Beliefs concerning the reliability of nuclear power plant in-service inspections; Uskomuksia ydinvoimalaitoksissa suoritettavien tarkastusten luotettavuudesta

Energy Technology Data Exchange (ETDEWEB)

Kettunen, J. [VTT Automation, Espoo (Finland)

1997-01-01

The aim of the study was to investigate belief systems held by the officials responsible for the planning and supervision of NDT operations within the Finnish nuclear industry. They were asked to express their opinions on (1) the reliability of NDT methods in general, (2) the factors influencing the reliability of in-service inspections, and (3) the degree of reliability of the current inspections operations conducted by means of NDT methods in the Finnish nuclear power plants. Another goal of the study was to assess the adequacy of officials` beliefs (or belief systems). The research data was collected by interviewing representatives from Finnish power companies (Imatran Voima Oy and Teollisuuden Voima Oy), independent inspection organisations, and the Finnish Centre for Radiation and Nuclear Safety (STUK). The adequacy of the beliefs expressed was assessed by means of the results obtained from international NDT reliability studies and on the basis of interviewees` own justification. (refs.).

Beliefs concerning the reliability of nuclear power plant in-service inspections

International Nuclear Information System (INIS)

Kettunen, J.

1997-01-01

The aim of the study was to investigate belief systems held by the officials responsible for the planning and supervision of NDT operations within the Finnish nuclear industry. They were asked to express their opinions on (1) the reliability of NDT methods in general, (2) the factors influencing the reliability of in-service inspections, and (3) the degree of reliability of the current inspections operations conducted by means of NDT methods in the Finnish nuclear power plants. Another goal of the study was to assess the adequacy of officials' beliefs (or belief systems). The research data was collected by interviewing representatives from Finnish power companies (Imatran Voima Oy and Teollisuuden Voima Oy), independent inspection organisations, and the Finnish Centre for Radiation and Nuclear Safety (STUK). The adequacy of the beliefs expressed was assessed by means of the results obtained from international NDT reliability studies and on the basis of interviewees' own justification. (refs.)

Human and organisational factors influencing the reliability of non-destructive testing. An international literary survey

International Nuclear Information System (INIS)

Kettunen, J.; Norros, L.

1996-04-01

The aim of the study is to chart human and organisational factors influencing the reliability of non-destructive testing (NDT). The emphasis will be in ultrasonic testing (UT) and in the planning and execution of in-service inspections during nuclear power plant maintenance outages. Being a literary survey this study is mainly based on the foreign and domestic research available on the topic. In consequence, the results presented in this report reflect the ideas of international research community. In addition to this, Finnish nuclear power plant operators (Imatran Voima Oy and Teollisuuden Voima Oy), independent inspection organisations and the Finnish Centre for Radiation and Nuclear Safety have provided us with valuable information on NDT theory and practice. Especially, a kind of 'big picture' of non-destructive testing has been pursued in the study. (6 figs., 2 tabs.)

Human and organisational factors influencing the reliability of non-destructive testing. An international literary survey; Inhimillisten ja organisatoristen tekijoeiden yhteys NDT- tarkastusten luotettavuuteen. Katsaus kansainvaeliseen kirjallisuuteen

Energy Technology Data Exchange (ETDEWEB)

Kettunen, J.; Norros, L.

1996-04-01

The aim of the study is to chart human and organisational factors influencing the reliability of non-destructive testing (NDT). The emphasis will be in ultrasonic testing (UT) and in the planning and execution of in-service inspections during nuclear power plant maintenance outages. Being a literary survey this study is mainly based on the foreign and domestic research available on the topic. In consequence, the results presented in this report reflect the ideas of international research community. In addition to this, Finnish nuclear power plant operators (Imatran Voima Oy and Teollisuuden Voima Oy), independent inspection organisations and the Finnish Centre for Radiation and Nuclear Safety have provided us with valuable information on NDT theory and practice. Especially, a kind of `big picture` of non-destructive testing has been pursued in the study. (6 figs., 2 tabs.).

Means of achieving high load factors at Olkiluoto 1 and 2

International Nuclear Information System (INIS)

Patrakka, E.

2001-01-01

Teollisuuden Voima Oy operates two BWR units Olkiluoto 1 and 2 that have achieved load factors typically higher than 90%. The operating experiences gained in the 1990s is summarised and the factors contributing to the high capacity factors are addressed. These include the general objectives for operation and maintenance, plant modernisation programme, maintenance principles, and outage policy and experiences. Finally, the international evaluations performed at Olkiluoto are mentioned. (author)

3D-mainosvideo teollisuusyritykselle : Case: Leppäkosken Lämpö

OpenAIRE

Utriainen, Sari

2015-01-01

Toimeksiantajana opinnäytetyölle toimi biolämmitysjärjestelmiä valmistava ja markkinoiva Ariterm Oy. Tavoitteena oli tuottaa messu- ja markkinointikäyttöön 3D-mainosvideo Ariterm Oy:n toimittamasta, elokuussa 2014 käyttöönotetusta Leppäkosken Lämpö Oy:n pellettilämpölaitoksesta. Opinnäytetyössä käsiteltiin 3D-mallien hyödyntämistä eri teollisuuden aloilla sekä tutustuttiin 3D-mainosvideoprojektissa käytettyihin ohjelmistoihin ja teollisuuden 3D-suunnitteluohjelmien eri tiedostomuotoihin se...

A critical review of published groundwater flow models for safety of nuclear waste disposal

International Nuclear Information System (INIS)

Laine, E.

1997-04-01

Flow models have been simulated for the potential nuclear waste sites in Precambrian bedrock of Finland in the Technical Research Centre of Finland (VTT). The work had been commissioned by Teollisuuden Voima Oy. In the present study, the published flow models are critically reviewed. The work concentrates on qualitative evaluation of the applied equivalent continuum approach applied to crystalline bedrock. Special attention is paid to the use of the geological information in connection with flow modelling. (35 refs., 6 figs.)

Environmental impact assessments of a fifth nuclear power plant unit in Finland

International Nuclear Information System (INIS)

Aurela, Jorma; Koivisto, Katarina

2000-01-01

This paper presents the results of president questionnaires and media monitoring of press cuttings concerned with siting of the new fifth in a row Finnish NPP. Two years ago both Fortum Power and Heat Oy and Teollisuuden Voima Oy (TVO) launched their Environmental impact assessment (EIA) procedures of a new nuclear power unit in Finland. The EIA procedures were launched to investigate the environmental impacts of a fifth nuclear power plant which possibly will be built in Loviisa or at Olkiluoto. In Finland there are four operating NPP units, two in Loviisa (Fortum) and two in Eurajoki, Olkiluoto (TVO). In the EIA procedure citizens and various associations and authorities have an opportunity to express their views on the matters related to the project. The Ministry of Trade and Industry (MTI) as the coordination authority arranges the organisation of the EIA hearings and the collection of statements and opinions. The EIA procedure in Finland takes place in two stages. The first stage i.e. the EIA programme describes the project and presents the plan on how the environmental effects are investigated and assessed. In the second stage the actual assessment of the environmental effects of the project will be submitted. Both Fortum and Teollisuuden Voima Oy (TVO) launched in spring 1998 their EIA procedures. The main alternatives are the Loviisa 3 project includes two plant type alternatives. The size of the plant is between 1000 and 1700 MWe, and the extension project of the Olkiluoto NPP to build a NPP unit of about 1000-1500 MWe at Olkiluoto. The EIA reports were submitted to the MTI in August 1999 and after that they were on display for two months for opinions and statements

Local opinion - the impact of visitor centre two studies at Olkiluoto

International Nuclear Information System (INIS)

Aemmaelae, V.M.

1993-01-01

Teollisuuden Voima Oy (TVO) has recently commissioned two opinion polls about attitudes towards nuclear power. The target group of the first research was teachers in the nearby community. The subject of the poll was attitudes towards nuclear power and the impact of a visit. The second research handles the company image of TVO target groups being local elected officials, press, teachers and local representatives of trade organizations. The aim of these two studies was to measure the impact of the company's information and visits activities to the attitudes. (author)

Application of the REMIX thermal mixing calculation program for the Loviisa reactor

International Nuclear Information System (INIS)

Kokkonen, I.; Tuomisto, H.

1987-08-01

The REMIX computer program has been validated to be used in the pressurized thermal shock study of the Loviisa reactor pressure vessel. The program has been verified against the data from the thermal and fluid mixing experiments. These experiments have been carried out in Imatran voima Oy to study thermal mixing of the high-pressure safety injection water in the Loviisa VVER-440 type pressurized water reactor. The verified REMIX-versions were applied to reactor calculations in the probabilistic pressurized thermal shock study of the Loviisa Plant

SAFIR. The Finnish research programme on nuclear power plant safety 2003-2006. Executive summary

International Nuclear Information System (INIS)

Puska, E.

2006-12-01

Major part of Finnish public research on nuclear power plant safety during the years 2003-2006 has been carried out in the SAFIR programme. The programme has been administrated by the steering group that was nominated by the Ministry of Trade and Industry (KTM). The steering group of SAFIR has consisted of representatives from Radiation and Nuclear Safety Authority (STUK), Ministry of Trade and Industry (KTM), Technical Research Centre of Finland (VTT), Teollisuuden Voima Oy (TVO), Fortum Power and Heat Oy, Fortum Nuclear Services Oy (Fortum), Finnish Funding Agency for Technology and Innovation (Tekes), Helsinki University of Technology (TKK) and Lappeenranta University of Technology (LTY). The key research areas of SAFIR have been (1) reactor fuel and core, (2) reactor circuit and structural safety, (3) containment and process safety functions, that was divided in 2005 into (3a) thermal hydraulics and (3b) severe accidents, (4) automation, control room and IT, (5) organisations and safety management and (6) risk-informed safety management. The research programme has included annually from 20 up to 24 research projects, whose volume has varied from a few person months to several person years. The total volume of the programme during the four year period 2003-2006 has been 19.7 million euros and 148 person years. The research in the programme has been carried out primarily by Technical Research Centre of Finland (VTT). Other research units responsible for the projects include Lappeenranta University of Technology, Fortum Nuclear Services Oy, Helsinki University of Technology and RAMSE Consulting Oy. In addition, there have been a few minor subcontractors in some projects. The programme management structure has consisted of the steering group, a reference group in each of the seven research areas and a number of ad hoc groups in the various research areas. This report gives a short summary of the results of the SAFIR programme for the period January 2003 - November

Fuel cycle management in Finland

International Nuclear Information System (INIS)

Vaeyrynen, H.; Mikkola, I.

1987-01-01

Both Finnish utilities producing nuclear power - Imatran Voima Oy (IVO) and Teollisuuden Voima Oy (Industrial Power Co. Ltd, TVO) - have created efficient fuel cycle management systems. The systems however differ in almost all respects. The reason is that the principal supplier for IVO is the Soviet Union and for TVO is Sweden. A common feature of both systems at the front end of the cycle is the building of stockpiles in order to provide for interruptions in fuel deliveries. Quality assurance supervision at the fuel factory for IVO is regulated by the Soviet Chamber of Commerce and Industry and a final control is made in Finland. The in-core fuel management is done by IVO using codes developed in Finland. The whole IVO fuel cycle is basically a leasing arrangement. The spent fuel is returned to the USSR after five years cooling. TVO carries out the in-core fuel management using a computer code system supplied by Asea-Atom. TVO is responsable for the back end of the cycle and makes preparations for the final disposal of the spent fuel in Finland. 6 refs., 2 figs

Operation set for 2018 as regulator considers olkiluoto-3 licence application

Energy Technology Data Exchange (ETDEWEB)

Kraev, Kamen [NucNet, Brussels (Belgium)

2017-02-15

Finland has announced progress with its delayed nuclear project and has confirmed it will not be affected by anomalies discovered in some components manufactured for EPRs in France. The Olkiluoto-3 European Pressurised Reactor (EPR) nuclear plant under construction in Finland is on schedule to begin commercial operation in 2018 with the country's regulator preparing a safety assessment that will pave the way for fuel loading. Jouni Silvennoinen, senior vice-president for Olkiluoto-3 at Teollisuuden Voima Oyj (TVO), told NucNet that fuel loading at the EPR plant, which is nine years behind schedule, is expected in the spring of 2018. He said construction and licensing of the plant are progressing.

Regulatory control of nuclear safety in Finland. Annual report 1998

Energy Technology Data Exchange (ETDEWEB)

Tossavainen, K. [ed.

1999-10-01

The report describes regulatory control of the safe use of nuclear energy by the Radiation and Nuclear Safety Authority (STUK) in 1998. STUK is the Finnish nuclear safety authority. The submission of this report to the Ministry of Trade and Industry is stipulated in Section 121 of the Nuclear Energy Decree. It was verified by regulatory control that the operation of Finnish NPPs was in compliance with conditions set out in the operating licences of the plants and with regulations currently in force. In addition to supervising the normal operation of the plants, STUK oversaw projects carried out at the plant units, which related to the uprating of their power and the improvement of their safety. STUK issued to the Ministry of Trade and Industry a statement about applications for the renewal of the operating licences of Loviisa and Olkiluoto NPPs, which had been submitted by Imatran Voima Oy and Teollisuuden Voima Oy. Regulatory activities in the field of nuclear waste management were focused on the storage and final disposal of spent fuel as well as the treatment, storage and final disposal of reactor waste. STUK issued a statement to the Ministry of Trade and Industry about an environmental impact assessment programme pertaining to a spent fuel repository project, which had been submitted by Posiva Oy, as well as on Imatran Voima Oy's application concerning the operation of a repository for medium- and low-level reactor waste from Loviisa NPP. The use of nuclear materials was in compliance with the regulations currently in force and also the whereabouts of every batch of nuclear material were ensured by safeguards control. In international safeguards, important changes took place, which were reflected also in safeguards activities at national level. International co-operation continued based on financing both from STUK's budget and from additional sources. The focus of co-operation funded from outside sources was as follows: improvement of the safety of

Regulatory control of nuclear safety in Finland. Annual report 1998

International Nuclear Information System (INIS)

Tossavainen, K.

1999-10-01

The report describes regulatory control of the safe use of nuclear energy by the Radiation and Nuclear Safety Authority (STUK) in 1998. STUK is the Finnish nuclear safety authority. The submission of this report to the Ministry of Trade and Industry is stipulated in Section 121 of the Nuclear Energy Decree. It was verified by regulatory control that the operation of Finnish NPPs was in compliance with conditions set out in the operating licences of the plants and with regulations currently in force. In addition to supervising the normal operation of the plants, STUK oversaw projects carried out at the plant units, which related to the uprating of their power and the improvement of their safety. STUK issued to the Ministry of Trade and Industry a statement about applications for the renewal of the operating licences of Loviisa and Olkiluoto NPPs, which had been submitted by Imatran Voima Oy and Teollisuuden Voima Oy. Regulatory activities in the field of nuclear waste management were focused on the storage and final disposal of spent fuel as well as the treatment, storage and final disposal of reactor waste. STUK issued a statement to the Ministry of Trade and Industry about an environmental impact assessment programme pertaining to a spent fuel repository project, which had been submitted by Posiva Oy, as well as on Imatran Voima Oy's application concerning the operation of a repository for medium- and low-level reactor waste from Loviisa NPP. The use of nuclear materials was in compliance with the regulations currently in force and also the whereabouts of every batch of nuclear material were ensured by safeguards control. In international safeguards, important changes took place, which were reflected also in safeguards activities at national level. International co-operation continued based on financing both from STUK's budget and from additional sources. The focus of co-operation funded from outside sources was as follows: improvement of the safety of Kola and

Parliament's decision specified the energy policy in Finland

International Nuclear Information System (INIS)

Aurela, J.

2002-01-01

The Finnish Parliament decided on Friday 24 May 2002, to ratify the favourable decision-in-principle on the fifth nuclear power plant unit made by the Government last January, with 107 votes in favour and 92 votes against. Thus Teollisuuden Voima Oy (TVO), the Finnish privately owned electricity generation company responsible for the project, is authorised to continue the preparations for the construction of a new nuclear power plant unit. The article handles the process in the coming years if TVO will go on with the process. The most important steps according to the nuclear legislation are to have construction and operating licenses in due time. (author)

Bedrock model of the Veitsivaara area

International Nuclear Information System (INIS)

Saksa, P.; Kuivamaeki, A.; Kurimo, M.; Anttila, P.; Front, K.; Pitkaenen, P.; Korkealaakso, J.; Vaittinen, T.

1993-07-01

Site investigations were carried out at Veitsivaara, in 1987-1991 in accordance with an investigation programme for radioactive waste disposal drawn up by Teollisuuden Voima Oy (TVO). The site was modelled in terms of rock types, fracturing, fracture structures and geophysical conditions, the main focus of examination was on fracturing and associated hydraulic conductivity. The various properties of the bedrock structures were classified by means a three-dimensional model. The descriptions of the models were stored in a computer system for illustration purposes. The rock types at Veitsivaara are tonalite gneiss, Tuliniemet potassium granite, amphipolite, granite porphyry and metadiabase, the last two of which occur in dykes

Bedrock model of the Olkiluoto area

International Nuclear Information System (INIS)

Saksa, P.; Paananen, M.; Paulamaeki, S.; Anttila, P.; Front, K.; Pitkaenen, P.; Hassinen, P.; Ylinen, A.

1993-07-01

Site investigations were carried out at Olkiluoto (in Finland) in 1987-1992 in accordance with an investigation programme drawn up by Teollisuuden Voima Oy (TVO). The site was modelled in terms of rock types, fracturing, fracture structures and geohydrological conditions, the main focus of examination was on fracturing and associated hydraulic conductivity. The various properties of the bedrock structures were classified by means of a three-dimensional model. The descriptions of the models were gathered in a computer system for illustration and storage purposes. The rock types at Olkiluoto are migmatite, which may be divided into mica gneiss and veined gneiss, and also tonalite and coarse-grained migmatite granite (pegmatite). (64 refs., 65 figs.)

Finnish participation in the European utility requirements work

International Nuclear Information System (INIS)

Patrakka, E.

2000-01-01

The Finnish participation in the EUR process started already in April 1994 when IVO (Imatran Voima Oy presently Fortum Oyj) and TVO (Teollisuuden Voima Oy) were asked to comment EUR Volume 1 and 2 Revision A in April 1994. A formal application for the Finnish membership in the EUR organisation was sent on 20 November 1995, and Finland was accepted as an associated member on the next day. The Finnish representatives in the various EUR bodies were appointed in March 1996, at which time the formal participation in these bodies commenced. On 7 November 1996, EUR Steering Committee approved a full membership of IVO and TVO that are joint EUR members representing Finland together. A major Finnish contribution was made in 1997 when IVO and TVO performed a comparison between the EUR document and YVL guides. The period of the Finnish membership has been characterised by the compilation of EUR Volume 3 subsets, in which process IVO/Fortum and TVO have been actively participating. From the Finnish point of view, the EUR work can also be seen as a part of getting prepared to proceed with a possible new NPP project. The EUR document is a substantial aid when preparing the technical specifications for a NPP bid inquiry. The information received in connection with the detailed assessment work for Volume 3 subsets is very valuable when considering the feasibility of NPP concepts. In addition. the experiences gained in the Volume 3 activities enable to develop even better requirements that are manifested by Revision C of Volumes 1 and 2. (author)

TVO-92 safety analysis of spent fuel disposal

International Nuclear Information System (INIS)

Vieno, T.; Hautojaervi, A.; Koskinen, L.; Nordman, H.

1993-08-01

The spent fuel from the TVO I and TVO II reactors at the Olkiluoto nuclear power plant is planned to be disposed in a repository constructed at a depth of about 500 meters in crystalline bedrock. Teollisuuden Voima Oy (TVO) has carried out preliminary site investigations for spent fuel disposal between 1987 and 1992 at five areas in Finland (Olkiluoto, Kivetty, Romuvaara, Syyry and Veitsivaara). The Safety analysis of the disposal system is presented in the report. Spent fuel will be encapsulated in composite copper-steel canisters. The canister design (ACP canister) consists of an inner container of steel as a load-bearing element and an outer container of oxygen-free copper to provide a shield against corrosion. In the repository the canisters will be emplaced in vertical holes drilled in the floors of horizontal deposition tunnels. The annulus between the canister and the rock is filled with compacted bentonite. The results of the safety analysis attest that the planned disposal system fulfils the safety requirements. Suitable places for the repository can be found at each of the five investigation sites

The Geology of the Kivetty area

International Nuclear Information System (INIS)

Anttila, P.; Paulamaeki, S.; Lindberg, A.; Paananen, M.; Koistinen, T.; Front, K.; Pitkaenen, P.

1992-05-01

Teollisuuden Voima Oy (TVO) is preparing for the final disposal of spent nuclear fuel from the Olkiluoto nuclear power plant (TVO-I and TVO-II) deep in the Finnish bedrock. Kivetty in Konginkangas was one of the five areas selected in 1987 for preliminary site investigations for this purpose. The Kivetty area in Central Finland is located in a Svecokarelian granitoid environment consisting of a complex of synorogenic granitoids 1900 - 1860 million years in age. The bedrock consists almost entirely of plutonic rocks, i.e. gabbro, porphyritic granodiorite and granite, eguigranular granodiorite and granite, listed in order of age. The majority of the rock types are porphyritic in character, and supracrustal rocks such as quartz-feldspar schist and gneiss are found occasionally in small xenoliths

The international INTRAVAL project. Summary and conclusions by the TVO/VTT Team

International Nuclear Information System (INIS)

Hautojaervi, A.

1994-12-01

Teollisuuden Voima Oy (TVO) participated the international cooperation project INTRAVAL and VTT Energy acted as a project team. The Finnish participation focused on flow and transport in crystalline fractured rock and six test cases out of thirteen were tackled. The experimental results were evaluated mainly by means of analytical transport models. The report presents a short review of the experience obtained in the course of the project. It concentrates on the issues revealed in the discussions and analyses of the six test cases in which the TVO/VTT team actively participated but some of the conclusions are even more general in nature. Some suggestions are made for future experimental and theoretical work in the field of geosphere. (15 refs., 2 tabs.)

Design principle of TVO's final repository and preliminary adaptation to site specific conditions

International Nuclear Information System (INIS)

Salo, J-P.; Reikkola, R.

1995-01-01

Teollisuuden Voima Oy (TVO) is responsible for the management of spent fuel produced by the Olkiluoto power plant. TVO's current programme of spent fuel management is based on the guidelines and time schedule set by the Finnish Government. TVO has studied a final disposal concept in which the spent fuel bundles are encapsulated in copper canisters and emplaced in Finnish bedrock. According to the plan the final repository for spent fuel will be in operation by 2020. TVO's updated technical plans for the disposal of spent fuel together with a performance analysis (TVO-92) were submitted to the authorities in 1992. The paper describes the design principle of TVO's final repository and preliminary adaptation of the repository to site specific conditions. (author). 10 refs., 5 figs

Bedrock model of the Kivetty area

International Nuclear Information System (INIS)

Saksa, P.; Paulamaeki, S.; Paananen, M.; Anttila, P.; Front, K.; Pitkaenen, P.; Korkealaakso, J.; Okko, O.

1993-07-01

Preliminary site investigations were carried out at Kivetty (in Finland), in 1987-1992 in accordance with the investigation programme drawn up by Teollisuuden Voima Oy (TVO). The site was modelled in terms of rock type, fracturing, fracture structures and geohydrological conditions, with the main emphasis being placed on fracturing and associated hydraulic conductivity. The various properties of the bedrock structures were classified in relation to a three-dimensional model. The descriptions of the models were stored in a computer system for the purpose of illustration. The principal rock types encountered at the Kivetty site are porphyritic granodiorite and porphyritic granite, in addition to which even-grained granite and granodiorite, gabbro, and small felsic and mafic veins occur. The rocks have undergone two distinct phases of deformation. (41 refs., 50 figs.)

Geophysical borehole logging. Final disposal of spent fuel

International Nuclear Information System (INIS)

Rouhiainen, P.

1984-01-01

Teollisuuden Voima Oy (Industrial Power Company Ltd.) will take precautions for final disposal of spent fuel in the Finnish bedrock. The first stage of the site selection studies includes drilling of a deep borehole down to approximately 1000 meters in the year 1984. The report deals with geophysical borehole logging methods, which could be used for the studies. The aim of geophysical borehole logging methods is to descripe specially hydrogeological and structural features. Only the most essential methods are dealt with in this report. Attention is paid to the information produced with the methods, derscription of the methods, interpretation and limitations. The feasibility and possibilities for the aims are evaluated. The evaluations are based mainly on the results from Sweden, England, Canada and USA as well as experiencies gained in Finland

Diffusion of water, cesium and neptunium in pores of rocks

International Nuclear Information System (INIS)

Puukko, E.; Heikkinen, T.; Hakanen, M.

1993-10-01

Teollisuuden Voima Oy (TVO) is investigating the feasibility to dispose of spent nuclear fuel within Finland. The present plan calls for the repository to be located in crystalline rock at a depth of several hundred meters. The safety assessment of the repository includes calculations of migration of waste nuclides. The flow of waste elements in groundwater will be retarded through sorption interaction with minerals and through diffusion into rock. Diffusion is the only mechanism retarding the migration of non-sorbing species and, it is expected to be the dominating retardation mechanism of many of the sorbing elements. In the investigation the simultaneous diffusion of tritiated water (HTO), cesium and neptunium in rocks of TVO investigation sites at Kivetty, Olkiluoto and Romuvaara were studied. (11 refs., 33 figs., 9 tabs.)

Experimental testing of an ABB Master application

International Nuclear Information System (INIS)

Haapanen, P.; Maskuniitty, M.; Korhonen, J.; Tuulari, E.

1995-10-01

A prototype dynamic testing harness for programmable automation systems has been specified and implemented at the Technical Research Centre of Finland (VTT). In order to get experience on the methodology and equipment for the testing of systems important to the safety of nuclear power plants, where the safety and reliability requirements often are very high, two different pilot systems have been tested. One system was an ABB Master application, which was loaned for testing from ABB Atom by Teollisuuden Voima Oy (TVO). Another system, loaned from Siemens AG (SAG) by IVO International Oy (IVO), was an application realized with SAG's digital SILT technology. The report describes the experiences gained in testing an APRM pilot system realized with ABB Master technology. The testing of the pilot application took place in the VTT Automation laboratory in Otaniemi in September-October 1994. The purpose of the testing was not to assess the quality of the pilot system, but to get experience in the testing methodology and find out the further development needs and potentials of the test methodology and equipment. (7 refs., 14 figs., 9 tabs.)

An approach to quality classification of deep groundwaters in Sweden and Finland

International Nuclear Information System (INIS)

Laaksoharju, M.; Smellie, J.; Ruotsalainen, P.; Snellman, M.

1993-11-01

In Sweden and Finland high quality groundwater samples are required in the site characterization programmes relating to safe disposal of spent nuclear fuel. SKB (Swedish Nuclear Fuel and Waste Management Co.) and TVO (Teollisuuden Voima Oy, Finland) initiated a cooperative task to critically evaluate the quality of the earlier sampling programmes and to further develop the understanding of quality or representativeness of the groundwater samples. The major aim in this report has been, therefore, to make an attempt to classify groundwaters from site investigations in Sweden and Finland based on quality. Different classification systems have been tested and developed. These can be divided in two main groups; manual methods and computer-based mathematical methods. Manual, statistical, mixing ratio and scoring systems have all been used to illustrate the difficulty in judging groundwater quality. (28 refs., 19 figs., 11 tabs.)

The geology of the Romuvaara area

International Nuclear Information System (INIS)

Anttila, P.; Paulamaeki, S.; Lindberg, A.; Paananen, M.; Pitkaenen, P.; Front, K.

1990-12-01

Teollisuuden Voima Oy (TVO) is preparing for the final disposal of spent uranium fuel from the Olkiluoto nuclear power plant deep in the Finnish bedrock. The report presents a summary of the geological conditions at Romuvaara in Kuhmo, which was one of the five areas selected in 1987 for the preliminary site investigations. The Romuvaara site and its surroundings belong to the Archaean basement complex, the age of the oldest parts of which is over 2800 Ma. The bedrock consists mainly of migmatic banded gneisses (tonalite, leucotonalite and mica gneiss). These rock types are intersected by granodiorite and metadiabase dykes. Proterozoic metadiabases represent the youngest rock unit in the area. Except for the metadiabase, the rocks have undergone a multiphase Archaean deformation. The bedrock structures are interpreted as representing six deformation phases, after which sharp faults developed during at least four further movement phases

Operation of Finnish nuclear power plants. Quarterly report, 3rd quarter 1996

International Nuclear Information System (INIS)

Sillanpaeae, T.

1997-02-01

Quarterly Reports on the operation of Finnish nuclear power plants describe events and observations relating to nuclear and radiation safety which the Finnish Centre for Radiation and Nuclear Safety (STUK) considers safety significant. Safety improvements at the plants are also described. The Report also includes a summary of the radiation safety of plant personnel and of the environment and tabulated data on the plants' production and load factors. In the third quarter of 1996, the Finnish nuclear power plant units were in power operation except for the annual maintenance outages of Loviisa plant units and a shutdown at Olkiluoto 1 to identify and repair malfunctions of a high pressure turbine control valve. The load factor average of all plant units was 77.2%. Events in the third quarter of 1996 were classified level 0 on the International Nuclear Event Scale. Occupational doses and radioactive releases off-site were below authorised limits. Radioactive substances were measurable in samples collected around the plants in such quantities only as have no bearing on the radiation exposure of the population. The names of Teollisuuden Voima Oy's plant units have changed. Olkiluoto 1 and Olkiluoto 2 now replace the names TVO I and TVO II previously used in quarterly reports. (orig.)

Isothermal Kinetics of Diesel Soot Oxidation over La0.7K0.3ZnOy Catalysts

Directory of Open Access Journals (Sweden)

Ram Prasad

2014-10-01

Full Text Available This paper describes the kinetics of catalytic oxidation of diesel soot with air under isothermal conditions (320-350 oC. Isothermal kinetics data were collected in a mini-semi-batch reactor. Experiments were performed over the best selected catalyst composition La0.7K0.3ZnOy prepared by sol-gel method. Characterization of the catalyst by XRD and FTIR confirmed that La1-xKxZnOy did not exhibit perovskite phase but formed mixed metal oxides. 110 mg of the catalyst-soot mixture in tight contact (10:1 ratio was taken in order to determine the kinetic model, activation energy and Arrhenius constant of the oxidation reaction under the high air flow rate assuming pseudo first order reaction. The activation energy and Arrhenius constant were found to be 138 kJ/mol and 6.46x1010 min-1, respectively. © 2014 BCREC UNDIP. All rights reservedReceived: 26th April 2014; Revised: 27th May 2014; Accepted: 28th June 2014How to Cite: Prasad, R., Kumar, A., Mishra, A. (2014. Isothermal Kinetics of Diesel Soot Oxidation over La0.7K0.3ZnOy Catalysts. Bulletin of Chemical Reaction Engineering & Catalysis, 9(3: 192-200. (doi: 10.9767/bcrec.9.3.6773.192-200Permalink/DOI: http://dx.doi.org/10.9767/bcrec.9.3.6773.192-200

Commensurability oscillations in NdBa2Cu3Oy single crystals

Indian Academy of Sciences (India)

gated by angular dependent magnetization in very pure twinned and twin-free NdBa2 Cu3 Oy single ... The layered structure and the c-axis coherence length, ξc ≈ 4 ˚A, smaller than the lattice ... The high quality of both crystals is demonstrated by ... Commensurability oscillations in NdBa2Cu3Oy single crystals. 2. 3. 4. 5. 6.

Research into basic rocks types

International Nuclear Information System (INIS)

1993-06-01

Teollisuuden Voima Oy (TVO) has carried out research into basic rock types in Finland. The research programme has been implemented in parallel with the preliminary site investigations for radioactive waste disposal in 1991-1993. The program contained two main objectives: firstly, to study the properties of the basic rock types and compare those with the other rock types under the investigation; secondly, to carry out an inventory of rock formations consisting of basic rock types and suitable in question for final disposal. A study of environmental factors important to know regarding the final disposal was made of formations identified. In total 159 formations exceeding the size of 4 km 2 were identified in the inventory. Of these formations 97 were intrusive igneous rock types and 62 originally extrusive volcanic rock types. Deposits consisting of ore minerals, industrial minerals or building stones related to these formations were studied. Environmental factors like natural resources, protected areas or potential for restrictions in land use were also studied

Summary report of the experiences from TVO's site investigations

International Nuclear Information System (INIS)

Oehberg, A.; Saksa, P.; Ahokas, H.; Ruotsalainen, P.; Snellman, M.

1994-09-01

In 1992 Teollisuuden Voima Oy (TVO) completed preliminary site investigations for radioactive waste disposal at five sites in Finland. The aim of this report was the compilation of the experiences from TVO's site investigations. The main interest was focused on investigation strategies and the most important investigation methods for the conceptual modelling. The objective of the preliminary site investigations was to obtain data on the bedrock properties in order to evaluate the areas. The programme was divided into four stages, each stage having its own subobjective. The site-specific investigation programme for each site included a large common part and a small site-specific part. The strategies (objectives) and experiences from different disciplines, geology, hydrogeochemistry, geophysics and geohydrology, are presented in the report. The conceptual modelling work procedure including both bedrock and groundwater modelling is described briefly using the Olkiluoto site as an example. Each of the other areas has undergone similar phases of work. (52 refs., 45 figs., 5 tabs.)

Simulation of the groundwater flow of the Kivetty area

International Nuclear Information System (INIS)

Taivassalo, V.; Meszaros, F.

1994-02-01

Teollisuuden Voima Oy (TVO) is preparing for the final disposal of spent nuclear fuel into crystalline bedrock in Finland. Groundwater flow modelling is a part of the preliminary site investigation work. The aim is to simulate groundwater flow as realistically as possible in view of the experimental data available. Three dimensional groundwater flow modelling is based on a conceptual bedrock model. The modelling results will be used in the site evaluation process. Observations from flow simulations will also be used to identify and study uncertainties included in the site characterization. First a conceptual flow model for the Kivetty site in Konginkangas was developed. As a second stage the flow model was calibrated. The goal was to increase the reality of the model. To evaluate the reality of the flow model, the values of the input and output parameters were compared with the field data. Finally groundwater flow simulation results were computed and groundwater flow at the Kivetty area was analysed. (50 refs., 78 figs., 7 tabs.)

Bedrock Model of the Syyry area

International Nuclear Information System (INIS)

Saksa, P.; Kuivamaeki, A.; Kurimo, M.; Paananen, M.; Anttila, P.; Front, K.; Pitkaenen, P.; Hassinen, P.; Ylinen, A.

1993-09-01

Preliminary site investigations implemented in accordance with the research programme drawn up by Teollisuuden Voima Oy (TVO) were carried out at Syyry (in Finland) in 1987-1992. Models of the site were compiled and used for describing the rock types, fracturing, fracture structures and geohydrological conditions, the main emphasis being on the examination of the bedrock fracturing and related hydraulic conductivity. Three-dimensional models were used for the classification of the various properties of the bedrock structures. The descriptive models were gathered into a computer system to facilitate illustration and storage. The main rock type at Syyry is tonalite. A mica gneiss formation SE of the investigation site dips towards the NW and delimits the tonalite as far as the central part of the investigation site. The miga gneiss has a heterogeneous composition and includes intermediate layers consisting of quartz feldspar schist and amphibolite. There are mafic formations in the vicinity of the investigation site. The intrusive rocks have been deformed during three plastic and three mainly brittle deformation stages. (47 refs., 61 figs.)

Local negotiation on compensation siting of the spent nuclear fuel repository in Finland

International Nuclear Information System (INIS)

Kojo, Matti

2007-01-01

The aim of the paper is to analyse the local negotiation process between the Municipality of Eurajoki and the nuclear power company Teollisuuden Voima (TVO) and the nuclear waste management company Posiva Oy. The aim of the negotiations was to find an acceptable form of compensation for siting a spent nuclear fuel repository in Olkiluoto, Finland. The paper includes background information on the siting process in Finland, the local political setting in the Municipality of Eurajoki and a description of the negotiation process. The analysis of the negotiations on compensation is important for better understanding the progress of the Finnish siting process. The paper describes the picture of the contest to host the spent nuclear fuel repository. It also provides more information on the relationship between the Municipality of Eurajoki and the power company TVO. The negotiations on compensation and the roles of various players in the negotiations have not been studied in detail because the minutes of the Vuojoki liaison group were not available before the decision of the Supreme Administrative Court in May 2006. (author)

Compiling Utility Requirements For New Nuclear Power Plant Project

International Nuclear Information System (INIS)

Patrakka, Eero

2002-01-01

Teollisuuden Voima Oy (TVO) submitted in November 2000 to the Finnish Government an application for a Decision-in-Principle concerning the construction of a new nuclear power plant in Finland. The actual investment decision can be made first after a positive decision has been made by the Government and the Parliament. Parallel to the licensing process, technical preparedness has been upheld so that the procurement process can be commenced without delay, when needed. This includes the definition of requirements for the plant and preliminary preparation of bid inquiry specifications. The core of the technical requirements corresponds to the specifications presented in the European Utility Requirement (EUR) document, compiled by major European electricity producers. Quite naturally, an amount of modifications to the EUR document are needed that take into account the country- and site-specific conditions as well as the experiences gained in the operation of the existing NPP units. Along with the EUR-related requirements concerning the nuclear island and power generation plant, requirements are specified for scope of supply as well as for a variety of issues related to project implementation. (author)

Geophysical investigations in the Veitsivaara area, Finland summary report

International Nuclear Information System (INIS)

Heikkinen, E.; Saksa, P.; Hinkkanen, H.

1991-10-01

Teollisuuden Voima Oy (TVO carries out site investigations in Finland for final disposal of nuclear high level waste during 1987-2000. Investigations by geological, geophysical, geohydrological and geochemical methods were carried out in the Veitsivaara area in 1987-90 to determine the suitability of the bedrock for the final disposal of spent nuclear fuel. Airborne, ground and borehole geophysical methods were used to study the rock type distribution, fracturing and hydraulic conductivity. Airborne surveys were performed by magnetic, radiometric and two electromagnetic methods and ground investigations by VLF magnetic and resistivity, magnetic and impulse radar methods. Electromagnetic and seismic refraction surveys were used to locate crushed and fracture zones. The properties of weak electrical conductors, e.g. their depth dimensions, were studied by direct current resistivity measurements. The rock type distribution was studied by single-hole logging of susceptibility, natural γ-radiation and radiometric γ γ-density. Electrical and acoustic logging allowed water bearing fractures to be mapped and the results of water injection tests to be interpreted. Flow conditions in the boreholes were studied by both fluid logging and tube wave sounding

Final disposal of spent fuel in the Finnish bedrock

International Nuclear Information System (INIS)

1992-12-01

Teollisuuden Voima Oy (TVO) is preparing for the final disposal of spent nuclear fuel from the Olkiluoto nuclear power plant (TVO-I and TVO-II reactors). According to present estimates, a total of 1840 tU of spent fuel will be accumulated during the 40-year lifetime of the power plant. An interim storage facility for spent fuel (TVO-KPA Store) has operated at Olkiluoto since 1987. The spent fuel will be held in storage for several decades before it is shipped to the repository site. Both train and road transportation are possible. The spent fuel will be encapsulated in composite copper and steel canisters (ACP Canister) in a facility that will be build above the ground on the site where the repository is located. The repository will be constructed at the depth of several hundreds of meters in the bedrock. In 1987 five areas were selected for preliminary site investigations. The safety analysis (TVO-92) that was carried out shows that the proposed safety criteria would be met at each of the candidate sites. In future expected conditions there would never be significant releases of radioactive substances to the biosphere. The site investigations will be continued in the period 1993 to 2000. In parallel, a R and D programme will be devoted to the safety and technology of final disposal. The site for final disposal will be selected in the year 2000 with the aim of having the capability to start the disposal operations in 2020

Influence of calcium on transport properties, band spectrum and superconductivity of YBa2Cu3O(y) and YBa(1.5)La(0.5)Cu3O(y)

Science.gov (United States)

Gasumyants, V. E.; Vladimirskaya, E. V.; Patrina, I. B.

1995-01-01

The comparative investigation of transport phenomena in Y(1-x)Ca(x)Ba2Cu3O(y) (0 is less than x is less than 0.25; 6.96 is greater than y is greater than 6.87 and 6.73 is less than x is less than 6.53); Y(1-x)Ca(x)Ba(1.5)La(0.5)Cu3O(y) (0 is less than x is less than 0.5; 7.12 is greater than y is greater than 6.96) and YBa(2-x)La(x)Cu3O(y) (0 is less than x is less than 0.5; 6.95 is less than y is less than 7.21) systems have been carried out. The temperature dependencies of resistivity and thermopower have been measured. It was found that the S(T) dependencies take some additional features with Ca content increase. The results obtained have been analyzed on the basis of the phenomenological theory of electron transport in the case of the narrow conductive band. The main parameters of the band spectrum (the band filling with electrons degree and the total effective band width) have been determined. The dependencies of these from contents of substituting elements are discussed. Analyzing the results obtained simultaneously with the tendencies in oxygen content and critical temperature change we have confirmed the conclusion that the oxygen sublattice disordering has a determinant effect on band structure parameters and superconductive properties of YBa2Cu3O(y). The results obtained suggest that Ca gives rise to some peculiarities in band spectrum of this compound.

Finnish research programmes on nuclear power plant safety

International Nuclear Information System (INIS)

Puska, E. K.

2010-01-01

The current Finnish national research programme on nuclear power plant safety SAFIR2010 for the years 2007-2010 as well as the coming SAFIR2014 programme for the years 2011-2014 are based on the chapter 7a, 'Ensuring expertise', of the Finnish Nuclear Energy Act. The objective of this chapter is realised in the research work and education of experts in the projects of these research programmes. SAFIR2010 research programme is divided in eight research areas that are Organisation and human, Automation and control room, Fuel and reactor physics, Thermal hydraulics, Severe accidents, Structural safety of reactor circuit, Construction safety, and Probabilistic Safety Analysis (PSA). All the research areas include both projects in their own area and interdisciplinary co-operational projects. Research projects of the programme are chosen on the basis of annual call for proposals. In 2010 research is carried out in 33 projects in SAFIR2010. VTT is the responsible research organisation in 26 of these projects and VTT is also the coordination unit of SAFIR2010 and SAFIR2014. In 2007-2009 SAFIR2010 produced 497 Specified research results (Deliverables), 618 Publications, and 33 Academic degrees. SAFIR2010 programme covers approximately half of the reactor safety research volume in Finland currently. In 2010 the programme volume is EUR 7.1 million and 47 person years. The major funding partners are VYR with EUR 2.96 million, VTT with EUR 2.66 million, Fortum with EUR 0.28 million, TVO with EUR 0.19 million, NKS with EUR 0.15 million, EU with only EUR 0.03 million and other partners with EUR 0.85 million. The new decisions-in-principle on Olkiluoto unit 4 for Teollisuuden Voima and new nuclear power plant for Fennovoima ratified by the Finnish Parliament on 1 July 2010 increase the annual funding collected according to the Finnish Nuclear Energy Act from Fennovoima, Fortum and Teollisuuden Voima for the SAFIR2014 programme to EUR 5.2 million from the current level of EUR 3

Determination of interstitial oxygen atom position in U2N3+xOy by near edge structure study

Science.gov (United States)

Jiang, A. K.; Zhao, Y. W.; Long, Z.; Hu, Y.; Wang, X. F.; Yang, R. L.; Bao, H. L.; Zeng, R. G.; Liu, K. Z.

2018-06-01

The determination of interstitial oxygen atom site in U2N3+xOy film could facilitate the understanding of the oxidation mechanism of α-U2N3 and the effect of U2N3+xOy on anti-oxidation. By comparing the similarities and variances between N K edge and O K edge electron energy loss spectra (EELS) for oxidized α-U2N3 and UO2, the present work looks at the local structure of nitrogen and oxygen atoms in U2N3+xOy film, identifying the most possible position of interstitial O atom.

Finnish experiences on licensing and using of programmable digital systems in nuclear power plants

International Nuclear Information System (INIS)

Haapanen, P.; Maskuniitty, M.; Heimburger, H.; Hall, L.E.; Manninen, T.

1993-01-01

Finnish utility companies, Imatran Voima Oy (IVO) and Teollisuuden Voima (TVO), and the licensing authority, the Finnish Centre for Radiation and Nuclear Safety (STUK), are preparing for a new nuclear power plant in Finland. Plant vendors are proposing programmable digital automation systems for both the safety-related and the operational I and C (instrumentation and control) systems in this new unit. Also in existing plant units the replacement of certain old analog systems with state-of-the-art digital ones will become necessary in the years to come. Licensing of programmable systems for safety critical applications requires a new approach due to the special properties and failure modes of these systems. The major difficulties seem to be in the assessment and quantification of software reliability. The Technical Research Centre of Finland has in co-operation with the authority and the utilities conducted a project (AJA) to develop domestically applicable licensing requirements, guidelines and practices. International standards, guidelines and licensing practices have been analyzed in order to specify national licensing requirements. The paper describes and discusses the findings and experiences of the AJA project so far. The experience in introducing advanced programmable digital control and computer systems in the operating nuclear power plants will be covered briefly. Although these systems are not safety-related but systems of more general interest regarding nuclear safety, some routines regarding the licensing of safety- related systems have been followed. In these backfitting and replacement projects some experience have been gained in how to license safety-related programmable systems. (Author) 31 refs., 2 figs

SAFIR2014. The Finnish Research Programme on Nuclear Power Plant Safety 2011-2014. Interim Report

Energy Technology Data Exchange (ETDEWEB)

Simola, K. [ed.

2013-02-15

The Finnish Nuclear Power Plant Safety Research Programme 2011-2014, SAFIR2014, is a 4-year publicly funded national technical and scientific research programme on the safety of nuclear power plants. The programme is funded by the State Nuclear Waste Management Fund (VYR), as well as other key organisations operating in the area of nuclear energy. The programme provides the necessary conditions for retaining knowledge needed for ensuring the continuance of safe use of nuclear power, for developing new know-how and for participation in international co-operation. The SAFIR2014 Steering Group, responsible of the strategic alignements of the programme, consists of representatives of the Finnish Nuclear Safety Authority (STUK), Ministry of Employment and the Economy (MEE), Technical Research Centre of Finland (VTT), Teollisuuden Voima Oyj (TVO), Fortum Power and Heat Oy (Fortum), Fennovoima Oy, Lappeenranta University of Technology (LUT), Aalto University (Aalto), Finnish Funding Agency for Technology and Innovation (Tekes), Finnish Institute of Occupational Health (TTL) and the Swedish Radiation Safety Authority (SSM). The research programme is divided into nine areas: Man, organisation and society, Automation and control room, Fuel research and reactor analysis, Thermal hydraulics, Severe accidents, Structural safety of reactor circuits, Construction safety, Probabilistic risk analysis (PRA), and Development of research infrastructure. A reference group is assigned to each of these areas to respond for the strategic planning and to supervise the projects in its respective field. Research projects are selected annually based on a public call for proposals. Most of the projects are planned for the entire duration of the programme, but there can also be shorter one- or two-year projects. The annual volume of the SAFIR2014 programme in 2011-2012 has been 9,5-9,9 M euro. Main funding organisations were the State Nuclear Waste Management Fund (VYR) with over 5 M euro and

SAFIR2014. The Finnish Research Programme on Nuclear Power Plant Safety 2011-2014. Interim Report

International Nuclear Information System (INIS)

Simola, K.

2013-02-01

The Finnish Nuclear Power Plant Safety Research Programme 2011-2014, SAFIR2014, is a 4-year publicly funded national technical and scientific research programme on the safety of nuclear power plants. The programme is funded by the State Nuclear Waste Management Fund (VYR), as well as other key organisations operating in the area of nuclear energy. The programme provides the necessary conditions for retaining knowledge needed for ensuring the continuance of safe use of nuclear power, for developing new know-how and for participation in international co-operation. The SAFIR2014 Steering Group, responsible of the strategic alignements of the programme, consists of representatives of the Finnish Nuclear Safety Authority (STUK), Ministry of Employment and the Economy (MEE), Technical Research Centre of Finland (VTT), Teollisuuden Voima Oyj (TVO), Fortum Power and Heat Oy (Fortum), Fennovoima Oy, Lappeenranta University of Technology (LUT), Aalto University (Aalto), Finnish Funding Agency for Technology and Innovation (Tekes), Finnish Institute of Occupational Health (TTL) and the Swedish Radiation Safety Authority (SSM). The research programme is divided into nine areas: Man, organisation and society, Automation and control room, Fuel research and reactor analysis, Thermal hydraulics, Severe accidents, Structural safety of reactor circuits, Construction safety, Probabilistic risk analysis (PRA), and Development of research infrastructure. A reference group is assigned to each of these areas to respond for the strategic planning and to supervise the projects in its respective field. Research projects are selected annually based on a public call for proposals. Most of the projects are planned for the entire duration of the programme, but there can also be shorter one- or two-year projects. The annual volume of the SAFIR2014 programme in 2011-2012 has been 9,5-9,9 M euro. Main funding organisations were the State Nuclear Waste Management Fund (VYR) with over 5 M euro and

Final disposal of spent nuclear fuel in the Finnish bedrock

International Nuclear Information System (INIS)

1992-12-01

Teollisuuden Voima Oy (TVO) studies Finnish bedrock for the final disposal of the spent nuclear fuel from the Olkiluoto nuclear power plant. The study is in accordance with the decision in principle by Finnish government in 1983. The report is the summary of the preliminary site investigations carried out during the years 1987-1992. On the basis of these investigations a few areas will be selected for detailed site investigation. The characterization comprises five areas selected from the shortlist of potential candidate areas resulted in the earlier study during 1983-1985. Areas are located in different parts of Finland and they represent the main formations of the Finnish bedrock. Romuvaara area in Kuhmo and Veitsivaara area in Hyrynsalmi represent the Archean basement. Kivetty area in Konginkangas consists of mainly younger granitic rocks. Syyry in Sievi is located in transition area of Svecofennidic rocks and granitic rocks. Olkiluoto in Eurajoki represents migmatites in southern Finland. For the field investigations area-specific programs were planned and executed. The field investigations have comprised airborne survey by helicopter, geophysical surveys, geological mappings and samplings, deep and shallow core drillings, geophysical and hydrological borehole measurements and groundwater samplings

Ore potential of basic rocks in Finland

International Nuclear Information System (INIS)

Reino, J.; Ekberg, M.; Heinonen, P.; Karppanen, T.; Hakapaeae, A.; Sandberg, E.

1993-02-01

The report is associated with a study programme on basic rocks, which has the aim to complement the preliminary site investigations on repository for TVO's (Teollisuuden Voima Oy) spent nuclear fuel. The report comprises a mining enterprise's view of the ore potential of basic plutonic rocks in Finland. The ores associated with basic plutonic rocks are globally known and constitute a significant share of the global mining industry. The ores comprise chromium, vanadium-titanium-iron, nickel-copper and platinum group element ores. The resources of the metals in question and their mining industry are examined globally. A review of the use of these metals in the industry is presented as well. General factors affecting the mining industry, such as metal prices, political conjunctures, transport facilities, environmental requirements and raw material sources for the Finnish smelters have been observed from the point of view of their future effect on exploration activity and industrial development in Finland. Information on ores and mineralizations associated with Finnish basic rocks have been compiled in the report. The file comprises 4 chromium occurrences, 8 vanadium-titanium-iron occurrences, 13 PGE occurrences and 38 nickel-copper occurrences

Colloids or artefacts? A TVO/SKB cooperation project in Olkiluoto, Finland

International Nuclear Information System (INIS)

Laaksoharju, M.; Vuorinen, U.; Snellman, M.; Helenius, J.; Allard, B.; Pettersson, C.; Hinkkanen, H.

1993-12-01

TVO (Teollisuuden Voima Oy, Finland) initiated a co-operative task with SKB (Swedish Nuclear Fuel and Waste Management Co.) to critically evaluate colloid sampling methods at the test site in Olkiluoto, SW Finland. Three different colloid sampling methods were compared when sampling borehole OL-KR1 at 613-618 m depth. One possible way to make a conservative in-situ colloid estimation is to omit the contribution from calcite precipitation which is considered to be the main artefact. When this is made the inorganic colloid content (size 1-1000 nm) in Olkiluoto is 184 ±177 ppb consisting of clay minerals, silica, pyrite, goethite and magnesium oxide; the concentration of organic substances are around 100 ppb. The in-situ colloid concentration seems to be low which is in good agreement with experiences from years of sampling in similar environment and depths. The exercise shows the many difficulties encountered when sampling colloids. Small error in the planning, pump rate selection, a lack of precautionary measures, artefact sensitivity of the method etc have a tendency to affect significantly the results on the measured ppb colliod level

Review of the sorption of radionuclides on the bedrock of Haestholmen and on construction and backfill materials of a final repository for reactor wastes

International Nuclear Information System (INIS)

Kulmala, S.; Hakanen, M.

1992-10-01

Imatran Voima Oy (IVO) has plans to build a final repository for reactor wastes in the bedrock of the nuclear power plant site at Haestholmen, Loviisa. This report summarizes the sorption studies of radionuclides in Finnish bedrock performed at the Department of Radiochemistry, University of Helsinki. The values of mass distribution ratios, K d , and surface distribution ratios, K a ; of carbon, calsium, Zirconium, niobium, cobalt, nickel, strontium, cesium, uranium, plutonium, americium, thorium, chlorine, iodine and technetium are surveyed. Special attention is paid to the sorption data for construction and backfill materials of rector waste repository and the bedrock of Haestholmen. Safety assessment of a repository includes calculations of migration of the waste element in construction materials and backfill in the nearfield and in bedrock. Retardation by sorption of waste nuclides compared to groundwater flow is described by using distribution ratios between solid materials and water. (orig.)

Regression methodology in groundwater composition estimation with composition predictions for Romuvaara borehole KR10

Energy Technology Data Exchange (ETDEWEB)

Luukkonen, A.; Korkealaakso, J.; Pitkaenen, P. [VTT Communities and Infrastructure, Espoo (Finland)

1997-11-01

Teollisuuden Voima Oy selected five investigation areas for preliminary site studies (1987Ae1992). The more detailed site investigation project, launched at the beginning of 1993 and presently supervised by Posiva Oy, is concentrated to three investigation areas. Romuvaara at Kuhmo is one of the present target areas, and the geochemical, structural and hydrological data used in this study are extracted from there. The aim of the study is to develop suitable methods for groundwater composition estimation based on a group of known hydrogeological variables. The input variables used are related to the host type of groundwater, hydrological conditions around the host location, mixing potentials between different types of groundwater, and minerals equilibrated with the groundwater. The output variables are electrical conductivity, Ca, Mg, Mn, Na, K, Fe, Cl, S, HS, SO{sub 4}, alkalinity, {sup 3}H, {sup 14}C, {sup 13}C, Al, Sr, F, Br and I concentrations, and pH of the groundwater. The methodology is to associate the known hydrogeological conditions (i.e. input variables), with the known water compositions (output variables), and to evaluate mathematical relations between these groups. Output estimations are done with two separate procedures: partial least squares regressions on the principal components of input variables, and by training neural networks with input-output pairs. Coefficients of linear equations and trained networks are optional methods for actual predictions. The quality of output predictions are monitored with confidence limit estimations, evaluated from input variable covariances and output variances, and with charge balance calculations. Groundwater compositions in Romuvaara borehole KR10 are predicted at 10 metre intervals with both prediction methods. 46 refs.

Yritysmarkkinointisuunnitelma : -Case Kakadu Oy

OpenAIRE

Lamberg, Leena

2009-01-01

Opinnäytetyön aihe on yritysmarkkinointisuunnitelman kehittäminen tilasuunnitteluyritys Kakadu Oy:lle. Työn tavoitteena on segmentoida Kakadu Oy:n asiakaskunta ja määrittää tärkein asiakassegmentti sekä tehdä valitulle, potentiaalisimmalle asiakassegmentille kohdistettu markkinointisuunnitelma. Markkinointisuunnitelman tarkoitus on auttaa Kakadu Oy:tä suuntaamaan erittäin rajalliset resurssit mahdollisimman perustellusti ja tuottavasti. Yrityksessä ei ole aikaisemmin toteutettu markkinoin...

Nuclear power and Imatran Voima in the future

International Nuclear Information System (INIS)

Numminen, K.

1995-01-01

As the owner of the Loviisa NPS with two VVER-440 units, Imatran Voima (IVO) has worked with nuclear power for more than twenty years. After the negative decision of the Finnish Parliament in 1993 there are no possibilities to build nuclear power in Finland in the near future. However, the preparation work for increasing the produced power of all four operating NPP's of Finland is going on. The emphasis in the work with new nuclear energy is on the supporting programs in Eastern Europe and the preparation of a building contract of a new NPS to China together with the Russians. With a new decision of the Finnish Parliament, the nuclear option could still be an important part of the future energy strategy of Finland. (orig.)

AquaBuOY

DEFF Research Database (Denmark)

Weinstein, Alla; Fredrikson, Göran; Claeson, Lennart

2003-01-01

BuOY in five representative generic sea states. Ocean energy and offshore wave energy conversion in the United States is at a significant milestone. During the next year, ocean energy technology developers and energy officials have the potential to deploy pilot scale ocean power plants and transition......This paper describes development of the mathematical model simulating ocean performance of an offshore wave energy point absorber device-AquaBuOY. The AquaBuOY is the next generation of the technology, based on the IPS point absorber system and the hose pump, both of Sweden. AquaEnergy Group Ltd......, engineers, and developers can continue to lay the groundwork for government spending and interest in ocean energies....

Experimental testing of a SAG digital SILT application

International Nuclear Information System (INIS)

Haapanen, P.; Maskuniitty, M.; Heikkinen, J.; Korhonen, J.

1995-10-01

A prototype dynamic testing harness for programmable automation systems has been specified and implemented at the Technical Research Centre of Finland (VTT). In order to get experience on the methodology and equipment for the testing of systems important to the safety of nuclear power plants, where the safety and reliability requirements often are very high, two different pilot systems have been tested. One system was an ABB Master application, which was loaned for testing from ABB Atom by Teollisuuden Voima Oy (TVO). Another system, loaned from Siemens AG(SAG) by IVO International Oy (IVO), was an application realized with SAG's digital SILT technology. The report describes the testing of the SAG application. The purpose of the testing was not to assess the pilot system, but to get experience in the testing methodology and find out the further development needs and potentials of the test methodology and equipment. The experience show that dynamic testing is one feasible way to get more confidence about the safety and reliability of a programmable system that would be hard to achieve by other means. It also shows that more development of the test harness is still needed, especially concerning the comparison of the obtained test response to the expected response provided by the logical model of the system. Also the user interface of the on-line part of the test harness needs development. Methods for generation of the test cases also need further development eg. for achieving statistical significance for the reliability estimates. (10 refs., 90 figs., 9 tabs.)

Operation of Finnish nuclear power plants. Quarterly report 4th quarter, 1994 and annual summary

Energy Technology Data Exchange (ETDEWEB)

Tossavainen, K [ed.

1995-05-01

The Loviisa NPP units were in power operation the whole last quarter, with the exception of a reactor scram at Loviisa 1. The load factor average of all Finnish plant units was 100.2 %. The annual average was 90.0 %. All events in the fourth annual quarter were assigned level 0 (no safety significance) on the international INES scale. Four events in 1994 were classified level 1 (an anomaly). The Finnish Centre for Radiation and Nuclear Safety in December approved Imatran Voima Oy`s application to extend the operation of the reactor pressure vessel of Loviisa 2 until the annual maintenance outage of 2010. During this quarter, a batch of spent fuel from Loviisa power plant was transported to Russia. Occupational doses and radioactive releases off-site were below authorised limits. Only such quantities of plant-based radioactive materials were measurable in samples collected around the plants as have no bearing on the radiation exposure of the population. The report includes a summary of all the items described in the Quarterly Reports of 1994. (8 figs., 4 tabs.).

Customer database for Watrec Oy

OpenAIRE

Melnichikhina, Ekaterina

2016-01-01

This thesis is a development project for Watrec Oy. Watrec Oy is a Finnish company specializes in “waste-to-energy” issues. Customer Relation Management (CRM) strategies are now being applied within the company. The customer database is the first and trial step towards CRM strategy in Watrec Oy. The reasons for database project lie in lacking of clear customers’ data. The main objectives are: - To integrate the customers’ and project data; - To improve the level of sales and mar...

Workshop on iodine aspects of severe accident management. Summary and conclusions

International Nuclear Information System (INIS)

2000-03-01

Following a recommendation of the OECD Workshop on the Chemistry of Iodine in Reactor Safety held in Wuerenlingen (Switzerland) in June 1996 [Summary and Conclusions of the Workshop, Report NEA/CSNI/R(96)7], the CSNI decided to sponsor a Workshop on Iodine Aspects of Severe Accident Management, and their planned or effective implementation. The starting point for this conclusion was the realization that the consolidation of the accumulated iodine chemistry knowledge into accident management guidelines and procedures remained, to a large extent, to be done. The purpose of the meeting was therefore to help build a bridge between iodine research and the application of its results in nuclear power plants, with particular emphasis on severe accident management. Specifically, the Workshop was expected to answer the following questions: - what is the role of iodine in severe accident management? - what are the needs of the utilities? - how can research fulfill these needs? The Workshop was organized in Vantaa (Helsinki), Finland, from 18 to 20 May 1999, in collaboration with Fortum Engineering Ltd. It was attended by forty-six specialists representing fifteen Member countries and the European Commission. Twenty-eight papers were presented. These included four utility papers, representing the views of Electricite de France (EDF), Teollisuuden Voima Oy and Fortum Engineering Ltd (Finland), the Nuclear Energy Institute (USA), and Japanese utilities. The papers were presented in five sessions: - iodine speciation; - organic compound control; - iodine control; - modeling; - iodine management; A sixth session was devoted to a general discussion on iodine management under severe accident conditions. This report summarizes the content of the papers and the conclusions of the workshop

Seismic VSP and HSP surveys on preliminary investigation areas in Finland for final disposal of spent nuclear fuel

International Nuclear Information System (INIS)

Keskinen, J.; Cosma, C.; Heikkinen, P.

1992-10-01

Seismic reflection surveys in boreholes were carried out for Teollisuuden Voima Oy at five sites in Finland (Eurajoki Olkiluoto, Hyrynsalmi Veitsivaara, Konginkangas Kivetty, Kuhmo Romuvaara and Sievi Syyry). The vertical Seismic Profiling (VSP) surveys were a part of the investigation programme for the final disposal of spent nuclear fuel. The purpose was to detect fractured zones, lithological contacts and other anomalies in the structure of the rockmass and to determine their position and orientation. Horizontal Seismic Profiling (HSP) was used at the Olkiluoto site, additionally to VSP. The data has been organized in profiles containing seismograms recorded from the same shotpoint (shot gathers). One of the most powerful processing methods used with this project has been the Image Space Filtering, a new technique, which has been developed (in the project) for seismic reflection studies in crystalline rock. The method can be applied with other rock types where steeply inclined or vertical anomalies are of interest. It acts like a multichannel filter, enhancing the reflected events and also as an interpretation tool, to estimate the strength and position of the reflectors. This approach has been of great help in emphasizing the weak reflections from uneven and sometimes vanishing interfaces encountered in crystalline

Sähkösuunnittelu osana suunnitteluprojektia

OpenAIRE

Grönroos, Roope

2016-01-01

Insinöörityö tehtiin Sweco Industry Oy:n automaatio-osastolle. Työssä tutkittiin, kuinka teollisuuden sähkösuunnittelu liittyy suunnitteluprojekteissa muihin suunnittelualoihin isomman suunnittelukonsernin näkökulmasta. Alussa perehdytään muihin suunnittelualoihin ja siihen, kuinka ne ovat sidoksissa teollisuuden sähkösuunnitteluun. Alussa käydään läpi suunnitteluprojektin eri vaiheita ja mitä niihin sisältyy. Työtä varten haastateltiin eri suunnittelualojen kokeneita ammattil...

--- Oy:n tavoitteellinen markkinointi Facebookissa

OpenAIRE

Järventaus, Aino

2012-01-01

Tämä opinnäytetyö on laadittu toimeksiantona ohjelmatoimisto ---- Oy:lle. Toimeksiantosopimus syntyi opinnäytetyön tekijän ehdotuksesta kehittää --- Oy:n digitaalista markkinointia yrityksen toivomasta näkökulmasta. Työn fokukseksi valittiin Facebook koska sillä on merkittävä osa musiikkiteollisuuden markkinoinnissa. Toimeksiantajan Facebook-markkinoinnin hyödyntäminen ja ylläpito oli jäänyt liian vähälle huomiolle. Näin ollen kehittämiskohteeksi valittiin --- Oy:n Facebook-markkinointi ...

Markkinointiviestintäsuunnitelma : Classic Coffee Oy

OpenAIRE

Eerola, Laura

2015-01-01

Opinnäytetyön aiheena oli laatia markkinointiviestintäsuunnitelma kalenterivuodelle 2016 vuosikellon muodossa, toimintansa jo vakiinnuttaneelle Classic Coffee Oy:lle. Classic Coffee Oy on vuonna 2011 perustettu, Tampereella toimiva kahvila-alan yritys joka tarjoaa lounaskahvilatoiminnan lisäksi laadukkaita konditoria-palveluita, yritys- ja kokoustarjoiluja sekä tilavuokrausta. Classic Coffee Oy:llä on yksi kahvila, Classic Coffee Tampella. Kahvila sijaitsee Tampellassa, Tampereen keskustan vä...

Protecting the endangered lake salmon

International Nuclear Information System (INIS)

Soimakallio, H.; Oesch, P.

1997-01-01

In addition to the Ringed Seal, the labyrinthine Saimaa lake system created after the Ice Age also trapped a species of salmon, whose entire life cycle became adapted to fresh water. In order to improve the living conditions of this lake salmon which - like the ringed seal - is today classified as an endangered species, an intensive research programme has been launched. The partners include the Finnish Game and Fisheries Research Institute, fishing and environmental authorities and - in collaboration with UPM-Kymmene Oy and Kuurnan Voima Oy - the IVO subsidiary Pamilo Oy

Protecting the endangered lake salmon

Energy Technology Data Exchange (ETDEWEB)

Soimakallio, H.; Oesch, P. [ed.

1997-11-01

In addition to the Ringed Seal, the labyrinthine Saimaa lake system created after the Ice Age also trapped a species of salmon, whose entire life cycle became adapted to fresh water. In order to improve the living conditions of this lake salmon which - like the ringed seal - is today classified as an endangered species, an intensive research programme has been launched. The partners include the Finnish Game and Fisheries Research Institute, fishing and environmental authorities and - in collaboration with UPM-Kymmene Oy and Kuurnan Voima Oy - the IVO subsidiary Pamilo Oy

Evaluation of incident analysis practices in the Finnish nuclear power industry

Energy Technology Data Exchange (ETDEWEB)

Kettunen, J.; Laakso, K

1999-12-01

This report provides an analysis and evaluation of incident analysis methods and practices applied by the Finnish regulator Radiation and Nuclear Safety Authority (STUK) and the two Finnish nuclear power plant operators Teollisuuden Voima Oy (TVO) and Fortum Power and Heat Oy (Fortum). The study was conducted in 1998-99. The research material was based on tape-recorded interviews as well as internal directions and event investigation reports provided by the three participating organisations. A framework for analysis and evaluation was developed as part of the study on the basis of referenced root cause analysis and operating experience review methods, selected (foreign) inspection reports, scientific papers and research literature. Well-known inspection methods and principles, such as ASSET and MTO/HPES, provided important guidance to this work. This study shows that although all the evaluated organisations had rather comprehensive incident analysis arrangements, more focus and priorisation is needed. Deficiencies were identified mostly in the areas of recording, assessment and classification of new events and observations, use of existing operating experience data, utilisation of information technology based tools, and allocation of work and resources. In general the direct causes of identified events can be detected and removed, but more emphasis should be given to the prevention of recurrence. This requires a more efficient feedback loop that can be created and maintained by focusing on the root causes of significant events, tasks and activities in which the originating errors occurred, and weaknesses of defensive barriers, and by implementing periodic operational experience reviews. A strategy document for the operating experience feedback process, and firm procedures for the initial assessment of new events and the carrying out of data analyses would help. (orig.)

Synthesis and photoluminescence properties of CaSixOy:Tb3+ phosphors prepared using solution-combustion method

CSIR Research Space (South Africa)

Dejene, FB

2011-07-01

Full Text Available Effect of Tb3+ ion concentrations on the structural and persistence luminescence properties of CaSixOy:Tb3+ crystals were evaluated using powders grown by the solution combustion technique. The XRD study indicates the change of phase from CaSiO3...

Geological and geophysical investigations in the selection and characterization of the disposal site for high-level nuclear waste in Finland

Energy Technology Data Exchange (ETDEWEB)

Paulamaki, S.; Paananen, M.; Kuivamaki, A. [Geological Survey of Finland, Espoo (Finland); Wikstrom, L. [Posiva Oy, Olkiluoto (Finland)], e-mail: seppo.paulamaki@gtk.fi

2011-07-01

Two power companies, Teollisuuden Voima Oy (TVO) and Fortum Power and Heat Oy, are preparing for the final disposal of spent nuclear fuel deep in the Finnish bedrock. In the initial phase of the site selection process in the late 1970s and early 1980s, the Geological Survey of Finland (GTK) examined the general bedrock factors that would have to be taken into account in connection with final disposal with reference to the international guidelines adapted to Finnish conditions. On the basis of extensive basic research data, it was concluded that it is possible to find a potential disposal site that fulfils the geological safety criteria. In the subsequent site selection survey covering the whole of Finland, carried out by GTK in 1983-1985, 101 potential investigation areas were discovered. Eventually, five areas were selected by TVO for preliminary site investigations: Romuvaara and Veitsivaara in the Archaean basement complex, Kivetty and Syyry in the Proterozoic granitoid area, and Olkiluoto (TVO's NPP site) in the Proterozoic migmatite area. The preliminary site investigations at the selected sites in 1987-1992 comprised deep drillings together with geological, geophysical, hydrogeological and hydrogeochemical investigations. A conceptual geological bedrock model was constructed for each site, including lithology, fracturing, fracture zones and hydrogeological conditions. On the basis of preliminary site investigations, TVO selected Romuvaara, Kivetty and Olkiluoto for detailed site investigations to be carried out during 1993-2000. After the feasibility studies, the island of Haestholmen, where Fortum's Loviisa nuclear power plant is located, was added to the list of potential disposal sites. In the detailed site investigations, additional data on bedrock were gathered, the previous conceptual geological, hydrogeological and hydrogeochemical models were complemented, the rock mechanical properties of the bedrock were examined, and the constructability

Results of monitoring at Olkiluoto in 2009. Environment

Energy Technology Data Exchange (ETDEWEB)

Haapanen, A. (ed.) (Haapanen Forest Consulting, Vanhakylae (Finland))

2010-10-15

This Working Report presents the main results of Posiva Oy's environmental monitoring programme on Olkiluoto Island in 2009. These summary reports have been published since 2005. The environmental monitoring system supervised by Posiva Oy produces input for biosphere modelling for long-term safety purposes as well as for monitoring the state of the environment during the construction (and later operation) of ONKALO underground characterization facility. Part of the monitoring is performed by the company running the nuclear power plants on the island, Teollisuuden Voima Oy (TVO). Monitoring has been carried out for varying periods of time depending on the sector: some monitoring activities performed by TVO originate from the 1970s and the repository-related environmental monitoring of Olkiluoto from the early 2000s. The monitoring programme evolves according to the experiences gained from the modelling work and an increased understanding of the site. Augmentations in 2009 include e.g. establishment of a new forest intensive monitoring plot (FIP14), continuation of studies on fine roots and on the species composition and abundances of small mammals. Line transect samplings of ants, terrestrial snails and earthworms were carried out and a systematic monitoring of island birds was started. In addition, a project was started where the sediment load and factors affecting the sediment transportation into Eurajoensalmi bay is examined. Dust produced during construction of the third nuclear power unit (OL3), ONKALO and related infrastructure can be seen in the soil solution and deposition results. Furthermore, the construction works and road traffic have a raising effect on the noise levels of the immediate surroundings. The land-use continues to change, but the remaining natural environment resembles other coastal locations. The young age of the soils and the closeness of the sea are reflected in the soil properties. Mammalian fauna on the island is typical of coastal

Results of monitoring at Olkiluoto in 2009. Environment

International Nuclear Information System (INIS)

Haapanen, A.

2010-10-01

This Working Report presents the main results of Posiva Oy's environmental monitoring programme on Olkiluoto Island in 2009. These summary reports have been published since 2005. The environmental monitoring system supervised by Posiva Oy produces input for biosphere modelling for long-term safety purposes as well as for monitoring the state of the environment during the construction (and later operation) of ONKALO underground characterization facility. Part of the monitoring is performed by the company running the nuclear power plants on the island, Teollisuuden Voima Oy (TVO). Monitoring has been carried out for varying periods of time depending on the sector: some monitoring activities performed by TVO originate from the 1970s and the repository-related environmental monitoring of Olkiluoto from the early 2000s. The monitoring programme evolves according to the experiences gained from the modelling work and an increased understanding of the site. Augmentations in 2009 include e.g. establishment of a new forest intensive monitoring plot (FIP14), continuation of studies on fine roots and on the species composition and abundances of small mammals. Line transect samplings of ants, terrestrial snails and earthworms were carried out and a systematic monitoring of island birds was started. In addition, a project was started where the sediment load and factors affecting the sediment transportation into Eurajoensalmi bay is examined. Dust produced during construction of the third nuclear power unit (OL3), ONKALO and related infrastructure can be seen in the soil solution and deposition results. Furthermore, the construction works and road traffic have a raising effect on the noise levels of the immediate surroundings. The land-use continues to change, but the remaining natural environment resembles other coastal locations. The young age of the soils and the closeness of the sea are reflected in the soil properties. Mammalian fauna on the island is typical of coastal

Geological and geophysical investigations in the selection and characterization of the disposal site for high-level nuclear waste in Finland

Energy Technology Data Exchange (ETDEWEB)

Paulamaki, S; Paananen, M; Kuivamaki, A [Geological Survey of Finland, Espoo (Finland); Wikstrom, L. [Posiva Oy, Olkiluoto (Finland)], e-mail: seppo.paulamaki@gtk.fi

2011-07-01

Two power companies, Teollisuuden Voima Oy (TVO) and Fortum Power and Heat Oy, are preparing for the final disposal of spent nuclear fuel deep in the Finnish bedrock. In the initial phase of the site selection process in the late 1970s and early 1980s, the Geological Survey of Finland (GTK) examined the general bedrock factors that would have to be taken into account in connection with final disposal with reference to the international guidelines adapted to Finnish conditions. On the basis of extensive basic research data, it was concluded that it is possible to find a potential disposal site that fulfils the geological safety criteria. In the subsequent site selection survey covering the whole of Finland, carried out by GTK in 1983-1985, 101 potential investigation areas were discovered. Eventually, five areas were selected by TVO for preliminary site investigations: Romuvaara and Veitsivaara in the Archaean basement complex, Kivetty and Syyry in the Proterozoic granitoid area, and Olkiluoto (TVO's NPP site) in the Proterozoic migmatite area. The preliminary site investigations at the selected sites in 1987-1992 comprised deep drillings together with geological, geophysical, hydrogeological and hydrogeochemical investigations. A conceptual geological bedrock model was constructed for each site, including lithology, fracturing, fracture zones and hydrogeological conditions. On the basis of preliminary site investigations, TVO selected Romuvaara, Kivetty and Olkiluoto for detailed site investigations to be carried out during 1993-2000. After the feasibility studies, the island of Haestholmen, where Fortum's Loviisa nuclear power plant is located, was added to the list of potential disposal sites. In the detailed site investigations, additional data on bedrock were gathered, the previous conceptual geological, hydrogeological and hydrogeochemical models were complemented, the rock mechanical properties of the bedrock were examined, and the constructability and the

Summary report on groundwater chemistry

International Nuclear Information System (INIS)

Lampen, P.; Snellman, M.

1993-07-01

The preliminary site investigations for radioactive waste disposal (in Finland) carried out by Teollisuuden Voima Oy (TVO) during the period 1987 to 1992 yielded data on hydrogeochemistry from a total 337 water samples. The main objective of the groundwater chemistry studies was to characterize groundwaters at the investigation sites and, specifically, to create a concept for the mean residence times and evolution of groundwater by means of isotopic analyses. Moreover, the studies yielded input data for geochemical modelling and the performance assessment. Samples were taken from deep boreholes (with a depth of 500 to 1000 m), percussion-drilled boreholes (depth approx. 200 m), flushing-water wells (approx. 100 m) and multi-level pietzometers (approx. 100 m) used in the hydrological tests. The water used for drilling the deep boreholes was taken from local flushing-water wells, whose water was also analyzed in detail. The flushing water used in drilling was marked with two tracers, iodine and uranine, analyzed with two different methods. For reference purposes, samples were also taken from surficial and groundwaters over a large area surrounding the investigation site. Precipitation over a period of at least one year was collected at all the five investigation sites and the samples were analyzed in great detail, particularly with regard to isotopes. Similarly, snow profile samples representing precipitation during the entire winter was taken from each site at least once

ROCK-CAD - computer aided geological modelling system

International Nuclear Information System (INIS)

Saksa, P.

1995-12-01

The study discusses surface and solid modelling methods, their use and interfacing with geodata. Application software named ROCK-CAD suitable for geological bedrock modelling has been developed with support from Teollisuuden Voima Oy (TVO). It has been utilized in the Finnish site characterization programme for spent nuclear fuel waste disposal during the 1980s and 1990s. The system is based on the solid modelling technique. It comprises also rich functionality for the particular geological modelling scheme. The ROCK-CAD system provides, among other things, varying graphical vertical and horizontal intersections and perspective illustrations. The specially developed features are the application of the boundary representation modelling method, parametric object generation language and the discipline approach. The ROCK-CAD system has been utilized in modelling spatial distribution of rock types and fracturing structures in TVO's site characterization. The Olkiluoto site at Eurajoki serves as an example case. The study comprises the description of the modelling process, models and illustration examples. The utilization of bedrock models in site characterization, in tentative repository siting as well as in groundwater flow simulation is depicted. The application software has improved the assessment of the sites studied, given a new basis for the documentation of interpretation and modelling work, substituted hand-drawing and enabled digital transfer to numerical analysis. Finally, aspects of presentation graphics in geological modelling are considered. (84 refs., 30 figs., 11 tabs.)

Assessment of health risks brought about by transportation of spent fuel

International Nuclear Information System (INIS)

Suolanen, V.; Lautkaski, R.; Rossi, J.

1999-03-01

In the study health risks caused by transportation of spent fuel from Olkiluoto and from Loviisa NPP's to the planned disposal site have been evaluated. The Olkiluoto NPP is owned by Teollisuuden Voima Oy (TVO) and the Loviisa NPP, situated at Haestholmen, by Fortum Power and Heat Oy. According to the base scenario of 40 years use of the current NPP's the total amount of spent fuel will be 1840 tU (TVO) and 860 tU (Fortum). Annually, 110 tU on the average and at most 250 tU will be transported to the disposal site. The considered transportation routes are from Olkiluoto to Haestholmen, from Olkiluoto to Kivetty, from Olkiluoto to Romuvaara, from Haestholmen to Olkiluoto, from Haestholmen to Kivetty and from Haestholmen to Romuvaara. The considered transportation modes are truck, rail or ship, or combinations of these modes. Each transportation route has been divided into homogenised sequences with respect to population density and/or route type. Total amount of analysed route options were 40, some route sequences are overlapping. Radiation exposures to the population along the routes have been calculated in normal, incident and accident situations during transportation. Occupational radiation doses to the personnel have been estimated for normal transportation only. The consequences of normal transportation have been evaluated based on RADTRAN-model, developed by the Sandia National Laboratories. As incidents, stopping of spent fuel transportation for an exceptionally long period of time, and in another case contamination of outer surface of spent fuel cask have been considered. Expected collective doses and health risks of transportation accidents connected to the routes have been calculated with RADTRAN-model. Single hypothetical transport accidents with pessimistic release assumptions have been further analysed in more detail with the ARANO-model, developed by VTT (Technical Research Centre of Finland). (orig.)

IAEA Mission Sees Safety Commitment at Finland's New Olkiluoto Reactor Before Planned Start in December

International Nuclear Information System (INIS)

2018-01-01

An International Atomic Energy Agency (IAEA) team of experts observed a commitment to safety by the operator of Unit 3 at Finland’s Olkiluoto Nuclear Power Plant, ahead of the Evolutionary Pressurised Water Reactor’s (EPR) planned connection to the grid in December. The team also identified areas for further enhancements as the operator prepares to put the reactor online. The Pre-Operational Safety Review Team (Pre-OSART) concluded an 18-day mission today to assess operational safety at the 1600 MW reactor, located about 280 km northwest of the capital, Helsinki. Finland has engaged France’s Areva SA together with Germany’s Siemens to construct and commission the unit. The operator is Teollisuuden Voima (TVO). Pre-OSART missions aim to improve operational safety by objectively assessing safety performance using the IAEA’s safety standards and proposing recommendations for improvement where appropriate. The review covered the areas of leadership and management for safety; training and qualification; operations; maintenance; technical support; operating experience; radiation protection; chemistry; emergency preparedness and response; accident management; and commissioning. The team identified a number of good practices that will be shared with the nuclear industry globally, including: • The plant has developed and implemented an efficient system for improving knowledge and skills of staff members. • The plant has developed and validated a unique method for performing suspended solids analysis using a microscope, imaging software and a digital camera. • The plant has introduced a system for systematically assessing nuclear safety culture in the plant supplier organization during construction and commissioning. The mission made several recommendations to improve operational safety, including: • Plant management should set appropriate expectations, communicate them to staff and reinforce them in the field. • The plant should improve the

HPK Liiga Oy:n Facebook-markkinoinnin mittaaminen

OpenAIRE

Anttila, Pasi

2016-01-01

Opinnäytetyön toimeksiantaja on HPK Liiga Oy. Työn taustalla toimeksiantajan osalta oli lähteä kehittämään sosiaalisen median mittaamista, ja opinnäytetyön tekijä oli myös kiinnostunut tämän asian tutkimisesta. Taustalla olivat myös HPK Liiga Oy:n strategiset muutokset sekä yhtiöittäminen rekisteröidystä yhdistyksestä osakeyhtiöksi. Työn tavoitteena oli selvittää, millä mittareilla HPK Liiga Oy:n kannattaisi mitata Facebook-markkinointiaan ydintavoitteiden ollessa asiakkaiden sitouttamin...

Structural and superconducting properties of YBa2Cu3-xMxOy (M=Ag, Al

Directory of Open Access Journals (Sweden)

S Falahati

2009-08-01

Full Text Available   Samples of YBa2Cu3-xAgxOy with x=0, 0.1, 0.15, 0.2, 0.3 and samples of YBa2Cu3-xAlxOy with x=0, 0.01, 0.02, 0.03 and 0.045 are prepared by the sol-gel method. Structural and superconducting properties of samples are studied by electrical resistivity (R-T, X-ray diffraction (XRD and scanning electron microscopy (SEM. All the samples show transition to superconducting state and the transition temperatures of the samples increased with increasing Ag doping up to x=0.15. R-T measurements show a small decrease of TC (zero with increasing Al doping up to x=0.02, and followed by a faster decrease with increasing doping concentration. YBCO grains are better linked with increasing Ag doping. So, Ag has positive effects in superconducting properties of YBCO. The crystal structure of samples was refined by MAUD. These results show tha, Ag is substituted for Cu(1 in YBCO. According to these analysis, we introduce x=0.15 as the optimum value for doping concentration .

Social media marketing communication plan for Hauskafe Oy

OpenAIRE

Uzunova, Aleksandra; Franko, Jan

2017-01-01

The case company of this project-based thesis is Hauskafe Oy, an SME company located in Espoo, Finland. Following examples of best practices in the field, Hauskafe Oy recognised the need of designing a social media marketing plan that will allow to build the brand, improve customer loyalty and as a final goal – increase the sales. A social media marketing communication plan for Hauskafe Oy is the outcome of this thesis. The theoretical framework of the thesis is a desktop study that discu...

Fuel reliability experience in Finland

International Nuclear Information System (INIS)

Kekkonen, L.

2015-01-01

Four nuclear reactors have operated in Finland now for 35-38 years. The two VVER-440 units at Loviisa Nuclear Power Plant are operated by Fortum and two BWR’s in Olkiluoto are operated by Teollisuuden Voima Oyj (TVO). The fuel reliability experience of the four reactors operating currently in Finland has been very good and the fuel failure rates have been very low. Systematic inspection of spent fuel assemblies, and especially all failed assemblies, is a good practice that is employed in Finland in order to improve fuel reliability and operational safety. Investigation of the root cause of fuel failures is important in developing ways to prevent similar failures in the future. The operational and fuel reliability experience at the Loviisa Nuclear Power Plant has been reported also earlier in the international seminars on WWER Fuel Performance, Modelling and Experimental Support. In this paper the information on fuel reliability experience at Loviisa NPP is updated and also a short summary of the fuel reliability experience at Olkiluoto NPP is given. Keywords: VVER-440, fuel reliability, operational experience, poolside inspections, fuel failure identification. (author)

Developing and design a website for mc kalla oy

OpenAIRE

Bekele, Henok

2013-01-01

This bachelor thesis is about Website development and design. I have a chance to work with Mc kalla Oy. Mc kalla Oy is a construction company from Kempele which was founded 2011. They have projects in Central-Finland, through Northern Finland to Lapland. This thesis is to develop and design a new website to Mc kalla Oy. Wordpress is used to develop the new website. For the development process I use school server (.opiskelijaprojektit.net). The Thesis contains two main parts designing the ...

Sunray project - A long-term, nationwide educational process

International Nuclear Information System (INIS)

Nikula, Anneli

2000-01-01

The Sunray project is a nationwide educational process coordinated by the Economic Information Bureau (TaT Group) for ninth graders in Finnish comprehensive schools. The project aims at giving thorough and versatile information on radiation within the framework of various subjects (physics, biology, domestic science, history, European languages, mother tongue, health education etc.). The Sunray project covers all ninth graders of the existing 600 Finnish comprehensive schools; in all involving some 65 000 pupils. The project, which has been repeated five times, was initiated as part of the European Science and Technology week in 1995. During the first two years it was strongly linked with the science week as natural sciences were seen as a good framework for the chosen perspective. Since 1997, the project has been run as an event in its own right. The project has applied the method of processing integrated groups of themes, which is an objective of the comprehensive school system and the experimental method of science. As schools make their own decisions about the educational programmes to be adopted every semester, the project has been marketed to schools at the beginning of May. The TaT Group has arranged marketing events in some 10 localities in Finland. The Economic Information Bureau of Finland coordinates the project and in 1995-2000 the Radiation and Nuclear Safety Authority (STUK), the Finnish Energy Industries' Federation, the Finnish Electricity Association, Fortum Oyj and Teollisuuden Voima Oy have participated in the project

Summary report of the experiences from TVO's site investigations

International Nuclear Information System (INIS)

Oehberg, A.; Saksa, P.; Ahokas, H.; Ruotsalainen, P.; Snellman, M.

1994-05-01

Teollisuuden Voima Oy (TVO) has completed preliminary site investigations at five sites in Finland. At the end of 1992 TVO presented the final report to the authorities. The preliminary site investigation phase 1986-1992 was conducted according to the investigation programme compiled by TVO. The aim of this report was to compile a report on experiences from TVOs site investigations. The main interest was focused on investigation strategies and the most important investigation methods for the conceptual modelling. The objective of the preliminary site investigations was to obtain data on the bedrock properties in order to evaluate the areas. The programme was divided into four stages, each stage having its own sub-objective. The site-specific investigation programme for each site included a large common part and a small site-specific part. The strategies (objectives) and experiences from different disciplines, geology, hydrogeochemistry, geophysics and geohydrology, are presented in the report. The conceptual modelling work procedure including both bedrock and groundwater modelling is described briefly using the Olkiluoto site as an example. Each of the other areas has undergone similar phases of work. The uncertainties associated with conceptual modelling are also discussed. The usefulness of the investigation strategy and the investigation methods for conceptual modelling is discussed in the report. Some new equipment, methods or enhancements that have not yet been used in TVOs site investigations have become new tools in site characterisation and are briefly presented in the report. 52 refs, 35 figs, 1 tab

Bedrock model of the Romuvaara area

International Nuclear Information System (INIS)

Saksa, P.; Paananen, M.; Paulamaeki, S.; Anttila, P.; Pitkaenen, P.; Front, K.; Vaittinen, T.

1992-05-01

Site for the final disposal of the spent nuclear fuel investigations implemented in accordance with the research programme drawn up by Teollisuuden Voima Oy were carried out at Romuvaara, Kuhmo, in 1987 - 1991. Model of the site were compiled and used for describing the rock types, fractures, fracturing structures and geohydrological conditions, the main emphasis being on the examination of the bedrock fracturing and related hydraulic conductivity. Three-dimensional models were used for the classification of the various properties of the bedrock structures. The descriptive models were gathered together in a computerized system to facilitate illustration and strage. The rock types at Romuvaara are gneiss, mica gneiss, leucotonalite gneiss, amphibolite, granodiorite and metadiabase. The structural model for fracturing at the site contains 19 zones described in terms of a number of properties. The fracturing observed at Romuvaara ranges from local occurences of dence fracturing to significant, altered fracture zones. The structural model includes deduced values for hydraulic conductivity, deduced points of flow in the boreholes and measured hydraulic heads.Various classifications were used for assessment of hydraulic conductivity in the zones and solid bedrock, and in both cases conductivity was found to diminish with depth. Measured hydraulic heads were mostly found to support structural interpretation. The results were used for estimation of a three-dimensional hydraulic head distribution. Results from pumping tests carried out in the significant flow zone support the geometric interpretation

In-situ experiments to investigate rock matrix retention properties in ONKALO, Olkiluoto, Finland

Energy Technology Data Exchange (ETDEWEB)

Voutilainen, Mikko; Helariutta, Kerttuli [Helsinki Univ. (Finland). Dept. of Chemistry; Poteri, Antti [Technical Research Centre of Finland VTT (Finland); and others

2015-07-01

Spent nuclear fuel from nuclear power plants, owned by TVO (Teollisuuden Voima Oy) and Fortum, is planned to be disposed to a repository at a depth of more than 400 meters in the bedrock of Olkiluoto (Eurajoki, Finland). The repository system of multiple release barriers consists of both manmade and natural barriers. The surrounding rock acts as the last barrier if other barriers fail during passage of the millennia. Therefore, safe disposal of spent nuclear fuel requires information on the radionuclide transport and retention properties within the porous and water-containing rock matrix along the water conducting flow paths. To this end, various types of experiments are being performed and planned within ONKALO, the underground rock characterization facility in Olkiluoto, as part of the project @''rock matrix REtention PROperties'' (REPRO). The research site is located at a depth of 420 meters close to the repository site. The aim is to study the diffusion and sorption properties of nuclear compounds in the rock matrix under real in-situ conditions. The first in-situ experiment was performed during 2012 using HTO, Na-22, Cl-36 and I-125 as tracer nuclides. Breakthrough curves show retention and asymptotic behavior that are in-line with those caused by matrix diffusion and sorption were observed in their breakthrough curves. Weak sorption was also observed in the breakthrough curves of Na-22 and I-125.

Results of monitoring at Olkiluoto in 2010 - Environment

Energy Technology Data Exchange (ETDEWEB)

Haapanen, A. (ed.) [Haapanen Forest Consulting, Vanhakylae (Finland)

2011-10-15

This Working Report presents the main results of Posiva Oy's environmental monitoring programme on Olkiluoto Island in 2010. These summary reports have been published since 2005. The environmental monitoring system supervised by Posiva Oy produces input for biosphere modelling for long-term safety purposes as well as for monitoring the state of the environment during the construction (and later operation) of ONKALO underground rock characterization facility. Part of the monitoring is performed by the company running the nuclear power plants on the island, Teollisuuden Voima Oy (TVO). Monitoring has been carried out for varying periods of time depending on the sector: some monitoring activities performed by TVO originate from the 1970s and the repository-related environmental monitoring of Olkiluoto from the early 2000s. The monitoring programme evolves according to experiences gained from the modelling work and increased understanding of the site. Augmentations in 2010 include one previously unmonitored private drilled well, and sampling of crop plants, aquatic macrophytes, and bottom fauna, as well as soil and water in order to obtain more data on site-specific concentration ratios. In addition to Olkiluoto Island, two so called reference lakes have been included in the sampling. Studies have been going on on one reference mire, as well. Bottom fauna studies of River Eurajoki exist from late 1970s, but have not been presented here before. Dust produced during construction of the third nuclear power unit (OL3), ONKALO and related infrastructure can be seen in the analysis results of needle litter. The construction works and road traffic have a raising effect on the noise levels of the immediate surroundings. The land-use continues to change, but the remaining natural environment resembles other coastal locations. The young age of the soils and the closeness of the sea are reflected in the soil properties. Mammalian fauna on the island is typical of coastal

Results of monitoring at Olkiluoto in 2010 - Environment

International Nuclear Information System (INIS)

Haapanen, A.

2011-10-01

This Working Report presents the main results of Posiva Oy's environmental monitoring programme on Olkiluoto Island in 2010. These summary reports have been published since 2005. The environmental monitoring system supervised by Posiva Oy produces input for biosphere modelling for long-term safety purposes as well as for monitoring the state of the environment during the construction (and later operation) of ONKALO underground rock characterization facility. Part of the monitoring is performed by the company running the nuclear power plants on the island, Teollisuuden Voima Oy (TVO). Monitoring has been carried out for varying periods of time depending on the sector: some monitoring activities performed by TVO originate from the 1970s and the repository-related environmental monitoring of Olkiluoto from the early 2000s. The monitoring programme evolves according to experiences gained from the modelling work and increased understanding of the site. Augmentations in 2010 include one previously unmonitored private drilled well, and sampling of crop plants, aquatic macrophytes, and bottom fauna, as well as soil and water in order to obtain more data on site-specific concentration ratios. In addition to Olkiluoto Island, two so called reference lakes have been included in the sampling. Studies have been going on on one reference mire, as well. Bottom fauna studies of River Eurajoki exist from late 1970s, but have not been presented here before. Dust produced during construction of the third nuclear power unit (OL3), ONKALO and related infrastructure can be seen in the analysis results of needle litter. The construction works and road traffic have a raising effect on the noise levels of the immediate surroundings. The land-use continues to change, but the remaining natural environment resembles other coastal locations. The young age of the soils and the closeness of the sea are reflected in the soil properties. Mammalian fauna on the island is typical of coastal areas in

KUSTANNUS JA KANNATTAVUUSLASKENNAN KEHITTÄMINEN CASE KOTEK FACTORY SERVICE OY

OpenAIRE

Ilmonen, Hanna

2009-01-01

Opinnäytetyön toimeksiantajana oli Kotek Factory Service Oy ja kohteena oli Kotek Factory Service Oy:n maalaamo. Kotek Factory Service on Chesterton ARC- pinnoitteiden maahantuoja ja myyjä, joka vastaa tuotteidensa erilaisista urakoinneista, täydellisinä palvelukokonaisuuksina. Opinnäytetyön tavoitteena oli saada aikaiseksi toimintaehdotus siitä, kuinka Kotek Factory Service Oy voisi kehittää kustannus ja kannattavuuslaskentaansa maalaamon ja koko yrityksen osalta. Ongelmaan haettiin ratk...

Mortality, fertility, and the OY ratio in a model hunter-gatherer system.

Science.gov (United States)

White, Andrew A

2014-06-01

An agent-based model (ABM) is used to explore how the ratio of old to young adults (the OY ratio) in a sample of dead individuals is related to aspects of mortality, fertility, and longevity experienced by the living population from which the sample was drawn. The ABM features representations of rules, behaviors, and constraints that affect person- and household-level decisions about marriage, reproduction, and infant mortality in hunter-gatherer systems. The demographic characteristics of the larger model system emerge through human-level interactions playing out in the context of "global" parameters that can be adjusted to produce a range of mortality and fertility conditions. Model data show a relationship between the OY ratios of living populations (the living OY ratio) and assemblages of dead individuals drawn from those populations (the dead OY ratio) that is consistent with that from empirically known ethnographic hunter-gatherer cases. The dead OY ratio is clearly related to the mean ages, mean adult mortality rates, and mean total fertility rates experienced by living populations in the model. Sample size exerts a strong effect on the accuracy with which the calculated dead OY ratio reflects the actual dead OY ratio of the complete assemblage. These results demonstrate that the dead OY ratio is a potentially useful metric for paleodemographic analysis of changes in mortality and mean age, and suggest that, in general, hunter-gatherer populations with higher mortality, higher fertility, and lower mean ages are characterized by lower dead OY ratios. Copyright © 2014 Wiley Periodicals, Inc.

PUHELINPALVELUN FACEBOOK-MAINOS : Case: Harmonia Care Oy

OpenAIRE

Tirkkonen, Mari

2014-01-01

Opinnäytetyön toimeksiantaja oli Harmonia Care Oy. Tutkimuksessa käytettiin kvantitatiivista tutkimusmenetelmää. Opinnäytetyö oli luonteeltaan toiminnallinen opinnäytetyö, jossa samalla suunniteltiin, toteutettiin ja kehitettiin yrityksen sähköistä markkinointia. Tutkimuksessa haettiin vastausta seuraaviin kysymyksiin: ” Onko Facebook-mainonnasta hyötyä Harmonia Care Oy:lle?” ja ”Millainen markkinointikampanja tukee yrityksen strategiaa parhaiten?” Opinnäytetyössä suunniteltiin ja toteute...

Ilmanpuhdistimen tuotekehitys : case: Mecastep Oy

OpenAIRE

Arasola, Aku

2015-01-01

Tämän opinnäytetyön tarkoituksena oli olla osana tuotekehitysprojektia, joka koskee Mecastep Oy:n ilmanpuhdistin innovaatiota. Kehitystä vaativia kohtia olivat komponenttien valinta ja mekaniikan saaminen toimivaksi. Edelläkävijän innovaatiosta tekee se, että samanlaista tuotetta ei ole vielä olemassa. Mekaniikan mallintaminen tapahtui 3D-ohjelmalla nimeltä KeyCreator. Ilmanpuhdistin tullaan sijoittamaan keittiön jätekaappiin. Puhdistuksella tässä tapauksessa tarkoitetaan epäpuhtaan ilman...

Condensation driven water hammer studies for feedwater distribution pipe

Energy Technology Data Exchange (ETDEWEB)

Savolainen, S.; Katajala, S.; Elsing, B.; Nurkkala, P.; Hoikkanen, J. [Imatran Voima Oy, Vantaa (Finland); Pullinen, J. [IVO Power Engineering Ltd., Vantaa (Finland); Logvinov, S.A.; Trunov, N.B.; Sitnik, J.K. [EDO Gidropress (Russian Federation)

1997-12-31

Imatran Voima Oy, IVO, operates two VVER 440 reactors. Unit 1 has been operating since 1977 and unit 2 since 1981. First damages of the feed water distribution (FWD) pipes were observed in 1989. In closer examinations FWD-pipe T-connection turned out to suffer from severe erosion corrosion damages. Similar damages have been found also in other VVER 440 type NPPs. In 1994 the first new FWD-pipe was replaced and in 1996 extensive water hammer experiments were carried out together with EDO Gidropress in Podolsk. After the first phase of the experiments some fundamental changes were made to the construction of the FWD-pipe. The object of this paper is to give short insight to the design of the new FWD-pipe concentrating on water hammer experiments. (orig.).

Condensation driven water hammer studies for feedwater distribution pipe

International Nuclear Information System (INIS)

Savolainen, S.; Katajala, S.; Elsing, B.; Nurkkala, P.; Hoikkanen, J.; Pullinen, J.; Logvinov, S.A.; Trunov, N.B.; Sitnik, J.K.

1997-01-01

Imatran Voima Oy, IVO, operates two VVER 440 reactors. Unit 1 has been operating since 1977 and unit 2 since 1981. First damages of the feed water distribution (FWD) pipes were observed in 1989. In closer examinations FWD-pipe T-connection turned out to suffer from severe erosion corrosion damages. Similar damages have been found also in other VVER 440 type NPPs. In 1994 the first new FWD-pipe was replaced and in 1996 extensive water hammer experiments were carried out together with EDO Gidropress in Podolsk. After the first phase of the experiments some fundamental changes were made to the construction of the FWD-pipe. The object of this paper is to give short insight to the design of the new FWD-pipe concentrating on water hammer experiments. (orig.)

Condensation driven water hammer studies for feedwater distribution pipe

Energy Technology Data Exchange (ETDEWEB)

Savolainen, S; Katajala, S; Elsing, B; Nurkkala, P; Hoikkanen, J [Imatran Voima Oy, Vantaa (Finland); Pullinen, J [IVO Power Engineering Ltd., Vantaa (Finland); Logvinov, S A; Trunov, N B; Sitnik, J K [EDO Gidropress (Russian Federation)

1998-12-31

Imatran Voima Oy, IVO, operates two VVER 440 reactors. Unit 1 has been operating since 1977 and unit 2 since 1981. First damages of the feed water distribution (FWD) pipes were observed in 1989. In closer examinations FWD-pipe T-connection turned out to suffer from severe erosion corrosion damages. Similar damages have been found also in other VVER 440 type NPPs. In 1994 the first new FWD-pipe was replaced and in 1996 extensive water hammer experiments were carried out together with EDO Gidropress in Podolsk. After the first phase of the experiments some fundamental changes were made to the construction of the FWD-pipe. The object of this paper is to give short insight to the design of the new FWD-pipe concentrating on water hammer experiments. (orig.).

Developing an e-commerce Website for Spicetown Oy, using Drupal.

OpenAIRE

Inegbedion, Usunobun

2015-01-01

This project is aimed at creating a website to be used for e-commerce by Spicetown Oy, a wholesaler of African and Asian food products based in Helsinki. Spicetown Oy is a traditional brick-and-mortar business in the process of transitioning into a business model combining tra-ditional business carried out on physical premises with running an online shop. The main aim of this project is to design and build a website with an online store for Spicetown Oy. The website is implemented using Drup...

Degradation modelling for the concrete silo in TVO's VLJ repository

International Nuclear Information System (INIS)

Alcorn, S.R.; Christian-Frear, T.L.; Wallace, M.

1991-05-01

Teollisuuden Voima Oy (TVO) is currendy construcing in Finland an underground repository (the VLJ repository) for storage of low- and intermediate-level radioactive wastes generated at the Olkiluoto (TVO I and TVO II) nuclear power plant. Intermediate level wastes will be emplaced inside a large concrete silo, which is the principal engineered barrier in the repository. The primary objective of the investigation is to develop an estimate of the length of time it will take for the silo to degrade due to interaction with groundwater to the point that it fails to perform as designed. A secondary objective is to develop a methodology to estimate the length of time required for radio nuclides to migrate from the region inside the silo through the silo wall and floor to the accessible environment as a function of cement and concrete properties. Chemical modeling techniques using the codes EQ3NR/EQ6 were employed to model the degradation of the repository concrete due to interaction with groundwater, and porous flow and diffusion modeling approaches were taken to: (1) estimate the time it would take groundwater and ions to travel into and out of the silo concrete, and (2) determine how these travel times change as the concrete degrades. The results of the investigation suggest that the hydraulic conductivity of the concrete will decrease over time because of the considerable net volume increase (net porosity decrease) from the chemical interactions. Therefore, it appears likely, based on the geochemical and mass transport models, that the silo win perform as required for at least its 500-year design life, and possibly much longer

Local extinction and reignition of the flame; Liekin paikallinen sammuminen ja uudelleen syttyminen

Energy Technology Data Exchange (ETDEWEB)

Kjaeldman, L. [VTT Energia, Espoo (Finland); Brink, A. [Aabo Akademi, Turku (Finland)

1996-12-01

A model of the local extinction and reignition of the flame suitable to be used in computational fluid dynamic analysis of primarily multi-burner furnaces is developed. The model is implemented in the computational environment Ardemus of VTT and Imatran Voima Oy, and tested against well defined experiments. The model makes the simulation of especially the near burner processes more realistic. (author)

Most advanced HTP fuel assembly design for EPR

International Nuclear Information System (INIS)

Francillon, Eric; Kiehlmann, Horst-Dieter

2006-01-01

End 2003, the Finnish electricity utility Teollisuuden Voima Oy (TVO) signed the contract for building an EPR in Olkiluoto (Finland). Mid 2004, the French electricity utility EDF selected an EPR to be built in France. In 2005, Framatome ANP, an AREVA and Siemens company, announced that they will be pursuing a design certification in the U.S. The EPR development is based on the latest PWR product lines of former Framatome (N4) and Siemens Nuklear (Konvoi). As an introductory part, different aspects of the EPR core characteristics connected to fuel assembly design are presented. It includes means of ensuring reactivity control like hybrid AIC/B4C control rod absorbers and gadolinium as burnable absorber integrated in fuel rods, and specific options for in-core instrumentation, such as Aeroball type instrumentation. Then the design requirements for the EPR fuel assembly are presented in term of very high burnup capacity, rod cladding and fuel assembly reliability. Framatome ANP fuel assembly product characteristics meeting these requirements are then described. EPR fuel assembly design characteristics benefit from the experience feedback of the latest fuel assembly products designed within Framatome ANP, leading to resistance to assembly deformation, high fuel rod restraint and prevention of handling hazards. EPR fuel assembly design features the best components composing the cornerstones of the upgraded family of fuel assemblies that FRAMATOME ANP proposes today. This family is based on a set of common characteristics and associated features, which include the HMP grid as bottom end spacer, the MONOBLOC guide tube and the Robust FUELGUARD as lower tie plate, the use of the M5 Alloy, as cladding and structure material. This fully re-crystallized, ternary Zr-Nb-O alloy produces radically improved in-reactor corrosion, very low hydrogen uptake and growth and an excellent creep behavior, which are described there. EPR fuel assembly description also includes fuel rod

Finnish Research Programme on Nuclear Waste Management (KYT). Framework Programme for 2002-2005

International Nuclear Information System (INIS)

Rasilainen, K.

2002-12-01

The new Finnish research programme on nuclear waste management (KYT) will be conducted in 2002 - 2005. This framework programme describes the starting point, the basic aims and the organisation of the research programme. The starting point of the KYT programme is derived from the present state and future challenges of Finnish nuclear waste management. The research programme is funded mainly by the Ministry of Trade and Industry (KTM), the Radiation and Nuclear Safety Authority (STUK), Posiva Oy, Fortum Oyj, Teollisuuden Voima Oy (TVO), and the National Technology Agency (Tekes). As both regulators and implementors are involved, the research programme concentrates on neutral research topics that must be studied in any case. Methods and tools for experimental and theoretical studies fall in this category. State of the art -reviews on relevant topics also create national know-how. Topics that directly belong to licensing activities of nuclear waste management are excluded from the research programme. KYT carries out technical studies that increase national know-how in the area of nuclear waste management. The aim is to maintain and develop basic expertise needed in the operations derived from the national nuclear waste management plan. The studies have been divided into strategic studies and studies enhancing the long-term safety of spent nuclear fuel disposal. Strategic studies support the overall feasibility of Finnish nuclear waste management. These studies include basic options and overall safety principles related to nuclear fuel cycle and nuclear waste management. In addition, general cost estimates as well as general safety considerations related to transportations, low- and medium level wastes, and decommissioning are included in strategic studies. Studies supporting the long-term safety of spent fuel disposal include issues related to performance assessment methodology, release of radionuclides from the repository, behaviour of bedrock and groundwater

Small scale power production

Energy Technology Data Exchange (ETDEWEB)

Muoniovaara, M [IVO International Ltd, Vantaa (Finland)

1997-12-31

IVO International is a major constructor of biomass power plants in Finland and abroad. As a subsidiary of Imatran Voima Oy, the largest power utility in Finland, it has designed and constructed ten power plants owned by IVO Group or others capable of burning biomasses. Sizes of the plants vary from the world`s largest condensing peat-fired power plant of 155 MWe to a 6 MWe combined heat and power producing unit. This article describes the biomass power plants designed and constructed by IVO Group 3 refs.

Small scale power production

Energy Technology Data Exchange (ETDEWEB)

Muoniovaara, M. [IVO International Ltd, Vantaa (Finland)

1996-12-31

IVO International is a major constructor of biomass power plants in Finland and abroad. As a subsidiary of Imatran Voima Oy, the largest power utility in Finland, it has designed and constructed ten power plants owned by IVO Group or others capable of burning biomasses. Sizes of the plants vary from the world`s largest condensing peat-fired power plant of 155 MWe to a 6 MWe combined heat and power producing unit. This article describes the biomass power plants designed and constructed by IVO Group 3 refs.

Evaluation of the dose assessment models for routine radioactive releases to the environment

International Nuclear Information System (INIS)

Rossi, J.

1998-05-01

The aim of the work was to evaluate the needs of development concerning the dose calculation models for routine releases and application of the models for exceptional release situations at the NPP plants operated by Imatran Voima Ltd. and Teollisuuden Voima Ltd. in Finland. First, the differences of the calculation models concerning input data, models themselves and output are considered. Subsequently some single features like importance of nuclides in exposure pathways due to change of the release composition, dose calculation for children and importance of time period of particle releases are considered. The existing dose calculation model used by the radiation safety authorities is aimed at a tool for checking the results from calculations of doses arising from routine releases by the power companies. Characteristics of an independent, foreign model and its suitability for safety authorities for dose calculations of releases in normal operation is also assessed. The needs of improvements in the existing calculation models and characteristics of a comprehensive model for safety authorities are discussed as well

Ion-Molecule Reaction of Gas-Phase Chromium Oxyanions: CrxOyHz- + H2O

International Nuclear Information System (INIS)

Gianotto, Anita Kay; Hodges, Brittany DM; Benson, Michael Timothy; Harrington, Peter Boves; Appelhans, Anthony David; Olson, John Eric; Groenewold, Gary Steven

2003-01-01

Chromium oxyanions having the general formula CrxOyHz- play a key role in many industrial, environmental, and analytical processes, which motivated investigations of their intrinsic reactivity. Reactions with water are perhaps the most significant, and were studied by generating CrxOyHz- in the gas phase using a quadrupole ion trap secondary ion mass spectrometer. Of the ions in the Cr1OyHz envelope (y = 2, 3, 4; z = 0, 1), only CrO2- was observed to react with H2O, producing the hydrated CrO3H2- at a slow rate (∼0.07% of the ion-molecule collision constant at 310 K). CrO3-, CrO4-, and CrO4H- were unreactive. In contrast, Cr2O4-, Cr2O5-, and Cr2O5H2- displayed a considerable tendency to react with H2O. Cr2O4- underwent sequential reactions with H2O, initially producing Cr2O5H2- at a rate that was ∼7% efficient. Cr2O5H2- then reacted with a second H2O by addition to form Cr2O6H4- (1.8% efficient) and by OH abstraction to form Cr2O6H3- (0.6% efficient). The reactions of Cr2O5- were similar to those of Cr2O5H2-: Cr2O5- underwent addition to form Cr2O6H2- (3% efficient) and OH abstraction to form Cr2O6H- (<1% efficient). By comparison, Cr2O6- was unreactive with H2O, and in fact, no further H2O addition could be observed for any of the Cr2O6Hz- anions. Hartree-Fock ab initio calculations showed that reactive CrxOyHz- species underwent nucleophilic attack by the incoming H2O molecules, which produced an initially formed adduct in which the water O was bound to a Cr center. The experimental and computational studies suggested that Cr2OyHz- species that have bi- or tricoordinated Cr centers are susceptible to attack by H2O; however, when the metal becomes tetracoordinate, reactivity stops. For the Cr2OyHz- anions the lowest energy structures all contained rhombic Cr2O2 rings with pendant O atoms and/or OH groups. The initially formed [Cr2Oy- + H2O] adducts underwent H rearrangement to a gem O atom to produce stable dihydroxy structures. The calculations indicated that

Environmental Impact Assessment for Olkiluoto 4 Nuclear Power Plant Unit in Finland

International Nuclear Information System (INIS)

Dersten, Riitta; Gahmberg, Sini; Takala, Jenni

2008-01-01

In order to improve its readiness for constructing additional production capacity, Teollisuuden Voima Oyj (TVO) initiated in spring 2007 the environmental impact assessment procedure (EIA procedure) concerning a new nuclear power plant unit that would possibly be located at Olkiluoto. When assessing the environmental impacts of the Olkiluoto nuclear power plant extension project, the present state of the environment was first examined, and after that, the changes caused by the projects as well as their significance were assessed, taking into account the combined impacts of the operations at Olkiluoto. The environmental impact assessment for the planned nuclear power plant unit covers the entire life cycle of the plant unit. (authors)

Environmental Impact Assessment for Olkiluoto 4 Nuclear Power Plant Unit in Finland

Energy Technology Data Exchange (ETDEWEB)

Dersten, Riitta; Gahmberg, Sini; Takala, Jenni [Teollisuuden Voima Oyj, Olkiluoto, FI-27160 Eurajoki (Finland)

2008-07-01

In order to improve its readiness for constructing additional production capacity, Teollisuuden Voima Oyj (TVO) initiated in spring 2007 the environmental impact assessment procedure (EIA procedure) concerning a new nuclear power plant unit that would possibly be located at Olkiluoto. When assessing the environmental impacts of the Olkiluoto nuclear power plant extension project, the present state of the environment was first examined, and after that, the changes caused by the projects as well as their significance were assessed, taking into account the combined impacts of the operations at Olkiluoto. The environmental impact assessment for the planned nuclear power plant unit covers the entire life cycle of the plant unit. (authors)

Markkinointisuunnitelma, URHOtv Oy

OpenAIRE

Pesonen, Matti

2012-01-01

Opinnäytetyön tarkoituksena on paneutua suomalaiseen maksu-tv markkinaan ja kilpailutilanteeseen jonka jälkeen laatia yritykselle vuoden 2012 markkinointisuunnitelma. URHOtv Oy on suurin kotimaiseen urheiluun keskittyä maksu-tv kanava. URHOtv:llä on noin 300 000 asiakasta. Opinnäytetyön pohjana on käytetty Jobberin markkinointisuunnitelman tekoprosessin teoriaa hiukan mukailtuna URHOtv:n tarpeisiin. Markkinointisuunnitelman tueksi opinnäytetyössä tutkitaan URHOtv:n tilaa analyysie...

Prospects of power ramping and cycling supervision in Finnish power reactors

Energy Technology Data Exchange (ETDEWEB)

Antila, M; Kaikkonen, H T [Imatran Voima Oy, Helsinki (Finland); Mannola, E [Teollisuuden Voima Oy Industries Kraft Ab, Helsinki (Finland)

1983-06-01

Since 1977 2x440 MWe PWR and 2x660 MWe BWR nuclear power has been taken in operation in Finland, which until the middle of 1982 has given favourable fuel operating experiences from 10 reactor years. This paper describes the core supervision systems of the plants especially from the viewpoint of ramp surveillance and the potentials and needs to improve the supervision capability to meet the future needs in case more load follow operation is required. As a special feature for Imatran Voima is the demand of general basic understanding of the behaviour of Loviisa reactors' fuel in different operating conditions. A possibility to investigate the fuel seem to be power cycling tests in Loviisa reactors. (author)

Prospects of power ramping and cycling supervision in Finnish power reactors

International Nuclear Information System (INIS)

Antila, M.; Kaikkonen, H.T.; Mannola, E.

1983-01-01

Since 1977 2x440 MWe PWR and 2x660 MWe BWR nuclear power has been taken in operation in Finland, which until the middle of 1982 has given favourable fuel operating experiences from 10 reactor years. This paper describes the core supervision systems of the plants especially from the viewpoint of ramp surveillance and the potentials and needs to improve the supervision capability to meet the future needs in case more load follow operation is required. As a special feature for Imatran Voima is the demand of general basic understanding of the behaviour of Loviisa reactors' fuel in different operating conditions. A possibility to investigate the fuel seem to be power cycling tests in Loviisa reactors. (author)

Drillings at Kivetty in Konginkangas

International Nuclear Information System (INIS)

Hinkkanen, H.; Oehberg, A.

1990-05-01

According to Government's decision in principle Teollisuuden Voima Oy is obliged to make bedrock investigations for the final disposal of the spent fuel produced by its power plant in Olkiluoto. Areas in Kuhmo, Hyrynsalmi, Sievi, Konginkangas and Olkiluoto were selected for the preliminary site investigations to be carried out during years 1987-1992. In Kivetty, Konginkangas the investigation program was started in spring 1988. During years 1988-1989 a deep borehole (1019 m) and 4 about 500 m deep additional boreholes were core drilled in the area. The structure of the holes makes it possible to carry out many investigations in the holes. Various parameters were measured from the flushing water during the drilling. Corelogging included collecting detailed data of fractures and determining the weathering degree and petrographical properties. Rock mechanical properties, uniaxial compressive strength, Young's modulus and Poisson's ratio were measured from core samples. The flushing water needed in the drillings was pumped from 100 m deep borehole wells drilled with down-the-hole method in the vicinity of the borehole. The water was labeled with 2 tracers before use. 30 vertical holes were core drilled down to the depth of 10 m in bedrock with a light drilling unit. Drilling was carried out in order to determine the thickness of the overburden to investigate the geophysical anomaly sources and to support geological mapping in areas covered with overburden. Groundwater hydraulics is one of the main subjects during the preliminary site investigation phase. For that reason 7 multilevel piezometers were installed on the site to monitore hydraulic head in 3 levels in the uppermost part of bedrock. The work consisted of borehole drillings to the depth of 100 m, geophysical borehole loggings and installation of piezometers. In addition about 65 shotholes were drilled for VSP-, tubewave and seismic measurements

Drillings at Syyry in Sievi

International Nuclear Information System (INIS)

Hinkkanen, H.; Oehberg, A.

1990-10-01

According to Government's decision in principle Teollisuuden Voima Oy is obliged to make bedrock investigations for the final disposal of the spent fuel produced by its power plant in Olkiluoto. Areas in Kuhmo, Hyrynsalmi, Sievi, Konginkangas and Olkiluoto were selected for the preliminary site investigations to be carried out during years 1987-1992. In Syyry, Sievi the investigation program was started in spring 1988. During years 1988-1989 a deep borehole (1022 m) and 4 about 500-700 m deep additional boreholes were core drilled in the area. The structure of the holes makes it possible to carry out many investigations in the holes. Various parameters were measured from the flushing water during the drilling. Corelogging included collecting detailed data of fractures and determining the weathering degree and petrographical properties. Rock mechanical properties, uniaxial compressive strength, Young's modulus and Poisson's ratio were measured from core samples. The flushing water needed in the drillings was pumped from 100 m deep borehole wells drilled with down-the-hole method in the vicinity of the borehole. The water was labeled with 2 tracers before use. 35 vertical holes were core drilled down to the depth of 10-20 m in bedrock with a light drilling unit. Drilling was carried out in order to determine the thickness of the overburden, to investigate the geophysical anomaly sources and to support geological mapping in areas covered with overburden. Groundwater hydraulics is one of the main subjects during the preliminary site investigation phase. For that reason 7 multilevel piezometers were installed on the site to monitore hydraulic head in 3 levels in the uppermost part of bedrock. The work consisted of borehole drillings to the depth of 100 m, geophysical borehole loggings and installation of piezometers. In addition about 85 shotholes were drilled for VSP-, tubewave and seismic measurements

RFID- TEKNIIKAN MAHDOLLISTAMAT TYÖAIKASÄÄSTÖT HUONEKALUTEOLLISUUDESSA : - CASE INCAP FURNITURE OY

OpenAIRE

Taskinen, Jarkko

2009-01-01

Opinnäytetyön tarkoituksena oli selvittää RFID- tekniikan eli radiotaajuudella toimivan etätunnistamisen mahdollistamia työaikasäästöjä huonekaluteollisuudessa. Toimeksiantaja oli Incap Furniture Oy. Incap Furniture Oy on yksi suurimmista mäntyhuonekalusopimusvalmistajista maailmassa. Pääasiakas on Ikea. Incap Furniture Oy on aloittanut projektin, jonka tarkoitus on tutkia RFID- tekniikan mahdollisuuksia ja hyötyjä tehtaiden toiminnassa. Opinnäytetyö oli osa projektia. Opinnäyt...

Modernization and power upgrading of the Loviisa NPP

Energy Technology Data Exchange (ETDEWEB)

Keskinen, A. [IVO Power Engineering Ltd., Vantaa (Finland)

1997-12-31

In 1995, Imatran Voima Oy (IVO) started a project for modernization and power upgrading of the Loviisa NPP. The main objectives of the project are to ensure plant safety, to increase electricity production and to improve the expertise of the IVO staff. The total electricity output of Loviisa 1 and 2 units is planned to be increased by about 100 MW. This will be achieved through renovation of the steam turbines and through gradual increase in the thermal reactor power up to 1,500 MW from the present level of 1,375 MW. The Loviisa NPP Final Safety Analysis Report has been revised to a great extent in connection with the licensing process of the reactor power upgrading. The project also includes certain improvements in the primary and safety systems to ensure plant safety. The total cost estimate of the project is around 200 million FIM. The project implementation started in 1995 and in accordance with the plans in 2000 after several phases the last measures at power plant will be completed. (orig.). 4 refs.

Modernization and power upgrading of the Loviisa NPP

International Nuclear Information System (INIS)

Keskinen, A.

1997-01-01

In 1995, Imatran Voima Oy (IVO) started a project for modernization and power upgrading of the Loviisa NPP. The main objectives of the project are to ensure plant safety, to increase electricity production and to improve the expertise of the IVO staff. The total electricity output of Loviisa 1 and 2 units is planned to be increased by about 100 MW. This will be achieved through renovation of the steam turbines and through gradual increase in the thermal reactor power up to 1,500 MW from the present level of 1,375 MW. The Loviisa NPP Final Safety Analysis Report has been revised to a great extent in connection with the licensing process of the reactor power upgrading. The project also includes certain improvements in the primary and safety systems to ensure plant safety. The total cost estimate of the project is around 200 million FIM. The project implementation started in 1995 and in accordance with the plans in 2000 after several phases the last measures at power plant will be completed. (orig.)

Operation of Finnish nuclear power plants. Quarterly report 4th quarter, 1994 and annual summary

International Nuclear Information System (INIS)

Tossavainen, K.

1995-05-01

The Loviisa NPP units were in power operation the whole last quarter, with the exception of a reactor scram at Loviisa 1. The load factor average of all Finnish plant units was 100.2 %. The annual average was 90.0 %. All events in the fourth annual quarter were assigned level 0 (no safety significance) on the international INES scale. Four events in 1994 were classified level 1 (an anomaly). The Finnish Centre for Radiation and Nuclear Safety in December approved Imatran Voima Oy's application to extend the operation of the reactor pressure vessel of Loviisa 2 until the annual maintenance outage of 2010. During this quarter, a batch of spent fuel from Loviisa power plant was transported to Russia. Occupational doses and radioactive releases off-site were below authorised limits. Only such quantities of plant-based radioactive materials were measurable in samples collected around the plants as have no bearing on the radiation exposure of the population. The report includes a summary of all the items described in the Quarterly Reports of 1994. (8 figs., 4 tabs.)

Modernization and power upgrading of the Loviisa NPP

Energy Technology Data Exchange (ETDEWEB)

Keskinen, A [IVO Power Engineering Ltd., Vantaa (Finland)

1998-12-31

In 1995, Imatran Voima Oy (IVO) started a project for modernization and power upgrading of the Loviisa NPP. The main objectives of the project are to ensure plant safety, to increase electricity production and to improve the expertise of the IVO staff. The total electricity output of Loviisa 1 and 2 units is planned to be increased by about 100 MW. This will be achieved through renovation of the steam turbines and through gradual increase in the thermal reactor power up to 1,500 MW from the present level of 1,375 MW. The Loviisa NPP Final Safety Analysis Report has been revised to a great extent in connection with the licensing process of the reactor power upgrading. The project also includes certain improvements in the primary and safety systems to ensure plant safety. The total cost estimate of the project is around 200 million FIM. The project implementation started in 1995 and in accordance with the plans in 2000 after several phases the last measures at power plant will be completed. (orig.). 4 refs.

Digitaalisen markkinointiviestinnän mahdollisuudet : Case Kelloliike EA Lahti Oy

OpenAIRE

Harmaala, Tiina

2011-01-01

TIIVISTELMÄ Harmaala, Tiina. 2011. Digitaalisen markkinointiviestinnän mahdollisuudet – Case Kelloliike EA Lahti Oy. Opinnäytetyö. Kemi-Tornion ammattikorkeakoulu. Kaupan ja kulttuurin toimiala. Sivuja 62. Opinnäytetyön tavoitteena on Kelloliike EA Lahti Oy:n digitaalisen markkinointiviestinnän suunnittelu. Tutkimuksessa tarkastelen digitaalista markkinointiviestintää integroidun markkinointiviestinnän näkökulmasta. Tutkimuksen tarkoituksena on selvittää toimeksiantajalle sen markkino...

Adidas Suomi Oy: Brändikuvan rakentaminen markkinoinnin keinoin

OpenAIRE

Leinonen, Miira

2015-01-01

Tämä opinnäytetyö tehtiin toimeksiantona Adidas Suomi Oy:lle syksyllä 2014. Opinnäytetyön tavoitteena oli selvittää muotialan mielipidevaikuttajien mielikuvat Adidaksesta ja sen merkittävimmistä kilpailijoista. Vertailtaviksi brändeiksi valittiin kuusi tärkeintä kilpailijaa, jotka ovat Converse, Karhu, New Balance, Nike, Puma ja Reebok. Opinnäytetyön toimeksiantaja Adidas Suomi Oy on osa kansainvälistä Adidas Group konsernia. Yritys on yksi maailman tunnetuimmista urheilumerkeistä ja toim...

SIEVIN JALKINE OY:N AVAINASIAKKAIDEN KÄSITYKSIÄ YRITYKSEN TOIMINNASTA

OpenAIRE

Haapakoski, Sanna

2014-01-01

Opinnäytetyön toimeksiantaja on Sievin Jalkine Oy. Yritys on Pohjoismaiden johtava turva- ja ammattijalkineiden valmistaja. Opinnäytetyön tavoitteena oli selvittää, kuinka tyytyväisiä avainasiakkaat ovat Sievin Jalkine Oy:n toimintaan ja miten toimintaa voidaan kehittää asiakkaiden hyväksi. Opinnäytetyössä selvitettiin, miten tyytyväisiä avainasiakkaat ovat tuotetarjontaan, palvelujen tarjontaan, markkinointiviestintään, asiakaspalveluun ja henkilöstöön. Opinnäytetyön tietoperustassa...

Tilinpäätösanalyysi : case Ilmatuote Oy

OpenAIRE

Luotola, Elina

2014-01-01

Opinnäytetyön tavoitteena oli saada selville Ilmatuote Oy:n taloudellinen tilanne tarkastellen tilinpäätösinformaatiota vuosien 2009-2012 väliseltä ajanjaksolta. Ilmatuote Oy:n erikoisalana on jo yli 40 vuoden ajan ollut talotekniikkajärjestelmien asennus, urakointi- ja huoltotoiminta. Tarkoituksena oli tehdä tilinpäätösanalyysi yhtiön tilinpäätöksistä ja selvittää, millä tasolla yritys taloudellisesti on ja mihin suuntaan yrityksen taloudellinen tila on kehittynyt. Teoreettisessa osassa t...

Results of monitoring at Olkiluoto in 2012. Environment

International Nuclear Information System (INIS)

Haapanen, A.

2014-04-01

In 2003, Posiva Oy presented a programme for monitoring at Olkiluoto during construction and operation of ONKALO. In 2012 the monitoring programme was updated to concern the years 2012-2018. Part of the monitoring is performed by the company running the nuclear power plants on the island, Teollisuuden Voima Oy (TVO). This Working Report presents the main results of Posiva's environmental monitoring programme on Olkiluoto Island in 2012. Results are presented under five topics: 1. Evolution of geosphere, 2. Biosphere modelling input data, 3. Interaction between surface environment and groundwater in bedrock, 4. Environmental impact and 5. Baseline of monitoring of radioactive releases. Concerning the evolution of geosphere, LIDAR-scannings were done in the Olkiluoto area in 2012. The acquired data can be used for elevation and other modelling purposes. The soil solution quality in 2012 was quite comparable to that in earlier years. Proximity of the sea and the young age of soils are seen in soil solution results. Biosphere modelling input data in 2012 included e.g. continuous tree litterfall and transpiration data, as well as updated game statistics and population estimates of fauna, a fishery survey from the River Eurajoki (2011) and basic monitoring data from Olkiluoto offshore properties. Interaction between surface environment and groundwater in bedrock includes e.g. weather and surface water monitoring data. Environmental impact analyses included e.g. monitoring of noise, air quality, effluent waters and private drilled wells. Noise monitoring in the vicinity of ONKALO showed that in the case of raised noise levels the sources are mainly the traffic on Olkiluodontie road, the air conditioning of ONKALO and occasional sources such as springtime bird sounds. Construction activities in the area were seen in increased amount of NO 3 -N in the bulk deposition, and Al and Fe accumulating on needle surfaces in areas close to the rock piling and crushing area. Scots

Results of monitoring at Olkiluoto in 2012. Environment

Energy Technology Data Exchange (ETDEWEB)

Haapanen, A. (ed.) [Haapanen Forest Consulting, Vanhakylae (Finland)

2014-04-15

In 2003, Posiva Oy presented a programme for monitoring at Olkiluoto during construction and operation of ONKALO. In 2012 the monitoring programme was updated to concern the years 2012-2018. Part of the monitoring is performed by the company running the nuclear power plants on the island, Teollisuuden Voima Oy (TVO). This Working Report presents the main results of Posiva's environmental monitoring programme on Olkiluoto Island in 2012. Results are presented under five topics: 1. Evolution of geosphere, 2. Biosphere modelling input data, 3. Interaction between surface environment and groundwater in bedrock, 4. Environmental impact and 5. Baseline of monitoring of radioactive releases. Concerning the evolution of geosphere, LIDAR-scannings were done in the Olkiluoto area in 2012. The acquired data can be used for elevation and other modelling purposes. The soil solution quality in 2012 was quite comparable to that in earlier years. Proximity of the sea and the young age of soils are seen in soil solution results. Biosphere modelling input data in 2012 included e.g. continuous tree litterfall and transpiration data, as well as updated game statistics and population estimates of fauna, a fishery survey from the River Eurajoki (2011) and basic monitoring data from Olkiluoto offshore properties. Interaction between surface environment and groundwater in bedrock includes e.g. weather and surface water monitoring data. Environmental impact analyses included e.g. monitoring of noise, air quality, effluent waters and private drilled wells. Noise monitoring in the vicinity of ONKALO showed that in the case of raised noise levels the sources are mainly the traffic on Olkiluodontie road, the air conditioning of ONKALO and occasional sources such as springtime bird sounds. Construction activities in the area were seen in increased amount of NO{sub 3}-N in the bulk deposition, and Al and Fe accumulating on needle surfaces in areas close to the rock piling and crushing area

Taloushallintoprosessien kehittäminen Iisalmen Putkiasennus Oy:ssä

OpenAIRE

Juntunen, Tiia

2015-01-01

Tämä opinnäytetyö käsittelee taloushallintoprosessien kehittämistä Iisalmen Putkiasennus Oy:ssä. Kehittämistyötä päädyttiin rajaamaan keskittyen myynti- ja ostolaskuprosessien kehittämiseen, koska niissä oli havaittu olevan eniten kehitettävää. Iisalmen Putkiasennus Oy on LVI-alalla työskentelevä perheyritys, joka on ollut toiminnassa jo yli 35 vuotta. Yritys sijaitsee Iisalmessa ja työllistää noin 50 henkilöä. Tässä opinnäytetyössä taloushallintoprosessien kehittäminen alkoi tilanteesta...

OpenOffice.org -toimisto-ohjelmiston soveltuvuus Sastamalan Tukipalvelu Oy:lle : selvitys

OpenAIRE

Ojanen, Olli-Veikko

2009-01-01

Tässä opinnäytetyössä selvitettiin OpenOffice.org toimisto-ohjelmiston soveltuvuutta Sastamalan Tukipalvelu Oy:lle, joka huolehtii osakkaittensa tieto-, talous- ja henkilöstöhallinnon palveluista. Sastamalan Tukipalvelu Oy:n osakkaina ovat Kiikoisten, Lavian ja Punkalaitumen kunnat, Sastamalan kaupunki, Sastamalan perusturvakuntayhtymä sekä Sastamalan koulutuskuntayhtymä. Selvityksen tavoitteena oli tuoda esille OpenOffice.org -toimisto-ohjelmistoon siirtymisen hyötyjä, joi...

Building a new nuclear power plant in Finland? Studies performed. Annex 2

International Nuclear Information System (INIS)

Patrakka, E.

2002-01-01

The electricity consumption per capita is high in Finland due to the country's industrial structure and to the climatic conditions. Industry consumes 55% of the electricity in Finland. The demand of electricity is expected to grow at a rate of 1.5% a year until 2010, and further at a yearly rate of 1% until 2015. This will require 3800 MW of new generating capacity by 2015. A recent study indicates that in base-load power production in Finland the generating costs of a nuclear plant are the lowest in comparison with generation using coal, natural gas or peat. The difference to coal would be 9%, to gas 18% and to peat 40%. The target for Finland to reduce greenhouse gas emissions under the EU burden sharing is 0%. In comparison with business as usual scenarios, however, the reduction need is of the magnitude of 20%, one of the hardest in the EU. Finland already has taken into use the methods, which now are considered essential within the EU for reducing the CO 2 releases. Teollisuuden Voima Oy (TVO) submitted on 15 November 2000 to the Council of State an application for a decision in principle concerning the construction of additional nuclear capacity. The submission of the application is reasoned by the shareholders' need for additional electricity. Furthermore, nuclear power, together with renewable energy sources, makes it possible to comply with the Kyoto protocol commitments. The actual investment decision can be made first after a positive decision in principle has been received from the Council of State and the Parliament. The submission of the application was preceded by a number of studies, the contents of which are summarised. (author)

Rock mechanical, thermomechanical and hydraulic behaviour of the near field for spent nuclear fuel

International Nuclear Information System (INIS)

Johansson, E.; Hakala, M.; Lorig, L.J.

1991-10-01

Teollisuuden Voima Oy (TVO) is investigating the feasibility of disposing high level nuclear waste in crystalline rock at depths of 400 to 600 meters below the ground surface. Two explicit distinct element computer codes UDEC and 3DEC were used to simulate the mechanical response associated with excavation and the thermomechanical response associated with waste emplacement. Model input data are mostly based on preliminary design of the repository and on field data from on-going site investigations in Finland. The results showed that the overall stability of the repository near-field appears to be good during the studied time period 0 - 900 years. The maximum displacements after excavation are about 2 mm on the walls of the disposal tunnel. Joint openings are only a few micrometers. The hydraulic conductivity increases by 4 to 6 times within the zone of 0,3 m around the tunnel and emplacement hole, and farther away the average increase in conductivity is 1,2 to 1,7 times. After 60 years the heating increases the stresses in the vicinity of the excavated rooms, and closes the joints decreasing the hydraulic conductivity by 93 - 99 % when assuming 10 μm in-situ hydraulic aperture. However, when assuming 50 μm in-situ hydraulic aperture the hydraulic conductivity increases 10 - 40 % because the change in dynamic viscosity of water has a larger effect than the joint aperture change. After 900 years in the cooling stage the stresses and displacements come back almost to the same level as after the excavation. Some permanent displacements remain in the joints due to the shearing. The hydraulic conductivity at 900 years is 10 - 70 % of the conductivity after the excavation. The comparisons between the 2-D and 3-D results show that the two-dimensional modeling, if sufficient cross-sections have been analyzed, is enough to describe mechanical behaviour of the near-field, whereas the three-dimensional modeling is needed in some cases to assess the thermomechanical behaviour

Markkinointisuunnitelma sosiaalista mediaa hyödyntäen : Case Branda Oy

OpenAIRE

Hämeenniemi, Henna-Riika

2016-01-01

Tämän opinnäytetyön tarkoituksena oli tehdä markkinointisuunnitelma toimintaansa aloittelevalle yritykselle nimeltä Branda Oy. Branda Oy perusti vuoden 2015 alussa verkkokaupan, jonka kautta se myy naisten urheiluvaatteita suomalaisille kuluttajille. Opinnäytetyön tavoitteena oli antaa yritykselle välineet markkinoida itseään sosiaalisessa mediassa parhaalla mahdollisella tavalla. Työ toteutettiin kahdessa vaiheessa, josta ensimmäisessä toteutettiin itse markkinointisuunnitelma Branda O...

Superconducting transition temperature in the Y(1-x)M(x)Ba2Cu3O(y) system

Science.gov (United States)

Suzuki, Takeyuki; Yamazaki, Tsutomu; Sekine, Ryuuta; Koukitsu, Akinori; Seki, Hisashi

1989-04-01

Experimental results are presented for the inclusion of compositional additives, M, to the sintered high-temperature superconductor Y(1-x)M(x)Ba2Cu3O(y); M can be the oxides of Mg, Ce, Gd, Yb, Ti, Zr, V, Nb, Ta, Cr, Mo, W, Mn, Fe, Co, Ni, Zn, B, Al, Ga, In, Si, Ge, Sn, Pb, Sb, Bi, and Te, as well as Li, Na, K, Ca, Sr, and La carbonates. Temperature dependence of the electrical resistance was measured down to about 80 K. Attention is given to the influence of ionic radius and the valence of the M species.

Jatkuvan koulutuksen antama tuki myyntityölle : Case: Systemair Oy

OpenAIRE

Mattila, Jeremias

2017-01-01

Opinnäytetyössä pyritään selvittämään, kokevatko myyjät koulutuksen edistävän heidän itse-varmuuttaan työtä kohden tai kokevatko he koulutuksen antaman tuen jollain muulla tavalla? Tulokset perustuvat laadulliseen Systemair Oy:n myyntihenkilöstön teemahaastatteluun. Kaikki tulokset ovat siis subjektiivisia Systemair Oy:n toimialaa ja sisäistä yrityskulttuuria kos-kien. Sovellettavuus muihin yrityksiin ja toimialoihin vaatii uuden tutkimuksen, mutta työ voi antaa osviittaa siihen, millaiset ja...

Situated Word Learning: Words of the Year (WsOY) and Social Studies Inquiry

Science.gov (United States)

Heafner, Tina L.; Triplett, Nicholas; Handler, Laura; Massey, Dixie

2018-01-01

Current events influence public interest and drive Internet word searches. For over a decade, linguists and dictionary publishers have analyzed big data from Internet word searches to designate "Words of the Year" (WsOY). In this study, we examine how WsOY can foster critical digital literacy and illuminate essential aspects of inquiry…

Site specific information in site selection

International Nuclear Information System (INIS)

Aeikaes, T.; Hautojaervi, A.

1998-01-01

The programme for the siting of a deep repository for final disposal of spent nuclear fuel was started already in 1983 and is carried out today by Posiva Oy which continues the work started by Teollisuuden Voima Oy. The programme aims at site selection by the end of the year 2000. The programme has progressed in successive interim stages with defined goals. After an early phase for site identification, five sites were selected in 1987 for preliminary site characterisation. Three of these were selected and judged to be best suited for the more detailed characterisation in 1992. An additional new site was included into the programme based on a separate feasibility study in the beginning of 1997. Since the year 1983 several safety assessments together with technical plans of the facility have been completed. When approaching the site selection the needs for more detailed consideration of the site specific properties in the safety assessment have been increased. The Finnish regulator STUK has published a proposal for general safety requirements for the final disposal of spent nuclear fuel in Finland. This set of requirements has been projected to be used in conjunction of the decision making by the end 2000. Based on the site evaluation all sites can provide a stable environment and there is evidence that the requirements for the longevity of the canister can be fulfilled at each site. In this manner the four candidate sites do not differ too much from each other. The main difference between the sites is in the salinity of the deep groundwater. The significance of differences in the salinity for the long-term safety cannot be defined yet. The differences may contribute to the discussion of the longevity of the bentonite buffer and also to the modelling of the groundwater flow and transport. The use of the geosphere as a transport barrier is basically culminated on the questions about sparse but fast flow routes and 'how bad channeling can be'. To answer these questions

RESEARCH ON THE EFFECTIVENESS OF ADVERTISING CHANNELS AT VIITAMAA CARAVAN OY

OpenAIRE

Huumonen, Anna-Sofia

2012-01-01

The topic of this study is the effectiveness of advertising channels.The topic was chosen by the request of Viitamaa Caravan Oy. The company wants this study to give more information about the effectiveness of different advertising channels in their own advertising. Viitamaa Caravan Oy also wanted some insight into the fact how much money in the budget should be reserved for each channel. The research problem can be defined as: Which of the advertising channels already used in Viitamaa Carav...

Hydrogeological modelling for assessment of radionuclide release scenarios for the repository system 2012

Energy Technology Data Exchange (ETDEWEB)

Hartley, L.; Hoek, J.; Swan, D.; Appleyard, P.; Baxter, S.; Roberts, D.; Simpson, T. [AMEC (United Kingdom)

2013-07-15

Posiva Oy is responsible for implementing the programme for geological disposal of spent nuclear fuel produced by its owners Teollisuuden Voima Oyj (TVO) and Fortum Power and Heat Oy in Finland. Olkiluoto in Eurajoki has been selected as the primary site for the repository, subject to further detailed investigation which is currently focused on the construction of an underground rock characterisation and research facility (the ONKALO). An essential part of the assessment of long-term safety of a repository is the analysis of groundwater flow since it is the only means of transport of radionuclides to the biosphere (besides human intrusion). The analysis of long-term safety for a KBS-3 concept requires as input a description of details of the groundwater flow around and through components of the engineered barrier system as well as details of the groundwater pathway to the biosphere during the current temperate climate period, as well as indications of behaviour under future climate periods such as glacial conditions. This report describes the groundwater flow modelling study performed to provide some of the necessary inputs required by Safety Assessment (i.e. radionuclide transport analysis). Underlying this study is the understanding of the site developed during the site investigations as summarised in the site descriptive model (SDM), and in particular the description of Olkiluoto Hydrogeological DFN model (Hydro-DFN). The main focus of this study is the temperate climate period, i.e. the evolution over the next 10,000 years, but the hydrogeological situation under various glacial climate conditions is also evaluated. Primary outputs of the study are repository performance measures relating to: the distributions of groundwater flow around the deposition holes; deposition tunnels and through the EDZ; flow-related transport resistance along groundwater pathways from the repository to the surface; and their the exit locations. Other analyses consider the

Hydrogeological modelling for assessment of radionuclide release scenarios for the repository system 2012

International Nuclear Information System (INIS)

Hartley, L.; Hoek, J.; Swan, D.; Appleyard, P.; Baxter, S.; Roberts, D.; Simpson, T.

2013-07-01

Posiva Oy is responsible for implementing the programme for geological disposal of spent nuclear fuel produced by its owners Teollisuuden Voima Oyj (TVO) and Fortum Power and Heat Oy in Finland. Olkiluoto in Eurajoki has been selected as the primary site for the repository, subject to further detailed investigation which is currently focused on the construction of an underground rock characterisation and research facility (the ONKALO). An essential part of the assessment of long-term safety of a repository is the analysis of groundwater flow since it is the only means of transport of radionuclides to the biosphere (besides human intrusion). The analysis of long-term safety for a KBS-3 concept requires as input a description of details of the groundwater flow around and through components of the engineered barrier system as well as details of the groundwater pathway to the biosphere during the current temperate climate period, as well as indications of behaviour under future climate periods such as glacial conditions. This report describes the groundwater flow modelling study performed to provide some of the necessary inputs required by Safety Assessment (i.e. radionuclide transport analysis). Underlying this study is the understanding of the site developed during the site investigations as summarised in the site descriptive model (SDM), and in particular the description of Olkiluoto Hydrogeological DFN model (Hydro-DFN). The main focus of this study is the temperate climate period, i.e. the evolution over the next 10,000 years, but the hydrogeological situation under various glacial climate conditions is also evaluated. Primary outputs of the study are repository performance measures relating to: the distributions of groundwater flow around the deposition holes; deposition tunnels and through the EDZ; flow-related transport resistance along groundwater pathways from the repository to the surface; and their the exit locations. Other analyses consider the

Afrikkaviikko Antell Catering Oy:ssä

OpenAIRE

Bask, Tiina

2011-01-01

Ruokapalvelualalla erilaisten teemaviikkojen toteuttaminen on keino tarjota asiakkaille ainutkertaisia elämyksiä ja tuoda vaihtelua normaaliin arkiruokailuun. Tämän opinnäytetyön tavoitteena oli suunnitella Afrikka-aiheinen teemaviikko Antell Catering Oy:n lounasravintoloihin. Afrikka-teeman mukaisesti suunniteltiin vakioruokaohjeet yhden viikon lounasvaihtoehdoiksi sekä markkinointimateriaali somistuksineen. Opinnäytetyö on toiminnallinen työ, jossa Afrikka-viikon lounasruokaohjeet mu...

Customer Satisfaction in Internal Customer Service : Case: Abloy Oy Internal Customer Service

OpenAIRE

Turunen, Susanna

2011-01-01

ABSTRACT Turunen, Susanna Marita 2011. Customer Satisfaction in Internal Customer Service. Case: Abloy Oy Internal Customer Service. Master’s thesis. Kemi-Tornio University of Applied Sciences. Business and Culture. Pages 73. Appendix 1. This thesis discusses and studies service quality and customer satisfaction in internal customer service. The main objective is to find out what the service quality level in the internal customer service at Abloy Oy is and whether there exists a diffe...

R20 Programme: Grout setting and strength development in ONKALO. Literature review, observations and experiments

International Nuclear Information System (INIS)

Karttunen, P.; Raivio, P.

2008-12-01

ONKALO is an underground rock characterisation facility planned to be a part of nuclear waste repository in future. ONKALO is located in Olkiluoto Finland. Posiva Oy owned by Teollisuuden Voima Oy and Fortum Power and Heat Oy is responsible for the repository, research, construction and use of the ONKALO and closing of the underground facility after use. During construction of ONKALO it has been observed that the setting and strength of grouting materials have not sporadically developed as expected (in ONKALO). The phenomenon has been observed for the first time in the year 2005. The observations examined in this report are made in the grouting field tests and in ordinary grouting during the year 2007. The phenomenon has been observed with low pH and standard grouts and bolt grouting mortars. The reasons for this phenomenon are studied based on literature review, observations and tests in the field and laboratory. The effect of reactions between groundwater and grout, the effect of the raw materials as well as curing conditions, temperature and pressure are studied. There are several potential factors that can cause observed phenomenon. Some factors are more probable than others. Laboratory experiments for the samples of poor strength development were done. These samples were taken from the grouting holes or packers in which the strength of the grout was not developed as expected. The results of these experiments were compared to the results gained from the samples cast from the same grout batches and cured in the tunnel conditions. The purpose was to find out the factor causing slow strength development of the grouted mixes. One single reason, which can slow the setting of the grouts in ONKALO is the low temperature in the rock, but the temperature cannot cause the phenomenon alone. Locally groundwater contains compounds that can create chemically aggressive environment for (the Portland) cement based grouts. The groundwater chemistry in ONKALO has not been proved

B2B-uutiskirjeen visuaalisen ilmeen kehitys — Case Tamtron Solutions Oy

OpenAIRE

Salo, Sari

2014-01-01

Opinnäytetyön tarkoituksena oli kehittää Tamtron Solutions Oy:n uutiskirjeelle uusi visuaalinen ilme. Yrityksen lähettämästä uutiskirjeestä yritysasiakkaille puhutaan B2B-uutiskirjeenä. Tamtron Solutions Oy halusi B2B-uutiskirjeen avulla tehostaa markkinointia, kasvattaa myyntiä ja luoda uusia asiakassuhteita. Yritys halusi toteuttaa markkinointikampanjan B2B-uutiskirjeen muodossa rakennusalan yrityksille. Yritys valitsi kampanjan kohderyhmäksi rakennusalan yritykset, koska lakimuutos rakennu...

Mixed pinning landscape in nanoparticle-introduced YGdBa2Cu3Oy films grown by metal organic deposition

Science.gov (United States)

Miura, M.; Maiorov, B.; Baily, S. A.; Haberkorn, N.; Willis, J. O.; Marken, K.; Izumi, T.; Shiohara, Y.; Civale, L.

2011-05-01

We study the field (H) and temperature (T) dependence of the critical current density (Jc) and irreversibility field (Hirr) at different field orientations in Y0.77Gd0.23Ba2Cu3Oy with randomly distributed BaZrO3 nanoparticles (YGdBCO+BZO) and YBa2Cu3Oy (YBCO) films. Both MOD films have large RE2Cu2O5 (225) nanoparticles (˜80 nm in diameter) and a high density of twin boundaries (TB). In addition, YGdBCO+BZO films have a high density of BZO nanoparticles (˜25 nm in diameter). At high temperatures (T > 40 K), the superconducting properties, such as Jc, Hirr, and flux creep rates, are greatly affected by the BZO nanoparticles, while at low temperatures the superconducting properties of both the YBCO and YGdBCO+BZO films show similar field and temperature dependencies. In particular, while the Jc of YBCO films follow a power-law dependence (∝H-α) at all measured T, this dependence is only followed at low T for YGdBCO+BZO films. As a function of T, the YGdBCO+BZO film shows Jc(T,0.01T)~[1-(T/Tc)2]n with n ˜ 1.24 ± 0.05, which points to “δTc pinning.” We analyze the role of different types of defects in the different temperature regimes and find that the strong pinning of the BZO nanoparticles yields a higher Hirr and improved Jc along the c axis and at intermediate orientations at high T. The mixed pinning landscapes due to the presence of disorder of various dimensionalities have an important role in the improvement of in-field properties.

The effect of CO2 on the plasma remediation of NxOy

Science.gov (United States)

Gentile, Ann C.; Kushner, Mark J.

1996-04-01

Plasma remediation is being investigated for the removal of oxides of nitrogen (NxOy) from atmospheric pressure gas streams. In previous works we have investigated the plasma remediation of NxOy from N2/O2/H2O mixtures using repetitively pulsed dielectric barrier discharges. As combustion effluents contain large percentages of CO2, in this paper we discuss the consequences of CO2 in the gas mixture on the efficiency of remediation and on the end products. We find that there is a small increase in the efficiency of total NxOy remediation (molecules/eV) with increasing CO2 fraction, however the efficiency of NO remediation alone generally decreases with increasing CO2. This differential is more pronounced at low energy deposition per pulse. More remediation occurs through the reduction channel with increasing CO2 while less NO2 and HNOx are produced through the oxidation channel. CO is produced by electron impact of CO2 though negligible amounts of cyanides are generated.

Entering market in St. Petersburg, Russia– Kärävä Oy

OpenAIRE

Tsirkunov, Aleksandr

2013-01-01

The company commissioning this thesis is Kärävä Oy. This company was founded in 1988 and since then they focus on producing timber of different quality and size. Their clients are wholesalers of timber and wood products, along with sauna manufacturers. Produced products are used for interior finishing in commercial and public areas. More-over, Kärävä Oy is also producing interior materials for saunas. Main purpose for this thesis is to analyse if entering Russian market will be profitable...

Development and testing of VTT approach to risk-informed in-service inspection methodology. Final report of SAFIR INTELI INPUT Project RI-ISI

International Nuclear Information System (INIS)

Cronvall, O.; Maennistoe, I.; Simola, K.

2007-04-01

This report summarises the results of a research project on risk-informed in-service inspection (RI-ISI) methodology conducted in the Finnish national nuclear energy research programme SAFIR (2003-2006). The purpose of this work was to increase the accuracy of risk estimates used in RI-ISI analyses of nuclear power plant (NPP) piping systems, and to quantitatively evaluate the effects of different piping inspection strategies on risk. Piping failure occurrences were sampled by using probabilistic fracture mechanics (PFM) analyses. The PFM results for crack growth were used to construct transition matrices for a discrete-time Markov process model, which in turn was applied to examine the effects of various inspection strategies on the failure probabilities and risks. The applicability of the developed quantitative risk matrix approach was examined as a pilot study performed to the Shut-down cooling piping system 321 in NPP unit OL1 of Teollisuuden Voima Oy (TVO). The analysed degradation mechanisms were stress corrosion cracking (SCC) and thermal fatigue induced cracking (in the mixing points). Here a new and rather straightforward approach was developed to model thermal fatigue induced cracking, which degradation mechanism is much more difficult to model than SCC. This study further demonstrated the usefulness of Markov analysis procedure development by VTT in RI-ISI applications. The most important results are the quantified comparisons of different inspections strategies. It was shown in this study that Markov models are useful for this purpose, when combined with PFM analyses. While the numerical results could benefit from further considerations of inspection reliability, this does not affect the feasibility of the method itself. The approach can be used to identify an optimal inspection strategy for achieving a balanced risk profile of piping segments. (orig.)

Applicability of living PSA in NPP modernization

International Nuclear Information System (INIS)

Himanen, R.

1999-01-01

Recently the utility Teollisuuden Voima Oy (TVO) has modernized the Olkiluoto 1 and 2 nuclear units and increased the net electric power by 18 per cent. Level 2 PSA was performed during the modernization project and the living level 1 PSA was used to support the design of the plant modifications. The plant specific living PSA model was a powerful tool when evaluating modernization alternatives. Successive support of safety management with the PSA model requires, that both the utility and the Regulatory Body understand capability and limitations of the model in details. TVO has prepared an internal procedure that presents in detail the practices and responsibilities concerning living PSA. The procedure is based on general guidelines and requirements on probabilistic safety analysis of nuclear power plants in Finland, released by the Regulatory Body. Living PSA requires that also the procedure for the use of living PSA is living. The recently published USNRC Regulatory Guides on PSA will be taken into account in the next version of the TVO PSA procedure. The PSA Peer Review Certification Process is one way to evaluate the quality of PSA in general, but also to detect the weaknesses of the PSA. However, the Certification Process cover only limited scope of PSA omitting e.g. all other external events except internal floods. This paper gives an overview on the scope of living PSA for Olkiluoto 1 and 2, and presents some examples on the real use of PSA concerning the modernization of the plant. Definition of quantitative dependability requirements for renovated systems is possible, but on the other hand, proving of these targets is in some cases extremely difficult, because of lacking dependability data. The problems are mainly concerned in systems with of programmable logic control. (au)

Drilling of deep boreholes and associated geological investigations. Final disposal of spent fuel

International Nuclear Information System (INIS)

Anttila, P.

1983-12-01

Teollisuuden Voima Oy (Industrial Power Company Ltd.) will take precautions for the final disposal of spent fuel in the Finnish bedrock. The first stage of the site selection studies includes drilling of a deep borehole down to approximately 1000 metres in the winter of 1984. The choice of drilling method and equipment depends on the geological circumstances and the target of the investigation. The most common drilling methods used with the investigations of nuclear waste disposal are diamond core drilling and percussion drilling. The Precambrian bedrock outcropping in Finland exists also in Sweden and Canada, where deep boreholes have been done down to more than 1000 metres using diamond core drilling. This method can be also used in Finland and equipment for the drilling are available. One of the main targets of the investigation is to clarify the true strike and dip of fractures and other discontinuities. The methods used abroad are taking of oriented cores, borehole television survey and geophysical measurements. TV-survey and geophysical methods seem to be most favourable in deep boreholes. Also the accurate position (inclination, bearing) of the borehole is essential to know and many techniques are used for measuring of it. Investigations performed on the core samples include core logging and laboratory tests. For the core logging there is no uniform practice concerning the nuclear waste investigations. Different counries use their own classifications. All of these, however, are based on the petrography and fracture properties of the rock samples. Laboratory tests (petrographical and rock mechanical tests) are generally performed according to the recommendations of international standards. The large volumes of data obtained during investigations require computer techniques which allow more comprehensive collection, storage and processing of data. This kind of systems are already used in Sweden and Canada, for instance, and they could be utilize in Finland

Equilibrium Structures and Absorption Spectra for SixOy Molecular Clusters using Density Functional Theory

Science.gov (United States)

2017-05-05

Naval Research Laboratory Washington, DC 20375-5320 NRL/MR/6390--17-9724 Equilibrium Structures and Absorption Spectra for SixOy Molecular Clusters...TELEPHONE NUMBER (include area code) b. ABSTRACT c. THIS PAGE 18. NUMBER OF PAGES 17. LIMITATION OF ABSTRACT Equilibrium Structures and Absorption...and electronic excited-state absorption spectra for eqilibrium structures of SixOy molecular clusters using density function theory (DFT) and time

Online video marketing plan for a product launch : Case company: Altal Oy

OpenAIRE

Ngo, Quan

2015-01-01

This bachelor’s thesis aims to create a detailed guideline for using marketing videos in various online channels for Altal Oy. The commissioning company, Altal Oy is a Finn-ish start-up founded by three employees in 2014. It operates in the smart home tech-nology sector. The thesis topic is based on the current needs of the company to pro-mote for the launch of its brand new products in the Finnish market. A real project of making a promotion video is conducted and reported in parallel with t...

Markkinointiviestintäsuunnitelma: Charlotta-Production Oy

OpenAIRE

Wikman, Carita

2015-01-01

Tämän opinnäytetyön tarkoituksena on luoda markkinointiviestintäsuunnitelma hollolalaiselle askartelutukulle ja vähittäismyymälälle Charlotta-Production Oy:lle. Suunnitelman tavoitteena on aktivoida etenkin vanhoja yritysasiakkaita, mutta myös yksityisasiakkaita, sekä tavoittaa uusia potentiaalisia asiakkaita erityisesti digitaalisiin markkinointikeinoihin painostaen. Opinnäytetyö jakaantuu teoreettiseen ja empiiriseen osuuteen. Teoriaosuudessa käsitellään markkinointiviestintää kokonaisu...

Viestintäsuunnitelma : Lippupalvelu Oy

OpenAIRE

Olin, Noora

2015-01-01

Tämän opinnäytetyön tavoitteena oli rakentaa viestintäsuunnitelma Lippupalvelu Oy:lle. Viestintäsuunnitelma alkaa ajallisesti siitä hetkestä, kun asiakas ostaa lipun verkkokaupasta, ja päättyy tapahtuman jälkimarkkinointiin. Tähän ajanjaksoon liittyvä uusi viestintäprosessi otettiin käyttöön vuoden 2015 aikana, mutta sitä haluttiin vielä kehittää eteenpäin. Tutkimuksessa selvitettiin tapahtumajärjestäjien mielipiteitä ja kehitysehdotuksia viestintäprosessista. Tavoitteena oli aikaansaada tark...

Drillings at Veitsivaara in Hyrynsalmi

International Nuclear Information System (INIS)

Hinkkanen, H.; Oehberg, A.

1990-04-01

According to Governmen's decision in principle Teollisuuden Voima Oy is obliged to make bedrock investigations for the final disposal of the spent fuel produced by its power plant in Olkiluoto. Areas in Kuhmo, Hyrynsalmi, Sievi, Konginkangas and Olkiluoto were selected for the preliminary site investigations to be carried out during years 1987-1992. In Veitsivaara, Hyrynsalmi the investigation program was started in April 1987. During years 1987-1988 a deep borehole (1002 m) and 4 and 500 m deep additional boreholes were core drilled in the area. Various parameters were measured from the flushing water during the drilling. Corelogging included collecting detailed data of fractures and determining the weathering degree and petrographical properties. Rock mechanical properties, uniaxial compressive strength, Young's modulus and Poisso's ratio were measured from core samples. The flushing water needed in the drillings was pumped from 100 m deep borehole wells drilled with down-the-hole method in the vicinity of the borehole. The water was labeled with 2 tracers before use. About 75 m deep hole was percussion drilled near the borehole KR1. The spreading of the flushing water in the upper part of bedrock and the quality off the ground of the groundwater were studied by taking watersamples from the hole. 30 vertical holes were core drilled down to the depth of 10 m in bedrock with a light drilling unit. Drilling was carried out in order to determine the thickness of the overburden, to investigate the geophysical anomaly sources and to support geological mapping in areas covered with overburden. Groundwater hydraulics is one of the main subjects during the preliminary site investigation phase. For that reason 7 multilevel piezometers were installed on the site to monitore hydraulic head in 3 levels in the uppermost part of bedrock. The work consisted of borehole drillings to the depth of 100 m, geophysical borehole loggings and installation of piezometers. In addition

Loviisa nuclear power plant analyzer

International Nuclear Information System (INIS)

Porkholm, K.; Nurmilaukas, P.; Tiihonen, O.; Haenninen, M.; Puska, E.

1992-12-01

The APROS Simulation Environment has been developed since 1986 by Imatran Voima Oy (IVO) and the Technical Research Centre of Finland (VTT). It provides tools, solution algorithms and process components for use in different simulation systems for design, analysis and training purposes. One of its main nuclear applications is the Loviisa Nuclear Power Plant Analyzer (LPA). The Loviisa Plant Analyzer includes all the important plant components both in the primary and in the secondary circuits. In addition, all the main control systems, the protection system and the high voltage electrical systems are included. (orig.)

Investigation of the submodels for combustion; Polton osamallien kaeytettaevyys

Energy Technology Data Exchange (ETDEWEB)

Kjaeldman, L.; Huttunen, M.; Kyttaelae, J. [VTT Energy, Espoo (Finland)

1997-10-01

The capability for numerical analysis of flow, combustion and heat transfer in furnaces has been developed by improving the knowledge of the sensitivity of computed results on submodels recently implemented to the computational environment Ardemus owned by VTT Energy and Imatran Voima Oy. The submodels studied include models for combustion of gaseous (pyrolysed) fuel and for nitric oxide. The cases investigated included a gas flame and pulverized coal and peat combustion in single burner furnaces. The effect of grid refinement on the results was investigated for a corner fired power station furnace. (orig.)

Mean sea level and change in the hydrological regime off Loviisa power plant around the year 2050

International Nuclear Information System (INIS)

Maelkki, P.; Voipio, A.

1985-03-01

On the request of Imatran Voima Oy, the Institute of Marine Research has made an estimate on the future sea level off Loviisa Power Plant. The estimate is based on observationsof mean sea level in the Gulf of Finland. The stations used are Helsinki (observations since 1904) and Hamina (observations since 1928). A litterature review was made in order to estimate impact of climate change on environmental conditions. The results presented are mainly based on various estimates of meterorological Global Circulation Models (GCM). Their usefulness in the connection is briefly discussed

Salon Patarouva Oy : kertomus lounasravintolan perustamisesta

OpenAIRE

Aronen, Miina

2011-01-01

Salon Patarouva Oy on Salossa Meriniityn teollisuusalueella toimiva lounasravintola. Yrityksen perustaminen ja toiminnan alkuunsaattaminen oli lähtökohta tämän toiminnallisen opinnäytetyön tekemiseen. Omat lähtökohdat yrittäjäksi ryhtymiseen olivat vahva ammattitaito ja kokemus suurkeittiöympäristöistä. Vuokrasimme toimitilat ja ostimme uudet laitteet. Yrityslainan saimme Finnverasta ja pankista. Lainopillisia ja muita neuvoja saimme Yrityssalosta. Salon Pararouvan liikeidea on palvella...

Turvajalkineiden vaikutus jalkaterveyteen Suur-Savon Sähkötyö Oy:ssä

OpenAIRE

Pulkkinen, Riikka; Tossavainen, Inka; Ryynänen, Elina

2014-01-01

Opinnäytetyön tarkoituksena oli selvittää, millaisia vaikutuksia turvajalkineilla voi olla jalkaterveyteen, minkälaisia jalkaongelmia turvajalkineiden käyttö voi aiheuttaa, voivatko työntekijät itse vaikuttaa turvajalkineiden hankintaan ja millainen on turvajalkineiden hankintaprosessi Suur-Savon Sähkötyö Oy:ssä. Opinnäytetyömme toimeksiantaja oli Suur-Savon Sähkötyö Oy. Tutkimusmenetelmänä käytettiin pääasiassa kvantitatiivista tutkimusta ja aineisto kerättiin puolistrukturoidulla kysely...

5S-MENETELMÄN KÄYTTÖÖNOTTO SERVICEPOINTILLA

OpenAIRE

Ovaskainen, Tatu-Pekka

2014-01-01

Opinnäytetyö on tehty Servicepoint Kuopio Oy:n toimeksiannosta. Servicepoint Kuopio on teollisuuden kunnossapitoon automaatio- ja sähköistysprojekteihin sekä robottisovelluksiin erikoistunut yritys. Yrityksen kahden toimipisteen yhdistyminen lisäsi toimipisteellä henkilöstön ja työkalujen määrää. Yrityksessä toteutetaan sovelletusti Lean-ajattelumallia. Yksi Leanin työkaluista on 5S, joka haluttiin ottaa käyttöön yrityksessä. Opinnäytetyön tavoitteena oli ottaa käyttöön 5S-menetelmä Servi...

Pienpanimon tuotekehitys : Mathildedalin Panimo Oy

OpenAIRE

Lilja, Henri

2014-01-01

Tässä opinnäytetyössä selvitettiin, minkälaisia asioita Mathildedalin Panimo Oy:n tulee tuotekehityksessään ottaa huomioon. Opinnäytetyön kirjallisuusosa käsittelee oluen määritelmää ja historiaa, valmistusprosessia pienpanimossa, sekä lainsäädäntöä ja myyntikanavia. Työn teoreettisena pohjana on käytetty Business Model Canvas -nimistä liiketoiminnan suunnittelutyökalua, joka toimii tuotekehityksen tukena. Verotus, tiukka lainsäädäntö ja kiristyvä kilpailu ovat pienpanimon toiminnan suuri...

Hoivia Oy:n ruokapalvelujen asiakaskunnan laajentaminen

OpenAIRE

Sipilä, Satu

2014-01-01

Tämä opinnäytetyö syntyi Hoivia Oy:n tarpeesta markkinoida omia ruokapalveluitaan talon ulkopuolisille ja samalla kehittää lounasravintola Ludvigin palveluliiketoimintaa. Keskusteluissa toimitusjohtajan kanssa päätettiin selvityksen kohderyhmäksi ottaa kokonaisia palveluyksiköitä eikä yksittäisiä henkilöitä. Markkinatutkimukseen haluttiin mukaan kehitysvammaisten asuntoloita, päivätoimintayksiköitä, yksityisiä vanhainkoteja ja palvelutaloja sekä päiväkoteja. Ludvigissa on resursseja valmistaa...

Balanced scorecard:Case ConnectedDay Oy

OpenAIRE

Alatalo, Teija

2009-01-01

Markkinoiden kasvaminen ja jatkuvasti koveneva kilpailu ovat luoneet yrityksille paineita tulevaisuuden kilpailukyvyn ylläpitämiselle, mikä edellyttää yrityksiltä uudenlaisia työkaluja liiketoiminnan seuraamiselle. Tämän opinnäytetyön aiheena oli soveltaa yritysmaailmassa yleisesti käytettyä seurantajärjestelmää, balanced scorecardia case –yrityksenä käytetyn ConnectedDay Oy:n liiketoimintaan. Robert Kaplan ja David Norton suunnittelivat 1990 –luvun alussa balanced scorecard -mallin eli tasap...

Quasiparticle Scattering off Defects and Possible Bound States in Charge-Ordered YBa_{2}Cu_{3}O_{y}.

Science.gov (United States)

Zhou, R; Hirata, M; Wu, T; Vinograd, I; Mayaffre, H; Krämer, S; Horvatić, M; Berthier, C; Reyes, A P; Kuhns, P L; Liang, R; Hardy, W N; Bonn, D A; Julien, M-H

2017-01-06

We report the NMR observation of a skewed distribution of ^{17}O Knight shifts when a magnetic field quenches superconductivity and induces long-range charge-density-wave (CDW) order in YBa_{2}Cu_{3}O_{y}. This distribution is explained by an inhomogeneous pattern of the local density of states N(E_{F}) arising from quasiparticle scattering off, yet unidentified, defects in the CDW state. We argue that the effect is most likely related to the formation of quasiparticle bound states, as is known to occur, under specific circumstances, in some metals and superconductors (but not in the CDW state, in general, except for very few cases in 1D materials). These observations should provide insight into the microscopic nature of the CDW, especially regarding the reconstructed band structure and the sensitivity to disorder.

Ohjelmapalveluiden tuotekortit Lahti Travel Oy:n kokouspalveluhakemistoon

OpenAIRE

Lahti, Kirsi; Lehtola, Senni

2010-01-01

Tämä toiminnallinen opinnäytetyö käsittelee matkailun ohjelmapalveluiden markkinointia ja tuotteistamista. Opinnäytetyön tuloksena syntyi ohjelmapalveluiden tuotekortteja toimeksiantajan uudistetuille Internet-sivuille. Toimeksiantajana on Lahti Travel Oy, ja työn produkti eli tuotekortit ovat osa toimeksiantajan Internet-sivujen uudistusta. Tuotekortit sijoittuvat Internet-sivuilla kokouspalveluiden palveluhakemistoon ja siellä tarkemmin oheispalveluiden osioon. Työssä on toiminnallisen työn...

Internetmarkkinoinnin mahdollisuudet : case NaNi Group Oy

OpenAIRE

Arvonen, Johanna

2014-01-01

Tämän opinnäytetyön tarkoituksena oli tutkia internetmarkkinoinnin mahdollisuuksia kolmessa internetpalvelussa pienten koirien vaatteita valmistavalle NaNi Group Oy:lle. Opinnäytetyö toimii hyödyllisenä oppaana toimeksiantajan internetmarkkinoinnin suunnittelussa. Lisäksi opinnäytetyössä arvioidaan toimeksiantajan Facebookissa ja YouTubessa suorittamaa mainontaa. Kampanjoiden tulosten arviointi ja vertailu antavat suuntaa sille, miten kampanjoita kannattaa toteuttaa tulevaisuudessa. Opinn...

Customer Orientation vs. Customer Orientation Perception : Case J & J Lakkapää Oy Tornio

OpenAIRE

Angeria, Heli

2011-01-01

Heli, Angeria 2011. Customer Orientation vs. Customer Orientation Perception. Case: J & J Lakkapää Oy Tornio. Kemi-Tornio University of Applied Sciences. Business and Culture. Pages 51. Appendices 5. The objective of this thesis is to study customer orientation with the help of a widely adapted Selling-Orientation-Customer Orientation (SOCO) scale, in order to find out what is the extent to which J & J Lakkapää Oy Tornio and its consumer customers agree or disagree about the company’s cus...

Laadun parantaminen jalkinetuotannossa : case: Reima Oy

OpenAIRE

Huurre, Helena

2014-01-01

Tämä opinnäytetyö käsittelee laatua ja laadun kehittämisen työkaluja yrityksissä. Opinnäytetyön tarkoituksena oli tehdä kotimaiselle lastenvaatevalmistajalle, Reima Oy:lle, laatukäsikirja jalkinetuotantoa varten sekä laatukäsikirjaan sisältyvä laatuvaatimustaulukko, mistä jalkineiden laadulliset vaatimukset on löydettävissä. Laatukäsikirjan tavoitteena oli yhdistää Reiman jalkineen tuotekehitysprosessin kuvaus ja yleiset jalkinevaatimukset koko tuotantoprosessin ajalta yhtenäiseksi dokume...

Assessing demand for physical objects among marketing agencies : market research for Alphaform RPI Oy

OpenAIRE

Popova, Oxana

2012-01-01

The market for 3D printing services is projected to grow significantly. A service provider of 3D printing and rapid prototyping services, Alphaform RPI Oy recognizes that there are numerous growth opportunities that can be exploited. A market research was initiated by the case company to look for any emerging trends to use physical objects. The study was focused on exploring marketing agencies and their vision on 3D printing and rapid prototyping services. Theoretical framework was built ...

Sosiaalisen median markkinoinnin vuosikello Weecos Oy:lle

OpenAIRE

Heinämäki, Lotta; Huuskonen, Leena

2015-01-01

Opinnäytetyön tarkoitus oli luoda kokonaisvaltainen ja selkeä suunnitelma Weecos Oy:n markkinointitoimenpiteille valituissa sosiaalisen median kanavissa. Weecos on vuonna 2012 perustettu ekologisia yrityksiä yhteen keräävä verkkokauppa-alusta. Pienestä koostaan johtuen se ei ole pystynyt toteuttamaan sosiaalisen median markkinointia toivomallaan tavalla ja markkinoinnin suunnittelu ja toteutus on ollut epäsäännöllistä. Markkinointisuunnitelman tavoitteena oli helpottaa yrityksen markkinoi...

Tuotannon layoutin suunnittelu Flinkenberg OY:lle

OpenAIRE

Puotiniemi, Olli

2013-01-01

Opinnäytetyön aiheena on layoutsuunnittelu Flinkenberg Oy:lle. Yrityksen on tarkoitus tulevaisuudessa hankkia uusi työstökone nykyisiin toimitiloihin. Siitä seurasi tarve saada uusi layoutsuunnitelma tuotantohallista vanhan tilalle. Myös vanha layout kaipasi päivitystä. Opinnäytetyössä on pyritty soveltamaan jo olemassa olevaa teoriaa ja käytäntöä layoutsuunnittelussa. Teoriaa oli tarjolla paljon, ja olikin tärkeää osata perehtyä vain olennaiseen. Siihen tässäkin tutkimuksessa on pyritty....

M Road Oy: Arvo-ja laatudokumentti

OpenAIRE

Röynä, Mira

2017-01-01

Opinnäytetyön aiheena on yrityksen arvot ja laatutekijät. Opinnäytetyön tutkimuksellinen näkökulma syntyy konstruktiivisesta tutkimuksesta sekä kyselystä ja tiimityöskentelystä. Työssä tutustutaan tarkasti arvojen ja laatutekijöiden teoriaan, joiden pohjalta analysoidaan työntekijäkyselyn ja tiimityöskentelyn tuloksia. Työn tavoitteena on saada yhdistettyä franchaising-yritys M Road Oy:n arvot koko M Room –ketjun arvoja vastaavaksi, sekä parantaa yrityksen tuottamaa laatua. Toimeksiantaja...

High-speed photometry of the eclipsing dwarf nova OY Carinae

Science.gov (United States)

Cook, M. C.

1985-01-01

High-speed photometry of the eclipsing dwarf nova OY Car in the quiescent state is presented. OY Car becomes highly reddened during eclipse, with minimum flux colours inconsistent with optically thick emission in the U and B bandpasses. Mass ratios in the range 6.5 to 12 are required to reconcile the eclipse structure with theoretical gas stream trajectories. Primary eclipse timings reveal a significant decrease in the orbital period and the duration of primary eclipse indicates the presence of a luminous ring about the white dwarf. The hotspot eclipse reveals a hotspot which is elongated along the rim of the accretion disc, with optical emission being non-uniformly distributed along the rim. The location of the hotspot in the accretion disc implies a disc radius larger than that of an inviscid disc, with variation in the position of the hotspot being consistent with a fixed stream trajectory.

Results of Monitoring at Olkiluoto in 2011. Environment

Energy Technology Data Exchange (ETDEWEB)

Haapanen, A. [Haapanen Forest Consulting, Vanhakylae (Finland)

2012-11-15

This Working Report presents the main results of Posiva Oy's environmental monitoring programme on Olkiluoto Island in 2011. These summary reports have been published since 2005. The environmental monitoring system supervised by Posiva Oy produces input for biosphere modelling for long-term safety purposes as well as for monitoring the state of the environment during the construction (and later operation) of ONKALO underground rock characterization facility. Part of the monitoring is performed by the company running the nuclear power plants on the island, Teollisuuden Voima Oy (TVO). Monitoring has been carried out for varying periods of time depending on the sector: some monitoring activities performed by TVO originate from the 1970s and the repository-related environmental monitoring of Olkiluoto from the early 2000s. The monitoring programme evolves according to experiences gained from the modelling work and increased understanding of the site. Monitoring activities in 2011 proceeded according to the plans. The land-use of the island continues to change due to the construction work of OL3, ONKALO and related infrastructure, but the remaining natural environment resembles other coastal locations. The amount of nitrogen in the bulk deposition increased in 2011, whereas that of sulphur decreased. Some litterfall fractions showed higher Al and Fe values than earlier, likely caused by soil dust. Proximity of the sea is seen in wet deposition and soil solution results. Soil solution also reflects the young age of soils. Undestorey vegetation has shown no essential changes during the monitoring period. Mammalian fauna on the island is typical of coastal areas in Southwestern Finland. Game catches vary according to hunting pressure and natural variation in populations. The condition of the nearby sea is affected by the continuous land uplift, the shallowness of the area, the weather conditions, the general condition of the Bothnian Sea, the nutrient and sediment loads

Results of Monitoring at Olkiluoto in 2011. Environment

International Nuclear Information System (INIS)

Haapanen, A.

2012-11-01

This Working Report presents the main results of Posiva Oy's environmental monitoring programme on Olkiluoto Island in 2011. These summary reports have been published since 2005. The environmental monitoring system supervised by Posiva Oy produces input for biosphere modelling for long-term safety purposes as well as for monitoring the state of the environment during the construction (and later operation) of ONKALO underground rock characterization facility. Part of the monitoring is performed by the company running the nuclear power plants on the island, Teollisuuden Voima Oy (TVO). Monitoring has been carried out for varying periods of time depending on the sector: some monitoring activities performed by TVO originate from the 1970s and the repository-related environmental monitoring of Olkiluoto from the early 2000s. The monitoring programme evolves according to experiences gained from the modelling work and increased understanding of the site. Monitoring activities in 2011 proceeded according to the plans. The land-use of the island continues to change due to the construction work of OL3, ONKALO and related infrastructure, but the remaining natural environment resembles other coastal locations. The amount of nitrogen in the bulk deposition increased in 2011, whereas that of sulphur decreased. Some litterfall fractions showed higher Al and Fe values than earlier, likely caused by soil dust. Proximity of the sea is seen in wet deposition and soil solution results. Soil solution also reflects the young age of soils. Undestorey vegetation has shown no essential changes during the monitoring period. Mammalian fauna on the island is typical of coastal areas in Southwestern Finland. Game catches vary according to hunting pressure and natural variation in populations. The condition of the nearby sea is affected by the continuous land uplift, the shallowness of the area, the weather conditions, the general condition of the Bothnian Sea, the nutrient and sediment loads

Results of Monitoring at Olkiluoto in 2011. Environment

Energy Technology Data Exchange (ETDEWEB)

Haapanen, A. (ed.) [Haapanen Forest Consulting, Vanhakylae (Finland)

2012-11-15

This Working Report presents the main results of Posiva Oy's environmental monitoring programme on Olkiluoto Island in 2011. These summary reports have been published since 2005. The environmental monitoring system supervised by Posiva Oy produces input for biosphere modelling for long-term safety purposes as well as for monitoring the state of the environment during the construction (and later operation) of ONKALO underground rock characterization facility. Part of the monitoring is performed by the company running the nuclear power plants on the island, Teollisuuden Voima Oy (TVO). Monitoring has been carried out for varying periods of time depending on the sector: some monitoring activities performed by TVO originate from the 1970s and the repository-related environmental monitoring of Olkiluoto from the early 2000s. The monitoring programme evolves according to experiences gained from the modelling work and increased understanding of the site. Monitoring activities in 2011 proceeded according to the plans. The land-use of the island continues to change due to the construction work of OL3, ONKALO and related infrastructure, but the remaining natural environment resembles other coastal locations. The amount of nitrogen in the bulk deposition increased in 2011, whereas that of sulphur decreased. Some litterfall fractions showed higher Al and Fe values than earlier, likely caused by soil dust. Proximity of the sea is seen in wet deposition and soil solution results. Soil solution also reflects the young age of soils. Undestorey vegetation has shown no essential changes during the monitoring period. Mammalian fauna on the island is typical of coastal areas in Southwestern Finland. Game catches vary according to hunting pressure and natural variation in populations. The condition of the nearby sea is affected by the continuous land uplift, the shallowness of the area, the weather conditions, the general condition of the Bothnian Sea, the nutrient and sediment

Transpro Oy:n koulutustenhallintajärjestelmä

OpenAIRE

Pöntinen, Mika

2011-01-01

Opinnäytetyössä toteutetaan koulutustenhallintajärjestelmä tuusulalaiselle Transpro Oy:lle, joka on kuljettajien ammattipätevyyskoulutuksiin erikoistunut yritys. Järjestelmän tehtävänä on tarjota työkalu, jolla voidaan hallita tietoja koulutustapahtumista, koulutustapahtumiin osallistuvista oppilaista, kouluttajista ja asiakkaista. Järjestelmä on toteutettu WWW-sovelluksena. Pääohjelmointikieli on PHP ja järjestelmän dynaamiset toiminnot on toteutettu JavaScript-kielellä. Sivuston haku j...

Selvitys kulttuuriyrityksen laajentumis- mahdollisuuksista. Saiffa Oy:n tapaus

OpenAIRE

Sirviö, Jussi

2013-01-01

Opinnäytetyö selvittää kulttuuriyrityksen laajentumismahdollisuuksia. Teoreettinen viitekehys on franchisingtoimintamalli, mutta selvitys on avoin myös muille toimivaksi osoittuville malleille. Opinnäytetyö on tapaustutkimus katutanssikeskus Saiffa – Flow Mo Dance Schoolille, jota johtaa Saiffa oy. Vuonna 2010 perustettu Saiffa on helsinkiläinen tanssikoulu, joka on erikoistunut katutanssilajeihin, kuten breakdance, popping locking ja hip hop. Sen toimialaan kuuluu tanssiopetuksen lis...

Transfer of NPP technology from Finland fo Hungary

Energy Technology Data Exchange (ETDEWEB)

Varis, M. V.K. [Imatran Voima Oy, Vantaa (Finland); Frigyesi, F. [Paksi Atomeroemue Vallalat (Hungary)

1989-07-15

Imatran Voima Oy (IVO), which accounts for 45% of the total Finnish electricity supply, have their own architect-engineering capacity. This know-how is also available internationally (IVO International). This report explains how technology is transferred to the client's organisation using the advantages of the client's own organization culture, supplemented by IVO's experience. The technology transferred to the Hungarian Paks Nuclear Power Company (PAV) regarding project management services is a good example. A materials management example explains the method. The customer is familiarized via wall chart on which the useful features in IVO's system are added.

Miesten muodin verkkokauppakonsepti : case: FRENN Company Oy

OpenAIRE

Hautaniemi, Tapio

2015-01-01

Tämä opinnäytetyö käsittelee miesten muodin verkkokaupan perustamista ja kehittämistä. Työ on tehty toimeksiantona vuoden 2013 syksyllä lanseeratulle suomalaisella miesten vaatemerkille Frenn Company Oy:lle. Opinnäytetyön aihe valikoitui toimeksiantajan toiveiden pohjalta sekä omasta mielenkiinnostani ja kokemuksestani aiheeseen. Työn tavoitteena on selvittää, millainen on kansainvälinen miesten muodin verkkokauppa ja miten alan muut toimijat ovat verkkokauppansa toteuttaneet. Opinnäytet...

KOKEMUKSET ASIAKKUUKSIEN JOHTAMISESTA : Case: Buildercom Oy partneriasiakkaat

OpenAIRE

Porkka, Elina

2013-01-01

Tässä opinnäytetyössä tutkittiin, miten Buildercom Oy:n partneritason asiakkaat olivat kokeneet heihin kohdistetun asiakkuuden hoitomallin. Tavoitteena oli tutkia, miten asiakkaat oli tällä hetkellä jaettu erilaisiin ryhmiin, millaista lisäarvoa hoitomallilla on mahdollisesti kyetty asiakkaille tuottamaan, ja mitkä tekijät ovat vaikuttaneet asiakkaiden kokemusten laatuun. Tutkimus toteutettiin kokonaisuudessaan kevään 2012 ja kevään 2013 välisenä aikana. Tutkimus suoritettiin, koska toimeksia...

Markkinointiviestintäsuunnitelma Case: MK Kivipiha Oy

OpenAIRE

Ruohomaa, Sami

2011-01-01

Tämän opinnäytetyön tarkoituksena on ollut markkinointiviestintäsuunnitelman laatiminen MK Kivipiha Oy:lle. Opinnäytetyö toteutettiin projektityönä, jonka lisäksi benchmarkkaus osiossa hyödynnettiin kvalitatiivista eli laadullista analyysiä. Lähtökohtana pidettiin suunnitelman realistisuutta ja käytännön toteuttamisen mahdollisuutta. Opinnäytetyö rakentuu kahdesta eri osiosta: teoreettisesta viitekehyksestä sekä empiirisestä osuudesta. Teoriana käytettiin katsausta perinteisen markkinointivie...

Vähittäismyymälän markkinoinnin kehittäminen : case Mattomies Oy

OpenAIRE

Hellstén, Maren

2015-01-01

Opinnäytetyön tavoitteena oli löytää toimeksiantajana toimineelle Mattomies Oy Ruohola & Ruoholalle konkreettisia keinoja yrityksen vähittäismyymälän markkinoinnin kehittämiseksi. Teoriaosassa käytiin läpi 4P-mallin mukaiset markkinoinnin kilpailukeinot. Toteutusosassa analysoitiin Mattomies Oy Ruohola & Ruoholan myymälän nykytilaa, myymälän merkitystä liiketoiminnalle sekä koottiin myymälälle markkinointimix, jossa annettiin konkreettisia ehdotuksia markkinoinnin kehittämiseksi. Ehdo...

Työohjeistus Perälän Turve Oy

OpenAIRE

Haukka, Arto

2014-01-01

Perälän Turve Oy:n uusille työntekijöille laadittu työohjeistus ja siitä raportointi. Työhön kuuluu kaksi osaa, word-raportti sekä powerpoint-työohjeistuskansio. Powerpoint-tiedosto upotettuna tiedostona alkuperäisessä word-tiedostossa.

Asiakkuudenhallinnan avulla lisäarvoa asiakkaalle ja lisämyyntiä yritykselle : case Naaantalin Matkailu Oy

OpenAIRE

Korpiranta, Mirka

2012-01-01

Tässä opinnäytetyössä on selvitetty, miten Naantalin Matkailu Oy voi asiakkuudenhallinnan avulla luoda lisäarvoa asiakkaille ja tuottaa lisänmyyntiä yritykselle. Lisäksi on selvitetty, miten Winres-matkatoimistojärjestelmää voidaan hyödyntää asiakkuudenhallinnassa tavoitteiden saavuttamiseksi. Vertailukohteina käytettiin muita matkailutoimistoja, jotka muistuttavat Naantalin Matkailu Oy:tä ja Naantalia kohteena. Opinnäytetyön vertailukartoitus toteutettiin kyselynä, jonka tarkoituksena oli...

Työterveys- ja työturvallisuusjohtamisjärjestelmä : Case Tapojärvi Oy

OpenAIRE

Honkanen, Päivi

2011-01-01

Honkanen, Päivi. 2011. Työterveys- ja työturvallisuusjohtamisjärjestelmä. Case Tapo-järvi Oy. Opinnäytetyö. Kemi-Tornion ammattikorkeakoulu. Kaupan ja kulttuurin toimiala. Sivuja 33. Tämän opinnäytetyön tavoitteena on selvittää Tapojärvi Oy:n työterveys- ja työturvalli-suusjohtamisjärjestelmän nykytilaa ja arvioida kuinka se vastaa standardin OHSAS 18001:2007 vaatimuksia. Opinnäytetyö koostuu teoriaosuudesta ja empiirisestä tutkimuksesta. Teoriaosassa käsit-telen turvallisuusjohtamis...

Hissihuollon kehittäminen ThyssenKrupp hissit Oy

OpenAIRE

Nieminen, Jani

2013-01-01

Insinöörityössä kartoitettiin ThyssenKrupp Hissit Oy:n kunnossapidon tämänhetkinen toi-mintamalli ja sen pohjalta laadittiin kunnossapidon kehityssuunnitelman runko. Thyssen-Krupp Hissien huoltotoiminta on kasvanut voimakkaasti muutamassa vuodessa pienestä helposti johdettavasta yksiköstä usean henkilön johtamaksi organisaatioksi. Haastattelujen, vikatilastojen ja oman kokemukseni avulla päädyin siihen, että nykyinen hissien kunnossapito perustuu voimakkaasti korjaavaan kunnossapitoon ja ...

Characteristics of NixFe1−xOy Electrocatalyst on Hematite as Photoanode for Solar Hydrogen Production

Directory of Open Access Journals (Sweden)

Chih-Ping Yen

2017-11-01

Full Text Available The use of hematite as the photoanode for photoelectrochemical hydrogen production by solar energy has been actively studied due to its abundance, stability, and adequate optical properties. Deposition of an electrocatalyst overlayer on the hematite may increase kinetics and lower the onset potential for water splitting. NixFe1−xOy is one of the most effective electrocatalysts reported for this purpose. However, the condition and results of the previous reports vary significantly, and a comprehensive model for NixFe1−xOy/hematite is lacking. Here, we report a simple and novel chemical bath deposition method for depositing low-onset-potential NixFe1−xOy electrocatalyst on hematite. With a Ni percentage of 80% and an immersion time of 2 min, the as-prepared NixFe1−xOy overlayer raised the photovoltage from 0.2 V to 0.7 V, leading to a cathodic shift of the onset potential by 400 mV, while maintaining the same level of current density. The dependence of the electrochemical and photoelectrochemical characteristics of the photoanode on the condition of the electrocatalyst was studied systematically and explained based on energy level diagrams and kinetics.

Effects of Ni-5%RExOy Composite Additives on Electrochemical Hydrogen Storage Performances of Mg2Ni

Directory of Open Access Journals (Sweden)

ZHANG Guo-fang

2017-11-01

Full Text Available The Ni-5%RExOy (CeO2, La2O3, Eu2O3 as composite additives, Mg2Ni-Ni-5%RExOy composites were prepared by the ball milling method. The effects of different additives on the structure, morphology, electrochemistry and kinetic properties of Mg2Ni alloy were studied systematically. The results show that composite additives can improve the proportion of amorphous and nanocrystalline structure of Mg2Ni alloy. The particle size is homogeneous but the agglomeration is observed in the sample with Ni-5%CeO2 additives. The composites with additives show higher maximum discharge capacity and better cycle stabilities. All of these three kinds of composite additives can improve the kinetic properties of the composites effectively, including optimizing the charge-transfer ability, the reversibility of the electrochemical reaction on the alloy surface, and enhancing the diffusion coefficients of H atoms in the bulk of alloy. Among these three kinds of additives, Ni-5%CeO2 additive shows the best catalysis effect on promoting the kinetic properties of the composites.

Social Media Marketing : CASE: OY SUOMEN LYYRA AB

OpenAIRE

Eriksson, Irene

2012-01-01

This bachelor thesis was commissioned by Oy Suomen Lyyra Ab, the largest student online media and student card producer for higher education students in Finland. The the-sis objective was to understanding the current social media situation and activity among the students of higher education in Finland, the social media networks that the case company currently uses as well as understanding how to use these networks for successful marketing activities. The quantitative research was conducted in...

Integrated Brand Promotion – Advertisement for STADIUM OY

OpenAIRE

Mai, Tung

2014-01-01

This project-based thesis is an advertisement for Stadium Oy. In 23 years of leading marketing department, Stadium’s marketing manager had to say this is a pioneer time when Stadium’s outsources an advertising project to an external individual resource. The project, therefore, consists a number of business partners so that professional quality is guaranteed to deliver. STADIUM’s advertising strategy stays committed to the company’s business model and its mission. This directly affec...

Henkilöstötilinpäätös Yritys Oy

OpenAIRE

Turunen, Tytti

2011-01-01

Tämän opinnäytetyön aiheena oli Yritys Oy:n henkilöstötilinpäätös vuodelta 2009. Henkilöstötilinpäätös jaetaan kahteen osaan, henkilöstötuloslaskelmaan ja henkilöstökertomukseen. Tavoitteena oli laatia Yritys Oy:lle henkilöstötilinpäätöksestä malli, jota yritys voi käyttää ja hyödyntää tulevaisuudessa henkilöstötilinpäätöstä laatiessa. Tavoitteena oli myös antaa kehittämisehdotuksia, minkälaista tietoa kannattaa kerätä seuraavia henkilöstötilinpäätöksiä varten. Henkilöstötilinpäätöksen tavoit...

Sisäisen viestinnän kehittäminen osana parempia kampanjatoteutuksia : Case: Nike Finland Oy

OpenAIRE

Kankkunen, Eeva

2014-01-01

Tämä opinnäytetyö tehtiin Nike Finland Oy:lle, ja sen aiheena oli sisäisen viestinnän kehittäminen osana parempia kampanjatoteutuksia. Nike Finland Oy on maailman tunnetuimman urheilubrändin Niken Suomen organisaatio, ja tällä hetkellä se on toisena vastaavien tuotteiden valmistajana Suomen markkinoilla. Yritys tunnetaan niin Suomessa kuin maailmallakin kuuluisasta ja uniikista Niken logosta, ja sloganista ”Just do it”. Koska yritys on aina enemmän kuin pelkästään sen hyvä yritysimag...

Hierarchical α-MnO2 nanowires@Ni1-x Mnx Oy nanoflakes core-shell nanostructures for supercapacitors.

Science.gov (United States)

Wang, Hsin-Yi; Xiao, Fang-Xing; Yu, Le; Liu, Bin; Lou, Xiong Wen David

2014-08-13

A facile two-step solution-phase method has been developed for the preparation of hierarchical α-MnO2 nanowires@Ni1-x Mnx Oy nanoflakes core-shell nanostructures. Ultralong α-MnO2 nanowires were synthesized by a hydrothermal method in the first step. Subsequently, Ni1-x Mnx Oy nanoflakes were grown on α-MnO2 nanowires to form core-shell nanostructures using chemical bath deposition followed by thermal annealing. Both solution-phase methods can be easily scaled up for mass production. We have evaluated their application in supercapacitors. The ultralong one-dimensional (1D) α-MnO2 nanowires in hierarchical core-shell nanostructures offer a stable and efficient backbone for charge transport; while the two-dimensional (2D) Ni1-x Mnx Oy nanoflakes on α-MnO2 nanowires provide high accessible surface to ions in the electrolyte. These beneficial features enable the electrode with high capacitance and reliable stability. The capacitance of the core-shell α-MnO2 @Ni1-x Mnx Oy nanostructures (x = 0.75) is as high as 657 F g(-1) at a current density of 250 mA g(-1) , and stable charging-discharging cycling over 1000 times at a current density of 2000 mA g(-1) has been realized. © 2014 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

Creating strategic brand management manual : Case: Tukikallio Oy

OpenAIRE

Vepsäläinen, Sini

2010-01-01

The purpose of this Thesis was to create a Strategic Brand Management Manual for the use of a new company called Tukikallio Oy. The manual was composed of the elements that the company wanted to include in it. A qualitative research method was used in collecting information. The theoretical part of the thesis includes theories about brand building. Brand identity tells what the brand really is with its positive and negative sides, image reflects on what kind of things the consumer associa...

JENDL-3.3 thermal reactor benchmark test

International Nuclear Information System (INIS)

Akie, Hiroshi

2001-01-01

Integral tests of JENDL-3.2 nuclear data library have been carried out by Reactor Integral Test WG of Japanese Nuclear Data Committee. The most important problem in the thermal reactor benchmark testing was the overestimation of the multiplication factor of the U fueled cores. With several revisions of the data of 235 U and the other nuclides, JENDL-3.3 data library gives a good estimation of multiplication factors both for U and Pu fueled thermal reactors. (author)

Integrated production of merchantable wood and wood fuels in industry; Teollisuuden ainespuun ja puupolttoaineen integroitu tuotanto

Energy Technology Data Exchange (ETDEWEB)

Kuvaja, K [Enso Oy, Imatra (Finland). Forest Dept.

1997-12-01

The aim of this project is the economically profitable integrated harvesting of industrial wood and firewood especially in harvesting of small-diameter first thinning wood. The research in 1994 was concentrated on improvement of the quality of the chipping methods based on chain-flail debarking chipping method, and on determination of the possible utilisation targets for the fuel fraction. A reasonably large drum debarking test was also carried out at the industrial scale debarking station of the Enocell Oy. More than 80 000 m{sup 3} of first thinning wood was delivered by Enocell during this project. The quality of wood chips, produced using the chain-flail delimbing method, could be improved in the case of pine nearly to the required quality level, but additional measures are still needed in the case of birch. The fuel fraction deliveries to different points of utilisation was started. The particle size of the fuel fraction appeared to be good after crushing. In 1995 a chain-flail-drum debarking chipping unit was developed to improve and homogenise the quality of chips. (orig.)

Maintaining staff competence-a NPP operator viewpoint

International Nuclear Information System (INIS)

Patrakka, E.

2000-01-01

For a nuclear power plant operator, it is crucial to guarantee the safe and economic operation of the power plant as well as to look after the general acceptability of nuclear power. As to human resources management, this requires continuous maintenance and enhancement of the performance of the individuals and organisation. To this end, several development projects have recently been implemented by Teollisuuden Voima Oy (TVO) at the Olkiluoto nuclear power plant, which consists of twin 840 MWe BWR units that commenced their operation in 1978 and 1980. Systematic initial and continuing training programmes are needed to maintain the technical and managerial skills and know-how at a high level. The present stabile state of nuclear power, i.e. operation of ageing plants with personnel ageing as well, requires a variety of actions to reinforce the training efforts. At Olkiluoto NPP, we have carried out an extensive modernization programme that allowed the personnel to strengthen their knowledge and supplement it with the most recent results of development. We have also closely monitored the NPP development projects of the vendors, which has added to the preservation of know-how and understanding of advanced nuclear power technology. We have close contacts to the research institutes and universities, and have performed R and D activities to limited extent. In addition to the projects mentioned above, a co-ordinated development programme, 'TVO 2002', was initiated last year. The main objective of this programme is to ensure the functional preconditions and the competitiveness of the company in a changing environment. The management and operational procedures will be developed in such a way that the goals set for year 2002 will be achieved. The programme is organised as ten projects, which cover a variety of development subjects. One of the focal areas includes projects that can be characterised with the words 'Survey of competencies' and 'Preservation of know

Distribution strategy: : A case study of Plantui Oy

OpenAIRE

Korpinen, Sakari; Siltala, Roope

2015-01-01

Distribution strategy: A case study of Plantui Oy Year 2015 Pages 57 Plantui is a Finnish start-up company focusing on enabling customers to grow plants at home. Plantui operates in the field of design and food tech. Plantui’s product offering consists of the Smart Garden, a device in which the plants are grown. Another product which Plantui offers to its customers is the plant capsules from which the plants are grown, with the help of the Smart Garden device. Plantui’s selling propos...

OL1/OL2 License renewal for extended life-time: Class 1 piping load and strength analyses

International Nuclear Information System (INIS)

Lemettinen, P.

2015-01-01

Teollisuuden Voima Oyj (TVO) operates two NPP units Olkiluoto 1 (OL1) and Olkiluoto 2 (OL2), that are identical 880 MWe BWRs. The units were originally designed for 40 years life-time. TVO is applying license renewal for extended life-time for 60 years plant life. Part of the license renewal project is to evaluate and update all Class 1 piping load and strength analyses. These analyses are done with the help of TVO's in-house Piping and Component Analysis and Monitoring System (PAMS). PAMS is basically a database system, consisting of separate geometry, material, loading, result and document databases. The thermo hydraulic analysis program RELAP5 is used to obtain temperature, pressure and mass flow for the piping loading areas. The piping strength analysis are carried out mainly with the Finnish FPIPE FEA program for all related thermal transient and dynamic load cases. The slides of the presentation have been added to the paper

RFID-tunnisteiden havaitseminen liikkuvasta ajoneuvosta : case: Fidera Oy

OpenAIRE

Artukka, Riku

2016-01-01

Tämän opinnäytetyön tarkoituksena on testata ja dokumentoida suoritettuja RFID- ajoneuvomittauksia eli ajoneuvojen ja henkilöiden automaattista tunnistusta hyödyntämällä RFID-tunnisteita. Opinnäytetyö on tehty toimeksiantona Fidera Oy:lle. Yrityksellä on jo käytössä automaattinen henkilötunnistusjärjestelmä, tarve olikin soveltaa samaa tekniikkaa ajoneuvojen ja niiden sisällä olevien henkilöiden tunnistukseen. Opinnäytetyön tavoitteena on luoda kuva RFID-tekniikan perusteista ja toimia s...

Mainonnan huomioarvotutkimus. Case Nelipyörä Oy

OpenAIRE

Survonen, Juha

2013-01-01

Tämä opinnäytetyö käsittelee mainontaa markkinointiviestinnän keinona. Toimeksiantajana tällä opinnäytetyöllä on Hämeenlinnassa ja Hyvinkäällä toimiva autoliike, Nelipyörä Oy. Opinnäytetyön tarkoituksena on antaa toimeksiantajalle tietoa heidän mainontansa huomioarvoista käytettyjen mainosmedioiden osalta. Opinnäytetyön teoriaosuus koostuu markkinointiviestinnästä sekä mainonnasta. Markkinointiviestinnän osuus keskittyy markkinointiviestinnän suunnitteluun sekä tavoitteisiin. Mainonnan os...

Mazda kiintotyövaiheiden perustaminen : Delta Auto Oy

OpenAIRE

Schreck, Ville

2012-01-01

Täyden palvelun autoliikeketju Delta Auto myy ja huoltaa Kia-, Mitsubishi- ja Mazda-merkkisiä autoja paikkakuntakohtaisesti kahdessakymmenessäviidessä toimipisteessä. Opinnäytetyön tarkoituksena oli tehdä toimiva ratkaisu Mazdan työvaiheiden sekä työpakettien myyntiin Automaster-ohjelmistoon. Työvaiheet sekä ohjeajat tulevat käyttöön kaikkiin Delta Auton toimipisteisiin ympäri Suomea. Työn lähdemateriaalin sain Inchcape Motor Finland Oy:ltä, joka toimii Suomessa Mazdan maahantuojana. Tarve tä...

Insta Trust Oy:n tuottamien kriisiharjoitusten kehittäminen

OpenAIRE

Väinölä, Marko

2016-01-01

Insta Trust Oy toteuttaa asiakkaidensa kanssa kymmeniä kriisiharjoituksia vuosittain. Kriisiharjoitusten avulla pyritään kehittämään harjoittelevien tahojen kriisijohtamisen ja kriisiviestinnän osaamista, tunnistamaan kehittämistä vaativia toimintamalleja ja ohjeita sekä lisäämään organisaation avainhenkilöiden motivaatiota ja tietoisuutta kriisijohtamisesta ja ennakoivaa turvallisuustyötä kohtaan. Kriisiharjoitusten kehittämisessä on osattava kuunnella asiakkaiden tarpeita ja pysyttävä mukan...

Selvitys HR House Oy:n vuokratyöntekijöiden työtyytyväisyydestä

OpenAIRE

Kaskela, Leeni

2011-01-01

Opinnäytetyön tarkoituksena oli selvittää HR House Oy:n palveluksessa olevien vuokratyöntekijöiden työtyytyväisyyteen vaikuttavia asioita. Opinnäytetyö on laadullinen tutkielma, jossa käytän tutkimusmenetelmänä teemahaastattelua. Tutkimuskysymykseni selvittävät, mitkä asiat vaikuttavat työntekijöiden työ- tyytyväisyyteen ja mitä he kertovat työtyytyväisyyteen vaikuttavista tekijöistä. Toimeksianto opinnäytetyön tekemiseen on tullut HR House Oy Henkilöstöpalveluilta. Tavoitteena oli löytää kei...

Kanta-asiakasohjelman kehittäminen : case Crocs Stores Oy

OpenAIRE

Manninen, Mirka

2014-01-01

Kanta-asiakkuuden tavoitteena on molemminpuolinen arvon tuottaminen asiakkaan ja yrityksen välillä. Tämän tutkimuksen tavoitteena on selvittää, millainen kanta-asiakasohjelma olisi toimeksiantajayrityksen kannalta toimivin. Tutkimus on tapaus- eli case -tutkimus, jonka toimeksiantaja on Crocs Stores Oy. Toimeksiantajalla on tällä hetkellä käytössään kanta-asiakasohjelma, joka ei palvele yrityksen tarpeita. Paremman kanta-asiakasohjelman kehittämisen lisäksi tavoitteina on selvittää asiakkaide...

Dynamics of TRIGA-3 Salazar Reactor

International Nuclear Information System (INIS)

Gallardo S, L.F.

1990-01-01

The theoretical study of temporal behavior of a nuclear reactor is of great importance, since it allows to know, in advance, the conditions to which a reactor is going to be submitted. The reliability of two computer codes (AIREK-JEN and PLANKIN) designed to reproduce the temporal behavior of nuclear reactors, generally power reactors, when they are applied to reproduce the dynamic behavior of TRIGA-3 Salazar Reactor is analyzed. In the first chapters, the fundamental equations that solve this computer codes are deduced, and also the main characteristics of TRIGA-3 Salazar Reactor and the necessary data to run the programs are presented; later the results obtained with the computer codes and the experimental results reported in the operational logbook of the reactor are compared, with the result that such computer codes are applicable to the temporal study of TRIGA-3 Salazar Reactor. (Author)

Integrated production of merchantable wood and wood fuels in industry; Teollisuuden ainespuun ja puupoltto-aineen integroitu tuotanto

Energy Technology Data Exchange (ETDEWEB)

Kuvaja, K [Enso-Gutzeit Oy, Imatra (Finland). Forest Dept.

1997-12-31

The aim of this project is the economically profitable integrated harvesting of industrial wood and firewood especially in harvesting of small-diameter first thinning wood. The research in 1994 was concentrated on improvement of the quality of the chipping methods based on chain-flail debarking chipping method, and on determination of the possible utilization targets for the fuel fraction. A reasonably large drum debarking test was also carried out at the industrial scale debarking station of the Enocell Oy. More than 80 000 m{sup 3} of first thinning wood was delivered by Enocell during this project. The quality of wood chips, produced using the chain-flail delimbing method, could be improved in the case of pine nearly to the required quality level, but additional measures are still needed in the case of birch. The fuel fraction deliveries to different points of utilization was started. The particle size of the fuel fraction appeared to be good after crushing. In 1995 a chain-flail-dry drum debarking chipping unit was developed to improve and homogenize the quality of chips

Integrated production of merchantable wood and wood fuels in industry; Teollisuuden ainespuun ja puupoltto-aineen integroitu tuotanto

Energy Technology Data Exchange (ETDEWEB)

Kuvaja, K. [Enso-Gutzeit Oy, Imatra (Finland). Forest Dept.

1996-12-31

The aim of this project is the economically profitable integrated harvesting of industrial wood and firewood especially in harvesting of small-diameter first thinning wood. The research in 1994 was concentrated on improvement of the quality of the chipping methods based on chain-flail debarking chipping method, and on determination of the possible utilization targets for the fuel fraction. A reasonably large drum debarking test was also carried out at the industrial scale debarking station of the Enocell Oy. More than 80 000 m{sup 3} of first thinning wood was delivered by Enocell during this project. The quality of wood chips, produced using the chain-flail delimbing method, could be improved in the case of pine nearly to the required quality level, but additional measures are still needed in the case of birch. The fuel fraction deliveries to different points of utilization was started. The particle size of the fuel fraction appeared to be good after crushing. In 1995 a chain-flail-dry drum debarking chipping unit was developed to improve and homogenize the quality of chips

Sosiaalisen median merkitys mikroyrityksille - Case: TallFits Oy

OpenAIRE

Nousiainen, Ari; Koskivuori, Timo

2011-01-01

Tämän opinnäytetyön aiheena on sosiaalinen media mikroyritysten näkökulmasta. Työssä käsi-tellään sosiaalista mediaa ja markkinoinnissa tapahtunutta muutosta sekä niitä toimintatapoja, joita interaktiivinen mediaympäristö yrityksille ja sen asiakkaille tarjoaa. Työn tavoitteena on tuottaa tietoa sosiaalisen median hyödyistä ja haasteista mikroyrityksille ja lisätä koh-deyrityksemme TallFits Oy:n myyntiä ja löydettävyyttä. Tavoitteiden saavuttamiseksi opinnäytetyömme tarkoituksena oli tehd...

Kankaan tuotekehittäminen erikoisolosuhteisiin K&H Annala Oy:lle

OpenAIRE

Koski, Marjo

2017-01-01

Opinnäytetyö tehtiin yhteistyössä Suomen ainoan huonekalukankaita valmistavan yrityksen K&H Annala Oy:n kanssa. Työn aiheena oli suunnitella ja tuotteistaa kangasmallisto erikoisolosuhteisiin innovatiivisesta, Italiassa valmistettavasta Flyer-langasta. Projektin lähtökohtana oli opinnäytetyöntekijän kasvava kiinnostuneisuus kudottuja tekstiilejä kohtaan ja kankaan suunnitteluprosessin vieminen teollisen tuotannon puolelle. Projektissa tähdättiin uudentyyppiseen kankaaseen ilmeen ja materiaali...

The using of the control room automation against human errors

International Nuclear Information System (INIS)

Kautto, A.

1993-01-01

The control room automation has developed very strongly during the 80's in IVO (Imatran Voima Oy). The former work expanded strongly with building of the full scope training simulator to the Loviisa plant. The important milestones has been, for example the testing of the Critical Function Monitoring System, a concept developed by Combustion Eng. Inc., in Loviisa training simulator 1982, the replacing of the process and simulator computers in Loviisa 1989, and 1990 and the presenting the use of the computer based procedures in training of operators 1993. With developing of automation and procedures it is possible to minimize the probability of human error. However, it is not possible totally eliminate the risks caused by human errors. (orig.)

Enhanced thermoelectric properties of nano SiC dispersed Bi2Sr2Co2Oy Ceramics

Science.gov (United States)

Hu, Qiujun; Wang, Kunlun; Zhang, Yingjiu; Li, Xinjian; Song, Hongzhang

2018-04-01

The thermoelectric properties of Bi2Sr2Co2Oy + x wt% nano SiC (x = 0.00, 0.025, 0.05, 0.1, 0.2, and 0.3) prepared by the solid-state reaction method were investigated from 300 K to 923 K. The resistivity can be reduced effectively by adding a small amount of SiC nano particles, which is attributed to the increase of the carrier concentration. At the same time, the Seebeck coefficients can be improved effectively due to the energy filtering effect that low energy carriers are strongly dispersed at the interface between the SiC nano particles and the matrix. The decrease of thermal conductivity is due to the increase of the scattering ability of the phonons by the SiC nanoparticles distributed at the boundary of the matrix. As a result, the Bi2Sr2Co2Oy + x wt% SiC composites exhibit better thermoelectric properties. The maximum ZT value 0.24 is obtained when x = 0.05 at 923 K. Compared with the sample without SiC nano particles, the ZT value is increased by about 59.7%.

Organizational Challenge of Posiva’s Final Disposal Programme: From an R&D Organization to a Project Organization, and Further Towards an Operational Organization

International Nuclear Information System (INIS)

Mokka, J.

2016-01-01

Full text: Posiva Oy is an expert organization established in 1995 and responsible for the final disposal of the spent nuclear fuel of its owners. Posiva currently employs around 100 people and has a turnover of some 63 million (2015). The company headquarters are located in Olkiluoto in the municipality of Eurajoki, Finland. Posiva is owned by two Finnish NPP operators Teollisuuden Voima Oyj (60%) (TVO) and Fortum Power & Heat Oy (40%), both of which are responsible for their costs of nuclear waste management. The Finnish final disposal programme has a long history. When NPP unit Olkiluoto 1 renewed its operating licence for the first time in 1983, TVO presented a programme showing final disposal to commence in the 2020s. In the 1980s and 1990s, the programme concentrated on concept development and site selection activities. After 2003, when Posiva received the decision in principle from the Finnish Government, a new phase began in the programme. Since 2004, Posiva Oy has constructed an underground rock characterization facility on the repository site in Olkiluoto, in western Finland. This facility, called ONKALO, has provided an opportunity to carry out further site investigations, develop construction techniques, and test and demonstrate the engineered barrier system in an actual repository environment. As a result of these investigations and development efforts, the application for a licence to construct the encapsulation plant and the geological repository was submitted in 2012. The Radiation and Nuclear Safety Authority in Finland (STUK) first gave a positive review on the safety of the facility, and consequently the Finnish Government granted the construction licence in November 2015. After receiving the construction licence as the first disposal programme in the world, the next phase in the program will be the construction project of the final disposal facilities required for the disposal operations. A significant first-of-a-kind construction project like

Period changes of cataclysmic variables below the period gap: V2051 Oph, OY Car and Z Cha

Science.gov (United States)

Pilarčík, L.; Wolf, M.; Zasche, P.; Vraštil, J.

2018-04-01

We present our results of a long-term monitoring of cataclysmic variable stars (CVs). About 40 new eclipses were measured for the three southern SU UMa-type eclipsing CVs: V2051 Oph, OY Car and Z Cha. Based on the current O - C diagrams we confirmed earlier findings that V2051 Oph and OY Car present cyclic changes of their orbital periods lasting 25 and 29 years, respectively. In case of Z Cha we propose the light-time effect caused probably by a presence of the third component orbiting the eclipsing CV with the period of 43.5 years. The minimal mass of this companion results about 15 MJup.

Managing Customer Relationships in the Social Media : Case: Diamo oy

OpenAIRE

Laakso, Heidi

2013-01-01

The purpose of this study was to increase the amount of new likes of Diamo Oy in the social media network site Facebook, eventually aiming for the creation of new customer relationships, increased visibility of the company and into strictly increasing purchasing of their products. The theoretical part of the study consists of the role of content production in the social media field and customer relationship management, of which particularly customer acquisition and retention are discussed mor...

Loviisa starts low-level operating waste disposal in 1997

International Nuclear Information System (INIS)

Snellman, J.

1996-01-01

At an early stage Imatran Voima Oy (IVO) decided to construct a waste repository for Loviisa NPP. The suitability of the power plant site for final disposal of low- and intermediate- level operating waste was studied. In the site report in 1982 the plant site was found to be geologically suitable and economically feasible for construction. The necessary preparations started in 1992. The repository will be constructed in three phases. The first phase will cover the transport tunnel, construction of one maintenance waste tunnel and the excavation of another maintenance waste tunnel together with a hall for solidified wastes. This phase will be finished by the end of 1996. During the second phase in the beginning of next century the remaining already excavated rooms will be furnished. Finally in the third phase the repository will be extended for the decommissioning waste somewhere around years 2020-2025. (3 figs., 1 tab.)

Structure and Electrical Properties of NdBa2Cu3Oy Thin Films by Laser Ablation at Low Oxygen Partial Pressure

DEFF Research Database (Denmark)

Mozhaev, Peter B.; Mozhaeva, Julia; Khoryushin, Alexey

2017-01-01

in the film can be suppressed by an increase of the deposition temperature or by a decrease of the oxygen partial pressure during deposition. The presence of Nd/Ba disorder during deposition stimulates the introduction of oxygen into the growing film. A simple model is proposed for estimation of oxygen......A deposition process for NdBa2Cu3Oy thin films by laser ablation at decreased deposition temperature was developed using substitution of oxygen with argon in the chamber during deposition. A low deposition rate is the crucial factor to obtain high-quality NBCO films. The Nd/Ba cation disorder...... contents in the film using structural parameters measured with XRD techniques. Studies of the post-deposition annealing process showed ordering of the Nd/Ba sub-lattice and intense oxygen in- and out-diffusion. The temperature of the post-deposition annealing step should be chosen low enough (∼400 °C...

Environmentally friendly, high-performance generation

International Nuclear Information System (INIS)

Kalmari, A.

2003-01-01

The project developer, owner, and operator of the new 45 MWth BFB-based cogeneration plant in Iisalmi is Termia Oy, part of the Atro Group (formerly Savon Voima Oy). Fired on peat and wood waste and handed over to the customer in November 2002, the plant's electrical output is sold to the parent company and heat locally to customers in Iisalmi. When the construction decision was made, one of the main objectives was to utilise as high a level of indigenous fuels (peat and biomass) as possible, at a high level of efficiency. An environmental impact analysis was carried out, taking into account the impact of various fuels and emissions in terms of combustion and logistics. One main benefit of the type of plant ultimately selected was that the bulk of the fuel can be supplied from the surrounding area. This is very important in terms of fuel supply security and local employment. The government provided a EUR 2.7 million grant for the project, equivalent to 13% of the total EUR 21 million investment budget. Before the plant was built, Termia used approximately 95 GWh of indigenous fuels annually. Today, this figure is 220 GWh. The main fuel used is milled peat. Up to 30% green chips from logging residues can be used. Recycled waste fuel can cover up to 3% of the total fuel requirement

Valmistalokauppa : Toimintatapa ja omavalvonnan kehitystarpeet Muurametalot Oy:ssä

OpenAIRE

Kantola, Jussi

2012-01-01

Valmistalokauppa on kehittynyt paljon viimeisten vuosien aikana. Taloja valmistavia yrityksiä on perustettu runsaasti ja valmistalojen kysyntä on kasvanut. Työn tarkoituksena oli päivittää Muurametalot Oy:n omavalvontajärjestelmä vastaamaan paremmin tämän päivän tarpeita. Tavoitteena oli muokata omavalvonnassa käytettävät tarkastuspöytäkirjat nykypäivän asetusten tasolle. Ideaalitilanne olisi, että jokainen työntekijä osallistuu omavalvonnan käyttöön. Järjestelmän periaatteena on tarjota ...

Yritys X Oy:n myyntijohtolankojen hallintaprosessin kehittäminen

OpenAIRE

Jokinen, Mika

2014-01-01

Tässä kehittämistehtävässä tutkittiin miten Yritys X Oy:n myyntijohtolankojen hallintaprosessia voidaan kehittää niin, että se tuottaisi enemmän ja laadukkaampia myyntijohtolankoja myynnin käsiteltäväksi. Yritys X on kansainvälisiä logistiikkapalveluja tuottava yritys ja myyntijohtolankojen avulla voidaan hankkia uusia asiakkaita ja lisämyyntiä. Tutkimuksen lähtöaineisto kerättiin kyselytutkimuksella kuljettajilta ja ajojärjestelijöiltä. Kehittämistyön tueksi aineistoa kerättiin myös myy...

Todellista asiakasarvoa luomassa – myynnistä arvomyyntiin : case Mediamaisteri Oy

OpenAIRE

Brander, Heli

2017-01-01

Tämän opinnäytetyön tavoite oli case-yritys Mediamaisteri Oy:n B2B-myyntiprosessin kehittäminen. Tarkoitus oli hyödyntää palvelumuotoilua lisäämään asiakasymmärrystä ja tukemaan asiakkaiden kokeman arvon muodostumista sekä positiivisten asiakaskokemusten syntymistä myyntiprosessin alkuvaiheen asiakaskohtaamisissa. Lisäksi tarkoitus oli kiteyttää empiirisen aineiston pohjalta saatu tieto asiakasprofiileiksi, joita voidaan hyödyntää asiakaslähtöisten ratkaisujen suunnittelussa heti asiakkuuden ...

Toimintolaskennan soveltaminen : Case Hyvönen Yhtiöt Oy

OpenAIRE

Hyvönen, Krista

2014-01-01

Tässä opinnäytetyössä kehitetään toimeksiantajayrityksenä toimivalle Hyvönen Yhtiöt Oy:lle toimintolaskentaa hyödyntävä kustannuslaskentamalli. Tehdyn kehitystyön tavoitteena on luoda kohdeyritykselle täysin uudenlainen kustannuslaskentamalli, joka näyttää kustannukset uudesta näkökulmasta erilaisen ryhmittelyn ansiosta sekä selvittää toimintolaskennan soveltuvuutta yrityksen palvelukohtaisten kustannusten laskemiseen. Tutkimustehtävänä on selvittää, miten toimintolaskentaa voidaan soveltaa H...

Electronic structure and fine structural features of the air-grown UNxOy on nitrogen-rich uranium nitride

Science.gov (United States)

Long, Zhong; Zeng, Rongguang; Hu, Yin; Liu, Jing; Wang, Wenyuan; Zhao, Yawen; Luo, Zhipeng; Bai, Bin; Wang, Xiaofang; Liu, Kezhao

2018-06-01

Oxide formation on surface of nitrogen-rich uranium nitride film/particles was investigated using X-ray photoelectron spectroscopy (XPS), auger electron spectroscopy (AES), aberration-corrected transmission electron microscopy (TEM), and high-angle annular dark-field scanning transmission electron microscopy (HAADF-STEM) coupled with electron energy-loss spectroscopy (EELS). XPS and AES studies indicated that the oxidized layer on UN2-x film is ternary compound uranium oxynitride (UNxOy) in 5-10 nm thickness. TEM/HAADF-STEM and EELS studies revealed the UNxOy crystallizes in the FCC CaF2-type structure with the lattice parameter close to the CaF2-type UN2-x matrix. The work can provide further information to the oxidation mechanism of uranium nitride.

Part II: Oxidative Thermal Aging of Pd/Al2O3 and Pd/CexOy-ZrO2 in Automotive Three Way Catalysts: The Effects of Fuel Shutoff and Attempted Fuel Rich Regeneration

Directory of Open Access Journals (Sweden)

Qinghe Zheng

2015-10-01

Full Text Available The Pd component in the automotive three way catalyst (TWC experiences deactivation during fuel shutoff, a process employed by automobile companies for enhancing fuel economy when the vehicle is coasting downhill. The process exposes the TWC to a severe oxidative aging environment with the flow of hot (800 °C–1050 °C air. Simulated fuel shutoff aging at 1050 °C leads to Pd metal sintering, the main cause of irreversible deactivation of 3% Pd/Al2O3 and 3% Pd/CexOy-ZrO2 (CZO as model catalysts. The effect on the Rh component was presented in our companion paper Part I. Moderate support sintering and Pd-CexOy interactions were also experienced upon aging, but had a minimal effect on the catalyst activity losses. Cooling in air, following aging, was not able to reverse the metallic Pd sintering by re-dispersing to PdO. Unlike the aged Rh-TWCs (Part I, reduction via in situ steam reforming (SR of exhaust HCs was not effective in reversing the deactivation of aged Pd/Al2O3, but did show a slight recovery of the Pd activity when CZO was the carrier. The Pd+/Pd0 and Ce3+/Ce4+ couples in Pd/CZO are reported to promote the catalytic SR by improving the redox efficiency during the regeneration, while no such promoting effect was observed for Pd/Al2O3. A suggestion is made for improving the catalyst performance.

Toimenpide-ehdotuksia sosiaaliseen mediaan : case: Sony Music Entertainment Finland Oy

OpenAIRE

Marttila, Noora

2016-01-01

Tämä opinnäytetyö toteutettiin toimeksiantona Sony Music Entertainment Finland Oy levy-yhtiölle. Toimeksiantaja on kansainvälinen yritys joka on yksi Suomen suurimmista levy-yhtiöistä. Yritys käyttää sosiaalista mediaa osana digitaalista markkinointiaan. Sitä käytetään erityisesti artistien sekä julkaisujen tunnettuuden lisäämiseen. Sosiaalinen media on kasvava markkinoinnin keino, joka oikein hyödynnettynä voi johtaa hyviin lopputuloksiin. Tämän työn tarkoituksena on antaa yrityksell...

Synthesis of vanadium oxides 5 wt.%VO2–MxOy by sol–gel process ...

Indian Academy of Sciences (India)

Experimental results indicated that the VO2–SiO2 catalysts .... crucible, which was supported by the beam of a balance in the oven. ... Epoxidation of cyclohexene on VO2–Mx Oy (M = Si, Al, Ti). 1189 ... tially amorphous nature of silica.

Markkinointi murroksessa: Mistä on toimiva Facebook-markkinointi tehty? : Case: Infokone Oy

OpenAIRE

Korhonen, Taneli

2010-01-01

Opinnäytetyön tavoitteena oli kartoittaa Facebook-markkinoinnin toimivia käytänteitä B2B-yritykselle. Tutkimuksessa saatuja tuloksia on tarkoitus käyttää apuna toimeksiantaja Infokone Oy:n Facebook-strategian suunnittelussa. Käytetty tutkimusote oli kvalitatiivinen eli laadullinen tutkimus ja tutkimusmenetelmä teemahaastattelu. Tutkimukseen haastateltiin syys–lokakuussa 2010 kolmea B2B-yrityksen edustajaa, jotka olivat olleet mukana toteuttamassa edustamansa yrityksen Facebook-markkinoint...

Kanta-asiakkuuksien hankinta paikallisessa yö-kerhossa, Case: Giggling Marlin Oy

OpenAIRE

Ylärakkola, Emmi

2016-01-01

Tämän opinnäytetyön tarkoituksena oli saada käsitys siitä, saadaanko työssä toteutettujen keinojen avulla uusia kanta-asiakkaita ja aiempia kanta-asiakkaita uusimaan jäsenyytensä paikallisessa yökerhossa. Työn toimeksiantaja oli Giggling Marlin Oy, joka kuuluu Night People Group -ravintolayhtiöön. Giggling Marlin sijaitsee Lappeenrannan keskustassa ja on perustettu alkukeväällä 2006. Toiminnallisena osana suunniteltiin ja toteutettiin mainosmateriaalia kampanja-ajankohdalle sekä valmisteltiin...

A visible and infrared study of the eclipsing dwarf nova OY Carinae

International Nuclear Information System (INIS)

Berriman, G.

1984-01-01

This paper presents four visible light curves of the highly inclined, short-period cataclysmic binary star OY Carinae in quiescence. These light curves show that the red dwarf eclipses both its white dwarf companion and the accretion disc and hotspot, which originate from material transferred from the red dwarf to the white dwarf. The consequences of the findings are discussed in the light of current ideas about the evolution of cataclysmic variable stars. (author)

Decommissioning of a small reactor (BR3 reactor, Belgium)

International Nuclear Information System (INIS)

Dadoumont, J.; Massaut, V.; Klein, M.; Demeulemeester, Y.

2002-01-01

Since 1989, SCK-CEN has been dismantling its PWR reactor BR3 (Belgian Reactor No. 3). After gaining a great deal of experience in remote dismantling of highly radioactive components during the actual dismantling of the two sets of internals, the BR3 team completed the cutting of its reactor pressure vessel (RPV). During the feasibility phase of the RPV dismantling, a decision was made to cut it under water in the refuelling pool of the plant, after having removed it from its cavity. The RPV was cut into segments using a milling cutter and a bandsaw machine. These mechanical techniques have shown their ability for this kind of operations. Prior to the segmentation, the thermal insulation situated around the RPV was remotely removed and disposed of. The paper will describe all these operations. The BR3 decommissioning activities also include the dismantling of contaminated loops and equipment. After a careful sorting of the pieces, optimized management routes are selected in order to minimize the final amount of radioactive waste to be disposed of. Some development of different methods of decontamination were carried out: abrasive blasting (or sand blasting), chemical decontamination (Oxidizing-Reducing process using Cerium). The main goal of the decontamination program is to recycle most of the metallic materials either in the nuclear world or in the industrial world by reaching the respective recycling or clearance level. Overall the decommissioning of the BR3 reactor has shown the feasibility of performing such a project in a safe and economical way. Moreover, BR3 has developed methodologies and decontamination processes to economically reduce the amount of radwaste produced. (author)

Content Marketing Plan for Logistics Company : Case: NetLogistic JVM Oy

OpenAIRE

Riabochkina, Elizaveta

2017-01-01

This project aim is to create a Content Marketing Strategy for a Finnish logistics company. The aim of the thesis is to develop an action plan that will help the commissioning company increase sales and gain new customers. The commissioner for the study is NetLogistic JVM Oy. The Company specializes in handling trade of transit and export shipments to Russia and CIS-countries via Finland. The company does not have huge resources to run large campaign, but is looking for ways to reach new...

Eclipse studies of the dwarf nova Oy Carinae in quiescence

International Nuclear Information System (INIS)

Wood, J.H.; Horne, K.; Berriman, G.; Wade, R.A.

1989-01-01

High-speed photometry of OY Car have been obtained which cover 20 eclipses in white light and seven eclipses in UBR. The results show the red dwarf to have a mass of 0.070 + or - 0.002 solar masses and a radius of 0.127 + or - 0.002 solar radii, and the white dwarf to have a temperature of several thousand degrees below 15,000 K. The bright spot is found to have a compact 15,000-K core and a tail that extends along the rim but does not penetrate far into the disk. 31 refs

Establishing a pricing structure for software products : Case study: Viope Solutions Oy

OpenAIRE

Nguyen, Tram

2013-01-01

This thesis is a case study that explores how to establish a pricing structure for software products. The objective is to provide a guideline to establish a pricing structure for Viope Solutions Oy. A new pricing structure is crucial for the company due to recent changes in its business such as internationalisation and new product launches. The literature review introduces five attributes of a pricing structure. They are the unit definition, price determination, price segmentation, versio...

Electrochemically Obtained TiO2/CuxOy Nanotube Arrays Presenting a Photocatalytic Response in Processes of Pollutants Degradation and Bacteria Inactivation in Aqueous Phase

Directory of Open Access Journals (Sweden)

Magda Kozak

2018-06-01

Full Text Available TiO2/CuxOy nanotube (NT arrays were synthesized using the anodization method in the presence of ethylene glycol and different parameters applied. The presence, morphology, and chemical character of the obtained structures was characterized using a variety of methods—SEM (scanning electron microscopy, XPS (X-ray photoelectron spectroscopy, XRD (X-ray crystallography, PL (photoluminescence, and EDX (energy-dispersive X-ray spectroscopy. A p-n mixed oxide heterojunction of Ti-Cu was created with a proved response to the visible light range and the stable form that were in contact with Ti. TiO2/CuxOy NTs presented the appearance of both Cu2O (mainly and CuO components influencing the dimensions of the NTs (1.1–1.3 µm. Additionally, changes in voltage have been proven to affect the NTs’ length, which reached a value of 3.5 µm for Ti90Cu10_50V. Degradation of phenol in the aqueous phase was observed in 16% of Ti85Cu15_30V after 1 h of visible light irradiation (λ > 420 nm. Scavenger tests for phenol degradation process in presence of NT samples exposed the responsibility of superoxide radicals for degradation of organic compounds in Vis light region. Inactivation of bacteria strains Escherichia coli (E. coli, Bacillus subtilis (B. subtilis, and Clostridium sp. in presence of obtained TiO2/CuxOy NT photocatalysts, and Vis light has been studied showing a great improvement in inactivation efficiency with a response rate of 97% inactivation for E. coli and 98% for Clostridium sp. in 60 min. Evidently, TEM (transmission electron microscopy images confirmed the bacteria cells’ damage.

5S-MENETELMÄN KÄYTTÖÖNOTTO: LUVATA PORI OY

OpenAIRE

Rantanen, Suvi

2015-01-01

Opinnäytetyön aiheena oli 5S-menetelmän käyttöönotto Luvata Pori Oy:n tuotannos-sa. Tutkimuksen lähtökohtana ja ongelmana oli siisteyden ja järjestyksen puute. Ta-voitteena oli aikaansaada valittuun pilottikohteeseen toimivat, siistit ja turvalliset tuotantotilat sekä luoda järjestelmä, jonka avulla ylläpidetään saavutettu taso. Lisäksi tavoiteltiin hukan poistamista ja Luvatan Lean toimintatavan syventämistä. Projekti toteutettiin toimintatutkimuksena syksyn 2015 aikana. Tutkimuksen teor...

Endoscopic inspection of steam turbines

International Nuclear Information System (INIS)

Maliniemi, H.; Muukka, E.

1990-01-01

For over ten years, Imatran Voima Oy (IVO) has developed, complementary inspection methods for steam turbine condition monitoring, which can be applied both during operation and shutdown. One important method used periodically during outages is endoscopic inspection. The inspection is based on the method where the internal parts of the turbine is inspected through access borings with endoscope and where the magnified figures of the internal parts is seen on video screen. To improve inspection assurance, an image-processing based pattern recognition method for cracks has been developed for the endoscopic inspection of turbine blades. It is based on the deduction conditions derived from the crack shape. The computer gives an alarm of a crack detection and prints a simulated image of the crack, which is then checked manually

The development of Facebook marketing utilizing Content marketing. Case: Pirjon Pakari Oy.

OpenAIRE

Lehtikangas, Pauliina

2015-01-01

The purpose of this thesis was to learn how to improve Pirjon Pakari Oy’s brand visibility and overall visibility on Facebook. This was done by trying to add value to the end customer by producing interesting and engaging posts to Pirjon Pakari Oy’s Facebook page. This case study utilized theory from content marketing in attempts to improve the marketing of “Pirjon” Facebook page and therefore a Facebook content marketing strategy was created for Pirjon Pakari Oy. This strategy was implemente...

Integral test of JENDL-3.3 for fast reactors

International Nuclear Information System (INIS)

Chiba, Gou

2003-01-01

An integral test of JENDL-3.3 was performed for fast reactors. Various types of fast reactors were analyzed. Calculation values of the nuclear characteristics were greatly especially affected by the revisions of the cross sections of U-235 capture and elastic scattering reactions. The C/E values were improved for ZPPR cross where plutonium is mainly fueled, but not for BFS cores where uranium is mainly fueled. (author)

Anisotropy of the Seebeck Coefficient in the Cuprate Superconductor YBa_{2}Cu_{3}O_{y}: Fermi-Surface Reconstruction by Bidirectional Charge Order

Directory of Open Access Journals (Sweden)

O. Cyr-Choinière

2017-09-01

Full Text Available The Seebeck coefficient S of the cuprate YBa_{2}Cu_{3}O_{y} is measured in magnetic fields large enough to suppress superconductivity, at hole dopings p=0.11 and p=0.12, for heat currents along the a and b directions of the orthorhombic crystal structure. For both directions, S/T decreases and becomes negative at low temperature, a signature that the Fermi surface undergoes a reconstruction due to broken translational symmetry. Above a clear threshold field, a strong new feature appears in S_{b}, for conduction along the b axis only. We attribute this feature to the onset of 3D-coherent unidirectional charge-density-wave modulations seen by x-ray diffraction, also along the b axis only. Because these modulations have a sharp onset temperature well below the temperature where S/T starts to drop towards negative values, we infer that they are not the cause of Fermi-surface reconstruction. Instead, the reconstruction must be caused by the quasi-2D bidirectional modulations that develop at significantly higher temperature. The unidirectional order only confers an additional anisotropy to the already reconstructed Fermi surface, also manifest as an in-plane anisotropy of the resistivity.

Entering Chinese market for Finnish fashion jewellery company : case: Ninja Finland Oy

OpenAIRE

Wang, Yujue

2014-01-01

As one of the fastest-growing economies in the world, China is expected to become the largest fashion market in the world in a few years. As a result, more and more foreign companies have entered the Chinese market and the market has been more competitive than ever. This requires any fashion company to adopt a strategic development plan when entering the Chinese market. Ninja Finland Oy is a fashion jewellery company which offers a wide range of fashion jewellery and accessories. The comp...

Tuotteiden ohjeistaminen osana hankintatoimintaa ja tuotetiedonhallintaa : case: Seppälä Oy

OpenAIRE

Heikkonen, Veera

2010-01-01

Tuoteohjeistusten tekeminen on arkipäivää suomalaisissa vaatetusalan yrityksissä. Tuotteiden hankinta ulkomailta on yleistynyt huomattavasti, joka taas on tehnyt tuoteohjeistuksista tärkeitä työkaluja yrityksen toiminnassa. Tuoteohjeistusten avulla suunnittelijan ideat ja ajatukset toteutetaan oikeaksi tuotteeksi. Tämän opinnäytetyön tarkoitus oli kehittää Seppälä Oy:lle yhtenäinen tuoteohjeis-tuspohja, joka on selkeä ja toimiva. Projektin lähtökohtainen tarkoitus on, että Seppälä voi hy...

Electronic states of SiO2-MxOy (MxOy=P205, TiO2 and ZrO2) glasses

Energy Technology Data Exchange (ETDEWEB)

Kowada, Y [Hyogo Univ. of Teacher Education, Hyogo (Japan); Adachi, H [Kyoto Univ. (Japan). Faculty of Engineering; Minami, T [Univ. of Osaka Prefecture, Osaka (Japan). Faculty of Engineering

1993-12-01

Using the sol-gel method the surface of metal and glass substrates can be modified. For example, stainless steel sheets coated with the SiO2-ZrO2 glass films have higher resistance to corrosion and oxidation. The coating films contain high concentration of alkali ions diffusing from the glass substrates. It suggests that the sodium ions are trapped strongly within the coating films and are blocked to further diffuse to the surface. This behavior must be associated with the chemical bonding around the sodium ions in the SiO2-TiO2 and SiO2-ZrO2 films. For better understanding of the chemical bonding in the glasses, the electronic states of the SiO2-MxOy glasses were calculated by means of the DV-Xa cluster method. In this paper, the calculation method is explained, the results are discussed and the conclusion is stated. 17 refs., 6 figs.

Digitaalinen markkinointi pk-yrityksessä inbound-markkinoinnin keinoin CASE: Customer Intelligence Finland Oy

OpenAIRE

Lassila, Anna-Sofia

2016-01-01

Tämän opinnäytetyön tarkoituksena oli tutkia kuinka digitaalista markkinointia kannattaa tehdä inbound-markkinoinnin keinoin. Tutkimuksen ensisijaisena tavoitteena oli löytää tehokkaimmat keinot inbound-markkinoinnin toteuttamiseen. Toisena tavoitteena oli antaa tuloksiin pohjautuen suosituksia digitaalisen markkinoinnin jatkotoimenpiteistä inbound-metodia hyödyntäen. Tämän opinnäytetyön toimeksiantajana toimi Customer Intelligence Finland Oy (CIFI). Tämän opinnäytetyön tutkimusmenetelmän...

Corporate Social Responsibility: A Case Study of BSCI Implementation at Best Friend Group Oy

OpenAIRE

Qi, Yuan

2012-01-01

With the current polarized debate over CSR issues worldwide, ever increasing number of companies believe that incorporating ethical values into the business operation is an attractive option to upgrade the brand value, maintain the competitive advantage and cater for the social demand of various stakeholders. Business Social Compliance Initiative, a division of FTA, serves as a platform to assist companies in achieving the goal. The study was commissioned by Best Friend Group Oy, a medium...

Instrumentation renewal at the FIR 1 research reactor in Finland

International Nuclear Information System (INIS)

Bars, Bruno; Kall, Leif

1982-01-01

The Finnish TRIGA Mark II reactor (FIR 1 100 kW, later 250 kW steady state power and pulsing capability up to 250 MW) has been in operation for 20 years. The reactor is the only research reactor in Finland and is an important research training and service facility, which obviously will be operated for 10...20 years ahead. The mechanical parts of the reactor are in good shape. Some minor modifications have previously been made in the instrumentation. However, the original instrumentation could hardly have been used for 10...20 years ahead without extensive modifications and modernization. After a careful evaluation and planning process the whole reactor instrumentation was renewed in 1981 at a cost of about 400 000 dollar. The renewal was carried out in cooperation with the Central Research Institute for Physics (KFKI) at the Hungarian Academy of Sciences, which delivered the nuclear part of the instrumentation and with the Finnish company Valmet Oy Instrument Works, which delivered the conventional instrumentation, including the automatic power control system and the control console. The instrumentation, which is located in-a new isolated control room is based on modern industrial standard modular units with standardized signal ranges, electronic testing possibilities, galvanically isolated outputs etc. The instrument renewal project was brought successfully to completion in November 1981 after only about 10 working days of shut down time. The reactor is now in routine operation and the experiences gained from the new instrumentation are excellent. (author)

Dynamics of TRIGA-3 Salazar Reactor.; Dinamica del Reactor TRIGA Mark III del Centro Nuclear de Mexico.

Energy Technology Data Exchange (ETDEWEB)

Gallardo S, L F

1991-12-31

The theoretical study of temporal behavior of a nuclear reactor is of great importance, since it allows to know, in advance, the conditions to which a reactor is going to be submitted. The reliability of two computer codes (AIREK-JEN and PLANKIN) designed to reproduce the temporal behavior of nuclear reactors, generally power reactors, when they are applied to reproduce the dynamic behavior of TRIGA-3 Salazar Reactor is analyzed. In the first chapters, the fundamental equations that solve this computer codes are deduced, and also the main characteristics of TRIGA-3 Salazar Reactor and the necessary data to run the programs are presented; later the results obtained with the computer codes and the experimental results reported in the operational logbook of the reactor are compared, with the result that such computer codes are applicable to the temporal study of TRIGA-3 Salazar Reactor. (Author).

3D CAD model of the subcritical nuclear reactor of IPN; Modelo CAD 3D del reactor nuclear subcritico del IPN

Energy Technology Data Exchange (ETDEWEB)

Pahuamba V, F. de J.; Delfin L, A.; Gomez T, A. [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico); Ibarra R, G.; Del Valle G, E.; Sanchez R, A., E-mail: narehc@hotmail.com [IPN, Escuela Superior de Fisica y Matematicas, Av. IPN, Edif. 9, Unidad Profesional Adolfo Lopez Mateos, San Pedro Zacatenco, 07738 Ciudad de Mexico (Mexico)

2016-09-15

The three-dimensional (3D) CAD model of the subcritical reactor Chicago model 9000 of Instituto Politecnico Nacional (IPN) allows obtaining a 3D view with the dimensions of each of its components, such as: natural uranium cylindrical rods, fuel elements, hexagonal reactor core arrangement, cylindrical stainless steel tank containing the core, fuel element support grids and reactor water cleaning system. As a starting point for the development of the model, the Chicago model 9000 subcritical reactor manual provided by the manufacturer was used, the measurement and verification of the components to adapt the geometric, physical and mechanical characteristics was carried out and materials standards were used to obtain a design that allows to elaborate a new manual according to the specifications. In addition, the 3D models of the building of the Advanced Physics Laboratory, neutron generator, cobalt source and the corridors connecting to the subcritical reactor facility were developed, allowing an animated ride, developed by computer-aided design software. The manual provided by the company Nuclear Chicago, dates from the year 1959 and presents diverse deviations in the design and dimensions of the reactor components. The model developed; in addition to supporting the development of the new manual represents a learning tool to visualize the reactor components. (Author)

Direct harvesting of Helium-3 (3He) from heavy water nuclear reactors

International Nuclear Information System (INIS)

Bentoumi, G.; Didsbury, R.; Jonkmans, G.; Rodrigo, L.; Sur, B.

2013-01-01

The thermal neutron activation of deuterium inside a heavy-water-moderated or -cooled nuclear reactor produces a build-up of tritium in the heavy water. The in situ decay of tritium can, for certain reactor types and operating conditions, produce potentially useable amounts of 3 He, which can be directly extracted via the heavy-water cover gas without first separating, collecting and storing tritium outside the reactor. It is estimated that the amount of 3 He available for recovery from the moderator cover gas of a 700 MWe class Pressurized Heavy Water Reactor (PHWR) ranges from 0.1 to 0.7 m 3 (STP) per annum, varying with the tritium activity buildup in the moderator. The harvesting of 3 He would generate approximately 12.7 m 3 (STP) of 3 He, worth more than $30M at current market rates, over a typical 25-year operating cycle of the PHWR. This paper discusses the production of 3 He in the moderator of a PHWR and its extraction from the 4 He moderator cover gas system using conventional methods. (author)

A/B Testing in Improving Conversion on a Website : Case: Sanoma Entertainment Oy

OpenAIRE

Arento, Thomas

2010-01-01

The purpose of this thesis is to study marketing possibilities of improved conversion rates on websites. The study was made for Sanoma Entertainment Oy’s Gaming & Online unit. The main objective was to explore A/B testing as a tool to improve conversion rates by increasing click-through rates. The secondary objective was to test Google Website Optimizer as an A/B testing tool in comparison to current methods of A/B testing in Sanoma Entertainment Oy. The results of this study will be used as ...

D-3He fuel cycles for neutron lean reactors

International Nuclear Information System (INIS)

Kernbichler, W.; Miley, G.H.; Heindler, M.

1989-01-01

The intrinsic potential of D-3He as a reactor fuel is investigated for a large range of 3He to D density ratios. A steady-state zero-dimensional reactor model is developed in which much care is attributed to a proper treatment of fast fusion products. Useful ranges of reactor parameters as well as temperature-density windows for driven and ignited operation are identified. Various figures of merit are calculated, such as power densities, net power production, neutron production, tritium load and radiative power. These results suggest several optimistic conclusions about the performance of D-3He as a reactor fuel

5S-menetelmän käyttöönotto Case: Planmeca Oy

OpenAIRE

Velic, Adel

2014-01-01

Opinnäytetyön tavoitteena on tutkia Planmeca Oy:n panoraamaröntgentuotannossa syksyllä 2012 toteutetun 5S-projektin onnistuneisuutta. Työ kattaa sekä 5S-periaatteiden mukaisesti suunniteltujen ja toteutettujen panoraamaröntgentuotannon Elko-, Pystytys 1 -, Pystytys 2 - sekä Potilastuki- ja skannausmekanismitiimin työpisteiden että projektin toteutustavan onnistuneisuuden mittaamisen. Tutkimusmenetelmänä käytetään Elko-, Pystytys 1 -, Pystytys 2 - sekä Potilastuki- ja skannausmekanismitiim...

Neste Oy starts the production of extra high viscosity index lubricating oil in Porvoo

International Nuclear Information System (INIS)

Kilander, H.

1997-01-01

Neste Oy is starting the manufacture of basic oil, used in advanced motor lubricants, in Finland. The plant will start the manufacture of the EHVI (Extra High Viscosity) by the end of 1997. The EHVI basic oil is a synthetic-like oil product, suitable for manufacture of high-quality lubricants. In the beginning the production of the basic oil will be about 50 000 tons/a. The investment costs of the plants are 180 million FIM

Markkinoinnin kehittäminen, Verkkomarkkinointi ja sosiaalinen media : JA-KI Muutto Oy

OpenAIRE

Kahanpää, Timo

2015-01-01

Tämän opinnäytetyön tavoitteena oli kehittää ja tehostaa helsinkiläisen muuttoalan yrityksen JA-KI Muutto Oy:n markkinointia verkossa ja siellä nimenomaan sosiaalisessa mediassa. Työn tavoitteena oli löytää ja tuoda esille keinoja ja kehittämisehdotuksia, joiden avulla markkinointia verkossa voitaisiin tehostaa entisestään. Opinnäytetyö tehtiin kehittämistehtävänä ja se jakautuu kahteen osaan: teoriaosaan ja kehittämisosaan. Teoriaosiossa käydään läpi eri lähteiden kautta ensin markkinoin...

Spin-Glass Transition and Giant Paramagnetism in Heavily Hole-Doped Bi2Sr2Co2Oy

Science.gov (United States)

Hsu, Hung Chang; Lee, Wei-Li; Lin, Jiunn-Yuan; Young, Ben-Li; Kung, Hsiang-Hsi; Huang, Jian; Chou, Fang Cheng

2014-02-01

Hole-doped single crystals of misfit-layered cobaltate Bi2-xPbxSr2-zCo2Oy (x = 0-0.61, y = 8.28-8.62, and z = 0.01-0.22) have been successfully grown using the optical floating-zone method. Heavier hole doping has been achieved through both Pb substitution in the Bi site and the more effective Sr vacancy formation. The Co4+ : Co3+ ratio can be raised significantly from its original ˜1 : 1 to 4.5 : 1, as confirmed by iodometric titration. A spin-glass transition temperature of Tg ˜ 70 K is confirmed by ac susceptibility measurement when the Co4+ : Co3+ ratio becomes higher than 2 : 1, presumably owing to the significantly increased probability of triangular geometrical frustration among antiferromagnetically coupled localized Co4+ spins.

Conceptual design of D-3He FRC reactor 'ARTEMIS'

International Nuclear Information System (INIS)

Momota, H.; Ishida, A.; Kohzaki, Y.

1991-07-01

A comprehensive design study of the D- 3 He fueled field-reversed configuration (FRC) reactor 'ARTEMIS' is carried out for the purpose of proving its attractive characteristics and clarifying the critical issues for a commercial fusion reactor. The FRC burning plasma is stabilized and sustained in a steady equilibrium by means of a preferential trapping of D- 3 He fusion-produced energetic protons. A novel direct energy converter for 15MeV protons is also presented. On the bases of a consistent scenario of the fusion plasma production and simple engineering, a compact and simple reactor concept is presented. The design of the D- 3 He FRC power plant definitely offers the most attractive prospect for energy development. It is environmentally acceptable in view of radio-activity and fuel resources; and the estimated cost of electricity is low compared to a light water reactor. Critical issues concerning physics or engineering for the development of the D- 3 He FRC reactor are clarified. (author)

Ostotoiminnan laadunvarmistus ja toiminnanohjauksen kehittäminen Mantsinen Group Ltd Oy:ssä

OpenAIRE

Hiltunen, Jarmo

2012-01-01

Opinnäytetyön aiheena oli ostotoiminnan laadunvarmistuksen ja toiminnanohjauksen kehittäminen Mantsinen Group Ltd Oy:ssä. Työssä on perehdytty erityisesti niihin ostotoiminnan laadun kehittämisen haasteisiin mitkä liittyvät alihankintaostamisen osa-alueeseen. Opinnäytetyön päätavoitteena oli luoda työkaluja hankinnan laadun jatkuvan parantamisen avuksi. Lisäksi tavoitteena oli perehtyä hankintojen mittaamisen teoriaan. Opinnäytetyön tutkimusmenetelmänä on käytetty käytännön tietoa ostam...

Ympäristöjärjestelmän rakentaminen : Case: Inno Interior Oy

OpenAIRE

Pirinen, Liisa

2009-01-01

Tavoitteena oli rakentaa huonekalualalla toimivalle Inno Interior Oy:lle, joka valmistuttaa tuotteet alihankkijoilla, sertifiointikelpoinen ympäristönhallintajärjestelmä. Yrityksellä ei ollut ympäristösertifikaattia. Asiakkaat vaativat yritykseltä ympäristöasioiden hoitamisen tason osoittamista. Ympäristöhallintajärjestelmän rakentaminen ja sertifiointi oli sopivin ratkaisu ongelmaan. Ympäristöjärjestelmän rakentamisen lähtökohtana olivat toiminnan aiheuttamat ympäristö-vaikutukset ja...

Removal in a lump of JRR-3 nuclear reactor

International Nuclear Information System (INIS)

Ohnishi, Nobuaki; Suzuki, Masanori; Nagase, Tetsuo; Watanabe, Morinari.

1989-01-01

The research reactor JRR-3 in Japan Atomic Energy Research Institute is called 'Home made No.1 reactor' as all except fuel and heavy water as the moderator and coolant were manufactured in Japan. The JRR-3 attained the criticality in 1962, and the cumulative time of operation reached 47135.5 hours, and the cumulative power output reached 419073.5 MWh. It was stopped in 1983. During the period, it was utilized for beam experiment, irradiation of fuel and materials, RI production and others. In order to cope with the expansion of utilization and the advance of utilizing technology of the research reactor, the reconstruction works are in progress, and the criticality of the reconstructed reactor is expected in 1990. On the site where the old reactor is removed, the reactor of different type is installed, and the first large cold neutron source is equipped. In this report, as to the removal of the old reactor proper, the method of working and the results are described. Considering the period of working, the cost and the management of the removed reactor, in the case of the JRR-3, the method of carrying it out in a lump was adopted as the optimum removal method. The plan, procedure and results of the removal working are reported. (K.I.)

3D simulation of CANDU reactor regulating system

International Nuclear Information System (INIS)

Venescu, B.; Zevedei, D.; Jurian, M.

2013-01-01

Present paper shows the evaluation of the performance of the 3-D modal synthesis based reactor kinetic model in a closed-loop environment in a MATLAB/SIMULINK based Reactor Regulating System (RRS) simulation platform. A notable advantage of the 3-D model is the level of details that it can reveal as compared to the coupled point kinetic model. Using the developed RRS simulation platform, the reactor internal behaviours can be revealed during load-following tests. The test results are also benchmarked against measurements from an existing (CANDU) power plant. It can be concluded that the 3-D reactor model produces more realistic view of the core neutron flux distribution, which is closer to the real plant measurements than that from a coupled point kinetic model. It is also shown that, through a vectorization process, the computational load of the 3-D model is comparable with that of the 14-zone coupled point kinetic model. Furthermore, the developed Graphical User Interface (GUI) software package for RRS implementation represents a user friendly and independent application environment for education training and industrial utilizations. (authors)

Analyses for MARIA Research Reactor with RELAP/MOD3 code

International Nuclear Information System (INIS)

Szczurek, J.; Czerski, P.

2004-01-01

This paper deals with the application of the RELAP5/MOD3 code to the transient analyses for MARIA research reactor. Poland's MARIA Research Reactor is water and beryllium moderated, water-cooled reactor of a pool type with pressurized fuel channels containing concentric multi-tube assemblies of highly enriched uranium clad in aluminium. The RELAP5/MOD3 input data model includes the whole primary cooling circuit of the MARIA reactor. The model was qualified against the reactor data at steady state conditions and additionally against the existing reliable experimental data for a transient initiated by the reactor scram. The RELAP transient simulation was performed for loss of forced flow accidents including two scenarios with protected and unprotected (no scram) reactor core. Calculations allow estimating time margin for reactor scram initiation and reactivity feedbacks contribution to the results. (author)

Biomass CFB gasifier connected to a 350 MW{sub t}h steam boiler fired with coal and natural gas - THERMIE demonstration project in Lahti in Finland

Energy Technology Data Exchange (ETDEWEB)

Palonen, J. [Foster Wheeler Energia Oy, Varkaus (Finland). Varkaus Global New Products

1997-12-31

The successful experience in developing the advanced Foster Wheeler Energi Oy`s (former Ahlstroem Pyropower) Circulating Fluidized Bed combustion system subsequently led to the development of the CFB gasification technology in the early 1980s. The driving force for the developing work was the dramatic increase in oil price during the oil crises. The primary advantage of CFB gasification technology is that the it enables the substitution of expensive fuels e.g. oil or gas with cheap solid fuels. These cheap fuels are typically different types of waste woods, bark or other biofuels. In the CFB gasifier these solid fuels are converted to gaseous fuel which can be used instead of other expensive fuels. In some cases this also solves a waste disposal problem, providing a secondary economic and environmental benefit. Foster Wheeler Energia Oy has supplied four commercial scale atmospheric CFB gasifiers in the mid 80s to the pulp and paper industry with capacities from 17 to 35 MW based on fuel input. These applications utilize waste wood as feedstock and the units are still successfully operation today. Lahden Laempoevoima Oy is a Finnish power company producing power and district heat for the city of Lahti. The company is 50 % owned by the city of Lahti and 50 % by Imatran Voima Oy, which is the largest utility power company in Finland. Lahden Laempoevoima Oy operates the Kymijaervi power plant locating nearby the city of Lahti in Southern Finland. To keep the energy prices as low as possible, Lahden Laempoevoima is continuously looking for the most economical fuel sources, and simultaneously, trying to improve the environmental acceptability of the energy production. At the moment, about 300 GWh/a different type of biofuels and refuse fuels are available in the Lahti area. On an annual basis, the available amount of biofuels and refuse fuels is enough to substitute for about 15 % of the fuels burned in the main boiler equalling max 30 % of coal. The aim in this Lahden

Biomass CFB gasifier connected to a 350 MW{sub t}h steam boiler fired with coal and natural gas - THERMIE demonstration project in Lahti in Finland

Energy Technology Data Exchange (ETDEWEB)

Palonen, J [Foster Wheeler Energia Oy, Varkaus (Finland). Varkaus Global New Products

1998-12-31

The successful experience in developing the advanced Foster Wheeler Energi Oy`s (former Ahlstroem Pyropower) Circulating Fluidized Bed combustion system subsequently led to the development of the CFB gasification technology in the early 1980s. The driving force for the developing work was the dramatic increase in oil price during the oil crises. The primary advantage of CFB gasification technology is that the it enables the substitution of expensive fuels e.g. oil or gas with cheap solid fuels. These cheap fuels are typically different types of waste woods, bark or other biofuels. In the CFB gasifier these solid fuels are converted to gaseous fuel which can be used instead of other expensive fuels. In some cases this also solves a waste disposal problem, providing a secondary economic and environmental benefit. Foster Wheeler Energia Oy has supplied four commercial scale atmospheric CFB gasifiers in the mid 80s to the pulp and paper industry with capacities from 17 to 35 MW based on fuel input. These applications utilize waste wood as feedstock and the units are still successfully operation today. Lahden Laempoevoima Oy is a Finnish power company producing power and district heat for the city of Lahti. The company is 50 % owned by the city of Lahti and 50 % by Imatran Voima Oy, which is the largest utility power company in Finland. Lahden Laempoevoima Oy operates the Kymijaervi power plant locating nearby the city of Lahti in Southern Finland. To keep the energy prices as low as possible, Lahden Laempoevoima is continuously looking for the most economical fuel sources, and simultaneously, trying to improve the environmental acceptability of the energy production. At the moment, about 300 GWh/a different type of biofuels and refuse fuels are available in the Lahti area. On an annual basis, the available amount of biofuels and refuse fuels is enough to substitute for about 15 % of the fuels burned in the main boiler equalling max 30 % of coal. The aim in this Lahden

Imaging the risks - risking the image: Social impact assessment of the final disposal facility

International Nuclear Information System (INIS)

Avolahti, J.; Vira, J.

1999-01-01

Preparations for the final disposal of spent nuclear fuel in Finland started about twenty years ago. At present the work is carried out by Posiva Oy, which in 1996 took over the programme managed earlier by Teollisuuden Voima Oy, one of the country's nuclear power companies. From 1996 on the preparations have been made for all the spent fuel from Finnish nuclear power stations. The site for the final disposal facility will be selected among four alternatives by the end of 2000 and - assuming that the technical approach proposed by Posiva is accepted by the Government and the Parliament - the construction of the repository will start in the 2010s. The disposal operations are planned to be started in 2020. The alternative four sites have gone through a systematic site selection process based on geologic siting criteria and on environmental and cultural considerations. One of the objectives of the process was to avoid inhabited areas, agricultural fields, valuable groundwater or preservation areas as well as areas which might draw interest as regards the potential for ore deposits. The idea was that the field investigations and later the possible disposal facility should not cause any harm to local people. Two of the candidate sites are at present nuclear power plant sites situated at the coast, the two other candidates are inland sites with no nuclear activities. The geologic siting investigations were started in 1987. Interim assessments of the results so far have been made in 1992 and 1996 and a final report of all the investigations will be published before the end of 2000. The present view is that all four candidates are geologically suitable for siting the repository. Posiva's EIA for the final disposal of spent fuel in Finland is nearing completion. A considerable effort was made to involve local groups and individuals in the assessment process. Yet the participation remained limited and consisted mainly of active opponents of the project and of those who were

Imaging the risks - risking the image: Social impact assessment of the final disposal facility

Energy Technology Data Exchange (ETDEWEB)

Avolahti, J.; Vira, J. [Posiva Oy, Helsinki (Finland)

1999-12-01

Preparations for the final disposal of spent nuclear fuel in Finland started about twenty years ago. At present the work is carried out by Posiva Oy, which in 1996 took over the programme managed earlier by Teollisuuden Voima Oy, one of the country's nuclear power companies. From 1996 on the preparations have been made for all the spent fuel from Finnish nuclear power stations. The site for the final disposal facility will be selected among four alternatives by the end of 2000 and - assuming that the technical approach proposed by Posiva is accepted by the Government and the Parliament - the construction of the repository will start in the 2010s. The disposal operations are planned to be started in 2020. The alternative four sites have gone through a systematic site selection process based on geologic siting criteria and on environmental and cultural considerations. One of the objectives of the process was to avoid inhabited areas, agricultural fields, valuable groundwater or preservation areas as well as areas which might draw interest as regards the potential for ore deposits. The idea was that the field investigations and later the possible disposal facility should not cause any harm to local people. Two of the candidate sites are at present nuclear power plant sites situated at the coast, the two other candidates are inland sites with no nuclear activities. The geologic siting investigations were started in 1987. Interim assessments of the results so far have been made in 1992 and 1996 and a final report of all the investigations will be published before the end of 2000. The present view is that all four candidates are geologically suitable for siting the repository. Posiva's EIA for the final disposal of spent fuel in Finland is nearing completion. A considerable effort was made to involve local groups and individuals in the assessment process. Yet the participation remained limited and consisted mainly of active opponents of the project and of those

Completion of reconstruction for Japan Research Reactor No.3

International Nuclear Information System (INIS)

Kakefuda, K.; Tani, M.; Isshiki, M.

1992-01-01

The works of the reconstruction for the Japan Research Reactor No.3 (JRR-3) started in 1985 and initial criticality of the new reactor achieved in March, 1990. After commissioning test, the new JRR-3 has been operated some operational cycles since November, 1990. This paper presents outline of the removal work on the old JRR-3 and the new JRR-3. (author)

Development of a new WWER-440 fuel design

International Nuclear Information System (INIS)

Coucil, D.; Totev, T.

1998-01-01

In March 1996 British Nuclear Fuel Limited signed a contract with Imatran Voima and Paks Nuclear Power Plant to design, develop, license and supply 5 Lead Test Assemblies to the WWER-440 reactor at Loviisa in Finland. In June 1998 the manufacture of these 5 assemblies (4 fixed assemblies and 1 follower assembly) was completed. The fuel is expected to be loaded into Loviisa Unit 2 reactor during the shutdown scheduled for September of this year. (Authors)

Internationalization Strategies for Multinational Companies (MNCs) : In the case of KWH Mirka Oy to Ethiopia

OpenAIRE

Yigzaw, Tamirat

2015-01-01

This thesis is done for the fulfilment of Masters of Business Administration (MBA) degree program in Business Management and Entrepreneurship at Häme University of Applied Sciences (HAMK). The main focus of the study is internationalization of companies to a new market as a development task for the company. The commissioner of the thesis is Mirka Oy, a Finnish multinational company. The purpose of the thesis was to study factors that should be considered in internationalization strategies...

Feed water distribution pipe replacement at Loviisa NPP

Energy Technology Data Exchange (ETDEWEB)

Savolainen, S.; Elsing, B. [Imatran Voima Loviisa NPP (Finland)

1995-12-31

Imatran Voima Oy operates two WWER-440 reactors. Unit 1 has been operating since 1977 and unit 2 since 1981. First damages of feed water distribution (FWD) pipe were observed in 1989. The FWD-pipe T-connection had suffered from severe erosion corrosion damages. Similar damages have been been found also in other WWER-440 type NPPs. In 1989 the nozzles of the steam generator YB11 were inspected. No signs of the damages or signs of erosion were detected. The first damaged nozzles were found in 1992 in steam generators of both units. In 1992 it was started studying different possibilities to either repair or replace the damaged FWD-pipes. Due to the difficult conditions for repairing the damaged nozzles it was decided to study different FWD-pipe constructions. In 1991 two new feedwater distributors had been implemented at Dukovany NPP designed by Vitckovice company. Additionally OKB Gidropress had presented their design for new collector. In spring 1994 all the six steam generators of Rovno NPP unit 1 were replaced with FWD-pipes designed by OKB Gidropress. After the implementation an experimental program with the new systems was carried out. Due to the successful experiments at Rovno NPP Unit 1 it was decided to implement `Gidropress solution` during 1994 refueling outage into the steam generator YB52 at Loviisa 2. The object of this paper is to discuss the new FWD-pipe and its effects on the plant safety during normal and accident conditions. (orig.).

Feed water distribution pipe replacement at Loviisa NPP

Energy Technology Data Exchange (ETDEWEB)

Savolainen, S; Elsing, B [Imatran Voima Loviisa NPP (Finland)

1996-12-31

Imatran Voima Oy operates two WWER-440 reactors. Unit 1 has been operating since 1977 and unit 2 since 1981. First damages of feed water distribution (FWD) pipe were observed in 1989. The FWD-pipe T-connection had suffered from severe erosion corrosion damages. Similar damages have been been found also in other WWER-440 type NPPs. In 1989 the nozzles of the steam generator YB11 were inspected. No signs of the damages or signs of erosion were detected. The first damaged nozzles were found in 1992 in steam generators of both units. In 1992 it was started studying different possibilities to either repair or replace the damaged FWD-pipes. Due to the difficult conditions for repairing the damaged nozzles it was decided to study different FWD-pipe constructions. In 1991 two new feedwater distributors had been implemented at Dukovany NPP designed by Vitckovice company. Additionally OKB Gidropress had presented their design for new collector. In spring 1994 all the six steam generators of Rovno NPP unit 1 were replaced with FWD-pipes designed by OKB Gidropress. After the implementation an experimental program with the new systems was carried out. Due to the successful experiments at Rovno NPP Unit 1 it was decided to implement `Gidropress solution` during 1994 refueling outage into the steam generator YB52 at Loviisa 2. The object of this paper is to discuss the new FWD-pipe and its effects on the plant safety during normal and accident conditions. (orig.).

Inter- and intragranular properties of bismuth calcium strontium copper oxide (Bi2CaSr2Cu2Oy) superconductors

NARCIS (Netherlands)

Emmen, J.H.P.M.; Brabers, V.A.M.; Steen, van der C.; Dalderop, J.H.J.; Lenczowski, S.K.J.; Jonge, de W.J.M.

1989-01-01

The granular behaviour of sintered bulk Bi2CaSr2Cu2Oy superconductor is investigated by resistivity and ¿ac measurements. The observed temperature and magnetic field dependence is discussed within the framework of a granular model. The frequency dependence of the intragranular losses leads to a flux

EL3 reactor description and safety analysis report

International Nuclear Information System (INIS)

1969-02-01

The EL-3 reactor is an experimental pile. Heterogenous type reactor, water moderated and cooled it uses slightly enriched uranium oxide as fuel (4.5 percent) distributed in vertical cells that constitute the core (the maximum number of cells is 99). It is conceived to function at a maximal thermal power of 20 MW. It supplies a maximum thermal neutron flux of 10 14 neutrons/cm 2 /sec. It has several experimental devices. The EL-3 reactor is surrounded by auxiliary circuits of fluids, in a sealed containment, slightly depressed. The primary heavy water coolant circuit is completely included in this containment. Its cooling is made by the intermediary of a light water secondary circuit by atmospheric refrigerants. The ventilation circuits of the sealed containment and the reactor block do not release air outside, under nornal functioning, by a particularly studied chimney only after filtering and eventually dilution. The eventual contamination of the light water or air by active products is permanently monitored to allow the reactor shutdown and avoid the release in atmosphere of dangerous products. The EL-3 reactor, laying down in may 1955, has diverged in july 1957, made its first ascending in power in december 1957 and reached its complete power in april 1958. The positioning of actual fuel (snow crystal) was made during summer 1964. Reactor with an experimental aim, it is used for theoretical and technological studies by material irradiation in the experimental channels and the core cells, with possibilities to constitute independent loops (relative to the cooling fluids). Thirty vertical channels are devoted to the fabrication of artificial radioelements [fr

Instrumentation of the model in scaled 1:10 to prototype of the AquaBuOY wave energy converter

DEFF Research Database (Denmark)

Margheritini, Lucia; Frigaard, Peter

The objective of this report is to provide guidelines for the instrumentation of a model in scale 1:10 to prototype of the AquaBuOY wave energy converter. The model will be located in Nissum Bredning area: this is an important waterway already used by Aalborg University for real sea tests of wave...... energy converters....

Course of operators of the RA-3 reactor

International Nuclear Information System (INIS)

Caligiuri, G.A.

1983-01-01

Description of the fundamental principles of the nuclear reactors' control systems. The RA-3 reactor's control and measurement systems are principally described, without setting aside the basic criteria for the design of an appropriate instrumentation for the control of a nuclear reactor, as well as the theory on which the functioning of the several detectors and equipments used in a nuclear instrumentation are based. The main purpose of this course is that of serving, preferentially as a text, for the training of personnel which shall perform operation tasks in this reactor. The work includes three well-defined sections. The first two ones make an introduction to the subject, while the third one, extending to more than half-work, deals with the general description of the system in which the control and operation logic of RA-3 are included. (R.J.S) [es

New version of the reactor dynamics code DYN3D for Sodium cooled Fast Reactor analyses

Energy Technology Data Exchange (ETDEWEB)

Nikitin, Evgeny [Ecole Polytechnique Federale de Lausanne (Switzerland); Helmholtz-Zentrum Dresden-Rossendorf (HZDR) e.V., Dresden (Germany); Fridman, Emil; Bilodid, Yuri; Kliem, Soeren [Helmholtz-Zentrum Dresden-Rossendorf (HZDR) e.V., Dresden (Germany)

2017-07-15

The reactor dynamics code DYN3D being developed at the Helmholtz-Zentrum Dresden-Rossendorf is currently under extension for Sodium cooled Fast Reactor analyses. This paper provides an overview on the new version of DYN3D to be used for SFR core calculations. The current article shortly describes the newly implemented thermal mechanical models, which can account for thermal expansion effects of the reactor core. Furthermore, the methodology used in Sodium cooled Fast Reactor analyses to generate homogenized few-group cross sections is summarized. The conducted and planned verification and validation studies are briefly presented. Related publications containing more detailed descriptions are outlined for the completeness of this overview.

Measurement of reinforcement corrosion in concrete structures. Betonirakenteiden raudoituksen korroosion tutkiminen

Energy Technology Data Exchange (ETDEWEB)

Meuronen, A

1992-03-01

Ageing and aggressive enviromental conditions of concrete structures will result in deterioration of concrete and corrosion of steel in concrete. Corrosion of steel will in time result in the end of the service life or expensive renovations, unless corrosion of steel is noticed and renovated in time. Corrosion of steel in concrete can be found out by the present corrosion measurement methods, so that renovation can be started in right time. The report presents mainly on the basis of the literature references the following corrosion measurement methods: polarisation resistance, AC-impedance, electrical resistance probe, electrochemical noice and half-cell potential mapping. The half-cell potential mapping will be presented more precisely than the other corrosion measurement methods, for the potential mapping is the most used method. Concrete and Soils Laboratory of Imatran Voima Oy uses in the measurement of reinforcement corrosion the English, eight channel potential measuring equipment.

Measurement of reinforcement corrosion in concrete structures; Betonirakenteiden raudoituksen korroosion tutkiminen

Energy Technology Data Exchange (ETDEWEB)

Meuronen, A

1992-03-01

Ageing and aggressive enviromental conditions of concrete structures will result in deterioration of concrete and corrosion of steel in concrete. Corrosion of steel will in time result in the end of the service life or expensive renovations, unless corrosion of steel is noticed and renovated in time. Corrosion of steel in concrete can be found out by the present corrosion measurement methods, so that renovation can be started in right time. The report presents mainly on the basis of the literature references the following corrosion measurement methods: polarisation resistance, AC-impedance, electrical resistance probe, electrochemical noice and half-cell potential mapping. The half-cell potential mapping will be presented more precisely than the other corrosion measurement methods, for the potential mapping is the most used method. Concrete and Soils Laboratory of Imatran Voima Oy uses in the measurement of reinforcement corrosion the English, eight channel potential measuring equipment.

Applied strain dependence of critical current and internal lattice strain for BaHfO_3-doped GdBa_2Cu_3O_y coated conductors

International Nuclear Information System (INIS)

Usami, Takashi; Yoshida, Yutaka; Ichino, Yusuke; Sugano, Michinaka; Machiya, Shutaro; Ibi, Akira; Izumi, Teruo

2016-01-01

The strain effect of REBa_2Cu_3O_y (REBCO: RE = Y, Gd, Sm)-coated conductors (CCs) on critical current (I_c) is one of the most fundamental factors for superconducting coil applications. In this study, we aim to clarify the effect of artificial pinning center shapes on the strain effect in BHO-doped GdBCO CCs. To achieve this, we fabricated a Pure-GdBCO CC, a BHO nanorod-doped GdBCO CC and a multilayered-GdBCO (ML-GdBCO) CC, and carried out bending tests. As the result, the strain dependence of I_c for each CC showed an upward convex and the peak strain of the BHO-doped GdBCO CC shifts towards the compressive strain independent of the BHO shapes. In addition, the strain sensitivity of I_c in the GdBCO CCs including BHO becomes smaller. To clarify the difference between the strain sensitivity of I_c and the peak strain among the CCs, we evaluated the residual strain and the slopes of the internal lattice strains against the applied tensile strain (β). From this measurement, the residual strains for the Pure-GdBCO CC and the ML-GdBCO CC were almost the same. In addition, there was no change in the β value between the Pure-GdBCO and ML-GdBCO CCs. These results suggest that the changes in peak strain and strain sensitivity were not related to the internal lattice strain. (author)

Ecosystem characterization strategy at a repository site - Olkiluoto as a case study

Energy Technology Data Exchange (ETDEWEB)

Pere, Tuomas [Posiva Oy, Olkiluoto, 27160 Eurajoki (Finland); Kangasniemi, Ville [Environmental Research and Assessment EnviroCase, Ltd., Hallituskatu 1 D 4, 28100 Pori (Finland); Lahdenperae, Anne-Maj [Saanio and Riekkola Oy, Laulukuja 4, 00420 Helsinki (Finland); Aro, Lasse [Finnish Forest Research Institute, Kaironiementie 15, 39700 Parkano (Finland)

2014-07-01

Posiva Oy is constructing an underground research facility ONKALO in Olkiluoto, located in the municipality of Eurajoki, Finland. This is part of the plan for final disposal of spent nuclear fuel produced by Posiva's owners: Teollisuuden Voima Oyj and Fortum Power and Heat Oyj. Posiva has applied for construction license for an underground repository, which is planned to start its operation around the year 2020. The final disposal of high-level radioactive waste poses questions of long-term safety and possible processes of radionuclide release in the EBS (Engineered Barrier System) and the geosphere are also modelled as well as possible transport routes in the bedrock. Posiva has also established a monitoring program for the environment and is also conducting modelling of the surface environment and biosphere in Olkiluoto and the surrounding reference area. These serve the purpose of both, site description and modelling of the transport and accumulation of possible radionuclide releases in the surface environment and biosphere. The process of modelling used by Posiva requires the division of the surface environment and biosphere into several categories. In Posiva's classification, ecosystems are divided to two categories: terrestrial and aquatic with terrestrial divided to forests and agricultural areas. These categories are further divided to ecosystem types which include: lake, river, forest, cropland and sea (with coastal sea as a separate type). These types are even further divided to 14 ecosystem sub-types and behind these sub-types, a total of 24 biotopes exist. Soil and sediment types are also classified to 7 classes. The methodology behind the selection of these biotopes and their connection to the modelling of radionuclide transport in the surface environment is further described in the text. Some of the ecosystem types and biotopes are absent in present-day Olkiluoto area, which necessitates the use of reference targets, such as lakes, rivers

Signs of growth in the industry

International Nuclear Information System (INIS)

Taipale, Tellervo

2002-01-01

Full scale of methods in use - long-term work is the basis. Long-term work in communication has been proved to be an excellent basis for a nuclear power company. Teollisuuden Voima Oy (TVO), which has been producing electricity by nuclear power for the Finnish society for 23 years, bears its responsibilities for environmental issues and as member of society. Maintaining good relations to the local municipalities and reporters is one of our strengths. Nuclear power is a way to produce electricity in an environmentally friendly, safe and economical way. TVO, as a modem nuclear power company, acknowledges its responsibilities for the environment and the entire society. In dealing with these responsibilities, a culture of effective, open and honest communication is crucial. The concept of our communication work was already laid down in the 70's We do our best to inform our neighbours and interest groups openly about all that happens at Olkiluoto NPP. We do not only tell the good news, we also tell if there are any problems, e.g. in the production. We apply several different methods to allow everyone to choose their favourite way of finding the information they want. We produce annual publications and videos, organise visits to the plant, participate in fairs and have an up-to-date www-site. A lot of information is available and the doors are open for visitors. Annually, we produce almost 60 publications, e.g. magazines, newspapers, environmental and economical reports and different kinds of brochures. A lot of people visit Olkiluoto NPP - we receive about 16,000 guests each year. We have four basic arguments in our messages to promote the new NPP unit: it covers additional electricity demand and replaces old fossil-fuel power plants, it enables Finland to meet the Kyoto commitments, it secures a stable and predictable electricity price and reduces the dependency on imported electricity. All these messages are included in the printing materials and www-sites, as well as

Array-level stability enhancement of 50 nm AlxOy ReRAM

Science.gov (United States)

Iwasaki, Tomoko Ogura; Ning, Sheyang; Yamazawa, Hiroki; Takeuchi, Ken

2015-12-01

ReRAM's low voltage and low current programmability are attractive features to solve the scaling issues of conventional floating gate Flash. However, read instability in ReRAM is a critical issue, due to random telegraph noise (RTN), sensitivity to disturb and retention. In this work, the array-level characteristics of read stability in 50 nm AlxOy ReRAM are investigated and a circuit technique to improve stability is proposed and evaluated. First, in order to quantitatively assess memory cell stability, a method of stability characterization is defined. Next, based on this methodology, a proposal to improve read stability, called ;stability check loop; is evaluated. The stability check loop is a stability verification procedure, by which, instability improvement of 7×, and read error rate improvement of 40% are obtained.

3 Investment Scenarios for Fast Reactors

International Nuclear Information System (INIS)

Shoai Tehrani, Bianka; Da Costa, Pascal

2013-01-01

Results: • 4 families of scenarios: – In each of them, 3 options for national nuclear policy → 12 scenarios; – 3 favorable to FRs: - “climate constraint” with strong pro-nuclear policy - “climate constraint” with moderate pro-nuclear policy - “totally green” with strong pro-nuclear policy. • Business As Usual is not favorable to Fast Reactors; Fast reactors deployment: - Needs strong climate policy - Is viable in case of important renewable progress as long as climate policy is strong. International perspective: • Results are valid for Europe, other drivers being likely to be more important in other countries : high growth and demand (Asia); • With strong contrasts between European countries. Further research: • Finer modeling of drivers with unclear influence (clustered and excluded variables): Influence of weak signals

D-3He fueled FRC reactor 'ARTEMIS-L'

International Nuclear Information System (INIS)

Momota, Hiromu; Tomita, Yukihiro; Ishida, Akio; Kohzaki, Yasuji; Nakao, Yasuyuki; Nishikawa, Masabumi; Ohi, Shoichi; Ohnishi, Masami.

1992-09-01

A neutron-lean D- 3 He fueled field reversed configuration (FRC) fusion reactor is studied on the bases of former high-efficiency ARTEMIS design. Certain improvements such as effective axial contracting plasma heating and cusp-type direct energy converters as well as an empirical scale of the energy confinement are introduced. The resultant total neutron load onto the first wall of the plasma chamber is as low as 0.1 MW/m 2 , which enable the life of the first wall or the structural materials to be longer than the whole life of the reactor. The attractive characteristics of the neutron-lean reactor follow in the ARTEMIS design: it is socially acceptable in views of radioactivity and fuel resources, and the cost of electricity appears to be cheap compared with that from a light water reactor. Critical physics and engineering issues for performing the ARTEMIS-L reactor are clarified. (author)

Lasten ulkovaatteiden kevät-kesämalliston seinäesillepano-ohje : case: Reima Oy

OpenAIRE

Järvenoja, Annika

2014-01-01

Opinnäytetyö tehtiin Reima Oy:lle. Työn tavoitteena oli kerätä asiakkaiden mielipiteitä lasten ulkovaatteiden seinäesillepanoista. Tutkimuksen pohjalta tehtiin seinäesillepano-ohje kevät-kesämallistolle. Reiman omat myymälät sijaitsevat pääosin Kiinassa, Suomessa ja Venäjällä. Kevät-kesämalliston tuotteiden vaihtelevuus välikausivaatteista ohuisiin aurinkosuojavaatteisiin koettiin myymälässä haasteelliseksi. Tavoitteena on tutkia lisäksi, mistä maiden kulttuurien eroavaisuudet johtuvat, ja mi...

HENKILÖSTÖVUOKRAUSPALVELUN SÄHKÖISEN MARKKINOINTIVIESTINNÄN KEHITTÄMINEN : Case Nordic JobCentre Oy

OpenAIRE

Salmela, Heli

2016-01-01

Vuokratyön käyttö on ollut kasvussa jo pitkään. Erityisesti rakennus- ja teollisuusaloilla vuokratyön käyttö on hyvin yleistä, vaikka edelleen toimialan maine on joiltain osin negatiivinen. Kilpailu on kuitenkin kovaa, ja erityisesti Etelä-Suomen alueella vuokratyöfirmoja on syntynyt viime vuosikymmenen aikana runsaasti. Opinnäytetyöni on sähköisen markkinointiviestinnän kehitystehtävä, jonka toimeksiantajana on oululaislähtöinen henkilöstövuokrausyritys Nordic JobCentre Oy. Kehitystehtäv...

Description of the RA-3 research reactor as a model facility

International Nuclear Information System (INIS)

Vicens, Hugo E.; Quintana, Jorge A.

2001-01-01

The Argentine RA-3 reactor is described as a model facility for the information to be provided to the IAEA in accordance with the requirements of the Model Additional Protocol. RA-3 reactor was designed as a 5 MW swimming pool reactor, moderated and cooled with light water. Its fuel was 90% enriched uranium. The reactor started its operation in 1967, has been modified and improved in many components, including the core, that now is fueled with moderately enriched uranium

A visible and infrared study of the eclipsing dwarf nova Oy Carinae

International Nuclear Information System (INIS)

Berriman, G.

1984-01-01

This paper presents three simultaneous visible (V) and infrared (J,H,K) light curves of the eclipsing dwarf nova binary system OY Carinae in quiescence. The infrared light curves show a secondary minimum, not seen in the visible, which is the ellipsoidal variations of the red dwarf and its eclipse by the accretion disc surrounding the white dwarf companion. The red star, an M dwarf, supplies between 30 and 60 per cent of the total light at J,H and K. This requires that the system is between 100 and 300 pc away. The infrared continuum of the accretion disc around the white dwarf companion comes largely from the optically thin gas giving rise to the emission lines seen in the visible and ultraviolet. (author)

Effects of Dăoyĭn Qìgōng in postpolio syndrome patients with cold intolerance

Directory of Open Access Journals (Sweden)

Paulo Eduardo Ramos

2012-09-01

Full Text Available Postpolio syndrome (PPS is characterized by progressive muscle weakness due to former infection with poliomyelitis and can be associated with other symptoms such as cold intolerance (CI. Dăoyĭn Qìgōng (DQ is a technique in Traditional Chinese Medicine that impacts the circulation of energy and blood. OBJECTIVE: It was to verify the effects of DQ in PPS patients complaining of cold intolerance. METHODS: Ten PPS patients were assessed using the visual analogue scale (VAS adapted for CI before and after intervention with DQ; patients practiced it in a sitting position for 40 minutes, 3 times per week over 3 consecutive months. Patients were reassessed three months after ceasing DQ. RESULTS: There was a statistically significant difference in local and systemic VAS-Cold both at the end of DQ training and three months past the end of this. CONCLUSION: The DQ technique ameliorated CI complaints in patients with PPS.

The simplified P3 approach on a trigonal geometry in the nodal reactor code DYN3D

International Nuclear Information System (INIS)

Duerigen, S.; Fridman, E.

2011-01-01

DYN3D is a three-dimensional nodal diffusion code for steady-state and transient analyses of Light-Water Reactors with square and hexagonal fuel assembly geometries. Currently, several versions of the DYN3D code are available including a multi-group diffusion and a simplified P 3 (SP 3 ) neutron transport option. In this work, the multi-group SP 3 method based on trigonal-z geometry was developed. The method is applicable to the analysis of reactor cores with hexagonal fuel assemblies and allows flexible mesh refinement, which is of particular importance for WWER-type Pressurized Water Reactors as well as for innovative reactor concepts including block type High-Temperature Reactors and Sodium Fast Reactors. In this paper, the theoretical background for the trigonal SP 3 methodology is outlined and the results of a preliminary verification analysis are presented by means of a simplified WWER-440 core test example. The accordant cross sections and reference solutions were produced by the Monte Carlo code SERPENT. The DYN3D results are in good agreement with the reference solutions. The average deviation in the nodal power distribution is about 1%. (Authors)

3D computer visualization and animation of CANDU reactor core

International Nuclear Information System (INIS)

Qian, T.; Echlin, M.; Tonner, P.; Sur, B.

1999-01-01

Three-dimensional (3D) computer visualization and animation models of typical CANDU reactor cores (Darlington, Point Lepreau) have been developed using world-wide-web (WWW) browser based tools: JavaScript, hyper-text-markup language (HTML) and virtual reality modeling language (VRML). The 3D models provide three-dimensional views of internal control and monitoring structures in the reactor core, such as fuel channels, flux detectors, liquid zone controllers, zone boundaries, shutoff rods, poison injection tubes, ion chambers. Animations have been developed based on real in-core flux detector responses and rod position data from reactor shutdown. The animations show flux changing inside the reactor core with the drop of shutoff rods and/or the injection of liquid poison. The 3D models also provide hypertext links to documents giving specifications and historical data for particular components. Data in HTML format (or other format such as PDF, etc.) can be shown in text, tables, plots, drawings, etc., and further links to other sources of data can also be embedded. This paper summarizes the use of these WWW browser based tools, and describes the resulting 3D reactor core static and dynamic models. Potential applications of the models are discussed. (author)

Experiment for search for sterile neutrino at SM-3 reactor

Science.gov (United States)

Serebrov, A. P.; Ivochkin, V. G.; Samoylov, R. M.; Fomin, A. K.; Zinoviev, V. G.; Neustroev, P. V.; Golovtsov, V. L.; Gruzinsky, N. V.; Solovey, V. A.; Cherniy, A. V.; Zherebtsov, O. M.; Martemyanov, V. P.; Zinoev, V. G.; Tarasenkov, V. G.; Aleshin, V. I.; Petelin, A. L.; Pavlov, S. V.; Izhutov, A. L.; Sazontov, S. A.; Ryazanov, D. K.; Gromov, M. O.; Afanasiev, V. V.; Matrosov, L. N.; Matrosova, M. Yu.

2016-11-01

In connection with the question of possible existence of sterile neutrino the laboratory on the basis of SM-3 reactor was created to search for oscillations of reactor antineutrino. A prototype of a neutrino detector with scintillator volume of 400 l can be moved at the distance of 6-11 m from the reactor core. The measurements of background conditions have been made. It is shown that the main experimental problem is associated with cosmic radiation background. Test measurements of dependence of a reactor antineutrino flux on the distance from a reactor core have been made. The prospects of search for oscillations of reactor antineutrino at short distances are discussed.

Integral test of JENDL-3.3 on fast reactors

International Nuclear Information System (INIS)

Chiba, Gou; Hazama, Taira

2003-05-01

An integral test has been carried out to evaluate a performance of evaluated nuclear data library JENDL-3.3, which was newly released, in a view of applying neutronics analyses of fast reactors. Japan Nuclear Cycle Development Institute has a large amount of data of critical assembly experiments (ZPPR, BFS, MOZART and FCA) and power reactor tests (JOYO). The database was utilized in this test. In plutonium loaded cores, an improvement was observed about 0.3% ε k in criticality and 5% in the non-leakage term of sodium void reactivity by a revision form JENDL-3.2 to -3.3. These results shoed that the revision is valid in plutonium loaded cores. In uranium loaded cores, dependence of C/E values on control rod position became smaller in control rod worth in ZPPR cores. On the other hand, C/E values became worse both in criticality (0.6%εk) and in sodium void reactivity (30%) in BFS cores. The main cause was a revision of uranium-235 capture cross section, and it could not be concluded whether the revision is valid or not in uranium loaded cores. It is necessary to carry out a validation test at other independent critical experiments in which uranium fuel is used. (author)

TEHDASPROSESSIN 3D-MALLINNUS, PROSESSI- JA LAITEKUVAUS

OpenAIRE

Huhtala, Aki

2010-01-01

Opinnäytetyö tehtiin tilauksesta AkkuSer Oy:lle. Tarkoituksena oli kartoittaa tehdasprosessi, selvittää laitetiedot ja luoda 3D-malli tehdasprosessista. Prosessin toiminta kuvattiin. Selvitettiin, miten prosessi eteni ja millä tavoin AkkuSer Oy kierrättää ongelma jätettä. Selitetään, mitkä ovat prosessin lähtökohdat ja mitä tuotteita prosessissa syntyi. Tämän lisäksi mitattiin kuljettimien sekä taajuusmuuttajatoimisten kuljettimien pyörimisnopeuksia. Tehdasprosessi kuvattiin Visual Compon...

Maa-analyysi liiketoiminnallisesta näkökulmasta tukemassa kansainvälistymisstrategiaa : Case: SuperApp Oy, Yhdysvallat

OpenAIRE

Laiho, Tea

2016-01-01

Tämä opinnäytetyö toteutettiin toimeksiantona lahtelaisen KIBS-palveluja eli osaamisintensiivisiä liike-elämän palveluja tarjoavan SuperApp Oy:n käyttöön. Tutkimuksen tavoitteena oli kartoittaa kohdemaan potentiaali kansainvälisen liiketoiminnan ja palveluyrityksen näkökulmasta. Tarkoituksena oli antaa case-yritykselle työkaluja ja avaimia, sen laajentaessa liiketoimintaansa tulevaisuudessa ulkomaille, erityisesti Yhdysvaltoihin. Tämä tutkimus käsittelee osaa kansainvälistymisstrategiaa varte...

Yrittäjyyden myytit ja möröt : case: Yritystakomo Oy, Oulu

OpenAIRE

Merentie, Elisa

2012-01-01

Tämän tutkimuksen tarkoituksena on selvittää mitkä yrittäjyyden myytit ja möröt vaikuttavat aloittavan yrittäjän uskallukseen aloittaa yrittäjänä. Ilmiötä tarkastelin tapausyrityksen, Yritystakomo Oy:n kautta. Ovatko vuosia kestäneet yrittäjyyden myytit, uskomukset aiheellisia ja sovellettavissa tämän päivän aloittelevan yrittäjän ajatusmaailmaan? Aihe on valittu, koska yrittäjyydestä ja sen edistämisestä puhutaan paljon, myös sen helppoudesta, mutta siitä huolimatta yrittäjyysidea jää usein...

3D CAD model of the subcritical nuclear reactor of IPN

International Nuclear Information System (INIS)

Pahuamba V, F. de J.; Delfin L, A.; Gomez T, A.; Ibarra R, G.; Del Valle G, E.; Sanchez R, A.

2016-09-01

The three-dimensional (3D) CAD model of the subcritical reactor Chicago model 9000 of Instituto Politecnico Nacional (IPN) allows obtaining a 3D view with the dimensions of each of its components, such as: natural uranium cylindrical rods, fuel elements, hexagonal reactor core arrangement, cylindrical stainless steel tank containing the core, fuel element support grids and reactor water cleaning system. As a starting point for the development of the model, the Chicago model 9000 subcritical reactor manual provided by the manufacturer was used, the measurement and verification of the components to adapt the geometric, physical and mechanical characteristics was carried out and materials standards were used to obtain a design that allows to elaborate a new manual according to the specifications. In addition, the 3D models of the building of the Advanced Physics Laboratory, neutron generator, cobalt source and the corridors connecting to the subcritical reactor facility were developed, allowing an animated ride, developed by computer-aided design software. The manual provided by the company Nuclear Chicago, dates from the year 1959 and presents diverse deviations in the design and dimensions of the reactor components. The model developed; in addition to supporting the development of the new manual represents a learning tool to visualize the reactor components. (Author)

Pienyrityksen B2B-myyntiprosessi Venäjän markkinoilla: Case Perfect Getaway Oy

OpenAIRE

Prokkola, Joni

2014-01-01

Opinnäytetyön tavoitteena oli luoda Perfect Getaway Oy:lle B2B myyntiprosessi suoramyyntiä varten Venäjän markkinoille. Yritys on luksuselämyksiä tarjoava matkailualan yritys, jonka pääkohderyhmänä ovat suomalaiset sekä venäläiset yksityis- ja yritysasiakkaat. Tavoitteena oli selvittää miten kylmäsoittoja voitaisiin käyttää hyväksi suoramyynnissä. Toimeksiantajana toimi yrityksen toimitusjohtaja. Opinnäytetyön kirjoittaja on myös työsuhteessa kyseiseen yritykseen. Tuotoksena toimisivat kaksi ...

ABB Oy Motors and Generators -yksikön energiansyöttöjen erotus- ja lukitusohjeet

OpenAIRE

Chi, Henri

2017-01-01

Tämä insinöörityö toteutettiin Quant Finland Oy:lle, joka vastaa Helsingin Pitäjänmäen ABB Motors and Generators -tehtaan kunnossapidosta. Insinöörityön tavoitteena oli luoda energiansyöttöjen lukitus- ja erotusohjeet tehtaan koneistoille, laitteille ja järjestelmille. Tehtaalla on isoja koneita ja laitteita, jonka erotus- ja lukitustoimenpiteitä kaikki eivät osaa. Näin ollen on haluttu mahdollistaa turvallinen työskentely tarjoamalla selkeät erotus- ja lukitusohjeet. ABB Motors and Gene...

MS CRM 3.0 Dymamics SBS

OpenAIRE

Salmensuu, Jussi

2008-01-01

Tämä opinnäytetyö käsittelee kuinka Microsoftin Customer Relationship Management (CRM) 3.0 SBS toimisi Primanet Oy:lle, joka on pieni IT-ylläpitoa ja - palveluja tarjoava yritys. Opinnäytetyön tavoitteena oli luoda testiympäristö, joka kattaisi CRM - järjestelmän testaamisen seuraavin osa-aluein: toimeksiantojen käsittely, asiakastietokannan luominen ja ylläpitäminen, työjonojen hallinta ja niihin liittyvät toiminnot. Primanet Oy päätti lähteä kokeilemaan CRM -järjestelmän soveltuvuutta, kosk...

Study on the reactivity behavior partially loaded reactor cores using SIMULATE-3

International Nuclear Information System (INIS)

Holzer, Robert; Zeitz, Andreas; Grimminger, Werner; Lubczyk, Tobias

2009-01-01

The reactor core design for the NPP Gundremmingen unit B and C is performed since several years using the validated 3D reactor core calculation program SIMULATE-3. The authors describe a special application of the program to study the reactivity for different partial core loadings. Based on the comparison with results of the program CASMO-4 the program SIMULATE-3 was validated for the calculation of partially loaded reactor cores. For the planned reactor operation in NPP Gundremmingen using new MOX fuel elements the reactivity behavior was studied with respect to the KTA-Code requirements.

Massaräätälöityjen tuotteiden prosessin hallinta ja konfigurointi : Selvitystyö yritykselle Kera Group Oy

OpenAIRE

Saares, Sanna

2017-01-01

Opinnäytetyön aiheena oli selvittää, mikä ohjelmistokokonaisuus soveltuisi parhaiten terassituotteiden suunnittelun ja myynnin tueksi. Opinnäytetyö tehtiin Orimattilassa sijaitsevalle muovi- ja lasituotteisiin erikoistuneelle Kera Group Oy:lle. Tuotealueena olivat lasitetut valmisterassit, joita toimitetaan pientalojen yhteyteen. Vaikka valmiita terassityyppejä on rajattu määrä, joudutaan suurin osa tuotteista suunnittelemaan ja valmistamaan yksilöllisesti. Uuden tuotteen suunnittelu on ...

Estimation of reactor pool water temperature after shutdown in JRR-3M

International Nuclear Information System (INIS)

Yagi, Masahiro; Sato, Mitsugu; Kakefuda, Kazuhiro

1999-01-01

The reactor pool water temperature increasing by the decay heat was estimated by calculation. The reactor pool water temperature was calculated by increased enthalpy that was estimated by the reactor decay heat, the heat released from the reactor biological shielding concrete, reactor pool water surface, the heat conduction from the canal and the core inlet piping. These results of calculation were compared with the past measured data. As the results of estimation, after the JRR-3M shutdown, the calculated reactor pool temperature first increased sharply. This is because the decay heat was the major contribution. And then, rate of increased reactor pool temperature decreased. This is because the ratio of heat released from reactor biological shielding concrete and core inlet piping to the decay heat increased. Besides, the calculated reactor pool water temperature agreed with the past measured data in consequence of correcting the decay heat and the released heat. The corrected coefficient k 1 of decay heat was 0.74 - 0.80. And the corrected coefficient k 2 of heat released from the reactor biological shielding concrete was 3.5 - 4.5. (author)

Franchising - liiketoiminnan käynnistämiseen vaadittavat toimenpiteet ja haasteet organisaatiossa : case: Touring Cars Finland Oy

OpenAIRE

Aaltonen, Suvi

2013-01-01

Tämän opinnäytetyön tavoitteena oli selvittää franchise-liiketoiminnan käynnistämiseen liittyviä toimenpiteitä ja haasteita organisaatiossa. Opinnäytetyön teoriaosuus selvittää niitä asioita, joita franchise-antajan ja –ottajan on huomioitava franchise-liiketoimintaa aloittaessaan. Case-yrityksenä on suomalainen Touring Cars Finland Oy. Case-yritys siirtyy käyttämään franchise-liiketoimintamallia vuosien 2013–2018 aikana yrityksen kaikissa toimipisteissä Suomessa ja ulkomailla. Touring Ca...

RELAP5/MOD 3.3 analysis of Reactor Coolant Pump Trip event at NPP Krsko

International Nuclear Information System (INIS)

Bencik, V.; Debrecin, N.; Foretic, D.

2003-01-01

In the paper the results of the RELAP5/MOD 3.3 analysis of the Reactor Coolant Pump (RCP) Trip event at NPP Krsko are presented. The event was initiated by an operator action aimed to prevent the RCP 2 bearing damage. The action consisted of a power reduction, that lasted for 50 minutes, followed by a reactor and a subsequent RCP 2 trip when the reactor power was reduced to 28 %. Two minutes after reactor trip, the Main Steam Isolation Valves (MSIV) were isolated and the steam dump flow was closed. On the secondary side the Steam Generator (SG) pressure rose until SG 1 Safety Valve (SV) 1 opened. The realistic RELAP5/MOD 3.3 analysis has been performed in order to model the particular plant behavior caused by operator actions. The comparison of the RELAP5/MOD 3.3 results with the measurement for the power reduction transient has shown small differences for the major parameters (nuclear power, average temperature, secondary pressure). The main trends and physical phenomena following the RCP Trip event were well reproduced in the analysis. The parameters that have the major influence on transient results have been identified. In the paper the influence of SG 1 relief and SV valves on transient results was investigated more closely. (author)

Olkiluoto surface hydrological modelling: Update 2012 including salt transport modelling

International Nuclear Information System (INIS)

Karvonen, T.

2013-11-01

Posiva Oy is responsible for implementing a final disposal program for spent nuclear fuel of its owners Teollisuuden Voima Oyj and Fortum Power and Heat Oy. The spent nuclear fuel is planned to be disposed at a depth of about 400-450 meters in the crystalline bedrock at the Olkiluoto site. Leakages located at or close to spent fuel repository may give rise to the upconing of deep highly saline groundwater and this is a concern with regard to the performance of the tunnel backfill material after the closure of the tunnels. Therefore a salt transport sub-model was added to the Olkiluoto surface hydrological model (SHYD). The other improvements include update of the particle tracking algorithm and possibility to estimate the influence of open drillholes in a case where overpressure in inflatable packers decreases causing a hydraulic short-circuit between hydrogeological zones HZ19 and HZ20 along the drillhole. Four new hydrogeological zones HZ056, HZ146, BFZ100 and HZ039 were added to the model. In addition, zones HZ20A and HZ20B intersect with each other in the new structure model, which influences salinity upconing caused by leakages in shafts. The aim of the modelling of long-term influence of ONKALO, shafts and repository tunnels provide computational results that can be used to suggest limits for allowed leakages. The model input data included all the existing leakages into ONKALO (35-38 l/min) and shafts in the present day conditions. The influence of shafts was computed using eight different values for total shaft leakage: 5, 11, 20, 30, 40, 50, 60 and 70 l/min. The selection of the leakage criteria for shafts was influenced by the fact that upconing of saline water increases TDS-values close to the repository areas although HZ20B does not intersect any deposition tunnels. The total limit for all leakages was suggested to be 120 l/min. The limit for HZ20 zones was proposed to be 40 l/min: about 5 l/min the present day leakages to access tunnel, 25 l/min from

Olkiluoto surface hydrological modelling: Update 2012 including salt transport modelling

Energy Technology Data Exchange (ETDEWEB)

Karvonen, T. [WaterHope, Helsinki (Finland)

2013-11-15

Posiva Oy is responsible for implementing a final disposal program for spent nuclear fuel of its owners Teollisuuden Voima Oyj and Fortum Power and Heat Oy. The spent nuclear fuel is planned to be disposed at a depth of about 400-450 meters in the crystalline bedrock at the Olkiluoto site. Leakages located at or close to spent fuel repository may give rise to the upconing of deep highly saline groundwater and this is a concern with regard to the performance of the tunnel backfill material after the closure of the tunnels. Therefore a salt transport sub-model was added to the Olkiluoto surface hydrological model (SHYD). The other improvements include update of the particle tracking algorithm and possibility to estimate the influence of open drillholes in a case where overpressure in inflatable packers decreases causing a hydraulic short-circuit between hydrogeological zones HZ19 and HZ20 along the drillhole. Four new hydrogeological zones HZ056, HZ146, BFZ100 and HZ039 were added to the model. In addition, zones HZ20A and HZ20B intersect with each other in the new structure model, which influences salinity upconing caused by leakages in shafts. The aim of the modelling of long-term influence of ONKALO, shafts and repository tunnels provide computational results that can be used to suggest limits for allowed leakages. The model input data included all the existing leakages into ONKALO (35-38 l/min) and shafts in the present day conditions. The influence of shafts was computed using eight different values for total shaft leakage: 5, 11, 20, 30, 40, 50, 60 and 70 l/min. The selection of the leakage criteria for shafts was influenced by the fact that upconing of saline water increases TDS-values close to the repository areas although HZ20B does not intersect any deposition tunnels. The total limit for all leakages was suggested to be 120 l/min. The limit for HZ20 zones was proposed to be 40 l/min: about 5 l/min the present day leakages to access tunnel, 25 l/min from

Modelling of MOCVD Reactor: New 3D Approach

Science.gov (United States)

Raj, E.; Lisik, Z.; Niedzielski, P.; Ruta, L.; Turczynski, M.; Wang, X.; Waag, A.

2014-04-01

The paper presents comparison of two different 3D models of vertical, rotating disc MOCVD reactor used for 3D GaN structure growth. The first one is based on the reactor symmetry, while the second, novel one incorporates only single line of showerhead nozzles. It is shown that both of them can be applied interchangeably regarding the phenomena taking place within the processing area. Moreover, the importance of boundary conditions regarding proper modelling of showerhead cooling and the significance of thermal radiation on temperature field within the modelled structure are presented and analysed. The last phenomenon is erroneously neglected in most of the hitherto studies.

Modelling of MOCVD reactor: new 3D approach

International Nuclear Information System (INIS)

Raj, E; Lisik, Z; Niedzielski, P; Ruta, L; Turczynski, M; Wang, X; Waag, A

2014-01-01

The paper presents comparison of two different 3D models of vertical, rotating disc MOCVD reactor used for 3D GaN structure growth. The first one is based on the reactor symmetry, while the second, novel one incorporates only single line of showerhead nozzles. It is shown that both of them can be applied interchangeably regarding the phenomena taking place within the processing area. Moreover, the importance of boundary conditions regarding proper modelling of showerhead cooling and the significance of thermal radiation on temperature field within the modelled structure are presented and analysed. The last phenomenon is erroneously neglected in most of the hitherto studies.

Structure and charge transfer correlated with oxygen content for a Y0.8Ca0.2Ba2Cu3Oy (y = 6.84 6.32) system: a positron study

Science.gov (United States)

Cao, Shixun; Li, Lingwei; Liu, Fen; Li, Wenfeng; Chi, Changyun; Jing, Chao; Zhang, Jincang

2005-05-01

The structure and charge transfer correlated with oxygen content are studied by measuring the positron lifetime parameters of the Y0.8Ca0.2Ba2Cu3Oy system with a large range of oxygen content (y = 6.84-6.32). The local electron density ne is evaluated from the positron lifetime data. The positron lifetime parameters show a clear change around y = 6.50 where the compounds undergo the orthorhombic-tetragonal phase transition. The effect of ne and oxygen content on the structure, charge transfer and superconductivity are discussed. With the decrease of oxygen content y, O(4) tends to the Cu(1) site, causing carrier localization, and accordingly, the decrease of ne. This would prove that the localized carriers (electrons and holes) in the Cu-O chain region have great influence on the superconductivity by affecting the charge transfer between the reservoir layers and the conducting layers. The positron annihilation mechanism and its relation with superconductivity are also discussed.

Characteristics of D(-3)He fueled FRC reactor: ARTEMIS-L

Science.gov (United States)

Momota, H.; Motojima, O.; Okamoto, M.; Sudo, S.; Tomita, Y.; Yamaguchi, S.; Iiyoshi, A.; Onozuka, M.; Ohnishi, M.; Uenosono, C.

1993-11-01

The characteristics of D(-3)He fueled commercial fusion reactor ARTEMIS-L are discussed. By using favorable characteristics of a field-reversed configuration, the fusion plasma of ARTEMIS-L becomes compact and its veta-value is extremely high. Consequently, it is possible to construct an economical fusion power plant based on this concept. The life of the structural materials is found during the full reactor life (30 years) and the safety of the reactor is intrinsic to D(-3)He fuels. The amount of disposed materials is rather small and the level of the intruder dose is so low that the plant appears to be acceptable in regards to the environment.

PR-EDB: Power Reactor Embrittlement Database Version 3

International Nuclear Information System (INIS)

Wang, Jy-An John; Subramani, Ranjit

2008-01-01

The aging and degradation of light-water reactor pressure vessels is of particular concern because of their relevance to plant integrity and the magnitude of the expected irradiation embrittlement. The radiation embrittlement of reactor pressure vessel materials depends on many factors, such as neutron fluence, flux, and energy spectrum, irradiation temperature, and preirradiation material history and chemical compositions. These factors must be considered to reliably predict pressure vessel embrittlement and to ensure the safe operation of the reactor. Large amounts of data from surveillance capsules are needed to develop a generally applicable damage prediction model that can be used for industry standards and regulatory guides. Furthermore, the investigations of regulatory issues such as vessel integrity over plant life, vessel failure, and sufficiency of current codes, Standard Review Plans (SRPs), and Guides for license renewal can be greatly expedited by the use of a well-designed computerized database. The Power Reactor Embrittlement Database (PR-EDB) is such a comprehensive collection of data for U.S. designed commercial nuclear reactors. The current version of the PR-EDB lists the test results of 104 heat-affected-zone (HAZ) materials, 115 weld materials, and 141 base materials, including 103 plates, 35 forgings, and 3 correlation monitor materials that were irradiated in 321 capsules from 106 commercial power reactors. The data files are given in dBASE format and can be accessed with any personal computer using the Windows operating system. 'User-friendly' utility programs have been written to investigate radiation embrittlement using this database. Utility programs allow the user to retrieve, select and manipulate specific data, display data to the screen or printer, and fit and plot Charpy impact data. The PR-EDB Version 3.0 upgrades Version 2.0. The package was developed based on the Microsoft .NET framework technology and uses Microsoft Access for

PR-EDB: Power Reactor Embrittlement Database - Version 3

Energy Technology Data Exchange (ETDEWEB)

Wang, Jy-An John [ORNL; Subramani, Ranjit [ORNL

2008-03-01

The aging and degradation of light-water reactor pressure vessels is of particular concern because of their relevance to plant integrity and the magnitude of the expected irradiation embrittlement. The radiation embrittlement of reactor pressure vessel materials depends on many factors, such as neutron fluence, flux, and energy spectrum, irradiation temperature, and preirradiation material history and chemical compositions. These factors must be considered to reliably predict pressure vessel embrittlement and to ensure the safe operation of the reactor. Large amounts of data from surveillance capsules are needed to develop a generally applicable damage prediction model that can be used for industry standards and regulatory guides. Furthermore, the investigations of regulatory issues such as vessel integrity over plant life, vessel failure, and sufficiency of current codes, Standard Review Plans (SRPs), and Guides for license renewal can be greatly expedited by the use of a well-designed computerized database. The Power Reactor Embrittlement Database (PR-EDB) is such a comprehensive collection of data for U.S. designed commercial nuclear reactors. The current version of the PR-EDB lists the test results of 104 heat-affected-zone (HAZ) materials, 115 weld materials, and 141 base materials, including 103 plates, 35 forgings, and 3 correlation monitor materials that were irradiated in 321 capsules from 106 commercial power reactors. The data files are given in dBASE format and can be accessed with any personal computer using the Windows operating system. "User-friendly" utility programs have been written to investigate radiation embrittlement using this database. Utility programs allow the user to retrieve, select and manipulate specific data, display data to the screen or printer, and fit and plot Charpy impact data. The PR-EDB Version 3.0 upgrades Version 2.0. The package was developed based on the Microsoft .NET framework technology and uses Microsoft Access for

Operating reactors licensing actions summary. Vol. 3, No. 3

International Nuclear Information System (INIS)

1983-04-01

The operating reactors licensing actions summary is designed to provide the management of the Nuclear Regularory Commission (NRC) with an overview of licensing actions dealing with operating power and nonpower reactors. These reports utilize data collected from the Division of Licensing in the Office of Nuclear Reactor Regulation and are prepared by the Office of Management and Program Analysis. This summary report is published primarily for internal NRC use in managing the operating reactors licensing actions program

Report on generation IV technical working group 3 : liquid metal reactors

International Nuclear Information System (INIS)

Lineberry, M. J.; Rosen, S. L.; Sagayama, Y.

2002-01-01

This paper reports on the first round of R and D roadmap activities of the Generation IV (Gen IV) Technical Working Group (TWG) 3, on liquid metal-cooled reactors. Liquid metal coolants give rise to fast spectrum systems, and thus the reactor systems considered in this TWG are all fast reactors. Gas-cooled fast reactors are considered in the context of TWG 2. As is noted in other Gen IV papers, this first round activity is termed ''screening for potential'', and includes collecting the most complete set of liquid metal reactor/fuel cycle system concepts possible and evaluating the concepts against the Gen IV principles and goals. Those concepts or concept groups that meet the Gen IV principles and which are deemed to have reasonable potential to meet the Gen IV goals will pass to the next round of evaluation. Although we sometimes use the terms ''reactor'' or ''reactor system'' by themselves, the scope of the investigation by TWG 3 includes not only the reactor systems, but very importantly the closed fuel recycle system inevitably required by fast reactors. The response to the DOE Request for Information (RFI) on liquid metal reactor/fuel cycle systems from principal investigators, laboratories, corporations, and other institutions, was robust and gratifying. Thirty three liquid metal concept descriptions, from eight different countries, were ultimately received. The variation in the scope, depth, and completeness of the responses created a significant challenge for the group, but the TWG made a very significant effort not to screen out concepts early in the process

Benchmark tests of JENDL-3.2 for thermal and fast reactors

International Nuclear Information System (INIS)

Takano, Hideki

1995-01-01

Benchmark calculations for a variety of thermal and fast reactors have been performed by using the newly evaluated JENDL-3 Version-2 (JENDL-3.2) file. In the thermal reactor calculations for the uranium and plutonium fueled cores of TRX and TCA, the k eff and lattice parameters were well predicted. The fast reactor calculations for ZPPR-9 and FCA assemblies showed that the k eff , reactivity worth of Doppler, sodium void and control rod, and reaction rate distribution were in a very good agreement with the experiments. (author)

Nuclear research reactor 0.5 to 3 MW

Energy Technology Data Exchange (ETDEWEB)

NONE

1992-05-15

This nuclear reactor has been designed for radioisotope production, basic and applied research in reactor physics and nuclear engineering, neutron-beam experimentation, irradiation of various materials and training of scientific and technical personnel. It is located in the 'Production Area' of the Nuclear Technology Center. It is equipped with the necessary facilities for large-scale production of radioisotopes to be used in medicine as well as for other scientific and industrial purposes. In addition, it has a Neutronography Facility and the required equipment to perform Neutron-Activation Analysis. It is an open pool-type reactor, moderated and cooled with light water, fuelled with 20% enriched uranium. Its reflector are graphite and water. It has plate-type fuel elements clad in aluminium. The reactor core is located near the bottom of the demineralized water pool. It includes fuel elements, reflector and sample-holding devices for materials to be irradiated. This kind of configuration, which is widely used in research reactors, provides a high degree of safety since it prevents the core from becoming exposed under any circumstance and does not require any cooling system during reactor shutdown. Power output is between 0.5 to 3 MW{sub TH}, with a minimum thermal neutron flux of approx, 10{sup 13} n/cm{sup 2}{center_dot}sec, at irradiation zone almost with no modifications. Heat extraction is achieved by means of a cooling circuit which comprises two circulation pumps and a plate-type heat exchanger. Final heat dissipation to the atmosphere is performed through another cooling circuit which includes two circulation pumps and a cooling tower. Reactor control is accomplished with five neutron-absorbing rods positioned by means of especially designed elements and governed by the reactor's instrumentation and control system. Should an abnormal situation arise, gravity causes the rods to fall automatically, thus extinguishing the nuclear reaction. The reactor

Nuclear research reactor 0.5 to 3 MW

International Nuclear Information System (INIS)

1992-05-01

This nuclear reactor has been designed for radioisotope production, basic and applied research in reactor physics and nuclear engineering, neutron-beam experimentation, irradiation of various materials and training of scientific and technical personnel. It is located in the 'Production Area' of the Nuclear Technology Center. It is equipped with the necessary facilities for large-scale production of radioisotopes to be used in medicine as well as for other scientific and industrial purposes. In addition, it has a Neutronography Facility and the required equipment to perform Neutron-Activation Analysis. It is an open pool-type reactor, moderated and cooled with light water, fuelled with 20% enriched uranium. Its reflector are graphite and water. It has plate-type fuel elements clad in aluminium. The reactor core is located near the bottom of the demineralized water pool. It includes fuel elements, reflector and sample-holding devices for materials to be irradiated. This kind of configuration, which is widely used in research reactors, provides a high degree of safety since it prevents the core from becoming exposed under any circumstance and does not require any cooling system during reactor shutdown. Power output is between 0.5 to 3 MW TH , with a minimum thermal neutron flux of approx, 10 13 n/cm 2 ·sec, at irradiation zone almost with no modifications. Heat extraction is achieved by means of a cooling circuit which comprises two circulation pumps and a plate-type heat exchanger. Final heat dissipation to the atmosphere is performed through another cooling circuit which includes two circulation pumps and a cooling tower. Reactor control is accomplished with five neutron-absorbing rods positioned by means of especially designed elements and governed by the reactor's instrumentation and control system. Should an abnormal situation arise, gravity causes the rods to fall automatically, thus extinguishing the nuclear reaction. The reactor building has a ventilation

EL-3 dismantling of an experimental reactor

International Nuclear Information System (INIS)

1989-01-01

The EL3 experimental reactor has been definitively stopped in march 1979. Its decommissioning has been pronounced in the end of 1982. This article is consecrated at decontamination and dismantling works necessited by its passage at the dismantling level 2 [fr

The benefits of Outsourcing facility services when selecting right service provider for a hotel:Case Kämp Group Oy

OpenAIRE

Paudyal, Manoj; Acharya, Saroj

2015-01-01

This research paper examines about the outsourcing of facility services in the Kämp group of hotels. The scope of the study includes Facility Management, outsourcing facilities services, and the selection process of the service providers for a hotel. The research was carried at the hotels of Kämp group Oy in the Metropolitan Area of Helsinki. Facility management includes wide ranges of non-core functions such as Property management, real estates, design and technology. Activities such as secu...

Use of plate fuel elements for the RA3 reactor

International Nuclear Information System (INIS)

Parodi, C.; Parkanski, D.; Higa, M.; Marajofsky, A.

1992-01-01

The RA3 reactor is a pool reactor, redesigned for 5 MW dissipation. Nineteen plates are used in each fuel element. The utilization of 20% enriched U, gives the possibility of the development of rod type fuel with Al/U 3 O 8 cermets. The thermohydraulic and neutronic conditions are studied in this work in order to satisfy the stipulated power. In addition, the fabrication conditions of Al/U 3 O 8 and Al/U 3 O 8 /Zr H 2 cermets with densities within the limits imposed by the thermohydraulics and neutronics conditions are studied. (author)

Characteristics of D-3He fueled frc reactor: ARTEMIS-L

International Nuclear Information System (INIS)

Momota, H.; Motojima, O.; Okamoto, M.; Sudo, S.; Tomita, Y.; Yamaguchi, S.; Iiyoshi, A.; Onozuka, M.; Ohnishi, M.; Uenosono, C.

1993-11-01

The paper introduces briefly the scenario and discuss the attractive characteristics of D-3He fueled commercial fusion reactor ARTEMIS-L. By using favorable characteristics of a field-reversed configuration, the fusion plasma of ARTEMIS-L is compact and its beta-value is extremely high. One find consequently a possibility of constructing an economical fusion power power plant on this prospect. The life of the structural materials is sound during the full reactor life (30 years) and the safety of the reactor is intrinsic to D-3He fuels. The amount of disposed materials is rather small and the level of these intruder dose is so low that the plant appears to be acceptable in view of the environment. (author)

Summary of the 3rd workshop on the reduced-moderation water reactor

International Nuclear Information System (INIS)

Ishikawa, Nobuyuki; Nakatsuka, Tohru; Iwamura, Takamichi

2000-06-01

The research activities of a Reduced-Moderation Water Reactor (RMWR) are being performed for a development of the next generation water-cooled reactor. A workshop on the RMWR was held on March 3rd 2000 aiming to exchange information between JAERI and other organizations such as universities, laboratories, utilities and vendors. This report summarizes the contents of lectures and discussions on the workshop. The 1st workshop was held on March 1998 focusing on the review of the research activities and future research plan. The succeeding 2nd workshop was held on March 1999 focusing on the topics of the plutonium utilization in water-cooled reactors. The 3rd workshop was held on March 3rd 2000, which was attended by 77 participants. The workshop began with a lecture titled 'Recent Situation Related to Reduced-Moderation Water Reactor (RMWR)', followed by 'Program on MOX Fuel Utilization in Light Water Reactors' which is the mainstream scenario of plutonium utilization by utilities, and 'Feasibility Studies on Commercialized Fast Breeder Reactor Cycle System' mainly conducted by Japan Nuclear Cycle Development Institute (JNC). Also, following lectures were given as the recent research activities in JAERI: 'Progress in Design Study on Reduced-Moderation Water Reactors', 'Long-Term Scenarios of Power Reactors and Fuel Cycle Development and the Role of Reduced Moderation Water Reactors', 'Experimental and Analytical Study on Thermal Hydraulics' and Reactor Physics Experiment Plan using TCA'. At the end of the workshop, a general discussion was performed about the research and development of the RMWR. This report includes the original papers presented at the workshop and summaries of the questions and answers for each lecture and general discussion, as well as presentation viewgraphs, program and participant list as appendixes. The 7 of the presented papers are indexed individually. (J.P.N.)

Summary of the 3rd workshop on the reduced-moderation water reactor

Energy Technology Data Exchange (ETDEWEB)

Ishikawa, Nobuyuki; Nakatsuka, Tohru; Iwamura, Takamichi [eds.

2000-06-01

The research activities of a Reduced-Moderation Water Reactor (RMWR) are being performed for a development of the next generation water-cooled reactor. A workshop on the RMWR was held on March 3rd 2000 aiming to exchange information between JAERI and other organizations such as universities, laboratories, utilities and vendors. This report summarizes the contents of lectures and discussions on the workshop. The 1st workshop was held on March 1998 focusing on the review of the research activities and future research plan. The succeeding 2nd workshop was held on March 1999 focusing on the topics of the plutonium utilization in water-cooled reactors. The 3rd workshop was held on March 3rd 2000, which was attended by 77 participants. The workshop began with a lecture titled 'Recent Situation Related to Reduced-Moderation Water Reactor (RMWR)', followed by 'Program on MOX Fuel Utilization in Light Water Reactors' which is the mainstream scenario of plutonium utilization by utilities, and 'Feasibility Studies on Commercialized Fast Breeder Reactor Cycle System' mainly conducted by Japan Nuclear Cycle Development Institute (JNC). Also, following lectures were given as the recent research activities in JAERI: 'Progress in Design Study on Reduced-Moderation Water Reactors', 'Long-Term Scenarios of Power Reactors and Fuel Cycle Development and the Role of Reduced Moderation Water Reactors', 'Experimental and Analytical Study on Thermal Hydraulics' and Reactor Physics Experiment Plan using TCA'. At the end of the workshop, a general discussion was performed about the research and development of the RMWR. This report includes the original papers presented at the workshop and summaries of the questions and answers for each lecture and general discussion, as well as presentation viewgraphs, program and participant list as appendixes. The 7 of the presented papers are indexed individually. (J.P.N.)

3Oy SUPERCONDUCTOR IN BIPOLARON MODEL

African Journals Online (AJOL)

D. Mani Kongnine, A. Hamrita and K. Napo

1 sept. 2016 ... ION supraconducteurs, le cuivre et l'oxygène jouent un rôle prépondéran supraconductivité. Pour le composé supraconducteur YBa2Cu3O (Y re sont à distinguer : le cuivre Cu(1) qui forme avec les atomes d'ox t le cuivre Cu(2) qui forme avec l'oxygène les plans CuO2 [1, 2]. la supraconductivité dans les ...

Energiantuotannon ratkaisut Penttilän Puu Oy:n höyläämön sivuvirrasta

OpenAIRE

Kortelainen, Toni

2017-01-01

Tässä teknistaloudellisessa opinnäytetyössä selvitettiin Penttilän Puu Oy:n Joensuun höyläämölle taloudellisesti kannattavin energiantuotantoratkaisu. Käyttöikänsä päähän tulleen nykyisen ruuvisyötteisen arinakattilan tilalle on harkittu pienen kokoluokan sähkön ja lämmön yhteistuotantoa (pien-CHP), mikäli sellainen osoittautuu kannattavaksi. Muussa tapauksessa arinakattila vaihdetaan uuteen biokattilaan lähivuosina. Opinnäytetyön teknisessä osiossa esitellään korvaavat lämmitysvaih...

Integral test of JENDL-3.3 for thermal reactors

International Nuclear Information System (INIS)

Okumura, Keisuke; Mori, Takamasa

2003-01-01

Criticality benchmark testing was carried out for 59 experiments in various thermal reactors using a continues-energy Monte Carlo code MVP and its different libraries generated from JENDL-3.2, JENDL-3.3, JEF-2.2 and ENDF/B-VI (R8). From the benchmark results, we can say JENDL-3.3 generally gives better k eff values compared with other nuclear data libraries. However, further modification of JENDL-3.3 is expected to solve the following problems: 1) systematic underestimation of k eff depending on 235 U enrichment for the cores with low (less than 3wt.%) enriched uranium fueled cores, 2) dependence of C/E value of k eff on neutron spectrum and plutonium composition for MOX fueled cores. These are common problems for all of the nuclear data libraries used in this study. (author)

Main messages for the new nuclear unit in Finland

International Nuclear Information System (INIS)

Nikula, Anneli

2001-01-01

Teollisuuden Voima Oy (TVO) submitted last year, on 15th November, to the Council of State an application for a decision in principle concerning the construction of additional nuclear capacity. In this presentation, the main messages to decision makers for the new unit are discussed. Also the studies, which the messages are based on, are described. According to the Finnish Nuclear Energy Act, a company considering a nuclear plant project must apply for a decision in principle from the Government beforehand. The Government then decides whether the project is in accordance with the overall good of the society. If the decision is positive, it needs ratification by the Parliament. Before the Government decision, various interested parties are heard, including the municipality of location and the Radiation and Nuclear Safety Authority. The phase of statements is underway and they all will be ready by the end of March 2001. The entire application process takes approximately 1-2 years. The main messages arguing for more nuclear power in Finland were as follows. The new nuclear power plant unit replaces old fossil power plants and satisfies the need for additional demand of electricity; makes possible to meet the requirements of the Kyoto agreement; reduces the dependence on the imported energy and assures a stable and predictable electricity price. Messaging concerning a new nuclear power plant unit is made in co-operation with several organisations. The most important of them, in addition to the company shareholders, are the companies which have shown interested in the new project as well as the Finnish Energy Industries Federation (Finergy) and the Confederation of Finnish Industry and Employers. A transparency folder 'Nuclear electricity 2000' was put together of the messages in collaboration with the organisations. The folder was distributed to about 400 opinion leaders in the energy field. To ensure the consistency of the messages, several interest group meetings

Extension of the reactor dynamics code MGT-3D for pebblebed and blocktype high-temperature-reactors

International Nuclear Information System (INIS)

Shi, Dunfu

2015-01-01

The High Temperature Gas cooled Reactor (HTGR) is an improved, gas cooled nuclear reactor. It was chosen as one of the candidates of generation IV nuclear plants [1]. The reactor can be shut down automatically because of the negative reactivity feedback due to the temperature's increasing in designed accidents. It is graphite moderated and Helium cooled. The residual heat can be transferred out of the reactor core by inactive ways as conduction, convection, and thermal radiation during the accident. In such a way, a fuel temperature does not go beyond a limit at which major fission product release begins. In this thesis, the coupled neutronics and fluid mechanics code MGT-3D used for the steady state and time-dependent simulation of HTGRs, is enhanced and validated [2]. The fluid mechanics part is validated by SANA experiments in steady state cases as well as transient cases. The fuel temperature calculation is optimized by solving the heat conduction equation of the coated particles. It is applied in the steady state and transient simulation of PBMR, and the results are compared to the simulation with the old overheating model. New approaches to calculate the temperature profile of the fuel element of block-type HTGRs, and the calculation of the homogeneous conductivity of composite materials are introduced. With these new developments, MGT-3D is able to simulate block-type HTGRs as well. This extended MGT-3D is used to simulate a cuboid ceramic block heating experiment in the NACOK-II facility. The extended MGT-3D is also applied to LOFC and DLOFC simulation of GT-MHR. It is a fluid mechanics calculation with a given heat source. This calculation result of MGT-3D is verified with the calculation results of other codes. The design of the Japanese HTTR is introduced. The deterministic simulation of the LOFC experiment of HTTR is conducted with the Monte-Carlo code Serpent and MGT-3D, which is the LOFC Project organized by OECD/NEA [3]. With Serpent the burnup

Development of a feeding device for solid material; Kiinteaen materiaalin syoettoelaitteen kehittaeminen

Energy Technology Data Exchange (ETDEWEB)

Heinonen, O.; Tiihonen, J. [Imatran Voima Oy, Vantaa (Finland). R and D Section

1995-12-31

Feeding of solid fuel into high pressure is an essential part of the pressurized power plant processes. A pilot scale fuel feeder meeting the requirements of these processes has been designed and built by Imatran Voima Oy (IVO). The fuel feeder is capable of feeding both relatively dry and wet solid material into high pressure. The object of this project was to develop the pilot scale fuel feeder to commercial level. The project was financed by IVO and Bioenergia -research programme. The project included testing of the previously built pilot-feeder at real operating conditions using peat and wood biomass as feedstocks. The testing consisted of short term and long term runs, which provided information about the operation and durability of the feeder with different materials. The tests were carried out partly in IVO`s laboratory, and partly in Jyvaeskylae at the pressurized steam drying pilot plant owned by IVO and VTT. The pilot-feeder operated well and reliably during the feeding tests. The feeder was dissembled and the parts were inspected between and after the test periods. No sign of excessive wear of the parts was noticed. Based on the good experiences from the pilot scale testing a commercial feeder with the capacity of 50 m{sup 3}/h was designed

Development of a feeding device for solid material; Kiinteaen materiaalin syoettoelaitteen kehittaeminen

Energy Technology Data Exchange (ETDEWEB)

Heinonen, O.; Tiihonen, J [Imatran Voima Oy, Vantaa (Finland). R and D Section

1996-12-31

Feeding of solid fuel into high pressure is an essential part of the pressurized power plant processes. A pilot scale fuel feeder meeting the requirements of these processes has been designed and built by Imatran Voima Oy (IVO). The fuel feeder is capable of feeding both relatively dry and wet solid material into high pressure. The object of this project was to develop the pilot scale fuel feeder to commercial level. The project was financed by IVO and Bioenergia -research programme. The project included testing of the previously built pilot-feeder at real operating conditions using peat and wood biomass as feedstocks. The testing consisted of short term and long term runs, which provided information about the operation and durability of the feeder with different materials. The tests were carried out partly in IVO`s laboratory, and partly in Jyvaeskylae at the pressurized steam drying pilot plant owned by IVO and VTT. The pilot-feeder operated well and reliably during the feeding tests. The feeder was dissembled and the parts were inspected between and after the test periods. No sign of excessive wear of the parts was noticed. Based on the good experiences from the pilot scale testing a commercial feeder with the capacity of 50 m{sup 3}/h was designed

Development of the fast reactor group constant set JFS-3-J3.2R based on the JENDL-3.2

CERN Document Server

Chiba, G

2002-01-01

It is reported that the fast reactor group constant set JFS-3-J3.2 based on the newest evaluated nuclear data library JENDL3.2 has a serious error in the process of applying the weighting function. As the error affects greatly nuclear characteristics, and a corrected version of the reactor constant set, JFS-3-J3.2R, was developed, as well as lumped FP cross sections. The use of JFS-3-J3.2R improves the results of analyses especially on sample Doppler reactivity and reaction rate in the blanket region in comparison with those obtained using the JFS-3-J3.2.

Development of 3D CFD simulation method in nuclear reactor safety analysis

International Nuclear Information System (INIS)

Rosli Darmawan; Mariah Adam

2012-01-01

One of the most prevailing issues in the operation of nuclear reactor is the safety of the system. Worldwide publicity on a few nuclear accidents as well as the notorious Hiroshima and Nagasaki bombing have always brought about public fear on anything related to nuclear. Most findings on the nuclear reactor accidents are closely related to the reactor cooling system. Thus, the understanding of the behaviour of reactor cooling system is very important to ensure the development and improvement on safety can be continuously done. Throughout the development of nuclear reactor technology, investigation and analysis on reactor safety have gone through several phases. In the early days, analytical and experimental methods were employed. For the last three decades 1D system level codes were widely used. The continuous development of nuclear reactor technology has brought about more complex system and processes of nuclear reactor operation. More detailed dimensional simulation codes are needed to assess these new reactors. This paper discusses the development of 3D CFD usage in nuclear reactor safety analysis worldwide. A brief review on the usage of CFD at Malaysia's Reactor TRIGA PUSPATI is also presented. (author)

FFUSION research programme 1993-1998. Final report of the Finnish fusion research programme

Energy Technology Data Exchange (ETDEWEB)

Karttunen, S.; Heikkinen, J.; Korhonen, R. [VTT Energy, Espoo (Finland)] [and others

1998-12-31

This report summarizes the results of the Fusion Energy Research Programme, FFUSION, during the period 1993-1998. After the planning phase the programme started in 1994, and later in March 1995 the FFUSION Programme was integrated into the EU Fusion Programme and the Association Euratom-Tekes was established. Research areas in the FFUSION Programme are (1) fusion physics and plasma engineering, (2) fusion reactor materials and (3) remote handling systems. In all research areas industry is involved. Recently, a project on environmental aspects of fusion and other future energy systems started as a part of the socio-economic research (SERF) in the Euratom Fusion Programme. A crucial component of the FFUSION programme is the close collaboration between VTT Research Institutes, universities and Finnish industry. This collaboration has guaranteed dynamic and versatile research teams, which are large enough to tackle challenging research and development projects. Regarding industrial fusion R and D activities, the major step was the membership of Imatran Voima Oy in the EFET Consortium (European Fusion Engineering and Technology), which further strengthened the position of industry in the engineering design activities of ITER. The number of FFUSION research projects was 66. In addition, there were 32 industrial R and D projects. The total cost of the FFUSION Programme in 1993-1998 amounted to FIM 54 million in research at VTT and universities and an additional FIM 21 million for R and D in Finnish industry. The main part of the funding was provided by Tekes, 36%. Since 1995, yearly Euratom funding has exceeded 25%. The FFUSION research teams have played an active role in the European Programme, receiving excellent recognition from the European partners. Theoretical and computational fusion physics has been at a high scientific level and the group collaborates with the leading experimental laboratories in Europe. Fusion technology is focused on reactor materials, joining

FFUSION research programme 1993-1998. Final report of the Finnish fusion research programme

International Nuclear Information System (INIS)

Karttunen, S.; Heikkinen, J.; Korhonen, R.

1998-01-01

This report summarizes the results of the Fusion Energy Research Programme, FFUSION, during the period 1993-1998. After the planning phase the programme started in 1994, and later in March 1995 the FFUSION Programme was integrated into the EU Fusion Programme and the Association Euratom-Tekes was established. Research areas in the FFUSION Programme are (1) fusion physics and plasma engineering, (2) fusion reactor materials and (3) remote handling systems. In all research areas industry is involved. Recently, a project on environmental aspects of fusion and other future energy systems started as a part of the socio-economic research (SERF) in the Euratom Fusion Programme. A crucial component of the FFUSION programme is the close collaboration between VTT Research Institutes, universities and Finnish industry. This collaboration has guaranteed dynamic and versatile research teams, which are large enough to tackle challenging research and development projects. Regarding industrial fusion R and D activities, the major step was the membership of Imatran Voima Oy in the EFET Consortium (European Fusion Engineering and Technology), which further strengthened the position of industry in the engineering design activities of ITER. The number of FFUSION research projects was 66. In addition, there were 32 industrial R and D projects. The total cost of the FFUSION Programme in 1993-1998 amounted to FIM 54 million in research at VTT and universities and an additional FIM 21 million for R and D in Finnish industry. The main part of the funding was provided by Tekes, 36%. Since 1995, yearly Euratom funding has exceeded 25%. The FFUSION research teams have played an active role in the European Programme, receiving excellent recognition from the European partners. Theoretical and computational fusion physics has been at a high scientific level and the group collaborates with the leading experimental laboratories in Europe. Fusion technology is focused on reactor materials, joining

Large-signal, dynamic simulation of the slowpoke-3 nuclear heating reactor

International Nuclear Information System (INIS)

Tseng, C.M.; Lepp, R.M.

1983-07-01

A 2 MWt nuclear reactor, called SLOWPOKE-3, is being developed at the Chalk River Nuclear Laboratories (CRNL). This reactor, which is cooled by natural circulation, is designed to produce hot water for commercial space heating and perhaps generate some electricity in remote locations where the costs of alternate forms of energy are high. A large-signal, dynamic simulation of this reactor, without closed-loop control, was developed and implemented on a hybrid computer, using the basic equations of conservation of mass, energy and momentum. The natural circulation of downcomer flow in the pool was simulated using a special filter, capable of modelling various flow conditions. The simulation was then used to study the intermediate and long-term transient response of SLOWPOKE-3 to large disturbances, such as loss of heat sink, loss of regulation, daily load following, and overcooling of the reactor coolant. Results of the simulation show that none of these disturbances produce hazardous transients

RA reactor exploitation, task 3.08/01

International Nuclear Information System (INIS)

Zecevic, V.

1963-01-01

During 1963 the RA reactor was operated for 1852 hours at mean power of 5.7 MW (total power production was 10716 MWh). Reactor was used for irradiation according to the demand of 356 users, and 15 experiments. The reason for decreased operation in comparison with the previous year was repair of all the reactor equipment and decontamination of the heavy water system. This report contains detailed data about reactor power, reactivity changes and fuel burnup. Mean monthly usage of the reactor experimental channels as well as samples which were irradiated are part of this report

Reinforced fluropolymer nanocomposites with high-temperature superconducting Bi2Sr2CaCu2Oy

Science.gov (United States)

Jayasree, T. K.

2014-10-01

Bismuth Strontium Calcium Copper Oxide (Bi2Sr2CaCu2Oy)/Polyvinylidene fluoride (PVDF) nanocomposite was prepared and their thermal properties were analyzed. The composite consists of the polyvinylidene fluoride (PVDF) as an insulating polymer matrix, and homogenously distributed Bismuth strontium calcium copperoxide (2212) nanoparticles. SEM data shows flaky grains of the superconductor coated and linked by polymer. Differential scanning calorimetry (DSC) results indicated that the melting point was not affected significantly by the addition of BSCCO. However, the addition of superconducting ceramic resulted in an extra melting peak at a lower temperature (145°C). Thermogravimetric analysis of the samples shows that the onset decomposition temperature of the PVDF matrix was decreased by the addition of SC filler.

The visual illustration of complex process information during abnormal incidents

International Nuclear Information System (INIS)

Heimbuerger, H.; Kautto, A.; Norros, L.; Ranta, J.

1985-01-01

One of the proposed solutions to the man-process interface problem in nuclear power plants is the integration of a system in the control room that can provide the operator with a display of a minimum set of critical plant parameters defining the safety status of the plant. Such a system has been experimentally validated using the Loviisa training simulator during the fall of 1982. The project was a joint effort between Combustion Engineering Inc., the Halden Reactor Project, Imatran Voima Oy and VTT. Alarm systems are used in nuclear power plants to tell the control room operators that an unexpected change in the plant operation state has occurred. One difficulty in using the alarms for checking the actions of the operator is that the conventional way of realizing the alarm systems implies that several alarms are active also during normal operation. The coding and representation of alarm information will be discussed in the paper. An important trend in control room design is the move away from direct, concrete indication of process parameters towards use of more abstract/logical representation of information as a basis for plant supervision. Recent advances in computer graphics provide the possibility that, in the future, visual information will be utilized to make the essential dynamics of the process more intelligible. A set of criteria for use of visual information will be necessary. The paper discusses practical aspects for the realisation of such criteria in the context of nuclear power plant. The criteria of the decomposition of the process information concerning the sub-goals safety and availability and also the tentative results of the conceptualization of a PWR-process are discussed in the paper

Ethylbenzene dehydrogenation over binary FeOx–MeOy/Mg(Al)O catalysts derived from hydrotalcites

KAUST Repository

Balasamy, Rabindran J.; Khurshid, Alam; Al-Ali, Ali A S; Atanda, Luqman A.; Sagata, Kunimasa; Asamoto, Makiko; Yahiro, Hidenori; Nomura, Kiyoshi; Sano, Tsuneji; Takehira, Katsuomi; Al-Khattaf, Sulaiman S.

2010-01-01

A series of FeOx-MeOy/Mg(Al)O catalysts were prepared from hydrotalcite-like compounds as precursors and were tested in the ethylbenzene dehydrogenation to styrene in He atmosphere at 550 °C. The hydrotalcite-like precursors of the metal compositions of Mg3Fe 0.25Me0.25Al0.5 (Me = Cu, Zn, Cr, Mn, Fe, Co and Ni) were coprecipitated from the nitrates of metal components and calcined to mixed oxides at 550 °C. After the calcination, the mixed oxides showed high surface area of 150-200 m2 gcat -1, and were mainly composed of (MgMe)(Fe3+Al)O periclase in the bulk, whereas the surface was enriched by (MgMe)(Fe3+Al)2O 4 pinel. Among the Me species tested, Co2+ was the most effective, followed by Ni2+. Co2+ addition increased the activity of original FeOx/Mg(Al)O catalyst, whereas Ni2+ increased the activity at the beginning of reaction, but deactivated the catalyst during the reaction. The other metals formed isolated MeOx species in the catalyst, resulting in a decrease in the activity compared to the original FeOx/Mg(Al)O catalyst. The active Fe species exists as metastable Fe3+ on the FeOx/Mg(Al)O catalyst. By the addition of Co2+, the reduction-oxidation between Fe3+ and Fe2+ was facilitated and, moreover, the active Fe3+ species was stabilized. It is likely that the dehydrogenation proceeds on the active Fe3+ species via its reduction-oxidation assisted by Co 2+. © 2010 Elsevier B.V.

Ethylbenzene dehydrogenation over binary FeOx–MeOy/Mg(Al)O catalysts derived from hydrotalcites

KAUST Repository

Balasamy, Rabindran J.

2010-12-20

A series of FeOx-MeOy/Mg(Al)O catalysts were prepared from hydrotalcite-like compounds as precursors and were tested in the ethylbenzene dehydrogenation to styrene in He atmosphere at 550 °C. The hydrotalcite-like precursors of the metal compositions of Mg3Fe 0.25Me0.25Al0.5 (Me = Cu, Zn, Cr, Mn, Fe, Co and Ni) were coprecipitated from the nitrates of metal components and calcined to mixed oxides at 550 °C. After the calcination, the mixed oxides showed high surface area of 150-200 m2 gcat -1, and were mainly composed of (MgMe)(Fe3+Al)O periclase in the bulk, whereas the surface was enriched by (MgMe)(Fe3+Al)2O 4 pinel. Among the Me species tested, Co2+ was the most effective, followed by Ni2+. Co2+ addition increased the activity of original FeOx/Mg(Al)O catalyst, whereas Ni2+ increased the activity at the beginning of reaction, but deactivated the catalyst during the reaction. The other metals formed isolated MeOx species in the catalyst, resulting in a decrease in the activity compared to the original FeOx/Mg(Al)O catalyst. The active Fe species exists as metastable Fe3+ on the FeOx/Mg(Al)O catalyst. By the addition of Co2+, the reduction-oxidation between Fe3+ and Fe2+ was facilitated and, moreover, the active Fe3+ species was stabilized. It is likely that the dehydrogenation proceeds on the active Fe3+ species via its reduction-oxidation assisted by Co 2+. © 2010 Elsevier B.V.

Characteristics of D-{sup 3}He fueled frc reactor: ARTEMIS-L

Energy Technology Data Exchange (ETDEWEB)

Momota, H.; Motojima, O.; Okamoto, M.; Sudo, S.; Tomita, Y.; Yamaguchi, S.; Iiyoshi, A.; Onozuka, M.; Ohnishi, M.; Uenosono, C.

1993-11-01

The paper introduces briefly the scenario and discuss the attractive characteristics of D-3He fueled commercial fusion reactor ARTEMIS-L. By using favorable characteristics of a field-reversed configuration, the fusion plasma of ARTEMIS-L is compact and its beta-value is extremely high. One find consequently a possibility of constructing an economical fusion power power plant on this prospect. The life of the structural materials is sound during the full reactor life (30 years) and the safety of the reactor is intrinsic to D-3He fuels. The amount of disposed materials is rather small and the level of these intruder dose is so low that the plant appears to be acceptable in view of the environment. (author).

Fusion reactor control study. Volume 3. Tandem mirror reactors. Final report

International Nuclear Information System (INIS)

Chang, F.R.; DeCanio, F.; Fisher, J.L.; Madden, P.A.

1982-03-01

A study of the control requirements of the Tandem Mirror Reactor concept is reported. The study describes the development of a control simulator that is based upon a spatially averaged physics code of the reactor concept. The simulator portrays the evolution of the plasma through the complete reactor operating cycle; it includes models of the control and measurement system, thus allowing the exploration of various strategies for reactor control. Startup, shutdown, and control during the quasi-steady-state power producing phase were explored. Configurations are described which use a variety of control effectors including modulation of the refueling rate, beam current, and electron cyclotron resonance heating. Multivariable design techniques were used to design the control laws and compensators for the feedback controllers and presume the practical measurement of only a subset of the plasma and machine variables. Performance of the various controllers is explored using the nonlinear control simulator. Derivative control strategies using new or developed sensors and effectors appropriate to a power reactor environment are postulated, based upon the results of the control configurations tested. Research and development requirements for these controls are delineated

Studies on the liquid fluoride thorium reactor: Comparative neutronics analysis of MCNP6 code with SRAC95 reactor analysis code based on FUJI-U3-(0)

Energy Technology Data Exchange (ETDEWEB)

Jaradat, S.Q., E-mail: sqjxv3@mst.edu; Alajo, A.B., E-mail: alajoa@mst.edu

2017-04-01

Highlights: • The verification for FUJI-U3-(0)—a molten salt reactor—was performed. • The MCNP6 was used to study the reactor physics characteristics for FUJI-U3 type. • The results from the MCNP6 were comparable with the ones obtained from literature. - Abstract: The verification for FUJI-U3-(0)—a molten salt reactor—was performed. The reactor used LiF-BeF2-ThF4-UF4 as the mixed liquid fuel salt, and the core was graphite moderated. The MCNP6 code was used to study the reactor physics characteristics for the FUJI-U3-(0) reactor. Results for reactor physics characteristic of the FUJI-U3-(0) exist in literature, which were used as reference. The reference results were obtained using SRAC95 (a reactor analysis code) coupled with ORIGEN2 (a depletion code). Some modifications were made in the reconstruction of the FUJI-U3-(0) reactor in MCNP due to unavailability of more detailed description of the reactor core. The assumptions resulted in two representative models of the reactor. The results from the MCNP6 models were compared with the reference results obtained from literature. The results were comparable with each other, but with some notable differences. The differences are because of the approximations that were done on the SRAC95 model of the FUJI-U3 to simplify the simulation. Based on the results, it is concluded that MCNP6 code predicts well the overall simulation of neutronics analysis to the previous simulation works using SRAC95 code.

Matkustussäännön toteuttaminen ja kehittäminen - Case Norilsk Nickel Harjavalta Oy

OpenAIRE

Sihvonen, Aino

2015-01-01

Tämä opinnäytetyö tehtiin toimeksiantona Norilsk Nickel Harjavalta Oy:lle. Opinnäytetyön tarkoituksena oli tutkia toimeksiantajan matkustussäännön vahvuuksia ja puutteita. Tutkimustehtävänä oli selvittää, miten yhtiön henkilöstö kokee nykyisen matkustussäännön sekä mistä johtuu, että matkustusasioissa ilmenee erilaisia käytänteitä. Tutkimuksen tavoitteena oli luoda kehitysehdotuksia yhtiön matkustussääntöön. Matkustussäännön kehittämisellä pyritään kustannustehokkuuden ja ympäristönäkökulmien...

Systems analysis of the CANDU 3 Reactor

Energy Technology Data Exchange (ETDEWEB)

Wolfgong, J.R.; Linn, M.A.; Wright, A.L.; Olszewski, M.; Fontana, M.H. [Oak Ridge National Lab., TN (United States)

1993-07-01

This report presents the results of a systems failure analysis study of the CANDU 3 reactor design; the study was performed for the US Nuclear Regulatory Commission. As part of the study a review of the CANDU 3 design documentation was performed, a plant assessment methodology was developed, representative plant initiating events were identified for detailed analysis, and a plant assessment was performed. The results of the plant assessment included classification of the CANDU 3 event sequences that were analyzed, determination of CANDU 3 systems that are ``significant to safety,`` and identification of key operator actions for the analyzed events.

Synthesis and Microstructure Properties of (Bi,Pb2Sr2Ca1Cu2Oy Ceramic Superconductor

Directory of Open Access Journals (Sweden)

nurmalita .

2015-11-01

Full Text Available Properties of (Bi, Pb2Sr2Ca1Cu2Oy ceramic superconductors were prepared by the melt textured growth methods in order to investigate the effects of the slow cooling time on the microstructur.  Phase analyses of the samples by X-ray diffraction (XRD has been carried out to assess the effects of the slow cooling time. From XRD analyses, the addition to the sample of  the slow cooling time degrades formation of the high-Tc Bi-2212 phase. The possible reasons for the observed degradation in the microstructure properties due to the slow cooling time addition were discussed.

CRM-tietojärjestelmän käyttöopas osana yrityksen asiakkuudenhallintaa : case: Fysioline Oy

OpenAIRE

Ilonen, Mikko

2012-01-01

Tämän opinnäytetyön aiheena on CRM-tietojärjestelmän käyttöopas osana yrityksen asiakkuudenhallintaa. Opinnäytetyön teoriaosuus käsittelee asiakkuudenhallintaa, CRM-tietojärjestelmiä ja CRM:n käyttöönottoa. Työn empiriaosuus muodostuu CRM-tietojärjestelmän käyttöoppaasta. CRM-tietojärjestelmän käyttöopas tehdään kohdeyritys Fysioline Oy:lle. Opinnäytetyön teoriaosuudessa selvitetään vastausta tutkimusongelmaan, miksi yrityksen tulee kiinnittää huomiota asiakkuudenhallintaan ja mitä hyötyä...

ISO 14001 -standardin mukaisen ympäristöjärjestelmän luominen : case: Plastep Oy

OpenAIRE

Harju, Siiri

2016-01-01

Opinnäytetyön tavoitteena oli luoda kattavan ympäristökartoituksen sisältävä ISO 14001 -standardin mukainen ympäristöjärjestelmä Plastep Oy:lle. Lisäksi tarkoituksena oli suunnitella yrityksen ympäristönsuojelun tilaa mahdollisimman luotettavasti kuvaava ympäristömittari sekä tarkastella muovin ruiskuvaluyritykseen kohdistuvia tulevaisuuden näkymiä ympäristönsuojelun näkökulmasta. Opinnäytetyössä esitellään ISO 14001 -ympäristöjärjestelmä sekä kyseisen standardin sisältä...

Reactor core for LMFBR type reactors

International Nuclear Information System (INIS)

Masumi, Ryoji; Azekura, Kazuo; Kurihara, Kunitoshi; Bando, Masaru; Watari, Yoshio.

1987-01-01

Purpose: To reduce the power distribution fluctuations and obtain flat and stable power distribution throughout the operation period in an LMFBR type reactor. Constitution: In the inner reactor core region and the outer reactor core region surrounding the same, the thickness of the inner region is made smaller than the axial height of the reactor core region and the radial width thereof is made smaller than that of the reactor core region and the volume thereof is made to 30 - 50 % for the reactor core region. Further, the amount of the fuel material per unit volume in the inner region is made to 70 - 90 % of that in the outer region. The difference in the neutron infinite multiplication factor between the inner region and the outer region is substantially constant irrespective of the burnup degree and the power distribution fluctuation can be reduced to about 2/3, by which the effect of thermal striping to the reactor core upper mechanisms can be moderated. Further, the maximum linear power during operation can be reduced by 3 %, by which the thermal margin in the reactor core is increased and the reactor core fuels can be saved by 3 %. (Kamimura, M.)

Ignition access in a D-3He helical reactor

International Nuclear Information System (INIS)

Mitarai, Osamu

2003-01-01

Ignition access in a D- 3 He helical reactor is studied based on 0-dimensional particle and power balance equations for deuterium, tritium, helium-3, alpha ash, proton ash, electron density and temperature. The calculations are based on the following experimental facts observed in LHD. (author)

Fusion blankets for catalyzed D--D and D--He3 reactors

International Nuclear Information System (INIS)

Fillo, J.A.; Powell, J.R.

1977-01-01

Blanket designs are presented for catalyzed D-D (Cat-D) and D-He 3 fusion reactors. Because of relatively low neutron wall loads and the flexibility due to non-tritium breeding, blankets potentially should operate for reactor life-times of approximately 30 years. Unscheduled replacement of failed blanket modules should be relatively rapid, due to very low residual activity, by operators working either through access ports in the shield (option 1) or directly in the plasma chamber (option 2). Cat-D blanket designs are presented for high (approximately 30%) and low (approximately 12%) β noncircular Tokamak reactors. The blankets are thick graphite screens, operating at high temperature to anneal radiation damage; the deposited neutron and gamma energy is thermally radiated along internal cavities and conducted to a bank of internal SiC coolant tubes (approximately 4 cm. ID) containing high pressure helium. In the D-He 3 Tokamak reactor design, the blanket consists of multiple layers (e.g., three) of thin (approximately 10 cm.) high strength aluminum (e.g., SAP), modular plates, cooled by organic terphynyl coolant

Fusion blankets for catalyzed D--D and D--3He reactors

International Nuclear Information System (INIS)

Fillo, J.A.; Powell, J.R.

1977-01-01

Blanket designs are presented for catalyzed D-D (Cat-D) and D-He 3 fusion reactors. Because of relatively low neutron wall loads and the flexibility due to non-tritium breeding, blankets potentially should operate for reactor life-times of approximately 30 years. Unscheduled replacement of failed blanket modules should be relatively rapid, due to very low residual activity, by operators working either through access ports in the shield (option 1) or directly in the plasma chamber (option 2). Cat-D blanket designs are presented for high (approximately 30%) and low (approximately 12%) β non-circular Tokamak reactors. The blankets are thick graphite screens, operating at high temperature to anneal radiation damage; the deposited neutron and gamma energy is thermally radiated along internal cavities and conducted to a bank of internal SiC coolant tubes (approximately 4 cm. ID) containing high pressure helium. In the D-He 3 Tokamak reactor design, the blanket consists of multiple layers (e.g., three) of thin (approximately 10 cm.) high strength aluminum (e.g., SAP), modular plates, cooled by organic terphenyl coolant

RELAP5-3D code validation of RBMK-1500 reactor reactivity measurement transients

International Nuclear Information System (INIS)

Kaliatka, Algirdas; Bubelis, Evaldas; Uspuras, Eugenijus

2003-01-01

This paper deals with the modeling of transients taking place during the measurements of the void and fast power reactivity coefficients performed at Ignalina NPP. The simulation of these transients was performed using RELAP5-3D code model of RBMK-1500 reactor. At the Ignalina NPP void and fast power reactivity coefficients are measured on a regular basis and, based on the total reactor power, reactivity, control and protection system control rods positions and the main circulation circuit parameter changes during the experiments, the actual values of these reactivity coefficients are determined. Following the simulation of the two above mentioned transients with RELAP5-3D code, a conclusion was made that the obtained calculation results demonstrate reasonable agreement with Ignalina NPP measured data. Behaviors of the separate MCC thermal-hydraulic parameters as well as physical processes are predicted reasonably well to the real processes, occurring in the primary circuit of RBMK-1500 reactor. The calculated reactivity and the total reactor core power behavior in time are also in reasonable agreement with the measured plant data. Despite of the small differences, RELAP5-3D code predicts reactivity and the total reactor core power behavior during the transients in a reasonable manner. Reasonable agreement of the measured and the calculated total reactor power change in time demonstrates the correct modeling of the neutronic processes taking place in RBMK-1500 reactor core

Recurring Events in the Finnish Nuclear Power Plants

International Nuclear Information System (INIS)

Suksi, Seija; Olander, Ronnie; Tiippana, Petteri

2003-01-01

An analysis and evaluation of event investigation methods applied by the Radiation and Nuclear Safety Authority (STUK), and the two Finnish nuclear power plant operators Teollisuuden Voima Oy (TVO) and Fortum Power and Heat Oy (Fortum) was carried out by the Technical Research Centre (VIT) on an assignment from STUK. The study aimed at providing a broad overview of the whole organisational framework to support event investigation practices at the regulatory body and at the utilities. The study was part of the IAEA Co-ordinated Research Programme (CRP) on 'Investigation of Methodologies for Incident Analysis'. The main objective of the research was to evaluate the adequacy and reliability of event investigation analysis methods and practices in the Finnish nuclear power industry and based on the results to further develop them. In general, the direct causes of identified events could be detected and eliminated, but more emphasis should be given to the prevention of recurrence of events and identification of common causes and latent failures. The study showed that the evaluated organisations had rather comprehensive incident analysis arrangements. The study also showed that more focus and prioritisation are needed. Deficiencies were identified mostly in the areas of recording, assessment and classification of new events, use of existing operating experience data, utilisation of information technology tools, and allocation of work and resources. Also the indicators or measures for the effectiveness of event investigation and operating experience feedback were missing. All organisations should maintain adequate resources in this area. The researchers suggested a more effective operating experience feedback loop. Especially more attention should be paid to root cause analysis of significant events, tasks and activities where the initial errors have occurred, and weaknesses of defensive barriers. It was also recommended that implementing periodic operational experience

GRIMH3: A new reactor calculation code at Savannah River Site

International Nuclear Information System (INIS)

Le, T.T.; Pevey, R.E.

1993-01-01

The GRIMHX reactor code currently in use at the Savannah River Site (SRS) was written at a time when computer processing speed and memory storage were very limited. Recently, a new reactor code (GRIMH3) was written to take advantage of the hardware improvements (vectorization and higher memory capacities) as well as the range of available computers at SRS (workstations and supercomputers). The GRIMH3 code computes the solution of the static multigroup neutron diffusion equation in one-, two-, and three-dimensional hexagonal geometry. Either direct or adjoint solutions can be computed for k eff searches, buckling searches, external neutron sources, power flattening searches, or power normalization factor calculations with 1, 6, 24, 54, or 96 points per hex. The GRIMHX reactor code currently in use at the Savannah River Site (SRS) was written at a time when computer processing speed and memory storage were very limited. Recently, a new reactor code (GRIMH3) was written to take advantage of the hardware improvements (vectorization and higher memory capacities) as well as the range of available computers at SRS (workstations and supercomputers). The GRIMH3 code computes the solution of the static multigroup neutron diffusion equation in one-, two-, and three-dimensional hexagonal geometry. Either direct or adjoint solutions can be computed for k eff searches, buckling searches, external neutron sources, power flattening searches, or power normalization factor calculations with 1, 6, 24, 54, or 96 points per hex

Valokuvauksen voima päihdekuntoutuksessa

OpenAIRE

Nieminen, Noora

2016-01-01

Tämän projektin tarkoituksena oli toteuttaa entisille päihteidenkäyttäjille päihdekuntoutusta voimauttava valokuva - menetelmällä. Projektin tavoitteena oli, että projektin lopuksi osallistujat kokevat onnistumisen tunteita toiminnastaan ja ymmärtävät luovien menetelmien merkityksen päihdekuntoutuksen tukena. Opinnäytetyöntekijän henkilökohtaisina tavoitteina oli kehittää ohjaustaitoja, oppia voimauttava valo-kuvamenetelmästä, oppia toteuttamaan projektiluontoinen opinnäytetyö ja nähdä t...

Verkkokauppa osaksi B2B-yrityksen sähköistä liike-toimintaa : case: Capcons Technology Oy

OpenAIRE

Kaitarinne, Tomi; Hannula, Tuomo

2010-01-01

Tämän opinnäytetyön tarkoituksena on kuvata monipuolisesti B2B-verkkokaupan perustamisprosessi nykyaikaisin menetelmin ja uusimpien vaatimusten mukaisesti. Opinnäytetyön case-yrityksenä toimii Capcons Technology Oy, jonka verkko-kaupan perustamisprosessissa on tarkoitus soveltaa työssä käsiteltyjä asioita. Tavoitteena on, että yritys pystyy hyödyntämään opinnäytetyötä kattavasti verkko-kaupan perustamisen eri vaiheissa. Lisäksi tätä työtä voidaan hyödyntää myös muita B2B-verkkokauppoja perust...

Irradiated graphite studies prior to decommissioning of G1, G2 and G3 reactors

International Nuclear Information System (INIS)

Bonal, J.P.; Vistoli, J.Ph.; Combes, C.

2005-01-01

G1 (46 MW th ), G2 (250 MW th ) and G3 (250 MW th ) are the first French plutonium production reactors owned by CEA (Commissariat a l'Energie Atomique). They started to be operated in 1956 (G1), 1959 (G2) and 1960 (G3); their final shutdown occurred in 1968, 1980 and 1984 respectively. Each reactor used about 1200 tons of graphite as moderator, moreover in G2 and G3, a 95 tons graphite wall is used to shield the rear side concrete from neutron irradiation. G1 is an air cooled reactor operated at a graphite temperature ranging from 30 C to 230 C; G2 and G3 are CO 2 cooled reactors and during operation the graphite temperature is higher (140 C to 400 C). These reactors are now partly decommissioned, but the graphite stacks are still inside the reactors. The graphite core radioactivity has decreased enough so that a full decommissioning stage may be considered. Conceming this decommissioning, the studies reported here are: (i) stored energy in graphite, (ii) graphite radioactivity measurements, (iii) leaching of radionuclide ( 14 C, 36 Cl, 63 Ni, 60 Co, 3 H) from graphite, (iv) chlorine diffusion through graphite. (authors)

Part I: A Comparative Thermal Aging Study on the Regenerability of Rh/Al2O3 and Rh/CexOy-ZrO2 as Model Catalysts for Automotive Three Way Catalysts

Directory of Open Access Journals (Sweden)

Qinghe Zheng

2015-10-01

Full Text Available The rhodium (Rh component in automotive three way catalysts (TWC experiences severe thermal deactivation during fuel shutoff, an engine mode (e.g., at downhill coasting used for enhancing fuel economy. In a subsequent switch to a slightly fuel rich condition, in situ catalyst regeneration is accomplished by reduction with H2 generated through steam reforming catalyzed by Rh0 sites. The present work reports the effects of the two processes on the activity and properties of 0.5% Rh/Al2O3 and 0.5% Rh/CexOy-ZrO2 (CZO as model catalysts for Rh-TWC. A very brief introduction of three way catalysts and system considerations is also given. During simulated fuel shutoff, catalyst deactivation is accelerated with increasing aging temperature from 800 °C to 1050 °C. Rh on a CZO support experiences less deactivation and faster regeneration than Rh on Al2O3. Catalyst characterization techniques including BET surface area, CO chemisorption, TPR, and XPS measurements were applied to examine the roles of metal-support interactions in each catalyst system. For Rh/Al2O3, strong metal-support interactions with the formation of stable rhodium aluminate (Rh(AlO2y complex dominates in fuel shutoff, leading to more difficult catalyst regeneration. For Rh/CZO, Rh sites were partially oxidized to Rh2O3 and were relatively easy to be reduced to active Rh0 during regeneration.

Threats and benefits updated information on local opinions regarding the spent nuclear fuel repository in Finland - 16128

International Nuclear Information System (INIS)

Kojo, Matti; Kari, Mika; Litmanen, Tapio

2009-01-01

The aim of the paper is to provide updated information on local opinion regarding the siting of a spent nuclear fuel repository in Finland. The main question is how the residents of the municipality perceive the threats and benefits of the repository. In accordance with the Decision in Principle by the Council of State passed in 2000, the Olkiluoto area in Municipality of Eurajoki was chosen as the location for the repository to accommodate spent nuclear fuel produced in Finland. Updated information on local opinions is needed as the siting process is approaching the next phase, the application for a construction license by 2012. The nuclear waste management company Posiva, owned by the utilities Teollisuuden Voima and Fortum Power and Heat, has also applied for a new Decision in Principle (DiP) for expansion of the repository. The data provided in this paper is based on a survey carried out in June 2008. The respondents were selected from the residents of the municipality of Eurajoki and the neighbouring municipalities using stratified random sampling (N=3000). The response rate of the survey was 20% (N=606). The paper is part of a joint research project between the University of Jyvaeskylae and the University of Tampere. The research project 'Follow-up research regarding socio-economic effects and communication of final disposal facility of spent nuclear fuel in Eurajoki and its neighbouring municipalities' is funded by the Finnish Research Programme on Nuclear Waste Management (KYT2010). (authors)

The 4th surveillance testing for Kori unit 3 reactor vessel materials

Energy Technology Data Exchange (ETDEWEB)

Chang, Kee Ok; Kim, Byoung Chul; Lee, Sam Lai; Choi, Kwun Jae; Gong, Un Sik; Chang, Jong Hwa; Joo, Yong Sun; Ahn, Sang Bok; Hong, Joon Hwa [Korea Atomic Energy Research Institute, Taejeon (Korea)

2000-10-01

Surveillance testing for reactor vessel materials is performed in order to evaluate the irradiation embrittlement due to neutrons during operation and set the condition of safe operation of nuclear reactor. The 4th surveillance testing was performed completely by Korea Atomic Energy Research Institute at Taejeon after the capsule was transported from Kori site including its removal from reactor. Fast neutron fluences for capsules were calculated and various testing including mechanical and chemistry analysis were performed in order to evaluate the integrity of Kori unit 3 reactor vessel during the operation until life time. The evaluation results are as follows; Fast neutron fluences for capsules U, V, X and W are 4.983E+18, 1.641E+19, 3.158E+19, and 4.469E+19n/cm{sup 2}, respectively. The bias factor, the ratio of calculation/measurement, was 0.840 for the 1st through 4th testing and the calculational uncertainty, 12% satisfied the requirement of USNRC Reg.Guide DG-1053, 20%. The best estimated neutron fluence for reactor vessel inside surface was 1.362E+19n/cm{sup 2} based on the end of 12th fuel cycle and it was predicted that the fluences of vessel inside surface at 32, 40, 48 and 56EFPY would reach 3.481E+19, 4.209E+19, 5.144E+19 and 5.974E+19n/cm{sup 2} based on the current calculation. The result through this analysis for Kori unit 3 showed that there would be no problem for the pressurized thermal shock(PTS) during the operation until design life. 48 refs., 35 figs., 41 tabs. (Author)

Comparison of 2D and 3D Neutron Transport Analyses on Yonggwang Unit 3 Reactor

International Nuclear Information System (INIS)

Maeng, Aoung Jae; Kim, Byoung Chul; Lim, Mi Joung; Kim, Kyung Sik; Jeon, Young Kyou; Yoo, Choon Sung

2012-01-01

10 CFR Part 50 Appendix H requires periodical surveillance program in the reactor vessel (RV) belt line region of light water nuclear power plant to check vessel integrity resulting from the exposure to neutron irradiation and thermal environment. Exact exposure analysis of the neutron fluence based on right modeling and simulations is the most important in the evaluation. Traditional 2 dimensional (D) and 1D synthesis methodologies have been widely applied to evaluate the fast neutron (E > 1.0 MeV) fluence exposure to RV. However, 2D and 1D methodologies have not provided accurate fast neutron fluence evaluation at elevations far above or below the active core region. RAPTOR-M3G (RApid Parallel Transport Of Radiation - Multiple 3D Geometries) program for 3D geometries calculation was therefore developed both by Westinghouse Electronic Company, USA and Korea Reactor Integrity Surveillance Technology (KRIST) for the analysis of In-Vessel Surveillance Test and Ex-Vessel Neutron Dosimetry (EVND). Especially EVND which is installed at active core height between biological shielding material and concrete also evaluates axial neutron fluence by placing three dosimetries each at Top, Middle and Bottom part of the angle representing maximum neutron fluence. The EVND programs have been applied to the Korea Nuclear Plants. The objective of this study is therefore to compare the 3D and the 2D Neutron Transport Calculations and Analyses on the Yonggwang unit 3 Reactor as an example

Safety in the ARIES-III D-3He tokamak reactor design

International Nuclear Information System (INIS)

Herring, J.S.; Dolan, T.J.

1992-01-01

This paper reports on the ARIES-III reactor study, an extensive examination of the viability of a D- 3 He-fueled commercial tokamak powder reactor. Because neutrons are produced only through side reactions (D+D- 3 HE+N; and D+D-T+p followed by D+T- 4 He+n), the reactor has the significant advantages of reduced activation of the first wall and shield, low afterheat and Class A or C low level waste disposal. Since no tritium is required for operation, no lithium-containing breeding blanket is necessary. A ferritic steel shield behind the first wall protects the magnets from gamma and neutron heating and from radiation damage. The authors explored the potential for isotopically tailoring the 4 mm tungsten layer on the divertor in order to reduce the offsite doses should a tungsten aerosol be released from the reactor after an accident. The authors also modeled a loss-of-cooling accident (LOCA) in which the organic coolant was burning in order to estimate the amount of radionuclides released from the first wall. Because the maximum temperature is low, degree C, release fractions are small. The authors analyzed the disposition of the 20 g/day of tritium that is produced by D-D reactions and removed by the vacuum pumps

The Flamanville 3 EPR reactor; Le reacteur EPR Flamanville 3

Energy Technology Data Exchange (ETDEWEB)

NONE

2007-07-01

On April 10. 2007, the french government authorized EDF to create on the site of Flamanville ( La Manche) a nuclear base installation containing a pressurized water EPR type reactor. This nuclear reactor, conceived by AREVA NP and EDF, is the first copy of a generation susceptible to replace later, at least partly, the French nuclear reactors at present in operation.Within the framework of its mission of technical support of the Authority of Nuclear Safety ( A.S.N.), the I.R.S.N. widely contributed successively: to define the general objectives of safety assigned to this new generation of pressurized water nuclear reactors; to analyze the options of safety proposed by EDF for the EPR project; To deepen, upstream to the authorization of creation, the evaluation of the step of safety and the measures of conception retained by EDF that have to allow to respect the objectives of safety which were notified to it. (N.C.)

3D-mallien muokkaus 3D-tulostamista varten CAD-ohjelmilla

OpenAIRE

Lehtimäki, Jarmo

2013-01-01

Insinöörityössäni käsitellään 3D-mallien tulostamista ja erityisesti 3D-mallien mallintamista niin, että kappaleiden valmistaminen 3D-tulostimella onnistuisi mahdollisimman hyvin. Työ tehtiin Prohoc Oy:lle, joka sijaitsee Vaasassa. 3D-tulostuspalveluun tuli jatkuvasti 3D-malleja, joiden tulostuksessa oli ongelmia. Työssäni tutkin näiden ongelmien syntyä ja tein ohjeita eri 3D-mallinnusohjelmille, joiden tarkoituksena on auttaa tekemään helpommin tulostettavia 3D-malleja. Työhön kuului myös et...

Reversed field pinch reactor study 3

International Nuclear Information System (INIS)

Hollis, A.A.; Mitchell, J.T.D.

1977-12-01

This report, the third of a series on the Reversed Field Pinch Reactor, describes a preliminary concept of the engineering design and layout of this pulsed toroidal reactor, which uses the stable plasma behaviour first observed in ZETA. The basic parameters of the 600 MW(e) reactor are taken from a companion study by Hancox and Spears. The plasma volume is 1.75m minor radius and 16m major radius surrounded by a 1.8m blanket-shield region - with the blanket divided into 14 removable segments for servicing. The magnetic confinement system consists of 28 toroidal field coils situated just outside the blanket and inside the poloidal and vertical field coils and all coils have normal copper conductors. The requirement to incorporate a conducting shell at the front of the blanket to provide a short-time plasma stability has a marked effect on the design. It sets the size of the blanket segment and the scale of the servicing operations, limits the breeding gain and complicates the blanket cooling and its integration with the heat engine. An extensive study will be required to confirm the overall reactor potential of the concept. (author)

Safety analysis of loss of flow transients in a typical research reactor by RELAP5/MOD3.3

International Nuclear Information System (INIS)

Di Maro, B.; Pierro, F.; Adorni, M.; Bousbia Salah, A.; D'Auria, F.

2003-01-01

The main aim of the following study is to assess the RELAP5/MOD3.3 code capability in simulating transient dynamic behaviour in nuclear research reactors. For this purpose typical loss of flow transient in a representative MTR (Metal Test Reactor) fuel type Research Reactor is considered. The transient herein considered is a sudden pump trip followed by the opening of a safety valve in order to allow passive decay heat removal by natural convection. During such transient the coolant flow decay, originally downward, leads to a flow reversal and the cooling process of the core passes from forced, mixed and finally to natural circulation. This fact makes it suitable for evaluating the new features of RELAP5 to simulate such specific operating conditions. The instantaneous reactor power is derived through the point kinetic calculation, both protected and unprotected cases are considered (with and without Scram). The results obtained from this analysis were also compared with previous results obtained by old version RELAP5/MOD2 code. (author)

The World's Reactors no. 70 - Forsmark 3, BWR-75

International Nuclear Information System (INIS)

Anon.

1976-01-01

A large pull-out wall chart is presented showing a coloured cut-away diagram of the Forsmark 3 station. It is accompanied by 2 small sketches one showing the layout of station buildings and the other the inside of the reactor vessel. Parameters are listed. (U.K.)

Gas Reactor International Cooperative program. Pebble bed reactor plant: screening evaluation. Volume 3. Appendix A. Equipment list

International Nuclear Information System (INIS)

1979-11-01

This report consists of three volumes which describe the design concepts and screening evaluation for a 3000 MW(t) Pebble Bed Reactor Multiplex Plant (PBR-MX). The Multiplex plant produces both electricity and transportable chemical energy via the thermochemical pipeline (TCP). The evaluation was limited to a direct cycle plant which has the steam generators and steam reformers in the primary circuit. Volume 1 reports the overall plant and reactor system and was prepared by the General Electric Company. Core scoping studies were performed which evaluated the effects of annular and cylindrical core configurations, radial blanket zones, burnup, and ball heavy metal loadings. The reactor system, including the PCRV, was investigated for both the annular and cylindrical core configurations. Volume 3 is an Appendix containing the equipment list for the plant and was also prepared by United Engineers and Constructors, Inc. It tabulates the major components of the plant and describes each in terms of quantity, type, orientation, etc., to provide a basis for cost estimation

Development of telerobotic systems for reactor decommissioning, (3)

International Nuclear Information System (INIS)

Usui, Hozumi; Fujii, Yoshio; Shinohara, Yoshikuni

1991-01-01

This paper describes the telerobotic system for reactor decommissioning in the scope of engineering demonstration of dismantling radioactive reactor internals of an experimental boiling water power reactor JPDR. The total system consists of a telerobotic manipulator system equipped with a multi-functional amphibious slave manipulator with a load capacity of 25 daN, a chain-driven transport system, and a computer-assisted monitoring and control system. Preceding to the application of the telerobotic system to actual dismantling operation, a mockup test was performed of dismantling the simulated reactor internals of actual-size by the method of underwater plasma arc cutting in order to study the performance of the telerobotic system in a realistic environment. The system was then successfully applied to dismantling the actual reactor internals according to the JPDR decommissioning program. (author)

On severe accident hydrogen behaviour in Loviisa

International Nuclear Information System (INIS)

Okkonen, T.

1996-02-01

This study is related to the hydrogen management strategy of the Loviisa ice-condenser containments. A synthetic survey is conducted of the various parts of the subject by using compact 'back-of-the-envelope' analysis methods. The analysed cases are consistent with the principal hydrogen management approaches proposed by the utility Imatran Voima Oy (IVO). The study begins by introduction of the Loviisa plant features and various severe accident types. Hydrogen generation characteristics are analysed mainly for the core degradation phase, but the hydrogen sources from molten fuel-coolant interactions and reflooding of a degraded core are discussed, as well. The hydrogen generation and release rates are compared with the overall gas convection and mixing conditions in order to estimate hydrogen concentrations in the containment. The natural convection currents are examined also from the scaling point of view, concerning the scaled-down VICTORIA tests of IVO. Finally, the potential for large deflagration loadings or local detonations is examined for the Loviisa containments. The study is concluded by preliminary subjective judgments about the most critical factors of the Loviisa hydrogen problematics and about any issues that may require additional confirmative research. (orig.) (47 refs., 4 figs., 24 tabs.)

On severe accident hydrogen behaviour in Loviisa

Energy Technology Data Exchange (ETDEWEB)

Okkonen, T. [OTO-Consulting Ay, Helsinki (Finland)

1996-02-01

This study is related to the hydrogen management strategy of the Loviisa ice-condenser containments. A synthetic survey is conducted of the various parts of the subject by using compact `back-of-the-envelope` analysis methods. The analysed cases are consistent with the principal hydrogen management approaches proposed by the utility Imatran Voima Oy (IVO). The study begins by introduction of the Loviisa plant features and various severe accident types. Hydrogen generation characteristics are analysed mainly for the core degradation phase, but the hydrogen sources from molten fuel-coolant interactions and reflooding of a degraded core are discussed, as well. The hydrogen generation and release rates are compared with the overall gas convection and mixing conditions in order to estimate hydrogen concentrations in the containment. The natural convection currents are examined also from the scaling point of view, concerning the scaled-down VICTORIA tests of IVO. Finally, the potential for large deflagration loadings or local detonations is examined for the Loviisa containments. The study is concluded by preliminary subjective judgments about the most critical factors of the Loviisa hydrogen problematics and about any issues that may require additional confirmative research. (orig.) (47 refs., 4 figs., 24 tabs.).

A Small-Animal Irradiation Facility for Neutron Capture Therapy Research at the RA-3 Research Reactor

Energy Technology Data Exchange (ETDEWEB)

Emiliano Pozzi; David W. Nigg; Marcelo Miller; Silvia I. Thorp; Amanda E. Schwint; Elisa M. Heber; Veronica A. Trivillin; Leandro Zarza; Guillermo Estryk

2007-11-01

The National Atomic Energy Commission of Argentina (CNEA) has constructed a thermal neutron source for use in Boron Neutron Capture Therapy (BNCT) applications at the RA-3 research reactor facility located in Buenos Aires. The Idaho National Laboratory (INL) and CNEA have jointly conducted some initial neutronic characterization measurements for one particular configuration of this source. The RA-3 reactor (Figure 1) is an open pool type reactor, with 20% enriched uranium plate-type fuel and light water coolant. A graphite thermal column is situated on one side of the reactor as shown. A tunnel penetrating the graphite structure enables the insertion of samples while the reactor is in normal operation. Samples up to 14 cm height and 15 cm width are accommodated.

CRM-järjestelmän suunnittelu ja käyttöönotto myynnin strategiseksi työkaluksi : case: Stala Oy

OpenAIRE

Pirkkalainen, Markus

2013-01-01

Tämän opinnäytetyön aiheena on asiakkuudenhallinta- eli Customer Relationship Management(CRM)-järjestelmän suunnittelu ja käyttöönotto myynnin strategiseksi työkaluksi. Opinnäytetyön teoriaosuudessa tutkitaan asiakkuudenhallintaa, CRM-järjestelmiä ja CRM-järjestelmien suunnitteluun sekä käyttöönottoon liittyviä haasteita. Työn empiirinen osuus koostuu CRM-järjestelmän käyttöoppaasta kohdeyritys Stala Oy:n myyntitiimille. Tutkimusongelma vastaa kysymykseen, miksi yrityksen tulisi panostaa a...

Station Blackout Analysis for a 3-Loop Westinghouse PWR Reactor Using Trace

International Nuclear Information System (INIS)

El-Sahlamy, N.M.

2017-01-01

One of the main concerns in the area of severe accidents in nuclear reactors is that of station blackout (SBO). The loss of offsite electrical power concurrent with the unavailability of the onsite emergency alternating current (AC) power system can result in loss of decay heat removal capability, leading to a potential core damage which may lead to undesirable consequences to the public and the environment. To cope with an SBO, nuclear reactors are provided with protection systems that automatically shut down the reactor, and with safety systems to remove the core residual heat. This paper provides a best estimate assessment of the SBO scenario in a 3-loop Westinghouse PWR reactor. The evaluation is performed using TRACE, a best estimate computer code for thermal-hydraulic calculations. Two sets of scenarios for SBO analyses are discussed in the current work. The first scenario is the short term SBO where it is assumed that in addition to the loss of AC power, there is no DC power; i.e., no batteries are available. In the second scenario, a long term SBO is considered. For this scenario, DC batteries are available for four hours. The aim of the current SBO analyses for the 3-loop pressurized water reactor presented in this paper is to focus on the effect of the availability of a DC power source to delay the time to core uncovers and heatup

The qualification of U3O8 as research reactor fuel

International Nuclear Information System (INIS)

Krull, W.

1983-01-01

This report summarizes the today knowledge of the qualification status of U 3 O 8 as low enriched ( 3 O 8 is so far qualified to start testing of ten (10) fuel elements with an U-density of 3.1 g U/cc in the FRG-2 research reactor. (orig.) [de

Operation experience with the 3 MW TRIGA Mark-II research reactor of Bangladesh

International Nuclear Information System (INIS)

Islam, M.S.; Haque, M.M.; Salam, M.A.; Rahman, M.M.; Khandokar, M.R.I.; Sardar, M.A.; Saha, P.K.; Haque, A.; Malek Sonar, M.A.; Uddin, M.M.; Hossain, S.M.S.; Zulquarnain, M.A.

2004-01-01

The 3 MW TRIGA Mark-II research reactor of Bangladesh Atomic Energy Commission (BAEC) has been operating since September 14, 1986. The reactor is used for radioisotope production ( 131 I, 99m Tc, 46 Sc), various R and D activities and manpower training. The reactor has been operated successfully since it's commissioning with the exception of a few reportable incidents. Of these, the decay tank leakage incident of 1997 is considered to be the most significant one. As a result of this incident, reactor operation at full power under forced-convection mode remained suspended for about 4 years. During that time, the reactor was operated at a power level of 250 kW so as to carry out experiments that require lower neutron flux. This was made possible by establishing a temporary by pass connection across the decay tank using local technology. The other incident was the contamination of the Dry Central Thimble (DCT) that took place in March 2002 when a pyrex vial containing 50 g of TeO 2 powder got melted inside the DCT. The vial was melted due to high heat generation on its surface while the reactor was operated for 8 hours at 3 MW for trial production of Iodine-131 ( 131 I). A Wet Central Thimble (WCT) was used to replace the damaged DCT in June 2002 such that the reactor operation could be resumed. The WCT was again replaced by a new DCT in June 2003 such that radioisotope production could be continued. A total of 873 irradiation requests (IRs) have been catered for different reactor uses. Out of these, 114 IRs were for radioisotope (RI) production and 759 IRs for different experiments. The total amount of RI produced stands at about 2100 GBq. The total amount of burn-up-fuel is about 6158 MWh. Efforts are on to undertake an ADP project so as to convert the analog console and I and C system of the reactor into digital one. The paper summarizes the reactor operation experiences focusing on troubleshooting, rectification, modification, RI production, various R and D

Computational Analysis of Nuclear Safety Parameters of 3 MW TRIGA Mark-II Research Reactor Based on Evaluated Nuclear Data Libraries JENDL-3.3 and ENDF/B-VII.0

International Nuclear Information System (INIS)

Khan, Jahirul Haque

2013-01-01

The objective of this study is to explain the main nuclear safety parameters of 3 MW TRIGA Mark-II Research Reactor at AERE, Savar, Dhaka, Bangladesh from the viewpoint of reactor safety and also reactor operator. The most important nuclear reactor physics safety parameters are power distribution, power peaking factors, shutdown margin, control rod worth, excess reactivity and fuel temperature reactivity coefficient. These parameters are calculated using the chain of the computer codes the SRAC-PIJ for cell calculation based on neutron transport theory and the SRAC-CITATION for core calculation based on neutron diffusion equation. To achieve this objective the TRIGA model is developed by the 3-D diffusion code SRAC-CITATION based on the group constants that come from the collision probability transport code SRAC-PIJ. In this study the evaluated nuclear data libraries JENDL-3.3 and ENDF/B-VII.0 are used. The calculated most important reactor physics parameters are compared to the safety analysis report (SAR) values as well as earlier published MCNP results (numerically benchmark). It was found that the calculated results show a good agreement between the said libraries. Besides, in most cases the calculated results reveal a reasonable agreement with the SAR values (by General Atomic) as well as the MCNP results. In addition, this analysis can be used as the inputs for thermal-hydraulic calculations of the TRIGA fresh core in the steady state and pulse mode operation. Because of power peaking factors, power distributions and temperature reactivity coefficients are the most important reactor safety parameters for normal operation and transient safety analysis in research as well as in power reactors. They form the basis for technical specifications and limitations for reactor operation such as loading pattern limitations for pulse operation (in TRIGA). Therefore, this analysis will be very important to develop the nuclear safety parameters data of 3 MW TRIGA Mark

Gas-cooled reactor thermal-hydraulics using CAST3M and CRONOS2 codes

International Nuclear Information System (INIS)

Studer, E.; Coulon, N.; Stietel, A.; Damian, F.; Golfier, H.; Raepsaet, X.

2003-01-01

The CEA R and D program on advanced Gas Cooled Reactors (GCR) relies on different concepts: modular High Temperature Reactor (HTR), its evolution dedicated to hydrogen production (Very High Temperature Reactor) and Gas Cooled Fast Reactors (GCFR). Some key safety questions are related to decay heat removal during potential accident. This is strongly connected to passive natural convection (including gas injection of Helium, CO 2 , Nitrogen or Argon) or forced convection using active safety systems (gas blowers, heat exchangers). To support this effort, thermal-hydraulics computer codes will be necessary tools to design, enhance the performance and ensure a high safety level of the different reactors. Accurate and efficient modeling of heat transfer by conduction, convection or thermal radiation as well as energy storage are necessary requirements to obtain a high level of confidence in the thermal-hydraulic simulations. To achieve that goal a thorough validation process has to ve conducted. CEA's CAST3M code dedicated to GCR thermal-hydraulics has been validated against different test cases: academic interaction between natural convection and thermal radiation, small scale in-house THERCE experiments and large scale High Temperature Test Reactor benchmarks such as HTTR-VC benchmark. Coupling with neutronics is also an important modeling aspect for the determination of neutronic parameters such as neutronic coefficient (Doppler, moderator,...), critical position of control rods...CEA's CAST3M and CRONOS2 computer codes allow this coupling and a first example of coupled thermal-hydraulics/neutronics calculations has been performed. Comparison with experimental data will be the next step with High Temperature Test Reactor experimental results at nominal power

3-DB, 3-D Multigroup Diffusion, X-Y-Z, R-Theta-Z, Triangular-Z Geometry, Fast Reactor Burnup

International Nuclear Information System (INIS)

Hardie, R.W.; Little, W.W. Jr.; Mroz, W.

1974-01-01

1 - Description of problem or function: 3DB is a three-dimensional (x-y-z, r-theta-z, triangular-z) multigroup diffusion code for use in detailed fast-reactor criticality and burnup analysis. The code can be used to - (a) compute k eff and perform criticality searches on time absorption, reactor composition, and reactor dimensions by means of either a flux or an adjoint model, (b) compute material burnup using a flexible material shuffling scheme, and (c) compute flux distributions for an arbitrary extraneous source. 2 - Method of solution: Eigenvalues are computed by standard source- iteration techniques. Group re-balancing and successive over-relaxation with line inversion are used to accelerate convergence. Adjoint solutions are obtained by inverting the input data and redefining the source terms. Material burnup is by reactor zone. The burnup rate is determined by the zone and energy-averaged cross sections which are recomputed after each time-step. The isotopic chains, which can contain any number of isotopes are formed by the user. The code does not contain built- in or internal chains. 3 - Restrictions on the complexity of the problem: Since variable dimensioning is employed, no simple bounds can be stated

A robot-automated work site for repair of the Chinon A3 reactor

International Nuclear Information System (INIS)

Raynal, A.

1987-01-01

In 1982, following degradation due to corrosion of low-carbon steel by carbon dioxide gas, the utility undertook to repair some of the support structures at Chinon A3. This involved consolidation and reinforcing thermocouples and gas monitor pipeworks supports. A welding process was selected and the use of robots became indispensable because of the large number of components to be replaced (200 per outage). Two robots, supplied with tool heads and replacement components from outside the reactor were used. The robots and their servers were coordinated by a central computer and monitored by a closed circuit television system. Each repair operation was performed after ''training'' on a full-scale mockup of the top of the reactor reconstructed from telemetry of the real reactor dimensions. Since becoming operational in June 1986, the robots have accumulated over 20 000 hours of operation and seventy parts have been welded to the reactor. A 3D CAD system has been adapted to simulate the robots and analyse long trajectories in order to reduce robot learning time [fr

EL3 reactor description and safety analysis report; Pile EL3, rapport descriptif et de surete

Energy Technology Data Exchange (ETDEWEB)

NONE

1969-02-01

The EL-3 reactor is an experimental pile. Heterogenous type reactor, water moderated and cooled it uses slightly enriched uranium oxide as fuel (4.5 percent) distributed in vertical cells that constitute the core (the maximum number of cells is 99). It is conceived to function at a maximal thermal power of 20 MW. It supplies a maximum thermal neutron flux of 10{sup 14} neutrons/cm{sup 2}/sec. It has several experimental devices. The EL-3 reactor is surrounded by auxiliary circuits of fluids, in a sealed containment, slightly depressed. The primary heavy water coolant circuit is completely included in this containment. Its cooling is made by the intermediary of a light water secondary circuit by atmospheric refrigerants. The ventilation circuits of the sealed containment and the reactor block do not release air outside, under nornal functioning, by a particularly studied chimney only after filtering and eventually dilution. The eventual contamination of the light water or air by active products is permanently monitored to allow the reactor shutdown and avoid the release in atmosphere of dangerous products. The EL-3 reactor, laying down in may 1955, has diverged in july 1957, made its first ascending in power in december 1957 and reached its complete power in april 1958. The positioning of actual fuel (snow crystal) was made during summer 1964. Reactor with an experimental aim, it is used for theoretical and technological studies by material irradiation in the experimental channels and the core cells, with possibilities to constitute independent loops (relative to the cooling fluids). Thirty vertical channels are devoted to the fabrication of artificial radioelements. [French] La pile EL-3 est une pile experimentale. Du type heterogene, moderee et refroidie a l'eau lourde elle utilise comme combustible de l'oxygene d'uranium faiblement enrichi (4,5 p.cent) reparti en cellules verticales qui constituent le coeur (le nombre maximal de cellules est de, 99). Elle est

EL3 reactor description and safety analysis report; Pile EL3, rapport descriptif et de surete

Energy Technology Data Exchange (ETDEWEB)

NONE

1969-02-01

The EL-3 reactor is an experimental pile. Heterogenous type reactor, water moderated and cooled it uses slightly enriched uranium oxide as fuel (4.5 percent) distributed in vertical cells that constitute the core (the maximum number of cells is 99). It is conceived to function at a maximal thermal power of 20 MW. It supplies a maximum thermal neutron flux of 10{sup 14} neutrons/cm{sup 2}/sec. It has several experimental devices. The EL-3 reactor is surrounded by auxiliary circuits of fluids, in a sealed containment, slightly depressed. The primary heavy water coolant circuit is completely included in this containment. Its cooling is made by the intermediary of a light water secondary circuit by atmospheric refrigerants. The ventilation circuits of the sealed containment and the reactor block do not release air outside, under nornal functioning, by a particularly studied chimney only after filtering and eventually dilution. The eventual contamination of the light water or air by active products is permanently monitored to allow the reactor shutdown and avoid the release in atmosphere of dangerous products. The EL-3 reactor, laying down in may 1955, has diverged in july 1957, made its first ascending in power in december 1957 and reached its complete power in april 1958. The positioning of actual fuel (snow crystal) was made during summer 1964. Reactor with an experimental aim, it is used for theoretical and technological studies by material irradiation in the experimental channels and the core cells, with possibilities to constitute independent loops (relative to the cooling fluids). Thirty vertical channels are devoted to the fabrication of artificial radioelements. [French] La pile EL-3 est une pile experimentale. Du type heterogene, moderee et refroidie a l'eau lourde elle utilise comme combustible de l'oxygene d'uranium faiblement enrichi (4,5 p.cent) reparti en cellules verticales qui constituent le coeur (le nombre maximal de cellules est de, 99

Uranium-fuel thermal reactor benchmark testing of CENDL-3

International Nuclear Information System (INIS)

Liu Ping

2001-01-01

CENDL-3, the new version of China Evaluated Nuclear Data Library are being processed, and distributed for thermal reactor benchmark analysis recently. The processing was carried out using the NJOY nuclear data processing system. The calculations and analyses of uranium-fuel thermal assemblies TRX-1,2, BAPL-1,2,3, ZEEP-1,2,3 were done with lattice code WIMSD5A. The results were compared with the experimental results, the results of the '1986'WIMS library and the results based on ENDF/B-VI. (author)

Investigations related to a one-piece removal of the reactor block in the frame of the JRR-3 reconstruction program

International Nuclear Information System (INIS)

Onishi, N.; Kanenari, A.; Futamura, Y.; Sakurai, H.; Suzuki, S.; Nagase, T.; Iwatani, A.; Otsubo, F.

1987-01-01

In the Japan Atomic Energy Research Institute (JAERI), an outdated research reactor (Japan Research Reactor No.3; JRR-3) was removed to a storage facility between October 14th and November 7th, 1986. The removal of the 2250-ton reactor block (10 x 10 x 10 m) was performed as a part of a program to replace the JRR-3's core (10-MW thermal) with an upgraded research reactor core. The heavy water and fuel elements were taken out from the JRR-3 before removal work began. The reactor block was raised about 3.7 meters, using a 12-cubic meter steel frame and a center-hole jack system. The reactor block was then transported horizontally about 34 meters on steel rails, using four 100-ton jacks, to a storage facility. Finally, the reactor block was lowered 14 meters into the storage facility. After the reactor block was stored, a new 20-MW thermal, light-water moderated and cooled JRR-3 core will be built, with criticality targeted for 1989

Decommissioning of the BR3 pressurized-water reactor

International Nuclear Information System (INIS)

Massaut, V.

1996-01-01

The dismantling and the decommissioning of nuclear installations at the end of their life-cycle is a new challenge to the nuclear industry. Different techniques and procedures for the dismantling of a nuclear power plant on an existing installation, the BR-3 pressurized-water reactor, are described. The scientific programme, objectives, achievements in this research area at the Belgian Nuclear Research Centre SCK-CEN for 1995 are summarized

Application of RELAP5-3D code for thermal analysis of the ADS reactor core

International Nuclear Information System (INIS)

Fernandes, Gustavo Henrique Nazareno

2018-01-01

Nuclear power is essential to supply global energy demand. Therefore, in order to use nuclear fuel more efficiently, more efficient nuclear reactors technologies researches have been intensified, such as hybrid systems, composed of particle accelerators coupled into nuclear reactors. In order to add knowledge to such studies, an innovative reactor design was considered where the RELAP5-3D thermal-hydraulic analysis code was used to perform a thermal analysis of the core, either in stationary operation or in situations transitory. The addition of new kind of coolants, such as, liquid salts, among them Flibe, lead, lead-bismuth, sodium, lithium-bismuth and lithium-lead was an important advance in this version of the code, making possible to do the thermal simulation of reactors that use these types of coolants. The reactor, object of study in this work, is an innovative reactor, due to its ability to operate in association with an Accelerator Driven System (ADS), considered a predecessor system of the next generation of nuclear reactors (GEN IV). The reactor selected was the MYRRHA (Multi-purpose Hybrid Research Reactor for High tech Applications) due to the availability of data to perform the simulation. In the modeling of the reactor with the code RELAP5-3D, the core was simulated using nodules with 1, 7, 15 and 51 thermohydraulic channels and eutectic lead-bismuth (LBE) as coolant. The parameters, such as, pressure, mass flow and coolant and heat structure temperature were analyzed. In addition, the thermal behavior of the core was evaluated by varying the type of coolant (sodium) in substitution for the LBE of the original design using the model with 7 thermohydraulic channels. The results of the steady-state calculations were compared with data from the literature and the proposed models were verified certifying the ability of the RELAP5-3D code to simulate this innovative reactor. After this step, it was analysed cases of transients with loss of coolant flow

Extension of the Repository Under Excavation. The Opinions of the Local Residents in the Municipality of Eurajoki

International Nuclear Information System (INIS)

Kojo, Matti; Kari, Mika; Litmanen, Tapio

2009-12-01

The aim of the paper is to provide updated information on the opinions of residents of Eurajoki municipality concerning the disposal facility for spent nuclear fuel (SNF) in Finland. The SNF facility project is approaching the construction licence phase by 2012. At the same time as it prepares for the next phase the nuclear waste company Posiva Oy is planning to extend the disposal capacity of the facility up to 12000 tU due to the revival of nuclear energy policy in Finland. It is not only the owners of Posiva, namely Teollisuuden Voima (TVO) and Fortum Power and Heat (FPH), who need more disposal capacity. A brand new nuclear operator Fennovoima is also interested in disposing of its SNF into Posiva's facility. The possible extension of the SNF facility needs to be approved by the council of Eurajoki municipality. According to the Nuclear Energy Act the council has the right of veto. The original application of Posiva was approved by the council in 2000. According to an opinion poll 59% of the residents of the Eurajoki municipality would have accepted the siting in 1999 if the facility were found safe by the investigations of the authorities. The Olkiluoto site in the municipality of Eurajoki was chosen to be the site for further investigations in accordance with the DiP of 2000 by the Council of State. The DiP was ratified by Parliament in May 2001. Thus the local residents have lived the post site selection phase for nearly one decade. During this phase Posiva, for example, has started excavations for the Underground Rock Characterization Facility Onkalo into the bedrock of Olkiluoto. The residents have also experienced years of risk communication after the site selection of 2001. However, two recent surveys indicate that the local attitudes are showing increasing reservations rather than confidence regarding the disposal of SNF in Olkiluoto. Furthermore, data show that over 50% of the residents perceived at least an explicit threat to the health, safety and

Proceedings of 2. Yugoslav symposium on reactor physics, Part 3, Herceg Novi (Yugoslavia), 27-29 Sep 1966

International Nuclear Information System (INIS)

1966-01-01

This Volume 3 of the Proceedings of 2. Yugoslav symposium on reactor physics includes three papers describing the following: model for spatial synthesis of automated control system of the GCR type reactor; model for analysis of hydrodynamic processes at the BHWR type reactors; mathematical model for safety analysis of heavy water power reactor

Reactor core of FBR type reactor

International Nuclear Information System (INIS)

Hayashi, Hideyuki; Ichimiya, Masakazu.

1994-01-01

A reactor core is a homogeneous reactor core divided into two regions of an inner reactor core region at the center and an outer reactor core region surrounding the outside of the inner reactor core region. In this case, the inner reactor core region has a lower plutonium enrichment degree and less amount of neutron leakage in the radial direction, and the outer reactor core region has higher plutonium enrichment degree and greater amount of neutron leakage in the radial direction. Moderator materials containing hydrogen are added only to the inner reactor core fuels in the inner reactor core region. Pins loaded with the fuels with addition of the moderator materials are inserted at a ratio of from 3 to 10% of the total number of the fuel pins. The moderator materials containing hydrogen comprise zirconium hydride, titanium hydride, or calcium hydride. With such a constitution, fluctuation of the power distribution in the radial direction along with burning is suppressed. In addition, an absolute value of the Doppler coefficient can be increased, and a temperature coefficient of coolants can be reduced. (I.N.)

Hybrid Reactor Simulation and 3-D Information Display of BWR Out-of-Phase Oscillation

International Nuclear Information System (INIS)

Edwards, Robert; Huang, Zhengyu

2001-01-01

The real-time hybrid reactor simulation (HRS) capability of the Penn State TRIGA reactor has been expanded for boiling water reactor (BWR) out-of-phase behavior. During BWR out-of-phase oscillation half of the core can significantly oscillate out of phase with the other half, while the average power reported by the neutronic instrumentation may show a much lower amplitude for the oscillations. A description of the new HRS is given; three computers are employed to handle all the computations required, including real-time data processing and graph generation. BWR out-of-phase oscillation was successfully simulated. By adjusting the reactivity feedback gains from boiling channels to the TRIGA reactor and to the first harmonic mode power simulation, limit cycle can be generated with both reactor power and the simulated first harmonic power. A 3-D display of spatial power distributions of fundamental mode, first harmonic, and total powers over the reactor cross section is shown

Simulation of channel blockage for the IEA-R1 research reactor using RELAP/MOD 3

International Nuclear Information System (INIS)

Oliveira, Eduardo C.F. de; Castrillo, Lazara Silveira

2015-01-01

Research reactors have great importance in the area of nuclear technology, such as radioisotope production, research in nuclear physics, development of new technologies and staff training for reactor operation. The IEA-R1 is a Brazilian research reactor type pool, located at the IPEN (Instituto de Pesquisas Energeticas e Nucleares). In this work is simulated with computer code RELAP5 / MOD 3.3.2 gamma, the effect caused by partial and complete blockage of a channel in MTR fuel element of the IEA-R1 core, in order to analyzed the thermal hydraulic parameters on adjacent channels. (author)

Simulation of channel blockage for the IEA-R1 research reactor using RELAP/MOD 3

Energy Technology Data Exchange (ETDEWEB)

Oliveira, Eduardo C.F. de; Castrillo, Lazara Silveira, E-mail: ecfoliveira@hotmail.com, E-mail: lazara.castrillo@upe.br [Universidade de Pernambuco (UPE), Recife, PE (Brazil). Escola Politecnica de Pernambuco

2015-07-01

Research reactors have great importance in the area of nuclear technology, such as radioisotope production, research in nuclear physics, development of new technologies and staff training for reactor operation. The IEA-R1 is a Brazilian research reactor type pool, located at the IPEN (Instituto de Pesquisas Energeticas e Nucleares). In this work is simulated with computer code RELAP5 / MOD 3.3.2 gamma, the effect caused by partial and complete blockage of a channel in MTR fuel element of the IEA-R1 core, in order to analyzed the thermal hydraulic parameters on adjacent channels. (author)

Advances in Reactor physics, mathematics and computation. Volume 3

Energy Technology Data Exchange (ETDEWEB)

1987-01-01

These proceedings of the international topical meeting on advances in reactor physics, mathematics and computation, volume 3, are divided into sessions bearing on: - poster sessions on benchmark and codes: 35 conferences - review of status of assembly spectrum codes: 9 conferences - Numerical methods in fluid mechanics and thermal hydraulics: 16 conferences - stochastic transport and methods: 7 conferences.

Simulation of a reactor FBR with hexagonal-Z geometry using the code PARCS 3.1; Simulacion de un reactor FBR con geometria hexagonal-Z usando el codigo PARCS 3.1

Energy Technology Data Exchange (ETDEWEB)

Reyes F, M. C.; Del Valle G, E. [IPN, Escuela Superior de Fisica y Matematicas, Av. Instituto Politecnico Nacional s/n, U.P. Adolfo Lopez Mateos, Edificio 9, Col. San Pedro Zacatenco, 07738 Mexico D. F. (Mexico); Filio L, C., E-mail: rf.melisa@gmail.com [Comision Nacional de Seguridad Nuclear y Salvaguardias, Dr. Jose Ma. Barragan No. 779, Col. Narvarte, 03020 Mexico D. F. (Mexico)

2013-10-15

The nuclear reactor core type FBR (Fast Breeder Reactor) was modeled in three dimensions of hexagonal-Z geometry using the code PARCS (Purdue Advanced Reactor Core Simulator) version 3.1 developed by Purdue University researchers. To carry out the modeling of the mentioned reactor was taken the corresponding information to one of the described benchmarks in the document NEACRP-L-330 (3-D Neutron Transport Benchmarks, 1991); fundamentally the corresponding to the geometric data and the cross sections. Being a quick reactor of breeding, known as the Knk-II, for which are considered 4 energy groups without dispersions up. The reactor core is formed by prismatic elements of hexagonal transversal cut where part of them only corresponds to nuclear fuel assemblies. This has four reflector rings and 6 identical control elements that together with the active part of the core is configured with 8 different types of elements.With the extracted information of the mentioned document the entrance file was prepared for PARCS 3.1 only considering a sixth part of the core due to the symmetry that presents their configuration. The NEACRP-L-330 shows a wide range of results reported by those who collaborated in its elaboration using different solution techniques that go from the Monte Carlo method to the approaches S{sub 2} and P{sub 1}. Of all the results were selected those obtained with the code HEXNOD, to which were carried out a comparison of the effective multiplication factor, being smaller differences to the 300 pcm, for three different scenarios: a) with the control bars extracted totally, b) with the semi-inserted control bars and c) with the control bars inserted completely and two different axial meshes, a thick mesh with 14 slices and another fine with 38, that which implies that the results can be considered very similar among if same. Radial maps and axial profiles are included, as much of the power as of the neutrons flow. (Author)

Air toxics emissions from an IGCC process

Energy Technology Data Exchange (ETDEWEB)

Mojtahedi, W.; Norrbacka, P. [Enviropower Inc., Espoo (Finland); Hinderson, A. [Vattenfall (Sweden); Rosenberg, R.; Zilliacus, R.; Kurkela, E.; Nieminen, M. [VTT Energy, Espoo (Finland); Hoffren, H. [IVO International Oy, Vantaa (Finland)

1996-12-01

The so-called simplified coal gasification combined cycle process, incorporating air gasification and hot gas cleanup, promises high power generation efficiency in an environmentally acceptable manner. Increasingly more stringent environmental regulations have focused attention on the emissions of not only SO{sub 2} and NO{sub x} but also on the so-called air toxics which include a number of toxic trace elements. As result of recent amendments to the United States Clean Air Act, IGCC emissions of eleven trace elements: antimony, arsenic, beryllium, cadmium, chromium, cobalt, lead, manganese, mercury, nickel, selenium - as well as the radionuclides uranium and thorium may be regulated. Similarly, air missions standards in Europe include a limit of 0.05 mg Nm{sup 3} for mercury and cadmium and 1.0 3/Nm{sup 3} for other class I trace elements. A suitable sampling/measuring system has been developed in this project (in cooperation with Imatran Voima Oy, Electric Power Research Institute (EPRI) and Radian Cooperation) which will be used in the pressurized gasification tests. This will enable an accurate measurement of the volatilized trace element species, at high temperature and pressure, which may be found in the vapour phase. Models are being developed that can be used to determine not only the chemical equilibrium composition of gaseous, liquid and solid phases, but also possible interactions of the gaseous species with aerosol particles and surfaces, These should be used to more accurately assess the impact of the toxic trace metals emitted from the simplified IGCC system

Preparations for the shipment of RA-3 reactor irradiated fuel

International Nuclear Information System (INIS)

Goldschmidt, Adrian; Novara, Oscar; Lafuente, Jose

2002-01-01

During the last quarter of 2000, in the Radioactive Waste Management Area of the Argentine National Commission of Atomic Energy (CNEA), located at Ezeiza Atomic Center (CAE), activities associated to the shipment of 207 MTR spent fuels containing high enrichment uranium were carried out within the Foreign Research Reactor/Domestic Research Reactor Receipt Program launched by the US Department of Energy (DOE). The MTR spent fuel shipped to Savannah River Site (SRS) was fabricated in Argentina with 90% enriched uranium of US origin and it was utilized in the operation of the research and radioisotope production reactor RA-3 from 1968 until 1987. After a cooling period at the reactor, the spent fuel was transferred to the Central Storage Facility (CSF) located in the waste management area of CAE for interim storage. The spent fuel (SF) inventory consisted of 166 standard assemblies (SA) and 41 control assemblies (CA). Basically, the activities performed were the fuel conditioning operations inside the storage facility (remote transference of the assemblies to the operation pool, fuel cropping, fuel re-identification, loading in transport baskets, etc.) conducted by CNEA. The loading of the filled baskets in the transport casks (NAC-LWT) by means of intermediate transfer systems and loaded casks final preparations were conducted by NAC personnel (DOE's contractor) with the support of CNEA personnel. (author)

Design and implementation of the control system for the new console of TRIGA-3-Salazar Reactor

International Nuclear Information System (INIS)

Gonzalez M, J.L.

1994-01-01

TRIGA-3-Salazar Reactor was set in operation in 1968 and the aging of its components has cause the increasing in the maintenance. In the presence of this, it becomes necessary to replace the reactor console using new technologies, considering the incorporation of a personal computer. The aim of this work is the design and construction of the equipment interfaces as well as the digital computer program for the automation and control of the TRIGA-3-Salazar Reactor by means of a personal computer. (Author)

On the major DYN3D developments for fast reactor design and transient analysis

International Nuclear Information System (INIS)

Merk, B.; Kliem, S.

2013-01-01

Due to the French project ASTRID, the European CP-ESFR project, and the MYRRHA/FASTEF project, the research work on fast reactors has got a new push in Europe. Additionally to this European projects a strong project is growing in Russia based on the lead cooled fast reactor design BREST. Following this trend, the Institute of Resource Ecology at the Helmholtz-Zentrum Dresden-Rossendorf has decided to start several projects dedicated to fast reactor technology, among them the extension of the well validated LWR core simulator DYN3D. The new developments, first validation results, and the next strategic steps for the adaption of the code for the improved simulation of fast reactor cores are presented. (orig.)

On the major DYN3D developments for fast reactor design and transient analysis

Energy Technology Data Exchange (ETDEWEB)

Merk, B.; Kliem, S. [Helmholtz-Zentrum Dresden-Rossendorf e.V., Dresden (Germany). Reactor Safety Div.

2013-07-01

Due to the French project ASTRID, the European CP-ESFR project, and the MYRRHA/FASTEF project, the research work on fast reactors has got a new push in Europe. Additionally to this European projects a strong project is growing in Russia based on the lead cooled fast reactor design BREST. Following this trend, the Institute of Resource Ecology at the Helmholtz-Zentrum Dresden-Rossendorf has decided to start several projects dedicated to fast reactor technology, among them the extension of the well validated LWR core simulator DYN3D. The new developments, first validation results, and the next strategic steps for the adaption of the code for the improved simulation of fast reactor cores are presented. (orig.)

Calculation of low-energy reactor neutrino spectra reactor for reactor neutrino experiments

Energy Technology Data Exchange (ETDEWEB)

Riyana, Eka Sapta; Suda, Shoya; Ishibashi, Kenji; Matsuura, Hideaki [Dept. of Applied Quantum Physics and Nuclear Engineering, Kyushu University, Kyushu (Japan); Katakura, Junichi [Dept. of Nuclear System Safety Engineering, Nagaoka University of Technology, Nagaoka (Japan)

2016-06-15

Nuclear reactors produce a great number of antielectron neutrinos mainly from beta-decay chains of fission products. Such neutrinos have energies mostly in MeV range. We are interested in neutrinos in a region of keV, since they may take part in special weak interactions. We calculate reactor antineutrino spectra especially in the low energy region. In this work we present neutrino spectrum from a typical pressurized water reactor (PWR) reactor core. To calculate neutrino spectra, we need information about all generated nuclides that emit neutrinos. They are mainly fission fragments, reaction products and trans-uranium nuclides that undergo negative beta decay. Information in relation to trans-uranium nuclide compositions and its evolution in time (burn-up process) were provided by a reactor code MVP-BURN. We used typical PWR parameter input for MVP-BURN code and assumed the reactor to be operated continuously for 1 year (12 months) in a steady thermal power (3.4 GWth). The PWR has three fuel compositions of 2.0, 3.5 and 4.1 wt% {sup 235}U contents. For preliminary calculation we adopted a standard burn-up chain model provided by MVP-BURN. The chain model treated 21 heavy nuclides and 50 fission products. The MVB-BURN code utilized JENDL 3.3 as nuclear data library. We confirm that the antielectron neutrino flux in the low energy region increases with burn-up of nuclear fuel. The antielectron-neutrino spectrum in low energy region is influenced by beta emitter nuclides with low Q value in beta decay (e.g. {sup 241}Pu) which is influenced by burp-up level: Low energy antielectron-neutrino spectra or emission rates increase when beta emitters with low Q value in beta decay accumulate. Our result shows the flux of low energy reactor neutrinos increases with burn-up of nuclear fuel.

RA reactor exploitation, task 3.08/01; Zadatak 3.08/01 - Eksploatacija reaktora RA

Energy Technology Data Exchange (ETDEWEB)

Zecevic, V [Institute of Nuclear Sciences Boris Kidric, Reaktor RA, Vinca, Beograd (Serbia and Montenegro)

1963-12-15

During 1963 the RA reactor was operated for 1852 hours at mean power of 5.7 MW (total power production was 10716 MWh). Reactor was used for irradiation according to the demand of 356 users, and 15 experiments. The reason for decreased operation in comparison with the previous year was repair of all the reactor equipment and decontamination of the heavy water system. This report contains detailed data about reactor power, reactivity changes and fuel burnup. Mean monthly usage of the reactor experimental channels as well as samples which were irradiated are part of this report.

Operating reactors licensing actions summary. Vol. 3, No. 6

International Nuclear Information System (INIS)

1983-07-01

The operating reactors licensing actions summary is designed to provide the management of the Nuclear Regulatory Commission (NRC) with an overview of licensing actions dealing with operating power and nonpower reactors. These reports utilize data collected from the Division of Licensing in the Office of Nuclear Reactor Regulation and are prepared by the Office of Management and Program Analysis. This summary report is published primarily for internal NRC use in managing the operating reactors licensing actions program. Its content will change based on NRC management informational requirements

Communications highlights of the Finnish nuclear industry in 1997

International Nuclear Information System (INIS)

Heininen-Ojanperae, Marke

1998-01-01

Full text: The Poster will contain two major issues i.e. description of the information Centre Renovation Project at Loviisa Power Plant and a presentation with material of the Sunray III Project. The Sunray Project: The Sunray Project is a national radiation project intended for all ninth graders in the Finnish school system. The project started two years ago. Attention has been focused on the topic of radiation. The topic has been dealt with in connection with different subject as history, English, Swedish, French, Finnish, mathematics, physics, chemistry, geography, health and home economics, as well as vocational counselling. The aim of the project is to provide extensive information on the subject of radiation and radioactivity, to investigate benefits and disadvantages and to help pupils understand units and see things in correct proportions. Sunray I focused on radon, Sunray II on light. Sunray III will start this autumn and its main theme will be risk perception. The programme of the Sunray III is structured as follows: Part One: Articles on risk provided by experts from different fields. Part Two: A risk management game Part Three: A special view on natural radiation. The project is coordinated by Economic Information Bureau in cooperation with the Finnish Centre for Radiation and Nuclear Safety and two major power companies Imatran Voima and Teollisuuden Voima. The coordinator of the Sunray III Project is Mr. Matti Lattu. The Information Centre Renovation Project (ICRP) The ICRP has been a two-year project which was completed for the 20th anniversary celebration of Loviisa Nuclear Power Plant in February 1997. The main emphasis was to harness the local natural environment as much as possible. The focal point of the project was presented by a new, large-sized map as a water-colour painting (a printed version), a bird-eye view of the Haestholinen Island environment. The information signs system was completely renewed. A large sea-water aquarium containing

Study of seismic responses of Candu-3 reactor building using isolator bearings

International Nuclear Information System (INIS)

Biswas, J.K.

1992-01-01

Seismic isolator bearings are known to increase reliability, reduce cost and increase the potential sitings for nuclear power plants located in regions of high seismicity. High seismic activities in Canada occur mainly in the western coast, the Grand Banks and regions of Quebec along the St. Lawrence river. In Canada, nuclear power plants are located in Ontario, Quebec and New Brunswick where the seismicity levels are low to moderate. Consequently, seismic isolator bearings have not been used in the existing nuclear power plants in Canada. The present paper examines the effect of using seismic isolator bearings in the design for the new CANDU3 which would be suitable for regions having high seismicity. The CANDU3 Nuclear Power Plant is rated at 450 MW of net output power and is a smaller version of its predecessor CANDU6 successfully operating in Canada and abroad. The design of CANDU3 is being developed by AECL CANDU. Advanced technologies for design, construction and plant operation have been utilized. During the conceptual development of the CANDU3 design, various design options including the use of isolator bearings were considered. The present paper presents an overview of seismic isolation technology and summarizes the analytical work for predicting the seismic behavior of the CANDU3 reactor building. A lumped-parameter dynamic model for the reactor building is used for the analysis. The characteristics of the bearings are utilized in the analysis work. The time-history modal analysis has been used to compute the seismic responses. Seismic responses of the reactor building with and without isolator bearings are compared. The isolator bearings are found to reduce the accelerations of the reactor building. As a result, a lower level of seismic qualification for components and systems would be required. The use of these bearings however increases rigid body seismic displacements of the structure requiring special considerations in the layout and interfaces for

TU Electric reactor physics model verification: Power reactor benchmark

International Nuclear Information System (INIS)

Willingham, C.E.; Killgore, M.R.

1988-01-01

Power reactor benchmark calculations using the advanced code package CASMO-3/SIMULATE-3 have been performed for six cycles of Prairie Island Unit 1. The reload fuel designs for the selected cycles included gadolinia as a burnable absorber, natural uranium axial blankets and increased water-to-fuel ratio. The calculated results for both startup reactor physics tests (boron endpoints, control rod worths, and isothermal temperature coefficients) and full power depletion results were compared to measured plant data. These comparisons show that the TU Electric reactor physics models accurately predict important measured parameters for power reactors

Reactor Dosimetry Applications Using RAPTOR-M3G:. a New Parallel 3-D Radiation Transport Code

Science.gov (United States)

Longoni, Gianluca; Anderson, Stanwood L.

2009-08-01

The numerical solution of the Linearized Boltzmann Equation (LBE) via the Discrete Ordinates method (SN) requires extensive computational resources for large 3-D neutron and gamma transport applications due to the concurrent discretization of the angular, spatial, and energy domains. This paper will discuss the development RAPTOR-M3G (RApid Parallel Transport Of Radiation - Multiple 3D Geometries), a new 3-D parallel radiation transport code, and its application to the calculation of ex-vessel neutron dosimetry responses in the cavity of a commercial 2-loop Pressurized Water Reactor (PWR). RAPTOR-M3G is based domain decomposition algorithms, where the spatial and angular domains are allocated and processed on multi-processor computer architectures. As compared to traditional single-processor applications, this approach reduces the computational load as well as the memory requirement per processor, yielding an efficient solution methodology for large 3-D problems. Measured neutron dosimetry responses in the reactor cavity air gap will be compared to the RAPTOR-M3G predictions. This paper is organized as follows: Section 1 discusses the RAPTOR-M3G methodology; Section 2 describes the 2-loop PWR model and the numerical results obtained. Section 3 addresses the parallel performance of the code, and Section 4 concludes this paper with final remarks and future work.

Research on intelligent monitor for 3D power distribution of reactor core

International Nuclear Information System (INIS)

Xia, Hong; Li, Bin; Liu, Jianxin

2014-01-01

Highlights: • Core power distribution of ex-core measurement system has been reconstructed. • Building up an artificial intelligence model for 3-D core power distribution. • Error of the experiments has been reduced to 0.76%. • Methods for improving the accuracy of the model have been obtained. - Abstract: A real-time monitor for 3D reactor power distribution is critical for nuclear safety and high efficiency of NPP’s operation as well as for optimizing the control system, especially when the nuclear power plant (NPP) works at a certain power level or it works in load following operation. This paper was based on analyzing the monitor for 3D reactor power distribution technologies used in modern NPPs. Furthermore, considering the latest research outcomes, the paper proposed a method based on using an ex-core neutron detector system and a neural network to set up a real time monitor system for reactor’s 3D power distribution supervision. The results of the experiments performed on a reactor simulation machine illustrated that the new monitor system worked very well for a certain burn-up range during the fuel cycle. In addition, this new model could reduce the errors associated with the fitting of the distribution effectively, and several optimization methods were also obtained to improve the accuracy of the simulation model

Käyttäjäystävällisen SharePoint 2013 -ohjeen laatiminen : Case VisualWeb Group Oy

OpenAIRE

Pynttäri, Henna

2014-01-01

Tässä toimintakeskeisessä opinnäytetyössä suunniteltiin ja toteutettiin VisualWeb Group Oy :n asiakkaille käyttäjäystävällinen ohje SharePoint 2013 -alustan käyttöön. Manuaali on suunnattu peruskäyttäjille, ja se sisältää ohjeet sellaisista toiminnoista, joita käyttäjä tarvitsee aloittaessaan SharePoint-alustan päälle rakennetun sivustonsa päivittämisen. Päätavoitteena oli laatia käyttöohjeesta mahdollisimman selkeä ja yksinkertainen, koska se on suunnattu käyttäjäryhmälle, jolla on vähän tai...

Chemical reactivities of the superconducting oxides, YBa2Cu3Oy and BiSrCaCu2Oy

International Nuclear Information System (INIS)

Toyama, Hisashi; Mizuno, Noritaka; Misono, Makoto

1989-01-01

The chemical reactivities of YBa 2 Cu 3 O y and BiSrCaCu 2 O y with various gases have been studied. It was found that large quantities of NO, CO, and NO 2 were rapidly absorbed (or intercalated) in the bulk of YBa 2 Cu 3 O y (T c : 90 K) at 573 K. The amount absorbed was in the order NO ∼ CO ∼ NO 2 > O 2 ∼ CO 2 > N 2 O ∼ 0. The amount for NO was more than two times the amount of YBa 2 Cu 3 O y in molar ratio and elongation by about 0.2 angstrom along c-axis was observed. NO absorbed was almost completely recovered as NO by the evacuation at 773 K. This absorption-desorption cycle proceeded reversively. The electronic resistivity at 573 K of YBa 2 Cu 3 O y increased upon the NO absorption and was restored by the evacuation at 773 K. CO was also absorbed rapidly accompanied by evolution of CO 2 . BiSrCaCu 2 O y did not absorb either NO or CO

Effect of 3-D moderator flow configurations on the reactivity of CANDU nuclear reactors

International Nuclear Information System (INIS)

Zadeh, Foad Mehdi; Etienne, Stephane; Chambon, Richard; Marleau, Guy; Teyssedou, Alberto

2017-01-01

Highlights: • 3-D CFD simulations of CANDU-6 moderator flows are presented. • A thermal-hydraulic code using thermal physical fluid properties is used. • The numerical approach and convergence is validated against available data. • Flow configurations are correlated using Richardson’s number. • The interaction between moderator temperatures with reactivity is determined. - Abstract: The reactivity of nuclear reactors can be affected by thermal conditions prevailing within the moderator. In CANDU reactors, the moderator and the coolant are mechanically separated but not necessarily thermally isolated. Hence, any variation of moderator flow properties may change the reactivity. Until now, nuclear reactor calculations have been performed by assuming uniform moderator flow temperature distribution. However, CFD simulations have predicted large time dependent flow fluctuations taking place inside the calandria, which can bring about local temperature variations that can exceed 50 °C. This paper presents robust CANDU 3-D CFD moderator simulations coupled to neutronic calculations. The proposed methodology makes it possible to study not only different moderator flow configurations but also their effects on the reactor reactivity coefficient.

G2 and G3 reactors design; Description des reacteurs G2 et G3

Energy Technology Data Exchange (ETDEWEB)

Herreng,; Ertaud,; Pasquet, [Societe Alsacienne de Constructions Mecaniques (France)

1958-07-01

'FRANCE ATOME' Manufacturers Party has been entrusted with the G2 and G3 reactors engineering by the french A.E.C., for the first-five-year french project. Although these reactors are essentially plutonium generators, everyone has been linked with a power station which is supposed to supply with 40 MW, 'Electricite de France' has taken the liability upon itself. The reactor core includes most of G1 reactor parts (central gap excluded): horizontal channels, graphite parallelepipedic bricks stacking, steel thermal shield. The cooling is provided with CO{sub 2} under a 15 atmospheres pressure. This pressure is kept steady in a press-stressed concrete packing-case which is a cylinder horizontally shaped. Steel strips tightened encircle the concrete cylinder; itself protected by sole-plates. The cylinder bottom has brought about unusual problems which have been solved by the choice of an hemispheric shape. Packing-case tightness is provided by a 30 mm iron-plate connected with the inner wall of concrete. One of the reactor's special characteristics is the possibility of loading and unloading while operating. On loading side, barrel locks, each weighting 50 tons, allow new cans, at a pressure of 15 atmospheres, to pass. The cans process almost in a steady way through the channel, and finally drop down through bent spouts, then through spiral toboggans into a new lock. The cooling CO{sub 2} flow is provided with 3 turbo-bellows, these are actuated by average pressure-steam, obtained from exchangers. Every reactor supplies 4 exchangers which have been very difficult to build and to set up. The secondary cycle is standard and contains 3 stages (pressure 10,3: 2 and 0,5 kg/cm{sup 2}). Steam can be condensed in the event of a group turbo-generator stopping, with no modifion for the normal operating conditions of the reactor. Auxiliary circuits have to assure the continuous purifying of cooling CO{sub 2}, its storage and drain. 49 boron carbide rods are used to control the

Role of PRA in new NPP projects

International Nuclear Information System (INIS)

Julin, A.; Sandberg, J.; Virolainen, R.

2012-01-01

In Finland, a plant specific, Level 1 and 2 Probabilistic Risk Analysis (PRA) is required as a prerequisite for issuing the construction license and operating license. The use of PRA in various applications and the main insights are presented. These applications include e.g. PRA support to the design of SSCs (Systems, Structures and Components), definition of pre-service and in-service inspection programs, evaluation of the safety classification of SSCs, development of procedures, training and in definition of risk informed technical specifications, periodic testing and on-line preventive maintenance programs. In addition, PRA shall be used to assess the adequacy and coverage of the phase and system commissioning programs. Also the potential risks related to commissioning tests during nuclear test phase, shall be assessed with the help of PRA. In OL3 project, risk informed approach has been applied on a large scale for the first time in the design, construction and commissioning of a new NPP unit. Pre-nuclear commissioning tests have started at OL3 site and the plant is foreseen to begin commercial operation in 2013. Decisions have been made to launch new NPP projects. Teollisuuden Voima Oyj (TVO) is planning to build a new unit (OL4) at Olkiluoto site and a new utility, Fennovoima, is planning to build one unit at one of two alternative green field sites in Northern parts of Finland. Insights from PRAs of operating NPPs have been used in the evaluation of possible new sites to ensure that the site specific concerns and environmental conditions are adequately taken into account in the design of SSCs. Although the seismic activity at the Olkiluoto site is low, a comprehensive seismic risk analysis is being conducted. Its results support the review of the deterministic seismic design. For new sites, a probabilistic seismic hazard analysis has been carried out for the determination of the design earthquake. Experiences from OL3 licensing have been utilized in the

3. International conference on catalysis in membrane reactors

Energy Technology Data Exchange (ETDEWEB)

NONE

1998-09-01

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

Thermal-hydraulic modelling of the SAFARI-1 research reactor using RELAP/SCDAPSIM/MOD3.4

International Nuclear Information System (INIS)

Sekhri, Abdelkrim; Graham, Andy; D'Arcy, Alan; Oliver, Melissa

2008-01-01

The SAFARI-1 reactor is a tank-in-pool MTR type research reactor operated at a nominal core power of 20 MW. It operates exclusively in the single phase liquid water regime with nominal water and fuel temperatures not exceeding 100 deg. C. RELAP/SCDAPSIM/MOD3.4 is a Best Estimate Code for light water reactors as well as for low pressure transients, as part of the code validation was done against low pressure facilities and research reactor experimental data. The code was used to simulate SAFARI-1 in normal and abnormal operation and validated against the experimental data in the plant and was used extensively in the upgrading of the Safety Analysis Report (SAR) of the reactor. The focus of the following study is the safety analysis of the SAFARI-1 research reactor and describes the thermal hydraulic modelling and analysis approach. Particular emphasis is placed on the modelling detail, the application of the no-boiling rule and predicting the Onset of Nucleate Boiling and Departure from Nucleate Boiling under Loss of Flow conditions. Such an event leads the reactor to switch to a natural convection regime which is an adequate mode to maintain the clad and fuel temperature within the safety margin. It is shown that the RELAP/SCDAPSIM/MOD3.4 model can provide accurate predictions as long as the clad temperature remains below the onset of nucleate boiling temperature and the DNB ratio is greater than 2. The results are very encouraging and the model is shown to be appropriate for the analysis of SAFARI-1 research reactor. (authors)

Application of Raptor-M3G to reactor dosimetry problems on massively parallel architectures - 026

International Nuclear Information System (INIS)

Longoni, G.

2010-01-01

The solution of complex 3-D radiation transport problems requires significant resources both in terms of computation time and memory availability. Therefore, parallel algorithms and multi-processor architectures are required to solve efficiently large 3-D radiation transport problems. This paper presents the application of RAPTOR-M3G (Rapid Parallel Transport Of Radiation - Multiple 3D Geometries) to reactor dosimetry problems. RAPTOR-M3G is a newly developed parallel computer code designed to solve the discrete ordinates (SN) equations on multi-processor computer architectures. This paper presents the results for a reactor dosimetry problem using a 3-D model of a commercial 2-loop pressurized water reactor (PWR). The accuracy and performance of RAPTOR-M3G will be analyzed and the numerical results obtained from the calculation will be compared directly to measurements of the neutron field in the reactor cavity air gap. The parallel performance of RAPTOR-M3G on massively parallel architectures, where the number of computing nodes is in the order of hundreds, will be analyzed up to four hundred processors. The performance results will be presented based on two supercomputing architectures: the POPLE supercomputer operated by the Pittsburgh Supercomputing Center and the Westinghouse computer cluster. The Westinghouse computer cluster is equipped with a standard Ethernet network connection and an InfiniBand R interconnects capable of a bandwidth in excess of 20 GBit/sec. Therefore, the impact of the network architecture on RAPTOR-M3G performance will be analyzed as well. (authors)

The design of a fuel element for the RA-3 reactor (Ezeiza Atomic Center)

International Nuclear Information System (INIS)

Agueda, Horacio C.; Estevez, Esteban; Gerding, Jose R.; Markiewicz, Mario E.

2003-01-01

Some features of the mechanical design of the low enrichment fuel element for the RA-3 reactor are described, with emphasis in those aspects of the original design that have been modified considering the experience acquired in the design of other fuel elements. The proposed modification is based fundamentally on the replacement of all welded joints by screwed joints, which facilitates the manufacture of the fuel element, avoiding the distortions produced by the welds used at present and contributing to the fulfillment of the foreseen tolerances. A basic characteristic of this design is a careful manufacture of the fuel element's structural components in order to assure an assembling of the fuel element that fulfills the tolerances intrinsically required. The fuel is designed for the RA-3 reactor and uses U 3 O 8 or U 3 Si 2 as carrying phase of the fissile material with an enrichment of 19.70% of 235 U. The design verification was performed by analytical and numerical methods, and is supported by testing of materials in laboratory, hydrodynamics tests and performance evaluations of the fuel elements in the RA-3 reactor. (author)

Technical report on implementation of reactor internal 3D modeling and visual database system

International Nuclear Information System (INIS)

Kim, Yeun Seung; Eom, Young Sam; Lee, Suk Hee; Ryu, Seung Hyun

1996-06-01

In this report was described a prototype of reactor internal 3D modeling and VDB system for NSSS design quality improvement. For improving NSSS design quality several cases of the nuclear developed nation's integrated computer aided engineering system, such as Mitsubishi's NUWINGS (Japan), AECL's CANDID (Canada) and Duke Power's PASCE (USA) were studied. On the basis of these studies the strategy for NSSS design improvement system was extracted and detail work scope was implemented as follows : 3D modelling of the reactor internals were implemented by using the parametric solid modeler, a prototype system of design document computerization and database was suggested, and walk-through simulation integrated with 3D modeling and VDB was accomplished. Major effects of NSSS design quality improvement system by using 3D modeling and VDB are the plant design optimization by simulation, improving the reliability through the single design database system and engineering cost reduction by improving productivity and efficiency. For applying the VDB to full scope of NSSS system design, 3D modelings of reactor coolant system and nuclear fuel assembly and fuel rod were attached as appendix. 2 tabs., 31 figs., 7 refs. (Author) .new

Technical report on implementation of reactor internal 3D modeling and visual database system

Energy Technology Data Exchange (ETDEWEB)

Kim, Yeun Seung; Eom, Young Sam; Lee, Suk Hee; Ryu, Seung Hyun [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1996-06-01

In this report was described a prototype of reactor internal 3D modeling and VDB system for NSSS design quality improvement. For improving NSSS design quality several cases of the nuclear developed nation`s integrated computer aided engineering system, such as Mitsubishi`s NUWINGS (Japan), AECL`s CANDID (Canada) and Duke Power`s PASCE (USA) were studied. On the basis of these studies the strategy for NSSS design improvement system was extracted and detail work scope was implemented as follows : 3D modelling of the reactor internals were implemented by using the parametric solid modeler, a prototype system of design document computerization and database was suggested, and walk-through simulation integrated with 3D modeling and VDB was accomplished. Major effects of NSSS design quality improvement system by using 3D modeling and VDB are the plant design optimization by simulation, improving the reliability through the single design database system and engineering cost reduction by improving productivity and efficiency. For applying the VDB to full scope of NSSS system design, 3D modelings of reactor coolant system and nuclear fuel assembly and fuel rod were attached as appendix. 2 tabs., 31 figs., 7 refs. (Author) .new.

Licensed reactor nuclear safety criteria applicable to DOE reactors

International Nuclear Information System (INIS)

1991-04-01

The Department of Energy (DOE) Order DOE 5480.6, Safety of Department of Energy-Owned Nuclear Reactors, establishes reactor safety requirements to assure that reactors are sited, designed, constructed, modified, operated, maintained, and decommissioned in a manner that adequately protects health and safety and is in accordance with uniform standards, guides, and codes which are consistent with those applied to comparable licensed reactors. This document identifies nuclear safety criteria applied to NRC [Nuclear Regulatory Commission] licensed reactors. The titles of the chapters and sections of USNRC Regulatory Guide 1.70, Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants, Rev. 3, are used as the format for compiling the NRC criteria applied to the various areas of nuclear safety addressed in a safety analysis report for a nuclear reactor. In each section the criteria are compiled in four groups: (1) Code of Federal Regulations, (2) US NRC Regulatory Guides, SRP Branch Technical Positions and Appendices, (3) Codes and Standards, and (4) Supplemental Information. The degree of application of these criteria to a DOE-owned reactor, consistent with their application to comparable licensed reactors, must be determined by the DOE and DOE contractor

RELAP5-3D Code for Supercritical-Pressure Light-Water-Cooled Reactors

Energy Technology Data Exchange (ETDEWEB)

Riemke, Richard Allan; Davis, Cliff Bybee; Schultz, Richard Raphael

2003-04-01

The RELAP5-3D computer program has been improved for analysis of supercritical-pressure, light-water-cooled reactors. Several code modifications were implemented to correct code execution failures. Changes were made to the steam table generation, steam table interpolation, metastable states, interfacial heat transfer coefficients, and transport properties (viscosity and thermal conductivity). The code modifications now allow the code to run slow transients above the critical pressure as well as blowdown transients (modified Edwards pipe and modified existing pressurized water reactor model) that pass near the critical point.

Development of a 3-D flow analysis computer program for integral reactor

International Nuclear Information System (INIS)

Youn, H. Y.; Lee, K. H.; Kim, H. K.; Whang, Y. D.; Kim, H. C.

2003-01-01

A 3-D computational fluid dynamics program TASS-3D is being developed for the flow analysis of primary coolant system consists of complex geometries such as SMART. A pre/post processor also is being developed to reduce the pre/post processing works such as a computational grid generation, set-up the analysis conditions and analysis of the calculated results. TASS-3D solver employs a non-orthogonal coordinate system and FVM based on the non-staggered grid system. The program includes the various models to simulate the physical phenomena expected to be occurred in the integral reactor and will be coupled with core dynamics code, core T/H code and the secondary system code modules. Currently, the application of TASS-3D is limited to the single phase of liquid, but the code will be further developed including 2-phase phenomena expected for the normal operation and the various transients of the integrator reactor in the next stage

Research reactors in Argentina

International Nuclear Information System (INIS)

Carlos Ruben Calabrese

1999-01-01

Argentine Nuclear Development started in early fifties. In 1957, it was decided to built the first a research reactor. RA-1 reactor (120 kw, today licensed to work at 40 kW) started operation in January 1958. Originally RA-1 was an Argonaut (American design) reactor. In early sixties, the RA-1 core was changed. Fuel rods (20% enrichment) was introduced instead the old Argonaut core design. For that reason, a critical facility named RA-0 was built. After that, the RA-3 project started, to build a multipurpose 5 MW nuclear reactor MTR pool type, to produce radioisotopes and research. For that reason and to define the characteristics of the RA-3 core, another critical facility was built, RA-2. Initially RA-3 was a 90 % enriched fuel reactor, and started operation in 1967. When Atucha I NPP project started, a German design Power Reactor, a small homogeneous reactor was donated by the German Government to Argentina (1969). This was RA-4 reactor (20% enrichment, 1W). In 1982, RA-6 pool reactor achieved criticality. This is a 500 kW reactor with 90% enriched MTR fuel elements. In 1990, RA-3 started to operate fueled by 20% enriched fuel. In 1997, the RA-8 (multipurpose critical facility located at Pilcaniyeu) started to operate. RA-3 reactor is the most important CNEA reactor for Argentine Research Reactors development. It is the first in a succession of Argentine MTR reactors built by CNEA (and INVAP SE ) in Argentina and other countries: RA-6 (500 kW, Bariloche-Argentina), RP-10 (10MW, Peru), NUR (500 kW, Algeria), MPR (22 MW, Egypt). The experience of Argentinian industry permits to compete with foreign developed countries as supplier of research reactors. Today, CNEA has six research reactors whose activities have a range from education and promotion of nuclear activity, to radioisotope production. For more than forty years, Argentine Research Reactors are working. The experience of Argentine is important, and argentine firms are able to compete in the design and

Hydrogen production using Rhodopseudomonas palustris WP 3-5 with hydrogen fermentation reactor effluent

International Nuclear Information System (INIS)

Chi-Mei Lee; Kuo-Tsang Hung

2006-01-01

The possibility of utilizing the dark hydrogen fermentation stage effluents for photo hydrogen production using purple non-sulfur bacteria should be elucidated. In the previous experiments, Rhodopseudomonas palustris WP3-5 was proven to efficiently produce hydrogen from the effluent of hydrogen fermentation reactors. The highest hydrogen production rate was obtained at a HRT value of 48 h when feeding a 5 fold effluent dilution from anaerobic hydrogen fermentation. Besides, hydrogen production occurred only when the NH 4 + concentration was below 17 mg-NH 4 + /l. Therefore, for successful fermentation effluent utilization, the most important things were to decrease the optimal HRT, increase the optimal substrate concentration and increase the tolerable ammonia concentration. In this study, a lab-scale serial photo-bioreactor was constructed. The reactor overall hydrogen production efficiency with synthetic wastewater exhibiting an organic acid profile identical to that of anaerobic hydrogen fermentation reactor effluent and with effluent from two anaerobic hydrogen fermentation reactors was evaluated. (authors)

Research reactor core conversion guidebook. V. 3: Analytical verification (Appendices G and H)

International Nuclear Information System (INIS)

1992-04-01

Volume 3 consists of Appendix G which contains detailed results of a safety-related benchmark problem for an idealized reactor and Appendix H which contains detailed comparisons of calculated and measured data for actual cores with moderately enriched uranium and low enriched uranium fuels. The results of the benchmark calculations in Appendix G are summarized in Chapter 7 of Volume 1 and the results of the comparisons between calculations and measurements are summarized in Chapter 8 of Volume 1. Both the approaches described in these appendices are very useful in ensuring that the calculational methods employed in the preparation of a Safety Report are accurate. As a first step, it is recommended that reactor operators/physicists use their own methods and codes to first calculate the benchmark problem, and then compare the results of calculations with measurements in their own reactor or in one of the reactors for which measured data is available in Appendix H. (author). Refs, figs and tabs

The ARIES-III D-3He tokamak reactor

International Nuclear Information System (INIS)

Bathke, C.G.; Werley, K.A.; Miller, R.L.; Krakowski, R.A.; Santarius, J.F.

1992-01-01

The multi-institutional ARIES study has generated a conceptual design of another tokamak fusion reactor in a series that varies the assumed advances in technology and physics. The ARIES-III design uses a D- 3 He fuel cycle and requires advances in technology and physics for economical attractiveness. The optimal design was characterized through systems analyses for eventual conceptual engineering design. In this paper, results from the systems analysis are summarized, and a comparison with the high-field, D-T fueled ARIES-I is included

Annealing of the BR3 reactor pressure vessel

International Nuclear Information System (INIS)

Fabry, A.; Motte, F.; Stiennon, G.; Debrue, J.; Gubel, P.; Van de Velde, J.; Minsart, G.; Van Asbroeck, P.

1985-01-01

The pressure vessel of the Belgian BR-3 plant, a small (11 MWe) PWR presently used for fuel testing programs and operated since 1962, was annealed during March, 1984. The anneal was performed under wet conditions for 168 hours at 650 0 F with core removal and within plant design margins justification for the anneal, summary of plant characteristics, description of materials sampling, summary of reactor physics and dosimetry, development of embrittlement trend curves, hypothesized pressurized and overcooling thermal shock accidents, and conclusions are provided in detail

RB Research nuclear reactor RB reactor, Annual report for 2000

International Nuclear Information System (INIS)

Milosevic, M.

2000-12-01

Report on RB reactor operation during 2000 contains 3 parts. Part one contains a brief description of reactor operation and reactor components, relevant dosimetry data and radiation protection issues, personnel and financial data. Part two is devoted to maintenance of the reactor components, namely, fuel, heavy water, reactor vessel, heavy water circulation system, absorption rods and heavy water level-meters, maintenance of electronic, mechanical, electrical and auxiliary equipment. Part three contains data concerned with reactor operation and utilization with a comprehensive list of publications resulting from experiments done at the RB reactor. It contains data about reactor operation during previous 14 years, i.e. from 1986 - 2000

Reactor science and technology: operation and control of reactors

International Nuclear Information System (INIS)

Qiu Junlong

1994-01-01

This article is a collection of short reports on reactor operation and research in China in 1991. The operation of and research activities linked with the Heavy Water Research Reactor, Swimming Pool Reactor and Miniature Neutron Source Reactor are briefly surveyed. A number of papers then follow on the developing strategies in Chinese fast breeder reactor technology including the conceptual design of an experimental fast reactor (FFR), theoretical studies of FFR thermo-hydraulics and a design for an immersed sodium flowmeter. Reactor physics studies cover a range of topics including several related to work on zero power reactors. The section on reactor safety analysis is concerned largely with the assessment of established, and the presentation of new, computer codes for use in PWR safety calculations. Experimental and theoretical studies of fuels and reactor materials for FBRs, PWRs, BWRs and fusion reactors are described. A final miscellaneous section covers Mo-Tc isotope production in the swimming pool reactor, convective heat transfer in tubes and diffusion of tritium through plastic/aluminium composite films and Li 2 SiO 3 . (UK)

Master-3.0: multi-purpose analyzer for static and transient effects of reactors

International Nuclear Information System (INIS)

Cho, Byung Oh; Joo, Han Gyu; Cho, Jin Young; Song, Jae Seung; Zee, Sung Quun

2002-03-01

MASTER-3.0 (Multi-purpose Analyzer for Static and Transient Effects of Reactors) is a nuclear design code based on the multi-group diffusion theory to calculate the steady-state and transient pressurized water reactor core in a 3-dimensional Cartesian or hexagonal geometry. Its neutronics model solves the space-time dependent neutron diffusion equations with NIM (Nodal Integration Method), NEM (Nodal Expansion Method), AFEN (Analytic Function Expansion Nodal Method)/NEM Hybrid Method, NNEM (Non-linear Nodal Expansion Method) or NANM (Non-linear Analytic Nodal Method) for a Cartesian geometry and with NTPEN (Non-linear Triangle-based Polynomial Expansion Nodal Method), AFEN (Analytic Function Expansion Nodal)/NEM Hybrid Method or NLFM (Non-linear Local Fine-Mesh Method) for a hexagonal one. Coarse mesh rebalancing, Krylov Subspace method, energy group restriction/prolongation method and asymptotic extrapolation method are implemented to accelerate the convergence of iteration process. MASTER-3.0 performs microscopic depletion calculations using microscopic cross sections provided by CASMO-3 or HELIOS and also has the reconstruction capability of pin information by use of MSS-IAS (Method of Successive Smoothing with Improved Analytic Solution). For the thermal-hydraulic calculation, fuel temperature table or COBRA3-C/P or MATRA model can be used selectively. In addition, MASTER-3.0 is designed to cover various PWRs including SMART as well as WH- and CE-type reactors, providing all data required in their design procedures

3 D flow computations under a reactor vessel closure head

International Nuclear Information System (INIS)

Daubert, O.; Bonnin, O.; Hofmann, F.; Hecker, M.

1995-12-01

The flow under a vessel cover of a pressurised water reactor is investigated by using several computations and a physical model. The case presented here is turbulent, isothermal and incompressible. Computations are made with N3S code using a k-epsilon model. Comparisons between numerical and experimental results are on the whole satisfying. Some local improvements are expected either with more sophisticated turbulence models or with mesh refinements automatically computed by using the adaptive meshing technique which has been just implemented in N3S for 3D cases. (authors). 6 refs., 7 figs

Review on the seismic safety of JRR-3 according to the revised regulatory code on seismic design for nuclear reactors

Energy Technology Data Exchange (ETDEWEB)

Kobayashi, Tetsuya; Araki, Masaaki; Ohba, Toshinobu; Torii, Yoshiya [Japan Atomic Energy Agency, Tokai, Ibaraki (Japan); Takeuchi, Masaki [Nuclear Safety Commission (Japan)

2012-03-15

JRR-3(Japan Research Reactor No.3) with the thermal power of 20MW is a light water moderated and cooled, swimming pool type research reactor. JRR-3 has been operated without major troubles. This paper presents about review on the seismic safety of JRR-3 according to the revised regulatory code on seismic design for nuclear reactors. In addition, some topics concerning damages in JRR-3 due to the Great East Japan Earthquake are presented. (author)

Neutron flux calculations for the Rossendorf research reactor in (hex)- and (hex,z)-geometry using SNAP-3D

International Nuclear Information System (INIS)

Koch, R.; Findeisen, A.

1986-04-01

The multigroup neutron diffusion theory code SNAP-3D has been used to perform time independent neutron flux and power calculations of the 10 MW Rossendorf research reactor of the type WWR-SM. The report describes these calculations, as well as the actual reactor configuration, some details of the code SNAP-3D, and two- and three-dimensional reactor models. For evaluating the calculations some flux values and control rod worths have been compared with those of measurements. (author)

Thermal fluid dynamics study of nuclear advanced reactors of high temperature using RELAP5-3D

International Nuclear Information System (INIS)

Scari, Maria Elizabeth

2017-01-01

Fourth Generation nuclear reactors (GEN-IV) are being designed with special features such as intrinsic safety, reduction of isotopic inventory and use of fuel in proliferation-resistant cycles. Therefore, the investigation and evaluation of operational and safety aspects of the GEN-IV reactors have been the subject of numerous studies by the international community and also in Brazil. In 2008, in Brazil, was created the National Institute of Science and Technology of Innovative Nuclear Reactors, focusing on studies of projects and systems of new generation reactors, which included GEN-IV reactors as well as advanced PWR (Pressurized Water Reactor) concepts. The Department of Nuclear Engineering of the Federal University of Minas Gerais (DEN-UFMG) is a partner of this Institute, having started studies on the GEN-IV reactors in the year 2007. Therefore, in order to add knowledge to these studies, in this work, three projects of advanced reactors were considered to verify the simulation capability of the thermo-hydraulic RELAP5-3D code for these systems, either in stationary operation or in transient situations. The addition of new working fluids such as ammonia, carbon dioxide, helium, hydrogen, various types of liquid salts, among them Flibe, lead, lithium-bismuth, lithium-lead, was a major breakthrough in this version of the code, allowing also the simulation of GEN-IV reactors. The modeling of the respective core of an HTTR (High Temperature Engineering Test Reactor), HTR-10 (High Temperature Test Module Reactor) and LS-VHTR (Liquid-Salt-Cooled Very-High-Temperature Reactor) were developed and verified in steady state comparing the values found through the calculations with reference data from other simulations, when it is possible. The first two reactors use helium gas as coolant and the LS-VHTR uses a mixture of 66% LiF and 34% of BeF 2 , the LiF-BeF 2 , also know as Flibe. All the studied reactors use enriched uranium as fuel, in form of TRISO (Tristructural

3D CFD for chemical transport profiles in a rotating disk CVD reactor

Science.gov (United States)

Han, Jong-Hyun; Yoon, Do-Young

2010-06-01

The RDCVD (Rotating Disk Chemical Vapor Deposition) technique is an appropriate method for uniform deposition of grains, such as compound semiconductior materials. The substrate temperature and rotation speed are the major factors, which determine the thickness uniformity of the deposited films. This paper investigates 3D CFD (3 Dimensional Computational Fluid Dynamics) simulation results of flow and heat transfer in a reactor of RDCVD using Fluent. In order to establish the reducibility of buoyancy effect on deposition quality, the chemical transport profile upon the disk heated is examined successfully in 3D domain for different rotating speeds. The resulting vortex flows due the simultaneous buoyance and centrifuge are discussed qualitatively in the 3D virtual system of a RDCVD reactor. 3D CFD is even more effective to describe the internal vortex flows due to the competitive inlet, buoyancy and centrifuge flows, which cannot be realized in the general 2D (2 Dimensional) CFD.[Figure not available: see fulltext.

Development of a version of the reactor dynamics code DYN3D applicable for High Temperature Reactors; Entwicklung einer Version des Reaktordynamikcodes DYN3D fuer Hochtemperaturreaktoren. Abschlussbericht

Energy Technology Data Exchange (ETDEWEB)

Rohde, Ulrich; Apanasevich, Pavel; Baier, Silvio; Duerigen, Susan; Fridman, Emil; Grahn, Alexander; Kliem, Soeren; Merk, Bruno

2012-07-15

Based on the reactor dynamics code DYN3D for the simulation of transient processes in Light Water Reactors, a code version DYN3D-HTR for application to graphitemoderated, gas-cooled block-type high temperature reactors has been developed. This development comprises: - the methodical improvement of the 3D steady-state neutron flux calculation for the hexagonal geometry of the HTR fuel element blocks - the development of methods for the generation of homogenised cross section data taking into account the double heterogeneity of the fuel element block structure - the implementation of a 3D model for heat conduction and heat transport in the graphite matrix. The nodal method for neutron flux calculation based on SP3 transport approximation was extended to hexagonal fuel element geometry, where the hexagons are subdivided into triangles, thus the method had finally to be derived for triangular geometry. In triangular geometry, a subsequent subdivision of the hexagonal elements can be considered, and therefore, the effect of systematic mesh refinement can be studied. The algorithm was verified by comparison with Monte Carlo reference solutions, on the node-wise level, as well as also on the pin-wise level. New procedures were developed for the homogenization of the double-heterogeneous fuel element structures. One the one hand, the so-called Reactivity equivalent Physical Transformation (RPT), the two-step homogenization method based on 2D deterministic lattice calculations, was extended to cells with different temperatures of the materials. On the other hand, the progress in development of Monte Carlo methods for spectral calculations, in particular the development of the code SERPENT, opened a new, fully consistent 3D approach, where all details of the structures on fuel particle, fuel compact and fuel block level can be taken into account within one step. Moreover, a 3D heat conduction and heat transport model was integrated into DYN3D to be able to simulate radial

Experimental and kinetic modeling study of 3-methylheptane in a jet-stirred reactor

KAUST Repository

Karsenty, Florent

2012-08-16

Improving the combustion of conventional and alternative fuels in practical applications requires the fundamental understanding of large hydrocarbon combustion chemistry. The focus of the present study is on a high-molecular-weight branched alkane, namely, 3-methylheptane, oxidized in a jet-stirred reactor. This fuel, along with 2-methylheptane, 2,5-dimethylhexane, and n-octane, are candidate surrogate components for conventional diesel fuels derived from petroleum, synthetic Fischer-Tropsch diesel and jet fuels derived from coal, natural gas, and/or biomass, and renewable diesel and jet fuels derived from the thermochemical treatment of bioderived fats and oils. This study presents new experimental results along with a low- and high-temperature chemical kinetic model for the oxidation of 3-methylheptane. The proposed model is validated against these new experimental data from a jet-stirred reactor operated at 10 atm, over the temperature range of 530-1220 K, and for equivalence ratios of 0.5, 1, and 2. Significant effort is placed on the understanding of the effects of methyl substitution on important combustion properties, such as fuel reactivity and species formation. It was found that 3-methylheptane reacts more slowly than 2-methylheptane at both low and high temperatures in the jet-stirred reactor. © 2012 American Chemical Society.

Development of a 3D-Multigroup program to simulate anomalous diffusion phenomena in the nuclear reactors

International Nuclear Information System (INIS)

Maleki Moghaddam, Nader; Afarideh, Hossein; Espinosa-Paredes, Gilberto

2015-01-01

Highlights: • The new version of neutron diffusion equation for simulating anomalous diffusion is presented. • Application of fractional calculus in the nuclear reactor is revealed. • A 3D-Multigroup program is developed based on the fractional operators. • The super-diffusion and sub-diffusion phenomena are modeled in the nuclear reactors core. - Abstract: The diffusion process is categorized in three parts, normal diffusion, super-diffusion and sub-diffusion. The classical neutron diffusion equation is used to model normal diffusion. A new scheme of derivatives is required to model anomalous diffusion phenomena. The fractional space derivatives are employed to model anomalous diffusion processes where a plume of particles spreads at an inconsistent rate with the classical Brownian motion model. In the fractional diffusion equation, the fractional Laplacians are used; therefore the statistical jump length of neutrons is unrestricted. It is clear that the fractional Laplacians are capable to model the anomalous phenomena in nuclear reactors. We have developed a NFDE-3D (neutron fractional diffusion equation) as a core calculation code to model normal and anomalous diffusion phenomena. The NFDE-3D is validated against the LMW-LWR reactor. The results demonstrate that reactors exhibit complex behavior versus order of the fractional derivatives which depends on the competition between neutron absorption and super-diffusion phenomenon

Study on effects of development of reactor constant in fast reactor analysis

International Nuclear Information System (INIS)

Chiba, Gou

2002-12-01

Evaluation was carried out about an effect of development of the new generation reactor constant system that substitutes for the JFS library in fast reactor analysis. Analyzed cores were ZPPR in JUPITER critical experiment and several power reactor cores that were designed in the feasibility study. In the JUPITER analysis, large effects, over 10%, were observed in sodium void reactivity and sample Doppler reactivity. The former resulted from several factors, while the latter was due to an accurate of a resonance interaction effect between Doppler sample and core fuel. In the previous study, the effect had been evaluated in power reactor cores. The effect included an effect of corrosion of weighting spectrum because JFS-3-J3.2, which had been made with the incorrect weighting spectrum, was used in the evaluation. In the present study, JFS-3-J3.2R, which had been made with the correct weighting spectrum, was used. It was confirmed that the effect of development of reactor constant in power reactor was not as large as that in critical assembly. (author)

Simulation in 3 dimensions of a cycle 18 months for an BWR type reactor using the Nod3D program; Simulacion en 3 dimensiones de un ciclo de 18 meses para un reactor BWR usando el programa Nod3D

Energy Technology Data Exchange (ETDEWEB)

Hernandez, N.; Alonso, G. [ININ, A.P. 18-1027, 11801 Mexico D.F. (Mexico)]. E-mail: nhm@nuclear.inin.mx; Valle, E. del [IPN, ESFM, 07738 Mexico D.F. (Mexico)

2004-07-01

The development of own codes that you/they allow the simulation in 3 dimensions of the nucleus of a reactor and be of easy maintenance, without the consequent payment of expensive use licenses, it can be a factor that propitiates the technological independence. In the Department of Nuclear Engineering (DIN) of the Superior School of Physics and Mathematics (ESFM) of the National Polytechnic Institute (IPN) a denominated program Nod3D has been developed with the one that one can simulate the operation of a reactor BWR in 3 dimensions calculating the effective multiplication factor (kJJ3, as well as the distribution of the flow neutronic and of the axial and radial profiles of the power, inside a means of well-known characteristics solving the equations of diffusion of neutrons numerically in stationary state and geometry XYZ using the mathematical nodal method RTN0 (Raviart-Thomas-Nedelec of index zero). One of the limitations of the program Nod3D is that it doesn't allow to consider the burnt of the fuel in an independent way considering feedback, this makes it in an implicit way considering the effective sections in each step of burnt and these sections are obtained of the code Core Master LEND. However even given this limitation, the results obtained in the simulation of a cycle of typical operation of a reactor of the type BWR are similar to those reported by the code Core Master LENDS. The results of the keJ - that were obtained with the program Nod3D they were compared with the results of the code Core Master LEND, presenting a difference smaller than 0.2% (200 pcm), and in the case of the axial profile of power, the maxim differs it was of 2.5%. (Author)

Synthesis of the IRSN report related to severe accidents and to the probabilistic level-2 safety study for the Flamanville EPR reactor. Referral of the Permanent Group of Experts for nuclear reactors (GPR), examination of probabilistic level-2 safety studies (EPS 2) and severe accidents (AG) of the Flamanville reactor nr 3. Opinion related to severe accidents and to the probabilistic level-2 safety study for the Flamanville EPR reactor (FA3). Electronuclear reactors - EDF - Flamanville 3 EPR reactor. Severe accidents and probabilistic level 2 studies

International Nuclear Information System (INIS)

2015-01-01

This document gathers several documents. The first one recalls the main arrangements implemented on the FA3 EPR reactor regarding accidents with core fusion, reports the analysis made by the IRSN about the sizing of these arrangements to reach a controlled status of the installation after a severe accident, regarding the probabilistic level-2 safety assessment, regarding the radiological impact of a severe accident on the population and on the environment, regarding those aimed at facing a total and long duration loss of electric power sources and cold sources, and about the situation of the reactor with respect to WENRA positions on severe accidents for new reactors. The second document is a letter sent by the ASN to the Permanent Group of Experts for nuclear reactors (GPR) to address probabilistic level-2 safety studies (EPS2) and severe accidents for the Flamanville 3 reactor. The third one reports the opinion of the GPR on these both issues and proposes a set of recommendations. The next document is a letter sent by the ASN to the Flamanville 3 project manager at EDF which recalls the related objectives, the ASN opinion on the implemented arrangements for severe accidents (de-pressurization of the primary circuit, management of hydrogen-related risks, corium recovery and cooling outside the vessel, limitation of vapour explosion risks outside the vessel, heat evacuation system, containment enclosure, management of the risk of a return to criticality), to face a total and long duration loss of electricity sources and cold sources, and other aspects addressed in the IRSN analysis. Requests and remarks formulated by the ASN are provided in an appendix to this last document

Extension of the GeN-Foam neutronic solver to SP3 analysis and application to the CROCUS experimental reactor

International Nuclear Information System (INIS)

Fiorina, Carlo; Hursin, Mathieu; Pautz, Andreas

2017-01-01

Highlights: • Development and verification of an SP 3 solver based on OpenFOAM. • Integration into the GeN-Foam multi-physics platform. • Application of the new GeN-Foam SP 3 solver to the CROCUS reactor. - Abstract: The Laboratory for Reactor Physics and Systems Behaviour at the PSI and at the EPFL has been developing since 2013 a multi-physics platform for coupled reactor analysis named GeN-Foam. The developed tool includes a solver for the eigenvalue and transient solution of multi-group neutron diffusion equations. Although frequently used in reactor analysis, the diffusion theory shows some limitations for core configurations involving strong anisotropies, which is the case for the CROCUS research reactor at the EPFL. The use of an SP 3 approximation to neutron transport can often lead to visible improvements in a code predictive capabilities, especially for one-directional anisotropies, with acceptable added computational cost vs diffusion. Following some modelling issues for the CROCUS reactor, and in order to improve the GeN-Foam modelling capabilities, the GeN-Foam diffusion solver has been extended to allow for SP 3 analyses. The present paper describes such extension and a preliminary verification using a mini-core PWR benchmark. The newly developed solver is then applied to the analysis of the CROCUS experimental reactor and results are compared to Monte Carlo calculations, as well as to the results of the diffusion solver.

A next-generation reactor concept: The Integral Fast Reactor (IFR)

Energy Technology Data Exchange (ETDEWEB)

Chang, Y.I.

1992-01-01

The Integral Fast Reactor (IFR) is an advanced liquid metal reactor concept being developed at Argonne National Laboratory as reactor technology for the 21st century. It seeks to specifically exploit the inherent properties of liquid metal cooling and metallic fuel in a way that leads to substantial improvements in the characteristics of the complete reactor system, in particular passive safety and waste management. The IFR concept consists of four technical features: (1) liquid sodium cooling, (2) pool-type reactor configuration, (3) metallic fuel, and (4) fuel cycle closure based on pyroprocessing.

A next-generation reactor concept: The Integral Fast Reactor (IFR)

Energy Technology Data Exchange (ETDEWEB)

Chang, Y.I.

1992-07-01

The Integral Fast Reactor (IFR) is an advanced liquid metal reactor concept being developed at Argonne National Laboratory as reactor technology for the 21st century. It seeks to specifically exploit the inherent properties of liquid metal cooling and metallic fuel in a way that leads to substantial improvements in the characteristics of the complete reactor system, in particular passive safety and waste management. The IFR concept consists of four technical features: (1) liquid sodium cooling, (2) pool-type reactor configuration, (3) metallic fuel, and (4) fuel cycle closure based on pyroprocessing.

A next-generation reactor concept: The Integral Fast Reactor (IFR)

International Nuclear Information System (INIS)

Chang, Y.I.

1992-01-01

The Integral Fast Reactor (IFR) is an advanced liquid metal reactor concept being developed at Argonne National Laboratory as reactor technology for the 21st century. It seeks to specifically exploit the inherent properties of liquid metal cooling and metallic fuel in a way that leads to substantial improvements in the characteristics of the complete reactor system, in particular passive safety and waste management. The IFR concept consists of four technical features: (1) liquid sodium cooling, (2) pool-type reactor configuration, (3) metallic fuel, and (4) fuel cycle closure based on pyroprocessing

Tritium production, management and its impact on safety for a D-3He fusion reactor

International Nuclear Information System (INIS)

Sze, D.K.; Herring, S.; Sawan, M.

1991-11-01

About three percent of the fusion energy produced by a D- 3 He reactor is in the form of neutrons. Those neutrons are generated by D-D and D-T reactions, with the tritium produced by the D-D fusion. The neutrons will react with structural steel, deuterium, 3 He and shielding material to produce tritium. About half of the tritium generated by the D-D reaction will not burn in the plasma and will exit as a part of the plasma exhaust. Thus, there is enough tritium produced in a D- 3 He reactor and careful management will be required. The tritium produced in the shield and plasma can be managed with an acceptable effect on cost and safety. 3 refs., 2 figs., 3 tabs

Advanced reactors transition fiscal year 1995 multi-year program plan WBS 7.3

International Nuclear Information System (INIS)

Loika, E.F.

1994-01-01

This document describes in detail the work to be accomplished in FY-1995 and the out years for the Advanced Reactors Transition (WBS 7.3). This document describes specific milestones and funding profiles. Based upon the Fiscal Year 1995 Multi-Year Program Plan, DOE will provide authorization to perform the work outlined in the FY 1995 MYPP. Following direction given by the US Department of Energy (DOE) on December 15, 1993, Advanced Reactors Transition (ART), previously known as Advanced Reactors, will provide the planning and perform the necessary activities for placing the Fast Flux Test Facility (FFTF) in a radiologically and industrially safe shutdown condition. The DOE goal is to accomplish the shutdown in approximately five years. The Advanced Reactors Transition Multi-Year Program Plan, and the supporting documents; i.e., the FFTF Shutdown Program Plan and the FFTF Shutdown Project Resource Loaded Schedule (RLS), are defined for the life of the Program. During the transition period to achieve the Shutdown end-state, the facilities and systems will continue to be maintained in a safe and environmentally sound condition. Additionally, facilities that were associated with the Office of Nuclear Energy (NE) Programs, and are no longer required to support the Liquid Metal Reactor Program will be deactivated and transferred to an alternate sponsor or the Decontamination and Decommissioning (D and D) Program for final disposition, as appropriate

Reactor Physics Programme

Energy Technology Data Exchange (ETDEWEB)

De Raedt, C

2000-07-01

The Reactor Physics and Department of SCK-CEN offers expertise in various areas of reactor physics, in particular in neutronics calculations, reactor dosimetry, reactor operation, reactor safety and control and non-destructive analysis on reactor fuel. This expertise is applied within the Reactor Physics and MYRRHA Research Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 reactor dosimetry, and the preparation and interpretation of irradiation experiments. Progress and achievements in 1999 in the following areas are reported on: (1) investigations on the use of military plutonium in commercial power reactors; (2) neutron and gamma calculations performed for BR-2 and for other reactors; (3) the updating of neutron and gamma cross-section libraries; (4) the implementation of reactor codes; (6) the management of the UNIX workstations; and (6) fuel cycle studies.

Reactor Physics Programme

International Nuclear Information System (INIS)

De Raedt, C.

2000-01-01

The Reactor Physics and Department of SCK-CEN offers expertise in various areas of reactor physics, in particular in neutronics calculations, reactor dosimetry, reactor operation, reactor safety and control and non-destructive analysis on reactor fuel. This expertise is applied within the Reactor Physics and MYRRHA Research Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 reactor dosimetry, and the preparation and interpretation of irradiation experiments. Progress and achievements in 1999 in the following areas are reported on: (1) investigations on the use of military plutonium in commercial power reactors; (2) neutron and gamma calculations performed for BR-2 and for other reactors; (3) the updating of neutron and gamma cross-section libraries; (4) the implementation of reactor codes; (6) the management of the UNIX workstations; and (6) fuel cycle studies

Characteristics of UV-MicroO3 Reactor and Its Application to Microcystins Degradation during Surface Water Treatment

Directory of Open Access Journals (Sweden)

Guangcan Zhu

2015-01-01

Full Text Available The UV-ozone (UV-O3 process is not widely applied in wastewater and potable water treatment partly for the relatively high cost since complicated UV radiation and ozone generating systems are utilized. The UV-microozone (UV-microO3, a new advanced process that can solve the abovementioned problems, was introduced in this study. The effects of air flux, air pressure, and air humidity on generation and concentration of O3 in UV-microO3 reactor were investigated. The utilization of this UV-microO3 reactor in microcystins (MCs degradation was also carried out. Experimental results indicated that the optimum air flux in the reactor equipped with 37 mm diameter quartz tube was determined to be 18∼25 L/h for efficient O3 generation. The air pressure and humidity in UV-microO3 reactor should be low enough in order to get optimum O3 output. Moreover, microcystin-RR, YR, and LR (MC-RR, MC-YR, and MC-LR could be degraded effectively by UV-microO3 process. The degradation of different MCs was characterized by first-order reaction kinetics. The pseudofirst-order kinetic constants for MC-RR, MC-YR, and MC-LR degradation were 0.0093, 0.0215, and 0.0286 min−1, respectively. Glucose had no influence on MC degradation through UV-microO3. The UV-microO3 process is hence recommended as a suitable advanced treatment method for dissolved MCs degradation.

Nuclear reactor (1960)

International Nuclear Information System (INIS)

Maillard, M.L.

1960-01-01

The first French plutonium-making reactors G1, G2 and G3 built at Marcoule research center are linked to a power plant. The G1 electrical output does not offset the energy needed for operating this reactor. On the contrary, reactors G2 and G3 will each generate a net power of 25 to 30 MW, which will go into the EDF grid. This power is relatively small, but the information obtained from operation is great and will be helpful for starting up the power reactor EDF1, EDF2 and EDF3. The paper describes how, previous to any starting-up operation, the tests performed, especially those concerned with the power plant and the pressure vessel, have helped to bring the commissioning date closer. (author) [fr

Construction of Research Reactors for Gen 3 and Gen 4 Reactors Development

International Nuclear Information System (INIS)

Behar, Christophe

2014-01-01

Christophe Behar, Director of the Nuclear Energy Division at CEA, detailed the different kind of research reactors and the issues in term of investment, use, side application such as the medical isotopes production

Simulation of a reactor FBR with hexagonal-Z geometry using the code PARCS 3.1

International Nuclear Information System (INIS)

Reyes F, M. C.; Del Valle G, E.; Filio L, C.

2013-10-01

The nuclear reactor core type FBR (Fast Breeder Reactor) was modeled in three dimensions of hexagonal-Z geometry using the code PARCS (Purdue Advanced Reactor Core Simulator) version 3.1 developed by Purdue University researchers. To carry out the modeling of the mentioned reactor was taken the corresponding information to one of the described benchmarks in the document NEACRP-L-330 (3-D Neutron Transport Benchmarks, 1991); fundamentally the corresponding to the geometric data and the cross sections. Being a quick reactor of breeding, known as the Knk-II, for which are considered 4 energy groups without dispersions up. The reactor core is formed by prismatic elements of hexagonal transversal cut where part of them only corresponds to nuclear fuel assemblies. This has four reflector rings and 6 identical control elements that together with the active part of the core is configured with 8 different types of elements.With the extracted information of the mentioned document the entrance file was prepared for PARCS 3.1 only considering a sixth part of the core due to the symmetry that presents their configuration. The NEACRP-L-330 shows a wide range of results reported by those who collaborated in its elaboration using different solution techniques that go from the Monte Carlo method to the approaches S 2 and P 1 . Of all the results were selected those obtained with the code HEXNOD, to which were carried out a comparison of the effective multiplication factor, being smaller differences to the 300 pcm, for three different scenarios: a) with the control bars extracted totally, b) with the semi-inserted control bars and c) with the control bars inserted completely and two different axial meshes, a thick mesh with 14 slices and another fine with 38, that which implies that the results can be considered very similar among if same. Radial maps and axial profiles are included, as much of the power as of the neutrons flow. (Author)

Bases for Decisions on Final Disposal in Finland

International Nuclear Information System (INIS)

Avolahti, Jaana

2001-01-01

The disposal of the spent nuclear fuel is approaching one of the significant milestones in Finland. Social debate on the nuclear waste management is going on aiming at a decision of principle on future directions of spent fuel management. The research so far has required no political decision. This current situation is preceded by preparations for two decades carried out by Posiva Oy who took over the programme managed earlier by Teollisuuden Voinia Oy, one of the country's nuclear power companies. The preparations comprise site investigations, technical concept development, research into long-term safety and an environmental impact assessment. The work carried out by Posiva is under regular assessment by the authorities. Research programmes are drawn up every year and reports are published for open review. The preparations in the next years to come aim at starting the construction of the repository in 2010 and the disposal operations are planned to be started in 2020. Various stakeholders in Finland are involved in the decision-making process on the disposal of spent nuclear fuel. The process started when Posiva as an implementer applied for the Government's Decision in Principle in 1999. The Government made a favourable decision in December 2000 on the basis of different considerations. Among the important bases were the preceding favourable decisions made by the proposed siting municipality and the regulatory authority for radiation safety. At the moment the members of the Parliament are discussing the principles of the disposal in order to be able to vote on the Government's decision in the springtime. This paper discusses similarities and differences between the decisions made so far as regards the deep repository. The objective is to present the significance of the decisions from the point of view of an implementer of the repository. The Decision in Principle does not give any consent to start constructing the repository. Licenses for construction and

Operating experiences and utilization programmes of the BAEC 3 MW TRIGA Mark-II research reactor of Bangladesh

International Nuclear Information System (INIS)

Haque, M.M.; Soner, M.A.M.; Saha, P.K.; Salam, M.A.; Zulquarnain, M.A.

2008-01-01

The 3 MW TRIGA Mark-II research reactor of Bangladesh Atomic Energy Commission (BAEC) has been operating since September 14, 1986. The reactor is used for radioisotope production ( 131 I, 99m Tc, 46 Sc), various R and D activities, manpower training and education. The reactor has been operated successfully since its commissioning with the exception of a few reportable incidents. Of these, the decay tank leakage incident of 1997 is considered to be the most significant one. As a result of this incident, reactor operation at full power remained suspended for about 4 years. However, the reactor operation was continued during this period at a power level of 250 kW to cater the needs of various R and D groups, which required lower neutron flux for their experiments. This was made possible by establishing a temporary by pass connection across the decay tank using local technology. The reactor was made operational again at full power after successful replacement of the damaged decay tank in August 2001. At that time, several modifications of the reactor cooling system along with its associated structures were also implemented and then necessary testing and commissioning of the newly installed component/equipment were carried out. The other incident was the contamination of the Dry Central Thimble (DCT) that took place in March 2002 when a pyrex vial containing 50g of TeO 2 powder got melted inside the DCT. The vial was melted due to high heat generation on its surface while the reactor was operated for 8 hours at 3 MW for trial production of Iodine-131 ( 131 I). A Wet Central Thimble (WCT) was used to replace the damaged DCT in June 2002 such that the reactor operation could be resumed. The WCT was again replaced by a new DCT in June 2003 such that radioisotope production could be continued. The facility has so far been used to train up a total of 27 personnel including several foreign nationals to the level of Senior Reactor Operator (SRO) and Reactor Operator (RO). The

Effect of fuel assembly when changing from AFA 2G to AFA 3G on seismic loads of reactor internal

International Nuclear Information System (INIS)

Liu Wenjin; Zeng Zhongxiu; Ye Xianhui; Wu Wanjun

2013-01-01

Nonlinear seismic model for reactor with fuel assemblies of AFA 2G and AFA 3G is established. Using ANSYS software, seismic nonlinear time -history analysis is completed and the effects on seismic loads of reactor system are obtained. The result shows that when the fuel assembly changing from AFA 2G to AFA 3G, it is necessary to reevaluate the fuel assembly itself, but not the reactor internal. (authors)

FMDP Reactor Alternative Summary Report: Volume 3 - partially complete LWR alternative

International Nuclear Information System (INIS)

Greene, S.R.; Fisher, S.E.; Bevard, B.B.

1996-09-01

The Department of Energy Office of Fissile Materials Disposition (DOE/MD) initiated a detailed analysis activity to evaluate each of ten plutonium disposition alternatives that survived an initial screening process. This document, Volume 3 of a four volume report summarizes the results of these analyses for the partially complete LWR (PCLWR) reactor based plutonium disposition alternative

FMDP Reactor Alternative Summary Report: Volume 3 - partially complete LWR alternative

Energy Technology Data Exchange (ETDEWEB)

Greene, S.R.; Fisher, S.E.; Bevard, B.B. [and others

1996-09-01

The Department of Energy Office of Fissile Materials Disposition (DOE/MD) initiated a detailed analysis activity to evaluate each of ten plutonium disposition alternatives that survived an initial screening process. This document, Volume 3 of a four volume report summarizes the results of these analyses for the partially complete LWR (PCLWR) reactor based plutonium disposition alternative.

Modelling and thermal hydraulic analysis of the Angra-2 nuclear reactor using RELAP5-3D code

International Nuclear Information System (INIS)

González Mantecón, Javier

2015-01-01

The evaluation of Nuclear Power Plants (NPPs) performance during steady-state and accident conditions has been one of the main research subjects in the nuclear field. In order to simulate the behavior of water-cooled reactors, several complex thermal-hydraulic codes systems have been developed. Particularly, the RELAP5 code, developed by the Idaho National Laboratory, is a best-estimate thermal-hydraulic analysis tool and one of the most used in nuclear industry. The RELAP5-3D 3.0.0 code was used to develop a detailed model of Angra 2 nuclear reactor using reference data from the Final Safety Analysis Report. Angra 2 is the second Brazilian NPP, which began commercial operation in 2001. The plant is equipped with a Pressurized Water Reactor (PWR) type with 3771.0 MWt. Simulations of the reactor behavior during normal operation conditions and postulated accident conditions were performed. Results achieved in the reactor steady-state simulation were compared with nominal parameters of the NPP. These results proved to be in good agreement, with relative errors less than 1%. In the transient simulation, the obtained results were coherent and satisfactory. This study demonstrates that the RELAP5-3D model is capable to reproduce the thermal-hydraulic behavior of the Angra-2 PWR during diverse operation conditions and it can contribute for the process of the plant safety analysis. (author)

Activity report on the utilization of research reactors (JRR-3 and JRR-4). Japanese fiscal year, 2009

International Nuclear Information System (INIS)

2014-02-01

JRR-3 is used for the purposes below; Experimental studies such as neutron scattering, prompt gamma-ray analyses, neutron radiography, Irradiation for activation analyses, radioisotope (RI) productions, fission tracks, Irradiation test of reactor materials, etc. JRR-4 is used for the purposes below; Medical irradiation (Boron Neutron Capture Therapy : BNCT), Prompt gamma-ray analyses, Sensitivity measurement of radiation detectors, Experiment in the nuclear reactor training, Practice of Reactor operation, Irradiation for activation analyses, RI productions, fission tracks, etc. In the fiscal year 2009, The research reactor JRR-3 was operated 7 cycles (cycle operation : 26days/cycle) for utilization sharing of the facility. And JRR-4 was operated 6 cycles (daily operation : 24 days). The volume contains 138 activity reports, which are categorized into the fields of neutron scattering (11 subcategories), neutron radiography, prompt gamma-ray analyses, neutron activation analyses, RI productions, and others submitted by the users in JAEA and from other organizations. (author)

Activity report on the utilization of research reactors (JRR-3 and JRR-4). Japanese fiscal year, 2005

International Nuclear Information System (INIS)

2007-03-01

In the fiscal year 2005, The research reactor JRR-3 was operated 7 cycles (cycle operation : 26days/cycle) for utilization sharing of the facility. And JRR-4 was operated 37 cycles (daily operation : 137 days). JRR-3 is used for the purposes below; Experimental studies such as neutron scattering, prompt gamma-ray analyses, neutron radiography. Irradiation for activation analyses, radioisotope (RI) productions, fission tracks. Irradiation test of reactor materials etc. JRR-4 is used for the purposes below; Medical irradiation (Boron Neutron Capture Therapy : BNCT). Prompt gamma-ray analyses. Sensitivity measurement of radiation detectors. Experiment in the nuclear reactor training. Practice of Reactor operation. Irradiation for activation analyses, RI productions, fission tracks etc. The volume contains 100 activity reports, which are categorized into the fields of neutron scattering (9 subcategories), neutron radiography, neutron activation analyses, RI productions, prompt gamma-ray analyses, and others submitted by the users in JAEA and from other organizations. (author)

Activity report on the utilization of research reactors (JRR-3 and JRR-4). Japanese fiscal year, 2006

International Nuclear Information System (INIS)

2009-01-01

In the fiscal year 2006, the research reactor JRR-3 was operated 7 cycles (cycle operation: 26 days/cycle) for utilization sharing of the facility. And JRR-4 was operated 37 cycles (daily operation: 151 days). JRR-3 is used for the purposes below; Experimental studies such as neutron scattering, prompt gamma-ray analyses, neutron radiography, Irradiation for activation analyses, radioisotope (RI) productions, fission tracks, Irradiation test of reactor materials, etc. JRR-4 is used for the purposes below; Medical irradiation (Boron Neutron Capture Therapy : BNCT), Prompt gamma-ray analyses, Sensitivity measurement of radiation detectors, Experiment in the nuclear reactor training, Practice of Reactor operation, Irradiation for activation analyses, RI productions, fission tracks, etc. The volume contains 294 activity reports, which are categorized into the fields of neutron scattering (11 subcategories), neutron radiography, neutron activation analyses, RI productions, prompt gamma-ray analyses, and others submitted by the users in JAEA and from other organizations. (author)

Activity report on the utilization of research reactors (JRR-3 and JRR-4). Japanese fiscal year, 2009

Energy Technology Data Exchange (ETDEWEB)

NONE

2014-02-15

JRR-3 is used for the purposes below; Experimental studies such as neutron scattering, prompt gamma-ray analyses, neutron radiography, Irradiation for activation analyses, radioisotope (RI) productions, fission tracks, Irradiation test of reactor materials, etc. JRR-4 is used for the purposes below; Medical irradiation (Boron Neutron Capture Therapy : BNCT), Prompt gamma-ray analyses, Sensitivity measurement of radiation detectors, Experiment in the nuclear reactor training, Practice of Reactor operation, Irradiation for activation analyses, RI productions, fission tracks, etc. In the fiscal year 2009, The research reactor JRR-3 was operated 7 cycles (cycle operation : 26days/cycle) for utilization sharing of the facility. And JRR-4 was operated 6 cycles (daily operation : 24 days). The volume contains 138 activity reports, which are categorized into the fields of neutron scattering (11 subcategories), neutron radiography, prompt gamma-ray analyses, neutron activation analyses, RI productions, and others submitted by the users in JAEA and from other organizations. (author)

A 3D transport-based core analysis code for research reactors with unstructured geometry

International Nuclear Information System (INIS)

Zhang, Tengfei; Wu, Hongchun; Zheng, Youqi; Cao, Liangzhi; Li, Yunzhao

2013-01-01

Highlights: • A core analysis code package based on 3D neutron transport calculation in complex geometry is developed. • The fine considerations on flux mapping, control rod effects and isotope depletion are modeled. • The code is proved to be with high accuracy and capable of handling flexible operational cases for research reactors. - Abstract: As an effort to enhance the accuracy in simulating the operations of research reactors, a 3D transport core analysis code system named REFT was developed. HELIOS is employed due to the flexibility of describing complex geometry. A 3D triangular nodal S N method transport solver, DNTR, endows the package the capability of modeling cores with unstructured geometry assemblies. A series of dedicated methods were introduced to meet the requirements of research reactor simulations. Afterwards, to make it more user friendly, a graphical user interface was also developed for REFT. In order to validate the developed code system, the calculated results were compared with the experimental results. Both the numerical and experimental results are in close agreement with each other, with the relative errors of k eff being less than 0.5%. Results for depletion calculations were also verified by comparing them with the experimental data and acceptable consistency was observed in results

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

International Nuclear Information System (INIS)

Moukhamadeev, R.; Suvorov, A.

2000-01-01

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

ASAMPSA2 best-practices guidelines for L2 PSA development and applications. Volume 3 - Extension to Gen IV reactors

International Nuclear Information System (INIS)

Bassi, C.; Bonneville, H.; Brinkman, H.; Burgazzi, L.; Polidoro, F.; Vincon, L.; Jouve, S.

2010-01-01

The main objective assigned to the Work Package 4 (WP4) of the 'ASAMPSA2' project (EC 7. FPRD) consist in the verification of the potential compliance of L2PSA guidelines based on PWR/BWR reactors (which are specific tasks of WP2 and WP3) with Generation IV representative concepts. Therefore, in order to exhibit potential discrepancies between LWRs and new reactor types, the following work was based on the up-to-date designs of: - The European Fast Reactor (EFR) which will be considered as prototypical of a pool-type Sodium-cooled Fast Reactor (SFR); - The ELSY design for the Lead-cooled Fast Reactor (LFR) technology; - The ANTARES project which could be representative of a Very-High Temperature Reactor (VHTR); - The CEA 2400 MWth Gas-cooled Fast Reactor (GFR). (authors)

Plutonium-burn high temperature gas-cooled reactor for 3E+3S

International Nuclear Information System (INIS)

Okamoto, Koji

2015-01-01

The Nuclear Energy Development in Japan is facing a very difficult conditions after Fukushima-Daiichi NPP Accident. Nuclear Energy has strong advantages on 3E, i.e., Energy security, Economical efficiency and Environment. However, people does not believe the Safety 'S' of Nuclear Energy, now. The disadvantage of 'S' overrides the advantages of '3E'. In Nuclear Energy, 'S' is expanded into 3S, i.e., Safety, Security and Safeguards. Especially, the management of Plutonium inventory in Spent Fuel generated by the NPP operation is very important in the viewpoints of non-proliferation. The high-temperature gas cooled reactor (HTGR) is the solution of these disadvantages of '3S' in Nuclear Energy. The fuel of HTGR is composed by 1 mm spherical fuel particle, i.e., TRISO made by fuel, graphite and silicon-carbide. The silicon-carbide can confine the fission products in any conditions of fuel life cycle, i.e., during operation, accidents and disposal for 1 million years. The confinement of the radioactive materials can be confirmed by the TRISO. The HTGR core has strong negative feedback for temperature. So, the fission automatically stopped at the accidental conditions, such as loss of flow and LOCA. Also, the residual heat can be cooled by the radiation heat transfer to reactor vessel wall. The HTGR system usually has passive vessel wall cooling system. When the passive cooling system had been failed, the heat can be transferred to the land by heat conductions, and fuel does not reach the SiC broken temperature. The fission chain reaction has been stopped automatically by negative feedback, i.e., physics. The residual heat had been cooled automatically by radiation. The radioactive materials had been confined automatically by silicon-carbide. The HTGR is superior for 'S' safety. Plutonium can be burned by the HTGR. In the viewpoints of non-proliferation, the fuel should be made by YSZ-PuO 2 , stabilized buffer

Mathematical Modelling of Catalytic Fixed-Bed Reactor for Carbon Dioxide Reforming of Methane over Rh/Al2O3 Catalyst

Directory of Open Access Journals (Sweden)

New Pei Yee

2008-04-01

Full Text Available A one-dimensional mathematical model was developed to simulate the performance of catalytic fixed bedreactor for carbon dioxide reforming of methane over Rh/Al2O3 catalyst at atmospheric pressure. The reactionsinvolved in the system are carbon dioxide reforming of methane (CORM and reverse water gas shiftreaction (RWGS. The profiles of CH4 and CO2 conversions, CO and H2 yields, molar flow rate and molefraction of all species as well as reactor temperature along the axial bed of catalyst were simulated. In addition,the effects of different reactor temperature on the reactor performance were also studied. The modelscan also be applied to analyze the performances of lab-scale micro reactor as well as pilot-plant scale reactorwith certain modifications and model verification with experimental data. © 2008 BCREC UNDIP. All rights reserved.[Received: 20 August 2008; Accepted: 25 September 2008][How to Cite: N.A.S. Amin, I. Istadi, N.P. Yee. (2008. Mathematical Modelling of Catalytic Fixed-Bed Reactor for Carbon Dioxide Reforming of Methane over Rh/Al2O3 Catalyst. Bulletin of Chemical Reaction Engineering and Catalysis, 3 (1-3: 21-29. doi:10.9767/bcrec.3.1-3.19.21-29