidraulico do canal quente de reator nuclear do tipo PWR

Energy Technology Data Exchange (ETDEWEB)

Belem, J A.T.

1993-09-01

The present work deals with the thermal-hydraulic analysis of the hot channel of a standard PWR type reactor utilizing a simplified mathematical model that considers constant the water mass flux during single-phase flow and reduction of the flow when the steam quality is increasing in the channel (two-phase flow). The model has been applied to the Angra-1 reactor and it has proved satisfactory when compared to other ones. (author). 25 refs, 15 figs, 3 tabs.

PWR vessel flaw distribution development

International Nuclear Information System (INIS)

Rosinski, S.T.; Kennedy, E.L.; Foulds, J.R.; Kinsman, K.M.

1990-01-01

This paper reports on PWR pressure vessels which operate under NRC rules and regulatory guides intended to prevent failure of the vessels. Plants failing to meet the operating criteria specified under these rules and regulations are required to analytically demonstrate fitness for service in order to continue operation. The initial flaw size or distribution of initial vessel flaws is a key input to the required vessel integrity analyses. However, the flaw distribution assumed in the development of the NRC Regulations and recommended for the plant specific analyses is potentially over-conservative. This is because the distribution is based on the limited amount of vessel inspection data available at the time the criteria were being developed and does not take full advantage of the more recent and reliable domestic vessel inspection results. The U.S. Department of Energy is funding an effort through Sandia National Laboratories to investigate the possibility of developing a new flaw distribution based on the increased amount and improved reliability of domestic vessel inspection data. Results of Phase I of the program indicate that state-of-the-art NDE systems' capabilities are sufficient for development of a new flaw distribution that could ultimately provide life extension benefits over the presently required operating practice

Open and closed fuel cycle of HWR and PWR. How large is the high-level radioactive wastes repository; Ciclos de combustible abierto y cerrado con HWR y PWR. ?Cuanto mas grande es el repositorio de residuos radiactivos de alta actividad?

Energy Technology Data Exchange (ETDEWEB)

Bevilacqua, Arturo M. [Comision Nacional de Energia Atomica, San Carlos de Bariloche (Argentina). Centro Atomico Bariloche

1996-07-01

A conceptual analysis was carried out on the size of a high-level wastes (HLW) repository for the waste arising from once-through and closed fuel cycles with (HLW) and PWR. The mass, the activity and thermal loading was calculated with the ORIGEN2.1 computer code for the spent fuel and for the high-level liquid wastes. It was considered a minimum burnup of 7.000 MW.d/t U and 33.000 MW.d/t U for HWR and PWR respectively, cooling times of 20 and 55 years, reprocessing recovery ratios of 99% and 99.7% and a total electricity production of 81.6 GW(e).a. It was concluded that the cooling time is the most important repository size reproduction parameter for the closed cycles. On the other hand, the spent fuel mass for the once-through cycles does not depend on the cooling time what prevents repository size reduction once a cooling time of 55 years is reached. The repository size reduction in the case of HWR is larger than in the case of PWR, owing to the larger fuel mass required to produce the specific electricity amount. (author)

For the Good of All : Integrating corporate social responsibility into the business strategy of small and medium sized enterprises in Finland: CASE: Oy Vallila Interior Ab

OpenAIRE

Anttila, Hannele

2010-01-01

Corporate social responsibility (CSR) can be defined as a concept that includes simultaneous fulfilment of a company’s economic, legal, ethical and philanthropic responsibilities, as well as companies integrating social and environmental concerns in their business operations and in their interaction with their stakeholders on a voluntary basis. The purpose of this dissertation is to investigate CSR from the perspective of small and medium sized enterprises (SMEs), more specifically from t...

PWR secondary water chemistry guidelines

International Nuclear Information System (INIS)

Bell, M.J.; Blomgren, J.C.; Fackelmann, J.M.

1982-10-01

Steam generators in pressurized water reactor (PWR) nuclear power plants have experienced tubing degradation by a variety of corrosion-related mechanisms which depend directly on secondary water chemistry. As a result of this experience, the Steam Generator Owners Group and EPRI have sponsored a major program to provide solutions to PWR steam generator problems. This report, PWR Secondary Water Chemistry Guidelines, in addition to presenting justification for water chemistry control parameters, discusses available analytical methods, data management and surveillance, and the management philosophy required to successfully implement the guidelines

Contribution to study and design of PWR plant simulation code

International Nuclear Information System (INIS)

Delourme, Didier.

1980-11-01

This paper presents an improvement of PICOLO, a package for PWR plants simulation. Its describes principally the integration to the code of a primary loop and pressurizer model and the corresponding control loops. Fast transients are tested on the packages and results are compared with real transients obtained on plants [fr

Conceptual study on advanced PWR system

International Nuclear Information System (INIS)

Bae, Yoon Young; Chang, M. H.; Yu, K. J.; Lee, D. J.; Cho, B. H.; Kim, H. Y.; Yoon, J. H.; Lee, Y. J.; Kim, J. P.; Park, C. T.; Seo, J. K.; Kang, H. S.; Kim, J. I.; Kim, Y. W.; Kim, Y. H.

1997-07-01

In this study, the adoptable essential technologies and reference design concept of the advanced reactor were developed and related basic experiments were performed. 1) Once-through Helical Steam Generator: a performance analysis computer code for heli-coiled steam generator was developed for thermal sizing of steam generator and determination of thermal-hydraulic parameters. 2) Self-pressurizing pressurizer : a performance analysis computer code for cold pressurizer was developed. 3) Control rod drive mechanism for fine control : type and function were surveyed. 4) CHF in passive PWR condition : development of the prediction model bundle CHF by introducing the correction factor from the data base. 5) Passive cooling concepts for concrete containment systems: development of the PCCS heat transfer coefficient. 6) Steam injector concepts: analysis and experiment were conducted. 7) Fluidic diode concepts : analysis and experiment were conducted. 8) Wet thermal insulator : tests for thin steel layers and assessment of materials. 9) Passive residual heat removal system : a performance analysis computer code for PRHRS was developed and the conformance to EPRI requirement was checked. (author). 18 refs., 55 tabs., 137 figs

Conceptual study on advanced PWR system

Energy Technology Data Exchange (ETDEWEB)

Bae, Yoon Young; Chang, M H; Yu, K J; Lee, D J; Cho, B H; Kim, H Y; Yoon, J H; Lee, Y J; Kim, J P; Park, C T; Seo, J K; Kang, H S; Kim, J I; Kim, Y W; Kim, Y H

1997-07-01

In this study, the adoptable essential technologies and reference design concept of the advanced reactor were developed and related basic experiments were performed. (1) Once-through Helical Steam Generator: a performance analysis computer code for heli-coiled steam generator was developed for thermal sizing of steam generator and determination of thermal-hydraulic parameters. (2) Self-pressurizing pressurizer : a performance analysis computer code for cold pressurizer was developed. (3) Control rod drive mechanism for fine control : type and function were surveyed. (4) CHF in passive PWR condition : development of the prediction model bundle CHF by introducing the correction factor from the data base. (5) Passive cooling concepts for concrete containment systems: development of the PCCS heat transfer coefficient. (6) Steam injector concepts: analysis and experiment were conducted. (7) Fluidic diode concepts : analysis and experiment were conducted. (8) Wet thermal insulator : tests for thin steel layers and assessment of materials. (9) Passive residual heat removal system : a performance analysis computer code for PRHRS was developed and the conformance to EPRI requirement was checked. (author). 18 refs., 55 tabs., 137 figs.

Modern methods of evaluation existing suppliers and suppliers selected by customer for small and medium-sized companies

Directory of Open Access Journals (Sweden)

Marie Jurová

2010-01-01

Full Text Available On existing conditions on global market (almost identical quality, almost identical purchasing values companies have to definite the total rating of importance of evaluative criteria. In post-crisis period the problem of suppliers’ evaluation is one of the biggest, because companies had to use all resources and all possibilities to develop their own business. Many authors wrote about collaborative planning and warehousing as one of possibilities to optimization work inside supply chain. For this paper define small and medium-sized enterprises (SME as enterprises with the size of maximum 250 em­ployers. In literature was read we couldn’t find information about own system of suppliers evaluation for small and medium-sized enterprises. SME can have some different types of business and in this case need the universal system of eva­lua­tion. The research of this paper is oriented on small and medium-sized enterprises with different types of business. The new theoretical universal method of calculation for evaluation existing suppliers for small and medium-sized enterprises will present in this paper. This theoretical method is based on average values. This method includes traditional evaluative criteria (quality, delivery time and other (mobility of supplier, possibilities of new level of partnership. This method of evaluation can not be used for continual manufacture. New method can improve the total evaluation of supplier in small and medium-sized enterprises.

AGR v PWR

International Nuclear Information System (INIS)

Green, D.

1986-01-01

When the Central Electricity Generating Board (CEGB) invited tenders and placed a contract for the Advanced Gas Cooled Reactor (AGR) at Dungeness B in 1965 -preferring it to the Pressurised Water Reactor (PWR) -the AGR was lamentably ill developed. The effects of the decision were widely felt, for it took the British nuclear industry off the light water reactor highway of world reactor business and up and idiosyncratic private highway of its own, excluding it altogether from any material export business in the two decades which followed. Yet although the UK may have made wrong decisions in rejecting the PWR in 1965, that does not mean that it can necessarily now either correct them, or redeem their consequence, by reversing the choice in 1985. In the 20 years since 1965 the whole world economic and energy picture has been transformed and the national picture with it. Picking up the PWR now could prove as big a disaster as rejecting it may have been in 1965. (author)

Industrial assessment of nonbackfittable PWR design modifications. Final report

International Nuclear Information System (INIS)

Matzie, R.A.; Daleas, R.S.; Miller, D.D.

1980-11-01

As part of the US Department of Energy's Advanced Reactor Design Study, various nonbackfittable PWR design modifications were evaluated to determine their potential for improved uranium utilization and commercial viability. Combustion Engineering, Inc. contributed to this effort through participation in the Battelle Pacific Northwest Laboratory industrial assessment of such design modifications. Seven modifications, including the use of higher primary system temperatures and pressures, rapid-frequent refueling, end-of-cycle stretchout, core periphery modifications, radial blankets, low power density cores, and small PWR assemblies, were evaluated with respect to uranium utilization, economics, technical and operational complexity, and several other subjective considerations. Rapid-frequent refueling was judged to have the highest potential although it would probably not be economical for the majority of reactors with the design assumptions used in this assessment

Bounding the conservatism in flaw-related variables for pressure vessel integrity analyses

International Nuclear Information System (INIS)

Foulds, J.R.; Kennedy, E.L.

1993-01-01

The fracture mechanics-based integrity analysis of a pressure vessel, whether performed deterministically or probabilistically, requires use of one or more flaw-related input variables, such as flaw size, number of flaws, flaw location, and flaw type. The specific values of these variables are generally selected with the intent to ensure conservative predictions of vessel integrity. These selected values, however, are largely independent of vessel-specific inspection results, or are, at best, deduced by ''conservative'' interpretation of vessel-specific inspection results without adequate consideration of the pertinent inspection system performance (reliability). In either case, the conservatism associated with the flaw-related variables chosen for analysis remains examination (NDE) technology and the recently formulated ASME Code procedures for qualifying NDE system capability and performance (as applied to selected nuclear power plant components) now provides a systematic means of bounding the conservatism in flaw-related input variables for pressure vessel integrity analyses. This is essentially achieved by establishing probabilistic (risk)-based limits on the assigned variable values, dependent upon the vessel inspection results and on the inspection system unreliability. Described herein is this probabilistic method and its potential application to: (i) defining a vessel-specific ''reference'' flaw for calculating pressure-temperature limit curves in the deterministic evaluation of pressurized water reactor (PWR) reactor vessels, and (ii) limiting the flaw distribution input to a PWR reactor vessel-specific, probabilistic integrity analysis for pressurized thermal shock loads

Application of the perturbation theory-differential formalism-for sensitivity analysis in steam generators of PWR type nuclear power plants

International Nuclear Information System (INIS)

Sanders, R.M.G.; Andrade Lima, F.R. de; Alvim, A.C.M.

1987-06-01

An homogeneous model which simulates the stationary behavior of steam generators of PWR type reactors and uses the differential formalism of perturbation theory for analysing sensibility of linear and non-linear responses, is presented. The PERGEVAP computer code to calculate the temperature distribution in the steam generator and associated importance function, is developed. The code also evaluates effects of the thermohydraulic parameter variation on selected functionals. The obtained results are compared with results obtained by GEVAP computer code . (M.C.K.) [pt

Reassessment of PWR pressure-vessel integrity during overcooling accidents

International Nuclear Information System (INIS)

Cheverton, R.D.; Ball, D.G.

1983-01-01

A continuing analysis of the PTS problem associated with PWR postuated OCA's indicates that the previously accepted degree of conservatism in the fracture-mechanics model needs to be more closely evaluated, and if excessive, reducted. One feature that was believed to be conservative was the use of two-dimensional as opposed to finite-length (three-dimensional) flaws. A flaw of particular interest is one that is located in an axial weld of a plate-type vessel. For those vessels that suffer relatively high radiation damage in the welds, the length of the flaw will be no greater than the length of the weld, and recent calculations indicate that a deep flaw of that length (approx. 2 m) is not effectively infinitely long, contrary to previous thinking. The benefit to be derived from consideration of the 2-m flaw and also a semielliptical flaw with a length-to-depth ratio of 6/1 was investigated by analyzing several postulated transients. In doing so the sensitivity of the benefit to a specified maximum crack arrest toughness and to the duration of the transient was investigated. Results of the analysis indicate that for some conditions the benefit in using the 2-m flaw is substantial, but it decreases with increasing pressure, and above a certain pressure there may be no benefit, depending on the duration of the transient and the limit on crack arrest toughness

77 FR 39442 - Receipts-Based, Small Business Size Standard

Science.gov (United States)

2012-07-03

... RIN 3150-AJ14 [NRC-2012-0062] Receipts-Based, Small Business Size Standard AGENCY: Nuclear Regulatory... Regulatory Flexibility Act of 1980, as amended. The NRC is proposing to increase its receipts-based, small business size standard from $6.5 million to $7 million to conform to the standard set by the Small Business...

78 FR 77334 - Small Business Size Standards: Construction

Science.gov (United States)

2013-12-23

... enrollment in the System of Award Management's (SAM) Dynamic Small Business Search database, and more firms... SMALL BUSINESS ADMINISTRATION 13 CFR Part 121 RIN 3245-AG37 Small Business Size Standards: Construction AGENCY: U.S. Small Business Administration. ACTION: Final rule. SUMMARY: The United States Small...

PWR burnable absorber evaluation

International Nuclear Information System (INIS)

Cacciapouti, R.J.; Weader, R.J.; Malone, J.P.

1995-01-01

The purpose of the study was to evaluate the relative neurotic efficiency and fuel cycle cost benefits of PWR burnable absorbers. Establishment of reference low-leakage equilibrium in-core fuel management plans for 12-, 18- and 24-month cycles. Review of the fuel management impact of the integral fuel burnable absorber (IFBA), erbium and gadolinium. Calculation of the U 3 O 8 , UF 6 , SWU, fuel fabrication, and burnable absorber requirements for the defined fuel management plans. Estimation of fuel cycle costs of each fuel management plan at spot market and long-term market fuel prices. Estimation of the comparative savings of the different burnable absorbers in dollar equivalent per kgU of fabricated fuel. (author)

PWR degraded core analysis

International Nuclear Information System (INIS)

Gittus, J.H.

1982-04-01

A review is presented of the various phenomena involved in degraded core accidents and the ensuing transport of fission products from the fuel to the primary circuit and the containment. The dominant accident sequences found in the PWR risk studies published to date are briefly described. Then chapters deal with the following topics: the condition and behaviour of water reactor fuel during normal operation and at the commencement of degraded core accidents; the generation of hydrogen from the Zircaloy-steam and the steel-steam reactions; the way in which the core deforms and finally melts following loss of coolant; debris relocation analysis; containment integrity; fission product behaviour during a degraded core accident. (U.K.)

Investigation of small and modular-sized fast reactor

International Nuclear Information System (INIS)

Kubota, Kenichi; Kawasaki, Nobuchika; Umetsu, Yoichiro; Akatsu, Minoru; Kasai, Shigeo; Konomura, Mamoru; Ichimiya, Masakazu

2000-06-01

In this paper, feasibility of the multipurpose small fast reactor, which could be used for requirements concerned with various utilization of electricity and energy and flexibility of power supply site, is discussed on the basis of examination of literatures of various small reactors. And also, a possibility of economic improvement by learning effect of fabrication cost is discussed for the modular-sized reactor which is expected to be a base load power supply system with lower initial investment. (1) Multipurpose small reactor (a) The small reactor with 10MWe-150MWe has a potential as a power source for large co-generation, a large island, a middle city, desalination and marine use. (b) Highly passive mechanism, long fuel exchange interval, and minimized maintenance activities are required for the multipurpose small reactor design. The reactor has a high potential for the long fuel exchange interval, since it is relatively easy for FR to obtain a long life core. (c) Current designs of small FRs in Japan and USA (NERI Project) are reviewed to obtain design requirements for the multipurpose small reactor. (2) Modular-sized reactor (a) In order that modular-sized reactor could be competitive to 3200MWe twin plant (two large monolithic reactor) with 200kyenWe, the target capital cost of FOAK is estimated to be 260kyen/yenWe for 800MWe modular, 280kyen/yenWe for 400MWe modular and 290kyen/yenWe for 200MWe by taking account of the leaning effect. (b) As the result of the review on the current designs of modular-sized FRs in Japan and USA (S-PRISM) from the viewpoint of economic improvement, since it only be necessary to make further effort for the target capital cost of FOAK, since the modular-sized FRs requires a large amount of material for shielding, vessels and heat exchangers essentially. (author)

76 FR 27935 - Small Business Size Standards: Transportation and Warehousing

Science.gov (United States)

2011-05-13

... assistance, SBA establishes small business definitions (referred to as size standards) for private sector... SMALL BUSINESS ADMINISTRATION 13 CFR Part 121 RIN 3245-AG08 Small Business Size Standards: Transportation and Warehousing AGENCY: U.S. Small Business Administration. ACTION: Proposed rule. SUMMARY: The U...

Chironomid midges (Diptera, chironomidae) show extremely small genome sizes.

Science.gov (United States)

Cornette, Richard; Gusev, Oleg; Nakahara, Yuichi; Shimura, Sachiko; Kikawada, Takahiro; Okuda, Takashi

2015-06-01

Chironomid midges (Diptera; Chironomidae) are found in various environments from the high Arctic to the Antarctic, including temperate and tropical regions. In many freshwater habitats, members of this family are among the most abundant invertebrates. In the present study, the genome sizes of 25 chironomid species were determined by flow cytometry and the resulting C-values ranged from 0.07 to 0.20 pg DNA (i.e. from about 68 to 195 Mbp). These genome sizes were uniformly very small and included, to our knowledge, the smallest genome sizes recorded to date among insects. Small proportion of transposable elements and short intron sizes were suggested to contribute to the reduction of genome sizes in chironomids. We discuss about the possible developmental and physiological advantages of having a small genome size and about putative implications for the ecological success of the family Chironomidae.

The causal effect of board size in the performance of small and medium-sized firms

DEFF Research Database (Denmark)

Bennedsen, Morten; Kongsted, Hans Christian; Meisner Nielsen, Kasper

2008-01-01

correlation between family size and board size and show this correlation to be driven by firms where the CEO's relatives serve on the board. Second, we find empirical evidence of a small adverse board size effect driven by the minority of small and medium-sized firms that are characterized by having......Empirical studies of large publicly traded firms have shown a robust negative relationship between board size and firm performance. The evidence on small and medium-sized firms is less clear; we show that existing work has been incomplete in analyzing the causal relationship due to weak...... identification strategies. Using a rich data set of almost 7000 closely held corporations we provide a causal analysis of board size effects on firm performance: We use a novel instrument given by the number of children of the chief executive officer (CEO) of the firms. First, we find a strong positive...

IRIS-50. A 50 MWe advanced PWR design for smaller, regional grids and specialized applications

International Nuclear Information System (INIS)

Petrovic, Bojan; Carelli, Mario; Conway, Larry; Hundal, Rolv; Barbaso, Enrico; Gamba, Federica; Centofante, Mario

2009-01-01

IRIS is an advanced, medium-power (1000 MWt or ∼335 MWe) advanced PWR design of integral configuration, that has gained wide recognition due to its innovative 'safety-by-design' safety approach. In spite of its smaller size compared to large monolithic nuclear power plants, it is economically competitive due to its simplicity and advantages of modular deployment. However, the optimum power level for a class of specific applications (e.g., power generation in small regional isolated grids; water desalination and biodiesel production at remote locations; autonomous power source for special applications, etc.) may be even lower, of the order of tens rather than hundreds of MWe. The simple and robust IRIS 335 MWe design provides a solid basis for establishing a 20-100 MWe design, utilizing the same safety and economics principles, so that it will retain economic attractiveness compared to other alternatives of the same power level. A conceptual 50 MWe design, IRIS-50, was initially developed and then assessed in a 2001 report to the US Congress on small and medium reactors, as a design mature enough to have deployment potential within a decade. In the meantime, while the main efforts have focused on the 335 MWe design completion and licensing, parallel efforts have progressed toward the preliminary design of IRIS-50. This paper summarizes the main IRIS-50 features and presents an update on its design status. (author)

Minor actinide transmutation on PWR burnable poison rods

International Nuclear Information System (INIS)

Hu, Wenchao; Liu, Bin; Ouyang, Xiaoping; Tu, Jing; Liu, Fang; Huang, Liming; Fu, Juan; Meng, Haiyan

2015-01-01

Highlights: • Key issues associated with MA transmutation are the appropriate loading pattern. • Commercial PWRs are the only choice to transmute MAs in large scale currently. • Considerable amount of MA can be loaded to PWR without disturbing k eff markedly. • Loading MA to PWR burnable poison rods for transmutation is an optimal loading pattern. - Abstract: Minor actinides are the primary contributors to long term radiotoxicity in spent fuel. The majority of commercial reactors in operation in the world are PWRs, so to study the minor actinide transmutation characteristics in the PWRs and ultimately realize the successful minor actinide transmutation in PWRs are crucial problem in the area of the nuclear waste disposal. The key issues associated with the minor actinide transmutation are the appropriate loading patterns when introducing minor actinides to the PWR core. We study two different minor actinide transmutation materials loading patterns on the PWR burnable poison rods, one is to coat a thin layer of minor actinide in the water gap between the zircaloy cladding and the stainless steel which is filled with water, another one is that minor actinides substitute for burnable poison directly within burnable poison rods. Simulation calculation indicates that the two loading patterns can load approximately equivalent to 5–6 PWR annual minor actinide yields without disturbing the PWR k eff markedly. The PWR k eff can return criticality again by slightly reducing the boric acid concentration in the coolant of PWR or removing some burnable poison rods without coating the minor actinide transmutation materials from PWR core. In other words, loading minor actinide transmutation material to PWR does not consume extra neutron, minor actinide just consumes the neutrons which absorbed by the removed control poisons. Both minor actinide loading patterns are technically feasible; most importantly do not need to modify the configuration of the PWR core and

The impact of business type and sector, location and annual turnover on competitive intelligence practice of small and medium-sized enterprises

Directory of Open Access Journals (Sweden)

Tshilidzi Eric Nenzhelele

2015-11-01

Full Text Available Small and Medium-sized Enterprises (SMEs contribute much to job creation, poverty eradication, economic development and growth. Countries rely on SMEs for wealth creation and social wellbeing of their citizens. Moreover, SMEs contribute to the global competitiveness of a country. For SMEs to continue offering all these benefits to the country, they must be competitive in the midst of global competitive pressure. To gain competitive advantage, enterprises of different sizes practice Competitive Intelligence (CI. While CI practise has been widely researched in large enterprise, there is lack of CI practice research in SMEs. This research establishes the impact of business type and sector, location and annual turnover on the competitive intelligence practise of SMEs. The research was quantitative in nature and a self-administered questionnaire was used to collect data from owners/managers of SMEs.

76 FR 62313 - Small Business Size and Status Integrity

Science.gov (United States)

2011-10-07

... business concern in the Online Representations and Certifications Application (ORCA) database (or any... Veteran-Owned business concern in the Online Representations and Certifications Application (ORCA... SMALL BUSINESS ADMINISTRATION 13 CFR Parts 121, 124, 125, 126, 127 RIN 3245-AG23 Small Business...

Aspects Regarding the Development and the Integration of the Corporate Social Responsibility Concept in Firms’ Behaviour. Particularities for Small and Medium-sized Enterprises

Directory of Open Access Journals (Sweden)

Georgeta Vintila

2009-07-01

Full Text Available Starting from the concept of Corporate Social Responsibility (CSR, the purpose of this article is to highlight some of the particularities, at Small and Medium-sized Enterprises (SMEs level, regarding: socially responsible behaviour and participation in actions related to CSR; type of CSR strategies adopted by SMEs as compared to larger firms; the motivations, benefits and constraints of the social involvement. The ultimate question refers to the existence, nature and implications of a link between the firm size and the aspects mentioned above. These aspects will be presented according to the conclusions of the previous studies conducted in this area of research, the present paper representing therefore a theoretical synthesis of the existant literature.

SMALL AND MEDIUM-SIZED ENTERPRISES IN REPUBLIC OF MOLDOVA

Directory of Open Access Journals (Sweden)

Igor G. SÎRODOEV

2009-12-01

Full Text Available The paper examines the role of SME sector in Moldova’s economy and its likely implications in the RD processes. There are emphasized several key features of the relationship between knowledge economy, RD and SMEs, and their effect on regional and local communities. Then, a summary of RD particular features in Moldova is presented.The role of SME sector, examined in terms of its performance by ownership and activity types, as a whole and by main firms’ categories (micro-, small, and medium-sized is analyzed. Small enterprises have been considered as the most adequate solution forpromoting RD in general case. National-level analysis is completed by regional insight into the sector as a whole. Finally, some critical aspects of SMEs’ role in RD in Moldova are discussed.

Dynamical analysis of a PWR internals using super-elements in an integrated 3-D model model. Part 1: model description and static tests

International Nuclear Information System (INIS)

Jesus Miranda, C.A. de.

1992-01-01

An integrated 3-D model of a research PWR reactor core support internals structures was developed for its dynamic analyses. The static tests for the validation of the model are presented. There are about 90 super-elements with, approximately, 85000 degrees of freedom (DoF), 8200 masters DoF, 12000 elements with about 8400 thin shell elements. A DEC VAX computer 11/785 model and the ANSYS program were used. If impacts occurs the spectral seismic analysis will be changed to a non-linear one with direct integration of the displacement pulse derived from the seismic accelerogram. This last will be obtained from the seismic acceleration response spectra. (author)

Assessment of TRAC-PF1/MOD1 code for large break LOCA in PWR

International Nuclear Information System (INIS)

Akimoto, Hajime; Ohnuki, Akira; Murao, Yoshio; Abe, Yutaka.

1993-03-01

As the first step of the REFLA/TRAC code development, the TRAC/PF1/MOD1 code has been assessed for various experiments that simulate postulated large-break loss-of-coolant accident (LBLOCA) in PWR to understand the predictive capability and to identify the problem areas of the code. The assessment calculations were performed for separate effect tests for critical flow, counter current flow, condensation at cold leg and reflood as well as integral tests to understand predictability for individual phenomena. This report summarizes results from the assessment calculations of the TRAC-PF1/MOD1 code for LBLOCA in PWR. The assessment calculations made clear the predictive capability and problem areas of the TRAC-PF1/MOD1 code for LBLOCA in PWR. The areas, listed below, should be improved for more realistic and effective simulation of LBLOCA in PWR: (1) core heat transfer model during blowdown, (2) ECC bypass model at downcomer during refill, (3) condensation model during accumulator injection, and (4) core thermal hydraulic model during reflood. (author) 57 refs

Aerosols behavior inside a PWR during an accident

International Nuclear Information System (INIS)

Hervouet, C.

1983-01-01

During very hypothetical accidents occurring in a pressurized water ractor, radioactive aerosols can be released, during core-melt, inside the reactor containment building. A good knowledge of their behavior in the humid containment atmosphere (mass concentration and size distribution) is essential in order to evaluate their harmfulness in case of environment contamination and to design possible filtration devices. Accordingly the Safety Analysis Department of the Atomic Energy Commission uses several computer models, describing the particle formation (BOIL/MARCH), then behavior in the primary circuits (TRAP-MELT), and in the reactor containment building (AEROSOLS-PARFDISEKO-III B). On the one hand, these models have been improved, in particular the one related to the aerosol formation (nature and mass of released particles) using recent experimental results. On the other hand, sensitivity analyses have been performed with the AEROSOLS code which emphasize the particle coagulation parameters: agglomerate shape factors and collision efficiency. Finally, the different computer models have been applied to the study of aerosol behavior during a 900 MWe PWR accident: loss-of-coolant-accident (small break with failure of all safety systems) [fr

A scheme of better utilization of PWR spent fuels

International Nuclear Information System (INIS)

Chung, Bum Jin; Kang, Chang Soon

1991-01-01

The recycle of PWR spent fuels in a CANDU reactor, so called the tandem fuel cycle is investigated in this study. This scheme of utilizing PWR spent fuels will ease the shortage of spent fuel storage capacity as well as will improve the use of uranium resources. The minimum modification the design of present CANDU reactor is seeked in the recycle. Nine different fuel types are considered in this work and are classified into two categories: refabrication and reconfiguration. For refabrication, PWR spent fuels are processed and refabricated into the present 37 rod lattice structure of fuel bundle, and for reconfiguration, meanwhile, spent fuels are simply disassembled and rods are cut to fit into the present grid configuration of fuel bundle without refabrication. For each fuel option, the neutronics calculation of lattice was conducted to evaluate the allowable burn up and distribution. The fuel cycle cost of each option was also computed to assess the economic justification. The results show that most tandem fuel cycle option considered in this study are technically feasible as well as economically viable. (Author)

Involvement of small and medium-sized enterprises (SMEs in elaborating and implementing public policies: Study case-Romanian small and medium-sized enterprises

Directory of Open Access Journals (Sweden)

Popescu Ruxandra

2017-07-01

Full Text Available Involvement and development of the private sector in boosting the economy nationwide is a main objective of the current program of the government, which means that a good cooperation between small and medium companies, private companies and multinationals and public environment including both public institutions and policies implemented and developed by them, becomes more than necessary. The paper summarizes the findings of a quantitative research based on a self-applied questionnaire which was aimed at Romanian small and medium-sized enterprises and also of a qualitative research that gives an overview of the process of elaborating and implementing a public policy. The involvement of small and medium-sized enterprises in the process of designing and implementing a public policy can become indispensable but it is well know that there is, in fact, a lack of initiative at this level, from both parties. One of the main research questions of this paper is to find out how much do representatives of small and medium-sized enterprises get involved in the process of elaborating a public policy and how much do these actions and measures impact the organizational policies of the companies themselves. A good cooperation between the business environment and the public institutions and a strong correlation of their joint efforts, should become a common practice between both parties, being crucial that this form of cooperation to be initiated from the very beginning. The contribution of this paper is a practical one, given the fact that the paper itself entails the direct responses of small and medium-sized enterprises on the current and future public policies that directly targets them, providing as well an analysis on the effects of public policies on small and medium-sized enterprises. Thus being said, the paper can also be a guide for both small and medium-sized enterprises in providing examples and measures of involvement and favorable public policies

International standard problem ISP36. Cora-W2 experiment on severe fuel damage for a Russian type PWR

Energy Technology Data Exchange (ETDEWEB)

NONE

1997-12-31

An OECD/NEA-CSNI International Standard Problem (ISP) has been performed on the experimental comparison basis of the severe fuel damage experiment CORA-W2. The out-of-pile experiment CORA-W2 was executed in February 1993 at he Forschungszentrum Karlsruhe. The objective of this experiment was the investigation of the behavior of a Russian type PWR fuel element (VVER-1000) during early core degradation. The main difference between a Western type and a Russian type PWR bundle is the B{sub 4}C absorber rod instead of AgInCd. Measured quantities ar boundary conditions, bundle temperature, hydrogen generation and the final bundle configurations after cooldown. The ISP was conducted as a blind exercise. Boundary conditions were estimated using ATHLET-CD. Six different severe accident codes were used. The comparisons between experimental and analytical results were grouped by codes and examined separately. The thermal behavior up to significant oxidation has been predicted quite well. Larger deviations have been observed for the oxidation-induced temperature escalation, both time of onset and maximum temperature as well. The bundle behavior is greatly influenced by chemical interactions involving B{sub 4}C absorber rod material, which failed relatively early at low temperature due to eutectic interaction between B{sub 4}C and SS cladding as well as the SS guide tube. Regarding the complex material interaction larger differences can be recognized between calculated and measured results because of inappropriate models for material relocation and solidification processes and the lack of models describing the interactions of absorber rod materials with the fuel rods. For the total amount of H{sub 2} generated, acceptable agreement could be achieved, if the total of oxidized zirconium was calculated correctly. The oxidation of stainless steel components and B{sub 4}C were not treated. In general the confidence in code predictions decreases with processing core damage. 36 refs.

International standard problem ISP36. Cora-W2 experiment on severe fuel damage for a Russian type PWR

International Nuclear Information System (INIS)

1996-01-01

An OECD/NEA-CSNI International Standard Problem (ISP) has been performed on the experimental comparison basis of the severe fuel damage experiment CORA-W2. The out-of-pile experiment CORA-W2 was executed in February 1993 at he Forschungszentrum Karlsruhe. The objective of this experiment was the investigation of the behavior of a Russian type PWR fuel element (VVER-1000) during early core degradation. The main difference between a Western type and a Russian type PWR bundle is the B 4 C absorber rod instead of AgInCd. Measured quantities ar boundary conditions, bundle temperature, hydrogen generation and the final bundle configurations after cooldown. The ISP was conducted as a blind exercise. Boundary conditions were estimated using ATHLET-CD. Six different severe accident codes were used. The comparisons between experimental and analytical results were grouped by codes and examined separately. The thermal behavior up to significant oxidation has been predicted quite well. Larger deviations have been observed for the oxidation-induced temperature escalation, both time of onset and maximum temperature as well. The bundle behavior is greatly influenced by chemical interactions involving B 4 C absorber rod material, which failed relatively early at low temperature due to eutectic interaction between B 4 C and SS cladding as well as the SS guide tube. Regarding the complex material interaction larger differences can be recognized between calculated and measured results because of inappropriate models for material relocation and solidification processes and the lack of models describing the interactions of absorber rod materials with the fuel rods. For the total amount of H 2 generated, acceptable agreement could be achieved, if the total of oxidized zirconium was calculated correctly. The oxidation of stainless steel components and B 4 C were not treated. In general the confidence in code predictions decreases with processing core damage. (N.T.)

Sensitivity Verification of PWR Monitoring System Using Neuro-Expert For LOCA Detection

International Nuclear Information System (INIS)

Muhammad Subekti

2009-01-01

Sensitivity Verification of PWR Monitoring System Using Neuro-Expert For LOCA Detection. The present research was done for verification of previous developed method on Loss of Coolant Accident (LOCA) detection and perform simulations for knowing the sensitivity of the PWR monitoring system that applied neuro-expert method. The previous research continuing on present research, has developed and has tested the neuro-expert method for several anomaly detections in Nuclear Power Plant (NPP) typed Pressurized Water Reactor (PWR). Neuro-expert can detect the LOCA anomaly with sensitivity of primary coolant leakage of 7 gallon/min and the conventional method could not detect the primary coolant leakage of 30 gallon/min. Neuro expert method detects significantly LOCA anomaly faster than conventional system in Surry-1 NPP as well so that the impact risk is reducible. (author)

A digital control and monitoring system for PWR waste-disposal systems

International Nuclear Information System (INIS)

Ueda, Toshiharu; Fuchigami, Kazuyuki; Shimozato, Masao; Takazawa, Kazuo

1982-01-01

Mitsubishi Electric has developed a digital control and monitoring system for PWR waste-disposal systems. This novel system has improved operability due to its automated operations and control, and integrated supervisory functions. The system includes other features to improve operability: sequence control by a control computer, direct-digital process control, integrated supervision of operation states by a supervisory computer and a high-speed dataway, and CRT interfacing between the computer and dataway. (author)

Conceptual design of small-sized HTGR system (3). Core thermal and hydraulic design

International Nuclear Information System (INIS)

Inaba, Yoshitomo; Sato, Hiroyuki; Goto, Minoru; Ohashi, Hirofumi; Tachibana, Yukio

2012-06-01

The Japan Atomic Energy Agency has started the conceptual designs of small-sized High Temperature Gas-cooled Reactor (HTGR) systems, aiming for the 2030s deployment into developing countries. The small-sized HTGR systems can provide power generation by steam turbine, high temperature steam for industry process and/or low temperature steam for district heating. As one of the conceptual designs in the first stage, the core thermal and hydraulic design of the power generation and steam supply small-sized HTGR system with a thermal power of 50 MW (HTR50S), which was a reference reactor system positioned as a first commercial or demonstration reactor system, was carried out. HTR50S in the first stage has the same coated particle fuel as HTTR. The purpose of the design is to make sure that the maximum fuel temperature in normal operation doesn't exceed the design target. Following the design, safety analysis assuming a depressurization accident was carried out. The fuel temperature in the normal operation and the fuel and reactor pressure vessel temperatures in the depressurization accident were evaluated. As a result, it was cleared that the thermal integrity of the fuel and the reactor coolant pressure boundary is not damaged. (author)

Design of the control room of the N4-type PWR: main features and feedback operating experience; La salle de commande du palier N4: principales caracteristiques et retour d'experience d'exploitation

Energy Technology Data Exchange (ETDEWEB)

Peyrouton, J.M.; Guillas, J.; Nougaret, Ch. [Electricite de France (EDF/DPN/CAPE), 93 - Saint-Denis (France)

2004-07-01

This article presents the design, specificities and innovating features of the control room of the N4-type PWR. A brief description of control rooms of previous 900 MW and 1300 MW -type PWR allows us to assess the change. The design of the first control room dates back to 1972, at that time 2 considerations were taken into account: first the design has to be similar to that of control rooms for thermal plants because plant operators were satisfied with it and secondly the normal operating situation has to be privileged to the prejudice of accidental situations just as it was in a thermal plant. The turning point was the TMI accident that showed the weight of human factor in accidental situations in terms of pilot team, training, procedures and the ergonomics of the work station. The impact of TMI can be seen in the design of 1300 MW-type PWR. In the beginning of the eighties EDF decided to launch a study for a complete overhaul of the control room concept, the aim was to continue reducing the human factor risk and to provide a better quality of piloting the plant in any situation. The result is the control room of the N4-type PWR. Today the cumulated feedback experience of N4 control rooms represents more than 20 years over a wide range of situations from normal to incidental, a survey shows that the N4 design has fulfilled its aims. (A.C.)

Evaluation of Approaches to Analyzing Continuous Correlated Eye Data When Sample Size Is Small.

Science.gov (United States)

Huang, Jing; Huang, Jiayan; Chen, Yong; Ying, Gui-Shuang

2018-02-01

To evaluate the performance of commonly used statistical methods for analyzing continuous correlated eye data when sample size is small. We simulated correlated continuous data from two designs: (1) two eyes of a subject in two comparison groups; (2) two eyes of a subject in the same comparison group, under various sample size (5-50), inter-eye correlation (0-0.75) and effect size (0-0.8). Simulated data were analyzed using paired t-test, two sample t-test, Wald test and score test using the generalized estimating equations (GEE) and F-test using linear mixed effects model (LMM). We compared type I error rates and statistical powers, and demonstrated analysis approaches through analyzing two real datasets. In design 1, paired t-test and LMM perform better than GEE, with nominal type 1 error rate and higher statistical power. In design 2, no test performs uniformly well: two sample t-test (average of two eyes or a random eye) achieves better control of type I error but yields lower statistical power. In both designs, the GEE Wald test inflates type I error rate and GEE score test has lower power. When sample size is small, some commonly used statistical methods do not perform well. Paired t-test and LMM perform best when two eyes of a subject are in two different comparison groups, and t-test using the average of two eyes performs best when the two eyes are in the same comparison group. When selecting the appropriate analysis approach the study design should be considered.

Preliminary LOCA analysis of the westinghouse small modular reactor using the WCOBRA/TRAC-TF2 thermal-hydraulics code

Energy Technology Data Exchange (ETDEWEB)

Liao, J.; Kucukboyaci, V. N.; Nguyen, L.; Frepoli, C. [Westinghouse Electric Company, 1000 Westinghouse Drive, Cranberry Township, PA 16066 (United States)

2012-07-01

The Westinghouse Small Modular Reactor (SMR) is an 800 MWt (> 225 MWe) integral pressurized water reactor (iPWR) with all primary components, including the steam generator and the pressurizer located inside the reactor vessel. The reactor core is based on a partial-height 17x17 fuel assembly design used in the AP1000{sup R} reactor core. The Westinghouse SMR utilizes passive safety systems and proven components from the AP1000 plant design with a compact containment that houses the integral reactor vessel and the passive safety systems. A preliminary loss of coolant accident (LOCA) analysis of the Westinghouse SMR has been performed using the WCOBRA/TRAC-TF2 code, simulating a transient caused by a double ended guillotine (DEG) break in the direct vessel injection (DVI) line. WCOBRA/TRAC-TF2 is a new generation Westinghouse LOCA thermal-hydraulics code evolving from the US NRC licensed WCOBRA/TRAC code. It is designed to simulate PWR LOCA events from the smallest break size to the largest break size (DEG cold leg). A significant number of fluid dynamics models and heat transfer models were developed or improved in WCOBRA/TRAC-TF2. A large number of separate effects and integral effects tests were performed for a rigorous code assessment and validation. WCOBRA/TRAC-TF2 was introduced into the Westinghouse SMR design phase to assist a quick and robust passive cooling system design and to identify thermal-hydraulic phenomena for the development of the SMR Phenomena Identification Ranking Table (PIRT). The LOCA analysis of the Westinghouse SMR demonstrates that the DEG DVI break LOCA is mitigated by the injection and venting from the Westinghouse SMR passive safety systems without core heat up, achieving long term core cooling. (authors)

SCC of Alloy 600 components in PWR primary loop

International Nuclear Information System (INIS)

Gomez-Briceno, Dolores; Lapena, Jesus; Castano, M. Luisa; Blazquez, Fernando

2002-01-01

Full text: Cracking due to PWSCC in PWR CRDM nozzles and other VHP nozzles fabricated from Alloy 600 is not a new issue. In 1991, a leak was discovered on one CRDM nozzle at Bugey 3 PWR plant in France. The cause of the cracking was identified as primary water stress corrosion cracking. From then, similar cracks have been found in other European and USA PWR plants. The cracks were predominantly axial in orientation and it was accepted that CRDM nozzles and weld cracking in PWR was not a immediate safety concern. However, this consideration has to be reassessed in light of the recent identification of circumferential cracking in CRDM nozzles at Oconee Nuclear Station Unit 2 and 3 along with axial cracking in the Alloy 182 J-groove welds at these two units and at Oconee Nuclear Station 1 and Arkansas Nuclear One Unit 1. Alloy 600 susceptibility in primary water has received an enormous research effort for many years since the Alloy 600 steam generators tube degradation started. A significant amount of information is available to characterise the susceptibility of Alloy 600. However, Alloy 600 susceptibility is strongly dependent on the heat thermomechanical history and both the crack initiation time and the crack growth rate data obtained from representative materials of the VHP nozzles seem to be necessary for the structural integrity assessment of cracking nozzles. An extensive experimental program has been performed at CIEMAT, to study the behaviour of Alloy 600 VHP nozzles in PWR primary conditions. Crack initiation and crack propagation tests have been performed using different types of products (forged bar, tube, plate and steam generator tubing). Long duration crack initiation tests have been carried out, at 330 deg. C and 360 deg. C in water and at 400 deg. C in steam, using ten Alloy 600 heats with yield strength ranging from 291 MPa to 489 MPa. The influence of several parameters (grain boundary carbide distribution, grain size and yield strength) on crack

Development of laser weld monitoring system for PWR space grid

International Nuclear Information System (INIS)

Chung, Chin Man; Kim, Cheol Jung; Kim, Min Suk

1998-06-01

The laser welding monitoring system was developed to inspect PWR space grid welding for KNFC. The demands for this optical monitoring system were applied to Q.C. and process control in space grid welding. The thermal radiation signal from weld pool can be get the variation of weld pool size. The weld pool size and depth are verified by analyzed wavelength signals from weld pool. Applied this monitoring system in space grid weld, improved the weld productivity. (author). 4 refs., 5 tabs., 31 figs

Analytical and sampling problems in primary coolant circuits of PWR-type reactors

International Nuclear Information System (INIS)

Illy, H.

1980-10-01

Details of recent analytical methods on the analysis and sampling of a PWR primary coolant are given in the order as follows: sampling and preparation; analysis of the gases dissolved in the water; monitoring of radiating substances; checking of boric acid concentration which controls the reactivity. The bibliography of this work and directions for its use are published in a separate report: KFKI-80-48 (1980). (author)

Aqueous Boric acid injection facility of PWR type reactor

International Nuclear Information System (INIS)

Matsuoka, Tsuyoshi; Iwami, Masao.

1996-01-01

If a rupture should be caused in a secondary system of a PWR type reactor, pressure of a primary coolant recycling system is lowered, and a back flow check valve is opened in response to the lowering of the pressure. Then, low temperature aqueous boric acid in the lower portion of a pressurized tank is flown into the primary coolant recycling system based on the pressure difference, and the aqueous boric acid reaches the reactor core together with coolants to suppress reactivity. If the injection is continued, high temperature aqueous boric acid in the upper portion boils under a reduced pressure, further urges the low temperature aqueous boric acid in the lower portion by the steam pressure and injects the same to the primary system. The aqueous boric acid stream from the pressurized tank flowing by self evaporation of the high temperature aqueous boric acid itself is rectified by a rectifying device to prevent occurrence of vortex flow, and the steam is injected in a state of uniform stream. When the pressure in the pressurized tank is lowered, a bypass valve is opened to introduce the high pressure fluid of primary system into the pressurized tank to keep the pressure to a predetermined value. When the pressure in the pressurized tank is elevated to higher than the pressure of the primary system, a back flow check valve is opened, and high pressure aqueous boric acid is flown out of the pressurized tank to keep the pressure to a predetermined value. (N.H.)

VALIDATION OF SIMBAT-PWR USING STANDARD CODE OF COBRA-EN ON REACTOR TRANSIENT CONDITION

Directory of Open Access Journals (Sweden)

Muhammad Darwis Isnaini

2016-03-01

Full Text Available The validation of Pressurized Water Reactor typed Nuclear Power Plant simulator developed by BATAN (SIMBAT-PWR using standard code of COBRA-EN on reactor transient condition has been done. The development of SIMBAT-PWR has accomplished several neutronics and thermal-hydraulic calculation modules. Therefore, the validation of the simulator is needed, especially in transient reactor operation condition. The research purpose is for characterizing the thermal-hydraulic parameters of PWR1000 core, which be able to be applied or as a comparison in developing the SIMBAT-PWR. The validation involves the calculation of the thermal-hydraulic parameters using COBRA-EN code. Furthermore, the calculation schemes are based on COBRA-EN with fixed material properties and dynamic properties that calculated by MATPRO subroutine (COBRA-EN+MATPRO for reactor condition of startup, power rise and power fluctuation from nominal to over power. The comparison of the temperature distribution at nominal 100% power shows that the fuel centerline temperature calculated by SIMBAT-PWR has 8.76% higher result than COBRA-EN result and 7.70% lower result than COBRA-EN+MATPRO. In general, SIMBAT-PWR calculation results on fuel temperature distribution are mostly between COBRA-EN and COBRA-EN+MATPRO results. The deviations of the fuel centerline, fuel surface, inner and outer cladding as well as coolant bulk temperature in the SIMBAT-PWR and the COBRA-EN calculation, are due to the value difference of the gap heat transfer coefficient and the cladding thermal conductivity.

76 FR 14323 - Small Business Size Standards: Professional, Scientific and Technical Services

Science.gov (United States)

2011-03-16

... government small business assistance programs, SBA establishes small business size definitions (referred to... its field of operation and (3) within a specific small business definition or size standard... SMALL BUSINESS ADMINISTRATION 13 CFR Part 121 RIN 3245-AG07 Small Business Size Standards...

The development of flow test technology for PWR fuel assembly

International Nuclear Information System (INIS)

Chung, Moon Ki; Cha, Chong Hee; Chung, Chang Hwan; Chun, Se Young; Song, Chul Hwa; Chung, Heung Joon; Won, Soon Yeun; Cho, Yeong Rho; Kim, Bok Deuk

1988-05-01

KAERI has an extensive program to develope PWR fuel assembly. In relation to the program, development of flow test technology is needed to evaluate the thermal hydraulic compactibility and mechanical integrity of domestically fabricated nuclear fuels. A high-pressure and high-temperature flow test facility was designed to test domestically fabricated fuel assembly. The test section of the facility has capacity of a 6x6 full length PWR fuel assembly. A flow test rig was designed and installed at Cold Test Loop to carry out model experiments with 5x5 rod assembly under atmosphere pressure to get information about the characteristics of pressure loss of spacer grids and velocity distribution in the subchannels. LDV measuring technology was established using TSI's Laser Dopper Velocimeter 9100-3 System

Review on research of small break loss of coolant accident

International Nuclear Information System (INIS)

Bo Jinhai; Wang Fei

1998-01-01

The Small Break Loss of Coolant Accident (SBLOCA) and its research art-of -work are reviewed. A typical SBLOCA process in Pressurized Water Reactor (PWR) and Boiling Water Reactor (BWR) and the influence of break size, break location and reactor coolant pump on the process are described. The existing papers are classified in two categories: experimental and numerical modeling, with the primary experimental apparatuses in the world listed and the research works on SBLOCA summarized

PWR secondary water chemistry guidelines: Revision 3

International Nuclear Information System (INIS)

Lurie, S.; Bucci, G.; Johnson, L.; King, M.; Lamanna, L.; Morgan, E.; Bates, J.; Burns, R.; Eaker, R.; Ward, G.; Linnenbom, V.; Millet, P.; Paine, J.P.; Wood, C.J.; Gatten, T.; Meatheany, D.; Seager, J.; Thompson, R.; Brobst, G.; Connor, W.; Lewis, G.; Shirmer, R.; Gillen, J.; Kerns, M.; Jones, V.; Lappegaard, S.; Sawochka, S.; Smith, F.; Spires, D.; Pagan, S.; Gardner, J.; Polidoroff, T.; Lambert, S.; Dahl, B.; Hundley, F.; Miller, B.; Andersson, P.; Briden, D.; Fellers, B.; Harvey, S.; Polchow, J.; Rootham, M.; Fredrichs, T.; Flint, W.

1993-05-01

An effective, state-of-the art secondary water chemistry control program is essential to maximize the availability and operating life of major PWR components. Furthermore, the costs related to maintaining secondary water chemistry will likely be less than the repair or replacement of steam generators or large turbine rotors, with resulting outages taken into account. The revised PWR secondary water chemistry guidelines in this report represent the latest field and laboratory data on steam generator corrosion phenomena. This document supersedes Interim PWR Secondary Water Chemistry Recommendations for IGA/SCC Control (EPRI report TR-101230) as well as PWR Secondary Water Chemistry Guidelines--Revision 2 (NP-6239)

Application of concrete filled steel bearing wall to inner concrete structure fro PWR nuclear power plant

International Nuclear Information System (INIS)

Sekimoto, Hisashi; Tanaka, Mamoru; Inoue, Kunio; Fukihara, Masaaki; Akiyama, Hiroshi.

1992-01-01

'Concrete filled steel bearing wall', applied to the inner concrete structure for PWR nuclear power plant, was developed for rationalization of construction procedure at site. It was concluded through preliminary studies that this new type of wall, where concrete is placed between steel plates, is best suited for the strength members of the above structure, due to the high strength and ductility of surface steel plates and the confinement effect of filled concrete. To verify the behavior from the elastic range to the inelastic range, the ultimate strength and the failure mechanism, and to clarify experimentally the structural integrity of the inner concrete structure, which was composed of a concrete filled steel bearing wall, against seismic lateral loads, horizontal loading tests using a 1/10th scale model of the inner concrete structure for PWR nuclear power plant were conducted. As a result of the tests, the inner concrete structure composed of a concrete filled steel bearing wall appeared to have a larger load carrying capacity and a higher ductility as compared with that composed of a reinforced concrete wall. (author)

Preliminary Analysis of Severe Accident Progression Initiated from Small Break LOCA of a SMART Reactor

International Nuclear Information System (INIS)

Jin, Young Ho; Park, Jong Hwa; Kim, Dong Ha; Cho, Seong Won

2010-01-01

SMART (System integrated Modular Advanced ReacTor), is under the development at Korea Atomic Energy Research Institute (KAERI). SMART is an integral type pressurized water reactor which contains a pressurizer, 4 reactor coolant pumps (RCPs), and 8 steam generator cassettes(S/Gs) in a single reactor vessel. This reactor has substantially enhanced its safety with an integral layout of its major components, 4 trains of safety injection systems (SISs), and an adoption of 4 trains of passive residual heat removal systems (PRHRS) instead of an active auxiliary feedwater system . The thermal power is 330 MWth. During the conceptual design stage, a preliminary PSA was performed. PSA results identified that a small break loss of coolant accident (SLOCA) with all safety injections unavailable is one of important severe core damage sequences. Clear understanding of this sequence helps in the developing accident mitigation strategies. MIDAS/SMR computer code is used to simulate the severe accident progression initiated from a small break LOCA in SMART reactor. This code has capability to model a helical steam generator which is adopted in SMART reactor. The important accident progression results for SMART reactor are then compared with the typical pressurized water reactor (PWR) result

The value chain of small-sized energy wood

Energy Technology Data Exchange (ETDEWEB)

Karttunen, K.; Foehr, J.; Ranta, T. (Lappeenranta Univ. of Technology, Mikkeli (Finland), LUT Energy), Email: kalle.karttunen@lut.fi, Email: jarno.fohr@lut.fi, Email: tapio.ranta@lut.fi; Ahtikoski, A. (The Finnish Forest Research Institute, Rovaniemi (Finland)), Email: anssi.ahtikoski@metla.fi; Valsta, L. (Helsinki Univ. (Finland), Dept. of Forest Economics), Email: lauri.valsta@helsinki.fi

2009-07-01

Finland has agreed to increase the share of renewable energy to the level of 38% by the end of 2020. Most of the increase is to be based on bioenergy. According to the National Climate and Energy Strategy, the need for forest biomass will come to more than 20 TWh, or some 10 million cubic meters per year. Energy wood from young stand thinnings are the biomass resource with the most potential at the moment. The purpose of this study was to compare cost differences between forest management incorporating energy wood thinning and forest management based on traditional roundwood thinning. In addition, alternative supply chain costs for small-sized wood were studied. The results of the study show that it is worth considering the following points if the demand and average price for forest chips remain high. 1. Forest-owners: Forest management including energy wood thinning is financially feasible. 2. Supply chain: A terminal chipping chain enables large-scale procurement of small-sized energy wood. 3. Power plants: Currently, subsidies, emission trading, and decreasing pulpwood prices together enable large-scale use of small-sized wood for energy purposes. The value chain of small-sized energy wood in large-scale power plants could be mobilised. (orig.)

Reliability analysis of 2 types of auxiliary feedwater system for PWR

International Nuclear Information System (INIS)

Ekariansyah, Andi Sofrany

2002-01-01

This paper will explain the application of Fault Three Method for analyzing the system reliability of Auxiliary Feedwater System with 2 different configurations taken from PWR type nuclear power plant (NPP) in the USA. The first configuration of Braidwood NPP (design A) basically consists of 1 motor driven pump and 1 diesel driven pump. The second configuration of Haddam Neck NPP (Design B) consists of 2 turbine driven pumps. Based on the P and ID and success criteria the fault trees are constructed to estimate the system failure probabilities quantified from software code PIRAS 1.0. The result shows the second configuration (Design B) with 2 turbine driven pumps have the higher failure probability of 1,06 x 10 - 2 compared with design A of 1,09 x 10 - 3 . The modification of both systems are also tried to analyze its effect to the end result. Qualitatively, the common cause failures of 2 turbine driven pumps contribute to the highest risk of system failure probability. Combination with 1 turbine driven pump and 1 motor driven pump or 1 diesel driven pump will increase the system reliability about 80% and 50% without considering if this configuration is possible to realize in a real plant

Maintenance service for major component of PWR plant. Replacement of pressurizer safe end weld

International Nuclear Information System (INIS)

Miyoshi, Yoshiyuki; Kobayashi, Yuki; Yamamoto, Kazuhide; Ueda, Takeshi; Suda, Naoki; Shintani, Takashi

2017-01-01

In October 2016, MHI completed the replacement of safe end weld of pressurizer (Pz) of Ringhals unit 3, which was the first maintenance work for main component of pressurized water reactor (PWR) plant in Europe. For higher reliability and longer lifetime of PWR plant, MHI has conducted many kinds of maintenance works of main components of PWR plants in Japan against stress corrosion cracking due to aging degradation. Technical process for replacement of Pz safe end weld were established by MHI. MHI has experienced the work for 21 PWR units in Japan. That of Ringhals unit 3 was planned and conducted based on the experiences. In this work, Alloy 600 used for welds of nozzles of Pz was replaced with Alloy 690. Alloy 690 is more corrosive-resistant than Alloy 600. Specially designed equipment and technical process were developed and established by MHI to replace safe end weld of Pz and applied for the Ringhals unit 3 as a first application in Europe. The application had been performed in success and achieved the planned replacement work duration and total radiation dose by using sophisticated machining and welding equipment designed to meet the requirements to be small, lightweight and remote-controlled and operating by well skilled MHI personnel experienced in maintenance activities for major components of PWR plant in Japan. The success shows that the experience, activities and technology developed in Japan for main components of PWR plant shall be applicable to contribute reliable operations of nuclear power plants in Europe and other countries. (author)

Information Security in Small and Medium-Sized Companies

OpenAIRE

David Kral

2011-01-01

Information security doesn’t involve only large organizations. Small and medium-sized companies must closely examine this issue too, because they are increasingly threatened by cyber attacks. Many of them mistakenly believe, that security of their valuable data is sufficient, or that the attackers are not interested in them. Existing standards and methodologies for implementation and management of information security are often hard to transfer to the environment of small and medium-sized bus...

78 FR 37404 - Small Business Size Standards: Support Activities for Mining

Science.gov (United States)

2013-06-20

... assistance programs, SBA establishes small business size definitions (referred to as size standards) for... million should be the limit of a small business definition and anything larger than that, such as that SBA... Business Act (15 U.S.C. 632(a)) (Act) requires that small business size definitions vary to reflect...

THE SOCIAL MEDIA IMPACT ON SMALL AND MEDIUM SIZED BUSINESSES

OpenAIRE

Mihai Alexandru Constantin Logofatu

2012-01-01

This paper aims to be a short introduction to social media and discusses on few ways in which small and medium sized businesses in Romania can take advantage of this hot topic. Through the use of social media every company can reach a global audience with less effort, time and money. In a world shaped more and more around social platforms the customer behaviour has completely and forever changed and those leaders and organizations that understand and embrace this new type of communication, co...

77 FR 72766 - Small Business Size Standards: Support Activities for Mining

Science.gov (United States)

2012-12-06

... eligibility for Federal small business assistance, SBA establishes small business size definitions (referred... operation; and (3) within a specific small business definition or size standard established by SBA... to SBA's Administrator the responsibility for establishing small business definitions. The Act also...

Re-irradiation and limit testing of the fuels PWR type reactors

International Nuclear Information System (INIS)

Roche, M.; Molvault, M.

1978-01-01

In view of investigating the neutron radiation behavior of PWR fuel pins, the S.P.S. (Services des Piles de Saclay) developed a set of experimental means used at OSIRIS in Saclay Nuclear Research Center. Said devices are shown to be able to meet present problems concerning can failures, power and temperature cycling, remote-control studies. These means can also be used either for statistical studies, they can then receive several samples, or for analytical studies in instrumented devices of large capacity and accelerated irradiation rate [fr

Comparative economic analysis of the Integral Molten Salt Reactor and an advanced PWR using the G4-ECONS methodology

International Nuclear Information System (INIS)

Samalova, Ludmila; Chvala, Ondrej; Maldonado, G. Ivan

2017-01-01

The assessment of economic viability of a new reactor concept is crucial particularly during the early stages of its concept development. The G4-ECONS methodology provides a standardized top-down estimate of electricity cost and parametric sensitivities, not specifically targeted toward an accurate prediction of the final cost when deployed, but rather seeking an approximation of cost variations relative to other systems. This study presents an analysis of the Integral Molten Salt Reactor (IMSR) concept in comparison with a consistent analysis of an advanced PWR reactor (represented by AP1000). Estimation of levelized unit electricity costs, as well as sensitivity analyses to the discount rate and uranium or SWU prices, are presented using this methodology.

Prognosis method to predict small-sized breast cancer affected by fibrocystic disease

Directory of Open Access Journals (Sweden)

S. A. Velichko

2017-01-01

Full Text Available The purpose of the study is to develop an effective radiological symptom-complex of small-sized breast cancer affected by fibrocystic breast disease by using multivariate statistical methods.Materials and methods. Radiological findings of small-sized breast cancer affected by fibrocystic mastopathy were analyzed in 100 patients with histologically verified diagnosis.Results. It was revealed that the conventional approach to the analysis of mammograms based on the detection of the primary, secondary and indirect mammographic signs of small-sized breast cancer is not effective enough - the sensitivity of mammography is only 62%. Fibrocystic disease and moderate-to-severe sclerosing adenosis make small-sized breast cancer hard to visualize by mammography. The detailed analysis of mammograms allowed us to identify the additional manifestations of small-sized breast cancer affected by mastopathy. The computer program allowing us to evaluate the risk of small-size breast cancer and the diagnostic algorithm for detecting small size breast cancer with sensitivity of 92% were developed. 

Small-size biofuel cell on paper.

Science.gov (United States)

Zhang, Lingling; Zhou, Ming; Wen, Dan; Bai, Lu; Lou, Baohua; Dong, Shaojun

2012-05-15

In this work, we demonstrated a novel paper-based mediator-less and compartment-less biofuel cell (BFC) with small size (1.5 cm × 1.5 cm). Ionic liquid functionalized carbon nanotubes (CNTs-IL) nanocomposite was used as support for both stably confining the anodic biocatalyst (i.e., NAD(+)-dependent glucose dehydrogenase, GDH) for glucose electrooxidation and for facilitating direct electrochemistry of the cathodic biocatalyst (i.e., bilirubin oxidase, BOD) for O(2) electroreduction. Such BFC provided a simple approach to fabricate low-cost and portable power devices on small-size paper, which can harvest energy from a wide range of commercial beverages containing glucose (e.g., Nescafe instant coffee, Maidong vitamin water, Watermelon fresh juice, and Minute Maid grape juice). These made the low-cost paper-based biodevice potential for broad energy applications. Copyright © 2012 Elsevier B.V. All rights reserved.

The Effect of Material Homogenization in Calculating the Gamma-Ray dose from Spent PWR Fuel Pins in an Air Medium

International Nuclear Information System (INIS)

TH Trumbull

2005-01-01

The effect of material homogenization on the calculated dose rate was studied for several arrangements of typical PWR spent fuel pins in an air medium using the Monte Carlo code, MCNP. The models analyzed increased in geometric complexity, beginning with a single fuel pin, progressing to ''small'' lattices, i.e., 3x3, 5x5, 7x7 fuel pins, and culminating with a full 17x17 pin PWR bundle analysis. The fuel pin dimensions and compositions were taken directly from a previous study and efforts were made to parallel this study by specifying identical flux-to-dose functions and gamma-ray source spectra. The analysis shows two competing components to the overall effect of material homogenization on calculated dose rate. Homogenization of pin lattices tends to lower the effect of radiation ''channeling'' but increase the effect of ''source redistribution.'' Depending on the size of the lattice and location of the detectors, the net effect of material homogenization on dose rate can be insignificant or range from a 6% decrease to a 35% increase relative to the detailed geometry model

Non-farm businesses local economic integration level: the case of six Portuguese small and medium-sized Markettowns - a sector approach

OpenAIRE

Diniz, Francisco

2004-01-01

Small and medium-sized towns in rural areas are of particular interest at this stage in the evolution of the European CAP. Serious consideration is being given to mechanisms that could transfer resources from the agricultural sector into a more diversified rural economy in order to safeguard the well-being of both the farming community and the wider rural population while still preserving the environmental assets which are such a valued feature of Europe's rural areas. Small and medium-sized ...

Techniques for Primary-to-Secondary Leak Monitoring in PWR Plants

International Nuclear Information System (INIS)

Sohn, Wook; Chi, Jun Hwa; Kang, Duck Won; Tae, Jeong Woo

2006-01-01

Historically, corrosion and mechanical damage have made steam generator tubes in PWR plants see various types of degradation from both the primary and secondary sides of the tubes. Since the tube degradation can lead to through-wall failure, the plant personnel should make efforts to prevent the failure. One of such preventive efforts is to monitor primary-to-secondary leakage (PSL) that usually precedes the tube rupture. Thus the objective of PSL monitoring is to make operators to determine when to shutdown the plant in order to minimize the likelihood of propagation of leaks to tube rupture under normal and faulted conditions This paper addresses briefly the status of techniques for PSL monitoring used in PWR plants

Examination on small-sized cogeneration HTGR for developing countries

International Nuclear Information System (INIS)

Sakaba, Nariaki; Tachibana, Yukio; Shimakawa, Satoshi; Ohashi, Hirofumi; Sato, Hiroyuki; Yan, Xing; Murakami, Tomoyuki; Ohashi, Kazutaka; Nakagawa, Shigeaki; Goto, Minoru; Ueta, Shohei; Mozumi, Yasuhiro; Imai, Yoshiyuki; Tanaka, Nobuyuki; Okuda, Hiroyuki; Iwatsuki, Jin; Kubo, Shinji; Takada, Shoji; Nishihara, Tetsuo; Kunitomi, Kazuhiko

2008-03-01

The small-sized and safe cogeneration High Temperature Gas-cooled Reactor (HTGR) that can be used not only for electric power generation but also for hydrogen production and district heating is considered one of the most promising nuclear reactors for developing countries where sufficient infrastructure such as power grids is not provided. Thus, the small-sized cogeneration HTGR, named High Temperature Reactor 50-Cogeneration (HTR50C), was studied assuming that it should be constructed in developing countries. Specification, equipment configuration, etc. of the HTR50C were determined, and economical evaluation was made. As a result, it was shown that the HTR50C is economically competitive with small-sized light water reactors. (author)

75 FR 61597 - Small Business Size Standards: Retail Trade

Science.gov (United States)

2010-10-06

... eligibility for Federal small business assistance programs, SBA establishes small business size definitions... authorizes the SBA Administrator to establish only one definition of small business for an industry. [[Page... SBA's Administrator the responsibility for establishing small business definitions. The Act also...

A nodal model for the simulation of a PWR core

International Nuclear Information System (INIS)

Souza Pinto, R. de.

1981-06-01

A computer program FORTRAN language was developed to simulate the neutronic and thermal-hydraulic transient behaviour of a PWR reactor core. The reator power is calculated using a point kinectics model with six groups of delayed neutron precursors. The fission product decay heat was considered assuming three effective decay heat groups. A nodal model was employed for the treatment of heat transfer in the fuel rod, with integration of the heat equation by the lumped parameter technique. Axial conduction was neglected. A single-channel nodal model was developed for the thermo-hydrodynamic simulation using mass and energy conservation equations for the control volumes. The effect of the axial pressure variation was neglected. The computer program was tested, with good results, through the simulation of the transient behaviour of postulated accidents in a typical PWR. (Author) [pt

Simulation of fission products behavior in severe accidents for advanced passive PWR

International Nuclear Information System (INIS)

Tong, L.L.; Huang, G.F.; Cao, X.W.

2015-01-01

Highlights: • A fission product analysis model based on thermal hydraulic module is developed. • An assessment method for fission product release and transport is constructed. • Fission products behavior during three modes of containment response is investigated. • Source term results for the three modes of containment response are obtained. - Abstract: Fission product behavior for common Pressurized Water Reactor (PWR) has been studied for many years, and some analytical tools have developed. However, studies specifically on the behavior of fission products related to advanced passive PWR is scarce. In the current study, design characteristics of advanced passive PWR influencing fission product behavior are investigated. An integrated fission products analysis model based on a thermal hydraulic module is developed, and the assessment method for fission products release and transport for advanced passive PWR is constructed. Three modes of containment response are simulated, including intact containment, containment bypass and containment overpressure failure. Fission products release from the core and corium, fission products transport and deposition in the Reactor Coolant System (RCS), fission products transport and deposition in the containment considering fission products retention in the in-containment refueling water storage tank (IRWST) and in the secondary side of steam generators (SGs) are simulated. Source term results of intact containment, containment bypass and containment overpressure failure are obtained, which can be utilized to evaluate the radiological consequences

77 FR 42211 - Small Business Size Standards: Arts, Entertainment, and Recreation

Science.gov (United States)

2012-07-18

..., SBA establishes small business definitions (referred to as size standards) for private sector... operated; (2) not dominant in its field of operation; and (3) within a specific small business definition... SMALL BUSINESS ADMINISTRATION 13 CFR Part 121 RIN 3245-AG36 Small Business Size Standards: Arts...

Burst protected nuclear reactor plant with PWR

International Nuclear Information System (INIS)

Harand, E.; Michel, E.

1978-01-01

In the PWR, several integrated components from the steam raising unit and the main coolant pump are grouped around the reactor pressure vessel in a multiloop circuit and in a vertical arrangement. For safety reasons all primary circuit components and pipelines are situated in burst protection covers. To reduce the area of the plant straight tube steam raising units with forced circulation are used as steam raising units. The boiler pumps are connected to the vertical tubes and to the pressure vessel via double pipelines made as twin chamber pipes. (DG) [de

Hygrometric measurement for on-line monitoring of PWR vessel head penetrations

International Nuclear Information System (INIS)

Germain, J.L.; Loisy, F.; Apolzan, S.

1994-06-01

In September 1991, a small leak was found on one of the reactor's upper vessel head penetrations. After inspection, other non-throughwall cracks were localized in the lower part of the vessel head adapter in questions. The same type of crack was later found inside some adapters on other French PWR units. After repairs, the safety authorities granted approval to continue unit operation, with the specific provision that a system for ongoing monitoring of the penetrations be set up. Two types of system were selected to detect leaks through any potential cracks: the first is based on nitrogen-13 detection and the second on steam detection. Both systems call for sampling the air in a confined space above the vessel head. The number and distribution of sampling taps in the circuit, and the balancing of their respective flow rates, are factors in proper monitoring of all vessel head penetrations. Gas-injection holes are also installed in the confined space. These holes are used during the sampling system qualification tests to simulate leaks in various positions and calculate the effective performance of the sampling system. Leaks are simulated using a helium-base gas tracer and measuring tracer concentrations in the sampling system. The system for measuring steam levels in air samples uses chilled-mirror hygrometers. A microcomputer takes regular readings, drives the various automatic functions of the measurement system and automatically analyses the readings so as to monitor operations and trigger an alarm at the first sign of a leak. This system has now been installed for a year and a half on three French PWR units and is functioning satisfactorily. (authors). 5 figs

Status of integration of small computers into NDE systems

International Nuclear Information System (INIS)

Dau, G.J.; Behravesh, M.M.

1988-01-01

Introduction of computers in nondestructive evaluations (NDE) has enabled data acquisition devices to provide a more thorough and complete coverage in the scanning process, and has aided human inspectors in their data analysis and decision making efforts. The price and size/weight of small computers, coupled with recent increases in processing and storage capacity, have made small personal computers (PC's) the most viable platform for NDE equipment. Several NDE systems using minicomputers and newer PC-based systems, capable of automatic data acquisition, and knowledge-based analysis of the test data, have been field tested in the nuclear power plant environment and are currently available through commercial sources. While computers have been in common use for several NDE methods during the last few years, their greatest impact, however, has been on ultrasonic testing. This paper discusses the evolution of small computers and their integration into the ultrasonic testing process

Mathematical modelling of plant transients in the PWR for simulator purposes

International Nuclear Information System (INIS)

Hartel, K.

1984-01-01

This chapter presents the results of the testing of anticipated and abnormal plant transients in pressurized water reactors (PWRs) of the type WWER 440 by means of the numerical simulation of 32 different, stationary and nonstationary, operational regimes. Topics considered include the formation of the PWR mathematical model, the physical approximation of the reactor core, the structure of the reactor core model, a mathematical approximation of the reactor model, the selection of numerical methods, and a computerized simulation system. The necessity of a PWR simulator in Czechoslovakia is justified by the present status and the outlook for the further development of the Czechoslovak nuclear power complex

Nonlinear Fuzzy Model Predictive Control for a PWR Nuclear Power Plant

Directory of Open Access Journals (Sweden)

Xiangjie Liu

2014-01-01

Full Text Available Reliable power and temperature control in pressurized water reactor (PWR nuclear power plant is necessary to guarantee high efficiency and plant safety. Since the nuclear plants are quite nonlinear, the paper presents nonlinear fuzzy model predictive control (MPC, by incorporating the realistic constraints, to realize the plant optimization. T-S fuzzy modeling on nuclear power plant is utilized to approximate the nonlinear plant, based on which the nonlinear MPC controller is devised via parallel distributed compensation (PDC scheme in order to solve the nonlinear constraint optimization problem. Improved performance compared to the traditional PID controller for a TMI-type PWR is obtained in the simulation.

Modular simulation of the dynamics of a 925 MWe PWR electronuclear type reactor and design of a multivariable control algorithm

International Nuclear Information System (INIS)

Mansouri, S.

1985-06-01

This work has been consecrated to the modular simulation of a PWR 925 MWe power plant's dynamic and to the design of a multivariable algorithm control: a mathematical model of a plant type was developed. The programs were written on a structured manner in order to maximize flexibility. A multivariable control algorithm based on pole placement with output feedback was elaborated together with its correspondent program. The simulation results for different normal transients were shown and the capabilities of the new method of multivariable control are illustrated through many examples

Results of safety analysis on PWR type nuclear power plants with two and three loops

International Nuclear Information System (INIS)

1979-01-01

The results of safety analysis on PWR type nuclear power plants with two and three loops are presented, which was conducted by the Resource and Energy Agency, in June, 1979. This analysis was made simulating the phenomenon relating to the pressurizer level gauge at the time of the TMI accident. The model plants were the Ikata nuclear power plant with two loops and the Takahama No. 1 nuclear power plant with three loops. The premise conditions for this safety analysis were as follows: 1) the main feed water flow is totally lost suddenly at the full power operation of the plants, and the feed water pump is started manually 15 minutes after the accident initiation, 2) the relief valve on the pressurizer is kept open even after the pressure drop in the primary cooling system, and the primary cooling water flows out into the containment vessel through the rupture disc of the pressurizer relief tank, and 3) the electric circuit, which sends out the signal of safety injection at the abnormal low pressure in the reactor vessel, is added from the view-point of starting the operation of the emergency core cooling system as early as possible. Relating to the analytical results, the pressure in the reactor vessels changes less, the water level in the pressurizers can be regulated, and the water level in the steam generators is recovered safely in both two and three-loop plants. It is recognized that the plants with both two- and three loops show the safe transient phenomena, and the integrity of the cores is kept under the premise conditions. The evaluation for each analyzed result was conducted in detail. (Nakai, Y.)

75 FR 1296 - Small Business Size Regulations; 8(a) Business Development/Small Disadvantaged Business Status...

Science.gov (United States)

2010-01-11

... SMALL BUSINESS ADMINISTRATION 13 CFR Parts 121 and 124 Small Business Size Regulations; 8(a) Business Development/Small Disadvantaged Business Status Determinations AGENCY: U.S. Small Business Administration. ACTION: Notice of public meetings; request for comments. SUMMARY: The U.S. Small Business...

Optimization of airfoil-type PCHE for the recuperator of small scale brayton cycle by cost-based objective function

International Nuclear Information System (INIS)

Kwon, Jin Gyu; Kim, Tae Ho; Park, Hyun Sun; Cha, Jae Eun; Kim, Moo Hwan

2016-01-01

Highlights: • Suggest the Nusselt number and Fanning friction factor correlation for airfoil-type PCHE. • Show that cost-based optimization is available to airfoil-type PCHE. • Suggest the recuperator design for SCIEL test loop at KAERI by cost-based objective function with correlations from numerical analysis. - Abstract: Supercritical carbon dioxide (SCO_2) Brayton cycle gives high efficiency of power cycle with small size. Printed circuit heat exchangers (PCHE) are proper selection for the Brayton cycle because their operability at high temperature and high pressure with small size. Airfoil fin PCHE was suggested by Kim et al. (2008b), it can provide high heat transfer-like zigzag channel PCHE with low pressure drop-like straight channel PCHE. Optimization of the airfoil fin PCHE was not performed like the zigzag channel PCHE. For optimization of the airfoil fin PCHE, the operating condition of the recuperator of SCO_2 Integral Experiment Loop (SCIEL) Brayton cycle test loop at Korea Atomic Energy Research Institute (KAERI) was used. We performed CFD analysis for various airfoil fin configurations using ANSYS CFX 15.0, and made correlations for predicting the Nusselt number and the Fanning friction factor. The recuperator was designed by the simple energy balance code with our correlations. Using the cost-based objective function with production cost and operation cost from size and pressure drop of the recuperator, we evaluated airfoil fin configuration by using total cost and suggested the optimization configuration of the airfoil fin PCHE.

Optimization of airfoil-type PCHE for the recuperator of small scale brayton cycle by cost-based objective function

Energy Technology Data Exchange (ETDEWEB)

Kwon, Jin Gyu [Division of Advanced Nuclear Engineering, POSTECH, Pohang 790-784 (Korea, Republic of); Kim, Tae Ho [Department of Mechanical Engineering, POSTECH, Pohang 790-784 (Korea, Republic of); Park, Hyun Sun, E-mail: hejsunny@postech.ac.kr [Division of Advanced Nuclear Engineering, POSTECH, Pohang 790-784 (Korea, Republic of); Cha, Jae Eun [Korea Atomic Energy Research Institute, Daejeon 305-353 (Korea, Republic of); Kim, Moo Hwan [Division of Advanced Nuclear Engineering, POSTECH, Pohang 790-784 (Korea, Republic of); Korea Institute of Nuclear Safety, Daejeon 305-338 (Korea, Republic of)

2016-03-15

Highlights: • Suggest the Nusselt number and Fanning friction factor correlation for airfoil-type PCHE. • Show that cost-based optimization is available to airfoil-type PCHE. • Suggest the recuperator design for SCIEL test loop at KAERI by cost-based objective function with correlations from numerical analysis. - Abstract: Supercritical carbon dioxide (SCO{sub 2}) Brayton cycle gives high efficiency of power cycle with small size. Printed circuit heat exchangers (PCHE) are proper selection for the Brayton cycle because their operability at high temperature and high pressure with small size. Airfoil fin PCHE was suggested by Kim et al. (2008b), it can provide high heat transfer-like zigzag channel PCHE with low pressure drop-like straight channel PCHE. Optimization of the airfoil fin PCHE was not performed like the zigzag channel PCHE. For optimization of the airfoil fin PCHE, the operating condition of the recuperator of SCO{sub 2} Integral Experiment Loop (SCIEL) Brayton cycle test loop at Korea Atomic Energy Research Institute (KAERI) was used. We performed CFD analysis for various airfoil fin configurations using ANSYS CFX 15.0, and made correlations for predicting the Nusselt number and the Fanning friction factor. The recuperator was designed by the simple energy balance code with our correlations. Using the cost-based objective function with production cost and operation cost from size and pressure drop of the recuperator, we evaluated airfoil fin configuration by using total cost and suggested the optimization configuration of the airfoil fin PCHE.

Small and medium-sized nuclear power plants

International Nuclear Information System (INIS)

Schmidt, R.

1986-01-01

Small and medium-sized nuclear power plants have long been under discussion as possible applications of nuclear power in countries with small transmission grid systems, in threshold countries and developing countries, and under special local supply conditions. IAEA has condensed and promoted this interest and tried to establish the demand, and possibilities of meeting it, in special events and campaigns. In recent years, considerable interest was registered even in industrialized countries, but here specially for heating and process heat generation applications and for special purposes and, in medium-sized units, also for combined supplies of electricity and heat. This corresponds to special reactor and plant concepts, some of which have already been developed to a stage at which construction work could begin. The analysis presented deals with necessary preconditions on the sides of the users and the vendors, with problems of economy, infrastructure and financing and with the market prospects of small nuclear power plants. (orig./HP) [de

Structure of Small and Medium-Sized Business: Results of Total Statistic Observations in Russia

Directory of Open Access Journals (Sweden)

Iuliia S. Pinkovetskaia

2018-03-01

Full Text Available The aim of the research is estimation of regularities and tendencies, characteristic for modern sectoral structure of small and mediumsized business in Russia. The subject of the research is a set of processes of structural changes on the types of economic activities of such enterprises, as well as the differentiation of the number of employees in enterprises. The research methodology included consideration of aggregates of subjects of small and medium-sized business, formed according to sectoral and territorial features. As the initial data used the official statistical information, which was obtain in the course of total observation of the activities of small and medium-sized businesses in 2010 and 2015. The study was conducted on indicators characterizing the full range of legal entities and individual entrepreneurs in the country. The materiality of structural changes was carried out on the basis of the Ryabtsev index. Modeling the differentiation of the values of the number of employees per enterprise was based on the development of density normal distribution functions. According to the hypothesis it is assumed that the differentiation of the number of employees working in enterprises depend on six main types of economic activity and on the subjects of Russia. Based on the results of the study was proved that there are no significant structural changes for the period from 2010 to 2015, both in terms of the number of enterprises and the number of their employees. Based on the results of the simulation, the average values of the number of employees for the six main types of activity were established, as well as the intervals for changing these indicators for the aggregates of small and medium-sized enterprises located in the majority of the country's subjects. The results of research can be used in the performance of scientific works related to the justification of the expected number and number of employees of enterprises, the formation of

THE POSSIBILITY OF USING INTERNATIONAL EXPERIENCE IN MICRO-CREDIT FOR SMALL AND MEDIUM-SIZED INDUSTRIAL ENTERPRISES

Directory of Open Access Journals (Sweden)

I. N. Klyukin

2016-01-01

Full Text Available Purpose of the study. Increasing the availability of funding for small and medium-sized enterprises of the industrial sector put among the most important tasks of economic development as the leading developed countries, and developing countries. In connection with the above, the purpose of this article is to study the micro-credit as an efficient mechanism to stimulate the development of small and medium-sized industrial enterprises and analyzing the possibility of using foreign experience in improving the process of micro-credit to stimulate their development.Research Methodology. The study was conducted on the material of publications on various aspects of the microcredit industry small and medium businesses, including international experience in micro-credit for small and medium-sized industrial enterprises.The article analyzes the functioning of the various models and micro-technologies, disclosed interoperability of commercial banks and microfinance institutions (MFIs in the framework of country-specific microcredit models, and formulated the immediate tasks and activities of the government and regulatory authorities of the Russian Federation aimed at improving the financing of small and medium-sized industrial enterprises.Sounding the findings suggest that the funds to support small and medium-sized industrial enterprises should be more actively attract private investment in the implementation of industrial and innovative development of their projects. In this case, the intensification of financial-credit and investment support to small and medium-sized industrial enterprises, integration and optimization of the different sources of financial resources create favorable conditions for their access to financial and credit resources, and improvement of financial and credit support mechanisms will enhance their responsibility for use granted resources and contribute to their development. At the same time, the main focus of the use of public

Ultrasonic testing results of fatigue cracks in PWR mock-up

International Nuclear Information System (INIS)

Gondard, C.

1990-01-01

The Ispra Joint Research Center has entered, since many years a study on fatigue crack propagation in PWR reactor vessels. The objective of this study is to establish a relation between the size and the location of defects and the lifetime of the vessel. For verifying the theoretical models validity a mockup has been built. This document gives the results of CEA for 6 in service inspection during 5 years [fr

Bank Size and Small- and Medium-sized Enterprise (SME) Lending: Evidence from China

Science.gov (United States)

SHEN, YAN; SHEN, MINGGAO; XU, ZHONG; BAI, YING

2014-01-01

Summary Using panel data collected in 2005, we evaluate how bank size, discretion over credit, incentive schemes, competition, and the institutional environment affect lending to small- and medium-sized enterprises in China. We deal with the endogeneity problem using instrumental variables, and a reduced-form approach is also applied to allow for weak instruments in estimation. We find that total bank asset is an insignificant factor for banks’ decision on small- and medium-enterprise (SME) lending, but more local lending authority, more competition, carefully designed incentive schemes, and stronger law enforcement encourage commercial banks to lend to SMEs. PMID:26052179

Bank Size and Small- and Medium-sized Enterprise (SME) Lending: Evidence from China.

Science.gov (United States)

Shen, Yan; Shen, Minggao; Xu, Zhong; Bai, Ying

2009-04-01

Using panel data collected in 2005, we evaluate how bank size, discretion over credit, incentive schemes, competition, and the institutional environment affect lending to small- and medium-sized enterprises in China. We deal with the endogeneity problem using instrumental variables, and a reduced-form approach is also applied to allow for weak instruments in estimation. We find that total bank asset is an insignificant factor for banks' decision on small- and medium-enterprise (SME) lending, but more local lending authority, more competition, carefully designed incentive schemes, and stronger law enforcement encourage commercial banks to lend to SMEs.

Horizontal loading test by whole model specimen simulating inner concrete structure of PWR type nuclear power plant

International Nuclear Information System (INIS)

Furuya, Noriyuki; Sekine, Masataka; Kimura, Kozo; Yamaguchi, Yoshihiro; Yamaguchi, Tsuneo; Takeda, Toshikazu

1985-01-01

The Nuclear Power Engineering Test Center has performed a horizontal loading test by a whole model specimen simulating the inner concrete structure of a PWR type nuclear power plant in order to investigate restoring force characteristics of reactor buildings. This report describes the results of examination of applicability to the test results of analysis methods based on elastic theory. The analysis results of elastic stiffness, concrete cracking load, rebar yielding loads and ultimate strength were compared with the test results. According to this examination, it is recognized that even these analysis methods based on elastic theory are comparatively effective for analysis of an inner concrete structure of fairly complex configuration, although there are limits of the scope of applicability. (author)

The role of independent intermediaries. The case of small and medium-sized exporters

DEFF Research Database (Denmark)

Madsen, Tage Koed; Moen, Øystein; Hammervold, Randi

2012-01-01

and participation rather than just discrete types of intermediary modes (agents, importers, dealers, etc.). Associations with performance as well as the role of product and distributor characteristics are analyzed. Empirical data based on a sample of product-market ventures in 227 small and medium-sized Norwegian...... export firms are analysed by a structural equation modelling approach. The article provides empirical evidence that managers keep control of decision making to an extent that may have a negative impact on export performance. The empirical study indicates that firms should participate more in task......The article examines how small and medium-sized exporters collaborate with intermediaries in foreign markets by studying the level of control, i.e. the delegation of decisions rights and task solution. It goes one step further than previous research, since it examines degrees of control...

Internationalization of small and medium-sized enterprises

Directory of Open Access Journals (Sweden)

Zoran Paunović

2010-06-01

Full Text Available Highly developed countries, which are similar to Croatia by size and population, highlight the importance of small and medium-sized enterprises (SMEs as holders of export activities and key factors in raising the competitiveness of the entire economy. In this paper authors research the concept of internationalization of SMEs. Analyzing the influence of decision makers on the process of internationalization and showing its advantages and disadvantages for the respective company and country, this research introduces the most common models on the basis of which this process is implemented in practice. A case study of a small export company from Croatia illustrates the process of internationalization to the U.S. market and provides useful information to the companies which are planning to enter new markets.

Small-sized reverberation chamber for the measurement of sound absorption

International Nuclear Information System (INIS)

Rey, R. del; Alba, J.; Bertó, L.; Gregori, A.

2017-01-01

This paper presents the design, construction, calibration and automation of a reverberation chamber for small samples. A balance has been sought between reducing sample size, to reduce the manufacturing costs of materials, and finding the appropriate volume of the chamber, to obtain reliable values at high and mid frequencies. The small-sized reverberation chamber, that was built, has a volume of 1.12 m3 and allows for the testing of samples of 0.3 m2. By using diffusers, to improve the diffusion degree, and automating measurements, we were able to improve the reliability of the results, thus reducing test errors. Several comparison studies of the measurements of the small-sized reverberation chamber and the standardised reverberation chamber are shown, and a good degree of adjustment can be seen between them, within the range of valid frequencies. This paper presents a small laboratory for comparing samples and making decisions before the manufacturing of larger sizes. [es

SIVAR - Computer code for simulation of fuel rod behavior in PWR during fast transients

International Nuclear Information System (INIS)

Dias, A.F.V.

1980-10-01

Fuel rod behavior during a stationary and a transitory operation, is studied. A computer code aiming at simulating PWR type rods, was developed; however, it can be adapted for simulating other type of rods. A finite difference method was used. (E.G.) [pt

Qualification tests for PWR control element drive mechanism

Energy Technology Data Exchange (ETDEWEB)

Kim, In Yong; Jin, Choon Eon; Choi Suhn [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1996-01-01

It is necessary to perform the qualification test for the magnetic jack type CEDM to show the design compatibility because the CEDM is composed of many mechanical and electrical components complicatedly. ABB-CE performed various qualification tests during the development of the System80 CEDM to which Korea Standard Nuclear Plant (KSNP) CEDM referred. The qualification test for the CEDM is classified into the performance test and the dynamic test. The performance test is to verify operability of the CEDM, and the dynamic test is to find dynamic characteristics and to verify the structural integrity if the CEDM for the seismic accidents. Described in this report are the test requirements, the test facilities and the test methods for the performance and the dynamic qualification tests of the PWR magnetic jack type CEDM. The impacts of the design changes in the Korea Next Generation Reactor (KNGR) on the KSNP CEDM were analyzed to present the necessity for the tests. This report also proposes the facilities to perform the tests in KAERI including reasonable schedule for the tests. Attached to this report is the summary of qualification tests of System 80 CEDM performed by ABB-CE. 20 figs., 16 tabs., 21 refs. (Author) .new.

Qualification tests for PWR control element drive mechanism

International Nuclear Information System (INIS)

Kim, In Yong; Jin, Choon Eon; Choi Suhn

1996-01-01

It is necessary to perform the qualification test for the magnetic jack type CEDM to show the design compatibility because the CEDM is composed of many mechanical and electrical components complicatedly. ABB-CE performed various qualification tests during the development of the System80 CEDM to which Korea Standard Nuclear Plant (KSNP) CEDM referred. The qualification test for the CEDM is classified into the performance test and the dynamic test. The performance test is to verify operability of the CEDM, and the dynamic test is to find dynamic characteristics and to verify the structural integrity if the CEDM for the seismic accidents. Described in this report are the test requirements, the test facilities and the test methods for the performance and the dynamic qualification tests of the PWR magnetic jack type CEDM. The impacts of the design changes in the Korea Next Generation Reactor (KNGR) on the KSNP CEDM were analyzed to present the necessity for the tests. This report also proposes the facilities to perform the tests in KAERI including reasonable schedule for the tests. Attached to this report is the summary of qualification tests of System 80 CEDM performed by ABB-CE. 20 figs., 16 tabs., 21 refs. (Author) .new

Decision Support on Small size Passive Samples

Directory of Open Access Journals (Sweden)

Vladimir Popukaylo

2018-05-01

Full Text Available A construction technique of adequate mathematical models for small size passive samples, in conditions when classical probabilistic-statis\\-tical methods do not allow obtaining valid conclusions was developed.

Optimal sizing of a run-of-river small hydropower plant

International Nuclear Information System (INIS)

Anagnostopoulos, John S.; Papantonis, Dimitris E.

2007-01-01

The sizing of a small hydropower plant of the run-of-river type is very critical for the cost effectiveness of the investment. In the present work, a numerical method is used for the optimal sizing of such a plant that comprises two hydraulic turbines operating in parallel, which can be of different type and size in order to improve its efficiency. The study and analysis of the plant performance is conducted using a newly developed evaluation algorithm that simulates in detail the plant operation during the year and computes its production results and economic indices. A parametric study is performed first in order to quantify the impact of some important construction and operation factors. Next, a stochastic evolutionary algorithm is implemented for the optimization process. The examined optimization problem uses data of a specific site and is solved in the single and two-objective modes, considering, together with economic, some additional objectives, as maximization of the produced energy and the best exploitation of the water stream potential. Analyzing the results of various optimizations runs, it becomes possible to identify the most advantageous design alternatives to realize the project. It was found that the use of two turbines of different size can enhance sufficiently both the energy production of the plant and the economic results of the investment. Finally, the sensitivity of the plant performance to other external parameters can be easily studied with the present method, and some indicative results are given for different financial or hydrologic conditions

First application of hollow fiber filter for PWR condensate polishing

International Nuclear Information System (INIS)

Tsuda, S.; Otoha, K.; Takiguchi, H.

2002-01-01

In Tsuruga Unit-2 (PWR 1160 MWe commenced commercial operation in 1987), current procedure for secondary system clean-up before start-up had prolonged outage time and had consumed a huge amount of de-ionized (DI) water. In addition, iron oxide in condensate had accelerated the degradation of condensate demineralizer (CD) resin. The corrosion product of iron could also influence the secondary side corrosion of steam generator (SG) tubing if it intruded into SG through CD. To solve these problems, Japan Atomic Power Company (JAPC) decided to introduce hollow fiber filter (HFF) type condensate filter into Tsuruga-2, as the first application to PWR in the world. Because of retro-fitted HFF in Tsuruga Unit-2, limitations for installation space and flow resistance in condensate system and cost reduction required new design for compact and low differential pressure system and for long life filter module. JAPC and ORGANO assessed methodologies to achieve these goals. An advanced HFF system, including a newly developed compact HFF module design, was installed at Tsuruga Unit-2 in 1997 based on the assessment. During the 5 years since the installation, the HFF system has provided excellent crud removal that enables to shorten the outage period and to reduce DI water consumption drastically. Stable differential pressure (dP) trend of the HFF system indicates an expected module life of more than 7 years, with backwash cleaning required only 2 or 3 times per year. In addition to providing the expected operating cost reduction and improved SG tube integrity, numerous additional benefits have resulted from the retrofit. (authors)

76 FR 70680 - Small Business Size Standards: Real Estate and Rental and Leasing

Science.gov (United States)

2011-11-15

... eligibility for Federal small business assistance, SBA establishes small business size definitions (referred... business definition or size standard established by the SBA Administrator. SBA considers as part of its... SMALL BUSINESS ADMINISTRATION 13 CFR Part 121 RIN 3245-AG28 Small Business Size Standards: Real...

77 FR 76215 - Small Business Size Regulations, Small Business Innovation Research (SBIR) Program and Small...

Science.gov (United States)

2012-12-27

... overall goal of simplification and maximization of benefits for small businesses, SBA proposed amendments... franchisee. F. Section 121.704--When SBA Determines Size and Eligibility SBA's proposed regulations for the...

76 FR 8221 - Small Business Size Regulations; 8(a) Business Development/Small Disadvantaged Business Status...

Science.gov (United States)

2011-02-11

... Vol. 76 Friday, No. 29 February 11, 2011 Part VII Small Business Administration 13 CFR Parts 121 and 124 Small Business Size Regulations; 8(a) Business Development/Small Disadvantaged Business Status... Regulations#0;#0; [[Page 8222

Microcomputer simulation of PWR power plant pressurizer

International Nuclear Information System (INIS)

Araujo, L.R.A. de; Calixto Neto, J.; Martinez, A.S.; Schirru, R.

1990-01-01

It is presented a method for the simulation of the pressurizer behavior of a PWR power plant. The method was implanted in a microcomputer, and it considers all the devices for the pressure control (spray and relief valves, heaters, controller, etc.). The physical phenomena and the PID (Proportional + Integral + Derivative) controller were mathematically represented by linear relations, uncoupled, discretized in the time. There are three different algorithms which take into account the non-linear effects introduced by the variation of the physical properties due to the temperature and pressure, and also the mutual effects between the physical phenomena and the PID controller. (author)

Pressure loss tests for DR-BEP of fullsize 17 x 17 PWR fuel assembly

International Nuclear Information System (INIS)

Chung, Moon Ki; Chun, Se Young; Chang, Seok Kyu; Won, Soon Youn; Cho, Young Rho; Kim, Bok Deuk; Min, Kyoung Ho

1993-01-01

This report describes the conditions, procedure and results in the pressure loss tests carried out for a double grid type debris resistance bottom end piece (DR-BEP) designed by KAERI. In this test, the pressure loss coefficients of the full size 17 x 17 PWR simulated fuel assembly with DR-BET and with standard-BEP were measured respectively, and the pressure loss coefficients of DR-BEP were compared with the coefficients of STD-BET. The test conditions fall within the ranges of loop pressure from 5.2 to 45 bar, loop temperature from 27 to 221 deg C and Reynolds number in fuel bundle from 2.17 x 10 4 to 3.85 x 10 5 . (Author) 5 refs., 18 figs., 5 tabs

Transient study of a PWR pressurizer

International Nuclear Information System (INIS)

Sotoma, H.

1973-01-01

An appropriate method for the calculation and transient performance of the pressurizer of a pressurized water reactor is presented. The study shows a digital program of simulation of pressurizer dynamics based on the First Law of Thermodynamic and Laws of Heat and Mass Transfer. The importance of the digital program that was written for a pressurizer of PWR, lies in the fact that, this can be of practical use in the safety analysis of a reactor of Angra dos Reis type with a power of about 500 M We. (author)

Study on thermal-hydraulics during a PWR reflood phase

Energy Technology Data Exchange (ETDEWEB)

Iguchi, Tadashi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

1998-10-01

In-core thermal-hydraulics during a PWR reflood phase following a large-break LOCA are quite unique in comparison with two-phase flow which has been studied widely in previous researches, because the geometry of the flow path is complicated (bundle geometry) and water is at extremely low superficial velocity and almost under stagnant condition. Hence, some phenomena realized during a PWR reflood phase are not understood enough and appropriate analytical models have not been developed, although they are important in a viewpoint of reactor safety evaluation. Therefore, author investigated some phenomena specified as important issues for quantitative prediction, i.e. (1) void fraction in a bundle during a PWR reflood phase, (2) effect of radial core power profile on reflood behavior, (3) effect of combined emergency core coolant injection on reflood behavior, and (4) the core separation into two thermal-hydraulically different regions and the in-core flow circulation behavior observed during a combined injection PWR reflood phase. Further, author made analytical models for these specified issues, and succeeded to predict reflood behaviors at representative types of PWRs, i.e.cold leg injection PWRs and Combined injection PWRs, in good accuracy. Above results were incorporated into REFLA code which is developed at JAERI, and they improved accuracy in prediction and enlarged applicability of the code. In the present study, models were intended to be utilized in a practical use, and hence these models are simplified ones. However, physical understanding on the specified issues in the present study is basic and principal for reflood behavior, and then it is considered to be used in a future advanced code development and improvement. (author). 110 refs.

An analysis of transients in the PWR downcomer

International Nuclear Information System (INIS)

Jovanovic, A.

1981-01-01

The paper deals with the problem of determining non-stationary temperature field in the downcomer of a PWR type reactor. For this purpose, an analytical model has been developed. The model covers five components of (PWR - Krsko) downcomer: the core-barrel, floor between the core-barrel and the thermal shield, the thermal shield, flow between the thermal shield and the reactor vessel wall, the reactor vessel wall. The model includes internal heat generation in metal structures. The governing equations of the model have been written in the finite difference explicit form. The system of resulting algebraic equations was solved bu Gauss-Seidel method, using a modular computer code. Several characteristic transients were examined (step and continuous change of fluid temperature at the inlet nozzle). Also, an analysis of main parameters (heat transfer coefficient and flow rate) has been performed. The model is intended to be used as basics for further development of a more realistic model that could be used for practical safety analysis. (author)

Evaluation of Kirkwood-Buff integrals via finite size scaling: a large scale molecular dynamics study

Science.gov (United States)

Dednam, W.; Botha, A. E.

2015-01-01

Solvation of bio-molecules in water is severely affected by the presence of co-solvent within the hydration shell of the solute structure. Furthermore, since solute molecules can range from small molecules, such as methane, to very large protein structures, it is imperative to understand the detailed structure-function relationship on the microscopic level. For example, it is useful know the conformational transitions that occur in protein structures. Although such an understanding can be obtained through large-scale molecular dynamic simulations, it is often the case that such simulations would require excessively large simulation times. In this context, Kirkwood-Buff theory, which connects the microscopic pair-wise molecular distributions to global thermodynamic properties, together with the recently developed technique, called finite size scaling, may provide a better method to reduce system sizes, and hence also the computational times. In this paper, we present molecular dynamics trial simulations of biologically relevant low-concentration solvents, solvated by aqueous co-solvent solutions. In particular we compare two different methods of calculating the relevant Kirkwood-Buff integrals. The first (traditional) method computes running integrals over the radial distribution functions, which must be obtained from large system-size NVT or NpT simulations. The second, newer method, employs finite size scaling to obtain the Kirkwood-Buff integrals directly by counting the particle number fluctuations in small, open sub-volumes embedded within a larger reservoir that can be well approximated by a much smaller simulation cell. In agreement with previous studies, which made a similar comparison for aqueous co-solvent solutions, without the additional solvent, we conclude that the finite size scaling method is also applicable to the present case, since it can produce computationally more efficient results which are equivalent to the more costly radial distribution

Evaluation of Kirkwood-Buff integrals via finite size scaling: a large scale molecular dynamics study

International Nuclear Information System (INIS)

Dednam, W; Botha, A E

2015-01-01

Solvation of bio-molecules in water is severely affected by the presence of co-solvent within the hydration shell of the solute structure. Furthermore, since solute molecules can range from small molecules, such as methane, to very large protein structures, it is imperative to understand the detailed structure-function relationship on the microscopic level. For example, it is useful know the conformational transitions that occur in protein structures. Although such an understanding can be obtained through large-scale molecular dynamic simulations, it is often the case that such simulations would require excessively large simulation times. In this context, Kirkwood-Buff theory, which connects the microscopic pair-wise molecular distributions to global thermodynamic properties, together with the recently developed technique, called finite size scaling, may provide a better method to reduce system sizes, and hence also the computational times. In this paper, we present molecular dynamics trial simulations of biologically relevant low-concentration solvents, solvated by aqueous co-solvent solutions. In particular we compare two different methods of calculating the relevant Kirkwood-Buff integrals. The first (traditional) method computes running integrals over the radial distribution functions, which must be obtained from large system-size NVT or NpT simulations. The second, newer method, employs finite size scaling to obtain the Kirkwood-Buff integrals directly by counting the particle number fluctuations in small, open sub-volumes embedded within a larger reservoir that can be well approximated by a much smaller simulation cell. In agreement with previous studies, which made a similar comparison for aqueous co-solvent solutions, without the additional solvent, we conclude that the finite size scaling method is also applicable to the present case, since it can produce computationally more efficient results which are equivalent to the more costly radial distribution

An analytical means of comparing the rates of different agglomeration mechanisms, and its application to a PWR containment

International Nuclear Information System (INIS)

Payne, J.F.B.; Skyrme, G.

1993-01-01

In this paper, agglomeration behaviour is described analytically and this quantitative description is used as a means of comparing the rates of different agglomeration mechanisms. The comparison depends on the average particle size but does not require detailed knowledge of the particle size distribution. The method is applied to the agglomeration of an aerosol suspended in the atmosphere of a pressurized water reactor (PWR) containment building. The rate of turbulent agglomeration in the PWR containment depends upon the degree of turbulence at each point in the fluid, expressed as the turbulent energy dissipation rate, ε. An approximate model of the containment flow is solved to obtain a satisfactory estimate of the functions of ε needed in evaluating turbulent agglomeration rates. (Author)

77 FR 11001 - Small Business Size Standards: Health Care and Social Assistance

Science.gov (United States)

2012-02-24

..., and (3) within a specific small business definition or size standard established by the SBA... SMALL BUSINESS ADMINISTRATION 13 CFR Part 121 RIN 3245-AG30 Small Business Size Standards: Health Care and Social Assistance AGENCY: U.S. Small Business Administration. ACTION: Proposed rule. SUMMARY...

77 FR 55755 - Small Business Size Standards: Agriculture, Forestry, Fishing, and Hunting

Science.gov (United States)

2012-09-11

... operation; and (3) within a specific small business definition or size standard established by SBA... SMALL BUSINESS ADMINISTRATION 13 CFR Part 121 RIN 3245-AG43 Small Business Size Standards: Agriculture, Forestry, Fishing, and Hunting AGENCY: U.S. Small Business Administration. ACTION: Proposed rule...

76 FR 27952 - Small Business Size Standards: Professional, Scientific and Technical Services.

Science.gov (United States)

2011-05-13

... Administration (SBA or Agency) proposed to increase small business size standards for 35 industries and one sub... SMALL BUSINESS ADMINISTRATION 13 CFR Part 121 RIN 3245-AG07 Small Business Size Standards: Professional, Scientific and Technical Services. AGENCY: U.S. Small Business Administration. ACTION: Proposed...

Prevention and mitigation of steam-generator water-hammer events in PWR plants

International Nuclear Information System (INIS)

Han, J.T.; Anderson, N.

1982-11-01

Water hammer in nuclear power plants is an unresolved safety issue under study at the NRC (USI A-1). One of the identified safety concerns is steam generator water hammer (SGWH) in pressurized-water reactor (PWR) plants. This report presents a summary of: (1) the causes of SGWH; (2) various fixes employed to prevent or mitigate SGWH; and (3) the nature and status of modifications that have been made at each operating PWR plant. The NRC staff considers that the issue of SGWH in top feedring designs has been technically resolved. This report does not address technical findings relevant to water hammer in preheat type steam generators. 10 figures, 2 tables

Visual SLAM and Moving-object Detection for a Small-size Humanoid Robot

Directory of Open Access Journals (Sweden)

Yin-Tien Wang

2010-09-01

Full Text Available In the paper, a novel moving object detection (MOD algorithm is developed and integrated with robot visual Simultaneous Localization and Mapping (vSLAM. The moving object is assumed to be a rigid body and its coordinate system in space is represented by a position vector and a rotation matrix. The MOD algorithm is composed of detection of image features, initialization of image features, and calculation of object coordinates. Experimentation is implemented on a small-size humanoid robot and the results show that the performance of the proposed algorithm is efficient for robot visual SLAM and moving object detection.

An integrated methodology to evaluate a spent nuclear fuel storage system

International Nuclear Information System (INIS)

Yoon, Jeong Hyoun

2008-02-01

This study introduced a methodology that can be applied for development of a dry storage system for spent nuclear fuels. It consisted of several design activities that includes development of a simplified program to analyze the amount of spent nuclear fuels from reflecting the practical situation in spent nuclear fuel management and a simplified program to evaluate the cost of 4 types of representing storage system to choose the most competitive option considering economic factor. As verification of the implementation of the reference module to practical purpose, a simplified thermal analysis code was suggested that can see fulfillment of limitation of temperature in long term storage and oxidation analysis. From the thermal related results, the reference module can accommodate full range of PHWR spent nuclear fuels and significant portion of PWR ones too. From the results, the reference storage system can be concluded that has fulfilled the important requirements in terms of long term integrity and radiological safety. Also for the purpose of solving scattered radiation along with deep penetration problems in cooling storage system, small but efficient design alternation was suggested together with its efficiency that can reduce scattered radiation by 1/3 from the original design. Along with the countermeasure for the shielding problem, in consideration of PWR spent nuclear fuels, simplified criticality analysis methodology retaining conservativeness was proposed. The results show the reference module is efficient low enrichment PWR spent nuclear fuel and even relatively high enrichment fuels too if burnup credit is taken. As conclusive remark, the methodology is simple but efficient to plan a concept design of convective cooling type of spent nuclear fuels storage. It can be also concluded that the methodology derived in this study and the reference module has feasibility in practical implementation to mitigate the current complex situation in spent fuel

Effects of delayed RCP trip during SBLOCA in PWR

International Nuclear Information System (INIS)

Montero-Mayorga, J.; Queral, C.; Gonzalez-Cadelo, J.

2014-01-01

Highlights: • Review of RCP trip issue in case of SBLOCA showing adequacy of present EOPs. • Risk assessment of a SBLOCA deterministic safety analysis by means of ISA methodology. • Evaluation of the probability of damage considering uncertainties in operator actuation times. • Application of ISA methodology to probabilistic safety analysis. • Obtaining of RCP trip available time as function of break size. - Abstract: After the Three Mile Island (TMI) accident, the issue of when to trip the Reactor Coolant Pumps (RCPs) in case of a Small Break Loss of Coolant Accident (SBLOCA) became very important. Several analyses were performed during the 1980s leading to the current Emergency Operating Procedures (EOPs). However these analyses have not been reviewed taking into account that several improvements have been performed in the last thirty years with respect to two phase-flow models, thermal–hydraulics codes and safety assessment methodologies. In this sense, this work has two main objectives: First of all, an assessment of the analyses carried out by Pressurizer Water Reactor (PWR) vendors after the TMI-2 accident with a model of Almaraz Nuclear Power Plant (NPP) for TRACE code (V 5.0 patch 1). On the other hand, Integrated Safety Assessment (ISA) methodology is applied to explore this matter. Such methodology has been developed by the Spanish Nuclear Safety Council (CSN) and it is an adequate method to perform analyses in nuclear safety in which the uncertainties in operator actuation time play an important role. The main conclusions obtained from this work are that, the current EOPs are adequate to manage a SBLOCA sequence in a suitable manner and that ISA methodology is a powerful tool that provides accurate information to the analyst in order to verify the robustness of the EOPs and to perform the safety assessment of both, deterministic and probabilistic safety analysis

Integration of Small-Diameter Wood Harvesting in Early Thinnings using the Two pile Cutting Method

Energy Technology Data Exchange (ETDEWEB)

Kaerhae, Kalle (Metsaeteho Oy, P.O. Box 101, FI-00171 Helsinki (Finland))

2008-10-15

Metsaeteho Oy studied the integrated harvesting of industrial roundwood (pulpwood) and energy wood based on a two-pile cutting method, i.e. pulpwood and energy wood fractions are stacked into two separate piles when cutting a first-thinning stand. The productivity and cost levels of the integrated, two-pile cutting method were determined, and the harvesting costs of the two-pile method were compared with those of conventional separate wood harvesting methods. In the time study, when the size of removal was 50 dm3, the productivity in conventional whole-tree cutting was 6% higher than in integrated cutting. With a stem size of 100 dm3, the productivity in whole-tree cutting was 7% higher than in integrated cutting. The results indicated, however, that integrated harvesting based on the two-pile method enables harvesting costs to be decreased to below the current cost level of separate pulpwood harvesting in first thinning stands. The greatest cost-saving potential lies in small-sized first thinnings. The results showed that, when integrated wood harvesting based on the two-pile method is applied, the removals of both energy wood and pulpwood should be more than 15-20 m3/ha at the harvesting sites in order to achieve economically viable integrated procurement

78 FR 45051 - Small Business Size Standards; Support Activities for Mining; Correction

Science.gov (United States)

2013-07-26

... Regulations by increasing small business size standards for three of the four industries in North American... SMALL BUSINESS ADMINISTRATION 13 CFR Part 121 RIN 3245-AG44 Small Business Size Standards; Support Activities for Mining; Correction AGENCY: U.S. Small Business Administration. ACTION: Final rule; correction...

Liquid radioactive waste processing improvement of PWR nuclear power plants; Melhorias no processamento de rejeitos liquidos radioativos de usinas nucleares PWR

Energy Technology Data Exchange (ETDEWEB)

Nery, Renata Wolter dos Reis; Martinez, Aquilino Senra; Monteiro, Jose Luiz Fontes [Universidade Federal, Rio de Janeiro, RJ (Brazil). Coordenacao dos Programas de Pos-graduacao de Engenharia. Programa de Engenharia Nuclear]. E-mail: wolter@eletronuclear.gov.br; monteiro@peq.coppe.ufrj.br; aquilinosenra@lmp.ufrj.br

2005-07-01

The study evaluate an inorganic ion exchange to process the low level liquid radwaste of PWR nuclear plants, so that the level of the radioactivity in the effluents and the solid waste produced during the treatment of these liquid radwaste can be reduced. The work compares two types of ion exchange materials, a strong acid cation exchange resin, that is the material typically used to remove radionuclides from PWR nuclear plants wastes, and a mordenite zeolite. These exchange material were used to remove cesium from a synthetic effluent containing only this ion and another effluent containing cesium and cobalt. The breakthrough curves of the zeolite and resin using a fix bed reactor were compared. The results demonstrated that the zeolite is more efficient than the resin in removing cesium from a solution containing cesium and cobalt. The results also showed that a bed combining zeolite and resin can process more volume of an effluent containing cesium and cobalt than a bed resin alone. (author)

Small renal size in newborns with spina bifida: possible causes.

Science.gov (United States)

Montaldo, Paolo; Montaldo, Luisa; Iossa, Azzurra Concetta; Cennamo, Marina; Caredda, Elisabetta; Del Gado, Roberto

2014-02-01

Previous studies reported that children with neural tube defects, but without any history of intrinsic renal diseases, have small kidneys when compared with age-matched standard renal growth. The aim of this study was to investigate the possible causes of small renal size in children with spina bifida by comparing growth hormone deficiency, physical limitations and hyperhomocysteinemia. The sample included 187 newborns with spina bifida. Renal sizes in the patients were assessed by using maximum measurement of renal length and the measurements were compared by using the Sutherland monogram. According to the results, the sample was divided into two groups--a group of 120 patients with small kidneys (under the third percentile) and a control group of 67 newborns with normal kidney size. Plasma total homocysteine was investigated in mothers and in their children. Serum insulin-like growth factor-1 (IGF-1) levels were measured. Serum IGF-1 levels were normal in both groups. Children and mothers with homocysteine levels >10 μmol/l were more than twice as likely to have small kidneys and to give to birth children with small kidneys, respectively, compared with newborns and mothers with homocysteine levels <10 μmol/l. An inverse correlation was also found between the homocysteine levels of mothers and kidney sizes of children (r = - 0.6109 P ≤ 0.01). It is highly important for mothers with hyperhomocysteinemia to be educated about benefits of folate supplementation in order to reduce the risk of small renal size and lower renal function in children.

Replacement of the control and instrumentation system with the microprocessor based systems in Japanese PWR plants

International Nuclear Information System (INIS)

Hayashi, N.

1998-01-01

In Ohi Units 3 and 4, Ikata Unit 3, and Genkai Units 3 and 4, the latest of PWR plants now under operation in Japan, the reactor control system and turbine control system employ the microprocessor base digital control systems with a view to improving reliability, operability and maintainability. In the next stage plants, another application of such digital system is also planned for the instrumentation rack for the reactor protection system for further improvement. On the other hand, in Mihama Unit 1, the first of domestic PWR plants, and later plants except for the latest 5 plants, analog control systems are employed for the instrumentation racks. For the analog control systems of these plants, FOXBORO H-Line instruments, equivalent domestic box type instruments or WH7300 Series card type instruments were initially employed, and later replaced with domestic card type control systems after 10-15 year operation. However, 8-12 years have passed since these replacements, so the 15th year generally quoted as an interval for replacing C and I systems is near at hand. This is the time to consider next replacement. This replacement will be based on the latest digital technology. However, it is not practical way for the existing plants to apply the same integrated digital C and I system configuration for the next stage plants, because it requires the drastic change of the C and I system configuration and significant cost-up. Therefore, we must investigate the optimum digital C and I system configuration for the existing system. (author)

Investigation of size effect on film type haptic actuator made with cellulose acetate

International Nuclear Information System (INIS)

Kim, Sang-Youn; Kim, Jaehwan; Kim, Ki-Baek

2014-01-01

The most important factor in haptic interaction with hand-held devices is to develop a thin film type actuator which can be easily inserted into the devices and create vibrotactile signals with wide frequency bandwidth. This paper reports a film type vibrotactile actuator which is tiny enough to be embedded into small hand-held devices. The vibration mechanism and experiment results for the suggested vibrotactile actuator are explained. The aim of the actuator is to convey a vibrotactile force greater than a human’s vibrotactile threshold with broad frequency bandwidth to users. To achieve the requirement, we fabricate a film type vibrotactile actuator with cellulose acetate. When an AC voltage is applied to the actuator, the cellulose acetate film gets charged and then generates vibration. The suggested vibrotactile actuator is fabricated in two sizes: 50 mm × 25 mm and 25 mm × 25 mm. For each size of actuator, three kinds of actuator are fabricated with different pillar materials to support the cellulose acetate films. An experiment for measuring vibrational amplitude is conducted over a wide frequency range of actuation voltage. It is known that the proposed film type actuator is feasible for haptic application in the small hand-held devices. (paper)

Improved liquid waste processing system of PWR plant

International Nuclear Information System (INIS)

Suehiro, Kazuyasu

1977-01-01

Mitsubishi Heavy Industries, Ltd. has engaged in the improvement and enhancement of waste-processing facilities for PWR power stations, and recently established the improved processing system. With this system, it becomes possible to contain radioactive waste gas semi-permanently within plants and to recycle waste liquid after the treatment, thus to make the release of radioactive wastes practically zero. The improved system has the following features, namely the recycling system is adopted, drain is separated and each separated drain is treated by specialized process, the reboiler type evaporator and the reverse osmosis equipment are used, and the leakless construction is adopted for the equipments. The radioactive liquid wastes in PWR power stations are classified into coolant drain, drain from general equipments, chemical drain and cleaning water. The outline of the improved processing system and the newly developed equipments such as the reboiler type evaporator and the reverse osmosis equipment are explained. With the evaporator, the concentration rate of waste liquid can be raised to about three times, and foaming waste can be treated efficiently. The decontamination performance is excellent. The reverse osmosis treatment is stable and reliable method, and is useful for the treatment of cleaning water. It is also effective for concentrating treatment. The unmanned automatic operation is possible. (Kako, I.)

VOLTTRON™: Tech-to-Market Best-Practices Guide for Small- and Medium-Sized Commercial Buildings

Energy Technology Data Exchange (ETDEWEB)

Cort, Katherine A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Haack, Jereme N. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Katipamula, Srinivas [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Nicholls, Andrew K. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

2016-07-11

VOLTTRON™ is an open-source distributed control and sensing platform developed by Pacific Northwest National Laboratory for the U.S. Department of Energy. It was developed to be used by the Office of Energy Efficiency and Renewable Energy to support transactive controls research and deployment activities. VOLTTRON is designed to be an overarching integration platform that could be used to bring together vendors, users, and developers and enable rapid application development and testing. The platform is designed to support modern control strategies, including the use of agent- and transaction-based controls. It also is designed to support the management of a wide range of applications, including heating, ventilation, and air-conditioning systems; electric vehicles; and distributed-energy and whole-building loads. This report was completed as part of the Building Technologies Office’s Technology-to-Market Initiative for VOLTTRON’s Market Validation and Business Case Development efforts. The report provides technology-to-market guidance and best practices related to VOLTTRON platform deployments and commercialization activities for use by entities serving small- and medium-sized commercial buildings. The report characterizes the platform ecosystem within the small- and medium-sized commercial building market and articulates the value proposition of VOLTTRON for three core participants in this ecosystem: 1) platform owners/adopters, 2) app developers, and 3) end-users. The report also identifies key market drivers and opportunities for open platform deployments in the small- and medium-sized commercial building market. Possible pathways to the market are described—laboratory testing to market adoption to commercialization. We also identify and address various technical and market barriers that could hinder deployment of VOLTTRON. Finally, we provide “best practice” tech-to-market guidance for building energy-related deployment efforts serving small- and

Sensitivity analysis on hot channel of PWR type reactors using matricial formalism

International Nuclear Information System (INIS)

Maciel, Edisson Savio G.; Andrade Lima, Fernando Roberto de; Lira, Carlos Alberto B.O.

1995-01-01

The matricial formalism of the perturbation theory is applied in a simplified model to study the hot channel of PWR reactors. Mass, linear momentum and energy conservation equations and appropriated heat transfer and fluid mechanics correlations describe the discretized system. After calculating system's thermalhydraulic properties, the matricial formalism is applied and the sensitivity coefficients are determined for each case of interest. Comparisons between perturbative method and direct results of the model have shown good agreement which demonstrates that the matricial formalism is an important tool for discretized system analysis. (author). 6 refs, 2 tabs

Lessons learned from OECD/CSNI ISP on small break LOCA: final report

International Nuclear Information System (INIS)

1996-07-01

This document presents an overview of the results obtained from five recent OECD/CSNI International Standard Problems (ISPs) dealing with phenomenologies typical of Small Break LOCA in PWR nuclear power plants of western design. The experiment in four Integral test Facilities, Lobi, Spes, Bethsy and Lstf and the recorded data from a steam generator tube rupture transient in the Belgian PWR of Doel, were taken as reference for the calculations. Relevant hardware characteristics of the facilities and of the plant are firstly given, including the correlation between key thermalhydraulic phenomena and the reference experimental scenarios. A statistical evaluation of the general data connected with each ISP is then presented. The lessons learned from the ISPs are then considered. Four areas have been identified: code deficiencies and capabilities, scaling of the data, progress in code capabilities and various additional aspects

Operation results of the secondary circuits of the French PWR type power plant park

International Nuclear Information System (INIS)

Mercier, J.P.

1984-01-01

Global results of performances realized since 1981 by the French PWR 900 MW power plants (installed power, availability, casual or planned shutdowns); analysis of the behaviour (casual unavailability) comparing together the performances of the different components in the secondary circuit; behaviour of the principal materials of the secondary circuit and their weight in the unavailabilities of the whole French nuclear park [fr

PREP-PWR-1.0: a WIMS-D/4 pre-processor code for the generation of data for PWR fuel assemblies

International Nuclear Information System (INIS)

Ball, G.

1991-06-01

The PREP-PWR-1.0 computer code is a substantially modified version of the PREWIM code which formed part of the original MARIA System (Report J.E.N. 543). PREP-PWR-1.0 is a comprehensive pre-processor code which generates input data for the WIMS-D/4.1 code (Report PEL 294) for PWR fuel assemblies, with or without control and burnable poison rods. This data is generated at various base and off-base conditions. The overall cross section generation methodology is described, followed by a brief overview of the model. Aspects of the base/off-base calculational scheme are outlined. Additional features of the code are described while the input data format of PREP-PWR-1.0 is listed. The sample problems and suggestions for further improvements to the code are also described. 2 figs., 2 tabs., 12 refs

Four-fluid model of PWR degraded cores

International Nuclear Information System (INIS)

Dearing, J.F.

1985-01-01

This paper describes the new two-dimensional, four-fluid fluid dynamics and heat transfer (FLUIDS) module of the MELPROG code. MELPROG is designed to give an integrated, mechanistic treatment of pressurized water reactor (PWR) core meltdown accidents from accident initiation to vessel melt-through. The code has a modular data storage and transfer structure, with each module providing the others with boundary conditions at each computational time step. Thus the FLUIDS module receives mass and energy source terms from the fuel pin module, the structures module, and the debris bed module, and radiation energy source terms from the radiation module. MELPROG, which models the reactor vessel, is also designed to model the vessel as a component in the TRAC/PF1 networking solution of a PWR reactor coolant system (RCS). The coupling between TRAC and MELPROG is implicit in the fluid dynamics of the reactor coolant (liquid water and steam) allowing an accurate simulation of the coupling between the vessel and the rest of the RCS during an accident. This paper deals specifically with the numerical model of fluid dynamics and heat transfer within the reactor vessel, which allows a much more realistic simulation (with less restrictive assumptions on physical behavior) of the accident than has been possible before

Feasibility of using a dose-area product ratio as beam quality specifier for photon beams with small field sizes.

Science.gov (United States)

Pimpinella, Maria; Caporali, Claudio; Guerra, Antonio Stefano; Silvi, Luca; De Coste, Vanessa; Petrucci, Assunta; Delaunay, Frank; Dufreneix, Stéphane; Gouriou, Jean; Ostrowsky, Aimé; Rapp, Benjamin; Bordy, Jean-Marc; Daures, Josiane; Le Roy, Maïwenn; Sommier, Line; Vermesse, Didier

2018-01-01

To investigate the feasibility of using the ratio of dose-area product at 20 cm and 10 cm water depths (DAPR 20,10 ) as a beam quality specifier for radiotherapy photon beams with field diameter below 2 cm. Dose-area product was determined as the integral of absorbed dose to water (D w ) over a surface larger than the beam size. 6 MV and 10 MV photon beams with field diameters from 0.75 cm to 2 cm were considered. Monte Carlo (MC) simulations were performed to calculate energy-dependent dosimetric parameters and to study the DAPR 20,10 properties. Aspects relevant to DAPR 20,10 measurement were explored using large-area plane-parallel ionization chambers with different diameters. DAPR 20,10 was nearly independent of field size in line with the small differences among the corresponding mean beam energies. Both MC and experimental results showed a dependence of DAPR 20,10 on the measurement setup and the surface over which D w is integrated. For a given setup, DAPR 20,10 values obtained using ionization chambers with different air-cavity diameters agreed with one another within 0.4%, after the application of MC correction factors accounting for effects due to the chamber size. DAPR 20,10 differences among the small field sizes were within 1% and sensitivity to the beam energy resulted similar to that of established beam quality specifiers based on the point measurement of D w . For a specific measurement setup and integration area, DAPR 20,10 proved suitable to specify the beam quality of small photon beams for the selection of energy-dependent dosimetric parameters. Copyright © 2017 Associazione Italiana di Fisica Medica. Published by Elsevier Ltd. All rights reserved.

Requirement analysis to promote small-sized E-waste collection from consumers.

Science.gov (United States)

Mishima, Kuniko; Nishimura, Hidekazu

2016-02-01

The collection and recycling of small-sized waste electrical and electronic equipment is an emerging problem, since these products contain certain amounts of critical metals and rare earths. Even if the amount is not large, having a few supply routes for such recycled resources could be a good strategy to be competitive in a world of finite resources. The small-sized e-waste sometimes contains personal information, therefore, consumers are often reluctant to put them into recycling bins. In order to promote the recycling of E-waste, collection of used products from the consumer becomes important. Effective methods involving incentives for consumers might be necessary. Without such methods, it will be difficult to achieve the critical amounts necessary for an efficient recycling system. This article focused on used mobile phones among information appliances as the first case study, since it contains relatively large amounts of valuable metals compared with other small-sized waste electrical and electronic equipment and there are a large number of products existing in the market. The article carried out surveys to determine what kind of recycled material collection services are preferred by consumers. The results clarify that incentive or reward money alone is not a driving force for recycling behaviour. The article discusses the types of effective services required to promote recycling behaviour. The article concludes that securing information, transferring data and providing proper information about resources and environment can be an effective tool to encourage a recycling behaviour strategy to promote recycling, plus the potential discount service on purchasing new products associated with the return of recycled mobile phones. © The Author(s) 2015.

Design Development and Verification of a System Integrated Modular PWR

International Nuclear Information System (INIS)

Kim, S.-H.; Kim, K. K.; Chang, M. H.; Kang, C. S.; Park, G.-C.

2002-01-01

An advanced PWR with a rated thermal power of 330 MW has been developed at the Korea Atomic Energy Research Institute (KAERI) for a dual purpose: seawater desalination and electricity generation. The conceptual design of SMART ( System-Integrated Modular Advanced ReacTor) with a desalination system was already completed in March of 1999. The basic design for the integrated nuclear desalination system is currently underway and will be finished by March of 2002. The SMART co-generation plant with the MED seawater desalination process is designed to supply forty thousand (40,000) tons of fresh water per day and ninety (90) MW of electricity to an area with approximately a ten thousand (100,000) population or an industrialized complex. This paper describes advanced design features adopted in the SMART design and also introduces the design and engineering verification program. In the beginning stage of the SMART development, top-level requirements for safety and economics were imposed for the SMART design features. To meet the requirements, highly advanced design features enhancing the safety, reliability, performance, and operability are introduced in the SMART design. The SMART consists of proven KOFA (Korea Optimized Fuel Assembly), helical once-through steam generators, a self-controlled pressurizer, control element drive mechanisms, and main coolant pumps in a single pressure vessel. In order to enhance safety characteristics, innovative design features adopted in the SMART system are low core power density, large negative Moderator Temperature Coefficient (MTC), high natural circulation capability and integral arrangement to eliminate large break loss of coolant accident, etc. The progression of emergency situations into accidents is prevented with a number of advanced engineered safety features such as passive residual heat removal system, passive emergency core cooling system, safeguard vessel, and passive containment over-pressure protection. The preliminary

75 FR 61604 - Small Business Size Standards; Accommodation and Food Services Industries

Science.gov (United States)

2010-10-06

... business assistance programs, SBA establishes small business size definitions (referred to as size... Administrator the responsibility for establishing small business definitions. The Act also requires that small business definitions vary to reflect industry differences. The supplementary information section of this...

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)

Small-Size High-Current Generators for X-Ray Backlighting

Science.gov (United States)

Chaikovsky, S. A.; Artyomov, A. P.; Zharova, N. V.; Zhigalin, A. S.; Lavrinovich, I. V.; Oreshkin, V. I.; Ratakhin, N. A.; Rousskikh, A. G.; Fedunin, A. V.; Fedushchak, V. F.; Erfort, A. A.

2017-12-01

The paper deals with the soft X-ray backlighting based on the X-pinch as a powerful tool for physical studies of fast processes. Proposed are the unique small-size pulsed power generators operating as a low-inductance capacitor bank. These pulse generators provide the X-pinch-based soft X-ray source (hν = 1-10 keV) of micron size at 2-3 ns pulse duration. The small size and weight of pulse generators allow them to be transported to any laboratory for conducting X-ray backlighting of test objects with micron space resolution and nanosecond exposure time. These generators also allow creating synchronized multi-frame radiographic complexes with frame delay variation in a broad range.

Features of digital photogrammetry methods application and image processing in small and medium-sized enterprises

Directory of Open Access Journals (Sweden)

Samsonova N. V.

2018-05-01

Full Text Available the paper discusses the methods of survey enterprises employees effective training to use modern measurement systems and the need for further photogrammetric processing of the results obtained. Attention is also paid to integrated learning, based primarily on web content, the introduction of a social component in the development and familiarization with new photogrammetric equipment and technologies in order to increase the competitiveness of engineering and research small and medium-sized enterprises.

PWR AXIAL BURNUP PROFILE ANALYSIS

International Nuclear Information System (INIS)

J.M. Acaglione

2003-01-01

The purpose of this activity is to develop a representative ''limiting'' axial burnup profile for pressurized water reactors (PWRs), which would encompass the isotopic axial variations caused by different assembly irradiation histories, and produce conservative isotopics with respect to criticality. The effect that the low burnup regions near the ends of spent fuel have on system reactivity is termed the ''end-effect''. This calculation will quantify the end-effects associated with Pressurized Water Reactor (PWR) fuel assemblies emplaced in a hypothetical 21 PWR waste package. The scope of this calculation covers an initial enrichment range of 3.0 through 5.0 wt% U-235 and a burnup range of 10 through 50 GWd/MTU. This activity supports the validation of the process for ensuring conservative generation of spent fuel isotopics with respect to criticality safety applications, and the use of burnup credit for commercial spent nuclear fuel. The intended use of these results will be in the development of PWR waste package loading curves, and applications involving burnup credit. Limitations of this evaluation are that the limiting profiles are only confirmed for use with the B andW 15 x 15 fuel assembly design. However, this assembly design is considered bounding of all other typical commercial PWR fuel assembly designs. This calculation is subject to the Quality Assurance Requirements and Description (QARD) because this activity supports investigations of items or barriers on the Q-list (YMP 2001)

Medium-sized water reactors for undeveloped regions

International Nuclear Information System (INIS)

Osmachkin, V. S.

2004-01-01

In the new century the growth of population and an increasing of energy demands together with the difficulties of fossil fuel supply are expected. It is important to find optimal ways in solving such problems without the climate warming. The nuclear power having many advantages in comparison with fossil fuel technologies could play the great role in near future. The Medium-Sized Nuclear Reactors for production of electricity, heat and fresh water are considered as a main direction of nuclear power applications in the developing world It is important to discuss the requirements to such nuclear plants for using in the Countries with Small and Medium Electricity Grids. Particularly, cost-benefit analysis of construction NPP has to include assessment of all type risks and effectiveness of plant. In the paper an attention is paid on Water Reactors designed on the basis of navy technology. Such compact PWR built on special mills and placed on special floating vessel could be used in undeveloped regions. Total plant can be transported to any point of World Ocean and return back to mill for repair or decommissioning after exhaustion of lifetime. It is expected that such reactors with innovative design approach, provision of high safety and proper economic efficiency, based on leasing procedures, could be very attractive for medium-sized and developing countries.(author)

Prognostic Importance of Small Prostate Size in Men Receiving Definitive Prostate Brachytherapy

International Nuclear Information System (INIS)

Taira, Al V.; Merrick, Gregory S.; Galbreath, Robert W.; Butler, Wayne M.; Adamovich, Edward; Wallner, Kent E.

2012-01-01

Purpose: To assess whether small prostate size is an adverse prognostic factor in men undergoing brachytherapy in the same manner in which it seems to be for men undergoing radical prostatectomy. Methods and Materials: From April 1995 to June 2008, 2024 patients underwent brachytherapy by a single brachytherapist. Median follow-up was 7.4 years. The role of small prostate size (≤20 cm 3 ) as a prognostic factor for biochemical progression-free survival, cause-specific survival, and all-cause mortality was investigated. The differences in survival between men with small and larger prostates were compared using Kaplan-Meier curves and log-rank tests. Results: Median prostate size for the entire cohort was 32.7 cm 3 . For the 167 men with small prostates, median prostate size was 17.4 cm 3 . There was no difference in biochemical progression-free survival (95.2% vs 96.2%, P=.603), cause-specific survival (97.7% vs 98.3%, P=.546), or all-cause mortality (78.0% vs 77.2%, P=.838) at 10 years for men with small prostates compared with men with larger prostates. On univariate and multivariate analysis, small prostate size was not associated with any of the primary outcome measures. Conclusion: Men with small prostates treated with brachytherapy have excellent outcomes and are at no higher risk of treatment failure than men with larger glands. High-quality implants with adequate margins seem sufficient to address the increased adverse risk factors associated with small prostate size.

Response of pressurized water reactor (PWR) to network power generation demands

International Nuclear Information System (INIS)

Schreiner, L.A.

1991-01-01

The flexibility of the PWR type reactor in terms of response to the variations of the network power demands, is demonstrated. The factors that affect the transitory flexibility and some design prospects that allow the reactor fits the requirements of the network power demands, are also discussed. (M.J.A.)

Digital Image Collections for Asian Religion and Art History in a Small-Sized Liberal Arts College

Directory of Open Access Journals (Sweden)

May Chang

2002-10-01

Full Text Available

頁次：6-15

size: small;">This paper reviews the establishment of a digital image library from 35mm slides to support faculty and student needs in a small-sized liberal arts college. The framework consists of central local resources, distributed faculty collections, and external resources. Standards and guidelines for digital preservation and access are also discussed. The pilot collections were multi-disciplinary resources in Middle East art and architecture and faculty slide collections in East Asian religions and Asian art history. Technical and management issues of integrating digital technology in the traditional slide library are also discussed.

COSMIC EVOLUTION OF SIZE AND VELOCITY DISPERSION FOR EARLY-TYPE GALAXIES

International Nuclear Information System (INIS)

Fan, L.; Lapi, A.; Bressan, A.; De Zotti, G.; Danese, L.; Bernardi, M.

2010-01-01

Massive (stellar mass M * ∼> 3 x 10 10 M sun ), passively evolving galaxies at redshifts z ∼> 1 exhibit on average physical sizes smaller, by factors ∼3, than local early-type galaxies (ETGs) endowed with the same stellar mass. Small sizes are in fact expected on theoretical grounds, if dissipative collapse occurs. Recent results show that the size evolution at z ∼ 1, where both compact and already extended galaxies are observed and the scatter in size is remarkably larger than it is locally. The presence at high redshift of a significant number of ETGs with the same size as their local counterparts, as well as ETGs with quite small size (∼ H (z). We demonstrate that the projected mass of compact, high-redshift galaxies and that of local ETGs within the same physical radius, the nominal half-luminosity radius of high-redshift ETGs, differ substantially in that the high-redshift ETGs are on average significantly denser. This result suggests that the physical mechanism responsible for the size increase should also remove mass from central galaxy regions (r ∼ 1, we predict the local velocity dispersion distribution function. On comparing it to the observed one, we show that velocity dispersion evolution of massive ETGs is fully compatible with the observed average evolution in size at constant stellar mass. Less massive ETGs (with stellar masses M * ∼ 10 M sun ) are expected to evolve less both in size and in velocity dispersion, because their evolution is essentially determined by supernova feedback, which cannot yield winds as powerful as those triggered by quasars. The differential evolution is expected to leave imprints in the size versus luminosity/mass, velocity dispersion versus luminosity/mass, and central black hole mass versus velocity dispersion relationships, as observed in local ETGs.

Innovativeness of small and medium-sized enterprises in the Republic of Serbia and countries of the European Union

Directory of Open Access Journals (Sweden)

Nikolić Miroljub

2015-01-01

Full Text Available The paper provides a comparative analysis of the results of innovation activities in enterprises of different size from EU member and candidate states during the period 2008-2010. Particularly, the paper considers position of Serbia compared to the EU average and average of some neighboring countries (Hungary, Slovenia, Croatia, Romania and Bulgaria. The results confirmed the existence of a high correlation between size of enterprise and its innovation activities. The percentage of innovation activities in large enterprises is higher than in small enterprises. Serbia is lagging behind the EU average, regarding innovativeness level in all types of enterprises (classified by size. However, when Serbia is compared to neighboring countries, the situation is much different. Small enterprises from Serbia are more innovative than small enterprises from neighboring countries. The situation is similar in medium-sized enterprises. Large enterprises from Serbia are more innovative than enterprises from Bulgaria, Hungary and Romania, and less innovative than large enterprises from Slovenia and Croatia.

Product Evaluation Task Force Phase Two report for BWR/PWR dissolver wastes

International Nuclear Information System (INIS)

Francis, A.J.

1990-01-01

It has been proposed that all Intermediate Level Wastes arising at Sellafield should be encapsulated prior to ultimate disposal. The Product Evaluation Task Force (PETF) was set up to investigate possible encapsulants and to produce an adequate data base to justify the preferred matrices. This report details the work carried out, under Phase 2 of the Product Evaluation Task Force programme, on BWR/PWR Dissolver Wastes. Three possible types of encapsulants for BWR/PWR Dissolver Wastes:- Inorganic cements, Polymer cements and Polymers are evaluated using the Kepner Tregoe decision analysis technique. This technique provides a methodology for scoring and ranking alternative options and evaluating any risks associated with an option. The analysis shows that for all four stages of waste management operations ie Storage, Transport, handling and emplacement, Disposal and Process, cement matrices are considerably superior to other potential matrices. A matrix, consisting of three parts Blast Furnace Slag (BFS) to one part Ordinary Portland Cement (OPC), is recommended for Phase 3 studies on BWR/PWR Dissolver Wastes. (author)

The analysis by several neutron transport methods of a small PWR model problem

International Nuclear Information System (INIS)

Halsall, M.J.

1980-09-01

A small model problem in x-y co-ordinate geometry is specified in detail to permit readers to make their own calculations. The problem is analysed using diffusion theory, differential and integral transport methods and a Monte Carlo code, and a best estimate eigenvalue is deduced. (author)

Modelling of pellet-cladding interaction in PWR's

International Nuclear Information System (INIS)

Esteves, A.M.; Silva, A.T. e.

1992-01-01

The pellet-cladding interaction that can occur in a PWR fuel rod design is modelled with the computer codes FRAPCON-1 and ANSYS. The fuel performance code FRAPCON-1 analyses the fuel rod irradiation behavior and generates the initial conditions for the localized fuel rod thermal and mechanical modelling in two and three-dimensional finite elements with ANSYS. In the mechanical modelling, a pellet fragment is placed in the fuel rod gap. Two types of fuel rod cladding materials are considered: Zircaloy and austenitic stainless steel. (author)

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

Water-hammer experimental set-up and water-hammer experimental study for new types of check valve applied to PWR

International Nuclear Information System (INIS)

Liu Hanxun.

1985-01-01

This paper describes a self-designed constant temperature water-hammer shock test rig with stainless steel loop in which deionized water is used as working medium. To conduct water-hammer shock simulation tests for the countercurrent phenomenon occurred in the process of shutting, stopping, parallelling and switching the coolant loops of nuclear reactor, a specially designed four-way switching valve and its pneumatic mechanism are used. Water-hammer experimental study is performed for two types of PWR's nonshock check valve with diameter of 150 mm and 200 mm simultaneously. Transient performance of the shock waves, magnitude of their peaks and durations of their fluctuation, is obtained. Some analyses for existing calculational method on water-hammer are made

Advancing PWR fuel to meet customer needs

Energy Technology Data Exchange (ETDEWEB)

Kramer, F W

1987-03-01

Since the introduction of the Optimized Fuel Assembly (OFA) for PWRs in the late 1970s, Westinghouse has continued to work with the utility customers to identify the greatest needs for further advance in fuel performance and reliability. The major customer requirements include longer fuel cycle at lower costs, increased fuel discharge burn-up, enhanced operating flexibility, all accompanied by even greater reliability. In response to these needs, Westinghouse developed Vantage 5 PWR fuel. To optimize reactor operations, Vantage 5 fuel features distinct advantages: integral fuel burnable absorbers, axial and radial blankets, intermediate flow mixers, a removable top nozzle, and assembly modifications to accommodate increased discharge burn-up.

Seismic qualification of PWR plant auxiliary feedwater systems

International Nuclear Information System (INIS)

Lu, S.C.; Tsai, N.C.

1983-08-01

The NRC Standard Review Plan specifies that the auxiliary feedwater (AFW) system of a pressurized water reactor (PWR) is a safeguard system that functions in the event of a Safe Shutdown Earthquake (SSE) to remove the decay heat via the steam generator. Only recently licensed PWR plants have an AFW system designed to the current Standard Review Plan specifications. The NRC devised the Multiplant Action Plan C-14 in order to make a survey of the seismic capability of the AFW systems of operating PWR plants. The purpose of this survey is to enable the NRC to make decisions regarding the need of requiring the licensees to upgrade the AFW systems to an SSE level of seismic capability. To implement the first phase of the C-14 plan, the NRC issued a Generic Letter (GL) 81-14 to all operating PWR licensees requesting information on the seismic capability of their AFW systems. This report summarizes Lawrence Livermore National Laboratory's efforts to assist the NRC in evaluating the status of seismic qualification of the AFW systems in 40 PWR plants, by reviewing the licensees' responses to GL 81-14

Analyses of plant behaviors at the secondary side depressurization during LOCA of PWR

Energy Technology Data Exchange (ETDEWEB)

Kawabe, Yasuharu; Tamaki, Tomohiko; Kohriyama, Tamio; Ohtani, Masanori [Institute of Nuclear Safety System Inc., Mihama, Fukui (Japan)

2001-09-01

When high pressure injection systems failed during a small break loss-of-coolant-accident (LOCA) for a PWR, main steam relief valves are opened to operate accumulator systems. However, it is pointed out that the core can be exposed since so-called counter current flow limitation (CCFL) occurs in steam generator (SG) tubes. The possibility of the core exposure by CCFL in a PWR plant was evaluated. First, RELAP5/MOD2 code was modified to be able to calculate CCFL. And then the code was applied to evaluate a 4-loop PWR plant. The LOCA with a rupture 3 inches were analyzed with the following two cases: (1) Only the main steam relief valve of the loop with the rupture is opened. (2) all of the relief valves are opened. It is seen that the CCFL phenomenon occurs in the case (1), however, the core cooling was maintained by the accumulator systems that actuated during the core exposure. On the other hand, the core exposure by CCFL is not observed in the case (2). It is shown that core cooling is promoted by operation of main steam relief valves. (author)

Water Chemistry Control in Reducing Corrosion and Radiation Exposure at PWR Reactor

International Nuclear Information System (INIS)

Febrianto

2006-01-01

Water chemistry control plays an important role in relation to plant availability, reliability and occupational radiation exposures. Radiation exposures of nuclear plant workers are determined by the radiation rate dose and by the amount maintenance and repair work time Water chemistry has always been, from beginning of operation of power Pressurized Water Reactor, an important factor in determining the integrity of reactor components, fuel cladding integrity and minimize out of core radiation exposures. For primary system, the parameters to control the quality of water chemistry have been subject to change in time. Reactor water coolant pH need to be optimally controlled and be operated in range pH 6.9 to 7.4. At pH lower than 6.9, cause increasing the radiation exposure level and increasing coolant water pH higher than 7.4 will decrease radiation exposure level but increasing risk to fuel cladding and steam generator tube. Since beginning 90 decade, PWR water coolant pH tend to be operated at pH 7.4. This paper will discuss concerning water chemistry development in reducing corrosion and radiation exposure dose in PWR reactor. (author)

Brief account of the effect of overcooling accidents on the integrity of PWR pressure vessels

International Nuclear Information System (INIS)

Cheverton, R.D.

1982-01-01

The occurrence in recent years of several (PWR) accident initiating events that could lead to severe thermal shock to the reactor pressure vessel, and the growing awareness that copper and nickel in the vessel material significantly enhance radiation damage in the vessel, have resulted in a reevaluation of pressure-vessel integrity during postulated overcooling accidents. Analyses indicate that the accidents of concern are those involving both thermal shock and pressure loadings, and that an accident similar to that at Rancho Seco in 1978 could, under some circumstances and at a time late in the normal life of the vessel, result in propagation of preexistent flaws in the vessel wall to the extent that they might completely penetrate the wall. More severe accidents have been postulated that would result in even shorter permissible lifetimes. However, the state-of-the-art fracture-mechanics analysis may contain excessive conservatism, and this possibility is being investigated. Furthermore, there are several remedial measures, such as fuel shuffling, to reduce the damage rate, and vessel annealing, to restore favorable material properties, that may be practical and used if necessary. 5 figures

GOVERNMENT SUPPORT FOR SMALL AND MEDIUM-SIZED BUSINESS AND INNOVATIVE ACTIVITIES

Directory of Open Access Journals (Sweden)

Pоlina Kolisnichenko

2017-09-01

Full Text Available The purpose of the paper is to reveal the conditions of the innovative development of the small and mediumsized entrepreneurship in Ukraine; the problems that suppress the innovative activity and small and medium-sized enterprises development; peculiarities of the tax incentives for the development of the entrepreneurship in the advanced countries and in Ukraine; the main kinds and characteristics of the small and medium-sized enterprises public support. Methodology. The methods of scientific research include: analysis and generalization for studying the main kinds and characteristics of the government support of the small and medium-sized enterprises; dynamic and comparative analysis for studying the problems and factors influencing the development of small and medium enterprises and innovative activity, peculiarities of tax incentives for business development; systemanalytical method for studying the conditions of the innovative development of the small and medium-sized entrepreneurship. Results. The government's financial support priorities should be: optimal application of the fiscal regulation instruments (reduction of the amount of taxes, determination of the criteria for the maximum taxation amount, tax incentives etc., maintenance of the self-investment of small and medium-sized enterprises as well as investment, financial means of the public influence over the development of the enterprises, effective combination of both direct and indirect forms of the innovative development support. Practical implications. The obtained results can be used in the process of formation and implementation of the small and medium enterprise sector development strategy and innovative activity in the long-term perspective. Value/originality. The obtained data can provide a better understanding of the direction of innovative business development in Ukraine.

Assessment and limitation of radioactivity transfers in the event of a postulated severe PWR accident

International Nuclear Information System (INIS)

Gauvain, J.

1992-01-01

This report constitutes the supporting material for a lecture on severe accidents which could occur on PWR type nuclear reactors. It is assumed for present purposes that the reader has at least a rudimentary acquaintance with the basics of general physics if not with the operating processes of these reactors. After defining what is meant by a ''severe accident'' on a reactor, the possible phenomenology of such an accident is qualitatively described: loss of coolant and loss of containment integrity. A certain number of elements are then given for the quantitative assessment of these phenomena involving possible radioactivity transfers within and outside the plant. In conclusion, available means are indicated for the limitation and control of these environmental transfers. (author). 5 refs, figs

Development of a small-sized radon data logger

International Nuclear Information System (INIS)

Tasaka, Shigeki; Sasaki, Yoshimi

1996-01-01

A small-sized radon data logger and a electrostatic collecting radon monitor were developed for the continuous monitoring of environmental radon and radon daughters. A PIN photodiode (PD), an alpha particle defector, installed inside a container attracts radon daughters when charged electrostatically. Alpha particles are completely separated from each other according to the energy level. New logger has made it possible 10 analysts and save the radon data. Alpha particle count data from radon daughters are automatically integrated over preset time intervals and the energy regions. The desiccant P 2 O 5 was placed in the bottom of the monitor, since the collection efficiency of 218 Po atoms depends on the humidity of the air. We can get the 30 days continuous data logging at 30 min sampling frequency at any place with the car battery. We observed the radon concentration of the air inside the Super-Kamiokande dome from Jan-30 to Feb-8-1996. The average of radon concentration was found to be (46±13) Bq/m 3 . (author)

Characterization of Factors affecting IASCC of PWR Core Internals

Energy Technology Data Exchange (ETDEWEB)

Kim, Sung Woo; Hwang, Seong Sik; Kim, Won Sam [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2008-09-15

A lot works have been performed on IASCC in BWR. Recent efforts have been devoted to investigate IASCC in PWR, but the mechanism in PWR is not fully understood yet as compared with that in BWR due to a lack of data from laboratories and fields. Therefore it is strongly needed to review and analyse recent researches of IASCC in both BWR and PWR for establishing a proactive management technology for IASCC of core internals in Korean PWRs. This work is aimed to review mainly recent technical reports on IASCC of stainless steels for core internals in PWR. For comparison, the works on IASCC in BWR were also reviewed and briefly introduced in this report.

Integrative structural modeling with small angle X-ray scattering profiles

Directory of Open Access Journals (Sweden)

Schneidman-Duhovny Dina

2012-07-01

Full Text Available Abstract Recent technological advances enabled high-throughput collection of Small Angle X-ray Scattering (SAXS profiles of biological macromolecules. Thus, computational methods for integrating SAXS profiles into structural modeling are needed more than ever. Here, we review specifically the use of SAXS profiles for the structural modeling of proteins, nucleic acids, and their complexes. First, the approaches for computing theoretical SAXS profiles from structures are presented. Second, computational methods for predicting protein structures, dynamics of proteins in solution, and assembly structures are covered. Third, we discuss the use of SAXS profiles in integrative structure modeling approaches that depend simultaneously on several data types.

Ciclon: A neutronic fuel management program for PWR's consecutive cycles

International Nuclear Information System (INIS)

Aragones, J.M.

1977-01-01

The program description and user's manual of a new computer code is given. Ciclon performs the neutronic calculation of consecutive reload cycles for PWR's fuel management optimization. Fuel characteristics and burnup data, region or batch sizes, loading schemes and state of previously irradiated fuel are input to the code. Cycle lengths or feed enrichments and burnup sharing for each region or batch are calculate using different core neutronic models and printed or punched in standard fuel management format. (author) [es

Design study of a PWR of 1300 MWe of Angra-2 type operating in the thorium cycle

International Nuclear Information System (INIS)

Andrade, E.P.; Carneiro, F.A.N.; Schlosser, J.G.

1984-01-01

The utilization of the thorium-highly enriched uranium and of the thorium-plutonium mixed oxide fuels in an unmodified PWR is analysed. Reactor core design calculations were performed for both types of fuels considering once-through and recycle fuels. The calculations were performed with the KWU design codes FASER-3 and MEDIUM-2.2 after introduction of the thorium chain and some addition of nuclide data in FASER-3. A two-energy group scheme and a two-dimensional (XY) representation of the reactor core were utilized. No technical problem that precluded the utilization of any of the options analyzed was found. The savings in uranium ore introduced by the thorium cycle with fuel recycling ranges from 13% to 52% as compared with the usual uranium once-through cycle; the SWU savings goes from 13% to 22%. (Author) [pt

78 FR 37397 - Small Business Size Standards: Agriculture, Forestry, Fishing and Hunting

Science.gov (United States)

2013-06-20

... responsibility for establishing small business size definitions (15 U.S.C. 632(a)). The Act also requires that small business size definitions vary to reflect industry differences. The Jobs Act requires the... definition, after consultation with the Office of Advocacy of the U.S. Small Business Administration (5 U.S.C...

GOVERNMENT SUPPORT FOR SMALL AND MEDIUM-SIZED BUSINESS AND INNOVATIVE ACTIVITIES

OpenAIRE

Pоlina Kolisnichenko

2017-01-01

The purpose of the paper is to reveal the conditions of the innovative development of the small and mediumsized entrepreneurship in Ukraine; the problems that suppress the innovative activity and small and medium-sized enterprises development; peculiarities of the tax incentives for the development of the entrepreneurship in the advanced countries and in Ukraine; the main kinds and characteristics of the small and medium-sized enterprises public support. Methodology. The methods of scientific...

Dynamical analysis of a PWR internals using super-elements in an integrated 3-D model model. Part 2: dynamical tests and seismic analysis

International Nuclear Information System (INIS)

Jesus Miranda, C.A. de.

1992-01-01

The results of the test analysis (frequencies) for the isolated super-elements and for the developed 3-D model of the internals core support structures of a PWR research reactor are presented. Once certified of the model effectiveness for this type of analysis the seismic spectral analysis was performed. From the results can be seen that the structures are rigid for this load, isolated or together with the other in the 3-D model, and there are no impacts among them during the earthquake (OBE). (author)

The empirical intensity of PWR primary coolant pumps failure and repair

International Nuclear Information System (INIS)

Milivojevicj, S.; Riznicj, J.

1988-01-01

The wealth of operating experience concerning PWR type and nuclear reactors that has been regularly monitored and systematically processes since 1971, enabled an analysis of the PWR primary coolant pumps operation. Failure intensity α and repair intensity μ of the pump during its working life were calculated, as these values are necessary in order to determine the reliability and availability of the pump as the basis for analyzing its effect on the safety and efficiency of the nuclear power plant. The trend of failure intensity α follows the theoretically expected changes in α over time, and this is around 10 -5 in the majority of life-time. Repair intensity μ indicates a slow rise during life-time, i.e. its faster return to operation. (author).7 refs.; 5 figs

Swelling in cold-worked 316 stainless steels irradiated in a PWR

Energy Technology Data Exchange (ETDEWEB)

Fukuya, Koji; Fujii, Katsuhiko [Institute of Nuclear Safety System Inc., Mihama, Fukui (Japan)

2001-09-01

Swelling behavior in a cold-worked 316 stainless steel irradiated up to 53 dpa in a PWR at 290-320degC was examined using high resolution transmission electron microscopy. Small cavities with the average diameter of 1 nm were observed in the samples irradiated to doses above 3 dpa. The average diameter did not increase with increasing in dose. The maximum swelling was as low as 0.042%. The measured helium content and the cavity morphology led to the conclusion that the cavities were helium bubbles. A comparison of the observed cavity microstructure with data from FBR, HFIR and ATR irradiation showed that the cavity structure in PWR at 320degC or less was similar to those in HFIR and ATR irradiation but quite different from those in FBR condition. From a calculation based on the cavity data and kinetic models the incubation dose of swelling was estimated to be higher than 80dpa in the present irradiation condition. (author)

Sizewell 'B' PWR reference design

International Nuclear Information System (INIS)

1982-04-01

The reference design for a PWR power station to be constructed as Sizewell 'B' is presented in 3 volumes containing 14 chapters and in a volume of drawings. The report describes the proposed design and provides the basis upon which the safety case and the Pre-Construction Safety Report have been prepared. The station is based on a 3425MWt Westinghouse PWR providing steam to two turbine generators each of 600 MW. The layout and many of the systems are based on the SNUPPS design for Callaway which has been chosen as the US reference plant for the project. (U.K.)

Probable variations of a passive safety containment for a 1700 MWe class PWR with passive safety systems

International Nuclear Information System (INIS)

Sato, Takashi; Fujiki, Yasunobu; Oikawa, Hirohide; Ofstun, Richard P.

2009-01-01

The paper presents probable variations of a passive safety containment for a PWR. The passive safety containment is named Mark P containment tentatively. It is a pressure suppression type containment for a large scale PWR with a BWR type passive containment cooling system (PCCS). More than 3-day grace period can be achieved even for a 1700 MWe class large scale PWR owing to the PCCS. The containment is a reinforced concrete containment vessel (RCCV). The design pressure of the RCCV can be low owing to the suppression pool (S/P) and no prestressed tendon is necessary. It is a single barrier CV that can withstand a large airplane crash by itself. This simple configuration results in good economy and short construction term. The BWR type passive safety systems also include the Passive Cooling and Depressurization System (PCDS). The PCDS has 3-day grace period for the SBO induced by a giant earthquake and can practically eliminate the residual risk of a giant earthquake beyond the design basis earthquake of Ss. It also has a safety function to automatically depressurize the primary system at accidents such as SGTR and eliminate the need for operator actions. It is a large 1700 MWe passive safety PWR that has more than 3-day grace period for extremely severe natural disasters including a giant earthquake, a mega hurricane, tsunami and so on; no containment failure at a SA establishing a no evacuation plant; protection for a large airplane crash with the RCCV single barrier; good economy and short construction term. (author)

Low-density moderation in the storage of PWR fuel assemblies

International Nuclear Information System (INIS)

Alcorn, F.M.

1987-01-01

The nuclear criticality safety of PWR fuel storage arrays requires that the potential of low-density moderation within the array be considered. The calculated criticality effect of low-density moderation in a typical PWR fuel assembly array is described in this paper. Calculated reactivity due to low-density moderation can vary significantly between physics codes that have been validated for well moderated systems. The availability of appropriate benchmark experiments for low-density moderation is quite limited; attempts to validate against the one set of suitable experiments at low density have been disappointing. Calculations indicate that a typical array may be unacceptable should the array be subjected to interstitial moderation equivalent to 5 % of full density water. Array parameters (such as spacing and size) will dramatically affect the calculated maximum K-eff at low-density moderation. Administrative and engineered control may be necessary to assure maintenance of safety at low-density moderation. Potential sources for low-density moderation are discussed; in general, accidentally achieving degrees of low-density moderation which might lead to a compromise of safety are not credible. (author)

Robust Color Choice for Small-size League RoboCup Competition

Directory of Open Access Journals (Sweden)

Qiang Zhou

2004-10-01

Full Text Available In this paper, the problem of choosing a set of most separable colors in a given environment is discussed. The proposed method models the process of generating theoretically computed best colors, printing of these colors through a color printer, and imaging the printed colors through a camera into an integrated framework. Thus, it provides a feasible way to generate practically best separable colors for a given environment with a set of given equipment. A real world application (robust color choice for small-size league RoboCup competition is used as an example to illustrate the proposed method. Experimental results on this example show the competitiveness of the colors learned from our algorithm compared to the colors adopted by other teams which are chosen via an extensive trial and error process using standard color papers.

[Study on the types and water pollution driving forces of the typical and medium-small-sized cities in the southern China based on the analysis of water environment].

Science.gov (United States)

Jiao, Shi-Xing; Wang, La-Chun; Huo, Yu; Chen, Chang-Chun; Teng, Juan

2009-07-15

According to the major pollution sources of urban water environment, 10 indexes such as industrial sewage quantity were closen to establish evaluation indexes system about the types and influencing factors of the typical and medium-small-sized cities in the southern China. Case studies of 16 typical and medium-small-sized cities were taken in Jiangsu, Zhejiang, Hubei and Anhui provinces. Combined with SPSS 11.0 cluster analysis results, city types were divided in reference to the values of water resources comprehensive pollution indexes and economical development indexes. The driving forces about city water environment pollution were studied by principal component analysis method. The result indicates that the 16 cities belong to two categories and four sub-categories, which are rich economy as well as light pollution of water environment and poor economy as well as heavy pollution of water environment. The influencing factors of water environment pollution are in sequence of industrial water pollution, agricultural no-point source pollution and urban domestic water pollution. The main factors of water environment pollution influenced I category cities, II as well as IV category cities and III category cities are industrial water pollution, urban domestic pollution and agricultural no-point source pollution respectively.

Numerical simulation of thermohydraulic behavior of the steam generator of PWR type reactor

International Nuclear Information System (INIS)

Braga, C.V.M.; Carajilescov, P.

1981-01-01

Generally, 'U' tube steam generators with natural internal recirculation are used in PWR power stations. A thermalhydraulic model is developed for simulation of such components, in steady state. The flow of the secondary cycle fluid is divided in two parts individually homogeneous, allowing for heat and mass exchange between them. The secondary pressure is determined by defining the moisture of the vapor that feeds the turbine. This model is applied to the Angra II steam generator, operating in nominal conditions and with tubing partially plugged. (Author) [pt

17 CFR 232.308 - Type size and font; legibility.

Science.gov (United States)

2010-04-01

... 17 Commodity and Securities Exchanges 2 2010-04-01 2010-04-01 false Type size and font; legibility. 232.308 Section 232.308 Commodity and Securities Exchanges SECURITIES AND EXCHANGE COMMISSION... § 232.308 Type size and font; legibility. Provisions relating to type size, font and other legibility...

MELCOR/VISOR PWR desktop simulator

International Nuclear Information System (INIS)

With, Anka de; Wakker, Pieter

2010-01-01

Increasingly, there is a need for a learning support and training tool for nuclear engineers, utilities and students in order to broaden their understanding of advanced nuclear plant characteristics, dynamics, transients and safety features. Nuclear system analysis codes like ASTEC, RELAP5, RETRAN and MELCOR provide calculation results of and visualization tools can be used to graphically represent these results. However, for an efficient education and training a more interactive tool such as a simulator is needed. The simulator connects the graphical tool with the calculation tool in an interactive manner. A small number of desktop simulators exist [1-3]. The existing simulators are capable of representing different types of power plants and various accident conditions. However, they were found to be too general to be used as a reliable plant-specific accident analysis or training tool. A desktop simulator of the Pressurized Water Reactor (PWR) has been created under contract of the Dutch nuclear regulatory body (KFD). The desktop simulator is a software package that provides a close to real simulation of the Dutch nuclear power plant Borssele (KCB) and is used for training of the accident response. The simulator includes the majority of the power plant systems, necessary for the successful simulation of the KCB plant during normal operation, malfunctions and accident situations, and it has been successfully validated against the results of the safety evaluations from the KCB safety report. (orig.)

Study on virtual simulation technology for operation and control of PWR

International Nuclear Information System (INIS)

Fang Baoguo; Zhang Dafa; Lin Yajun

2006-01-01

The way to build graphical models of PWR with MultiGen Creator is discussed, and the three-dimensional model used in the virtual simulation is built. The mathematical simulation model for PWR based on the platform of MFC and Vega is built through the analysis of the mathematical simulation of PWR. The way to perform the virtual effect is introduced associating with the Pressurizer. And, all above parts are connected in one with VC++ to perform the whole virtual simulation of PWR. (authors)

Safety considerations of PWR's

International Nuclear Information System (INIS)

Arnold, W.H. Jr.

1977-01-01

The safety of the central station pressurized water reactor is well established and substantiated by its excellent operating record. Operating data from 55 reactors of this type have established a record of safe operating history unparalleled by any modern large scale industry. The 186 plants under construction require a continuing commitment to maintain this outstanding record. The safety of the PWR has been further verified by the recently completed Reactor Safety Study (''Rasmussen'' Report). Not only has this study confirmed the exceptionally low risk associated with PWR operation, it has also introduced a valuable new tool in the decision making process. PWR designs, utilizing the philosophy of defense in depth, provide the bases for evaluating margins of safety. The design of the reactor coolant system, the containment system, emergency core cooling system and other related systems and components provide substantial margins of safety under both normal and postulated accident conditions even considering simultaneous effects of earthquakes and other environmental phenomena. Margins of safety in the assessment of various postulated accident conditions, with emphasis on the postulated loss of reactor coolant accident (LOCA), have been evaluated in depth as exemplified by the comprehensive ECCS rulemaking hearings followed by imposition of very conservative Nuclear Regulatory Commission requirements. When evaluated on an engineering best estimate approach, the significant margins to safety for a LOCA become more apparent. Extensive test programs have also substantiated margins to safety limits. These programs have included both separate effects and systems tests. Component testing has also been performed to substantiate performance levels under adverse combinations of environmental stress. The importance of utilizing past experience and of optimizing the deployment of incremental resources is self evident. Recent safety concerns have included specific areas such

Development of the vacuum drying process for the PWR spent nuclear fuel dry storage

Energy Technology Data Exchange (ETDEWEB)

Baeg, Chagn Yeal; Cho, Chun Hyung [Korea Radioactive Waste Agency, Daejeon (Korea, Republic of)

2016-12-15

This paper describes the development of a dry operation process for PWR spent nuclear fuel, which is currently stored in the domestic NPP's storage pool, using a dual purpose metal cask. Domestic NNPs have had experience with wet type transportation of PWR spent nuclear fuel between neighboring NPPs since the early 1990s, but no experience with dry type operation. For this reason, we developed a specific operation process and also confirmed the safety of the major cask components and its spent nuclear fuel during the dual purpose metal cask operation process. We also describe the short term operation process that was established to be completed within 21 hours and propose the allowable working time for each step (15 hours for wet process, 3 hours for drain process and 3 hours for vacuum drying process)

Remote Monitoring and Instrumentation Strategies for Integral Reactors

International Nuclear Information System (INIS)

Upadhyaya, Belle R.; Lish, Matthew R.; Tarver, Rayan A.; Hines, J. Wesley

2014-01-01

The University of Tennessee is engaged in research and development projects related to instrumentation and controls for small modular reactors (SMR) and integral pressurized water reactors (iPWR). The approach incorporates the deployment of physics-based models for control design and parameter estimation, development of noncontact sensors for flow monitoring, and placement of sensors to maximize fault detection and isolation. The results of research and development illustrate the feasibility of sensor location in space-constrained environment. Major issues and challenges in I and C design are addressed

Remote Monitoring and Instrumentation Strategies for Integral Reactors

Energy Technology Data Exchange (ETDEWEB)

Upadhyaya, Belle R.; Lish, Matthew R.; Tarver, Rayan A.; Hines, J. Wesley [University of Tennessee, Knoxville (United States)

2014-08-15

The University of Tennessee is engaged in research and development projects related to instrumentation and controls for small modular reactors (SMR) and integral pressurized water reactors (iPWR). The approach incorporates the deployment of physics-based models for control design and parameter estimation, development of noncontact sensors for flow monitoring, and placement of sensors to maximize fault detection and isolation. The results of research and development illustrate the feasibility of sensor location in space-constrained environment. Major issues and challenges in I and C design are addressed.

NODAL3 Sensitivity Analysis for NEACRP 3D LWR Core Transient Benchmark (PWR

Directory of Open Access Journals (Sweden)

Surian Pinem

2016-01-01

Full Text Available This paper reports the results of sensitivity analysis of the multidimension, multigroup neutron diffusion NODAL3 code for the NEACRP 3D LWR core transient benchmarks (PWR. The code input parameters covered in the sensitivity analysis are the radial and axial node sizes (the number of radial node per fuel assembly and the number of axial layers, heat conduction node size in the fuel pellet and cladding, and the maximum time step. The output parameters considered in this analysis followed the above-mentioned core transient benchmarks, that is, power peak, time of power peak, power, averaged Doppler temperature, maximum fuel centerline temperature, and coolant outlet temperature at the end of simulation (5 s. The sensitivity analysis results showed that the radial node size and maximum time step give a significant effect on the transient parameters, especially the time of power peak, for the HZP and HFP conditions. The number of ring divisions for fuel pellet and cladding gives negligible effect on the transient solutions. For productive work of the PWR transient analysis, based on the present sensitivity analysis results, we recommend NODAL3 users to use 2×2 radial nodes per assembly, 1×18 axial layers per assembly, the maximum time step of 10 ms, and 9 and 1 ring divisions for fuel pellet and cladding, respectively.

Temporary core liquid level depression during cold-leg small-break LOCA effect of break size and power level

International Nuclear Information System (INIS)

Koizumi, Y.; Kumamaru, H.; Mimura, Y.; Kukita, Y.; Tasaka, K.

1989-01-01

Cold-leg small break LOCA experiments (0.5-10% break) were conducted at the large scale test facility (LSTF), a volumetrically-scaled (1/48) simulator of a PWR, of the ROSA-IV Program. When a break area was less than 2.5% of the scaled cold-leg flow area, the core liquid level was temporarily further depressed to the bottom elevation of the crossover leg during the loop seal clearing early in the transient only by the manometric pressure balance since no coolant remained in the upper portion of the primary system. When the break size was larger than 5%, the core liquid level was temporarily further depressed lower than the bottom elevation of the crossover leg during the loop seal clearing since coolant remained at the upper portion of the primary system; the steam generator (SG) U-tube upflow side and the SG inlet plenum, due to counter current flow limiting by updrafting steam while the coolant drained. The amount of coolant trapped there was dependent on the vapor velocity (core power); the larger the core power, the lower the minimum core liquid level. The RELAP5/MOD2 code reasonable predicted phenomena observed in the experiments. (orig./DG)

SCAR - Post-Accident Simulator SIPA with safety analysis code CATHARE-2 and PWR cold shutdown state simulation

International Nuclear Information System (INIS)

Farvacque, M.; Faydide, B.; Dufeil, Ph.; Raimond, E.

2003-01-01

The use of Cathare in the simulators of pressurized water reactors has been effective since the beginning of the nineties. Scar project is the second stage of the Cathare strategy for the simulators, its main objective is the extension of the field of simulation to the accident situations in cold shutdown states. Work was carried out in 3 major areas: modelling, optimization and integration in the simulator. Throughout the project, the developments were part of a 3 stages validation strategy: -) elementary tests of the developments of new model on the N4 (1450 MW PWR); -) analytical tests and systems to ensure non regression of the validation of the physical laws of the Cathare code during the modifications carried out within the optimization stage; and -) overall tests of the SIPA-CP1 (900 MW PWR) simulator, controlled automatically by programmed scenarios including the transients which are carried out in PWR, the transients of the Regulatory Guides and the accident transients

Water chemistry in PWR

International Nuclear Information System (INIS)

Abe, Kenji

1987-01-01

This article outlines major features and basic concept of the secondary system of PWR's and water properties control measures adopted in recent PWR plants. The secondary system of a PWR consists of a condenser cooling pipe (aluminum-brass, titanium, or stainless steel), low-pressure make-up water heating pipe (aluminum-brass or stainless steel), high-ressure make-up water heating pipe (cupro-nickel or stainless steel), steam generator heat-transfer pipe (Inconel 600 or 690), and bleed/drain pipe (carbon steel, low alloy steel or stainless steel). Other major pipes and equipment are made of carbon steel or stainless steel. Major troubles likely to be caused by water in the secondary system include reduction in wall thickness of the heat-transfer pipe, stress corrosion cracking in the heat-transfer pipe, and denting. All of these are caused by local corrosion due to concentration of purities contained in water. For controlling the water properties in the secondary system, it is necessary to prevent impurities from entering the system, to remove impurities and corrosion products from the system, and to prevent corrosion of apparatus making up the system. Measures widely adopted for controlling the formation of IGA include the addition of boric acid for decreasing the concentration of free alkali and high hydrazine operation for providing a highly reducing atmospere. (Nogami, K.)

Management of small and medium size enterprises as a carrier of economic growth

Directory of Open Access Journals (Sweden)

Ožegović Lazar

2012-01-01

Full Text Available Numerous studies in the world of market economy show that the share of small and medium size enterprises is constantly increasing compared to large enterprises. This does not decrease the significance of large enterprises, but the dependence between them gets larger every day. National economies which manage to find the optimal combination between small, medium size and large enterprises are more successful than the others. Management of small and medium size enterprises (SMEs in conceptual terms is similar to the process of management of large enterprises. An organization must be managed in order to function and this rule applies to small and medium size enterprises.

Measurement of grid spacer's enhanced droplet cooling under reflood condition in a PWR by LDA

International Nuclear Information System (INIS)

Lee, S.L.; Sheen, H.J.; Cho, S.K.; Issapour, I.; Hua, S.Q.

1984-01-01

Reported is an experiment designed for the measurements of grid spacer's enhanced droplet cooling under reflood condition at elevated temperatures in a steam environment. The flow channel consists of a simulated 1.60m-long pressurized water reactor (PWR) fuel rod bundle of 2 x 2 electrically heated rods. Embedded thermocouples are used to measure the rod cladding temperature at various axial levels and an unshielded Chromel-Alumel thermocouple sheathed by a small Inconel tube is traversed in the center of the subchannel to measure the temperatures of the water and steam coolant phases at various levels. The droplet dynamics across the grid spacer is directly obtained by a special laser-Doppler anemometry technique for the in situ simultaneous measurement of velocity and size of droplets through two observation windows on the test channel, one immediately before and one immediately after the grid spacer. Some results are presented and analyzed

PWR surveillance based on correspondence between empirical models and physical

International Nuclear Information System (INIS)

Zwingelstein, G.; Upadhyaya, B.R.; Kerlin, T.W.

1976-01-01

An on line surveillance method based on the correspondence between empirical models and physicals models is proposed for pressurized water reactors. Two types of empirical models are considered as well as the mathematical models defining the correspondence between the physical and empirical parameters. The efficiency of this method is illustrated for the surveillance of the Doppler coefficient for Oconee I (an 886 MWe PWR) [fr

Experimental results of the effective water head in downcomer during reflood phase of a PWR LOCA

International Nuclear Information System (INIS)

Sudo, Yukio; Murao, Yoshio; Akimoto, Hajime

1980-08-01

The results and analysis of an experiment for the effective water head in downcomer with 50mm gap size are described. The main objective of the experiment was to clarify the effect of gap size on reflooding in a PWR LOCA. The effective water head in downcomer is the driving force for feeding emergency coolant into the core during reflood phase of a PWR LOCA. Discussions presented here follow those of a previous report in which experimental results and analysis were described for the case of 200mm gap size. Experimental Conditions were: Initial Wall Temperature = 200 -- 300 0 C, Back Pressure = 1 atm., Coolant Temperature = 71 -- 100 0 C, Extraction Water Velocity = 0 -- 2 cm/s, Gap Size = 50 mm. The effective water head history obtained in the experiment was compared with those predicted with Sudo's void fraction correlation. In the prediction, heat input to coolant was calculated from the response of measured wall temperature with heat condition analysis. The experimental results and analysis reveals that: (1) The effects of the gap size and initial wall temperature are evident, (2) The effect of extraction water velocity is negligible, and (3) The predicted history of effective water head is in good agreement with the experimental results except during the transient period in which the effective water head is descreasing. (author)

The adoption of e-commerce in small and medium-sized enterprises in Vietnam : recommendations for building an e-commerce strategy

OpenAIRE

Ho, Huong

2013-01-01

It is highly recommended for small and medium-sized enterprises (SMEs) to integrate e-commerce into their business operations because it is an opportunity to increase profitability, enhance effectiveness, and increase corporate values. However, the number of Vietnamese SMEs adopting e-commerce is insignificant due to lack of understanding of the process of integrating it into business operations. This research is conducted for the purpose of providing SMEs attempting to move towards the e-com...

ABB advanced BWR and PWR fuel

International Nuclear Information System (INIS)

Junkrans, S.; Helmersson, S.; Andersson, S.

1999-01-01

Fuel designed and fabricated by ABB is now operating in 40 PWRs and BWRs in Europe, the United States and Korea. An excellent fuel reliability track record has been established. High burnups are proven for both BWR and PWR. Thermal margin improving features and advanced burnable absorber concepts enable the utilities to adopt demanding duty cycles to meet new economic objectives. In particular we note the excellent reliability record of ABB PWR fuel equipped with Guardian TM debris filter, proven to meet the -6 rod-cycles fuel failure goal, and the out-standing operating record of the SVEA 10x10 BWR fuel, where ABB is the only vendor to date with multi batch experience to high burnup. ABB is dedicated to maintain high fuel reliability as well as continually improve and develop a broad line of BWR and PWR products. ABB's development and fuel follow-up activities are performed in close co-operation with its customers. (orig.)

A Feasibility Study on Core Cooling of Reduced-Moderation PWR for the Large Break LOCA

International Nuclear Information System (INIS)

Hiroyuki Yoshida; Akira Ohnuki; Hajime Akimoto

2002-01-01

A design study of a reduced-moderation water reactor (RMWR) with tight lattice core is being carried out at the Japan Atomic Energy Research Institute (JAERI) as one candidate for future reactors. The concept is developed to achieve a conversion ratio greater than unity using the tight lattice core (volume ratio of moderator to fuel is around 0.5 and the gap spacing between the fuel rods is remarkably narrower than in a reactor currently operated). Under such tight configuration, the core thermal margin becomes smaller and should be evaluated in a normal operation and also during the reflood phase in a large break loss-of-coolant accident (LBLOCA) for PWR type reactors. In this study, we have performed a feasibility evaluation on core cooling of reduced moderation PWR for the LBLOCA (200% break). The evaluation was performed for the primary system after the break by the REFLA/TRAC code. The core thermal output of the reduced moderation PWR is 2900 MWt, the gap between adjacent fuel rods is 1 mm, and heavy water is used as the moderator and coolant. The present design adopts seed fuel assemblies (MOX fuel) and several blanket fuel assemblies. In the blanket fuel assemblies, power density is lower than that of the seed fuel assemblies. Then, we set a channel box to each fuel assembly in order to adjust the flow rate in each assembly, because the possibility that the coolant boils in the seed fuel assemblies is very high. The pressure vessel diameter is bigger in comparison with a current PWR and core height is smaller than the current one. The current 4-loop PWR system is used, and, however, to fit into the bigger pressure vessel volume (about 1.5 times), we set up the capacity of the accumulator (1.5 times of the current PWR). Although the maximum clad temperature reached at about 1200 K in the position of 0.6 m from the lower core support plate, it is sufficiently lower than the design criteria of the current PWR (1500 K). The core cooling of the reduced moderation

Parallel GPU implementation of PWR reactor burnup

International Nuclear Information System (INIS)

Heimlich, A.; Silva, F.C.; Martinez, A.S.

2016-01-01

Highlights: • Three GPU algorithms used to evaluate the burn-up in a PWR reactor. • Exhibit speed improvement exceeding 200 times over the sequential. • The C++ container is expansible to accept new nuclides chains. - Abstract: This paper surveys three methods, implemented for multi-core CPU and graphic processor unit (GPU), to evaluate the fuel burn-up in a pressurized light water nuclear reactor (PWR) using the solutions of a large system of coupled ordinary differential equations. The reactor physics simulation of a PWR reactor spends a long execution time with burnup calculations, so performance improvement using GPU can imply in better core design and thus extended fuel life cycle. The results of this study exhibit speed improvement exceeding 200 times over the sequential solver, within 1% accuracy.

Interface tracking simulations of bubbly flows in PWR relevant geometries

Energy Technology Data Exchange (ETDEWEB)

Fang, Jun, E-mail: jfang3@ncsu.edu [Department of Nuclear Engineering, North Carolina State University, Raleigh, NC 27695 (United States); Rasquin, Michel, E-mail: michel.rasquin@colorado.edu [Aerospace Engineering Department, University of Colorado, Boulder, CO 80309 (United States); Bolotnov, Igor A., E-mail: igor_bolotnov@ncsu.edu [Department of Nuclear Engineering, North Carolina State University, Raleigh, NC 27695 (United States)

2017-02-15

Highlights: • Simulations were performed for turbulent bubbly flows in PWR subchannel geometry. • Liquid turbulence is fully resolved by direct numerical simulation approach. • Bubble behavior is captured using level-set interface tracking method. • Time-averaged single- and two-phase turbulent flow statistical quantities are obtained. - Abstract: The advances in high performance computing (HPC) have allowed direct numerical simulation (DNS) approach coupled with interface tracking methods (ITM) to perform high fidelity simulations of turbulent bubbly flows in various complex geometries. In this work, we have chosen the geometry of the pressurized water reactor (PWR) core subchannel to perform a set of interface tracking simulations (ITS) with fully resolved liquid turbulence. The presented research utilizes a massively parallel finite-element based code, PHASTA, for the subchannel geometry simulations of bubbly flow turbulence. The main objective for this research is to demonstrate the ITS capabilities in gaining new insight into bubble/turbulence interactions and assisting the development of improved closure laws for multiphase computational fluid dynamics (M-CFD). Both single- and two-phase turbulent flows were studied within a single PWR subchannel. The analysis of numerical results includes the mean gas and liquid velocity profiles, void fraction distribution and turbulent kinetic energy profiles. Two sets of flow rates and bubble sizes were used in the simulations. The chosen flow rates corresponded to the Reynolds numbers of 29,079 and 80,775 based on channel hydraulic diameter (D{sub h}) and mean velocity. The finite element unstructured grids utilized for these simulations include 53.8 million and 1.11 billion elements, respectively. This has allowed to fully resolve all the turbulence scales and the deformable interfaces of individual bubbles. For the two-phase flow simulations, a 1% bubble volume fraction was used which resulted in 17 bubbles in

Topology and cellular localization of the small hydrophobic protein of avian metapneumovirus

Science.gov (United States)

The small hydrophobic protein (SH) is a type II integral membrane protein that is packaged into virions and is only present in certain paramyxoviruses including metapneumovirus. In addition to a highly divergent primary sequence, SH proteins vary significantly in size among the different viruses. Hu...

Experimental study of the solubilization of various elements likely to be emitted following a serious accident on a pressurized water reactor (P.W.R.)

International Nuclear Information System (INIS)

Monfort, M.; Picat, P.; Cartier, Y.

1988-09-01

The solubility of various constituents (Ru0 2 , Ce0 2 , Ag, In 2 0 3 , Fe 2 0 3 ) of aerosols released into the environment following a serious accident at a PWR have been studied using four types of natural waters (rain water and soil solutions). Very small quantities of each of the products studied pass into solution. The soluble fraction of Ru0 2 , composed of microparticles of oxide, appears to be more mobile in soils than that of Ag, consisting in part of Ag + ions [fr

Integral reactor vessel related to power reactor safety

International Nuclear Information System (INIS)

Widart, J.; Scailteur, A.

1978-01-01

Integral design applied to PWR pressure vessels allows to reach a high level of safety because: 1) it presents a better balance of the material in the geometry, resulting in an improved stress level (mainly faulted condition loadings); 2) location and geometry of the welds are designed in order to get a very sound pressure boundary of the upper part of the vessel; 3) the new location and geometry of the welds allow an easy ISI in such a way that ambiguity surrounding defect size or locaton is practically suppressed. (author)

Towards a PSA harmonization French-Belgian comparison of the level 1 PSA for two similar PWR types

International Nuclear Information System (INIS)

Dupuy, P.; Corenwinder, F.; Lanore, J.M.; Gryffroy, D.; Gelder, P. de; Hulsmans, M.

2002-06-01

In the framework of the cooperation between French and Belgian regulatory authorities, a PSA (Probabilistic Safety Assessment) comparison exercise has been carried out for several years. This comparison deals with two PSA level 1 studies for internal events, performed for both power and shutdown states: the French PSA of the 900 MWe-series PWR, and the Belgian PSA of the Tihange 1 PWR, which both concern PWRs with a similar Framatome design. The purpose of this paper is to describe the PSA comparison methodology and to present, in a qualitative way, an overview of the insights obtained up to now. It also shows that such an 'a posteriori' benchmark exercise turns out to be a step towards PSA harmonization, and gives more confidence in the results of plant specific PSA when used for applications like precursor analysis or evaluations of importance to safety. (authors)

PWR and BWR light water reactor systems in the USA and their fuel cycle

International Nuclear Information System (INIS)

Crawford, W.D.

1977-01-01

Light water reactor operating experience in the USA can be considered to date from the choice of the pressurized water reactor (PWR) for use in the naval reactor program and the subsequent construction and operation of the nuclear power plant at Shippingport, Pennsylvania in 1957. The development of the boiling water reactor (BWR) in 1954 and its selection for the plant at Dresden, Illinois in 1959 established this concept as the other major reactor type in the US nuclear power program. The subsequent growth profile is presented, leading to 31 PWR's and 23 BWR's currently in operation as well as to plants in the planning and construction phase. A significant operating record has been accumulated concerning the availability of each of these reactor types as determined by: (1) outage for refueling, (2) component reliability, (3) maintenance requirements, and (4) retrofitting required by government regulation. In addition, the use and performance of BWR's and PWR's in meeting system load requirements is discussed. The growing concern regarding possible terrorist activities and other potential threats has resulted in systems and procedures designed to assure effective safeguards at nuclear power installations. Safeguards measures currently in place are described. Environmental effects of operating plants are subject to both radiological and non-radiological monitoring to verify that results are within the limits established in the licensing process. The operating results achieved and the types of modifications that have been required of operating plants by the Nuclear Regulatory Commission are reviewed. The PWR and BWR Fuel Cycle is examined in terms of: (1) fuel burnup experience and prospects for improvement, (2) the status and outlook for natural uranium resources, (3) enrichment capacity, (4) reprocessing and recycle, and the interrelationships among the latter three factors. High level waste management currently involving on-site storage of spent fuel is discussed

Visual exposure to large and small portion sizes and perceptions of portion size normality: Three experimental studies.

Science.gov (United States)

Robinson, Eric; Oldham, Melissa; Cuckson, Imogen; Brunstrom, Jeffrey M; Rogers, Peter J; Hardman, Charlotte A

2016-03-01

Portion sizes of many foods have increased in recent times. In three studies we examined the effect that repeated visual exposure to larger versus smaller food portion sizes has on perceptions of what constitutes a normal-sized food portion and measures of portion size selection. In studies 1 and 2 participants were visually exposed to images of large or small portions of spaghetti bolognese, before making evaluations about an image of an intermediate sized portion of the same food. In study 3 participants were exposed to images of large or small portions of a snack food before selecting a portion size of snack food to consume. Across the three studies, visual exposure to larger as opposed to smaller portion sizes resulted in participants considering a normal portion of food to be larger than a reference intermediate sized portion. In studies 1 and 2 visual exposure to larger portion sizes also increased the size of self-reported ideal meal size. In study 3 visual exposure to larger portion sizes of a snack food did not affect how much of that food participants subsequently served themselves and ate. Visual exposure to larger portion sizes may adjust visual perceptions of what constitutes a 'normal' sized portion. However, we did not find evidence that visual exposure to larger portions altered snack food intake. Copyright © 2015 The Authors. Published by Elsevier Ltd.. All rights reserved.

Steam generator tube integrity program: Phase II, Final report

Energy Technology Data Exchange (ETDEWEB)

Kurtz, R.J.; Bickford, R.L.; Clark, R.A.; Morris, C.J.; Simonen, F.A.; Wheeler, K.R.

1988-08-01

The Steam Generator Tube Integrity Program (SGTIP) was a three phase program conducted for the US Nuclear Regulatory Commission (NRC) by Pacific Northwest Laboratory (PNL). The first phase involved burst and collapse testing of typical steam generator tubing with machined defects. The second phase of the SGTIP continued the integrity testing work of Phase I, but tube specimens were degraded by chemical means rather than machining methods. The third phase of the program used a removed-from-service steam generator as a test bed for investigating the reliability and effectiveness of in-service nondestructive eddy-current inspection methods and as a source of service degraded tubes for validating the Phase I and Phase II data on tube integrity. This report describes the results of Phase II of the SGTIP. The object of this effort included burst and collapse testing of chemically defected pressurized water reactor (PWR) steam generator tubing to validate empirical equations of remaining tube integrity developed during Phase I. Three types of defect geometries were investigated: stress corrosion cracking (SCC), uniform thinning and elliptical wastage. In addition, a review of the publicly available leak rate data for steam generator tubes with axial and circumferential SCC and a comparison with an analytical leak rate model is presented. Lastly, nondestructive eddy-current (EC) measurements to determine accuracy of defect depth sizing using conventional and alternate standards is described. To supplement the laboratory EC data and obtain an estimate of EC capability to detect and size SCC, a mini-round robin test utilizing several firms that routinely perform in-service inspections was conducted.

Steam generator tube integrity program: Phase II, Final report

International Nuclear Information System (INIS)

Kurtz, R.J.; Bickford, R.L.; Clark, R.A.; Morris, C.J.; Simonen, F.A.; Wheeler, K.R.

1988-08-01

The Steam Generator Tube Integrity Program (SGTIP) was a three phase program conducted for the US Nuclear Regulatory Commission (NRC) by Pacific Northwest Laboratory (PNL). The first phase involved burst and collapse testing of typical steam generator tubing with machined defects. The second phase of the SGTIP continued the integrity testing work of Phase I, but tube specimens were degraded by chemical means rather than machining methods. The third phase of the program used a removed-from-service steam generator as a test bed for investigating the reliability and effectiveness of in-service nondestructive eddy-current inspection methods and as a source of service degraded tubes for validating the Phase I and Phase II data on tube integrity. This report describes the results of Phase II of the SGTIP. The object of this effort included burst and collapse testing of chemically defected pressurized water reactor (PWR) steam generator tubing to validate empirical equations of remaining tube integrity developed during Phase I. Three types of defect geometries were investigated: stress corrosion cracking (SCC), uniform thinning and elliptical wastage. In addition, a review of the publicly available leak rate data for steam generator tubes with axial and circumferential SCC and a comparison with an analytical leak rate model is presented. Lastly, nondestructive eddy-current (EC) measurements to determine accuracy of defect depth sizing using conventional and alternate standards is described. To supplement the laboratory EC data and obtain an estimate of EC capability to detect and size SCC, a mini-round robin test utilizing several firms that routinely perform in-service inspections was conducted

PWR Users Group 10 CFR 61 Waste Form Requirements Compliance Test Program

International Nuclear Information System (INIS)

Rosenlof, R.C.

1985-01-01

In January of 1984, a PWR Users Group was formed to initiate a 10 CFR 61 Waste Form Requirements Compliance Test Program on a shared cost basis. The original Radwaste Solidification Systems sold by ATCOR ENGINEERED SYSTEMS, INC. to the utilities were required to produce a free-standing monolith with no free water. None of the other requirements of 10 CFR 61 had to be met. Current regulations, however, have substantially expanded the scope of the waste form acceptance criteria. These new criteria required that generators of radioactive waste demonstrate the ability to produce waste forms which meet certain chemical and physical requirements. This paper will present the test program used and the results obtained to insure 10 CFR 61 compliance of the three (3) typical waste streams generated by the ATCOR PWR Users Group's plants. The primary objective of the PWR Users Group was not to maximize waste loading within the masonry cement solidification media, but to insure that the users Radwaste Solidification System is capable of producing waste forms which meet the waste form criteria of 10 CFR 61. A description of the laboratory small sample certification program and the actual full scale pilot plant verification approach used is included in this paper. Also included is a discussion of the development of a Process Control Program to ensure the reproducibility of the test results with actual waste

Advanced PWR technology development -Development of advanced PWR system analysis technology-

Energy Technology Data Exchange (ETDEWEB)

Jang, Moon Heui; Hwang, Yung Dong; Kim, Sung Oh; Yoon, Joo Hyun; Jung, Bub Dong; Choi, Chul Jin; Lee, Yung Jin; Song, Jin Hoh [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1995-07-01

The primary scope of this study is to establish the analysis technology for the advanced reactor designed on the basis of the passive and inherent safety concepts. This study is extended to the application of these technology to the safety analysis of the passive reactor. The study was performed for the small and medium sized reactor and the large sized reactor by focusing on the development of the analysis technology for the passive components. Among the identified concepts the once-through steam generator, the natural circulation of the integral reactor, heat pipe for containment cooling, and hydraulic valve were selected as the high priority items to be developed and the related studies are being performed for these items. For the large sized passive reactor, the study plans to extend the applicability of the best estimate computer code RELAP5/MOD3 which is widely used for the safety analyses of the reactor system. The improvement and supplementation study of the analysis modeling and the methodology is planned to be carried out for these purpose. The newly developed technologies are expected to be applied to the domestic advanced reactor design and analysis and these technologies will play a key role in extending the domestic nuclear base technology and consolidating self-reliance in the essential nuclear technology. 72 figs, 15 tabs, 124 refs. (Author).

77 FR 53769 - Receipts-Based, Small Business Size Standard; Confirmation of Effective Date

Science.gov (United States)

2012-09-04

... Flexibility Act of 1980, as amended. The NRC is increasing its receipts-based, small business size standard from $6.5 million to $7 million to conform to the standard set by the Small Business Administration...-Based, Small Business Size Standard; Confirmation of Effective Date AGENCY: Nuclear Regulatory...

78 FR 30384 - Small Business Size Standards: Waiver of the Nonmanufacturer Rule

Science.gov (United States)

2013-05-22

... SBA's 8(a) Business Development (BD) program, or Women-Owned Small Businesses (WOSBs). DATES: Comments... small businesses, Service-Disabled Veteran-Owned (SDVO) small businesses, Women-Owned Small Businesses... SMALL BUSINESS ADMINISTRATION Small Business Size Standards: Waiver of the Nonmanufacturer Rule...

Shared or Integrated: Which Type of Integration is More Effective Improves Students’ Creativity?

Science.gov (United States)

Mariyam, M.; Kaniawati, I.; Sriyati, S.

2017-09-01

Integrated science learning has various types of integration. This study aims to apply shared and integrated type of integration with project based learning (PjBL) model to improve students’ creativity on waste recycling theme. The research method used is a quasi experiment with the matching-only pre test-post test design. The samples of this study are 108 students consisting of 36 students (experiment class 1st), 35 students (experiment class 2nd) and 37 students (control class 3rd) at one of Junior High School in Tanggamus, Lampung. The results show that there is difference of creativity improvement in the class applied by PjBL model with shared type of integration, integrated type of integration and without any integration in waste recycling theme. Class applied by PjBL model with shared type of integration has the higher creativity improvement than the PjBL model with integrated type of integration and without any integration. Integrated science learning using shared type only combines 2 lessons, hence an intact concept is resulted. So, PjBL model with shared type of integration more effective improves students’ creativity than integrated type.

Nondestructive testing of PWR type fuel rods by eddy currents and metrology in the OSIRIS reactor pool

International Nuclear Information System (INIS)

Faure, M.; Marchand, L.

1985-02-01

The Saclay Reactor Department has developed a nondestructive test bench, now installed above channel 1 of the OSIRIS reactor. As part of investigations into the dynamics of PWR fuel degradation, a number of fuel rods underwent metrological and eddy current inspection, after irradiation [fr

Successful E-Learning in Small and Medium-Sized Enterprises

Science.gov (United States)

Paulsen, Morten Flate

2009-01-01

So far, e-learning has primarily been used when there are many learners involved. The up-front investments related to e-learning are relatively high, and may be perceived as prohibitive for small and medium-sized enterprises (SMEs). Some e-learning is, however, getting less expensive, and some e-learning models are more suited for small-scale…

Regional statistical and economic analysis of small and medium-sized businesses development in Zhytomyr region

Directory of Open Access Journals (Sweden)